Vermont Yankee - Defueled Safety Analysis Report, Revision 1

VYNPS DSAR Revision 1

DEFUELED SAFETY ANALYSIS REPORT

VERMONT YANKEE NUCLEAR POWER STATION

DEFUELED SAFETY ANALYSIS REPORT

TABLE OF CONTENTS

VYNPS DSAR Revision 1 TOC-1 of 2

SECTION 1 UFSAR, REV 17, TOC

1.0 INTRODUCTION AND SUMMARY 1.1 INTRODUCTION 1.2 DESIGN CRITERIA 1.3 FACILITY DESCRIPTION 1.4 SUMMARY OF RADIATION EFFECTS 1.5 GENERAL CONCLUSIONS SECTION 2 2.0 STATION SITE AND ENVIRONS 2.1 SUMMARY DESCRIPTION 2.2 SITE DESCRIPTION 2.3 METEOROLOGY 2.4 HYDROLOGY AND BIOLOGY 2.5 GEOLOGY AND SEISMOLOGY 2.6 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SECTION 3 3.0 FACILITY DESIGN AND OPERATION 3.1 DESIGN CRITERIA 3.2 FACILITY STRUCTURES 3.3 SYSTEMS SECTION 4 4.0 RADIOACTIVE WASTE MANAGEMENT 4.1 SOURCE TERMS 4.2 RADIATION SHIELDING 4.3 HEALTH PHYSICS INSTRUMENTATION 4.4 RADIATION PROTECTION PROGRAM 4.5 LIQUID WASTE MANAGEMENT SYSTEMS 4.6 SOLID WASTE MANAGEMENT 4.7 EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING

DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Continued)

VYNPS DSAR Revision 1 TOC-2 of 2

SECTION 5 5.0 CONDUCT OF OPERATIONS 5.1 ORGANIZATION AND RESPONSIBILITY 5.2 TRAINING 5.3 EMERGENCY PLAN 5.4 QUALITY ASSURANCE PROGRAM 5.5 REVIEW AND AUDIT OF OPERATIONS 5.6 TECHNICAL REQUIREMENTS MANUAL SECTION 6 6.0 SAFETY ANALYSIS 6.1 INTRODUCTION 6.2 ACCEPTANCE CRITERIA 6.3 ACCIDENTS EVALUATED 6.4 SITE EVENTS EVALUATED 6.5 REFERENCES 6.6 APPENDICES SECTION 7 7.0 AGING MANAGEMENT 7.1 SUPPLEMENT FOR RENEWED OPERATING LICENSE 7.2 AGING MANAGEMENT PROGRAMS AND ACTIVITIES 7.3 REFERENCES 7.4 LIST OF LICENSE RENEWAL COMMITMENTS APPENDICES G.2 CURRENT ON-SITE METEOROLOGICAL PROGRAM

VYNPS DSAR Revision 1 1.0-1 of 17

INTRODUCTION AND SUMMARY TABLE OF CONTENTS

Section Title Page

1.1 INTRODUCTION .......................................................... 3

1.2 DESIGN CRITERIA ....................................................... 5

1.3 FACILITY DESCRIPTION .................................................. 7

1.3.1 General ...................................................... 7

1.3.1.1 Site and Environs ............................... 7

1.3.1.2 Facility Arrangement ............................ 9

1.3.2 Fuel Storage and Handling .................................... 9

1.3.2.1 Nuclear Fuel .................................... 9

1.3.2.2 Deleted ......................................... 9

1.3.2.3 Standby Fuel Pool Cooling and Demineralizer System ............................ 9

1.3.3 Radioactive Waste Management ................................. 9

1.3.3.1 Equipment and Floor Drainage Systems ........... 10

1.3.3.2 Liquid Radwaste System ......................... 10

1.3.3.3 Solid Radwaste System .......................... 10

1.3.4 Radiation Monitoring and Control ............................ 11

1.3.4.1 Reactor Building Ventilation Radiation Monitoring System .............................. 11

1.3.4.2 Process Radiation Monitoring ................... 11

1.3.4.3 Area Radiation Monitors ........................ 11

1.3.5 Auxiliary Systems ........................................... 12

1.3.5.1 Electrical Power Systems ....................... 12

1.3.5.2 Service Water System ........................... 12

1.3.5.3 Fire Protection System ......................... 13

1.3.5.4 Heating, Ventilating, and Air Conditioning Systems ........................... 13

1.3.5.5 Service and Instrument Air Systems ............. 13

1.3.5.6 Process Sampling System ........................ 14


1.3.6 Communications Systems ...................................... 14

1.3.6.1 Facility Communications System ................. 14

1.3.7 Station Water Purification, Treatment and Storage ........... 14

1.3.7.1 Deleted ........................................ 14

1.3.7.2 Potable and Sanitary Water System .............. 15

1.3.8 Shielding, Access Control, and Radiation Protection Procedures .................................................. 15

1.3.8.1 General ........................................ 15

1.3.9 Structural Loading Criteria ................................. 16

1.4 SUMMARY OF RADIATION EFFECTS ......................................... 17

1.4.1 Fuel Storage and Handling and Waste Management .............. 17

1.4.2 Accidents and Events ........................................ 17

1.5 GENERAL CONCLUSIONS .................................................. 17


1.1 INTRODUCTION

On January 12, 2015, Entergy Nuclear Operations (ENO) certified to the Nuclear

Regulatory Commission (NRC) that a determination to permanently cease operation

at the Vermont Yankee Nuclear Power Station (VYNPS) was made on December 29, 2014

which was the date on which operation ceased at VYNPS. ENO also certified that

the fuel has been permanently removed from the VYNPS reactor vessel and placed in

the spent fuel pool. ENO acknowledged that, following docketing, the VYNPS

license no longer authorized operation of the reactor or emplacement or retention

of fuel into the reactor vessel.

This Defueled Safety Analysis Report (DSAR) is derived from Revision 26 of the

VYNPS Updated Final Safety Analysis Report (UFSAR). The DSAR has been developed

as a licensing basis document that reflects the permanently defueled condition of

VYNPS. The DSAR serves the same function during SAFSTOR and decommissioning that

the UFSAR served during operation of the facility. An evaluation of the systems,

structures and components (SSCs) described in the UFSAR was performed to

determine the function, if any, these SSCs would perform in a defueled condition.

The criteria used to evaluate the major SSCs and the conclusions of the

evaluations are provided in appropriate station documents.

ENO acknowledged that the 10CFR50 operating license continues to remain in effect

until the Nuclear Regulatory Commission terminates the license.

The Vermont Yankee Nuclear Power Corporation was originally organized by ten New

England utilities in August, 1966, for the purpose of building and operating a

nuclear generating station in Vermont. At the time of application, Vermont

Yankee was similar in organization to the Yankee Atomic Electric Co. and the

Connecticut Yankee Atomic Power Co. Nine of the twelve Vermont Yankee sponsors

were also sponsors of Yankee and Connecticut Yankee. Thus, Vermont Yankee had

the benefit of the experience gained from the operation of these two plants.

The Vermont Yankee Nuclear Power Corporation was the sole applicant for an

operating license for a nuclear power station, located at the Vernon site in

Windham County, Vermont, for initial power levels up to 1593 MWt under Section

104(b) of the Atomic Energy Act of 1954, as amended, and the regulations of the

NRC set forth in Part 50 of Title 10 of the Code and Federal Regulations

(10CFR50).

The facility was designated as the Vermont Yankee Nuclear Power Station.

The Vermont Yankee Nuclear Power Corporation, as owner, was responsible for the

design, construction, operation and decommissioning of the station.


EBASCO Services, Inc. designed and constructed the station exclusive of the

nuclear steam supply system.

General Electric Company was awarded a contract to design, fabricate, and deliver

the nuclear steam supply system and nuclear fuel for the station, as well as to

provide technical direction for installation and startup of this equipment.

General Electric Company was also contracted to design, fabricate, deliver, and

install the turbine generator as well as to provide technical assistance for the

startup of this equipment.

In July 2002, the operating license was transferred to Entergy Nuclear Vermont

Yankee, LLC, a limited liability company and wholly owned subsidiary of Entergy

Nuclear Operations, Inc.


1.2 DESIGN CRITERIA

The principal architectural and engineering criteria for the design and

construction of the station, applicable in the permanently defueled state, are

summarized below.

General

The station design shall be in accordance with applicable codes and

regulations.

The station shall be designed in such a way that the release of radioactive

materials to the environment is limited so that the limits and guideline values

of Title 10 of the Code of Federal Regulations pertaining to the release of

radioactive materials are not exceeded.

Structural

Adequate strength and stiffness with appropriate safety factors shall be

provided so that a hazardous release of radioactive material shall not occur.

Nuclear Fuel

The fuel cladding shall be designed to retain integrity as a radioactive

material barrier.

The fuel cladding shall be designed to accommodate without loss of integrity

the pressures generated by the fission gases released from the fuel material

throughout the design life of the fuel.

The fuel cladding, in conjunction with other facility systems, shall be

designed to retain integrity throughout any abnormal operational transient.

Fuel Handling and Storage

Fuel handling and storage facilities shall be designed to maintain adequate

shielding and cooling for spent fuel.

Fuel handling and storage facilities shall be designed to preclude inadvertent

criticality.


Electrical Power Systems

The electric power system shall be designed to provide sufficient normal and

standby electrical power to assure proper operation of the spent fuel pool

cooling and support systems.

Transformers, switchgear, buses, and cables shall be designed to have adequate

current carrying capacity without exceeding the acceptable voltage drop of the

electrical loads.

Switchgear protective devices shall be provided to detect and interrupt

electrical malfunctions.

The rated capacity of interrupting devices shall exceed the maximum available

fault current.

Radioactive Waste Disposal Systems

Liquid and solid waste disposal facilities shall be designed so that the

discharge and off-site shipment of radioactive effluents can be made in

accordance with applicable regulations.

The design shall provide means to inform station operating personnel of an

approach to limits on the release of radioactive material.

Shielding and Access Control Radiation shielding shall be provided and access control patterns shall be

established to allow the staff to control radiation doses within the limits of

10CFR20.


1.3 FACILITY DESCRIPTION

1.3.1 General

1.3.1.1 Site and Environs

1.3.1.1.1 Location and Size of Site

The site is located on the west shore of the Connecticut River immediately

upstream of the Vernon Hydroelectric Station, in the town of Vernon, Vermont,

which is in Windham County. Site coordinates are approximately 4247' north,

7231' west. The facility is located on about 125 acres which are bounded by

privately owned land on the north, south, and west and by the Connecticut River

on the east. The site plot plan is shown on Drawing 5920-6245.

1.3.1.1.2 Site Ownership

Entergy Nuclear Vermont Yankee, LLC is the owner of the site, with the

exception of a narrow strip of land between the Connecticut River and the VYNPS

property for which it has perpetual rights and easements from its owner.

1.3.1.1.3 Activities at Site

All activities at the facility site will be under the control of Entergy

Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. at all times.

1.3.1.1.4 Access to the Site

The immediate area around the facility is completely enclosed by a fence with

access to the facility controlled at a security gate. Access to the site is

possible from either Governor Hunt Road, a local road, or from a spur of the

Central Vermont Railroad. Site boundaries are posted.

1.3.1.1.5 Description of Environs

The area adjacent to the facility is primarily farm and pasture land. Downstream

of the facility are the Vernon Hydroelectric Station and the town of Vernon,

Vermont. The area within a 5-mile radius is predominantly rural with the

exception of a portion of the city of Brattleboro, Vermont and the town of

Hinsdale, New Hampshire. Between 75% and 80% of the area within 5 miles of the

facility is wooded. The remainder is occupied by farms and small industries.


1.3.1.1.6 Geology

The major structures at the site are supported by bedrock. Compression tests

indicated minimum failure of the bedrock to be 16,000 psi (1,152 tons per

square foot). An allowable bearing pressure has been established at 50 tons

per square foot; however, actual loadings do not exceed 20 tons per square

foot.

1.3.1.1.7 Seismology

Based on a three-fold seismic evaluation, the site was found to be relatively

quiescent from a seismic standpoint. From these studies the design earthquake

has been established at 0.07g horizontal ground acceleration and the maximum

hypothetical earthquake at 0.14g horizontal ground acceleration. The seismic

evaluation consisted of a review of historical data from the New England area,

an analysis of instrument and historical records for the Vermont area, and a

study of earthquake intensity attenuation with distance for the northeast

United States.

1.3.1.1.8 Hydrology

The facility is on the Connecticut River in Vernon, Vermont, some 138.3 miles

from the river mouth. The river in the vicinity of the facility is comprised

of a series of ponds formed by dams constructed for the generation of

hydroelectric power. All local surface streams drain to the Connecticut River,

and the site is in the direct path of natural drainage to the east of the local

watershed. In the vicinity of the site there is also a considerable amount of

groundwater which several municipalities utilize as one source of water supply.

1.3.1.1.9 Regional and Site Meteorology

The general climatic regime is that of a continental type with some

modification from the maritime climate which prevails nearer the coast. For

the one-year period between August 1967 and July 1968, temperature inversions

occurred 39% of the total time. Seasonal inversion frequencies ranged between

36% and 42%. Wind distribution is biased in the direction of the river due to

the channeling effect of the valley.

Historical records show that annual snowfall varies between 30 inches and

118 inches. Temperature range is about 133F. Occasional heavy rains and ice

storms occur in the area.


1.3.1.2 Facility Arrangement

The facility arrangement is shown on Drawing 5920-6245. The principal

structures of the station are the reactor building and primary containment,

turbine building, control building, radwaste building, intake structure,

cooling towers, main stack, and an Independent Spent Fuel Storage Installation

(ISFSI) storage pad.

1.3.2 Fuel Storage and Handling

1.3.2.1 Nuclear Fuel

Nuclear fuel previously used for power generation consists of slightly enriched

uranium dioxide pellets contained in sealed Zircaloy tubes. These fuel rods

are assembled into individual fuel assemblies. On January 12, 2015, VYNPS

certified to the NRC that all nuclear fuel had been permanently removed from

the reactor vessel and placed in the spent fuel pool. Therefore, all nuclear

fuel is stored either in the Spent Fuel Pool (SFP) or at the Independent Spent

Fuel Storage Installation (ISFSI) Facility.

1.3.2.2 Deleted

1.3.2.3 Standby Fuel Pool Cooling and Demineralizer System

The Standby Fuel Pool Cooling (SFPCS) removes decay heat released from the

spent fuel to maintain fuel pool temperature within specified limits. The Fuel

Pool Demineralizer System (FPDS) maintains water clarity.

1.3.3 Radioactive Waste Management

The Radioactive Waste Systems are designed to control the release of

radioactive material to within the limits specified in 10CFR20 and within the

limits specified in technical specifications and the Off-Site Dose Calculation

Manual (ODCM). The methods employed for the controlled release of these

contaminants depends primarily upon the state of the material.


1.3.3.1 Equipment and Floor Drainage Systems

Drains and sumps are provided to ensure proper drainage and collection of all

reject liquids throughout the facility. The drain systems are:

1. The chemical waste sump and equipment and floor sumps in the Radwaste

Building and Reactor Building that contain or potentially contain

radioactive liquids are routed to the Torus.

2. Uncontaminated liquids are drained to storm sewers or other areas where

they can be discharged to the river.

1.3.3.2 Liquid Radwaste System

The Liquid Radwaste System is no longer in service. The system has been drained to the extent practical. The Torus-as-CST System processes water collected from the chemical waste sump and equipment and floor drains in the Radwaste Building and Reactor Building. This water is stored in the Torus and is normally used to control spent fuel pool inventory. Water stored in the torus may be disposed of offsite or discharged to the environs in accordance with applicable permits and regulatory approvals.

1.3.3.3 Solid Radwaste System

Solid radioactive wastes are collected, processed, and packaged for storage and subsequent off-site burial. Generally, these wastes are stored on-site until the short half-lived activities are insignificant. Solid wastes from equipment originating in the Nuclear System are stored for radioactive decay in the fuel storage pool and prepared for reprocessing or off-site burial in approved shipping containers. Examples of these wastes are spent fuel, spent control rods, in-core ion chambers, etc. Process solid wastes, such as resins or filter material, are collected, dewatered, and prepared for storage in shielded casks. Dry active waste such as paper, air filters, and used clothing is collected and temporarily stored in large shipping containers before being sent to a disposal site or to an off-site waste processor for volume reduction prior to disposal. The processed waste may be returned to VYNPS in strong tight packages, or sent directly to burial.


1.3.4 Radiation Monitoring and Control

1.3.4.1 Reactor Building Ventilation Radiation Monitoring System

The Reactor Building Ventilation Radiation Monitoring System consists of

radiation monitors arranged to monitor the activity level of the ventilation

exhaust from the Reactor Building.

1.3.4.2 Process Radiation Monitoring

Radiation monitors and monitoring systems are provided on process liquid and

gas lines that may serve as discharge routes for radioactive materials. The

monitors include the following:

Plant Stack Radiation Monitoring System

Process Liquid Radiation Monitoring System

Reactor Building Ventilation Radiation Monitoring System

1.3.4.3 Area Radiation Monitors

Radiation monitors are provided to monitor for abnormal radiation at various

locations. These monitors annunciate alarms when abnormal radiation levels are

detected.


1.3.5 Auxiliary Systems

1.3.5.1 Electrical Power Systems

At the 345 kV switchyard, a ring bus arrangement supplies the 115 kV switchyard

through a 345 kV/115 kV autotransformer. A line from the 115 kV switchyard also

interconnects with 115 kV transmission systems in New Hampshire. Off-site

power is supplied to the facility from the 115 kV switchyard via two startup

transformers.

The Auxiliary AC Power System provides adequate power for the safe storage and

handling of irradiated fuel and support activities.

Backup power is available from the Vernon Hydroelectric Station and the Station

Blackout Diesel.

The Main Battery System provides a reliable source of dc power for control

power to selected breakers and power to selected lighting systems.

1.3.5.2 Service Water System

The Service Water System supplies cooling water from the Connecticut River

directly to auxiliary equipment. Pumps supply the systems and equipment through

a dual header arrangement.


1.3.5.3 Fire Protection System

Water for the Fire Protection System is supplied by two vertical turbine-type

pumps, one diesel driven and one electric-motor driven, both located in the

intake structure. These pumps supply water to the facility fire loop with its

various hydrants and subsequently to the standpipe connections, sprinklers, and

deluge systems throughout portions of the facility. Supplementing these water

systems are a CO2 Fire Protection System for the cable vault and Switchgear

Rooms and portable fire extinguishers located throughout the facility.

The Heating Boiler Room is protected by automatic fire detection devices which

alarm in the Main Control Room.

Consideration has been given to the use of noncombustible and fire-resistant

materials throughout the facility.

1.3.5.4 Heating, Ventilating, and Air Conditioning Systems

The Heating, Ventilating, and Air Conditioning (HVAC) Systems normally provide

filtered air to the facility structures.

This air provides the appropriate temperature and humidity conditions as

required in these structures for personnel and equipment protection. It

provides for the effective protection of personnel against possible airborne

radioactive contaminants by maintaining flow direction and rate so that the

gaseous or particulate contaminants are effectively prevented from entering the

cleaner zones.

1.3.5.5 Service and Instrument Air Systems

The Instrument Air System provides the facility with a continuous supply of

dry, oil-free air for pneumatic instruments and controls through a dual header

system.

The Service Air System provides the facility with a continuous supply of air

where the air quality of the Instrument Air System is not required. Four 100%

capacity air compressors and two air receiver tanks comprise the pieces of

equipment for the two systems. Additionally, the Instrument Air System has a

filter and drier in each header to ensure air quality.


1.3.5.6 Process Sampling System

The Process Sampling System provides a means for sampling and testing various

process fluids in centralized locations, from which the performance of the

facility, items of equipment, and systems may be determined.

1.3.6 Communications Systems

1.3.6.1 Facility Communications System

The Communications System provides adequate means of communication throughout

the facility and from the facility to off-site locations. The on-site means of

communication are:

1. Intrasite dial telephone system

2. Intrastation public address system

3. Sound-powered telephone system

4. Intrastation radio communications system

Communications to off-site locations can be accomplished by means of:

1. Public telephones

2. Off-site radio communications system

3. Intersite microwave communications system

1.3.7 Station Water Purification, Treatment and Storage

1.3.7.1 Deleted


1.3.7.2 Potable and Sanitary Water System

Potable and sanitary water, filtered and treated as necessary, is provided in

sufficient quantity by this system to supply all facility drinking and sanitary

water requirements.

1.3.8 Shielding, Access Control, and Radiation Protection Procedures

1.3.8.1 General

Control of radiation exposure of facility personnel and people external to the

facility exclusion area is accomplished by a combination of radiation

shielding, control of access into certain areas, and administrative procedures.

The requirements of 10CFR20 are used as a basis for establishing the basic

criteria and objectives.

Shielding is used to reduce radiation dose rates in various parts of the

facility to acceptable limits. Access control and administrative procedure are

used to limit the integrated dose received by facility personnel to less than

that set forth in 10CFR20. Access control and procedures are also used to

limit the potential spread of contamination from various areas, particularly

areas where maintenance occurs.

Shielding is also used as necessary to protect equipment from radiation damage.

Of principal concern are organic materials such as insulation, linings, and

gaskets. The design levels are adjusted to accommodate the radiation damage

resistance of specific materials.


1.3.9 Structural Loading Criteria

Structures and equipment are designed to substantially resist mechanical damage

due to loads produced by mechanical and thermal forces. For the purpose of

categorizing mechanical strength designs for these loads, the following

definitions were established:

1. Class I

Class I includes those structures, equipment, and components whose failure

or malfunction might cause or increase the severity of an accident which

would endanger the public health and safety.

2. Class II

Class II includes those structures, and components which are important to

the safe storage and handling of irradiated fuel and radioactive waste, but

are not essential for preventing or mitigating the consequences of an

accident which would endanger the public health and safety.

The loading categories are generically described and their meaning is expanded

in Section 3.


1.4 SUMMARY OF RADIATION EFFECTS

1.4.1 Fuel Storage and Handling and Waste Management

Spent fuel storage and handling and waste management operations will be

conducted so that the dose to any off-site person, from external or internal

sources, will not exceed that permitted by 10 CFR 20.1301. It is expected that

during fuel storage and handling and waste management operations the dose to

any off-site person from gaseous waste discharge will not average more than

about 1% of the permissible dose, and that concentrations of liquid waste at

the point of discharge will average less than the concentrations permitted by

10 CFR 20. Both effects are only a small fraction of the effect of natural

background radiation.

For ISFSI operations, 10 CFR 72.106(b) defines the dose that any individual

located on or beyond the nearest boundary of the controlled area may receive

from any design basis accident associated with the ISFSI. For additional

information, see the VYNPS 10 CFR 72.212 Evaluation Report.

1.4.2 Accidents and Events

The ability of the station to withstand the consequences of accidents and

events without posing a hazard to the health and safety of the public is

evaluated by analyzing a fuel handling accident in the spent fuel pool and a

radwaste transfer cask drop event. The calculated consequences are

substantially below the dose limits given in 10 CFR 50.67 for the fuel handling

accident and 10CFR100 for the transfer cask drop event. A further description

is provided in Section 6.

1.5 GENERAL CONCLUSIONS

Based on the design of the facility and the analysis of credible events, there

is reasonable assurance that the facility can safely manage irradiated fuel and

radioactive waste without endangering the health and safety of the public.


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SITE AND ENVIRONS

TABLE OF CONTENTS Section Title Page

2.1 SUMMARY DESCRIPTION .................................................. 9

2.2 SITE DESCRIPTION ..................................................... 9

2.2.1 Location and Area ........................................... 9

2.2.2 Population ................................................. 10

2.2.3 Land Use ................................................... 10

2.2.4 Site Area Boundaries, Exclusion Area, and Low Population Zone ........................................ 12

2.2.5 Conclusions ................................................ 15

2.3 METEOROLOGY ......................................................... 22

2.3.1 General .................................................... 22

2.3.2 On-site Meteorological Programs ............................ 22

2.3.3 Diffusion Climatology ...................................... 22

2.3.4 Winds and Wind Loading ..................................... 23

2.3.5 Temperature and Precipitation .............................. 23

2.3.5.1 Temperature .................................... 23

2.3.5.2 Precipitation .................................. 24

2.3.5.3 Snowfall, Snow and Ice Loading ................. 24

2.3.6 Storms ..................................................... 26

2.3.6.1 Thunderstorms .................................. 26

2.3.6.2 Hurricanes ..................................... 27

2.3.6.3 Tornadoes ...................................... 27

2.3.7 Conclusions ................................................ 28

2.3.8 References ................................................. 29

2.4 HYDROLOGY AND BIOLOGY ............................................... 37

2.4.1 General .................................................... 37

2.4.2 Land Area Ground Hydrology ................................. 37


2.0-2 of 108

2.4.2.1 Introduction ................................... 37

2.4.2.2 Surface Water .................................. 37

2.4.2.3 Groundwater .................................... 37

2.4.3 Hydrology .................................................. 38

2.4.3.1 Introduction ................................... 38

2.4.3.2 Stream Flow .................................... 38

2.4.3.3 Temperature .................................... 39

2.4.3.4 Floods ......................................... 39

2.4.4 Uses of River .............................................. 46

2.4.4.1 Introduction ................................... 46

2.4.4.2 Industrial Use ................................. 46

2.4.4.3 Public Use ..................................... 46

2.4.5 Biology .................................................... 47

2.4.5.1 Commercial Fisheries ........................... 47

2.4.5.2 Sport Fisheries ................................ 48

2.4.5.3 Bottom Fauna ................................... 48

2.4.5.4 Aquatic Plants ................................. 49

2.4.5.5 Conclusions .................................... 49

2.4.6 Chemical and Bacteriological Quality of Water ...................................................... 49

2.4.7 River Field Program ........................................ 50

2.4.8 Conclusions ................................................ 50

2.4.9 References ................................................. 52

2.5 GEOLOGY AND SEISMOLOGY .............................................. 76

2.5.1 General .................................................... 76

2.5.2 Geology .................................................... 76

2.5.2.1 Introduction ................................... 76

2.5.2.2 Geological Investigation Program ............... 76

2.5.2.3 Regional Geology ............................... 77

2.5.2.4 Site Geology ................................... 79


2.0-3 of 108

2.5.2.5 River Geology .................................. 81

2.5.3 Seismology ................................................. 83

2.5.3.1 Introduction ................................... 83

2.5.3.2 Seismic Investigation Program .................. 83

2.5.3.3 Geologic and Tectonic Background ............... 83

2.5.3.4 Seismic History ................................ 83

2.5.3.5 Seismicity of Area ............................. 85

2.5.4 Conclusions ................................................ 86

2.6 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ...................... 104

2.6.1 Objectives ................................................ 104

2.6.2 Monitoring Network ........................................ 105

2.6.2.1 Direct Radiation .............................. 105

2.6.2.2 Airborne ...................................... 106

2.6.2.3 Waterborne .................................... 106

2.6.2.4 Ingestion ..................................... 107

2.6.3 Land Use Census ........................................... 107

2.6.4 Emergency Surveillance .................................... 107

2.6.5 Reports ................................................... 108


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STATION SITE AND ENVIRONS LIST OF TABLES Table No. Title 2.2.4 Urban Centers Within 30 Miles of Site 2.2.7 Hazardous Materials Railroad Traffic Through Vernon,

Vermont 2.3.1 Meteorology Record 2.3.4 Winds During Thunderstorms 2.3.5 Rainfall Data from Hurricane Connie 2.4.1 Average and Extreme Values of Stream Flow Connecticut

River at Vernon, Vermont Water Years 1944-1988 2.4.2 Vermont Yankee Nuclear Power Station, Daily Stream Flow

for October 1964 to September 1965, Connecticut River at Vernon, Vermont

2.4.3 Municipal and Industrial Groundwater Usage Within a

10-Mile Radius of the Vernon Site 2.4.4 Public Water Supplies Within a 10-Mile Radius of the

Vernon Site 2.4.5 Water Supplies Within a l-Mile Radius of the Site 2.4.6 Six-Hour PMP and Runoff Increments - Connecticut River

Basin above Vernon, Vermont 2.4.7 Maximum Annual Floods on Connecticut River at Vernon,

Vermont - Arranged in Descending Order (1927, 1936, 1938, 1945-1973)

2.4.8 Time - Varying PMF Stage - Discharge Table Vermont Yankee

Nuclear Plant Site 2.4.9 Time - Varying Modified PMF Stage - Discharge Table

Vermont Yankee Nuclear Plant Site 2.4.10 Checklist of Connecticut River Fishes Found Near Vernon,

Vermont 2.4.11 Fishes of the Connecticut River in the Vicinity of Vernon,

Vermont - All Collections, 1980


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STATION SITE AND ENVIRONS LIST OF TABLES (Cont'd) Table No. Title 2.5.1 Available Information Concerning Geology and Seismic

Activity Related to the Vermont Yankee Nuclear Power Station Site

2.5.2 Vernon Pluton: Estimated Mode of the Oliverian Magma

Series


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STATION SITE AND ENVIRONS LIST OF FIGURES Reference Figure No. Drawing No. Title 2.2-1 Location Map - 2-Mile Radius 2.2-2 Location Map - 10-Mile Radius 2.2-3 Location Map - 25-Mile Radius 2.2-5 5920-6245 Plan Showing Exclusion Area and

Restricted Area Boundaries 2.3-1 Station Site - Westover AFB,

Massachusetts Area - Annual Surface Windrose

2.3-2 Station Site - Westover AFB, Massachusetts Area - Seasonal Surface Windroses – (Winter, Spring, Summer, Fall)

2.3-3 Station Site - Concord, NH Area - Return Period of Rainfall (for extremely short intervals)

2.4-1 Station Site - Area Public Water

Supplies - 10-Mile Radius 2.4-2 Station Site - Area Private Water

Supplies - 1-Mile Radius 2.4-3 Enveloping Depth-Duration-Area Values

of PMP for Susquehannna River Basin 2.4-4 6-Hour Unit Hydrograph 2.4-5 Total SPF Hydrograph 2.4-6 Total PMF Hydrograph (Natural and

Modified) 2.4-8 Vermont Yankee Nuclear Plant - Location

of River Cross-Sections 2.4-9 Stage-Discharge Curve at the Vermont

Yankee Nuclear Plant


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STATION SITE AND ENVIRONS LIST OF FIGURES (Cont’d) Reference Figure No. Drawing No. Title 2.4-10 Cross Section of the Critical Fetch

2.4-11 Vermont Yankee Sample Stations on

Connecticut River 2.5-1 Not Used 2.5-2 Station Site - Geological Survey -

General Plan - Location of Test Borings 2.5-3 Station Site - Geological Survey -

Subsurface Profile - Log of Test Borings (1A, 2A, 3A, 4, 5, 8)

2.5-4 Station Site - Tectonic Map - State of

Vermont 2.5-5 Station Site - Tectonic Map - State of

New Hampshire 2.5-6 Station Site - Geological Survey - Area

Bedrock Geology 2.5-7 Station Site - Geological Survey - Area

Geological Section 2.5-8 Station Site - Geological Survey -

Subsurface Profile (Section AA) - Log of Test Borings (5, 8, S9, 11, and 21)

2.5-9 Station Site - Geological Survey -

Subsurface Profile (Section BB) - Log of Test Borings (2A, 3A, ST6-1/2, and S9)

2.5-10 Station Site - Geological Survey - Subsurface Profile (Section CC) - Log of Test Borings (2, 2A, 5, 7, 7A, 13, and 15)


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STATION SITE AND ENVIRONS LIST OF FIGURES (Cont’d) Reference Figure No. Drawing No. Title 2.5-11 Station Site - Geological Survey -

Subsurface Profile (Section DD) - Log of Test Borings (3, 3A, 4, 8, 8A, 12, and 16)

2.5-12 Station Site - Tectonic Map - New

England Area 2.5-13 Station Site - Compilation of

Earthquakes - New England Area 2.5-14 Station Site - Earthquake Intensity -

Modified Mercalli and Rossi - Forel Scales

2.5-15 Station Site - Compilation of

Earthquakes - Central New England Area


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2.1 SUMMARY DESCRIPTION

HISTORICAL INFORMATION BELOW NOT REQUIRED TO BE REVISED.

This section provides information about the site and environs of the

Vermont Yankee Nuclear Power Station (VYNPS) and summarizes the analyses

and studies which confirm the suitability of the site

The site of the VYNPS at Vernon, Vermont, was thoroughly investigated and

found to be suitable in 1967 when the construction permit was issued.

Since the issuance of the construction permit, further review has been

pursued in the areas of meteorology, hydrology, and marine ecology, geology

and seismology, and environmental radiation monitoring. The results of

this additional review confirmed the suitability of Vernon as a nuclear

power plant site.

2.2 SITE DESCRIPTION

2.2.1 Location and Area

The site is located in the town of Vernon, Vermont in Windham County on the

west shore of the Connecticut River immediately upstream of the Vernon

Hydroelectric Station. The site contains about 125 acres owned by Entergy

Nuclear Vermont Yankee, LLC and a narrow strip of land between the

Connecticut River and the east boundary of the VYNPS property to which

Entergy Nuclear Vermont Yankee, LLC has perpetual rights and easements from

its owner. This land is bounded on the north, south, and west by

privately-owned land and on the east by the Connecticut River. Site

coordinates are approximately 42o 47' north latitude and 72o 31' west

longitude. Figures 2.2-1 through 2.2-3 locate the site. The site plot

plan, exclusion area boundary and site area boundaries for both gaseous and

liquid effluents are shown on Drawing 5920-6245.


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2.2.2 Population


The population density for 1990 was estimated to be about 121 people per

square mile within a five-mile radius of the site. The population density

in this same area was estimated to be 126 people per square mile in 2000,

and projected to be about 131 people per square mile by 2010. In 1990, the

total population within 25 miles was estimated to be 189,038, or an average

density of 96 people per square mile. For 2000, the 25-mile radius

population has been estimated to be about 193,746, or an average density of

99 people per square mile. This represents a growth factor of about 2.5%

for 2000 area over the ten-year period 1990 to 2000. The total resident

population within 50 miles for 2000 is estimated to be about 1,467,343.

Based on this region's projected growth rate of 4% over the next 10 years,

the estimated 50-mile population for the year 2010 is 1,526,037.

The nearest towns with populations of 25,000 or more are Northampton,

Massachusetts (2000 population 28,978) at about 30 miles to the south; and

Amherst, Massachusetts (2000 population 34,874) at about 28 miles south.

Accordingly, 28 miles is the population center distance.

2.2.3 Land Use


About 80% of the land within a 25-mile radius of the site is undeveloped.

Most of the developed land is used for agriculture and dairying, with homes

scattered or grouped in small villages.

The primary agricultural crop in the immediate site area is silage corn

which is stored for year-round feed for milk cows.

The area within 10 miles of the site has only one urban area, the city of

Brattleboro, Vermont (2000 population 12,005), which is located about 5

miles upriver. The remainder of this area is rural and contains several

small villages with populations between 1,000 and 3,000. The area between

10 and 25 miles has only three urban centers with 2000 populations between

11,299 and 22,563 (see Table 2.2.4).


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The closest site boundary is 910 feet west of the Reactor Building. The

nearest homes are situated along the Governor Hunt Road just west of the

site. An annual land use census checks on the location of the nearest

resident and reports this finding as part of the Annual Radiological

Environmental Operating Report. The Vernon Elementary School, which has a

pupil enrollment of about 250 is on the other side of the road (Highway No.

4) about 1,500 feet from the Reactor Building.

The nearest hospital, Brattleboro Memorial, is approximately five (5) miles

from the site. The nearest dairy farm is approximately 1/2-mile

west-northwest of the site and there are several others within a 5-mile

radius of the plant. The nearest railroad line runs north-south through

the site area, and is approximately 0.5 miles west of the plant at its

closest approach. Table 2.2.7 lists the approximate quantities of

hazardous materials which are annually shipped past the site by the

Springfield Terminal Railway and the Central Vermont Railway which utilize

this track. No other significant off-site sources of hazardous materials

have been identified within five (5) miles of the site.

The land within a 1-mile radius of the site is occupied by rural homes and

is used for dairy feed products and pasture, except for a residential area

of about 75 houses located about 0.8 miles across the Connecticut River.

About 30% of this area consists of the river and undeveloped land adjacent

to it.


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2.2.4 Site Area Boundaries, Exclusion Area, and Low Population Zone

As defined in 10 CFR 20 and 10 CFR 100, the terms "unrestricted area,"

"controlled area," "restricted area," "exclusion area," and "low population

zone" each refer to a specific area about the site as a result of applying

different radiological health constraints. The "unrestricted area" refers

to all areas beyond the site's outer security fence access to which is

neither limited nor controlled by the licensee. The "controlled area"

refers to all plant areas inside the site boundary, but outside of any

restricted area, access to which is limited by the licensee for any reason.

Access to the controlled area can be limited to minimize exposures to

members of the public from routine radioactive releases from the plant and

fixed radiation sources. "Restricted area" refers to the inner most areas

of the plant site and facilities, access to which is limited by the

licensee for the purpose of protecting occupationally exposed individuals

against undue risks from radiation and radioactive materials. Exclusion

area means that area surrounding the reactor, as measured from the reactor

center line, in which the reactor licensee has the authority to determine

all activities including exclusion or removal of personnel and property

from the area. This area may be traversed by a highway, railroad, or

waterway, provided those are not so close to the facility as to interfere

with normal operations of the facility and provided appropriate and

effective arrangements are made to control traffic on the highway,

railroad, or waterway, in case of an emergency, to protect the public

health and safety. The exclusion area also includes part of the adjacent

waterway (Connecticut River) extending across to the opposite shoreline.

Finally, the low population zone is delineated by an area about the plant

which includes residential, farming, industrial, etc., activities to some

extent, but is not so large or populated to prevent orderly, effective

radiological control or evacuation in the event of an accident of an

environmentally significant nature.

Thus, these areas and zones are delineated for different purposes and vary

in the degree of control that the licensee can exercise from a radiation

protection standpoint. The following discussion presents an analysis of

each area in relation to the plant and its operations.


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1. Controlled Area

The controlled area for the VYNPS site consists of a significant portion

of the 125-acre property area owned by Entergy Nuclear Vermont Yankee,

LLC. The fenced boundaries of this area are delineated on Drawing 5920-

6245. The fence is a 6-foot high security fence topped by l foot of

barbed wire. In addition to the fence, signs are posted clearly

informing an individual that the area is private property and

unauthorized entry is strictly prohibited. Access to and activities

within this area are under the direct control of Entergy Nuclear Vermont

Yankee, LLC and Entergy Nuclear Operations, Inc. Access to the area is

from the Governor Hunt Road through the main gate. The fence and

location combine to afford access and activity control to the VYNPS

site.

Two normally locked gates exist in the northern corners of the

Controlled Area for access by Security Officers from the Controlled Area

into the Exclusion Area on the northern part of the property. One gate

is located along the east fence line and one gate is located along the

west fence line. The gate on the west fence may also be used for

alternate access to the site for fire trucks.

For ISFSI operations, 10 CFR 72.106(b) defines the dose that any

individual located on or beyond the nearest boundary of the controlled

area may receive from any design basis accident associated with the

ISFSI. For additional information, see the VYNPS 10 CFR 72.212

Evaluation Report.

2. Effluent Boundaries

In addition to the land area within the site's outer security fence,

VYNPS includes the river water area between the northern and southern

boundary fences, and extending out to the state border near the middle

of the river, as part of the site boundary for control of gaseous

effluents as regulated under the dose objectives of 10 CFR 50, Appendix

I. The low exposure rates involved and the zero or near zero occupancy

factor applicable to individuals in the river area combine to allow

VYNPS to include this region for the purpose of controlling plant

releases to levels as-low-as-reasonably achievable. The restricted area

boundary for liquid discharge concentration limits (10 CFR 20) is set at

the point of discharge from the plant to the river (see Drawing 5920-

6245). Thus, the overall boundary area for the plant is as shown on

Drawing 5920-6245.


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To ensure compliance with the constraints applicable to the unrestricted

and controlled areas as described, area dosimeter stations are provided

at strategic locations around the site. Measurements of integrated

gamma exposure are made to alert VYNPS to any condition that may produce

a greater exposure than necessary.

3. Exclusion Area

The exclusion area for the VYNPS site is also shown on Drawing 5920-6245

and includes the controlled area defined above. The minimum distance to

the boundary of the exclusion area, as measured from the reactor center

line, is 910 feet. In addition, the Connecticut River water area

between Vernon Dam and the northern VYNPS property line is included in

the exclusion area since it will be a controlled access region during an

accident condition. The means of controlling access on the river, and

evacuating it if necessary, have been worked out with the State of New

Hampshire officials who will coordinate control activities over the

river.

Passage on the Connecticut River to Vernon Pond is possible. The

licensee will at all times retain the complete authority to determine

and maintain sufficient control of all activities through ownership,

easement, contract and/or other legal instruments on property which is

closer to the reactor center line than 910 feet. This includes the

authority to exclude or remove personnel and property within the

exclusion area. Only facility related activities are permitted in the

exclusion area. No residences will be permitted in the exclusion area.

Control over activities within, and access to, the exclusion area assume

an entirely different form immediately following a condition that

produces, or threatens to produce, a radiological hazard to the site.

The VYNPS Emergency Plan describes the types and level of emergency

action that will be initiated at the plant in order to minimize

radiation exposure following an accidental release. The only addition

to that discussion is that, as previously mentioned, evacuation and

access control will be placed into effect for the Connecticut River area

included in the exclusion zone.

A normally locked gate on the northwest corner of the Exclusion Area

fence is used for access by Vermont Electric Power Co for access to

their switchyards, and is also used by VYNPS as an alternate access to

the site for fire trucks and emergency equipment.


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4. Low Population Zone

The low population zone for the VYNPS is the area included within a

5-mile radius of the site. It is outlined on Figure 2.2-2.

5. General

The boundaries for the unrestricted area, controlled area, restricted

area, exclusion area, and low population zone, as well as for control of

effluents to levels as-low-as-reasonably achievable, as described, are

fully consistent with the principles involved in ensuring the health and

safety of the public, together with the plant personnel. In addition,

the delineation yields an effective arrangement with regard to efficient

facility operation.

The complete perimeter fence described for the protected area, together

with the fact that the only facility access point is maintained by the

security force, afford the licensee with complete, continuous access and

activity control for every component of the facility. In addition,

fencing is provided for the 115 kV and 345 kV switchyards.

Thus, the responsibilities of the licensee are met from both

radiological protection and plant security standpoints.

2.2.5 Conclusions


About 80% of the land within 25 miles of the site is undeveloped. The 2000

census shows that about 489 people live within 1 mile of the site and about

9,919 live within 5 miles. The 2000 data also show that population density

in the vicinity is light, about 126 persons per square mile within a 5-mile

radius and 99 persons per square mile within a 25-mile radius. Population

projections to 2010 predict about a 4% increase above the 2000 figures.

However, the average population density is expected to remain low. The

location of the site provides good local isolation with light population

density in the surrounding area.

In summary, the site is suitable for the facility as designed from

population distribution and land usage considerations.


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TABLE 2.2.4

Urban Centers Within 30 Miles of Site

City Approximate Distance

From Site - Miles

1960 1970 1980 1990 2000

Brattleboro, VT 4 9,315 12,239 11,886 12,241 12,005

Greenfield, MA 12 14,389

18,116 18,415 18,666 18,168

Keene, NH 13 17,562

20,467 21,449 22,430 22,563

Athol, MA 19 10,161

11,185 10,619 11,451 11,299

Amherst, MA 28 13,718

26,331 33,210 35,228 34,874

Northampton, MA 30 30,058

29,669 29,128 29,289 28,978


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TABLE 2.2.7

Hazardous Materials Railroad Traffic Through Vernon, Vermont

Chemical(1) Central Vermont(2)

Springfield Track(2)

Total Per Year

Carbon Dioxide 395 96 491

Nitrogen 248 -- 248

Propane (LPG) 60 162 222

Chlorine 60 -- 60

Sulfuric Acid -- 24 24

Anhydrous Ammonia 1 6 7

Methyl Alcohol -- 4 4

Xylene -- 2 2

(1) Listed in either Regulatory Guide 1.78 or EPA's Extremely Hazardous

Substance List.

(2) Railcars per year.


Vermont Yankee Defueled Safety Analysis Report

Location Map – 2-Mile Radius

Figure 2.2-1


Vermont Yankee

Defueled Safety Analysis Report

Location Map – 10-Mile Radius

Figure 2.2-2


Vermont Yankee


Location Map – 25-Mile Radius Figure 2.2-3

DSAR Revision 1

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2.3 METEOROLOGY


2.3.1 General

The general climatic regime of the site area is that of a continental type

with some modification from the maritime climate which prevails nearer the

coast. Of special importance from an engineering standpoint is a temperature

range of 133oF for the period of record; extremes in annual snowfall, which

may be as little as 30 inches or as much as 118 inches; occasional ice storms;

occasional severe thunderstorms; occasional heavy rains due to hurricane

influences; and the possibility of an occasional tornado. These and other

pertinent meteorological data are presented in the following subsections.

Table 2.3.1 indicates the elements, station of record and lengths of record

that were utilized in the analyses.

The site meteorological monitoring program is the most important source of

additional information obtained since the submittal of the Plant Design and

Analysis Report (PDAR).

2.3.2 On-site Meteorological Programs

An initial data collection program was undertaken at the site of the Vermont

Yankee Atomic Power Station to provide information on meteorological

conditions for dispersion analysis for the PDAR. Data from one year, from

August 1, 1967 through July 31, 1968, were evaluated and formed the basis for

those analyses. Appendix G contains a discussion of the August 1967 - July

1968 data collected from the initial monitoring program.

An upgraded on-site monitoring program which meets the intent of Revision 0 to

Regulatory Guide 1.23 was installed in early 1976 and is currently in

operation. A description of this upgraded system is also presented in

Appendix G, along with wind and stability data summaries for one year of

operation.

2.3.3 Diffusion Climatology

The river valley location of the site exerts a strong influence on wind

distribution. As seen in the various wind roses of Appendix G, the channeling

effect of the valley is readily apparent. However, there is no appreciable

difference in wind distribution during poorer dispersion conditions.


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The PDAR stated inversion frequency estimates on the basis of C. R. Hosler's

tabulations(1). These are repeated below together with the corresponding

values determined from the site preoperational meteorological program. As

shown, the maximum difference occurred in the spring season.

Inversion Frequency (% of total hours)

Hosler's

Estimates

Site Meteorological

Program

Winter 33 37

Spring 26 42

Summer 31 37

Fall 36 36

2.3.4 Winds and Wind Loading

At the time the PDAR was submitted, no continuous wind records were available

for the Vernon area. Due to the similarity in terrain, the relatively close

location, and ready availability of information, wind data from Westover,

Massachusetts, was presented at that time. The annual and seasonal surface

wind roses from Westover are shown in Figures 2.3-1 and 2.3-2. The annual and

seasonal wind roses are based upon the total possible hours for each time

interval specified and, in each case, add to 100%.

The corresponding annual and seasonal wind roses obtained from the site

monitoring programs are shown in Appendix G. The several sets of wind roses

show the same channeling effect due to topographical similarities.

The minimum allowable resultant wind pressure(2) at 30 feet suggested by the

National Bureau of Standards for the Vernon area is 25 lb-ft-2. This value

was used as the general facility design basis.

2.3.5 Temperature and Precipitation

2.3.5.1 Temperature

Temperature data(3,4,5) from the records of Vernon (one-half mile south) and

Brattleboro (6 miles north) should be representative of the values for the

site.


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The mean number of days with temperatures greater than 90°F or less than 32°F

for Vernon (1951-1960) are as follows:

2.3.5.2 Precipitation

Precipitation(6) at the site averages 43 inches per year and is distributed

rather evenly throughout the 12-month period. Snowfall is moderately heavy on

the average; but there is considerable variation in amounts from season to

season. Nearly all winter precipitation is in frozen form, although not

entirely as snow. Sleet and freezing rain are not uncommon.

Intense rainfall will be produced by the occasional severe thunderstorm or

modified hurricane. The maximum (8,9) recorded rainfall (inches) for short

time intervals at Concord, New Hampshire, is given below:

Minutes Hours

5 10 15 30 60 2 3 6 12 24

0.66 1.12 1.60 2.53 2.71 2.73 3.56 3.82 5.53 5.97

The return period of extreme short-interval rainfall is a useful

design-and-planning guide. The nearest location for which return data are

available and which should be reasonably representative for the Vernon area is

Concord, New Hampshire. These data are shown in Figure 2.3-3.

2.3.5.3 Snowfall, Snow and Ice Loading

The site being located in the northeastern part of the United States is

subjected to a wide range of snowfall, which may be as little as 30 inches or

as much as 118 inches(5,10). Average snowfall statistics for Vernon (25 years

of record) are considered to be representative of the site.

* More than 0 but less than 0.5

Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Ann

>90° 0 0 0 0 * 3 6 3 1 0 0 0 13

<32° 30 28 29 14 6 * 0 0 2 13 23 30 175


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The most significant departure from the historical values occurred in the

amount of snowfall at Vernon between November 1968 and February 1969.

Snowfall during this period amounted to 80.2 inches compared with an average

for this period of 45.9 inches. The heaviest monthly snowfall was 42.7 inches

and occurred in February. This compares with a historical average value of

15.7 inches. However, the maximum annual snowfall of 118 inches was not

exceeded.

Average Monthly Snowfall (inches) for Vernon

Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Ann

16.4 15.7 12.1 2.1 0.1 0.0 0.0 0.0 0.0 T 3.3 10.5 60.0

T = Trace

Snow load data(11) available from a Housing and Home Finance Agency (HHFA)

study conducted in 1952 is as follows:

Weight of Seasonal Weight of Maximum Weight of Estimated

Snowpack Equaled or Snowpack Maximum Accumulation on Ground Plus

Exceeded 1 year in 10 of Record Weight of Maximum Possible Snowstorm

30 lb-ft-2 50 lb-ft-2 70 lb-ft-2

Data relating to freezing rain and resultant formation of glaze ice(12) on

highways and utility lines are available from the following studies:

American Telephone and Telegraph Company, 1917-18 to 1924-25

Edison Electric Institute, 1926-27 to 1937-38

Association of American Railroads, 1928-29 to 1936-37

Quartermaster Research and Engineering Command, U.S. Army, 1959

The U.S. Weather Bureau also maintains annual summaries. The conclusions

reached from these several sources are sometimes contradictory, but the

following is probably a fairly accurate description of the glaze-ice

climatology of southern Vermont.


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The most typical synoptic condition for glaze formation or freezing rain in

the northeastern United States is a polar front wave with an active warm front

moving in the north or northeasterly direction toward the region. A high

pressure area almost always is found north of New England, with the center of

the ridge or high pressure cell usually located somewhere northeast of

Newfoundland. This distribution causes a flow of cold continental-polar air

over the area from the north or east, and warm maritime-tropical air up from

the south behind the warm front.

In this situation, the over-running maritime-tropical air is frequently warmer

than 32oF, while the cold continental-polar air beneath the front has

temperatures from 20o to 30oF, and a situation almost ideal for the formation

of freezing rain or drizzle results. The Vermont site is situated on the

northern edge of the "glaze belt" which extends from southern New England

west-southwest to Ohio and then curving down into Texas. The following data

will apply:

1. Times of occurrence - November through April,

2. Average frequency without regard to ice thickness - 0-10 storms per year,

3. Duration of ice on utility lines - 20 hours (mean) to 55 hours (maximum

of record),

4. Return periods for freezing rain storms producing ice of various

thicknesses are:

Ice 0.25 inch every year

0.50 inch every year

0.75 inch at least every 3 years

A U.S. Weather Bureau summary for the years 1939-48 give the actual number of

days with freezing rain (without regard to ice formation) for Concord, New

Hampshire, as follows:

Total Days in

Nov Dec Jan Feb Mar Apr 10 Years

2 24 29 23 16 1 95

2.3.6 Storms

2.3.6.1 Thunderstorms

Some localized wind damage occurring with the passage of thunderstorm line

squalls may be experienced each year. Extreme wind data(13) for Westover,

Massachusetts, is shown in Table 2.3.4.


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The "Index of Wind Damage Potential" (excluding tornado, hurricane, and

tropical storm and hail)(defined in units of 1000ths of 1% of residential

property value per year) for the Vernon area is 12 compared to a value of 16

for the Oklahoma-Kansas area.

Heavy precipitation is usually associated with severe thunderstorms and

modified hurricanes. The maximum in 24 hours for Vernon (62 years of record)

is listed below.(8)

Maximum in 24 Hours (inches)

Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec

2.21 2.89 4.35 2.49 3.23 3.50 3.80 4.35 3.99 3.57 3.13 2.21

2.3.6.2 Hurricanes

Unusual heavy precipitation(14) was associated with hurricane Connie (August

11-14, 1955) and Diane (August 17-20, 1955). Mass rainfall tables for Birch

Hills, Massachusetts (approximately 28 miles south) are in Table 2.3.5.

In "Index of Hurricane and Tropical Storm Damage Potential" (defined in units

of 1000ths of 1% of residential property value per year) for the Vernon area

is 140 as compared to 337 for the Cape Cod area, 606 for the Cape Hatteras,

North Carolina area, and 633 for the Miami, Florida area. The decrease in the

index of hurricane potential as one moves northward is indicative of the

decreased intensity of the hurricane due to several physical reasons. Being

cut off from the major source of energy (the ocean) as a hurricane proceeds

northward, it diminishes in intensity. Topography also causes frictional drag

the farther the storm travels over land, thereby reducing the storm's

magnitude.

2.3.6.3 Tornadoes

Severe storms such as tornadoes(15) are not numerous, but they do occur

occasionally. Most tornadoes that occur in New England occur in

Massachusetts.

Massachusetts Vermont New Hampshire

Total Number of

Tornadoes (1916-1958) 56 11 15

(1959-1965) 23 12 22


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The apparent increase in tornado activity is probably due to increased

population and more and better observing and reporting facilities and

techniques.

The monthly distribution (1916-1965) of tornadoes for the tri-state area is as

follows:

Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Total

Massachusetts 12 15 26 9 6 6 3 2 79

Vermont 1 1 3 7 8 2 1 23

New Hampshire 6 8 14 6 2 1 37

In the period 1916 through 1965, Bennington County, Vermont, has reported only

2 tornadoes. Cheshire County, New Hampshire, reported 8, and Franklin County,

Massachusetts reported 9 for a total of 19 tornadoes for the immediate area.

The "Index of Tornado Damage Potential" (defined in units of 1000ths of 1%

residential property values per year) for the tri-county area is 1 as compared

to a value of 33 in "tornado alley" (Oklahoma-Kansas-Nebraska).

Thom(16) divides the United States into 1-degree squares and determines the

tornado frequency for each square. Using data from 1953-62, Thom records 12

tornadoes occurring within a 1-degree square (about 3 million acres)

encompassing the Vernon site. A mean recurrence interval for a tornado

striking a point within this 1-degree square was calculated to be 1040 years.

This seems reasonable if one considers that only 12 tornadoes were reported in

about 3 million acres in a 10-year period.

Even though the probability of a tornado at the site is small, all structures

and equipment necessary for the safe storage of irradiated fuel are designed

to withstand short-term loadings resulting from 300 mph tornadic winds and an

external pressure drop of 3 psi in 5 seconds.

2.3.7 Conclusions

The meteorology of the site is basically that of a continental type with some

modification from the maritime climate which prevails nearer the coast. The

annual frequency of inversion was determined to be 39%, within the 30% to 40%

range predicted in the PDAR.

The average annual wind speed for the site is 7.5 mph and the most frequent

direction is NNW, the downriver direction. The river valley location leads to

a channeling of the winds.


2.0-29 of 108

In summary, the site meteorological program substantiates the preliminary

conclusions. No changes in the previously described protection features were

necessary as a result of meteorological considerations.

2.3.8 References

1. "Low-Level Inversion Frequency in the Contiguous United States," Charles

R. Hosler, Monthly Weather Review, Vol. 89, No. 9, September 1961, pp.

319-339.

2. "Wind Pressures in Various Areas of the United States," Building

Materials and Structures, Report 152. National Bureau of Standards,

1959.

3. Climatological Data, New England, July Issue for 1962-1965 (four

publications), U.S. Weather Bureau.

4. Climatic Summary of the United States - Supplement for 1931 through 1952,

New England, U.S. Weather Bureau.

5. Climatic Summary of the United States - Supplement for 1951 through 1960,

New England, U.S. Weather Bureau.

6. "Rainfall Intensity - Duration - Frequency Curves", Technical Paper No.

25, U.S. Weather Bureau, 1955.

7. Deleted

8. "Maximum 24-Hour Precipitation in the United States," Technical Paper No.

16, U.S. Weather Bureau.

9. "Maximum Recorded United States Point Rainfall for 5 Minutes to 24

Hours," Technical Paper No. 2, U.S. Weather Bureau.

10. Climatological Data, New England, July Issues 1961-1965, U.S. Weather

Bureau (five publications).

11. "Snow Load Studies," Housing Research Paper 19, Housing and Home Finance

Agency, 1952.

12. "Glaze, Its Meteorology and Climatology, Geographical Distribution, and

Economic Effects," Quartermaster Research and Engineering Center, 1959.


2.0-30 of 108

13. Climatological Data, National Summaries (1959-60-61-62-63-64-65), U.S.

Weather Bureau.

14. "Hurricane Rains and Floods of August 1955, Carolinas to New England,"

Technical Paper No. 26, U.S. Weather Bureau.

15. "Tornado Occurrences in the United States," Technical Paper No. 20, U.S.

Weather Bureau.

16. "Tornado Probabilities," H.C.S. Thom, Monthly Weather Review, U.S.

Weather Bureau, Washington, D.C., October-December 1963, pp. 730-736


2.0-31 of 108

TABLE 2.3.1 METEOROLOGY RECORD Weather Element Station Record Period Temperature Brattleboro 11 years Vernon 10 years Precipitation Vernon 62 years Snowfall Vernon 25 years Surface Wind Westover, MA 22 years Surface Wind Vernon 1 year Stability Class Vernon 1 year


2.0-32 of 108

TABLE 2.3.4

WINDS DURING THUNDERSTORMS Westover, MA Max. Winds(1) Peak Gusts(2) (from hourly obs.) (from daily obs.) Speed Speed Month Direction (knots) Direction (knots) Jan S 52 NW 55 Feb NNW 40 S 58 Mar NE 63 NNW 50 Apr S 36 ENE 61 May SSE 38 NNW 48 June S 28 NNW 49 July NNW 28 W 47 Aug NNE 47 NNE 62 Sept NNE 52 N 60 Oct NW 37 SSE 49 Nov N 39 ENE 69 Dec N 39 NNW 54

(1) For period Apr 1941 through Dec 1963. (One minute sustained wind) (2) For period Jan-Apr 1946, Jan, Feb, Apr, June, July 1949, Jan, Apr 1950 through

Dec 1963


TABLE 2.3.5 RAINFALL DATA FROM HURRICANE CONNIE Birch Hill, MA Amherst, MA Time (Accumulative Inches) (Accumulative Inches) Aug. 11 - 6 AM 0.05 12 N 0.06 6 PM 0.14 2.15 12 M 0.24 2.25 Aug. 12 - 6 AM 1.00 3.35 12 N 1.40 3.90 6 PM 1.45 4.07 12 M 1.60 4.40 Aug. 13 - 6 AM 2.15 4.90 12 N 2.30 5.91 6 PM 4.30 7.65 12 M 6.30 7.70 Aug. 14 - 6 AM 6.39 7.70 12 N 6.40 7.70 6 PM 6.48 7.70 12 M 6.48 7.70 Aug. 15 - 6 AM 7.72 12 N 7.72 6 PM 7.72 12 M 7.73


Vermont Yankee


Station Site - Westover AFB,

Massachusetts Area- Annual Surface Windrose

i 2 3 1


Vermont Yankee


Station Site - Westover AFB,Massachusetts Area- Seasonal Surface Windroses (Winter – Spring – Summer – Fall)

Figure 2.3-2


Vermont Yankee Defueled Safety Analysis Report

Station Site – Concord, New Hampshire Area- Return Seasonal Surface Windroses

Period of Rainfall – (For extremely short intervals)

Figure 2.3-3


2.4 HYDROLOGY AND BIOLOGY


2.4.1 General

The site is at mile 138.3 above the mouth of the Connecticut River, located on

the west bank of the river, on the pond formed by the Vernon Dam and

Hydroelectric Station, licensed by the Federal Energy Regulatory Commission as

Project No. 1094. The site is about 3,500 feet upstream from the Vernon

Hydroelectric Station, on the same side of the river. The Vernon

Hydroelectric Station is the furthest downstream of a series of six

hydroelectric projects totaling over 456,000 kW on the river. Storage

reservoirs, whose contents total over 330,000 acre-feet, are also usable for

power generation.

Three of the dams, at 32, 75, and 132 miles above the site, are relatively low

structures developing heads of from 29 to 62 feet, with small amounts of

pondage. The large storage reservoirs are from 150 to 260 miles upstream from

Vernon.

2.4.2 Land Area Ground Hydrology

2.4.2.1 Introduction

The river in this general reach comprises a series of ponds formed by several

dams constructed for the generation of hydroelectric power.

There is sufficient groundwater in the area to provide wells for public as

well as private use.

2.4.2.2 Surface Water

All local streams in the area drain to the Connecticut River, and the site is

in the direct path of natural drainage to the east from the local watershed.

Surface drainage will flow toward the river.

2.4.2.3 Groundwater

2.4.2.3.1 Regional Area

There are several municipalities in the vicinity in which groundwater is

utilized as one source of water supply. These are listed in Tables 2.4.3 and

2.4.4 and shown in Figure 2.4-1. Private wells in the vicinity of the site

are listed in Table 2.4.5 and shown in Figure 2.4-2.


2.4.2.3.2 Site Area

The local water table level fluctuates differentially depending on the amount

of precipitation. It is affected by level changes in the Connecticut River.

River flooding will cause a temporary reversal in the flow direction of

groundwater, so that the local water table will be considerably higher than

usual during periods when the river level is high. Natural subsurface

drainage is over the rock surface.

In 1988 and 1989, groundwater monitoring wells were established throughout the

site area. Groundwater levels varied between about 5 feet to 18 feet below

ground surface in the northern portion of the site. In the vicinity of the

major plant structures, groundwater was determined to be about 20 feet below

ground surface. Along the southern portion of the site, depth to groundwater

was about 30 feet. Although these levels do vary throughout the year, they do

provide a general indication of site area groundwater levels.

Hydraulic gradients, as computed from water level elevations measured in

monitoring wells, bedrock water supply wells and the river, demonstrate that

groundwater flow in the overburden and bedrock is from west to east. Vertical

hydraulic gradients indicate vertically downward groundwater flow from the

shallow soils to the underlying lower sand deposit, and vertically upward flow

from the bedrock to the overlying lower sand deposit. These data indicate

that groundwater discharges into the river.

Current groundwater monitoring requirements are specified by the VYNPS

Radiological and Non-Radiological Environmental Monitoring Programs and

associated implementing procedures.

2.4.3 Hydrology


Under normal conditions, the flow of river water is largely determined by

operation of the hydroelectric stations and by the upstream reservoirs and

lakes.

2.4.3.2 Stream Flow

Connecticut River flow is monitored at Vernon Dam. Records were published by

the United States Geological Survey from 1944 to 1973 when their gage at

Vernon was discontinued. Drainage area at Vernon Dam is 6,266 square miles.

Nearby gages on the Connecticut River include N. Walpole, New Hampshire and

Turners Falls, Massachusetts. Their continuous periods of record are from

1942 and 1915 to the present at N. Walpole and Turners Falls, respectively.

The drainage areas at these two gages are 5,493 and 7,163 square miles.


Table 2.4.1 shows average and extreme values of monthly stream flow plus

minimum weekly flows for the Connecticut River below Vernon Dam for a 44-year

period of record (1944-1988). These stream flows were compiled using the

measured stream flows at Vernon from 1944 to 1973 and the generated stream

flows for Vernon using stream flow data from the two nearby gages for the

period 1973 to 1988.

Table 2.4.2 shows daily stream flows measured below Vernon for the period

October 1964 through September 1965.

2.4.3.3 Temperature

River temperatures have been measured along the river at six sampling

stations. The uppermost is at a point just downstream from Brattleboro, some

4-1/2 miles above the site, and the farthest downstream is at a point just

upstream of the Schell Bridge in Northfield, Massachusetts, some 6-1/2 miles

below the site.

Temperature measurements were made below Vernon Dam, in the general area of

the permanent monitoring Station 3, to determine how and where best to

establish this station for reliable, consistent results. A similar

temperature monitoring station (Station 7) has been established upstream of

the station circulating water intake. Continual records from these two

temperature monitoring stations are submitted annually to the Vermont Agency

of Natural Resources. The locations of these monitoring stations are

presented in Section 2.4.5.

In addition to the records presently being obtained at the site, temperature

records have been kept for a number of years at the Bellows Falls

Hydroelectric Station, some 32 miles upstream. River temperatures also have

been recorded over a period of several years at the Cabot Hydroelectric

Station, a unit of the Turners Falls Hydroelectric Project, downstream from

the site.

2.4.3.4 Floods

The flood of March 19, 1936, was the greatest and most destructive flood on

this reach of the river. The discharge on that day was 176,000 cfs, reaching

a river stage at Vernon of 231.4 feet MSL. Other major floods were those of

November 5, 1927, 155,000 cfs at elevation 229.0 feet MSL; and

September 22, 1938, 132,500 cfs at elevation 226.6 feet MSL.

Since the floods of 1936-1938, extensive flood control works, consisting of

some five projects with 247,800 acre-feet of flood storage, have been designed

and constructed by the Corps of Engineers in the Connecticut River Basin

upstream from the Vernon Dam.


The Probable Maximum Flood (PMF) on the Connecticut River Basin above Vernon,

Vermont (drainage area of 6,266 square miles) was determined using procedures

and information contained in the analytical studies for the Susquehanna River

(1). In addition, a stage-discharge curve was developed through step

backwater computations to determine the PMF elevation at the site.

In this study, the major emphasis was in the direction of conservatism. The

following conservative assumptions were made:

1. The maximum persisting 12-hour, 1000-millibar (mb) dew point temperature

of record is used as an index of the maximum precipitable water.

Furthermore, the 12-hour maximum persisting dew point was used throughout

the 72-hour rainfall period.

2. The unit of time selected for the unit hydrograph is 6 hours, although for

a basin area of 6,266 square miles and a lag time of 75 hours,

characteristic of the Connecticut River at Vernon, a more realistic unit

of time for the unit hydrograph would be 12 hours.

3. An infiltration rate of 0.05 inches per hour is assumed throughout the

rainfall period, although the recorded range for this particular basin is

0.05-0.10 inches per hour.

4. A baseflow of 58,800 cfs, which is about 5.7 times the average discharge

and greater than the annual peak discharge recorded in four of the 29-year

period of record, and about twice the value which is normally used.

Enveloping curves of PMP for 6, 12, 24, 48, and 72 hours were obtained by

adjustment of the depth-area-duration curves for the Susquehanna River Basin

(Figure 2.4-3). The adjustment is based on the precipitable water in the

1,000-200-mb air column (2) for the maximum persisting 12-hour, 1,000-mb dew

point of record (3). By applying the maximum persisting 12-hour, 1,000-mb dew

point of record and assuming this condition persists for an additional

60 hours (the PMP duration is 72 hours), there is a considerable amount of

conservatism in deriving the PMP for the basin.

At Harrisburg, Pennsylvania, the record maximum persisting 12-hour, 1,000-mb

dew point of 75.3F is equivalent to 2.93 inches of precipitable water, while

at Vernon, Vermont, it is 73.3F or 2.62 inches of precipitable water.

The 6-hour increments of PMP and runoff amounts, based on an infiltration rate

of 0.05 inches per hour, are presented in Table 2.4.6.

The 6-hour unit hydrograph (Figure 2.4-4) was derived from the Standard

Project Flood (SPF) hydrograph developed by the New England District of the

Corps of Engineers (Figure 2.4-5) by:


1. Separating the base flow and snowmelt from the total flow to obtain the

flood flow due to rainfall runoff.

2. Computation of the rainfall runoff (4.5 inches).

3. Dividing the ordinates of the SPF net flow by the rainfall runoff.

4. Conversion of the resulting 24-hour unit hydrograph to a six-hour unit

hydrograph by the S-curve technique (4).

The resulting hydrograph is a 6-hour unit hydrograph for the entire

6,266-square mile basin. The natural PMF hydrograph was then derived by

multiplying the ordinates of the 6-hour unit hydrograph and the 6-hour values

of rainfall runoff, summing the subtotals, and adding back the base flow

(58,800 cfs). The resulting natural PMF hydrograph is presented in

Figure 2.4-6.

There are five flood control storage reservoirs in the Connecticut River Basin

above Vernon. The total storage capacity of the reservoirs is

247,800 acre-feet, which represents a rainfall runoff over the total basin of

0.74 inches. The storage capacity of each reservoir is:

1. Union Village 38,000 acre-feet

2. North Hartland 71,400 acre-feet

3. North Springfield 50,600 acre-feet

4. Ball Mountain 54,600 acre-feet

5. Townshend 33,200 acre-feet

The operation of these flood control facilities has reduced the flood threat

in the basin. For instance, the Corps of Engineers estimates that the SPF

natural peak discharge of 263,700 cfs at Vernon has been reduced to

225,000 cfs for a net reduction of 38,700 cfs.

As stated above, the operation of current flood control facilities has reduced

the SPF at Vernon from a natural peak of 263,700 cfs to 225,000 cfs, or a

reduction of 38,700 cfs. If this same reduction were applied to the PMF, the

peak discharge would be decreased from 506,400 cfs to 469,700 cfs.

However, for conservatism, it is assumed that due to antecedent conditions,

the entire 247,800 acre-feet of storage capacity upstream is not available for

regulation of the PMF. Therefore, assuming that about 68% of the SPF

reduction would go into storage, the modified PMF discharge becomes

480,100 cfs. The resulting modified PMF hydrograph is shown in Figure 2.4-6.


The stage-discharge curve at the VYNPS site at Vernon was determined by the

standard step backwater method as described by Chow (5) utilizing the Ebasco

Backwater Calculation with Bridge Loss programmed for implementation on a

Burroughs 5500 computer. The recorded water surface profiles for applicable

floods of record (Table 2.4.7) were used as a basis for selecting roughness

coefficients, "n". The following "n" values were found to yield excellent

agreement with recorded flood profiles:

River Reach Low High Channel Overbank Above Vernon Dam 0.030 0.033 0.040 At Vernon Dam 0.013 0.013 0.013 Below Vernon Dam 0.030 0.033 0.050 River and valley cross sections upstream from Vernon Dam to the plant site and

downstream to the Central Vermont Railroad Bridge at Northfield,

Massachusetts, which were used for the step backwater computation are located

in Figure 2.4-8. The final rating curve for the plant site is shown in

Figure 2.4-9.

The time-varying PMF stage-discharge relationships are listed in Table 2.4.8

for the natural flood hydrograph and in Table 2.4.9 for the hydrograph as

modified by existing flood storage. Based on the PMF hydrograph modified for

existing flood storage, the PMF stillwater level at the site is

252.5 feet MSL.

As a check on the design flood for the site, failure of the largest upstream

flood control reservoir, Townshend Reservoir, was postulated to occur as a

result of an earthquake, which, in turn, occurs simultaneously with the SPF.

For conservatism, the maximum inflow of 71,000 cfs for this reservoir, which

is located about 22 miles upstream from Vernon, was considered to be

translated downstream and directly added onto the SPF peak discharge. This

coincident dam failure with the SPF modified peak discharge of 225,000 cfs

would produce a peak discharge of 296,000 cfs. From Figure 2.4-9, a peak

discharge of 296,000 cfs would produce a maximum stillwater elevation at the

site of 240.8 feet MSL.

The dam failure analysis described above was originally developed as a check

to ensure that the controlling flood for the site was the

precipitation-induced PMF. Since completion of the above upstream dam failure

analysis, additional information on flooding at the site due to failure of

upstream flood control and hydropower dams has been developed by the dam

owners and is summarized below. These more recent studies are based on

different criteria and analysis techniques than the previously described

analysis.


There are several large dams on the Connecticut River upstream of the VYNPS

site. The owners of these dams are required by the Federal Energy Regulatory

Commission to perform dam failure analysis as input to the development of

Emergency Action Plans. The only upstream dam failure flood that reaches the

VYNPS site for these Connecticut River dams is that for the Moore Dam. The

impacts for the other dam failures terminate well upstream of the site.

The hypothetical failure of Moore Dam was assumed to coincide with the peak of

the PMF inflow hydrograph. The dam is about 145 miles upstream from the VYNPS

site. Four downstream dams, Comerford, McIndoes, Dodge Falls and Wilder, were

assumed to fail in cascade. The results of the Moore Dam failure analyses at

Vernon Dam are a peak inflow of 305,600 cfs and a peak flood elevation of

240.1 feet MSL. The VYNPS site is subject to the same flood elevation as the

Vernon Dam. The arrival time at the site for the leading edge of the Moore

Dam failure flood wave is about 22 hours after the postulated failure of the

dam. The time of the peak flood at the site is about 47 hours after the

postulated dam failure.

There are also five flood control reservoirs on Connecticut River tributaries,

upstream of the VYNPS site. The owners have developed dam breach profiles for

each of the five dams. A review of these analyses showed that the impacts of

dam failure for three of the dams, Union Village, North Hartland, and North

Springfield do not reach the VYNPS site. Two of the dams, Townshend and Ball

Mountain, do produce flood levels downstream that reach the site. Both of

these dams are located on the West River, which is a tributary of the

Connecticut River.

For an assumed failure of Townshend Dam, the peak stage at Vernon Dam is

elevation 230 feet MSL. The time from the start of dam failure until the peak

stage is reached at the VYNPS site is 9.2 hours. The time from the start of

dam failure until the initial rise at the site is 5.2 hours. This analysis

used assumed pre-breach high flows in both the West and Connecticut Rivers.

For an assumed failure of Ball Mountain Dam, the peak stage at Vernon Dam is

elevation 235 feet MSL. The Ball Mountain Dam is upstream of the Townshend

Dam. The Townshend Dam fails as a result of the assumed failure of the Ball

Mountain Dam. The time from the start of dam failure until the peak stage is

reached at the VYNPS site is 10.0 hours. The time from the start of dam

failure until the initial rise at the site is 7.6 hours. This analysis also

assumed pre-breach high flows in both the West and Connecticut Rivers.

In summary, the flood levels at the VYNPS site due to upstream dam failures

are well below the PMF level at the site.


The maximum PMF stillwater level at the VYNPS site at Vernon, Vermont was

computed to be 252.5 feet MSL occurring 96 hours after the beginning of the

72-hour probable maximum precipitation period. Additional consideration is

now given to the problem of wave runup.

Atomic Energy Commission Safety Evaluation Docket No. 50-271 dated

June 1, 1971 has been reviewed during the NEI 12-07 Fukushima Flooding

evaluation and is considered the governing document. Page 12 of this document

concludes, “The PMF will produce a maximum discharge of 480,000 cfs at the

site and a corresponding stage of 252 feet 6 inches MSL. This maximum occurs

eight days after the start of the rainfall causing the flood. We consider it

possible that another storm or synoptic weather system with sustained winds of

at least 45 mph could follow the original storm and be at the site at the same

time that the peak discharge occurs. If the winds came from the most

effective direction, waves two to four feet high could result. These waves

would break at the river bank, but could produce plant flooding at elevations

as high as 254 feet MSL.

Nominal plant grade is 252.0 feet MSL. Accesses to the Turbine, Reactor,

Radwaste, and Control Buildings from out of doors are at grade 252.5 feet MSL.

In addition, direct access to the Reactor Building from out of doors is

through a pair of leak-tight doors.

Fuel pool cooling will be maintained during a maximum probable flood until

service water is lost due to river water leakage into the intake structure or

normal power is lost.

If normal electrical power is unavailable, the Vernon Hydroelectric Station

and the Station Blackout Diesel Generator are available to supply back-up

power.

If fuel pool cooling cannot be maintained during a maximum probable flood,

alternate fuel cooling strategies are available and will be implemented in

accordance with applicable facility procedures.

The PMF stillwater level is essentially equal to the top of most yard

electrical manholes. A potential avenue of water intrusion into the

Switchgear Room, Elevation 248.5 feet MSL exists through underground conduits

routed from manholes and handholes to the Switchgear Room floor. Should water

enter these manholes, the underground conduits could provide a path for water

to enter the Switchgear Room manholes. If the water level gets high enough,

flooding in the Switchgear Room and lower levels of the administration and

Turbine Building could occur. This flooding could affect the operability of

switchgear.


To preclude, or reduce the amount of water entering the Switchgear Room

manholes through the underground conduits which extend from the yard manholes,

these conduits have been sealed. In conjunction with the conduit sealing,

portable pumping capacity is available on-site to remove water which may enter

the Switchgear Room manholes. Additionally, facility procedures direct

personnel to remove this water as part of the site flood procedures.

Based on our review of these results of the flood analysis, we conclude that

acceptable measures will and can be taken to assure safe storage of irradiated

fuel even in the unlikely event that floods as large as the PMP should occur.”

The facility is, therefore, suitably protected against the maximum probable

flood and all lesser floods, including those due to the failure of upstream

dams.


2.4.4 Uses of River


The Connecticut River and its ponds are used by industry, chiefly for

hydroelectric power generation, and to some extent, by the public for

recreational purposes.

2.4.4.2 Industrial Use

The series of hydroelectric stations and their associated reservoirs on the

Connecticut River have been operated for many years to obtain maximum power

benefits for the power consumers of the New England region. This has required

operation of the river's hydroelectric stations as peak load facilities which

were shut down during the low load hours of each day and on weekends.

When river flow rates are less than 10,000 cfs, the Vernon Hydroelectric

Station is operated as a peak load facility. Often at such times, only one

hydroelectric unit is utilized during off-peak hours.

VYNPS's NPDES permit defines the maximum allowable thermal limits on the

Connecticut River.

Turners Falls Hydroelectric Project, FERC License No. 1889, is located

19.8 miles below Vernon Dam. This project, which utilizes water released from

the Vernon project, is owned and operated by the Western Massachusetts

Electric Company.

2.4.4.3 Public Use

Both Vernon Pond and Turners Falls Pond, next downstream, are used to some

extent for canoeing, boating, water skiing, and fishing. The utilization of

fishes resident in the Connecticut River has grown over the past few years.

Finfish have been studied in the Connecticut River in the area near VYNPS

since 1967. Fish were collected by various methods, including seining, gill

netting, minnow traps, fish traps (fyke nets), and electrofishing.

Table 2.4.10 lists, by scientific and common names, all of the species of

finfish taken through 1980 at Stations 2, 3, 4, and 5 on Figure 2.4-11. With

few exceptions, all specimens collected were identified, weighed, measured,

and released. Scale samples were taken from selected species for age-growth

studies. During the open cycle testing programs, similar data were collected

on all fish impinged on the traveling screens at the cooling water intake.

Fish data are presented in VYNPS's preoperational report (9), in subsequent

annual reports and in the reports of the open cycle testing programs.

A fish passage facility became operational at the Vernon Hydroelectric Station

in May 1981.


There are no direct municipal water intakes downstream of the VYNPS site.

Northeast Utilities operates a 1,000,000 kW pumped storage hydroelectric

generating plant at Northfield and Erving, Massachusetts. This plant obtains

water from the Connecticut River at a point 14 miles downstream from the

Vermont Hydroelectric Station and pumps the water to an upper reservoir.

During hours of peak electrical demand, this water is allowed to return to the

river through the reversible turbine pump units to provide peaking electrical

generating capacity.

The Metropolitan District Commission of Massachusetts has investigated the

feasibility of taking water from the Northfield Mountain Reservoir and

diverting it through a penstock and canal system to the Quabbin Reservoir,

approximately 10 miles distant. This reservoir supplies water to Metropolitan

Boston, Clinton, Marlboro, Southboro, Worcester, and other communities in

Massachusetts. No action has been taken on this proposal to date.

2.4.5 Biology

The location of VYNPS biological monitoring stations in the Connecticut River

are depicted in Figure 2.4-11. The approximate location of the eight

monitoring stations in river miles north and south of Vernon Dam are shown

below:

Station No. Location Relative to Vernon Dam

1 6.45 miles south 2 4.70 miles south 3 0.65 miles south 4 0.55 miles north 5 1.25 miles north 6 4.10 miles north 7 4.25 miles north 8 8.70 miles north

NOTE: Only Stations 3 and 8 are monitored subsequent to the permanent

cessation of power operation. 2.4.5.1 Commercial Fisheries

There are no commercial fisheries in the Connecticut River in the Vernon Pool

area.


2.4.5.2 Sport Fisheries

Thirty-three species of fishes have been found in the Connecticut River in the

vicinity of Vernon Dam. Some of the species, such as smallmouth bass and

yellow perch, are generally considered to be game fishes. Two anadromous

species, the Atlantic salmon and the American shad, have recently been

reintroduced into the Vernon area, as a direct result of the construction of

fish ladders at the Turners Falls and Vernon Dams. Successful passage of

these and other species have been recorded over the last few years. Other

species in the Connecticut River are either forage, coarse food, or "trash"

fish and are not generally sought by anglers. These species include the white

sucker and carp. Nearly all of the fish species present are warm-water

tolerant.

A 1980 survey shows that perch (both white and yellow), minnows, white sucker,

and bass are the most abundant fish species in the vicinity of the Vernon Dam

as seen in Table 2.4.11. These species comprised about 89% of the fish

population. The average weight of the smallmouth bass captured was

approximately 0.5 pounds, while the weight of the average sucker was nearly

1.5 pounds. Carp, white suckers, and minnows were shown to comprise

approximately 1/3 of the total number of all fishes caught, but accounted for

over 1/2 of the total weight.

2.4.5.3 Bottom Fauna

Monthly samples of Connecticut River benthic fauna were collected at

Stations 2, 3, 4, and 5 of Figure 2.4-11, from May through November with a

9-inch Ekman dredge and Henson traps (wire cages filled with 2 to 3-inch

diameter rocks). The following compares the number of samples and number of

genera of benthos collected by Ekman dredges over the years.

COMPARISON OF NUMBER OF SAMPLES AND NUMBER OF GENERA OF BENTHOS COLLECTED BY EKMAN DREDGE Station Number of Samples/Number of Genera Number 1969 1977 1978 1979 1980 2 6/33 8/20 8/22 7/27 7/36 3 6/24 8/25 8/13 7/26 7/39 4 7/16 8/19 8/17 7/26 7/30 5 8/18 8/20 6/14 7/28 7/25


As has been found in earlier years, caddis fly and chironomid larvae were the

predominant organisms in most of the spring and summer samples. Fall samples

showed a greater variety of dominant forms - fingernail clams, planarians,

oligochaetes. Chironomids and caddis flies were again dominant in the

November Henson trap samples. The very low Station 2 and 3 diversity indices

in that sample set were attributable to large percentages of a single

chironomid species, Tanytarsus sp., which accounted for 90% of the Station 2

sample and 94% of the Station 3 sample. Large percentages of the chironomid,

Glyptotendipes sp., in all three Henson trap samples of July and the Station 5

sample of September are evidenced in the relatively low diversity indices of

those samples.

2.4.5.4 Aquatic Plants

Few species of aquatic plants are found in the waters of the Connecticut River

in the Vernon Pool area. Marshes adjacent to the river are, however, rich in

vegetation. Cattails are the predominate vascular plant found in these

wetlands; other abundant species are rushes, sedges, grasses, horsetails, and

sweetflag.

2.4.5.5 Conclusions

The waters of the Connecticut River in the Vernon Pool area support a variety

of aquatic organisms. The fishes found in these waters are predominantly

those generally referred to as "warm-water" species. The benthic fauna are

generally sparse due to the silty nature of the river bottom. Marshes

adjacent to the river are rich in aquatic vegetation. Safe storage and

handling of irradiated fuel and radwaste management at the VYNPS does not

adversely affect the ecology of the Vernon Pool adjacent to the site.

2.4.6 Chemical and Bacteriological Quality of Water

Water quality monitoring requirements are established by the current National

Pollutant Discharge Elimination System (NPDES) Permit.

The water above Vernon Dam, as determined by the Vermont Water Resources Board

(effective July 2, 2000), has been classified as Class B waters and can be

described as follows:

Class B: The designated uses of Class B waters include aquatic biota,

wildlife, aquatic habitat, aesthetics, public water supply, irrigation of

crops and other agricultural uses, swimming and other primary contact

recreation, boating, fishing and other recreational uses.


2.4.7 River Field Program

Water quality parameters were monitored continuously from 1968 at Station 3,

downstream of VYNPS, and from 1970 at Station 7, upstream of the plant

(station locations are described in Section 2.4.5). In February 1980, the

requirement that conductivity and turbidity be monitored continuously was

deleted. The current ecological monitoring program is set forth in the NPDES

Permit which is issued every 5 years. Parameters to be monitored and

associated limits are ultimately established by the State of Vermont and may

be subject to revision within the course of each 5-year period. Data and

analysis from this monitoring program are presented in annual ecological

studies reports.

Biological studies, both qualitative and quantitative, are made to establish

the presence and amount of fish and benthic fauna. A comprehensive

environmental assessment is presented in Reference 11.

2.4.8 Conclusions

The station site nominal grade level is at elevation 252 feet Mean Sea Level

(MSL). The maximum river level that has occurred at the site was

elevation 231.4 feet MSL. The maximum Probable Maximum Flood Level at the

site is 252.5 feet MSL.

The PMF stillwater level is 6 inches above most yard electrical manholes. If

flood waters enter these manholes, potential flood pathways through conduits

which extend from the manholes into the Switchgear Room exist. This potential

water pathway through conduits is significantly reduced by the inclusion and

inspection of seals in conduits entering the Switchgear Room in addition to

measures in the flooding procedure which monitor and address any in-leakage.

Because the river is the natural low point and drainage channel for the

region, the groundwater table can be expected to slope toward the river.

Surface drainage also will flow toward the river. Thus, it is unlikely that

any liquids discharged to the river from the site would mix with domestic

water supplies in the area.


The Federal Energy Regulatory Commission (FERC) requires, under Order No. 122,

issued January 21, 1981, that the dams and related structures of all Licensed

Projects be inspected once every five years by an independent consulting

engineer and that they be certified as safe in their construction and

operation. In the event unsafe conditions of any nature are found, under the

order they must be called to the attention of the owner and the FERC and

necessary corrective measures must be carried out. In response to an

exception request dated June 26, 1997, FERC issued an exemption from filing an

Independent Consultant's Safety Inspection Report pursuant to the above

regulation for the Vernon, VT and Bellows Falls, VT dams by FERC letter dated

August 6, 1997, on the basis that those projects are "low hazard potential"

facilities. The dam owner retains the responsibility to provide an emergency

action plan and an inspection by an independent consultant in the event the

upstream or downstream circumstances of either project change such that

failure of a project structure would present a hazard to the public. The dams

operated by the Corps of Engineers are also subject to periodic safety

inspections. It is believed that these actions will assure the safety of all

dams on the river.


2.4.9 References

1. Probable Maximum Precipitation Susquehanna River Drainage above

Harrisburg, Pennsylvania, Hydrometeorological Report No. 40, U.S. Weather

Bureau, Washington, D.C., May 1965.

2. Seasonal Variation of the Probable Maximum Precipitation East of the

105th Meridian for Areas from 10 to 1,000 Square Miles and Durations of

6, 12, 24, and 48 hours, Hydrometeorological Report No. 33, U.S. Weather

Bureau, Washington, D.C., 1956.

3. Climatic Atlas of the United States, ESSA, U.S. Department of Commerce,

1968.

4. Flood Hydrograph Analysis and Computations, EM 1110-2-1405, U.S. Army

Corps of Engineers, 1959.

5. Chow, Ven Te, Open-Channel Hydraulics, Civil Engineering Series,

McGraw-Hill, 1959.

6. Technical Paper No. 55, Tropical Cyclones of the North Atlantic Ocean,

U.S. Department of Commerce, U.S. Weather Bureau, Washington, D.C., 1965.

7. Shore Protection, Planning and Design, Technical Report No. 4, Third

Edition, U.S. Army Coastal Engineering Research Center, Department of

Army, Corps of Engineers.

8. Computing Freeboard Allowances for Waves in Reservoirs, ETL No. 1210-2-8,

Department of Army, Corps of Engineers.

9. Webster-Martin, Incorporated, 1971. Ecological Studies of the

Connecticut River, Vernon, Vermont. Preoperational Report. Report

prepared for Vermont Yankee Nuclear Power Corporation.

10. U.S. Department of Commerce, 1978. "Tropical Cyclones of the North

Atlantic Ocean, 1871-1977," National Climatic Center, NOAA,

Asheville, N.C.

11. Aquatec, Incorporated, 1978. "316 Demonstration - Engineering,

Hydrological and Biological Information."

12. Aquatec, Incorporated, 1981. Ecological Studies of the Connecticut

River, Vernon, Vermont. Report X, January-December 1980. Report

prepared for Vermont Yankee Nuclear Power Corporation.

13. GZA GeoEnvironmental, Inc., 2011. Hydrogeologic Investigation of Tritium

in Groundwater, Vermont Yankee Nuclear Power Station Vernon, VT. Report

Prepared for Entergy Nuclear Operations, Vermont Yankee Nuclear Power

Station.


TABLE 2.4.1

Average and Extreme Values of Stream Flow Connecticut River at Vernon, Vermont Water Years 1944 - 1988 Highest Lowest Lowest Average Average Average Average Monthly Monthly Monthly Weekly Flow Flow Flow Flow Cfs Cfs Cfs Cfs October 6,571 20,201 1,646 1,475 November 9,033 20,450 3,366 2,159 December 9,486 24,326 2,934 2,494 January 7,655 17,338 2,589 2,283 February 8,187 24,428 2,935 2,135 March 15,544 36,245 5,308 4,373 April 30,799 51,210 14,980 11,523 May 18,047 38,790 7,262 3,118 June 8,768 21,890 3,387 2,424 July 4,911 21,790 1,841 1,033 August 4,005 13,615 1,805 1,223 September 4,159 15,610 1,650 1,138 NOTES: 1. All flows reflect regulation of upstream reservoirs for power purposes. 2. Flows measured at Vernon for 1944 - 1973, flows generated for Vernon from

United States Geological Survey gages at N. Walpole, New Hampshire, and Turners Falls, Massachusetts for 1973 - 1988.


TABLE 2.4.2 Vermont Yankee Nuclear Power Station, Daily Stream Flow for October 1964 to September 1965, Connecticut River at Vernon, Vermont Connecticut River Basin 1-1565. Connecticut River at Vernon, Vermont Location

Lat 4246'10", long 7230'50", on right bank just downstream from Vernon Dam at Vernon, Windham County, and 2 miles upstream from Ashuelot River. Drainage area 6,266 sq mi. Records available February to April 1936 (in WSP 798). September and October 1938 (in WSP 867), October 1944 to September 1965. Gage Water-stage recorder (digital). Datum of gage is at mean sea level, datum of 1929. Prior to January 20, 1948, at datum 94.13 ft higher. Average discharge 21 years (1944-65), 10,170 cfs (adjusted for storage). Extremes Maximum discharge during year, 32,000 cfs April 17 (gage height, 190.94 ft): minimum daily, 108 cfs September 6, 1936, 1938, 1944-65: Maximum discharge, 176,000 cfs March 19, 20, 1936 (gage height, 128.8 ft. datum then in use), from rating curve extended above 86,000 cfs: minimum daily, 99 cfs October 8, 1944. Remarks Records good except those below 1,000 cfs, which are fair. Flow regulated by powerplants and by First Connecticut and Second Connecticut Lakes, Lake Francis. Moore Reservoir and Comerford Station Pond (see Page 196), and other reservoirs (combined usable capacity, about 29 billion cubic feet).


2.0-55 of 108

TABLE 2.4.2

(Continued) DISCHARGE IN CUBIC FEET PER SECOND. WATER YEAR OCTOBER 1964 TO SEPTEMBER 1965 DAY OCT. NOV. DEC. JAN. FEB. MAR. APR. MAY JUNE JULY AUG. SEPT. 1. 1,220 164 10,900 6,730 4,500 7,000 *4,700 15,000 4,560 2,950 125 5,790 2. 1,900 3,070 5,580 5,450 3,500 6,600 4,390 13,600 7,620 2,830 1,710 8,810 3. 167 3,140 4,610 1,180 3,300 7,000 3,290 13,500 4,530 128 1,780 11,100 4. 157 3,050 6,740 4,890 3,400 7,400 826 13,000 *4,370 125 1,320 2,790 5. 3,270 2,210 1,760 7,300 4,900 8,000 4,970 12,300 1,100 125 1,900 134 6. 3,420 2,890 196 6,950 1,100 8,700 7,410 11,900 125 4,120 1,320 108 7. 3,720 725 4,150 6,420 1,000 9,800 8,910 10,200 4,650 2,650 296 3,760 8. 2,550 167 3,840 5,560 7,600 12,500 7,830 9,710 4,790 3,070 128 4,220 9. 2,620 3,400 4,190 5,250 6,800 13,000 10,200 4,630 5,600 3,270 2,640 2,860 10. 167 3,570 3,840 2,850 7,600 13,700 9,650 8,560 5,110 917 3,460 3,470 11. 160 811 3,990 6,770 8,400 13,800 11,900 9,660 4,890 567 3,460 2,210 12. 157 3,820 2,060 *6,760 9,000 12,800 14,300 10,400 1,060 2,290 2,390 628 13. 2,800 3,580 194 7,010 7,800 11,500 24,800 10,700 2,440 1,910 1,920 5,740 14. 2,850 2,920 4,340 6,960 7,400 9,650 23,600 8,840 10,800 2,390 125 6,540 15. 2,900 654 5,100 7,000 6,000 7,670 21,300 4,130 6,310 2,770 125 3,680 16. 2,200 4,990 5,330 2,800 6,300 9,260 24,100 2,400 7,400 2,720 3,900 2,270 17. 759 4,750 5,100 131 6,000 6,870 29,900 7,400 5,310 128 3,020 1,780 18. 737 4,750 5,630 4,300 6,500 7,500 25,400 8,730 5,300 125 4,690 125 19. 2,720 4,790 1,500 4,000 7,000 7,990 20,800 7,870 880 4,030 3,580 125 20. 3,550 5,250 174 3,700 3,500 7,250 19,400 7,050 2,550 3,810 3,080 1,940 21. 5,320 1,080 6,870 3,000 1,750 894 19,400 6,870 4,960 3,670 125 2,530 22. 3,580 167 5,120 3,800 5,000 7,320 *21,000 1,060 5,180 2,300 125 2,850 23. 5,470 5,260 6,980 2,000 5,900 6,610 24,400 134 5,450 2,530 2,280 3,050 24. 174 4,050 3,650 1,500 5,900 6,960 23,600 5,300 5,730 128 1,100 2,900 25. 167 4,460 3,660 4,600 7,000 6,440 19,100 4,910 4,450 125 1,430 9,360 26. 3,440 5,030 7,990 5,000 6,600 6,500 14,000 5,610 1,660 2,200 1,760 5,010 27. 3,620 8,580 15,500 4,100 1,850 3,920 15,400 5,960 759 1,730 1,740 7,960 28. 4,140 11,900 20,000 4,200 800 872 16,800 4,320 5,410 1,450 973 8,410 29. *3,870 12,100 17,400 4,500 ------ 6,770 16,300 1,120 4,940 859 885 7,040 30. 3,970 8,670 13,900 2,100 ------ 6,190 16,700 128 3,740 1,010 2,150 6,960 31. 969 ------ 12,400 600 ------ 5,530 ------ 125 ------ 131 2,420 ------ TOTAL 72,744 119,998 192,694 137,411 146,400 245,996 464,446 225,117 131,674 57,058 55,957 124,150 MEAN 2,347 4,000 6,216 4,433 5,229 7,935 15,480 7,262 4,389 1,841 1,805 4,138 MEAN** 1,942 4,335 6,040 3,867 3,834 6,385 16,490 8,924 4,518 1,809 2,174 4,383 CFSM** .310 .692 .964 .617 .612 1.02 2.63 1.42 .721 .289 .347 .699 IN** .36 .77 1.11 .71 .64 1.17 2.94 1.64 .80 .33 .40 .78 CALENDAR YEAR 1964 MAX 63,200 MIN 143 MEAN 7,840 MEAN** 7,822 CFSM** 1.25 IN 17.00 WATER YEAR 1964-65 MAX 29,900 MIN 108 MEAN 5,407 MEAN** 5,382 CFSM** .859 IN 11.65 Peak discharge (base, 50,000 cfs). - No peak above base. Note: Stage-discharge relation affected by ice January 15, 16, January 18 to March 9.

3* Discharge measurement made on this day. ** Adjusted for change in contents in all reservoirs from First Connecticut and Second Connecticut Lakes to reservoirs in West River basin listed on Page 196.


TABLE 2.4.3 Municipal and Industrial Groundwater Usage Within a 10-Mile Radius of the Vernon Site Minimum and Minimum and Static Number Maximum Depth Maximum Yield Level Town of Wells Feet GPM Feet Brattleboro, VT 49 10-540 1-40 9-35 Guilford, VT 30 50-540 1-50 8-41 Halifax, VT 12 34-345 1-30 8-10 Vernon, VT 14 36-565 1-75 25-115 Chesterfield, NH 26 41-470 0.2-25 12-50 Hinsdale, NH 5 72-280 3-30 -30 Winchester, NH 3 155-473 1.5-20 - Bernardston, MA 3 29-145 0.5-20 - Gill, MA 15 40-345 0.5-15 10-20 Leyden, MA 4 95-208 3-100 - Northfield, MA 19 52-345 0.3-40 14-85 Warwick, MA 5 52-372 1-9 - This information was obtained from records of the Green Mountain Well Company, Putney, Vermont.


TABLE 2.4.4

Public Water Supplies Within a 10-Mile Radius of the Vernon Site Location Town on Map Source Treatment Capacity Brattleboro, VT A Lake and Reservoir Filtration 3 MGD Plant Brattleboro B 3 Gravel-Packed Wells Filtration 970 GPM Supplementary 30 feet deep Plant 465 GPM 800 GPM Hinsdale, NH C-1 2 Gravel-Packed Wells Chlorine 220 GPM 74 feet and Sodium Phosphate 68 feet deep Sodium Hydroxide C-2 2 Gravel-Packed Wells Chlorine 500 GPM 47 feet and Sodium Phosphate 64 feet deep Sodium Hydroxide Winchester, NH D 3 Gravel-Packed Wells Phosphate 0.6 MGD Approx. 60 feet deep Northfield, MA No. 1 E 1 Gravel-Packed Well Sodium Hydroxide 100 GPM

No. 2 F Reservoir Chlorine 0.1 MGD Bernardston, MA G 2 Gravel-Packed Wells Potassium Hydroxide 480 GPM 69 feet and 74 feet 250 GPM deep G Sand and Gravel Well Chlorine 0.07 MGD 24 feet deep Sodium Hydroxide From 1963 Inventory of Municipal Water Facilities, U.S. Department of HEW, updated by local municipal water departments/town offices (2000).


TABLE 2.4.5 Water Supplies Within a 1-Mile Radius of the Site HINSDALE, NEW HAMPSHIRE Ground Feet Yield Elevation (ft) No. Depth GPM U.S.G.S. Use 1. 200 5 300 Domestic 2. 72 30 280 Domestic VERNON, VERMONT 1 198 15 295 Domestic 2 189 5.5 295 Domestic 3 368 11.5 275 School 4 356 25 208 Domestic & Homes 5 Spring Fed Supply Miller Farm Spring 6 125 30 320 Domestic 7 288 3.75 360 Domestic 8 19 275 Domestic & Farm 8A 18 Now Dry 275 9 17 3-4 275 Domestic 9A 25 Now Dry 275 Domestic 10 20 - 275 Domestic & Farm 10A 18 275 Domestic 11 30.5 4-5 275 Domestic 12 18 6 275 Domestic 13 16 275 Domestic 14 16 275 Domestic 15 14 275 Domestic 16 275 17 Spring 275 18 28 275 Domestic 19 20 15 275 Domestic 20 Spring 275 21 23 10 270 Domestic 22 19 270 Domestic 23 Spring 270 24 Spring 270 25 Spring 270 26 Spring 240 27 Spring 240 28 Spring 220 29 Spring 230 30 Going to 240 Drill 31 17 10 240 Domestic 32 20 240 Domestic 33 165 4 270 Domestic 34 70 20 270 Domestic 35 20 270 Domestic 36 31 270 Domestic & Farm 37 235 16.5 270 Domestic 38 23 275 Domestic 39 12 270 Domestic 40 18 270 Domestic 41 30 280 Domestic 42 20 280 Domestic 43 24 6 280 Domestic 44 35 Dry 280


Table 2.4.5 (Continued) Water Supplies Within a 1-Mile Radius of the Site Ground Feet Yield Elevation (ft) No. Depth GPM U.S.G.S. Use 45 275 Domestic 46 270 47 24 270 Domestic 48 23 275 Domestic 49 24 275 Domestic 50 22 275 Domestic 51 23 275 Domestic 52 20 275 Grange & Domestic 53 175 6 270 Domestic 54 18 270


TABLE 2.4.6 Six-Hour PMP and Runoff Increments Connecticut River Basin Above Vernon, Vermont Basin Arranged Shape Cumulative Incremental in PMP Time PMS Red. PMP PMP Critical Losses Runoff (hrs) (ins) Factor (ins) (ins) Order .05"/hr (ins) 6 6.1 .95 5.8 5.8 0.3 0.3 0.0 12 8.3 .95 7.9 2.1 0.7 0.3 0.4 18 9.7 .95 9.2 1.3 0.7 0.3 0.4 24 10.7 .95 10.2 1.0 0.9 0.3 0.6 30 11.7 .95 11.1 0.9 1.0 0.3 0.7 36 12.4 .95 11.8 0.7 2.1 0.3 1.8 42 13.2 .95 12.5 0.7 5.8 0.3 5.5 48 13.5 .95 12.8 0.3 1.3 0.3 1.0 54 13.8 .95 13.1 0.3 0.3 0.3 0.0 60 14.0 .95 13.3 0.2 0.2 0.3 0.0 66 14.2 .95 13.5 0.2 0.2 0.3 0.0 72 14.4 .95 13.7 0.2 0.2 0.3 0.0 TOTAL = 13.7 TOTAL = 10.4


TABLE 2.4.7 Maximum Annual Floods on Connecticut River at Vernon, Vermont Arranged in Descending Order (1927, 1936, 1938, 1945-1973) Year Peak Discharge, CFS Stage, Ft MSL 1936 176,000 222.9 1927 155,000 220.5 1938 132,500 214.8 1960 107,000 209.6 1973 102,000 - 1948 101,000 208.5 1953 98,800 208.0 1949 88,600 205.4 1968 88,100 205.3 1952 86,600 204.9 1958 84,200 204.3 1951 81,200 203.6 1969 81,200 203.5 1959 80,600 203.4 1947 79,600 203.3 1956 79,600 203.1 1972 78,600 201.7 1962 73,900 201.7 1955 70,500 200.9 1945 69,700 200.8 1950 68,300 200.3 1967 66,800 200.0 1964 66,100 199.8 1971 65,500 199.6 1954 65,000 199.5 1970 63,400 199.1 1946 62,700 198.8 1963 61,600 198.6 1961 57,900 197.7 1966 46,700 194.8 1965 32,000 190.9 1957 30,000 190.1


TABLE 2.4.8 Time-Varying PMF Stage-Discharge Table Vermont Yankee Nuclear Plant Site Time Discharge Stage Time Discharge Stage (hrs) (cfs) (ft MSL) (hrs) (cfs) (ft MSL) 0 58,800 221.3 156 247,000 237.0 6 58,800 221.3 162 231,000 235.8 12 58,900 221.4 168 217,000 234.7 18 59,000 221.5 174 205,300 233.8 24 59,300 221.5 180 192,300 233.0 30 59,800 221.6 186 182,000 232.1 36 60,800 221.8 192 172,000 231.4 42 66,300 222.0 198 162,400 230.7 48 78,300 223.4 204 154,000 230.1 54 95,200 225.0 210 146,000 229.4 60 117,500 227.2 216 139,400 229.0 66 158,200 230.4 222 133,000 228.5 72 255,000 237.5 228 128,400 228.1 78 367,000 245.7 234 122,000 227.5 84 417,000 248.8 240 117,300 227.2 90 464,000 251.5 246 112,000 226.6 96 506,400 253.9 252 107,500 226.2 102 465,000 251.6 258 104,000 226.0 108 430,000 249.5 264 100,000 225.5 114 403,000 248.0 270 97,000 225.2 120 380,000 246.5 276 93,700 225.0 126 351,000 244.8 282 91,400 224.7 132 321,000 242.7 288 88,500 224.5 138 294,000 240.8 294 86,000 224.2 144 283,000 239.6 300 83,700 224.0 150 265,000 238.3


TABLE 2.4.9 Time-Varying Modified PMF Stage-Discharge Table Vermont Yankee Nuclear Plant Site Time Discharge Stage Time Discharge Stage (hrs) (cfs) (ft MSL) (hrs) (cfs) (ft MSL) 0 58,800 221.3 156 235,100 236.1 6 58,800 221.3 162 220,600 235.0 12 58,900 221.4 168 207,400 234.1 18 59,000 221.5 174 196,500 233.2 24 59,200 221.5 180 184,200 232.4 30 59,600 221.6 186 174,300 231.5 36 60,500 221.9 192 164,900 230.9 42 65,900 222.3 198 156,000 230.3 48 74,100 223.0 204 148,200 229.6 54 87,900 224.4 210 140,600 229.1 60 106,400 226.3 216 134,500 228.6 66 142,300 229.2 222 128,200 228.1 72 235,000 236.1 228 122,300 227.5 78 343,800 244.1 234 118,100 227.2 84 390,700 247.2 240 113,800 226.8 90 437,700 250.0 246 108,800 226.4 96 480,100 252.5 252 104,500 226.0 102 439,700 250.1 258 101,200 225.6 108 407,300 248.2 264 97,700 225.3 114 381,900 246.8 270 94,500 225.0 120 360,500 245.5 276 91,500 224.6 126 333,500 243.5 282 89,460 224.4 132 305,100 241.4 288 86,630 224.2 138 283,900 239.6 294 84,280 224.0 144 269,000 238.5 300 82,140 223.9 150 252,900 237.4


TABLE 2.4.10 Checklist of Connecticut River Fishes Found Near Vernon, Vermont* Salmonidae - Trouts Salmo salar Linnaeus Atlantic Salmon Salmo trutta Linnaeus Brown Trout Salmo gairdneri Richardson Rainbow Trout Salvelinus fontinalis (Mitchill) Brook Trout Osmeridae - Smelts Osmerus mordax (Mitchill) Rainbow Smelt Catostomidae - Suckers Catostomus commersoni (Lacepede) White Sucker Catostomus catostomus (Forster) Longnose Sucker Cyprinidae - Minnows and Carps Cyprinus carpio Linnaeus Carp Semotilus corporalis (Mitchill) Fallfish Semotilus atromaculatus (Mitchill) Creek Chub Couesius plumbeus (Agassiz) Lake Chub Notemigonus crysoleucas (Mitchill) Golden Shiner Notropis cornutus (Mitchill) Common Shiner Notropis hudsonius (Clinton) Spottail Shiner Hybognathus nuchalis Agassiz Silvery Minnow Ictaluridae - Freshwater Catfishes Ictalurus nebulosus (LeSueur) Brown Bullhead Ictalurus natalis (LeSueur) Yellow Bullhead Esocidae - Pikes Esox lucius Linnaeus Northern Pike Esox niger LeSueur Chain Pickerel Anguillidae - Freshwater Eels Anguilla rostrata (LeSueur) American Eel Cyprinodontidae - Killifishes Fundulus diaphanus (LeSueur) Banded Killifish Percichthyidae - Temperate Basses Morone americana (Gmelin) White Perch

* Common names used in this checklist are those proposed by Bailey, Reeve M., et

al., 1970. "A List of Common Scientific Names of Fishes from the United States and Canada." Special Publication No. 6, American Fisheries Society, Washington.


TABLE 2.4.11 Fishes of the Connecticut River in the Vicinity of Vernon, Vermont All Collections - 1980 Total Total Weight Length Number Weight Extremes Extremes in Species Captured In Grams In Grams Millimeters Catostomus commersoni (Lacepede) White Sucker 190 129,514 0.5-1408 33-507 Cyprinus carpio Linnaeus Carp 19 91,765 138-8500 195-740 Semotilus corporalis (Mitchill) Fallfish 1 473 473 332 Notemigonus crysoleucas (Mitchill) Golden Shiner 12 913 35-170 137-225 Notropis hudsonius (Clinton) Spottail Shiner 195 2,062 7-15 73-128 Hybognathus nuchalis Agassiz Silvery Minnow 1 16 16 112 Juvenile Cyprinidae 133 208 0.05-2.6 17-68 Ictalurus nebulosus (LeSueur) Brown Bullhead 20 6,500 32-733 140-375 Ictalurus natalis (LeSueur) Yellow Bullhead 1 76 76 180 Esox lucius Linnaeus Northern Pike 1 400 400 400 Esox niger LeSueur Chain Pickerel 12 5,818 162-846 282-508 Anguilla rostrata (LeSueur) American Eel 1 1,360 1,360 750 Morone americana (Gmelin) White Perch 494 58,551 4-410 64-308 Perca flavescens (Mitchill) Yellow Perch 229 25,338 7-350 90-290 Stizostedion vitreum (Mitchill) Walleye 48 30,522 51-1156 185-490 Micropterus dolomieui Lacepede Smallmouth Bass 70 16,693 4-1470 70-490 Micropterus salmoides (Lacepede) Largemouth Bass 8 3,457 23-2040 110-507 Lepomis gibbosus (Linnaeus) Pumpkinseed 48 4,490 2.7-843 57-420 Lepomis macrochirus Rafinesque Bluegill 16 3,585 3-383 56-240 Ambloplites rupestris (Rafinesque) Rock Bass 103 13,246 2.1-302 51-250 TOTALS 1602 394,987 Source: See Reference 12.


Vermont Yankee


Station Site – Area Public Water Supplies 10 Mile Radius

Figure 2.4-1


Vermont Yankee


Station Site – Area Private Water Supplies

1 Mile Radius

Figure 2.4-2


Vermont Yankee


Enveloping Depth-Duration-Area ValuesOf PMP for Susquehanna River Basin

Figure 2.4-3


Vermont Yankee


6-Hour Unit Hydrograph

Figure 2.4-4


Vermont Yankee


Total SPF Hydrograph Figure 2.4-5


Vermont Yankee


Total PMF Hydrograph (Natural and Modified)

Figure 2.4-6


Vermont Yankee


Vermont Yankee Nuclear Plant

Location of River Cross-Sections Figure 2.4-8


Vermont Yankee


Stage Discharge Curve at The Vermont Yankee Nuclear Plant

Figure 2.4-9


Vermont Yankee


Cross Section of the Critical Fetch

Figure 2.4-10


Vermont Yankee


Vermont Yankee Sample StationsConn. River

Figure 2.4-11


2.5 GEOLOGY AND SEISMOLOGY


2.5.1 General

This subsection provides information related to geological and seismological

considerations at the site. Detailed site and laboratory investigations were

undertaken by independent consulting firms to obtain necessary design data

contained in this section. The consulting firms were Goldberg-Zoino and

Associates, Cambridge, Massachusetts, and Weston Geophysical Research, Inc., in

conjunction with the Vermont Yankee Nuclear Power Corporation and Ebasco Services

Incorporated. Evaluations of the investigations indicate that the proposed site

was adequate from the geological and seismological viewpoints and could safely

support the nuclear station installation.

2.5.2 Geology

The site is located on the west bank of the Connecticut River in the town of

Vernon, Vermont, which is in Windham County. Site coordinates are approximately

42 47' north latitude and 72 31' west longitude, in the extreme southeastern

corner of the state of Vermont.


All but one of the major structures of the facility, including the reactor

building and turbine building, are supported on rock. The storage pad for the

Independent Spent Fuel Storage Installation (ISFSI) is supported on engineered

fill placed on existing soils. Sixteen of the 93 borings at the site were made in

the immediate vicinity of the reactor building (see Figure 2.5-2). These borings

show that the area is overlaid by glacial deposits from the Pleistocene Age, with

an average 30 feet of glacial overburden above the local bedrock, which consists

of hard biotite gneiss. Rock outcroppings near the site are found along the river

bank. Bedrock exists at or near the foundation grades for the structures, namely

elevation 206 feet MSL for the reactor building, elevation 217 feet MSL for the

turbine building, elevation 227 feet MSL for the radwaste building, and elevation

187 feet MSL for the circulating water intake structure.

2.5.2.2 Geological Investigation Program

Standard geologic procedures were employed during the site investigation,

beginning with a complete search of available literature concerning geology and

seismic activity in the area including unpublished and published material (refer

to Table 2.5.1). A complete geologic field reconnaissance of the general area and

the immediate site was performed, employing United States Geodetic Survey

topographic maps, aerial photographs, and the state of Vermont geologic and

tectonic maps.


An extensive subsurface exploration project was undertaken at the site.

Ninety-three borings were made, 35 of which were from 32 to 100 feet in depth.

Thirty of these borings were AX (1-3/8 in. cores) and 5 were NX (2-1/8 in.

cores)(see Figure 2.5-2). All NX-size holes were logged in detail (see NX core

logs in Figure 2.5-3). The other cores were examined carefully to determine

general features and characteristics. Representative cores were taken at and

immediately below foundation grade in all NX core holes, and were submitted to

intense laboratory testing for determination of specific physical properties of

bedrock at the site. Several petrographic sections were made and analyzed to

ascertain the mineral composition and structure of bedrock.

A thorough seismic survey program was carried out to determine several of the

in-place physical properties of the site bedrock as it relates to earthquake

criteria for design of structures - such as compressional wave velocities (Vc),

shear wave velocities (Vs), subsurface rock contours, and the possibility of

extensive faulting and jointing.

The results of this investigation program are summarized in the following

paragraphs.

2.5.2.3 Regional Geology

Geologic structure of the region is complex, in that there are several sequences

of anticlinoria and synclinoria trending essentially in a northerly direction (see

Figures 2.5-4 and 2.5-5). The site is located geologically within the so-called

Brattleboro syncline, which is part of the Connecticut Valley-Gaspe Synclinorium.

Most of the region is underlain by Paleozoic metamorphic rocks and by a narrow

band of Triassic sedimentary rocks south of the site. The general outcrop pattern

of local Paleozoic formations indicates the presence of a major recumbent fold,

overturned to the west. The entire region has undergone extensive metamorphism

(mostly of the regional type) which apparently ranges from low-grade west of the

Vernon area to relatively high-grade east of the Vernon area.

Foliated igneous rocks of middle- and late-Devonian age underlie a large portion

of the region. These include three fairly large plutons of the Oliverian Magma

Series (Billings, 1935; Skehan, 1961), one of which is below the site, in the

towns of Vernon, Vermont and Hinsdale, New Hampshire - the Vernon Dome.


2.5.2.3.1 Regional and Local Stratigraphy

The Vernon pluton is a narrow elongate mass approximately 8 miles long and 2 miles

wide, striking approximately 10 degrees to the northwest and dipping steeply to

the east (see Figures 2.5-6 and 2.5-7). The Connecticut River flows southeasterly

across its central portion. Gneisses of the Oliverian Plutonic Series

(middle-Devonian) make up the pluton core (Billings, 1935)(see Figure 2.5-7).

Both the Vernon Dome and the Westmoreland Dome north of Brattleboro (see Figure

2.5-5), are part of the Bronson Hill Anticlinorium, a series of echelon gneissic

domes that extend northward into northern New Hampshire and southward into the

State of Connecticut. Except where the Connecticut River crosses the Vernon Dome,

the local topography reveals strikingly the distribution of the lithologic units.

2.5.2.3.2 Geological History

There appears to have been at least two distinct tectonic periods of folding after

formation of the Vernon Dome (late Paleozoic to pre-Triassic - over 70 million

years ago). Most normal faults on the flanks of the domes strike N30E. The two

faults at the northern terminus strike N10W. In all cases, the faults dip

steeply, and appear to be Triassic or younger in age. The Clough Quartzite and

the Littleton Formation have been intensely folded at the northern end of the

domal structure. Folds in the immediate area indicate differential movement with

reverse drag folds occurring along with recumbent structures (see Figure 2.5-4).

Faulting took place over 70 million years ago along the southeastern boundary of

the Vernon-Chesterfield area, particularly in the state of New Hampshire. The

Triassic Border Fault (see Figure 2.5-4) is the only fault structure of major

significance related to the site and there has been no apparent movement in it

during the last several million years. Rocks of the Oliverian Plutonic Series and

the mantle of a gneiss dome comprising the Ammonoosuc, Clough, and Littleton

Formations adjoin the fault on the east (Robinson, 1963). There has been relative

movement down on the west side of the fault. A crushed zone in gneiss on the east

side of the fault near Gill Station at the southern end of the Vernon area may be

associated with the fault (Bolk, 1956). The fault is exposed 4 miles south of the

Vernon-Chesterfield area where it dips steeply to the west (Keller and Brainard,

1940). It is difficult to match structures across the fault, and this prevents an

accurate estimate of throw on the fault. Recent movement along the fault is not

indicated. All minor faults in the region appear to be high-angle and Triassic or

younger in age.


2.5.2.3.3 Regional Structure

The three main structural features at the site are a relatively large, heavily

mantled gneissic dome, a major recumbent fold (Bernardson Nappe), and the eastern

margin of a regional syncline (Brattleboro Syncline).

The Bernardson Nappe apparently first formed during the main sequence of

deformation. No minor folds or lineations can be ascribed specifically to

movements that produced the nappe. The tectonic transport direction of the nappe

was generally from east to west. Formation of the gneissic domes followed

emplacement of the nappe and produced an early set of minor folds and lineations

on the mantle of the domes and on the rocks of the nappe (Trask, 1964). Early

folds at the north end of the Vernon Dome are overturned to the northeast and

northwest, with an apparent reverse drag. These early folds and lineations were

deformed then by still later folding in the synclinal area between the two domes.

Local isograds are essentially parallel to the regional structural trend. The

development of slip-cleavage in the Brattleboro Syncline, adjacent to the Vernon

Dome, was accompanied by extensive retrograde metamorphism (Moore, 1949).

Rocks of the Oliverian Plutonic Series, intrusive into the surrounding metamorphic

rocks, form the cores of elongate domes uplifts (Billings, 1935). The Vernon Dome

represents a southern counterpart of the Oliverian Magma Series. Foliation is

well developed around the margins of the Vernon pluton, but decreases slightly

toward its central portion. According to Moore (1949) and Skehan (1961), the

foliation is essentially parallel to the contact between the gneiss and the

overlying Ammonoosuc Volcanics. Available data (Moore, 1949), indicate that the

contact between the gneiss and the overlying Ammonoosuc Volcanics is concordant.

2.5.2.4 Site Geology

2.5.2.4.1 Physiography

The Connecticut River traverses the area near the site from north to south, along

the eastern side of the Vernon, Vermont area, geographically separating the states

of Vermont and New Hampshire at this point.

A strip of lowlands and terraces, about 1 mile in width, borders the river in the

area. There are naturally dissected uplands with an average local relief of

several hundred feet east and west of the lowlands. Wantastiguet Mountain, 0.5

mile east of Brattleboro, is the highest point in the area with an elevation of

1351 feet MSL. The lowest point is on the Connecticut River near Northfield,

Massachusetts, with an elevation of 175 feet MSL.


Rock composition of Vernon pluton is essentially a gneiss, grading from a

granodiorite (quartz-diorite) to granite. It is essentially light-gray to

pink-gray, slightly to moderately foliated, medium-grained, subporphyritic

quartz-diorite with hypidiomorphic to granoblastic texture. As tabulated in Table

2.5.2, the gneiss contains plagioclase feldspar (An12 to An44), quartz and biotite

mica, as essential minerals. Epidote, muscovite mica, K-feldspar, hornblende,

garnet, and magnetite are included as accessory minerals. Sericite, some

chlorite, and calcite are present also, as alteration products. Granulated and

flattened "quartz eyes" are present, and individual grains of these aggregates

show sutured and mortar textures. The quartz eyes and strings of biotite flakes

produce prominent lineation. As much as 1% zircon has been found in rocks of the

central region of the Vernon Dome.

An extensive subsurface exploration project was undertaken at the site. The

drilling program was carried out by the Raymond Concrete Pile Company during the

fall of 1966. Subsurface profiles of the borings in the vicinity of the station

structures are illustrated in Figures 2.5-8, 2.5-9, 2.5-10 and 2.5-11. Detailed

logs of deep borings in the reactor building area are provided in Figure 2.5-3.

During formation of the metamorphic plutonic body, various joints and minor

slippages occurred. Many of these joints served as avenues for solutions to

travel with subsequent mineralized fillings. Visual examination of rock cores

indicates that many joints were filled by hydrothermal solutions. A few joint

surfaces have a drusy appearance, some with crystal growth, and others with

mineral stainings left by ground water. Some fractures of joint surfaces appeared

weathered to highly weathered.

Pegmatitic quartzite veins were encountered in borings 1 at elevation 198 feet

MSL, 4 at elevation 208 feet MSL, and 5 at elevation 169 feet MSL. The

approximate strike of this vein is N60E and it dips 40NW. Apparent thickness of

this vein is 1-1/2 feet. Pegmatitic veins were encountered also in borings 6 and

21.

Dike or sill-like bodies of a dark green, fine grain diorite were found in boring

6 at elevation 194.5 feet MSL and elevation 191.7 feet MSL. They were found also

in boring 2A at elevation 210.1 feet MSL. Both units are approximately 5 inches

thick. Hard milky quartzite bands or veinlets with accessory magnetite were found

in many of the borings at various depths. Geologic relationships of these bands

have not been made, but it may be determined that they belong to a particular

joint set.

The rock is extensively jointed. Three or more joint sets may be present. These

joint sets appear reasonably tight.


2.5.2.4.2 Bedrock

Bedrock, although extremely hard and structurally competent, appears to be

fractured sufficiently to present occasional hydrostatic conditions in zones of

fracturing, as was observed in several of the drill holes. Water pressure tests

at the site were conducted to determine the "tightness" or permeability of certain

fractured zones. Tests proved that the formation is very tight.

2.5.2.4.3 Surficial Deposits

Rock types at the site are considered to be a metamorphosed igneous intrusive of

the Oliverian Plutonic Series. Generally, the rock is a quartzofeld

spathic-biotite gneiss, with variable amounts of orthoclase and plagioclase

feldspar. The attitude of this gneissic plutonic body is considered to trend

slightly west of north and dip to the east near the site.

At the site, the exposed outcrops along the edge of the river are massive, in some

instances intensively jointed due to mechanical weathering, and without any

visible gneissic structure. Foliation and lineation of the rock has been obscured

due to surface weathering. Rock cores reveal the gneissic structure. Foliation

is fairly well developed. The attitude of the rocks can be determined from the

cores by noting dip of foliation planes.

The gneiss is medium-grained, light-gray to slightly pinkish-gray rock, and its

texture somewhat approaches granoblastic. It is slightly subporphyritic and

rarely has a flaser fabric. Grains of white to gray glassy quartz, white to pink

feldspar, black biotite, with some muscovite and amphibole can be recognized.

Feldspar is quite variable. Minor constituents noted in the rock types are

magnetite, garnet, and possibly zircon and sphene.

2.5.2.5 River Geology

2.5.2.5.1 General

The Connecticut River at site lies within the New England upland. The basin is

maturely dissected with the river flowing throughout most of its course in an open

valley with well-developed flood plains above which rise glacial terraces tiered

on the valley walls. The main river in the upland section winds between rounded,

irregular hills and ridges.


The topography of the entire basin has been modified by glaciation which scraped

the tops from the bedrock hills and filled the valleys with glacial detritus with,

however, little actual diversion of drainage. The major effect of the glacial

fill was to raise the streams from their old beds, thereby permitting the

development of present channels which may or may not be related to the underlying

configuration of the old valleys in the bedrock.

2.5.2.5.2 Seismic Survey

The seismic tests resulted in conclusions as follows:

1. Seismic velocity measurements in boreholes and from surface studies are high

and indicate hard massive bedrock. Deeply weathered zones or faults were not

detected.

2. Bedrock surface is slightly irregular as evidenced by the borings and the

lines of seismic refraction investigations.

3. Elastic moduli values, based on seismic velocities, are high values of

approximately 4.16 x 106 lb/in2 for Young's Modulus and 1.53 x 106 lb/in2 for

the shear or rigidity modulus. Poisson's ratio is 0.347.

4. Compressional wave velocity was found to be 13,800 fps and the shear wave

velocity 6,500 fps.

2.5.2.5.3 Shoreline Retreat

The presence of natural outcrops of the bedrock at the various dam-sites such as

Vernon, Turners Falls, and Bellows Falls, coupled with the construction of dams at

these sites, have restricted the river's velocity and concentrated its potential

erosion at the sites themselves. The river banks, as a result, are relatively

stable, with erosion, if any, manifest only as a result of major floods. The long

intervening periods of placid flows provide ample opportunity for inspection and

stabilization of the river banks, should this be required.

At the site, the natural river banks have become well stabilized during the

60-year existence of the Vernon Hydroelectic Project immediately downstream.

There is little evidence of bank erosion.


2.5.3 Seismology


The evaluation of a nuclear power station site from a seismic standpoint is based

upon a combination of historical and instrumental data. Historical records before

1900 are somewhat misleading since observations are limited to population centers

and the untrained observer appears to sometimes exaggerate. Later historical

records, such as those of the early 1900's, appear to be more reliable.

Instrumentation for the detection of local earthquakes, which may or may not be

felt, has been operating in the New England area since the mid-1930's.

2.5.3.2 Seismic Investigation Program

The seismic evaluation of the station is a threefold study consisting of a review

of historical data from the New England area, an analysis of instrumental and

historical records for the Vernon area, and a study of earthquake intensity

attenuation with distance for northeastern United States.

2.5.3.3 Geologic and Tectonic Background

As described in detail in Subsection 2.5.2, "Geology", the southern parts of

Vermont and New Hampshire are composed of early Paleozoic sediments which have

been metamorphosed through intense folding. Some middle and late Paleozoic

igneous intrusives and extrusives are also present. The site itself is located on

the Vernon Dome, a middle Ordovician intrusive body of quartz-diorite gneiss. The

only post-Paleozoic tectonic feature present in the area is the eastern border

fault of the Triassic Basin of Massachusetts. The fault is present in extreme

southwestern New Hampshire where it strikes in a northeasterly direction passing

about 6 miles to the southwest of the station site. A tectonic map of the New

England area is shown in Figure 2.5-12.

2.5.3.4 Seismic History

Those earthquakes which have been strongly felt or have produced some damage in

the New England area are shown in Figure 2.5-13. Areas of some seismic activity

are noted in the vicinity of the following locations: Lake George, New York;

Concord, New Hampshire; Ossipee Mountains, New Hampshire; southeastern New

Hampshire and northeastern Massachusetts; and Haddam Connecticut. All of these

areas lie between 50 and 100 miles from the plant site and have experienced at

least one historical earthquake which has produced some minor damage (Modified

Mercalli Intensity VI or greater).


The nearest of these areas to the site is the Concord, New Hampshire area, about

50 miles to the northeast of the site. The earthquake of November 23, 1884, of

Intensity VI on the Modified Mercalli Scale, is the largest to have occurred at

Concord. This earthquake was felt over an 8000-square mile area which did not

include the Vernon, Vermont area.

The largest earthquake to have originated in the vicinity of Lake George, New

York, occurred on April 20, 1931. The epicenter of this earthquake is about 75

miles northwest of the site. It was reported to have been felt at Bellows Falls,

Vermont; Greenfield, Massachusetts; and Hinsdale, New Hampshire. The intensity at

Vernon can be estimated at about IV on the Modified Mercalli Scale (see Figure

2.5-14).

Modified Mercalli isoseismal lines for the Ossippe, New Hampshire, earthquakes of

December 20 and 24, 1940, which were of epicentral Intensity VII, show that the

intensity at Vernon, Vermont, was about IV. The epicenter of these earthquakes

was about 95 miles northeast of the site. Although the isoseismal lines show an

intensity of IV, reports from various localities in the area show that intensities

range from III to VI. In Keene, New Hampshire, a great part of which is located

on alluvium, an intensity of VI was noted, although just outside the town in the

surrounding highlands, the intensity was IV. Brattleboro, Vermont, reported an

Intensity V; Bellows Falls, Vermont, reported an Intensity IV; and Hinsdale, New

Hampshire across the Connecticut River from Vernon, reported an Intensity of III.

The earthquake of October 5, 1817, whose epicenter was near Woburn, Massachusetts,

was listed as Modified Mercalli Intensity VII by the United States Coast and

Geodetic Survey. The only report of damage is that "walls were thrown down at

Woburn". Since no other reports concerning this earthquake could be found, it is

doubtful that this earthquake had any effect on a site located 70 miles to the

west-northwest of Woburn, Massachusetts.

The earthquakes of November 9, 1727, at Newburyport, Massachusetts, and May 18,

1791, at East Haddam, Connecticut, are both listed by the United States Coast and

Geodetic Survey as Intensity VIII (Modified Mercalli). Both earthquakes occurred

between 85 and 90 miles from the plant site. Historical evidence shows that these

earthquakes were felt over wide areas of the northeastern United States, probably

including the Vernon, Vermont area. Although there is evidence that these

earthquakes were less than Intensity VIII, attenuation of earthquake intensity

with distance would probably have reduced these (even if they were of intensity

VIII) to Intensity IV or V at the plant site.


2.5.3.5 Seismicity of Area

A more detailed picture of the seismicity of the central New England area

surrounding Vernon, Vermont, is shown in Figure 2.5-15. This figure shows the

approximate epicentral location of all the earthquakes of record.

The nearest earthquake to Vernon, Vermont, for which instrumental records were

obtained, took place on June 1, 1963. The epicenter of this earthquake was

located near Shelburne Falls, Massachusetts, about 15 to 20 miles southwest of

Vernon, Vermont. The intensity was listed as Modified Mercalli Intensity II or

less.

The earthquake catalogue of Henry Fielding Reid lists some local activity in the

Keene, New Hampshire area on October 10, 1854, and December 1, 1875. No newspaper

accounts of these earthquakes could be found in the New Hampshire Sentinel,

published in Keene.

A report of the earthquake which was observed in Vernon, Vermont, on June 11,

1898, appeared in the Monthly Weather Review of June 1898. "Vernon, Vermont,

reports an earthquake on the 11th, at 1:25 a.m., which was distinctly felt and

jarred the house. This seems to be quite an isolated case, and it is worth

inquiring whether this jar was not due to something else than a true earthquake".

Local newspapers were studied for any accounts of this earthquake. The earthquake

was observed in Brattleboro, but apparently not observed in Keene, or Hinsdale,

New Hampshire or Greenfield, Massachusetts. The newspaper account of the

earthquake which appeared in the Brattleboro Evening Phoenix of June 13, 1898, is

as follows: "An earthquake shock was felt distinctly in Brattleboro at 1:45

Saturday morning. People who were awake say that houses were shaken, and that

doors were slammed by the shock. The nervous shock which one woman sustained was

sufficient to cause illness. Mr. Pratt, the night watchman at S. A. Smith and

Company's factory, says that almost the same moment that the earthquake was felt,

a brilliant meteor flashed across the sky and exploded with a loud report. People

who felt the earthquake also heard the report, but few saw the meteor".

It is possible that this event was a meteor, but in evaluating the seismic

history, we must consider it as a local earthquake of Modified Mercalli Intensity

IV in the Vernon-Brattleboro area (see Figure 2.5-14).


2.5.4 Conclusions

The nuclear installation is located on the west side of the Connecticut River near

Vernon, Vermont. The station is supported on rock at the site. Bedrock in the

area is hard, strong, competent gneiss with unconfined compressive strengths that

generally exceed 15,000 psi. The rock is moderately to highly jointed. The mass

of the rock has not been weakened structurally to any important degree by the

jointing. Seismic velocity measurements at the site verify the hard massive

nature of the bedrock. Deeply weathered or faulted zones were not detected at the

site. Geologic considerations do not preclude utilization of the site for a

nuclear station location.

The effects at the site resulting from a significant seismic disturbance have been

considered based upon local and regional geology, tectonics, and historical and

instrumental seismology.

It is indicated from geologic and tectonic history that the region is relatively

quiescent. Low magnitude seismic events can occur, but should be relatively

infrequent.

The seismic activity of the area is depicted on Figure 2.5-13. Some concentrated

areas of seismic activity may be noted 50 miles to the northeast of the site in

the Concord, New Hampshire area and at other localities 75 miles to 100 miles

distance from the site. The nearest earthquake to the site which produced damage

occurred near Concord, New Hampshire, and was of Intensity VI on the Modified

Mercalli Scale. Concord, New Hampshire, is 50 miles from the site.

Based on intensity attenuation with distance, the largest New England earthquakes,

which occurred some 85 to 90 miles from the site, would have been observed as

Modified Mercalli Intensity IV or V at the plant site.

The nearest earthquake to the site, which occurred during the instrumental

recordings of the last 30 years, had an epicentral location of approximately 15 to

20 miles from the site. The probable maximum intensity from an earthquake which

has been observed in the Vernon area is that of Intensity V on the Modified

Mercalli Scale. Based on extrapolated data from earthquakes which occurred on the

west coast of the United States, the maximum acceleration to be expected at the

bedrock surface of the plant site in Vernon, Vermont, would be from an earthquake

of Intensity V to low Intensity VI on the Modified Mercalli Scale. This

earthquake would produce an acceleration of approximately 0.03g to 0.04g.


It is believed that the earthquake accelerations developed for this site are

conservative. They result from detailed studies of the site and region by

consultants knowledgeable in the field of seismology. However, for design

purposes, a minimum ground acceleration of 0.07g was used. In addition,

structures and equipment have been examined for an acceleration of 0.14g to

ascertain that no failure could occur that would prevent safe storage of

irradiated fuel.


TABLE 2.5.1

AVAILABLE INFORMATION CONCERNING GEOLOGY AND SEISMIC ACTIVITY RELATED TO THE VERMONT YANKEE NUCLEAR POWER STATION SITE REFERENCES Balk, R., 1956. Bedrock Geology of the Massachusetts Portion of the

Northfield Quadrangle, Massachusetts - New Hampshire - Vermont: U.S. Geol. Survey, Geol. Quad. Map GQ92.

Billings, M.P., 1935. Geology of the Littleton - Moosilauke Quadrangles, New

Hampshire: New Hampshire State Planning and Development Comm., Concord, New Hampshire

Keeler, J., and Faulted Phyllite East of Greenfield, Brainard, C., 1940. Massachusetts: Am. Jour. Sci., V. 238, pp. 354-365. Moore, G. Em. Jr., 1949. Structure and Metamorphism of the Keene - Brattleboro

Area, New Hampshire - Vermont: Geol. Soc. Am., Bull., Vol. 60, pp. 1613-1670.

Robinson, P., 1963. Gneiss Domes of the Orange Area, Massachusetts and New

Hampshire: Doctoral Thesis, Harvard University. Skehan, J. W., 1961. The Green Mountain Anticlinorium in the Vicinity of

Wilmington and Woodford, Vermont: Bull: 17, Vermont Geol. Survey, Vermont Development Dept.

Trask, N.J., Jr., 1964. Stratigraphy and Structure in the Vernon -

Chesterfield Area, Massachusetts; New Hampshire; Vermont: Doctoral Thesis, Harvard University. Unpublished.

MAP REFERENCES 1. Centennial Geologic Map of Vermont, 1961, Compiled and Edited under the

direction of Dr. Charles G. Doll, State Geologist. 2. Geologic Map of New Hampshire, Marland P. Billings, Dept. of Geology, Harvard

University in cooperation with New Hampshire Planning and Development Commission and U.S. Geological Survey.

3. Geology Map of the Keene - Brattleboro Area; G.E. Moore, Jr., 1949, for New

Hampshire Planning and Development Commission.


TABLE 2.5.2 VERNON PLUTON: ESTIMATED MODE OF THE OLIVERIAN MAGMA SERIES 1 2 3 4 5 Phenocrysts Quartz - 2 5 8 - Plagioclase - 2 - 2 - Hornblende - - - 5 - Groundmass Plagioclase 54 52 58 54 53 K-feldspar tr tr - - 7 Quartz 32 36 31 30 36 Biotite 5 3 - - 3 Chlorite 2 1 2 tr - Muscovite 3 2 4 - tr Epidote 4 1 - - 1 Magnetite - tr tr tr tr Garnet - tr tr - - Zircon tr tr tr tr tr Apatite tr tr tr tr tr Sphene - - - - tr Pyrite - - tr - - Hematite - - tr - - Leucoxene - - tr - - Tourmaline - - tr - - Carbonate tr 1 - - - % of Anorthite in Plagioclase 41 37 12 44 41 Size of Groundmass (mm) 0.25-1.0 0.1-0.25 0.05-0.2 0.15-0.3 0.05-0.5 Size of Phenocrysts (mm) - 2.0-3.0 2.0-4.0 1.0-4.0 - Texture Gr* Gr Gr Gr Gr Subp" Subp M** Por' M** 1. Quartz-diorite 4. Hornblende quartz diorite *GR = Granoblastic 2. Quartz-diorite 5. Granodiorite "Subp = Subporphyritic 3. Quartz-diorite **M = Mortar 'Por = Porphyritic (After Moore, 1949)


2.0-90 of 108

Vermont YankeeDefueled Safety Analysis Report

Station Site - Geological Survey -

General Plan-Location of Test Borings Figure 2.5-2


2.0-91 of 108


Station Site - Geological Survey -Subsurface Profile – Log of Test

Borings (1A, 2A, 3A, 4, 5, 8)

Figure 2.5-3


2.0-92 of 108

Vermont Yankee


Station Site – Tectonic Map -

State of Vermont Figure 2.5-4


2.0-93 of 108

Vermont Yankee


Station Site – Tectonic Map -

State of New Hampshire Figure 2.5-5


2.0-94 of 108

Vermont Yankee


Station Site – Geological Survey Area Bedrock Geology

Figure 2.5-6


2.0-95 of 108

Vermont Yankee


Station Site – Geological Survey

Area Geological Section Figure 2.5-7


2.0-96 of 108

Vermont Yankee


Station Site – Geological Survey Subsurface Profile (Section AA)

Log of Test Borings (5, 8, S9, 11 and 21) Figure 2.5-8


2.0-97 of 108

Vermont Yankee


Station Site – Geological Survey Subsurface Profile (Section BB)

Log of Test Borings (2A, 3A, ST6-1/2 and S9) Figure 2.5-9


2.0-98 of 108

Vermont Yankee


Station Site – Geological Survey Subsurface Profile (Section CC)

Log of Test Borings (2, 2A, 5, 7, 7A, 13, 15)

Figure 2.5-10


2.0-99 of 108


Station Site – Geological Survey

Subsurface Profile (Section BB)

Log of Test Borings (3, 3A, 4, 8, 8A, 12 and

16)

Figure 2.5-11


2.0-100 of 108

Vermont Yankee


Station Site – Tectonic Map –

New England Area

Figure 2.5-12


2.0-101 of 108

Vermont Yankee


Station Site – Compilation of

Earthquakes-New England Area

Figure 2.5-13


Vermont Yankee


Station Site – Earthquake Intensity

Modified Mercalli and Rossi-Forel Scales

Figure 2.5-14


2.0-103 of 108

Vermont Yankee


Station Site – Compilation of Earthquakes

Central New England Area

Figure 2.5-15


2.6 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

2.6.1 Objectives

The radiological environmental monitoring program is designed to demonstrate

the adequacy of environmental safeguards inherent in station design, the

effectiveness of the Process Radiation and Area Radiation Monitoring Systems

in measuring the controlled releases of low levels of radioactive materials

and the impact, if any, on the environment as a result of facility operation.

Emphasis is placed on control at the source with follow-up and confirmation by

environmental radiological surveillance.

The program consists of two phases, preoperational and operational, each

having specific objectives. The preoperational phase was conducted over the

two-year (approximate) period preceding station operation to establish

background radiation levels and radioactivity concentrations at selected

locations, to assess the variability between sample locations, and to observe

any cyclical or seasonal trends in the environmental sample media. Although

VYNPS has certified permanent cessation of operation and permanent defueling

in accordance with 10 CFR 50.82, the facility will continue in the operational

phase of the radiological environmental monitoring program. The operational

phase of the program has the following objectives:

1. To assure that radiation levels and radioactivity concentrations in the

environment resulting from facility operation meet the applicable

regulatory and license requirements.

2. To make possible the prompt recognition of any significant increase in

environmental radiation or radioactivity levels and to identify the cause

of the change, whether it be station effluents, effluents from other

nuclear facilities, fallout from atmospheric nuclear weapons tests,

seasonal changes in natural background, or other sources.

3. Obtain information on the critical radionuclides and pathways leading to

the quantitative evaluation of the dose to man resulting from the

operation of the station.


2.6.2 Monitoring Network

The radiological environmental monitoring program compares measured radiation

levels and levels of radioactivity in samples from the area possibly

influenced by the station to levels found in areas not influenced by the

station. Sampling in both areas is done in accordance with the requirements

of Off-Site Dose Calculation Manual (ODCM) and Technical Specifications, with

the area outside the influence of the station serving as a background or

control for the area in the immediate vicinity of the station. A comparison

of survey data collected at control locations and locations within the range

of influence of the station (indicator locations) allows the determination of

any significant difference between the two areas. This method of

environmental sampling makes it possible to differentiate between facility

releases and other fluctuations in environmental radioactivity due to

atmospheric nuclear weapons test fallout, seasonal variations in natural

background, and other causes.

With the cessation of operations and reduced risk of radioactive releases from

the facility, the direct radiation monitoring network is reduced to each of

the 16 compass sectors around the facility with a land border with the state

of Vermont. Additional stations are situated at special interest and control

locations.

The types of sample media used for environmental surveillance are divided into

four categories, based on exposure pathways. These categories are direct

radiation, airborne, waterborne, and ingestion. Each of these is described

below. Specific and more detailed monitoring requirements may be found in

ODCM Section 3/4.5.1, and the identification of specific monitoring locations

may be found in Table 7.1 of the ODCM. The number of sampling locations and

the frequency of sampling discussed below reflect minimum ODCM requirements.

The actual sampling program may exceed these requirements.

2.6.2.1 Direct Radiation

Environmental direct radiation (gamma) measurements are continuously monitored

at approximately 40 locations. Either pressurized ion chambers or

Thermoluminescent Dosimeters (TLDs) are used to obtain an integrated gamma

radiation exposure at frequencies as prescribed in the ODCM. However, the

frequency of analysis readout is based upon the specific system used as

discussed in the ODCM.


2.6.2.2 Airborne

Air is sampled for particulates at offsite locations as described in the ODCM

(including one control). The samples are collected by passing the air through

a glass fiber filter in series with a charcoal cartridge. The sampling pumps

operate continuously, and a meter is incorporated into the sampling stream to

measure the total volume of air sampled during a given interval.

The air particulate filters are collected and analyzed weekly for gross beta

radioactivity. These filters are composited for each sampling station and are

analyzed quarterly for gamma-emitting radionuclides.

Increased sampling frequency or additional analyses may be required on air

particulate filters or charcoal cartridges if conditions warrant, pursuant to

the footnotes to Off-Site Dose Calculation Manual Table 3.5.1.

2.6.2.3 Waterborne

2.6.2.3.1 Surface Water

River water samples are collected from one upstream and one downstream

location. At the upstream (control) location, a grab sample is collected

monthly. At the downstream location, an automatic compositing water sampler

collects an aliquot of river water at time intervals that are very short

relative to the compositing period (monthly). These composited samples are

collected monthly.

A gamma isotopic analysis is required on each monthly sample. These samples

are also composited, by station, for a quarterly tritium analysis.

2.6.2.3.2 Ground Water

Grab samples of ground water are collected and analyzed in accordance with the

requirements of the Off-Site Dose Calculation Manual.

2.6.2.3.3 Sediment from Shoreline

Sediment grab samples are collected semiannually from two locations, one

downstream from the station and one at the North Storm Drain Outfall. Each

sample is analyzed for gamma-emitting radionuclides.


2.6.2.4 Ingestion

2.6.2.4.1 Deleted

2.6.2.4.2 Fish

Recreationally important species of fish are collected semiannually from two

locations, one upstream and one in the vicinity of the station discharge. The

edible portions of each sample are analyzed for gamma-emitting radionuclides.

2.6.2.4.3 Vegetation

A mixed grass sample is collected at each air sampling station on a quarterly

schedule, as available. Each sample is analyzed for gamma-emitting

radionuclides.

A silage sample is collected from each milk sampling station quarterly, as

available. Each sample is analyzed for gamma-emitting radionuclides.

2.6.3 Land Use Census

A Land Use Census is performed annually according to the Off-Site Dose

Calculation Manual 3/4.5.2. Analyses are done to ensure that the receptors

used for calculations done in accordance with the Off-Site Dose Calculation

Manual 3/4.3.3 are conservative.

2.6.4 Emergency Surveillance

The environmental monitoring program is designed to supplement emergency

monitoring functions as well as perform the routine surveillance activities.

The monitoring stations are strategically located and equipped to provide

radiation monitoring data essential to the rapid assessment of any accidental

radioactivity release.


2.6.5 Reports

An Annual Radiological Environmental Operating Report is submitted to the NRC.

The report contains a summary, interpretations, and an analysis of trends for

the results of the radiological environmental surveillance activities for the

report period. Included are comparisons with operational controls and

previous environmental surveillance reports, plus a description of the

radiological environmental program and a map of all sampling locations. An

assessment of the impact of the station operation on the environment is also

included.


FACILITY DESIGN AND OPERATION

TABLE OF CONTENTS

Section Title Page

3.1 DESIGN CRITERIA ....................................................... 7

3.1.1 Conformance with 10 CFR 50 Appendix A General Design Criteria .................................................... 7

3.1.2 Classification of Structures, Systems and Components ........ 9

3.1.3 Loading Considerations for Structures, Foundations, Equipment and Systems ...................................... 10

3.1.3.1. Seismic Classification ......................... 15

3.1.3.2 Seismic Design ................................. 17

3.1.4 References ................................................. 20

3.2 FACILITY STRUCTURES .................................................. 23

3.2.1 Reactor Building ........................................... 23

3.2.1.1 Function ....................................... 23

3.2.1.2 Description .................................... 23

3.2.1.3 Seismic Analysis ............................... 25

3.2.2 Turbine Building ........................................... 26

3.2.2.1 Function ....................................... 26

3.2.2.2 Description .................................... 26

3.2.3 Plant Stack ................................................ 27

3.2.3.1 Description .................................... 27


3.2.4 Control Room Building ...................................... 28

3.2.4.1 Description .................................... 28


3.2.5 Circulating Water Intake and Discharge Structures .......... 28

3.2.5.1 Intake Structure ............................... 28

3.2.5.2 Discharge and Aerating Structure ............... 29

3.2.6 Cooling Tower Deep Basin ................................... 29


3.2.7 Independent Spent Fuel Storage Installation ................ 30

3.2.7.1 Description .................................... 30


3.2.8 References ................................................. 32

3.3 SYSTEMS .............................................................. 35

3.3.1 Fuel Storage and Handling .................................. 36

3.3.1.1 Nuclear Fuel ................................... 36

3.3.1.2 Spent Fuel Storage ............................. 39

3.3.1.3 Standby Fuel Pool Cooling and Demineralizer Systems ........................................ 45

3.3.1.4 Tools and Servicing Equipment .................. 48

3.3.1.5 References ..................................... 51

3.3.2 Service Water System ....................................... 57

3.3.2.1 Objective ...................................... 57

3.3.2.2 Design Bases ................................... 57

3.3.2.3 Description .................................... 57

3.3.2.4 Evaluation ..................................... 59

3.3.2.5 Inspection and Testing ......................... 59

3.3.3 Electrical Power Systems ................................... 59

3.3.3.1 Transmission System ............................ 59

3.3.3.2 Auxiliary Power System ......................... 61

3.3.3.3 Deleted ........................................ 64

3.3.3.4 125 V DC System ................................ 64

3.3.3.5 24 V DC Power System .......................... 66

3.3.4 Fire Protection System ..................................... 67

3.3.4.1 Objective ...................................... 67

3.3.4.2 Design Basis ................................... 67

3.3.4.3 Description .................................... 68


3.3.4.5 References ..................................... 70

3.3.5 Heating, Ventilating and Air Conditioning Systems .......... 71


3.3.5.1 Objective ...................................... 71

3.3.5.2 Design Bases ................................... 71

3.3.5.3 Description .................................... 72


3.3.6 Instrument and Service Air Systems ......................... 76

3.3.6.1 Objective ...................................... 76

3.3.6.2 Design Basis ................................... 77

3.3.6.3 Description .................................... 77


3.3.7 Process Sampling ........................................... 78

3.3.7.1 Objective ...................................... 78

3.3.7.2 Design Basis ................................... 78

3.3.7.3 Description .................................... 78

3.3.8 Deleted .................................................... 79

3.3.9 Lighting Systems ........................................... 80

3.3.9.1 Objective ...................................... 80

3.3.9.2 Design Basis ................................... 80

3.3.9.3 Description .................................... 80


3.3.10 Communication Systems ...................................... 81

3.3.10.1 Objective ...................................... 81

3.3.10.2 Design Basis ................................... 81

3.3.10.3 Description .................................... 82


3.3.11 Process Computer System .................................... 83

3.3.11.1 Objectives ..................................... 83

3.3.11.2 Design Bases ................................... 83

3.3.11.3 Description .................................... 84


3.3.11.5 Cyber Security ................................. 86


3.3.11.6 Process Computer Data Feed to the Plant Data Server (PDS) ................................... 87

3.3.12 Torus-as-CST System ........................................ 87

3.3.12.1 Objective ...................................... 87

3.3.12.2 Design Basis ................................... 87

3.3.12.3 Description .................................... 87


FACILITY DESIGN AND OPERATION

LIST OF TABLES Table No. Title 3.1.1 Allowable Stresses for Class I Structures 3.1-2 Safety Margins for Several Critical Portions of Major Class I

Structures


FACILITY DESIGN AND OPERATION LIST OF FIGURES Reference Figure No. Drawing No. Title 3.2-18 Main Stack Geometry 3.3.1-1 Fuel Storage-Arrangement 3.3.1-2 5920-6893 POOL FUEL STORAGE RACK ARRANGEMENT 3.3.1-3 5920-12795 Pool Layout Spent Fuel Storage Racks 3.3.1-4 Fuel Storage Rack Assembly 3.3.1-5 HOLTEC Fuel Storage Rack Assembly (Partial)


3.1 DESIGN CRITERIA

3.1.1 Conformance with 10 CFR 50 Appendix A General Design Criteria

The final version of the General Design Criteria was published in the Federal

Register February 20, 1971 as 10CFR50 Appendix A. Differences between the proposed

and final versions of the criteria included a consolidation from 70 to 64 criteria

and general elaboration of design requirement details. At the time of issuance, the

Commission stressed that the final version of the criteria were not new requirements

and were promulgated to more clearly articulate the licensing requirements and

practices in effect at the time.

In a Staff Requirements Memorandum on SECY-92-223, the NRC approved a proposal in

which it was recognized that plants with construction permits issued before May 21,

1971 were not licensed to meet the final General Design Criteria. The memo

recognized that while compliance with the intent of the final General Design

Criteria was important, back fitting of these requirements to older plants would

provide little or no safety benefit.

Although VYNPS was not required to comply with the General Design Criteria, the

design and construction of VYNPS was reviewed against the intent of the General

Design Criteria proposed in July, 1967. That review was documented in the VYNPS

UFSAR, Appendix F.2, Revision 17, is historical, and is not included in the DSAR.

Although changes were made to the facility over the life of the plant that may have

invoked the final General Design Criteria as design criteria, such invocation was

not intended to constitute a regulatory commitment, unless specifically docketed as

such.

The original Appendix F information, except cross-reference to applicable FSAR

Sections, is retained here for historical significance. Indications of the present

or future tense should be understood as being related to the time frame during which

this Appendix was originally written. Refer to information elsewhere in the DSAR and

in other design basis documentation to determine current design configuration.

The proposed General Design Criteria that are considered to remain applicable in the

defueled condition include the following:

Criterion 1--Quality Standards The quality assurance program is presented in the VY Quality Assurance Program

Manual (VY QAPM). The description of the various systems and components includes

the codes and standards that are met in the design and their adequacy.


Criterion 2--Performance Standards Conformance to the applicable structural loading criteria ensures that those systems

and components affected by this criterion are designed and built to withstand the

forces that might be imposed by the occurrence of the various natural phenomena

mentioned in the criterion, and this presents no risk to the health and safety of

the public. The phenomena considered and margins of safety are also given.

Criterion 3--Fire Protection The materials and layout used in the station design have been chosen to minimize the

possibility and to mitigate the effects of fire. Sufficient fire protection

equipment is provided in the unlikely event of a fire.

Criterion 5--Records Requirement Complete records of the as-built design of the station, changes during operation and quality assurance records will be maintained throughout the life of the station.

Criterion 11--Control Room The facility is provided with a centralized control room having adequate shielding

to permit access and continuous occupancy under 10CFR20 dose limits during the

design basis accident situation.

Criterion 12--Instrumentation and Control Systems The necessary controls, instrumentation, and alarms for safe and orderly facility operation are located in the control room. These instruments and systems allow appropriate monitoring control of the facility. Sufficient instrumentation is provided to allow monitoring of all variables necessary for effective facility control.

Criterion 17--Monitoring Radioactive Releases The station process and area radiation monitoring systems are provided for

monitoring significant parameters from specific station process systems and specific

areas including the station effluents to the site environs and to provide alarms and

signals for appropriate corrective actions.

Criterion 18--Monitoring Fuel and Waste Storage The spent fuel storage areas have been analyzed to determine their safety, and

instrumentation is provided for monitoring where needed.


Criterion 66--Prevention of Fuel Storage Criticality Appropriate facility fuel handling and storage facilities are provided to preclude

accidental criticality for spent fuel.

Criterion 67--Fuel and Waste Storage Decay Heat The system used to cool the spent fuel pool is designed to remove sufficient decay heat to maintain the pool water temperature. The fuel storage pool contains sufficient water so that in the event of the failure of an active system component, sufficient time is available to either repair the component or provide alternate means of cooling the storage pool. Criterion 68--Fuel and Waste Storage Radiation Shielding The handling and storage of spent fuel is done in the spent fuel storage pool. Water depth in the pool is maintained at a level to provide sufficient shielding for normal reactor building occupancy by facility personnel. A demineralizer system is used to control water clarity and to reduce water radioactivity. Accessible portions of the reactor and radwaste buildings have sufficient shielding to maintain dose rates within the limits of 10CFR20. Criterion 69--Protection Against Radioactivity Release From Spent Fuel and Waste Storage The consequences of a fuel handling accident in the spent fuel pool are presented elsewhere in the DSAR. In this analysis, it is demonstrated that undue amounts of radioactivity are not released to the public. All spent fuel and waste storage systems are conservatively designed with ample margin to prevent the possibility of gross mechanical failure which could release significant amounts of radioactivity. Backup systems such as floor and trench drains are provided to collect potential leakages. Appropriate facility personnel are rigorously trained and administrative procedures are strictly followed to reduce the potential for human error. The radiation monitoring system is designed to provide facility personnel with early indication of possible malfunctions. Criterion 70--Control of Releases of Radioactivity to the Environment The station radioactive waste control systems (which include the liquid and solid radwaste systems) are designed to limit the off-site radiation exposure to levels below limits set forth in 10CFR20.

3.1.2 Classification of Structures, Systems and Components

Following certification of permanent defueling, VY is no longer authorized to

emplace or retain fuel in the reactor vessel in accordance with 10CFR50.82(a)(2).


Since it is no longer possible to load a nuclear core, power operations can no

longer occur and reactor related design basis accidents are no longer possible.

Consequently, it was determined that the remaining design basis accident possible at

VY is a fuel handling accident (FHA) consisting of a dropped fuel bundle in the

Spent Fuel Pool. As presented in the Safety Analysis chapter of the DSAR, the dose

consequences of the fuel handling accident are well within acceptance criteria, with

no reliance on either the Standby Gas Treatment System or the Secondary Containment

System.

Based on the changed conditions described above, an evaluation of the systems,

structures and components (SSCs) described in the UFSAR was performed to determine

the SSC safety classification based on the function, if any, each SSC would perform

in the permanently defueled condition. The process and criteria used to classify

the SSCs and the conclusions of the evaluation are provided in appropriate station

documents.

3.1.3 Loading Considerations for Structures, Foundations, Equipment and Systems

All structures have been designed to withstand the combinations of dead and live

loads which give the severest credible conditions of loading. Loading, including

seismic, wind, and impact loading, are in accordance with the applicable codes, and

incorporate the applicable provisions of the Uniform Building Code, Zone II, 1967

Edition; ACI Standard Building Code Requirements for Reinforced Concrete (ACI

318-63); ACI Standard Specification for the Design and Construction of Reinforced

Concrete Chimneys (ACI 505-54); AISC Specification for the Design, Fabrication, and

Erection of Structural Steel for Buildings (1963); American Water Works Association,

"AWWA Standard for Steel Tanks, Standpipes, Reservoirs, and Elevated Tanks for Water

Storage," AWWA D-100 (1967); USA Standards Institute ASA B96.1, "Welded Aluminum

Alloy Field-Erected Storage Tanks; National Fire Protection Association Standard

NFPA No. 30, "Flammable and Combustible Liquids Codes" (1966); Section III of the

ASME Boiler and Pressure Vessel Code, "Nuclear Vessels" (1968); and Section VIII of

the ASME Boiler and Pressure Vessel Code, "Unfired Pressure Vessels" (1968).

The Reactor Building and all other Class I structures except the main stack and

ISFSI storage pad are founded on firm bedrock. The main stack rests on end-bearing

steel piles which transfer stack loads to the bedrock. The ISFSI storage pad is

founded on engineered fill placed on existing soil.

The maximum allowable bearing pressure is 50 tons per square foot. The maximum

loading on the bedrock does not exceed 20 tons per square foot.

The maximum anticipated earthquake at the site would result in a maximum horizontal

ground acceleration of 0.07g. Facility design ensures that appropriate functions

remain available during or following a ground horizontal acceleration of 0.14g.


A strong motion, solid state digital accelerograph is floor mounted in the Reactor

Building southwest corner room, floor Elevation 213'-9". The unit is designed to

provide continuous monitoring for earthquakes by means of three orthogonal

accelerometers, two horizontal and one vertical, which sense earthquake ground

motion. Once triggered, the accelerograph will record the seismic event or a series

of events as long as the trigger levels are exceeded. The primary function of the

strong motion accelerograph is to provide data which will be of value in promptly

assessing the condition of the plant subsequent to an earthquake per 10CFR100,

Appendix A.

The maximum anticipated wind velocity that is anticipated at the site is 80 mph with

gusts to 100 mph. The station structures are designed to withstand the anticipated

wind loadings. The site is located in a geographic area which has a small

probability of being subjected to tornadic wind conditions.

Live loads, including construction loads, which are greater than the loads

prescribed under code, and loads from operating pressures and/or temperatures which

increase the stresses, have also been used in the design. Standard practice of use

and application in power plants determined the selection of the materials used in

the various structures and supports.

The loadings considered were as follows:

D is the dead load of structure and equipment plus any other permanent

loads contributing stress such as soil, hydrostatic pressure, temperature

loading, or operating pressures.

L is the live load from any nonpermanent loads such as equipment not fixed

in place, roof snow load, etc.

R is the jet force or pressure on the structure due to rupture of any one

pipe.

H is the force on the structure due to thermal expansion of pipes.

E is the design earthquake load.

E' is the maximum hypothetical earthquake load.

W is the load due to wind.

W' is the load due to tornado.


The loading considerations, using the postulated events, which have been followed

for all Class I structures and equipment to determine the controlling stress levels

to be used in design are:

Loading Consideration Allowable Stress A. Reactor Building and All Other Class I Structures, excluding the primary containment 1. D+L+R+E Normal allowable code stresses are used.

The customary increase in design stresses for the loading combinations considered is not permitted.

2. D+L+R+E' Stresses are allowed to approach the

3. D+L+W' yield point for ductile materials, and 0.85 times the ultimate strength for concrete.

4. D+L+W Normal allowable code stresses and

customary increases in stresses are used for these load combinations.

B. Class I Tanks 1. D+L+H+E Normal allowable code stresses are 2. D+L+H+W used. The customary increases in design

stresses for the load combinations considered are not permitted.

3. D+L+H+E' Stresses are allowed to approach the yield

point for ductile materials and 0.85 times the ultimate strength for concrete.

The load combination equations listed above are based on allowable stress design.

No plastic strength design for steel structures or ultimate strength design for

concrete was used for Vermont Yankee; therefore, no load factors were applied to the

subject equations.


To assure the required properties of concrete poured during cold weather, placing of

concrete with ambient temperatures around 15oF was done with several requirements

that included temperature control during the mixing, placing, and curing of the

concrete. The mixing water was heated to a temperature range of 100°F to 175°F

which was adequate to maintain a concrete temperature of ±65°F at the point of

discharge from the mixer. This temperature is within allowable limits for proper

concrete placement. No frozen lumps of material were allowed in the charging hopper

of the batching plant. When necessary, the area of concrete placement was sheltered

for protection against the weather and preheated. This precaution was taken to

assure that no concrete would be placed against frozen surfaces. During placement

of concrete floors, heat was provided for the underside as well as the top surface.

The ambient temperature in the area of the placement was maintained at a minimum

temperature of 45°F for at least 5 days, and special coverings or enclosures were

provided to permit proper curing conditions for the concrete. Concrete specialists

were retained to design the concrete mixes, perform testing as required, and to

assist in developing an overall concrete program for the project. They also

witnessed and reported on concrete placements and were encouraged to comment on all

phases of the program including cold weather concreting.

Table 3.1.1 gives the maximum allowable stresses used for the various loading

conditions for Class I structures.

Floor live loads are based on equipment and operating loads and applied in

accordance with the Uniform Building Code Zone II (UBC), 1967 Edition. Roof live

loads are 40 psf applied as specified in the UBC to obtain the worst condition of

stress.

The 40 psf design roof live loads (snow loading) was determined as follows:

The American National Standards Institute (formerly the American Standards

Institute) in their "Minimum Design Loads in Buildings and other Structures,"

specify the weight of seasonal snowpack equaled or exceeded 1 year in 10 as the

minimum snow load for design purposes. This figure for the Vermont Yankee Nuclear

Power Station is equal to 30 pounds per square foot. Forty pounds per square foot

or 10 psf more than specified, was conservatively used for the design of the

structures. The weight of the estimated maximum accumulation on the ground plus the

weight of the maximum possible snowstorm of 70 psf, as shown in Section 2.3.5.3, is

interpreted as applicable for the drifts on the ground where accumulation is

permitted by the terrain. Winds will not permit such accumulations to occur on

building roofs of the station; therefore, the 40 psf used in the design is

considered a conservative loading.


The station masonry wall design for Class I structures is analyzed to meet the NRC

Bulletin 80-11 guidelines. The design approach and analysis used were approved in

References 1 and 2.

Floor dead loads include the weight of the structural components and the

architectural appurtenances. Operating loads consist of gravity loads from all

equipment and piping. All structures satisfy the requirements of the UBC, Zone II,

35 psf basic wind as per American Standards Association (ASA) A58.1, 1955. In

addition, the following Class I structures have been designed to withstand

short-term tornado winds up to 300 mph: Control Room Building, Reactor Building

below the refueling level, intake structure (service bay area), Turbine Building

self-contained Diesel Generator Rooms, tornado walls around outdoor condensate

storage and fuel oil storage tanks. The effect of a 300 mph wind on a Class I

structure was analyzed by applying a uniformly distributed positive pressure of

185 psf on the windward side of the structure and a negative pressure of 115 psf on

the leeward side in accordance with ASCE Wind Forces on Structures. It is assumed

that there is a 3 psi pressure drop associated with the passage of a tornado. Only

those structures which are enclosed require design against the effect of this

pressure drop. In the Reactor Building, the internal overpressure is relieved by

providing that specified areas of the siding enclosure (blowout panels) above the

refueling level will fail at an overpressure falling in the designated range of 0.35

psi to 0.60 psi (Reference 3). Subsequent pressure equalization is obtained at each

successive level below the refuel floor by means of large open hatch areas on each

floor. In the Diesel Generator Rooms of the Turbine Building, dedicated tornado

pressure relief dampers are provided which will allow the room to vent through the

intake air supply to the exterior of the building. Since the siding on the Turbine

Building will blow off with winds, such as those associated with a tornado, the

Daytank Rooms are vented into the Turbine Building by the open space that is

provided beneath their doors. These dampers and openings will provide adequate

venting capacity to limit the pressure differential on the enclosure walls. The

Control Room Building has been designed to withstand a 3 psi pressure drop without

venting.

Class I structures are also designed against penetration by tornado-created

missiles. The missiles which have been considered are 4 x 4 inch x 16 foot-long

wood posts and 2 x 12 inch x 16 foot-long wood planks. For an analysis of the

effects of a tornado on the spent fuel storage pool, see APED-5696, "Tornado

Protection for the Spent Fuel Storage Pool."

For tornado loading, metals are allowed to approach their yield point, and concrete,

its ultimate strength.


3.1.3.1. Seismic Classification

The two classes of structures applicable to the earthquake design requirements are

as follows:

Class I - Structures and equipment whose failure could cause significant release of

radioactivity in excess of 10CFR100 for a low probability event or which are vital

to the removal of decay and sensible heat from the spent fuel pool.

The ISFSI storage pad (comprised of an East and West pad) is classified as Important

to Safety Class C (ITS-C) as defined in 10CFR72.3. The Important to Safety features

of the storage pad are to maintain the conditions required to store spent fuel

safely and prevent damage to the spent fuel container during storage.

Class II - Structures and equipment which may be essential or even nonessential to

the operation of the facility.

An analysis of the consequences of failure of several structures was performed.

This analysis showed that a condensate storage tank rupture could result in the

release of radioactivity resulting in potential doses in excess of the limits of

10CFR20 for unrestricted areas. It should be recognized, however, that this failure

constitutes an accident and that 10CFR50.67 rather than 10CFR20 applies. Within the

scope and bases used in this analysis, no Class I or II structures or equipment were

found which, upon failure, could result in doses in excess of the limits of

10CFR50.67 at the site boundary.

3.1.3.1.1 Class I Structures

The following is a listing of the Class I structures associated with the storage of

irradiated fuel:

Suppression Chamber (torus), including vents and penetrations

Reactor Building

Control Room Building

Plant Stack

Intake Structure (Service Water Pump Area)

Cooling Tower Deep Basin

Independent Spent Fuel Storage Installation (ISFSI) pad


3.1.3.1.2 Class I Equipment

The following is a list of Class I equipment associated with the storage of

irradiated fuel:

• Nuclear Steam Supply System

- Fuel Assemblies

• Station Service Water System (up to the main condenser discharge block)

• Fuel Storage Facilities, to include spent fuel storage equipment

• Instrumentation and Control Systems

- Radiation Monitoring System (partial)

• Fuel Oil Storage Tank

3.1.3.1.3 Class II Structures

Administration Building

Intake and Discharge Structures (except as noted under Class I structures)

All Other Structures, not listed in Paragraph 3.1.3.1.2, that have seismic

design requirements

3.1.3.1.4 Class II Equipment

Reactor Building Cranes

Condensate Storage Transfer System

Station Auxiliary Power Busses

Electrical Controls and Instrumentation (for above systems)

All Other Piping and Equipment, not listed in Paragraph 3.1.3.1.3, that have

seismic design requirements


3.1.3.2 Seismic Design

All Class I structures were designed conservatively so that under the worst loading

conditions the allowable stresses will not be exceeded. Several critical portions

of major Class I structures are listed in Table 3.1.2, showing the margins of safety

for the controlling loads for the listed structural member.

No. 1 shows (a) the circumferential stresses in the reactor pedestal due to a

jet force and (b) the vertical stresses at the base of the pedestal due to

direct load plus earthquake plus jet force.

No. 2 shows the stresses due to dead and live load plus design base or maximum

hypothetical earthquake at the face of the biological wall and at midspan of an

important beam in the Reactor Building. This beam supports part of the floor

deck at Elevation 280 plus an interior column which extends up to the refuel

floor.

No. 3 shows the stresses under the same loading conditions as in No. 2 in a

footing supporting a column of the Control Room.

No. 4 shows the stresses in the south wall of the housing of the service water

pumps in the intake structure under tornado wind load.

Based on seismological investigations, response spectra and dynamic analyses

established for the station, envelopes of maximum acceleration, displacement, shear,

and overturning moment versus height have been developed. The horizontal ground

acceleration for the design earthquake is 0.07 times gravity (0.07g), and the

vertical motion 2/3 that of the horizontal. Both motions are assumed to occur

simultaneously.

It is noted that Appendix A "Seismic Analysis" of the DSAR Revision 0, contained

historical information regarding the seismic design analysis for various SSCs.


3.1.3.2.1 Class I Structures

Mathematical models whose properties correspond to those of the structures or

equipment were formulated. The seismic design for the Class I structures and

equipment is based on dynamic analysis using the acceleration response spectrum

curves. The design is such that safe shutdown can be made during a ground motion of

0.14g, combined with the vertical accelerations assumed to be 2/3 of the horizontal

ground acceleration, with no variation of the vertical coefficients with height.

For the dynamic analysis of Class I structures, the damping factors used for

vibrations below the elastic limit are as follows:

Item Percent of Critical Damping

Reinforced Concrete Structures 5.0

Steel Frame Structure 2.0

Bolted or Riveted Assembly 2.0

Welded Assembly (equipment and supports) 1.0

Vital Piping System Various

Wood Structures with Bolted Joints 5.0

Summaries of the seismic analyses for the Reactor Building, Control Room Building,

plant stack, intake structure, and deep basin are given in the Facility Structures

section under the respective structure. Detailed analyses for the ISFSI storage pad

is contained in References 4 through 8 for the East Storage Pad and References 9

through 14 for the West Storage Pad.

3.1.3.2.2 Class II Structures

Design was in accordance with the provisions of the Uniform Building Code, Zone II.

Alternately, all such structures were designed to resist a minimum horizontal

seismic coefficient of 0.05, with a 1/3 allowable increase in basic stresses.


3.1.3.2.3 Equipment Seismic Design

Class I equipment analysis considers vertical and horizontal ground motions. The

coefficients for horizontal motion were adjusted to correct for equipment elevation

above grade, and also consider the stiffness of the equipment supports. The

magnitude of the vertical acceleration used was 2/3 of the horizontal ground

acceleration with no variation of the vertical coefficients with height. Allowable

stresses are in accordance with Table 3.1.1.

Stresses have also been checked for an earthquake with two times the seismic

coefficients. Class I equipment is bolted or fastened so that it will not be

displaced.

All Class I tanks have been analyzed for forces resulting from a horizontal

acceleration of 0.22g acting simultaneously with a vertical acceleration of 0.05g.

These accelerations take into account the height above grade of the various Class I

tanks. Stresses have been kept within the basic code allowables with no increase

for short-term loading. Further analysis was performed using twice the horizontal

and vertical accelerations, and for this condition of loading, stresses in the

ductile materials have been permitted to go to 0.90 of yield.

The selection of the horizontal seismic loading coefficient for the Class I tanks

was based on the maximum acceleration at the elevation of the tank in the supporting

structure. This permits maximum flexibility in arrangement of the vital tanks

within the structures and ensures no condition of overstress due to seismic loading.

For Class II equipment, the seismic analysis has assumed there is no vertical ground

motion. This is in accordance with the Uniform Building Code, Zone II. The

horizontal motion has been adjusted to correct for equipment elevation above grade.

Code allowable stresses with increase for short-term loading have been maintained.

Class II tanks are analyzed for forces resulting from a horizontal acceleration of

0.09g, with allowable stresses increased by 25% in accordance with the provisions of

the Uniform Building Code.

The selection of the seismic acceleration coefficients for the Class II tanks also

reflects the upper elevations of the structures where these tanks are located.

Class I equipment is principally supported on reinforced concrete. In the Reactor

Building, all supporting concrete has a minimum 4,000 psi, 28-day ultimate

compressive strength. All other supporting concrete has a minimum 3,000 psi, 28-day

ultimate compressive strength. All reinforcing has a minimum yield stress of 40,000

psi. Where structural steel is used to support Class I equipment, ASTM A36 standard

rolled shapes or other material analyzed to meet the requirements of this section

are used. The allowable stresses are as listed in Table 3.1.1.


3.1.4 References

1. Letter, G. Lainas (USNRC) to J. B. Sinclair (VYNPC), “Masonry Wall Design, IE

Bulletin 80-11,” NVY 83-262, dated November 15, 1983.

2. Letter, D. B. Vassallo (USNRC) to R. W. Capstick (VYNPC), “Masonry Wall Design

Supplement – Inspection and Enforcement Bulletin 80-11,” NVY 85-240, dated

November 18, 1985.

3. Calculation VYC-1828, “Reactor Building Masonry Wall Review for HELB

Loadings.”

4. Calculation VYC-2427, “Development of Acceleration Time Histories for Vermont

Yankee ISFSI Analysis.”

5. Calculation VYC-2428, “Development of Strain Compatible Soil Properties for

Vermont Yankee ISFSI Analysis.”

6. Calculation VYC-2433, “Soil Structure Interaction Analysis of the Vermont

Yankee ISFSI.”

7. Calculation VYC-2435, ”Vermont Yankee Nuclear Power Plant ISFSI Facility

Concrete Storage Pad Design”

8. Calculation VYC-2434, “Vermont Yankee ISFSI Cask Sliding Analysis.”

9. Calculation VYC-3175, "Determination of Soil Parameters for ISFSI Expansion

Concrete Storage Pad."

10.Calculation VYC-3176, "Development of Response Spectra Consistent Time

Histories for ISFSI Expansion Concrete Storage Pad."

11.Calculation VYC-3177, "Development of Strain Dependent Soil Properties for

ISFSI Expansion Concrete Storage Pad."

12.Calculation VYC-3178, "Soil Structure Interaction Analysis and Cask

Stability/Sliding of ISFSI Expansion Concrete Storage Pad."

13.Calculation VYC-3179, "Liquefaction Potential for ISFSI Expansion Concrete

Storage Pad."

14.Calculation VYC-3181, "Structural Concrete Design for ISFSI Expansion Concrete

Storage Pad."


TABLE 3.1.1 Allowable Stresses for Class I Structures

Loading Conditions

Reinforcing Steel

Maximum Allowable Stress

Concrete Maximum

Allowable Compressive

Stress

Concrete Maximum

Allowable Shear Stress

Concrete Maximum

Allowable Bearing Stress

Structural Steel

Tension on Net Section

Structural Steel Shear

on Gross

Section

Structural Steel

Compression on Gross Section

Structural Steel

Bending

1. Loading as defined without E', W and W'

0.50 Fy 0.45 f c

1.10 f c

0.25 f c

0.60 Fy 0.40 Fy Varies with slenderness ratio

0.66 Fy

to 0.60 Fy

2. Loading as defined excluding E, E' and W'

0.667 Fy 0.60 f c

1.467 f c

0.333 f c 0.80 Fy 0.53 Fy

Varies with slenderness ratio

0.88 Fy to

0.80 Fy

3. Loading as defined with E' or W' present

Seismic load (0.14g)

See Note A 0.85 f c -- -- See Note A 0.60 Fy

Varies with slenderness ratio

See Note A

* 25% Live Load is considered concurrent with seismic load. Fy is the minimum yield point of the steel used.

f c is the compressive strength of concrete.

Note A: Stresses permitted to approach but not exceed yield stress of the material.


TABLE 3.1.2 Safety Margins for Several Critical Portions

of Major Class I Structures

Controlling

Loading

Allowable Stress in psi

Actual Stresses in psi

Safety Margins (Allowable/Actual)

Structure Condition Concrete Reinforcing Concrete Reinforcing Concrete Reinforcing

1. RPV Pedestal a. Circumferential Stresses ......... b. Vertical Stresses ................

R

D+L+E+R

3400 1800

36,000 20,000

1720 877

34,000 19,644

1.97 2.06

1.06 1.02

2. RB Biological Wall Beam a. At face .......................... b. At face .......................... c. At midspan ....................... d. At midspan .......................

D+L+E D+L+E' D+L+E D+L+E'

1800 3400 1800 400

20,000 36,000 20,000 36,000

950 1080 1640 2380

18,100 20,800 16,400 23,600

1.90 3.14 1.16 1.43

1.10 1.73 1.22 1.53

3. Control Room Footing a. Face of column ................... b. Face of column ...................

D+L+E D+L+E'

1350 2550

20,000 36,000

450 810

15,600 28,200

3.00 3.15

1.28 1.28

4. Intake Structure Service Bay a. South enclosure ..................

D+L+W'

2550

36,000

300

15,000

8.50

2.40

NOTE: Loads as defined in Section 3.1.3.


3.2 FACILITY STRUCTURES

3.2.1 Reactor Building

3.2.1.1 Function

The Reactor Building encloses the spent fuel storage pool.

3.2.1.2 Description

The Reactor Building is constructed of monolithic reinforced concrete floors and

walls to the refueling level. Above the refueling level, the structure consists of

steel framing covered by insulated sealed siding and roof decking. The siding and

roofing can withstand a limited internal overpressure before pressure relief is

obtained by venting through the refuel floor blowout panels designed to release at

an overpressure falling in the designated range of 0.35 psi to 0.60 psi (Reference

1).

A 110/7.73 ton capacity overhead bridge crane provides services for the reactor and

refueling area. The crane is designed to remain on the rails and retain its load

with a 0.2g seismic loading. The Reactor Building bridge crane is of Class II

seismic design. Accordingly, the coefficient of 0.20g was specified based on the

building response of the level of crane supports under 0.07g minimum ground

acceleration. The crane supports are of Class I seismic design. The crane bridge

and trolley wheels are provided with seismic hold-down lugs to assure crane

stability in the event of a maximum hypothetical earthquake.

Reference 2 details the commitments to control the handling of heavy loads,

including the specific commitments made during the submittal process to the NRC, as

input to their Safety Evaluation Report, and how they are implemented at Vermont

Yankee. The Reactor Building overhead bridge crane trolley was modified to provide

redundancy in the load carrying path from the load to the crane itself, so that no

single failure would allow the load to drop. All components in the load path of the

main hoist are either redundant or designed with a large factor of safety, and are

structurally adequate to maintain the load capacity, as well as any transfer loads

should one path fail. Each load path for the main hook consists of a hook or

attachment point, load block, cable, reversing sheaves, drum, gear drive, and

brakes. Sheaves and blocks are captured so that failure would not result in

uncontrolled descent of the load. Redundant limit switches, of different types, are

provided to prevent over-hoisting, and a load indicating/limiting device prevents

overloading. An overspeed switch is provided on each load path to prevent runaway

lowering. Operating power and control for all crane motions are provided by a

control system which incorporates a torque limiter on the main hoist for additional

overload protection.


When moving a spent fuel shipping cask, the crane speeds are reduced and the travel

path limited to prevent the cask from passing over the stored spent fuel.

The crane was designed in accordance with the Electric Overhead Crane Institute

(EOCI) Specification No. 61 and, with minor exceptions, meets all requirements of

the Crane Manufacturers Association of America (CMAA) Specification No. 70.

The primary containment structure is an integral part of the Reactor Building and

occupies the core of the building. The spent fuel storage pool is located in the

Reactor Building. Access to the drywell and reactor head space is obtained by

removing a large segmented concrete plug in the refueling level floor by means of

the bridge crane. The crane also handles the drywell head, the reactor vessel head,

the segmented pool plugs, and the spent fuel shipping cask. A refueling platform,

with the requisite handling and grappling fixtures, services the spent fuel storage

pool. A passenger-freight elevator is provided for access to the various floors

above grade level.

The steel drywell vessel is fixed to the building along its lower portion, and is

laterally supported by the building along its upper portion. Within the drywell, a

cylindrical sacrificial shield structure surrounds the reactor vessel.

There is a remote possibility that the height of ground water during a given period

could exceed the elevation of the extreme lower portion of the drywell.

Nevertheless, it is not considered possible for this ground water to reach the steel

plating, assuming a crack in the foundation concrete. The bases for this conclusion

are as follows:

1. The monolithic foundation concrete structure is greater than 18 feet thick

below the drywell and is divided into three separate pours in the horizontal

plane. It is considered almost impossible for a crack to propagate completely

through any given pour because of the thicknesses involved and the bedrock

foundation. Even if this were to occur, it is not considered possible for any

given crack to propagate beyond the joint between pours.

2. Water-stop material is used at all foundation concrete joints between pours,

both in the horizontal and vertical planes. This design assures that water

will not propagate along any given joint in the concrete.

The possible effects of a given thermal gradient through the foundation concrete

have been considered. Based on the concrete thicknesses and possible temperature

differentials, it is not considered possible for any thermal gradients to exist

which would damage or otherwise affect the structural integrity of the concrete.

Therefore, thermal gradients are not considered a factor in the above discussion on

foundation cracking.


The reinforced concrete portion of the Reactor Building has been designed against

tornado missiles. Pressure relief below the refueling level is obtained through

large open hatches.

The general arrangement of the Reactor Building and the principal equipment is shown

on Drawings G-191148, G-191149 and G-191150.

3.2.1.3 Seismic Analysis

Dynamic earthquake analysis was made of the coupled Drywell/Reactor Building System

for an empty and flooded condition of the drywell. A separate analysis was made for

the pressure suppression chamber.

The effect of the adjacent Class II Turbine Building has been considered, and the

analysis shows that failure of the adjacent Turbine Building will not compromise the

integrity of the Class I Reactor Building in the event of a design basis or maximum

hypothetical earthquake.

The sacrificial shield wall and reactor pedestal are hollow cylinders of uniform

thickness connected by anchor bolts embedded in the top of the pedestal.

The pedestal carries the vertical load of the sacrificial shield wall including the

loads transmitted to it. The pedestal is supported at Elevation 238.0' by a

concrete foundation which rests on the lower part of the containment vessel.

Moments, vertical loads, and horizontal forces from the reactor pressure vessel,

pedestal, and drywell are transmitted to the supporting drywell foundations in the

following manner:

The reactor pressure vessel transmits vertical loads and shears directly to the

drywell foundation through the vessel skirt into the reactor pedestal via shear

rings welded to the inner skirt. The vertical and horizontal loads from the

pedestal are transferred to the interior and exterior surface of the drywell by a

combination of bond and friction forces between steel and concrete contact surfaces.

The contact between the exterior surface of the drywell and the supporting concrete

foundation is assured by the pressure grouting method used for the concreting of the

foundation itself. Additional resistance to shear is afforded by the physical

characteristics of the drywell which, in its lower portion, can be considered as a

bowl embedded in the supporting reinforced concrete foundation. No increase in

allowable stresses was permitted in any of the above considerations.

The stresses resulting from the maximum hypothetical earthquake were also checked to

make sure that their value was below allowable limits.


The interaction of the drywell base with the exterior concrete is comprised of

bonding and friction, and it is a result of these phenomena that the relative shears

are handled.

The phenomenon of bonding, although a significant contributory factor, is ignored

for conservatism. Extreme care is exercised in placing the grout between the

drywell base and the exterior concrete. This provides adequate assurance that there

are no significant voids in this area and that the actual drywell contact area is

high. In addition to providing significant bonding, this surface area also provides

a large contact area to resist relative shears through friction.

The vertical load transmitted through the drywell is approximately 8,230k. The

horizontal load resulting from a maximum hypothetical earthquake is 3,165k. To be

conservative, the calculations assume that vertical, horizontal and moment forces

are transmitted from the drywell to the foundation mat by the reactor vessel skirt

alone. It is further assumed that the reactor vessel skirt, welded to the drywell,

will transmit the horizontal forces by bearing against the fill concrete surrounding

it. For conservatism, only the top two feet of the skirt were considered as

transmitting the load.

The concrete stresses and welding stresses were checked against the allowable

stresses to determine if the skirt and the surrounding concrete can withstand the

horizontal forces. The concrete stress is 638 psi, which is less than the 1,000 psi

allowed by ACI 318, 1963. The unit shear stress on the skirt weld is 488 psi, which

is small in comparison with the load-carrying capability of the weld.

The ability of the foundation mat to resist shear forces was also investigated. No

credit was taken for the anchor bolts which fasten the skirt to the foundation mat,

and friction alone is assumed to resist shear forces. A coefficient of friction was

conservatively assumed to be 0.4, which results in a shear resisting force

capability of 3,292k. As the maximum horizontal load is 3,165k, the adequacy of the

foundation mat is demonstrated.

3.2.2 Turbine Building

3.2.2.1 Function

The Turbine Building SSCs have been abandoned.

3.2.2.2 Description

SSCs within the Turbine Building have been abandoned. Equipment and floor drain

sumps are routed to a batch tank. Tank contents are sampled prior to being

transferred, disposed of (via offsite shipments), or discharged to the environs in

accordance with applicable permits and regulatory approvals.


3.2.3 Plant Stack

3.2.3.1 Description

The plant stack provides an elevated point for the release of gases to the

atmosphere from portions of the Turbine Building, Reactor Building, and Radwaste

Building. Stack drainage is routed to the Liquid Radwaste Collection System via

loop seals.

The plant stack is designed for dead load, wind load, seismic load, and effects of

exhaust gas temperature. The plant stack is provided with appurtenances such as

aviation obstruction lights and isokinetic samplers for radiation monitoring, and is

designed in accordance with all applicable codes.

The unlined, freestanding, tapered, reinforced concrete stack has the following

dimensions:

Overall height above foundation 318 ft

Inside diameter at top 7 ft

Outside diameter at base 27.5 ft

Thickness at top 0.67 ft

A schematic of the stack geometry appears in Figure 3.2-18.


Dynamic analysis was made of the reinforced concrete ventilation stack.

The foundation material for the site is such that rocking effects are small and were

neglected for the dynamic analysis of the stack. Due to its geometry, the stack is

very flexible, with the natural period of vibration of the first mode equal to about

1.5 seconds. The spectral accelerations for periods higher than this decrease with

an increase in period. The model used for the dynamic analysis of the stack

conservatively assumed a fixed base. The damping value used in the analysis for the

responses to both the design basis and maximum hypothetical earthquakes was 5%.

The controlling loading conditions were the maximum hypothetical earthquake in the

region approximately 120 feet from top and the wind loading for the remaining

portion of the stack. For the maximum hypothetical earthquake, the maximum

calculated stress in the reinforcing steel was 35.4 ksi or 0.86 Fy, and the

calculated maximum stress in concrete was 1.32 ksi or 0.377 f c.


3.2.4 Control Room Building

3.2.4.1 Description

The Control Room Building houses all required instrumentation and controls. The

instrumentation is located in the Main Control Room. The cable vault and Switchgear

Room occupy the lower levels of the building. The location of the Control Room

Building is shown on Drawing G-191142. The building is a reinforced concrete

structure and is entirely of Class I seismic design.

Plan and elevation views of the building are shown on Drawings G-191592 and

G-191595.


A dynamic earthquake analysis was performed on the Control Room Building utilizing a

four-mass analytical model. The effect of the adjacent Class II Turbine Building

has been considered, and the results of the analysis show that failure of the

adjacent Class II structure will not compromise the integrity of the Class I Control

Room Building in the event of a design basis or maximum hypothetical earthquake.

3.2.5 Circulating Water Intake and Discharge Structures

3.2.5.1 Intake Structure

3.2.5.1.1 Description

A reinforced concrete, single unit intake structure on the riverbank east of the

station, is supported on rock. A partial enclosure is provided at the pumps. The

following equipment is provided at the intake: manually raked coarse trash racks,

regulating sluice gates; traveling screens; provisions for stoplogs; two fire water

pumps; two service water pumps and two radwaste dilution pumps.

The deck of the structure is at Elevation 237' MSL and the invert at Elevation 190'

MSL. The intake has service water bays for two service water pumps, two fire water

pumps, and two radwaste dilution pumps. Bays are provided with trash rack and

stoplog guides, traveling screens, and fine screen guides.

Water from the pond flows into service water bays at the north end of the intake

structure. These bays furnish water for the fire pumps, intake service water pumps,

and radwaste dilution pumps.

Retaining walls are provided at the front face of the intake structure to retain

fill.


The intake structure is shown on Drawings G-191451, G-191452 and G-191453.

3.2.5.1.2 Seismic Analysis

A dynamic earthquake analysis has been made of the intake structure. This analysis

verifies the adequacy of the design of the intake structure to withstand seismic

forces. The effect of adjacent Class II intake structures has been considered, and

the results of the analysis show that failure of an adjacent Class II structure will

not compromise the integrity of the Class I bay housing the service water pumps in

the event of a design basis or maximum hypothetical earthquake.

3.2.5.2 Discharge and Aerating Structure

A reinforced concrete discharge-aerating structure supported on rock and piles is

located near the riverbank south-southeast of the station. It is approximately 188

feet long by 108 feet wide by 46 feet deep. The top of the deck is at Elevation

248' MSL. Water elevation for siphon operation will be maintained by a reinforced

concrete weir. The top of the weir is at Elevation 225' MSL. An aerating spillway

concrete structure is adjacent and downstream of the discharge structure to provide

air entrainment, energy dissipation, and warm water dispersion of discharged water.

Sheet piling is used to prevent scour of the aerating apron.

The discharge and aerating structure is shown on Drawings G-191463, G-191461, Sh. 1

and G-200347.

3.2.6 Cooling Tower Deep Basin

The basin has been dynamically analyzed for 0.07g and 0.14g horizontal ground

accelerations; vertical accelerations were taken as 0.05g and 0.10g for the design

basis and maximum hypothetical earthquake, respectively.

The effect of adjacent Class II structures has been considered, and the analysis

show that a failure of the Class II adjacent cooling tower structures will not

compromise the integrity of the deep basin in the event of a design basis or maximum

hypothetical earthquake.


3.2.7 Independent Spent Fuel Storage Installation

3.2.7.1 Description

The ISFSI Storage Pad (comprised of an East and West pad) is monolithic reinforced

concrete slabs supported by compacted structural fill placed on existing soils. The

two storage pads provide structural support for up to 58 spent fuel storage casks

with four extra positions to provide sufficient room to be able to access any

individual cask should the need arise, and 3 spaces available for storage of

Greater-than-Class-C (GTCC) storage casks. The East Storage Pad can store up to 40

casks arranged in a 5 X 8 array. The West Storage Pad can store up to 25 casks in a

5 X 5 array. The spent fuel storage casks are free standing on the pad. There is

temperature monitoring available for each cask if desired. Each cask will be

grounded to plates embedded in the storage pad. The top of the pad elevation is

established at El. 254’-0” to ensure that the ventilation inlets at the bottom of

the spent fuel storage casks remain above the Probable Maximum Flood (PMF) elevation

including wave run-up.

The spent fuel cask manufacturer’s Final Safety Analysis Report (Reference 3)

requires that for free standing casks several criteria must be met to ensure that

the design features of the cask that protect the spent fuel from a cask drop or non-

mechanistic tip-over event are not jeopardized. These criteria are that the

thickness of the pad does not exceed 36 inches, the 28 day concrete compressive

strength must not be less than 3000 psi and must not exceed 4200 psi, the specified

minimum yield strength for the reinforcing steel be 60 ksi, and that the subgrade

modulus of elasticity not exceed 28,000 psi.



A dynamic analysis of each of the ISFSI storage pads was performed. This analysis

is composed of several parts. A subsurface investigation was performed to establish

bedrock elevations and soil properties beneath each pad (References 4 for the East

pad and 10 for the West pad). The design of the East pad meets the requirements of

Revision 3 of Section 3.7.1 of NUREG-0800 which was in effect at the time of its

design. A single set of three artificial time histories for the Design Basis

Earthquake was developed for input to the seismic analysis (Reference 5). The

design of the West pad meets the requirements of Revision 4 of Section 3.7.1 of

NUREG-0800 which was in affect the time of its design. Five sets of three

artificial time histories for the Design Basis Earthquake were developed for input

to the seismic analysis (Reference 12). These time histories envelope the design

response spectra for the site, the North 69º West component of the Taft Earthquake,

normalized to 0.14g for the Design Basis Earthquake. The earthquake(s) is applied

at the bedrock elevation under the storage pad. Analysis was then performed to

obtain strain compatible soil properties and to propagate the earthquake motion from

the bedrock to the ground surface. Since the bedrock under the storage pad is

sloping, this analysis was performed for two profiles, one profile to the deepest

bedrock depth under each pad and one profile to the shallowest bedrock depth under

each pad. This analysis is further described and provided in Reference 6 for the

East pad and 13 for the West pad. A soil structure interaction (SSI) analysis was

then performed to determine the acceleration at the center of gravity and at the

base of the casks. This analysis was performed using three separate soil cases

(upper bound, best estimate, and lower bound). The analysis also considered two

soil profiles to represent the sloping bedrock. The SSI analysis evaluates multiple

cask configurations to insure the maximum effect on the storage pad is enveloped.

The soil structure interaction analysis is further described and presented in

Reference 7 for the East pad and 14 for the West pad.

The results of the soil structure interaction analysis are used to perform a sliding

analysis and the storage pad design. The sliding analysis determines the potential

for the casks to:

(1) slide into each other, and

(2) uplift a seismic event.


The sliding analysis evaluated coefficients of friction ranging from 0.0 to

stimulate icing conditions on the pad up to a maximum of 0.8. The results of the

analysis show that the maximum horizontal displacements of the casks for any

condition are much smaller than half the free distance between the casks and much

less than the distance between the edge of the external casks and the edge of the

pad. This analysis also shows that the casks are stable and remain upright. The

sliding analysis is provided in Reference 8 for the East pad and 14 for the West

pad. References 9 (East pad) and 16 (West pad) provide the analysis to determine

the internal forces on the storage pad for all loading conditions, including

seismic, and the design of the reinforcement for the storage pad.

3.2.8 References

1. Calculation VYC-1828, “Reactor Building Masonry Wall Review for HELB Loadings.”

2. PP 7023, “Control of Heavy Loads Program Document.”

3. Final Safety Analysis Report for the Holtec International Storage and Transfer

Operation Reinforced Module Cask System (HI-STORM 100 Cask System), NRC Docket

No. 72-1014, Holtec Report HI-2002444, Volume I and II of II, prepared by Holtec

International, Marlton, New Jersey.

4. Geotechnical Engineering Report, Proposed ISFSI Pad and Haul Path – Vermont

Yankee, prepared by GZA GeoEnvironmental, Inc., Manchester, New Hampshire,

January 2004

5. Calculation VYC-2427, “Development of Acceleration Time Histories for Vermont

Yankee ISFSI Analysis.”

6. Calculation VYC-2428, “Development of Strain Compatible Soil Properties for

Vermont Yankee ISFSI Analysis.”

7. Calculation VYC-2433, “Soil Structure Interaction Analysis of the Vermont Yankee

ISFSI.”

8. Calculation VYC-2434, “Vermont Yankee ISFSI Cask Sliding Analysis.”

9. Calculation VYC-2435, ”Vermont Yankee Nuclear Power Plant ISFSI Facility Concrete

Storage Pad Design”

10 Report VY-ROT-14-00005, "Geotechnical Soils Report for DFS-PAD-2 – Data Report to

Support the Expansion of the Independent Spent Fuel Storage Installation

(ISFSI)."


11. Calculation VYC-3175, "Determination of Soil Parameters for ISFSI Expansion

Concrete Storage Pad."

12. Calculation VYC-3176, "Development of Response Spectra Consistent Time Histories

for ISFSI Expansion Concrete Storage Pad."

13. Calculation VYC-3177, "Development of Strain Dependent Soil Properties for ISFSI

Expansion Concrete Storage Pad."

14. Calculation VYC-3178, "Soil Structure Interaction Analysis and Cask

Stability/Sliding of ISFSI Expansion Concrete Storage Pad."

15. Calculation VYC-3179, "Liquefaction Potential for ISFSI Expansion Concrete

Storage Pad."

16. Calculation VYC-3181, "Structural Concrete Design for ISFSI Expansion Concrete

Storage Pad."


Vermont Yankee


Main Stack Geometry

Figure 3.2-18


3.3 SYSTEMS

The following systems have been or are in the process of being abandoned and removed

from service. Abandonment includes, where appropriate, draining piping and tanks,

removing electrical power, removal of combustible liquids and placing the abandoned

SSC in its lowest energy condition.

High Pressure Coolant Injection System

Main Steam

Heater Drains and Vents

Automatic Depressurization System

Air Evacuation, Auxiliary Steam, Advanced Off Gas

Condensate & Condensate Demineralizer System

Containment Air Dilution System

Circulating Water & Circulating Water Priming System (includes cooling tower

equipment)

Feedwater & Feedwater Controls Systems

Hydrogen, Hydrogen Water Chemistry, Nitrogen Supply & Oxygen Injection Systems

MG Lube Oil System

Reactor Protection and Primary Containment Isolation System

River Water Temperature and Toxic Gas Monitoring Systems

Reactor Core Isolation Cooling System

Recirculation Pumps, MG Sets & Flow Control System

Main Turbine Generator, TBCCW, Stator Cooling Seal Oil, Lube Oil, Isophase Bus

Cooling

Standby Liquid Control System

22K and 345K Volts AC Electrical System

Control Rod Drive & Hydraulic Control Unit Systems

Core Spray System

Nuclear Boiler and Nuclear Boiler Vessel Instrumentation Systems

Neutron Monitoring System

Reactor Building Closed Cooling Water System

Residual Heat Removal & RHR Service Water Systems

Radwaste System

Process Rad Monitor and Turbine Building Area Rad Monitor

Reactor Water Clean-Up System

Demineralized Water Transfer and Makeup Demineralizer Systems

Post Accident Sampling System

Primary Containment/Penetration System


Primary Containment Atmospheric Control

Emergency Diesel Generator and Fuel Oil Systems

Normal Fuel Pool Cooling System (FPCS)

Fire Protection System, Sprinklers/Detectors (Partial Abandonment)

Potable Water Reconfiguration For SAFSTOR

400V DC System

Service Water System (Partial Abandonment)

Removal Of Low Level Radwaste Site and Other Non-SSC Buildings

24V DC RPS Neutron Monitoring System Batteries

Vital MG Set MG-2-1A

Stack Gas III Radiation Monitor (RM-17-155)

Sentry Lights

3.3.1 Fuel Storage and Handling

3.3.1.1 Nuclear Fuel

3.3.1.1.1 Objective

The nuclear fuel provides a high integrity assembly containing fissionable material

which could be arranged in a critical array. The assembly efficiently transfers

decay heat to the spent fuel pool water while maintaining structural integrity and

containing the fission products.


A fuel assembly consists of a fuel bundle, channel fastener, and the channel which

surrounded it. Each fuel assembly was designed as Class I seismic design equipment.

A fuel bundle contains fuel rods and water rods, spaced and supported in a square

array by a lower tie plate, spacers, and an upper tie plate. The lower tie plate was

formed and machined to fit into the fuel support piece. The lower tie plate for the

GE13, GE14 and GNF2 fuel bundles also includes a debris filter. The upper tie plate

has a handle for transferring the fuel bundle from one location to another. The

identifying assembly number is engraved on the top of the handle and a boss projects

from one side of the handle to aid in assuring proper fuel assembly orientation.

The tie plates were fabricated from corrosion resistant materials. The fuel spacer

grids, which are positioned along the length of the fuel bundle, are made of

Zircaloy with Inconel springs. The GE13 and GE14 fuel spacer grids, which are

positioned along the length of the fuel bundle, are made of Zircaloy with alloy X750

springs. The GNF2 spacer is made entirely from alloy X750. The primary function of

the spacer grid is to provide lateral support and spacing of the fuel rods.


Each fuel rod consists of fuel pellets stacked in a Zircaloy cladding tube which is

evacuated, pressurized with helium, and sealed by welding Zircaloy end plugs in each

end. The fuel rod cladding thickness is adequate to be "free-standing", i.e.,

capable of withstanding external reactor pressure without collapsing onto the

pellets within. Although most fission products were retained within the UO2, a

fraction of the gaseous products were released from the pellet and accumulated in a

plenum and the gap between the pellet stack and the clad. Sufficient plenum volume

was provided to prevent excessive internal pressure from these fission gases or

other gases liberated over the design life of the fuel. A plenum spring, or

retainer, is provided in the top plenum space to minimize movement of the fuel

column during handling or shipping. Rigid precautions are taken to prevent cladding

damage due to excessive hydrogen bearing materials. These precautions may include a

hydrogen getter in the plenum to absorb hydrogen accidentally admitted during the

fabrication process.

Eight fuel rods (called tie rods) in each bundle have end plugs which thread into

the lower tie plate and extend through the upper tie plate. Stainless steel nuts

and locking tab washers are installed on the upper end plugs to hold the assembly

together. These tie rods support the weight of the assembly only during fuel

handling operations when the assembly hangs by the handle. The remaining fuel rods

in a bundle have end plug shanks which fit into locating holes in the tie plates.

An Inconel-expansion spring located over the top end plug shank of each full length

fuel rod keeps the fuel rods seated in the lower tie plate and allows them to expand

axially by sliding within the holes in the upper tie plate to accommodate

differential axial expansion. Part length rods use a threaded lower end plug which

screws into the lower tie plate. These rods terminate near one of the spacer grids

short of the upper tie plate.

Each fuel bundle may contain one or more empty Zircaloy tubes called water rods.

Perforations at each end of the water rod(s) permit coolant flow through the tube.

Tabs are fixed at axial intervals on one or more water rods to locate the spacer

grids. Water rods provide additional moderator throughout the height of the

assembly.

The fuel is in the form of cylindrical pellets manufactured by cold pressing and

sintering uranium dioxide powder. The average density of the pellets in the core is

approximately 96.5% of the theoretical density of UO2. Ceramic uranium dioxide is

chemically inert to the cladding at operating temperatures and is resistant to

attack by water.


Several different U-235 enrichments may be used in each fuel assembly. Fuel design,

manufacturing, and inspection procedures have been developed to prevent errors in

enrichment location within the fuel assembly. The fuel rods have unique

identification numbers. Rigid inspection techniques utilized during and following

assembly ensure that each fuel rod is in the correct position within the bundle.

Selected fuel rods contain gadolinia as a burnable poison for reactivity control.

The gadolinia is uniformly dispersed within the fuel pellets. However, the

gadolinia-bearing pellets are not uniformly distributed within the fuel rods, but

are grouped together into axial zones. These axially zoned regions of varying

gadolinia content provide reactivity control which enhances shutdown margin and/or

power distribution control to reduce axial peaking. U-235 enrichment is also zoned

axially to compliment the function of the gadolinia, and provide a more economical

fuel cycle.

The fuel channel enclosing the fuel bundle is fabricated from Zircaloy and, if

installed, performs the following functions:

1. Provides structural stiffness to the fuel bundle during lateral loading applied

from fuel rods through the fuel spacers.

2. Transmits fuel assembly seismic loadings to the top guide and fuel support of

the core internal structures.

The channel makes a sliding seal fit over finger springs attached to the lower tie

plate. The channel is attached to the upper tie plate by the channel fastener

assembly which is secured by a cap screw. Spacer buttons are located on the two

sides of the channel adjacent to the channel fastener assembly to maintain bundle

separation and form a path for the control blades in the core cell.

GNF2 fuel assemblies are arranged in a 10X10 array with two central water rods, as

well as both short and long partial length rods. Some of the design features

include the following:

Improved part-length rod configuration for improved Cold Shutdown Margin (CSDM)

and efficiency.

Modified fuel rod clad thickness to diameter ratio (T/D) with increased uranium

mass for increased bundle energy.

Modified channel that interacts with the LTP to control leakage flow while

eliminating finger springs for ease of channeling operations.

Improved Inconel X-750 grid type spacer with Flow Wings for increased margin to

Boiling Transition and reduced pressure drop.


Defender Debris Filter Lower Tie Plate for improved resistance to the intrusion

of foreign material.

High volume pellet for increased uranium mass and manufacturing quality control.

Locking retainer spring that restrains the fuel column during shipping and

supports a wide range of column lengths.

A non-Zircaloy 2 zirconium alloy, Ziron, is used for the fuel cladding material

for 24 rods in 2 of the 4 GNF2 LUAs.

The external envelope of GNF2 is virtually identical to GE14 and the nuclear

characteristics of the GNF2 are compatible with current vintage GE14. The thermal

hydraulic characteristics of GNF2 design closely match the overall pressure drop of

previous designs.

Licensing analyses of the GNF2 LUAs have been conducted using NRC approved methods,

which are capable of evaluating/analyzing all of the LUA features.

3.3.1.2 Spent Fuel Storage

3.3.1.2.1 Objective

The spent fuel storage arrangement provides specially designed underwater storage

space for the spent fuel assemblies which require shielding during storage and

handling.

Storage of spent fuel in dry casks at the Independent Spent Fuel Storage

Installation facility is licensed in accordance with 10CFR72 and is not within the

scope of the 10CFR50 Updated Final Safety Analysis Report.

3.3.1.2.2 Design Bases

1. The spent fuel pool is designed for a maximum of twelve spent fuel storage

racks with a maximum capacity of 3,353 spent fuel assemblies.

2. Spent fuel storage racks shall be designed and arranged so that the fuel

assemblies can be efficiently handled.

3. The fuel array in the fully loaded spent fuel racks shall be substantially

subcritical such that keff is less than or equal to 0.95.

4. Each spent fuel storage rack shall be designed to withstand earthquake loading

to prevent significant distortion of spent fuel storage arrangement when empty,

half-full, or fully loaded with fuel.



The spent fuel storage racks provide storage at the bottom of the fuel pool for the

spent fuel received from the reactor vessel, as shown in Figure 3.3.1-1. The racks

are full length, top entry, and designed to maintain the spent fuel in a space

geometry which precludes the possibility of criticality under normal and abnormal

conditions. Normal conditions exist when the spent fuel is stored at the bottom of

the fuel pool in the design storage position. Abnormal conditions may result from

an earthquake or mishandling of equipment.

The normal arrangement of the spent fuel storage racks consists of nine NES

manufactured racks (Drawing 5920-6893) and two Holtec manufactured racks (Drawing

5920–12795), giving a total capacity of 3087 assemblies. A twelfth rack can be

installed in the cask lay-down area as shown on Drawing 5920–12795 to provide

additional full core discharge capacity and a total pool capacity of 3,353

assemblies. The control rod blade (CRB) storage rack shown on Drawing 5920–6893

will be unloaded and removed if a twelfth rack needs to be installed or when the

cask pad must be used, such as for an irradiated hardware disposal campaign.

Partial plans depicting the Boral loading are provided in Figures 3.3.1-4 and 3.3.1-

5. The spent fuel storage racks are designated Safety Class 2.

Each rack consists of a welded assembly of individual storage cells in a staggered

checkerboard array. The storage cells are comprised of Type 304L stainless steel

boxes (5.922 inches square ID) welded to each other with corner angles to maintain a

pitch of 6.218 inches. The rack dimensions are 178.50 inches tall, 87.43 inches to

125.27 inches long, and 74.99 inches to 112.83 inches wide. Each storage cell has

an interior height of 168 inches. The construction of the storage cells provides

four vented (open to the pool) compartments in which B4C neutron absorber elements

are placed for criticality control. The neutron absorber elements are positioned on

the side of the storage cell at an elevation corresponding to the fuel region of a

spent fuel assembly placed within the cell. The bottom of each storage cell sits

on, and is welded to, the rack base plate which provides the level seating surface

required for each fuel assembly and also contains the openings necessary for

adequate cooling flow. Drawing 5920-6893 shows a schematic drawing of a typical

rack.

All materials used in the construction of the rack are specified in accordance with

the applicable ASME or equivalent ASTM specification, and all welds are in

accordance with ASME Section II, for materials used, and ASME Section IX. Materials

selected are corrosion-resistant or treated to provide the necessary corrosion

resistance.


The maximum number of assemblies stored in the pool cannot exceed 3,353.

Each rack is freestanding with no lateral restraints to the wall, and is supported

by a minimum of four steel feet that transfer load to the pool floor. Any lateral

loads on the racks will be transferred by friction between the feet and the pool

floor. The racks are designed such that a fuel assembly or grappling device cannot

become fouled during removal and, thereby, generate significant uplift loads.

No spaces exist between normal fuel storage positions so that it is not possible to

insert a fuel assembly, either deliberately or by accidental drop, in any position

not intended as a fuel storage position.

Each spent fuel storage rack loaded with fuel has been analyzed to determine its

continued operability during and after both design basis and safe shutdown

earthquakes. It has been determined that under the most severe seismic loading

condition, the rack will slide a maximum of 0.56 inches. A clear distance of 2

inches (minimum) is maintained between spent fuel storage racks, spent fuel storage

racks and walls, and spent fuel storage racks and any other objects in the pool. A

clear distance of 5 inches (minimum) is maintained between the Control Rod Blade

(CRB) storage racks and any other large or fixed objects in the pool. A clear

distance of 6.24 inches and 10.0 inches (minimum) is maintained between the CRB

storage racks and the NES and Holtec spent fuel storage racks respectively.

The fuel storage pool is designed so that no single failure of structures or

equipment will cause inability to (1) maintain irradiated fuel submerged in water,

(2) re-establish normal fuel pool water level, or (3) safely remove fuel from the

plant. In order to limit the possibility of pool leakage around pool penetrations,

the pool is lined with stainless steel. In addition to providing a high degree of

integrity, the lining is designed to withstand abuse that might occur when the

transport cask is moved about. No inlets, outlets, or drains are provided that

might permit the pool to be drained below approximately 10 feet above the top of the

active fuel. Lines extending below this level are equipped with valving.

Interconnected drainage paths are provided behind the liner welds. These paths are

designed to (1) prevent pressure buildup behind the liner plate, (2) prevent the

uncontrolled loss of contaminated pool water to other relatively cleaner locations

within the secondary containment, and (3) provide expedient liner leak detection and

measurement. These drainage paths are formed by welding channels behind the liner

weld joints and are designed to permit determination of liner weld leakage.


The spent fuel pool is 26 feet-0 inches wide by 40 feet-0 inches long by 39 feet-3/4

inches deep. The pool is completely lined with seam-welded ASTM-A240, Type 304

stainless steel. The floor plate is 1/4-inch thick and the wall plate is 3/16-inch

thick. Pipe sleeves are welded to the liner plate by full circumferential fillet

welds on both sides of the plate.

All welds above the waterline were visually examined. Those welds which could be

exposed to water were examined by liquid penetrant tests. In addition, all joint

welds and welds at penetrations in plates were tested for leaks using a vacuum box

and soap solution tests.

3.3.1.2.4 Safety Evaluation

Administrative controls ensure a sufficient level of water is maintained to ensure

shielding and/or cooling.

The design of the spent fuel storage racks provides for a subcritical multiplication

factor (keff) for both normal and abnormal storage conditions.

For all conditions, keff is equal to or less than 0.95. Normal conditions exist when

the fuel storage racks are located at the bottom of the pool covered with a normal

depth of water (about 23 feet above the stored fuel) for radiation shielding and

with the maximum number of fuel assemblies in their design storage position. The

spent fuel is covered with water at all times by a minimum depth required to provide

sufficient shielding. Abnormal conditions may result from an earthquake, accidental

dropping of equipment, or damage caused by the horizontal movement of fuel handling

equipment without first disengaging the fuel from the hoisting equipment.

Accidental dropping of large pieces of equipment, such as a spent fuel shipping

cask, is prevented by the use of an overhead bridge crane with redundant load

bearing equipment on the main hoist.

Criticality calculations were done using a two-dimensional, two-group diffusion

theory code with a water temperature of 39°F. Water temperatures were varied

between 39°F and 248°F to assure that 39°F was the more reactive under normal

conditions. Monte Carlo calculations and verifications assured the adequacy of the

diffusion theory representation.

In order to ensure that the design criteria stated above are met, the following

loading conditions have been analyzed. The results include allowance for

calculational uncertainty.


The off-normal conditions evaluated are:

1. Normal positioning in the NES and Holtec spent keff = 0.9469

fuel storage array

2. Eccentric positioning in the NES and Holtec spent less than 0.9469

fuel storage array

3. An assembly was placed tightly in the corner formed less than 0.9267

by an L-shaped junction of three racks (NES racks only)

Stress in a fully loaded rack will not exceed applicable stress limits for Seismic

Category I structures per requirements of the NRC Standard Review Plan,

Section 3.8.4. Horizontal acceleration time history data derived from a Bechtel

calculation and maximum vertical seismic acceleration were applied simultaneously.

Maximum vertical acceleration was taken from the applicable vertical spectra at the

fundamental vertical frequency of the rack. The stresses, due to partially loaded

and empty rack conditions, are smaller than the full loaded condition.

The storage rack structure is designed to absorb the vertical impact force imposed

by a fuel assembly dropped from a height of 36 inches above a rack onto any location

on the rack. Under this impact force, those members will remain intact whose

function it is to physically maintain the normal design subcritical spacing to

assure keff is less than 0.95.

GE topical report "Tornado Protection for the Spent Storage Pool," APED-5696,

November, 1968 investigated the potential effects of a tornado striking the fuel

storage pool of a boiling water reactor (BWR). Two key concerns were examined; (1)

whether sufficient water could be removed from the pool to prevent cooling of the

fuel, and (2) whether missiles could potentially enter the pool and damage the

stored fuel.

The fuel pool was designed with substantial capability for withstanding the effects

of a tornado. The design of the fuel pool makes the removal of five feet of water

due to tornado action highly improbable. With 25 feet of water covering the fuel

racks, the removal of five feet of water is of no concern. Protection against a

wide spectrum of tornado-generated missiles is provided by the water which covers

the fuel racks.

Protection is provided against all tornado-generated missiles having a probability

of hitting the pool greater than one per 1.4 billion reactor lifetimes. Typical

potential missiles in this category include a spectrum ranging up to a 3-inch

diameter steel cylinder 7 feet long or a 14-inch diameter wooden pole 12 feet long.


The General Electric Company concluded that adequate protection for the fuel pool

against the effects of a tornado was provided and no additional protection was

required.

NUREG-1738, “Technical Study of Spent Fuel Pool Accident Risk at Decommissioning

Nuclear Power Plants” (Reference 1) contains the results of an NRC staff evaluation

of the potential accident risk in spent fuel pools at decommissioning plants in the

United States. The study was undertaken to support development of a risk-informed

technical basis for reviewing exemption requests and a regulatory framework for

integrated rulemaking. The NRC staff performed analyses and sensitivity studies on

evacuation timing to assess the risk significance of relaxed offsite emergency

preparedness requirements during decommissioning. The staff based its sensitivity

assessment on the guidance in Regulatory Guide 1.174, "An Approach for Using

Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes

to the Licensing Basis” (Reference 2). The staff's analyses and conclusions apply to

decommissioning facilities with SFPs that meet the design and operational

characteristics assumed in the risk analysis.

The study found that the risk at decommissioning plants is low and well within the

Commission's Safety Goals. The risk is low because of the very low likelihood of a

zirconium fire (resulting from a postulated irrecoverable loss of SFP cooling water

inventory) even though the consequences from a zirconium fire could be serious.

NUREG-1738, Executive Summary, states in part, "the staff's analyses and conclusions

apply to decommissioning facilities with SFPs that meet the design and operational

characteristics assumed in the risk analysis. These characteristics are identified

in the study as IDCs and SDAs. Provisions for confirmation of these characteristics

would need to be an integral part of rulemaking."

Design and operation of the VY SFP has been evaluated against and confirmed to

comply with the industry decommissioning commitments (IDCs) and staff

decommissioning assumptions (SDAs) contained in NUREG-1738. The evaluation is

documented in BVY 14-009, Request for Exemptions from Portions of 10CFR50.47 and

10CFR50, Appendix E. (Reference 3)

3.3.1.2.5 Inspection and Testing

The spent fuel storage racks were tested at the plant site or visually inspected

during rack fabrication to ensure that the Boral sheets are in place and free of

voids. Since Boral absorbs neutrons, a neutron source and proportional counters

were used to verify the integrity of the Boral sheet.


An inspection mandrel was used to test each storage cell location. The insertion

and withdrawal of the mandrel was monitored over the entire length of the cell to

ensure that acceptable drag forces were not exceeded.

3.3.1.3 Standby Fuel Pool Cooling and Demineralizer Systems

3.3.1.3.1 Objective

The Standby Fuel Pool Cooling (SFPCS) removes decay heat released from the spent

fuel to maintain fuel pool temperature within specified limits. The Fuel Pool

Demineralizer System (FPDS) maintains water clarity.


1. The FPDS shall minimize corrosion product buildup within the spent fuel pool

and shall maintain proper water clarity so that the fuel assemblies can be

efficiently handled underwater.

2. The FPDS shall minimize fission product concentration in the spent fuel pool

water, thereby minimizing the radioactivity which could be released from the

pool to the Reactor Building environment.

3. The Fuel pool water level shall be maintained at a level above the fuel

sufficient to provide shielding for normal building occupancy.

4. The Standby Fuel Pool Cooling System shall be capable of maintaining the spent

fuel pool temperature below 150°F.


The SFPCS is shown on Drawing G-191173, Sheets 1 and 2.

Fuel Pool Structure

The fuel pool concrete structure, metal liner, spent fuel storage racks, and the

SFPCS are designed to withstand Seismic Class I earthquake loads.

FPDS

Fuel Pool clarity is maintained by the FPDS. The FPDS consists of submerged

underwater units which will be operated as required to minimize fission product

concentration and maintain water clarity through demineralization and filtration.


Fuel Pool Makeup and Letdown

Makeup to the pool is supplied by the Torus-as-CST System.

Water may be removed from the fuel pool, if required, via letdown to the Torus.

Fuel Pool Skimmers

Two skimmer pumps are provided which take suction from the top of the pool to remove

surface debris. These pumps circulate fuel pool water through cartridge filters and

return it to the pool through service boxes located around the pool.

SFPCS

The SFPCS functions to maintain pool temperature within specified limits.

An administrative limit of 125°F has been established for maximum fuel pool

temperature during normal cooling and filtering.

The operating temperature of the fuel pool is permitted to rise up to 25°F above the

administrative temperature limit (125°F) as specified in applicable procedures.

The SFPCS is a two train, Seismic Class I, non-safety related system, designed to be

remotely placed in operation from the control room. The SFPCS circulates the pool

water in a closed loop, taking suction from the spent fuel storage pool, through

heat exchangers and discharging the water back into the fuel pool. The SFPCS heat

exchangers transfer the spent fuel decay heat to the seismic Class I, non-

safety-related Station Service Water System (SWS).

The SFPCS includes two seismic Class I centrifugal pumps. All the parts of the pump

in contact with water are corrosion-resistant. A pump low discharge pressure alarm

annunciates in the Control Room. In addition, the pumps trip automatically on low

suction pressure.

The heat exchangers are shell and tube design; all parts in contact with water are

corrosion resistant. These heat exchangers are each sized to maintain fuel pool

water temperature below 150°F.


To minimize the potential for fuel pool water leakage into the Station SWS, service

water pressure is normally maintained greater than SFPCS pressure. The fuel pool water

side of the heat exchangers has a maximum operating pressure equivalent to the static

pressure head from the pool surface to the heat exchanger. The Station SWS side of the

heat exchangers has a minimum operating pressure which is normally greater than the

maximum pressure on the fuel pool side of the heat exchangers as long as the operating

SW pumps can maintain system header pressure above the pressure which results in NNS SW

header isolation valve closure. During events which result in low service water header

pressure, service water pressure may be lower than SFPCS pressure until the SW header

isolation valves are closed. By maintaining a positive differential pressure, leakage

of fuel pool water to the environment is prevented. The differential pressure across

each heat exchanger is monitored by a differential pressure sensor and displayed in the

control room.

Two motor operated throttling valves, V70-257A and 257B, provide service water flow

control through the respective SFPCS heat exchanger to control both pool temperature

and service water to SFPCS differential pressure.

Two motor operated isolation valves, V19-220 and 221, close on low pool level,

providing automatic pool isolation in case of a line break in the non-seismic

portion of the system.

SFPCS heat exchanger supply and return service water piping and SFPCS piping is

corrosion-resistant. The piping meets the requirements of ANSI B31.1-77.

Indication is provided in the control room and/or locally near the equipment.

Control Room indication for each train includes direct pool temperature, fuel pool

water temperature out of the heat exchangers (taken downstream of the pumps), pump

run lights, pump discharge pressures, service water flow, SWS to SFPCS heat

exchanger DP and valve position lights. Local indication includes fuel pool water

temperature into the heat exchangers, pump suction and discharge pressures, and heat

exchanger DP. Pool temperature is provided by redundant thermocouples located

within the pool. Pool level is provided by redundant transmitters located near the

pool. All other transmitters and sensors are located in or near the Fuel Pool

Cooling System cubicle.

Controls for the pumps and four MOVs are provided in the control room. Control room

controls include pump on/off switches, service water throttle valves control

switches, and V19-220 and V19-221 isolation valves control switches.


3.3.1.3.4 Evaluation

The SFPCS has a heat removal capability of 11 MBtu/hr with one pump and one heat

exchanger in service, and 22 MBtu/hr with both pumps and heat exchangers in service

(assuming 2% plugging).

Both trains of the SFPCS in operation have sufficient capacity to maintain fuel pool

temperature within specified limits with the maximum number of fuel assemblies in

the pool after a full core offload. Under these conditions, after a period of

approximately 40 days of fuel decay time following reactor shutdown, one train of

the SFPCS has sufficient capacity to maintain fuel pool temperature within specified

limits.


The SFPCS is normally in operation during all modes of facility operation.

Satisfactory operation is demonstrated continuously without the need for special

testing or inspection.

3.3.1.4 Tools and Servicing Equipment

3.3.1.4.1 Objective

To provide and use tools and servicing equipment in a way that ensures the bounds of

the design basis fuel handling accident are not exceeded.


1. The refueling platform shall withstand a seismic event without gross failure or

overturning.

2. Fuel handling equipment shall be classified in accordance with its potential

for damaging irradiated fuel.

3. Equipment weighing more than 700 pounds shall be classified as a heavy load and

handled in accordance with appropriate facility procedures.


3.3.1.4.3.1 Introduction

All tools and servicing equipment necessary are supplied for efficiency and safe

serviceability. The following is a listing of tools and servicing equipment.


Tools and Servicing Equipment

Quantity

Fuel Servicing Equipment

Fuel Preparation Machines 2 Channel Bolt Wrenches 2 Channel Handling Tool 1 Fuel Inspection Fixture 1 Channel Gauging Fixture 1 General Purpose Grapples 3 Channel Transfer Grapple 1 Fuel Pool Gates 2 Channel Handling Boom 1 Servicing Aids

Actuating Poles 3 General Area Underwater Lights 4 Local Area Underwater Lights 4 Drop Lights 4 Underwater TV Monitoring System 1 Underwater Vacuum Cleaner 1 Viewing Aids 4 Light Support Brackets 4 Jib Cranes 2 Refueling Equipment

Refueling Equipment Servicing Tools 1 Refueling Platform, main fuel grapple and contents 1

Storage Equipment

Control Curtain Transfer Basket 1 Spent Fuel Storage Racks 11 Channel Storage Rack 1 Control Rod Blade Storage Rack (30 Cavities per Rack) 1 Defective Fuel Storage Containers 8

3.3.1.4.3.2 Fuel Servicing Equipment

Two fuel preparation machines are used to remove the channels from and install

channels on fuel assemblies. These machines are designed to be removed from the

pool for servicing. A channel transfer grapple is provided for inserting or

withdrawing channels from storage racks.


An equipment support railing is provided around the pool periphery in order to tie

off miscellaneous equipment such as the fuel leak detector (sipper) and service

tools. Equipment lugs fabricated as part of the pool liner are required for

fixtures that might later be desired by facility personnel. In addition, a

4 x 4-inch curb with a 4-inch wide plate of 1-inch thick stainless steel on top is

provided around the entire periphery of the refueling volume. The plate provides a

suitable welding and drilling surface for mounting additional equipment. The curb

may be used as an additional support or tie-off area. Cable ways are recessed into

the floor around the pool periphery with openings to pass cables into the pool from

underneath this curbing.

A number of different grapples are available at Vermont Yankee for use during

maintenance activities. Grapples can be attached to the Reactor Building auxiliary

hoist, or the auxiliary hoists on the refueling platform. Grapples can be used to

shuffle fuel in the pool and to handle fuel during channeling.

A channel-handling boom with an electric hoist is used to assist the operator in

supporting the weight of a channel after the channel is removed from the fuel

assembly. The boom is set between the two fuel preparation machines. With the

channel-handling tool attached to the hoist, the channel may be conveniently moved

between the fuel preparation machines.

3.3.1.4.3.3 Servicing Aids

General area underwater lights are provided with a suitable reflector for general

downward illumination.

A portable underwater vacuum cleaner is provided to assist in removing crud and

miscellaneous objects from the pool floor. The pump and the filter unit are

completely submersible for extended periods.

3.3.1.4.3.4 Fuel Handling Equipment

The refueling platform is used as the principal means of transporting fuel

assemblies in the storage pool. The platform travels on tracks extending along each

side of fuel pool. The platform supports the main hoist and fuel grapple and two

auxiliary hoists. The grapple is suspended from a trolley system that can traverse

the width of the platform. Platform operations are controlled from either the

operator station on the trolley or auxiliary stations on the auxiliary hoist control

boxes. Refueling grapple operation and platform movement are controlled through a

Programmable Logic Controller (PLC). The PLC also limits platform velocity and

movement when the grapple is in close proximity to the perimeter of the storage

pool. A Personal Computer (PC), which is also part of the system, implements the

optional automatic mode of operation to allow a preprogrammed series of platform

movements corresponding to planned fuel assembly moves. The platform contains a

position-indicating system that indicates the position of the fuel grapple.


Mounted on both the reactor well side of the refueling platform and on the platform

trolley monorail are one-half-ton auxiliary hoists. These hoists normally can be

used with appropriate grapples to handle control rods, detectors, sources, and other

equipment. The auxiliary hoist can also serve as a means of shifting fuel elements

and other equipment within the pool.

All motions of the platform required to handle fuel assemblies may be controlled

from a single location.

3.3.1.4.3.5 Storage Equipment

A channel storage rack is located between the fuel preparation machines to permit a

logical work flow during channeling and de-channeling operations.

Racks are arranged so that fuel assemblies and control rod blades can be

conveniently positioned for storage. The racks can be removed without draining the

pool to allow inspection or replacement, should it become necessary. Capacity is

provided for a maximum of 38 control rod blades. One CRB storage rack provides a

capacity of 30 and one spent fuel storage rack provides an optional capacity of

eight.


The refueling platform can withstand a seismic event without gross failure or

overturning.

The safety classification of fuel handling equipment and tools is determined based

on their potential for damaging irradiated fuel and, as a result, exceeding

appropriate radiological dose criteria. Equipment weighing more than 700 pounds is

classified as a heavy load and is handled in accordance with Reference 4.

The design basis fuel handling accident is discussed in the Station Safety Analysis

Section of the DSAR.

3.3.1.5 References

1. NUREG-1738, “Technical Study of Spent Fuel Pool Accident Risk at Decommissioning

Nuclear Power Plants”

2. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In

Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis”

3. BVY 14-009, Request for Exemptions from Portions of 10CFR50.47 and 10CFR50,

Appendix E.

4. PP 7023, “Control of Heavy Loads Program Document”


Vermont Yankee


Fuel Storage-Arrangement

Figure 3.3.1-1

DSAR Revision 1

3.0-53 of 87

DSAR Revision 1

3.0-54 of 87


Vermont Yankee


Fuel Storage Rack Assembly

(Partial)

Figure 3.3.1-4


Vermont Yankee


HOLTEC Fuel Storage Rack Assembly

(Partial)

Figure 3.3.1-5


3.3.2 Service Water System

3.3.2.1 Objective

The objective of the Service Water System (SWS) is to provide water from the

Connecticut River for spent fuel pool cooling and other miscellaneous services.

3.3.2.2 Design Bases

The design bases of the Station SWS are:

1. To provide water for spent fuel pool cooling.

2. To minimize the probability of a release of radioactive contaminants to the

environs by monitoring the system discharge and maintaining sufficient

pressures at specific areas in the system.

3.3.2.3 Description

The flow diagram for the SWS is shown on Drawing G-191159, Shs. 1 and 2.

Two pumps located in the intake structure are provided to supply the SWS

requirements. The pumps are normally started and stopped by controls on the main

control board. With a design river water temperature of 85°F, one of the two pumps

is required to supply normal station cooling demands. Operating pumps will

continue to run until stopped from the Main Control Room.

The SWS is a dual header system using two parallel 24-inch supply headers. Two

automatic self-cleaning strainers, for removal of suspended matter from the river

water, serve the headers. The headers include cross-connect lines 20"SW-3, 3"SW-28

and 3"SW-28A, and 24"SW-8. Normally, the valves in the interconnecting lines are

open, permitting either pump to supply the cooling water through both strainers and

headers. In the event of a major malfunction in either header, it is possible to

isolate the portion of the system affected (using electrically-operated valves V70-

19A and 19B in line SW-8 and manual valves in the other cross-connect lines) and

maintain all essential cooling water services.


A pressurizing line to the Fire Water System and water for the Chlorination System

is supplied from lines tied into both headers. There is also a line, 12"SW-4A,

tying into the Fire Water System that can be supplied from either header. This

line contains a manual valve that is maintained closed and can only be opened under

specific procedural direction. The SFPCS heat exchangers and the water for the

backwash function of the traveling screens are supplied from the "B" service water

header.

The SWS discharges to the cooling tower deep basin.

A process radiation monitor is located in the station service water discharge

header. To prevent the release of radioactivity from the SFPC System to the SWS,

the system is designed such that the fuel pool side of the heat exchangers has a

maximum operating pressure equal to the static head developed by the difference in

elevation between the heat exchanger and the fuel pool surface. The minimum

operating service water pressure is normally greater than the pressure in the fuel

pool side of the heat exchanger. This positive differential pressure in the heat

exchangers will protect against any possible fuel pool water leakage into the

Station SWS.

SWS piping meets code standards, including ANSI B31.1. Those portions of the

Station SWS supplying spent fuel cooling are of Class I seismic design. Other

portions of the system, whose failure due to a seismic event could cause

unacceptable flooding, negatively impact flow to essential equipment, or impact the

ability to establish secondary containment closure, are either isolable via

automatic or manual valves or have been evaluated and determined to meet Class I

seismic requirements.

Vacuum breakers are installed in the following pipe lines to prevent water hammer

when service water flow is restored following a loss: supply line (8" SW-800) to

the standby fuel pool cooling heat exchanger, the supply line to the refuel floor

cooler lines, and the supply line (8" SW-33) to the Control Room chiller.

Operation of these valves will ensure adequate flow to essential components

following the events identified above and will prevent flooding in the Reactor

Building due to a water hammer event.

That portion of the intake structure, which houses the Station SWS equipment, is

Class I seismic design.


3.3.2.4 Evaluation

Service water piping meets code standards, including ANSI B31.1. Those portions

supplying fuel pool cooling equipment are of Class I seismic design. In addition,

the piping whose failure could (a) cause unacceptable flooding, or (b) negatively

impact flow to essential equipment, or (c) impact the ability to establish

secondary containment closure, is isolable via automatic or manual valves, has been

designed such that loss of water through failed piping is within the capability of

the system, or has been evaluated and determined to meet Class I seismic

requirements. Vacuum breakers have been installed on those pipe lines susceptible

to a water hammer event to ensure adequate flow to essential equipment and prevent

flooding in the Reactor Building. The maximum operating pressure of the service

water within these lines is typically on the order of 110 psig. This piping is

routed in the building such that it is not in the vicinity of any heavy equipment

movement during maintenance nor in the vicinity of vehicle traffic, and therefore,

is not vulnerable to damage from collisions.

Based upon the above discussion, it is highly improbable that the piping could fail

in such a manner as to cause flooding or interrupt SWS flow for spent fuel pool

cooling.

Maintaining a positive differential pressure between the service water side and the

fuel pool side of the SFPC heat exchanger protects against any possible fuel pool

water leakage into the SWS.

3.3.2.5 Inspection and Testing

The Station SWS is normally in operation during all modes of station operation.

Satisfactory operation is demonstrated continuously without the need for special

testing or inspection.

3.3.3 Electrical Power Systems

3.3.3.1 Transmission System

3.3.3.1.1 Objective

The objective of the Transmission System is to provide reliable power from off-site

to the facility Auxiliary Power System to facilitate the safe storage and handling

of irradiated fuel.


3.3.3.1.2 Design Basis

The Transmission System provides a reliable source of power from off-site to the

facility to facilitate the safe storage and handling of irradiated fuel.


There are two 345 kV switchyards and two 115 kV switchyards on site at VY. The

original 345 kV and 115 kV switchyards are now called the VY switchyards. New

Vernon 345 kV and Vernon 115 kV switchyards were installed by Vermont Electric

Power Company (VELCO) as part of their Southern Loop Project. These two

switchyards are on VY property north of the existing VY switchyards.

The VY 345 kV switchyard consists of four circuit breakers in a ring bus

configuration as shown on Drawing G-191298, Sh.3.

Electric power is supplied from off-site via the transmission network to the

on-site electric distribution system through either of two 345 kV/115 kV

autotransformers to the VY 115 kV switchyard. The VY 115 kV switchyard powers the

station startup transformers.

A portion of the 345kV system, utilized for power generation, has been abandoned.

The VY 345 kV switchyard north bus powers a 400 MVA autotransformer which supplies

power to the VY 115 kV switchyard.

The Vernon 115 kV switchyard supplies a second source of normal power to the VY 115

kV switchyard via a K-40 tie line to a second autotransformer.

The auto-transformers are operated in parallel; the loss of either source will not

cause the VY 115 kV switchyard to lose power.

An alternate circuit through the 115 kV K-186 transmission line may be made

available.

A 13.2 kV underground power line runs from the adjacent Vernon Hydroelectric

Station (VHS) to a 13.2-4.16 kV transformer near the cooling towers. From there, a

4160 V underground power line connects to the Station Blackout (SBO) Diesel

Generator (DG) switchgear and then goes on to the station switchgear. The SBO DG

or the VHS can be connected to selected 4160 V buses through manually operated

circuit breakers.


3.3.3.1.4 Evaluation of System Protection

Transmission system protection design meets the objectives of the Northeast Power

Coordinating Council, "Bulk Power System Protection Criteria."

The two tie lines between the VY switchyards and the Vernon switchyards each have

two redundant and diverse channels of the line differential, directional over-

current line and round impedance protection.

Four 345 kV and one 115 kV transmission lines are terminated in the Vernon

Switchyards. Protection in the Vernon Switchyards for the transmission lines

consists of Primary and Secondary (or backup) protection.

The 345 kV and 115 kV switchyards each have a primary and secondary bus

differential relay system. These systems are independent of one another and the

tripping of one will not cause tripping of breakers in the other substation, with

the exception that the 115 kV secondary side breaker on the 400 MVA autotransformer

will be tripped for a fault on the 345 kV switchyard's north bus.

In the unlikely event that the two sources to the 115 kV primary of the station

startup transformers, that is, the 345/115 kV autotransformer supply from the VY

345 kV switchyard, or the tie line to the Vernon 115 kV switchyard became

disconnected, the Station Blackout Diesel Generator Alternate ac source and the

line to the Vernon Hydroelectric Station, would also be available.

The 115KV switchyard contains three capacitor banks, one 30MVAr bank and two 15MVAr

banks. Each bank has its own breaker connecting it to the 115KV bus. These

breakers are individually controlled by the system operator via SCADA. Phase and

ground overcurrent, unbalance, over-voltage and breaker failure protection is

provided for each bank breaker.

3.3.3.2 Auxiliary Power System

3.3.3.2.1 Objective

The objective of the Auxiliary Power System is to provide a reliable power supply

to all station loads required for the safe storage and handling of irradiated fuel.


The Station Auxiliary Power System shall have the capacity and capability to supply

the required facility loads. Protective, control, and instrumentation devices

shall be provided to insure reliability and availability of the system.



The Station Auxiliary Power System is shown on Drawings G-191299, G-191300, Sh. 1

and 2, G-191301, Sheets 1 and 2. The system consists of six 4160 V buses, which

supply power to all 4000 V motors, and to the 4160-480 V station service

transformers, which supply power to the 480 V buses.

3.3.3.2.3.1 4160 V Switchgear

The normal supply for the 4160 V load is the startup transformers (T-3A and T-3B)

which are supplied from the 345/115 kV Transmission System.

The startup transformers have adequate capacity for all loads required for the safe

storage and handling of irradiated fuel.

The switchgear for the 4160 V Auxiliary System is of the metal-clad indoor type,

except 4160 V Buses 5A and 5B which are outdoor metal-clad units. Circuit breakers

are three pole, air break type, electrically-operated with control power supplied

from batteries.

3.3.3.2.3.2 480 V Buses

480 V auxiliary power is supplied from the 4160 V Auxiliary System through 4160-480

V station service transformers. The 480 V system consists of switchgear buses and

motor control centers.

The 480 V switchgear buses are self-supporting, metal-clad structures with draw-out

circuit breakers.

3.3.3.2.3.3 120/240 V Instrumentation Distribution System

A 120/240 V Single Phase Instrumentation Distribution System supplies selected

instrumentation and other loads. The system consists of the 120/240 V vital ac bus

and its subpanel and the 120/240 V instrumentation distribution panel and its

subpanels.

The bus arrangement is shown on Drawing G-191372, Sheets 4 and 5.


3.3.3.2.4 Cable Installation and Separation Criteria

1. Intermixing of Cables

Low-level instrumentation cables are routed in separate trays from control

cables.

The definition of "low level instrumentation cable" is:

A cable used for data, control, or instrumentation service. In

general, this service includes cable from thermocouples, resistance

temperature detectors, process instruments, and computer signals. As a

general rule, anything less than 50 V is considered low level.

The definition of "control cable" is:

A cable used for control, metering, relaying, and alarm circuits. In

general, these services include 125 V dc and 120 V ac control leads,

annunciator cables, PT cables, CT cables, and solenoid cables.

The following exception to the criteria for intermixing cables has been

justified as acceptable:

1. Instrumentation cable which only provides a Control Room indication function

may be run in the same tray as control cable.

2. Tray Loading and Cable Sizing

The general rules of 50% derating outside the drywell area was used in calculating

power cable sizes. The following design conditions were considered in arriving at

the 50% derating criteria for cables in tray: (1) load factor, (2) tray loading,

(3) short circuit capacity of cable, (4) ambient temperature, (5) grouping factor,

and (6) voltage drop.

For cable tray loading, the design is based on the IPCEA Code Bulletin

No. P-46-426, 1962.

For cables that are not routed in tray, derating is based on ampacities taken from

the appropriate tables in the IPCEA Code Bulletin No. P-46-426, 1962, which are a

function of the number of conductors in the conduit or duct bank.


3. Fire Protection Criteria

The criteria used for fire protection of cable installations are as follows:

Fire stops are provided in vertical tray runs at Reactor Building fire zone

boundary designations.

In areas other than the cable vault, tray covers are utilized on cable trays to

prevent fire from migrating from tray to tray in a vertical bank.

In the area where a high concentration of cables exists, such as the cable vault,

an automatic Fire Protection System is provided; and flame resistant cable

constructions are used to minimize the propagation of fire along horizontal runs of

cable trays. Cable tray covers are not used in the cable vault except where

required to achieve physical separation. This allows fire extinguishment by the

Fire Protection System or by manual means.


Periodic equipment tests are performed at scheduled intervals to detect any

deterioration of the system towards an unacceptable condition. The specific tests

and the frequency at which they are performed depend upon the specific components

installed, their function, and their environment.

3.3.3.3 Deleted

3.3.3.4 125 V DC System

3.3.3.4.1 Objective

The objective of the 125 V DC System is to provide a supply of 125 V dc power for

the operation of equipment.


The 125 V DC System shall consist of two dc systems, each capable of supplying its

required loads.



Two 125 V dc systems are provided to supply the station 125 V dc loads. One system

includes Main Station Battery A-1. The second system includes Main Station Battery

B-1. Each battery is of the central power station type, designed for continuous

duty at an operating voltage of 125 V dc.

Main Station Batteries A-1 and B-1 are located in the Reactor Building.

Each main station battery has a pair of constant voltage, current limiting

silicon-controlled rectifier type battery chargers, which are capable of supplying

normal continuous dc load and maintaining a float charge on the battery. Each

charger is also capable of recharging its associated battery to full charge if it

should become discharged to its minimum voltage.

The redundant distribution divisions of the 125 V dc systems are designated as the

DI and DII divisions. Division DI includes Distribution Panels DC-1, DC-1A, DC-1B,

and DC-1C connected to Main Station Battery A-1. Division DII includes

Distribution Panels DC-2, DC-2A and DC-2C connected to Main Station Battery B-1.

Distribution Panel DC-2D is powered from DII power supply DC-2.

Manual transfer switches are installed to provide a backup supply of power to

selected panels. Use of the transfer switches is administratively controlled by

procedure and limited to emergency situations or planned maintenance per

administrative control. Panel DC-3 is normally fed from Panel DC-2, but can be

connected to an alternative feed from Panel DC-1, by means of a manual transfer

switch. Other dc panels with manual transfer switches are DC-2A and DC-3A which

also have alternate feeds. The electrical arrangement of the batteries, chargers,

buses and switchgear is shown on Drawing G-191372, Sheets 1, 2 and 3.

The batteries, panels, and power feeds associated with Division DI are physically

isolated from the batteries, panels, and power feeds associated with Division DII

by a minimum of 15 feet. Where distribution circuits are separated by less than 15

feet, the cables are routed in rigid steel conduits, flexible steel conduits, or

enclosed in steel wireways.

The DC System is ungrounded and has a ground detection alarm system.



During normal operation, the continuous dc load is supplied by the battery chargers

which are connected to the dc buses. The batteries normally float on the system,

supplying any momentary high current control requirements. Upon loss of a charger,

the associated battery supplies its total dc load requirements. The normal ac

sources for the battery chargers are the 480 V emergency buses. These buses are

energized through transformers by normal auxiliary ac power.

Feeders from redundant dc sources are provided to control circuits for the 4160 V

bus switchgear (except for buses 4, 5A and 5B), and for certain 480 V emergency bus

switchgear and 125 V dc distribution panels. These alternate feeders are connected

through manual transfer switches such that only one dc source can be connected at a

time.

The Main Station Batteries A-1 and B-1 are located on elevation 318' of the Reactor

Building. Analysis assumes that the Reactor Building temperature will be

maintained at >40°F.

The accumulation of hydrogen from the batteries would not exceed 4% concentration

with an assumed complete loss of Reactor Building Ventilation.

3.3.3.4.5 Additional DC Systems

In addition to the above dc systems, two 125 V DC Systems in each of the switchyard

control houses which provide power for breaker operation and control and protective

relaying circuitry.


The batteries and other equipment associated with the 125 V DC System are easily

accessible for inspection and testing. Service and testing is performed on a

routine basis in accordance with approved station procedures/programs. Typical

periodic inspections will include visual examination for leaks and corrosion, and

check of all batteries for voltage, specific gravity of electrolyte, and

electrolyte level.

3.3.3.5 24 V DC Power System

3.3.3.5.1 Objective

The objective of the ±24 V DC Power System is to provide a supply of ±24 V DC power

for the operation of various process radiation monitoring instrumentation.



The ±24 V DC Power System shall supply all ±24 V dc power requirements.


A single ±24 V dc system is provided for operation of various process radiation

monitoring instruments. The system is a single channel system with no automatic or

manual transfer.

The ±24 V dc system consists of two 24 V dc power supplies connected to produce a

single ±24 V dc supply. The rating of each power supply is the same and is based

on the maximum load required by the process rad monitoring instrumentation.

The ±24 V dc system is fuse protected from high input or output current. Any

indication of a failure of the ±24 V dc power supply will be annunciated by the

associated rad monitors downscale/trouble alarms.


The components of this system are inspected and tested in accordance with approved

station procedures/programs.

3.3.4 Fire Protection System

3.3.4.1 Objective

This system is designed to provide fire protection for the station through the use

of water; CO2; FM-200; dry chemicals; detection and alarm systems; and rated fire

barriers, doors, and dampers.

3.3.4.2 Design Basis

The Fire Protection System shall prevent propagation of fire and isolate the areas

of the fire by:

1. Providing a reliable supply of fresh water for firefighting purposes.

2. Providing a reliable system for delivery of the water to potential fire

locations.

3. Providing automatic fire detection in those areas where the danger of fire is

more pronounced.


4. Providing fire extinguishment by fixed equipment activated automatically or

manually in those areas where danger of fire is most pronounced.

5. Providing manually operated fire extinguishing equipment for use by station

personnel at selected locations.

6. Providing means to isolate areas so that fires are prevented from propagating

from one area to another.

3.3.4.3 Description

The Vermont Yankee Fire Protection Program makes use of detection and suppression

systems, separation criteria, rated fire barriers and seals, fire stops, procedures

and fire watches, standpipe hose connections, and training.

The fire protection program for the permanently defueled state has been developed

based on the applicable requirements of 10CFR50.48 and BTP APCSB 9.5-1, Appendix A.

The Fire Hazards Analysis (FHA) documents existing plant configurations and defines

the resources available for the prevention and limitation of damage from fire

(Reference 1). In addition to plans and physical configurations for fire

protection, fire detection, fire suppression and limitation of fire damage, the FHA

also provides an overall description of the fire protection program.

The Fire Protection System is illustrated on Drawing G-191163, Sheets 1 and 2.

Water-type fire protection equipment has been limited in those areas where the

potential spread of radioactive contamination due to release of water for the

firefighting would result in more severe consequences than the results of a fire.

Fires in these areas will be primarily fought using portable dry chemical or carbon

dioxide extinguishers.

Water for the Fire Protection System is provided by two vertical turbine-type

pumps, one electric motor-driven and one diesel-driven. Each pump has a capacity

of 2,500 gpm at 125 psi discharge pressure. The pumps and drivers are located in

the intake structure. They discharge to an underground piping system which serves

the exterior and interior Fire Protection Systems.

The motor-driven pump is supplied from a 480 V bus. The diesel engine drive is

approved for fire pump service and is provided with its own fuel oil supply and

starting equipment.

The pressure in the Fire Main System is maintained at approximately 100 psig by an

interconnection to the Service Water System. An orifice in the 1.5 inch

pressurizing line limits pressure maintenance flow from the Service Water System to

30 gpm during normal operation. A check valve in the connecting pipe prevents

backflow.


Operation of the fire pumps is controlled from pressure switches in the discharge

piping. The motor-driven pump starts at a predesignated system pressure (typically

85 psig). The diesel-driven pump starts if the pressure continues to drop

(typically 75 psig). The motor-driven pump automatically shuts down when the Fire

System pressure is restored to the normal range (typically 100 psig) for

approximately seven minutes. The diesel-driven pump continues to operate until

shut down manually.

The yard piping consists of a 12-inch underground piping loop around the

entire station, with valved branches serving 10 fire hydrants. Valved branches

from the piping loop supply water for interior fire protection purposes.

Sectionalizing valves in the yard piping loop permit isolation of portions of the

loop, without interruption of service to the entire system.

A heat traced and insulated fire protection header in the Turbine Building supplies

an interior Reactor Building loop. This loop services two standpipes and fifteen

standpipe hose connections.

Selected abandoned water suppression deluge valves have retained the connected heat

actuated devices (HADs) for early indication of a fire event in areas which they

were previously installed. These valves are alarmed and signal the control room

fire alarm panel. Furthermore, the remaining turbine building water curtain,

condenser/heater bay and condensate demineralizer storage fire detection systems

were not originally installed to satisfy the requirements of 10 CFR 50.48 or Branch

Technical Position APCSB 9.5-1 Appendix A.

The cable vault and Switchgear Rooms are protected by fully automatic total

flooding CO2 suppression systems. The Cable Vault CO2 suppression system is

initiated by ionization detectors. The Switchgear Room CO2 suppression system is

initiated by ionization detectors coincident with thermal detection. Bottles

located in the West Switchgear Room System may also provide a backup or second shot

to the cable vault if desired. The Diesel Fire Pump Fuel Oil Storage Tank Room is

protected by a total flooding FM-200 suppression system initiated by an ionization

detector coincident with a thermal detector.

The yard loop supplies a wet pipe sprinkler system for the warehouse* and the

house-heating Boiler Room*. These systems are equipped with alarm check valves.

* Not required to satisfy the requirements of 10CFR50.48 or BTP APCSB9.5-1, Appendix A.


Fire detection devices are provided in areas which are not normally occupied, in

areas where substantial quantities of combustible materials are present, or in

other areas determined to be highly sensitive. These detection systems provide

local and remote alarms, as well as annunciation in the Main Control Room. In some

instances trip signals are provided directly to deluge systems or electrically

operated fire dampers.

Portable fire extinguishers are located throughout the buildings at the site.

Portable fire extinguishers use dry chemical, CO2 and water.

Buildings are constructed of steel and concrete with fire walls and/or shield walls

which isolate separate areas. Consideration has been given to the use of

noncombustible and fire-resistant materials throughout the facility, particularly

in the containment, Control Room, and areas containing critical portions of the

plant.

Fire barriers have been identified and their integrity assured by self-closing

doors (exception: RHR corner room doors at El. 213'-6" are not self-closing),

normally locked doors, alarmed doors, doors checked daily, automatic fire dampers,

and controlled procedures for penetration sealing and fire barrier repair. This

includes the northwest stairwell's ability to function as a fire exit.

Water flow alarms are provided in critical locations and annunciate in the Control

Room to provide positive indication of Fire Water System operation.


The fire pumps, water suppression systems, CO2 systems, FM-200 system, fire

barriers, fire doors, fire dampers, detection and alarm systems, and portable

extinguishers are inspected and tested periodically in accordance with approved

station procedures/programs. All equipment is accessible for periodic inspection.

3.3.4.5 References

1. Vermont Yankee Nuclear Power Station Fire Hazards Analysis


3.3.5 Heating, Ventilating and Air Conditioning Systems

3.3.5.1 Objective

The objective of the Heating, Ventilating, and Air Conditioning Systems is to

provide suitable environmental conditions for facility personnel and equipment.


The design bases of the Heating, Ventilating, and Air Conditioning Systems are as

follows:

1. Provide appropriate temperature and humidity conditions for personnel and

equipment.

2. Limit exposure of personnel to airborne contaminants by controlled migration

of air from radioactively clean areas to areas of progressively higher

contamination.

3. Normally, filter outside air to limit the introduction of particulate matter

to the plant. During winter operation, certain filter media may be removed to

prevent freezing.

4. Vent potentially contaminated leakage through systems that exhaust to the

plant stack.


3.3.5.3 Description

Flow diagrams for the Heating and Ventilation Systems are shown on Drawings G-

191237, Sheet 1 and 2, G-191236, G-19138 and G-191254.

The design temperatures used for the Heating and Ventilation Systems are provided

as follows:

Outdoor

Summer: 90°F dry bulb, 75°F wet bulb

Winter: -12°F dry bulb

Indoor

Reactor Building:

Maximum: 100°F (occupied areas)

Minimum: 65°F (refuel floor)

55°F (occupied areas other than refuel floor)

Control Room and Service Building:

Maximum: 78°F dry bulb, 50% relative humidity

Minimum: 72°F dry bulb

3.3.5.3.1 Reactor Building

The Reactor Building normal Heating, Ventilating, and Air Conditioning System

limits exposure of personnel to airborne contaminants and maintains appropriate

temperature conditions for personnel and equipment.

The Reactor Building normal HVAC System migrates air from clean accessible areas to

areas of progressively higher contamination or potential contamination, removes the

normal heat losses from all equipment and piping in the Reactor Building, limiting

the temperatures to approximately 100ºF, filters outside air to limit the

introduction of airborne particulate matter to the station, and exhausts

potentially contaminated air to the stack. During winter operation, certain filter

media may be removed to prevent freezing.


The Reactor Building normal HVAC System consists of a supply and exhaust side. See

Drawing G-191238.

The supply side includes in the direction of air flow, outside louvers, automatic

dampers, automatic roll-type filters, steam heating coils, and two double-width

centrifugal fans each sized for the full system capacity of 53,800 cfm. This

capacity provides approximately 1.5 net Reactor Building air changes per hour.

The exhaust side consists of two paralleled single-width centrifugal fans, each

having full system capacity of 55,800 cfm.

The excess of exhaust fan capacity over the supply fan capacity ensures against

building out leakages during normal operation. The main supply and exhaust ducts

penetrate the Reactor Building, each through two butterfly isolating valves in

series. The valves in the main supply duct are powered from different buses. This

is also true of the valves in the main exhaust duct. All four isolating valves

fail closed.

To permit maintenance of one fan while the other is in service without danger of

contamination, an isolation damper is provided at the inlet. Also, an isolation

outlet damper is provided to minimize the possibility of contamination through the

idle fan due to either stack backflow or recirculation from the active fan.

In addition, gravity dampers, i.e., non-return or backdraft dampers, are provided

to prevent reverse flow at all ventilating supply openings for areas having

contamination potential and in all branch exhaust ducts connecting with main ducts

which carry exhaust from areas having contamination potential.

Failure of a gravity damper to operate in a branch exhaust line will not result in

cross contamination. Each branch exhaust line consists of two 100% capacity

exhaust fans, a gravity damper on the discharge side of each fan, and a third

gravity damper in the branch line just prior to entering the main exhaust duct.

With the above arrangement, no backflow will occur through the branch exhaust lines

even in the event a gravity damper fails open and both exhaust fans are

inoperative.

Failure of a gravity damper to operate in a supply line could result in some

cross-contamination only if both redundant branch exhaust transfer fans are

inoperative. This is extremely unlikely.

Axial booster fans, each supported by an automatically cut-in standby unit, are

provided throughout the exhaust system to overcome air circuit losses.


In general, duct work is of galvanized steel. Duct work under positive pressure

exhausting to the main stack is of welded construction to minimize outleakage.

A purge exhaust fan permits exhausting the drywell or the suppression chamber. The

upstream end of the purge exhaust fan is connected to the Primary Containment

Atmospheric Control System through butterfly valves which are remotely actuated

from the Main Control Room panel.

The downstream end of the purge exhaust fan discharges into the Reactor Building

normal Exhaust System where exhaust fans direct the purged air to the main stack.

All equipment and components are accessible for inspection, adjustment, and

testing. The only moving parts in a backdraft (gravity) damper are pinned joints

and bearings (dry, oil-impregnated porous metal, Teflon, or Zytel).

Proper sequences of operation, as well as correct control point adjustments were

determined during station pre-operational tests to assure conformity to the

requirements and intent of the specifications and drawings.

3.3.5.3.2 Deleted

3.3.5.3.3 Main Control Room

The system serving the Main Control Room is designed to provide summer air

conditioning and heating during the winter.

The Supply System has a 12,500 cfm capacity and includes, in the direction of flow,

a wall louver, automatic outside air damper, filters, chilled water cooling coil,

steam heating coil, centrifugal fan section, a system of duct work, and air

outlets.

The Supply System chilled water coil is serviced by a double circuit refrigeration

plant to assure continuity of cooling. Refrigeration plant components are one

double circuit water chiller with a chilled water pump, two air-cooled condensers,

piping, and controls. A separate air-cooled chiller unit was installed to provide

equivalent or better primary, or backup, cooling for the Control Room.

A remote manual switch located in the Main Control Room permits closure of the

outside air damper, Control Room kitchen and bathroom exhaust dampers, and Computer

Room supply damper, in order to isolate the Control Room, if required.


SAC-1, which supplies the Control Room, contains a humidifier in the air supply

duct after the Computer Room duct. This unit is controlled by a humidity sensor in

the Control Room and has an alarm for high humidity level.

Upon a loss of the Control Room Ventilation System the SAC 1A/B dampers could fail

to the closed position. Operator actions, including manual control of appropriate

dampers, can be taken to restore system flow as discussed in plant procedures

The Control Room can be isolated by manually closing the fresh air inlet branch

damper, cable vault damper, and Control Room vent paths. This also puts the

Control Room ventilation in the recirculation mode of operation.

3.3.5.3.4 Service Building

The original portion of the Service Building is entirely air conditioned by an air

handling unit having 15,000 cfm capacity and which in the direction of flow,

consists of dampers, mixing box, filters, a chilled water cooling coil, and a fan

section. Electric zone reheat coils compensate for cooling load variations in

different areas.

The chilled water coil is served by a service water-cooled package chiller and

chilled water pump.

Air from spaces having potential contamination, such as the chemistry laboratory,

is not recirculated back to the air handling unit -- it is directly exhausted to

the plant stack by one of two full capacity fans.

The added portion of the Service Building on the north side is cooled and

ventilated by packaged units on the roof. Makeup and exhaust is local at each

unit. Heat is from the house boilers.

3.3.5.3.5 Deleted

3.3.5.3.6 Heating Boiler System

Drawing G-191254 shows the process flow diagram for the Station Heating Boiler

System. The Heating Boiler System is designed to provide a source of steam for

space heating and process requirements.

Each of the two boilers is rated for approximately 50% of the calculated heating

load.

All pressure containing parts have been designed to the ASME Boiler and Pressure

Vessel Code, Section I.


The boiler plant consists of two forced draft, four pass, 50% capacity, No. 2

oil-fired fire tube boilers, supplemented by a condensate return tank and three

cross-connected 50% capacity feedwater pumps. Each boiler is equipped with a

locally mounted control panel. Indicating lights will show "flame failure," "low

water," "fuel valve open," and "load demand."

Also provided is a Fuel Oil Pumping System, blowdown tank, and remotely located

condensate pump, receiver sets, Sampling System, and Chemical Addition Systems.

The Station Heating Boiler System is located in the south end of the Turbine

Building. The heating boiler feedwater is monitored by a process liquid radiation

monitor (see Drawing G-191254).

The unit is equipped with the necessary controls and safety devices to operate

automatically. The combustion safety control is provided with a safety lockout in

the event of flame failure or failure to start, which requires manual reset before

the automatic cycling can continue. The draft fan controls are interlocked with

the burner controls to prevent operation of the burner under improper draft

conditions. The flame requirements are designed to meet FIA and NEPIA

requirements.

3.3.5.3.7 Deleted


All equipment and components are accessible for inspection, adjustment, and

testing.

Absolute particulate filters will be factory tested with 0.3 micron monodisperse

thermally generated dioctyl phthalate (DOP) aerosol. Minimum acceptable efficiency

is 99.97% as measured by a light-scattering photometer.

To assure that gaskets and seals are properly installed and that no damage has

occurred to the filter during shipment or handling, in-place tests using a

polydisperse cold generated aerosol will be performed initially and then

periodically as required.

3.3.6 Instrument and Service Air Systems

3.3.6.1 Objective

The objective of the Instrument and Service Air Systems is to provide the station

with the compressed air requirements for pneumatic instruments and controls and

general station services.



The Instrument and Service Air Systems shall provide the plant with a continuous

supply of oil-free compressed air. Dry air shall be supplied to plant instruments

and controls as required. Undried air shall be provided for various station

services.

3.3.6.3 Description

The Compressed Air Systems are shown on Drawing G-191160, Shs.1 through 8. The

systems include nonlubricated air compressors connected in parallel, each with a

built-in intake filter-silencer, after-cooler, and moisture separator. The

compressors discharge to two vertically mounted air receivers.

Each compressor will function in either the lead or lag mode. Normally,

compressors which are selected to the Lead position will maintain pressure between

100 and 105 psi. The compressors which are in the lag position will start when

header pressure drops to a predetermined value below the normal operating range.

If the backup compressors run unloaded for a preset period of time, they will

automatically shut down and remain shutdown unless header pressure drops to the

predetermined value.

Separate piping is provided at the discharge of the air receivers for the

Instrument Air System and the Service Air System. The compressed air of the

Instrument Air System passes through two parallel branches both of which contain

the following equipment:

1. Prefilter - This unit filters the air to remove moisture droplets, particles

of dirt, rust, and scale of approximately 3 microns and larger through the use

of automatic traps.

2. Dryer - This unit regenerates (dries out) its desiccant by a heater-less

pressure-swing process. Each dryer is sized to provide 450 SCFM.

3. After Filter - This unit, like the prefilter, filters the air to remove

moisture droplets, particles of dirt, rust, and scale of approximately

3 microns and larger.

The Service and Instrument Air Systems meet all appropriate seismic criteria.

All original piping is designed in accordance with USAS B31.1, 1967. The air

compressor discharge piping to Valve V72-2B is designed to USAS B31.1, 1977.


The Service and Instrument Air Systems are designed to operate at a pressure of 100

psig and supply 322 scfm ±5% compressed air with one compressor operating.


The Instrument and Service Air Systems are normally in continuous operation.

Satisfactory performance of these systems is demonstrated continuously without the

need for any special inspection or testing.

3.3.7 Process Sampling

3.3.7.1 Objective

The process sampling systems provide representative samples for analysis.


The sampling systems shall be designed to ensure accuracy and sensitivity of

measurement of process fluids.

3.3.7.3 Description

3.3.7.3.1 General

For flow diagrams of the station liquid sampling system, refer to Drawings G-191164

and G-191165.

Fluids and gases are sampled continuously or periodically from selected equipment

or systems. Samples are taken either as grab samples or continuously. Grab

samples are taken from the collection area to the laboratory for analysis. The

continuous samples pass through analyzers and the results are recorded.

The following table lists the description, location, and purpose of the various

monitoring points associated with sampling process fluids as appropriate.


3.3.7.3.2 Radwaste Building Sample Panel

This panel includes conductivity elements, conductivity indicating transmitters,

and individual grab sample connections with the outlets enclosed in a hooded sink

provided with exhaust ventilation.

3.3.7.3.3 Gas Sampling and Monitoring

A list of gas samples, their locations, and purpose is provided below.

Description Location Purpose

Stack sample Stack Particulate and gaseous activity

Ventilation gases a) Reactor Building b) Radwaste Building

Fan discharge Fan discharge

Activity release Activity release

The capability exists to sample the ventilation gases, but these locations are not routinely sampled. The fan discharges from the Reactor Building and the Radwaste Building are routed to the stack which is sampled continuously. 3.3.8 Deleted

Description Location Purpose

Waste disposal a) Waste surge tank b) Waste collection tank c) Floor drain collection

tank d) Chemical waste tank e) Waste sample tank f) Floor drain sample tank g) Fuel pool filter

demineralizer influent h) Fuel pool filter

demineralizer effluent i) Floor drain filter

effluent j) Waste filter

demineralizer k) Waste demineralizer

Outlet pipe Pump discharge Pump discharge Pump discharge Pump discharge Pump discharge Inlet pipe Outlet pipe Outlet pipe Outlet pipe Outlet pipe

Process data Process data Process data Process data Discharge suitability Discharge suitability Fuel pool quality Filter demineralizer efficiency Filter efficiency Filter demineralizer efficiency Demineralizer efficiency

Makeup a) Condensate storage tank

Pump discharge

Water quality


3.3.9 Lighting Systems

3.3.9.1 Objective

The lighting system provides adequate lighting in all areas of the facility where

lighting is required.

During loss of power, the lighting in all areas essential to the safe storage of

irradiated fuel is provided by the following lighting systems:

1) The emergency lighting system - dc supplied.

2) Local emergency lights with self-contained rechargeable batteries.


The lighting systems shall provide adequate lighting for the safe storage of

irradiated fuel under all conditions.

3.3.9.3 Description

The safe storage and handling of irradiated fuel requires that adequate lighting be

available for the operation, control, and maintenance of equipment.

Buses which supply power for lighting are normally powered from the startup

transformers. Upon a loss of normal power, the buses are manually powered from a

standby ac power source, thereby restoring normal lighting.

Critical areas and access routes are illuminated by DC lighting during the power

transition.

Portable, battery-operated lighting fixtures are available to permit maintenance of

standby power sources. They are located at the standby ac supply equipment. These

units are wall-mounted and are connected to normal ac circuits which keep the

self-contained batteries in a fully charged condition. Upon loss of ac, the lights

automatically become lighted, using the self-contained batteries as a source.

In the control room, battery-operated emergency lights are available. These units

are wall mounted with remote lights mounted in the ceiling panels and are connected

to a normal emergency ac circuit. Upon the loss of the normal emergency circuits,

the lights automatically become lighted using self-contained batteries as a source.


The level of lighting intensity is consistent with the nature of the work likely to

be performed for maintenance and operation upon loss of normal ac power. Under

conditions where standby sources of power are limited, a minimum level of lighting

intensity is maintained in the egress routes from the main control room and other

areas, as necessary.


Normal lighting is used continuously. Emergency lighting systems are periodically

tested to check operation of standby sources. Portable lighting sets are tested

periodically to demonstrate their functional performance.

3.3.10 Communication Systems

3.3.10.1 Objective

The objective of the communications system is to provide a reliable, convenient,

and audible communication system which meets the requirements of facility operation

and maintenance.


The communications system shall consist of an intra-site operation and public

address system, a sound-powered telephone system, a dial telephone system, an

inter-site communication system, and an off-site radio communication system. To

ensure continued intra-site and off-site communications, power for these systems

shall be provided from the vital ac bus.


3.3.10.3 Description

The communications system consists of several types of communication media

including intra-site operation and public address communication, sound-powered

telephone communication, dial telephone communication, inter-site microwave

communication, and off-site radio communication.

The function of the communications system is to permit convenient and dependable

communications between all areas of the facility vital to operation and maintenance

and protection of personnel. The systems are as follows:

1. Intra-Site Operation and Public Address System

This system consists of speakers and microphones located throughout the

facility.

The system has four transistorized channels and provides separate and

independent page and party line channels. The page channel may be used to call

personnel over the speakers as well as issue facility-wide instructions. The

party line channels may be used to carry on inter-communication after the page

call is completed, thereby making the page channel available to others.

Simultaneous conversations can take place, one on each of the channels, without

interference. The system has an output adequate to be clearly audible in all

appropriate facility areas.

2. Sound-Powered Telephone System

This system allows private communications between specific areas and pieces of

equipment for maintenance purposes of either a routine or non-routine nature.

Two independent channels are provided at each location, and the system can be

used as a back-up communication system.

3. Dial Telephone System

A dial telephone system is provided for normal communications between offices

and work areas which are routinely occupied. Units are also located

conveniently within the reactor building for use by the facility personnel.

This system is connected to points outside the facility through both the

commercial telephone system and a microwave network.


4. Microwave Communication System

The microwave system provides for the interchange of information between the

facility and the electrical dispatcher. Microwave equipment is located in the

switchyard control house and, with its battery and charger, is independent of

the other plant communications systems.

5. Off-Site Radio Communication System

Radios located in the control room provide contact with the State Police, the

Utility Emergency Radio Network, and Mutual Aid.


Operational tests are frequently made as a result of constant use of the

communications systems.

3.3.11 Process Computer System

3.3.11.1 Objectives

The objectives of the process computer are to aid facility personnel by

continuously assessing the readout of instrumentation relative to permissible

limits, to provide data accumulation and logging functions, and to serve as a

Safety Parameter Display System (SPDS) which shall provide a display of selected

variables to aid facility personnel in determining the status of the plant


The Process Computer System (PCS) shall support the following SPDS functions:

Perform meaningful conversions and monitoring of radiation release

paths;

Monitor, calculate, and display meteorological data for emergency

response personnel.

Provide Data to the Plant Data System for emergency response.



3.3.11.3.1 Computer System Components

3.3.11.3.1.1 Central Processor

The central processor performs various calculations, makes necessary

interpretations, and provides for general input/output (I/O) device control and

buffered transmission between I/O devices and memory. To ensure data integrity,

the computer system has built-in testing checks and diagnostic facilities, such as

parity and error detection and correction in the processor, memories, and the

system bus, and automatic self-test at power-up. Real-time processing capability is

provided with battery backup to facilitate a rapid restart without loss of memory

or loss of processor clock time.

Power for the computer is supplied from an uninterruptible power source (UPS-2A)

which can supply power for a minimum of three hours while off-site power is not

available.

3.3.11.3.1.2 Auxiliary Memory Subsystem

Auxiliary memory consists of fixed disk drives.

The Auxiliary Memory Subsystem is designed for and provides the capability for

further expansion. The disk drives incorporate outstanding data reliability

characteristics, including Error Correction Code (ECC), microprocessor-controlled

diagnostics, and a modular design for easy maintenance.

3.3.11.3.1.3 Peripheral Input/Output Subsystem

The peripheral I/O equipment used to read programming data into and out of the

computer consists of a system console, terminals, printers, alarm typer, magnetic

tape and disk subsystems. The system console, magnetic tape and disk subsystems

are located in the Computer Room. The terminals, printers and alarm typer are

located in the Control Room.


3.3.11.3.1.4 Process Input/Output Subsystem

The process I/O hardware is a real time distributed, microprocessor based,

intelligent industrial I/O processor. The I/O processor with its processing

capability reduces the host computer signal linearization and raw data scaling

tasks. The process I/O hardware consists of analog/digital input cards,

pulse/sequence of events input cards, and analog/digital output cards all under

microprocessor control. The analog inputs accept analog signals from plant

instrumentation and provide signal conditioning for use in the computer system.

The digital input cards provide signal conditioning and filtering. Intermittent

signals and pulse-type inputs are handled by SOE/pulse input cards. These cards

have a programmable mode of operation, including interrupt on a specific count and

continuous count. This allows immediate response for processing of information

which otherwise might be lost if digital scanning techniques were used. The

process I/O hardware supports one second scan rates for digital inputs and sub-

second scan rates for analog inputs.

3.3.11.3.4 Monitor Alarm and Logging Functions

3.3.11.3.4.1 Analog Monitor and Alarm

The processor is capable of checking each analog input variable against three types

of limits for alarming purposes: (a) process alarm limits as determined by the

computer during computation or as preprogrammed at some fixed value by the user,

(b) a reasonableness limit of the analog -input signal level as determined and

programmed by the user, and (c) a rate of change alarm limit as determined and

programmed by the user.

The alarming sequence consists of a one-line message on the alarm typer for each

point exceeding process alarm limits. Alarm messages may also be displayed on a

video display as selected by the user. A variable that is returning to normal is

signified by a one-line message on the alarm typer. Actuation of the alarm typer

provides facility personnel with an audible cue that an alarm message has printed.

The processor provides the capability to alarm the Main Control Room Annunciator

System in the event of abnormal PCS operation. Abnormal conditions for alarm

include loss of power and stall conditions. Stall conditions can be caused by

software failure, hardware failure or PCS over-temperature conditions.


3.3.11.3.4.2 Digital Inputs Status Monitoring

1. Digital Input Status Logging

The status alarm function scans digital inputs at regular intervals and

provides a printed record of system alarms. The record includes point

description, state, and time of occurrence.

2. Sequence Annunciator Detection and Logging

Digital inputs associated with status changes of major plant equipment and

instrumentation are terminated on change-of-state detection sequence of

events input hardware.

To aid facility personnel in analysis of any plant event, this function

archives SOE messages into multiple files. The SOE messages will be logged

in the order of their detection, with one millisecond (msec) resolution

accuracy. The time logs will be synchronized to the PCS (ERFIS) internal

clock at the time of occurrence.

3.3.11.3.4.3 Alarm Logging

The alarm logs required by the associated process programs are typed by the alarm

typer. Alarm printouts, as well as alarm summary displays, are used to inform

facility personnel of computer system malfunction; system operation exceeding

acceptable limits; and potentially unreasonable, off-normal, or failed input

sensors.


The Process Computer System is self-checking. It performs diagnostic checks to

determine the operability of certain portions of the system hardware, and it

performs internal programming checks to verify that input signals and selected

program computations are either within specific limits or within reasonable bounds.

3.3.11.5 Cyber Security

The PCS is deterministically isolated from less secure digital components and

systems by a data diode. This device can transmit PCS data to the less secure

general user community, but there is no physical channel for data flow in the

reverse direction. This prevents malicious computer code from migrating to the

PCS.


3.3.11.6 Process Computer Data Feed to the Plant Data Server (PDS)

A datalink sends all data variables from the PPC, through the Data Diode, to PDS.

3.3.12 Torus-as-CST System

3.3.12.1 Objective

The objective of the Torus-as-CST System is to recirculate water in the Torus and

provide for Spent Fuel Pool water makeup and letdown.


The Torus-as-CST System utilizes the Torus for water storage. The system

recirculates water from the Torus and processes it through filters and

demineralizers. The system also provides for Spent Fuel Pool water makeup and

letdown.


The chemical waste sump and sumps (equipment and floor) in the Radwaste Building

and Reactor Building are routed to the Torus. The Torus-as-CST water treatment

system, installed after permanent plant shutdown, recirculates and cleans Torus

water. The system provides for suitable Spent Fuel Pool water makeup and letdown.

The System contains parallel paths of pumps, filters and demineralizers. Suction

is taken from the Torus and discharge is either to the SFP or back to the Torus.

Water stored in the torus may be disposed offsite or discharged to the environs in

accordance with applicable permits and regulatory approvals.


RADIOACTIVE WASTE MANAGEMENT

TABLE OF CONTENTS

Section Title Page

4.1 SOURCE TERMS .......................................................... 5

4.2 RADIATION SHIELDING ................................................... 5

4.2.1 Objective .................................................... 5

4.2.2 Design Basis ................................................. 5

4.2.3 Description .................................................. 6

4.2.3.1 Materials Description ........................... 6

4.2.3.2 Reactor Building ................................ 6

4.2.3.3 Main Control Room ............................... 6

4.2.4 Surveillance and Testing ..................................... 6

4.3 HEALTH PHYSICS INSTRUMENTATION ........................................ 7

4.3.1 Objective .................................................... 7

4.3.2 Description .................................................. 7

4.4 RADIATION PROTECTION .................................................. 9

4.4.1 Health Physics ............................................... 9

4.4.1.1 Personnel Monitoring Systems .................... 9

4.4.1.2 Personnel Protective Equipment .................. 9

4.4.1.3 Change Area and Shower Facilities ............... 9

4.4.1.4 Access Control ................................. 10

4.4.1.5 Laboratory Facilities .......................... 10

4.4.1.6 Bioassay Program ............................... 10

4.4.2 Radioactive Materials Safety Program ........................ 10

4.4.2.1 Facilities and Equipment ....................... 11

4.4.2.2 Personnel and Procedures ....................... 11

4.4.2.3 Required Materials ............................. 12

4.5 LIQUID WASTE MANAGEMENT SYSTEMS ...................................... 12


4.5.1 Equipment and Floor Drainage Systems ........................ 12

4.5.1.1 Objective ...................................... 12

4.5.1.2 Design Basis ................................... 12

4.5.1.3 Description .................................... 13


4.5.2 Liquid Radwaste System ...................................... 19

4.5.2.1 Objective ...................................... 19

4.5.2.2 Design Bases ................................... 19

4.5.2.3 Description .................................... 20

4.5.2.4 Evaluation ..................................... 24


4.6 SOLID WASTE MANAGEMENT ............................................... 35

4.6.1 Solid Radwaste System ....................................... 35

4.6.1.1 Objective ...................................... 35

4.6.1.2 Design Basis ................................... 35

4.6.1.3 Description .................................... 35


4.7 EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING ........................ 39

4.7.1 Process Radiation Monitoring Instrumentation ................ 39

4.7.1.1 Plant Stack Radiation Monitoring System ........ 39

4.7.1.2 Process Liquid Radiation Monitoring System ......................................... 41

4.7.1.3 Reactor Building Ventilation Radiation Monitoring System .............................. 42

4.7.2 Area Radiation Monitoring System ............................ 47

4.7.2.1 Objectives ..................................... 47

4.7.2.2 Design Basis ................................... 47

4.7.2.3 Description .................................... 47



RADIOACTIVE WASTE MANAGEMENT

LIST OF TABLES Table No. Title 4.5.2.1 Vermont Yankee Radioactive Liquid Waste Processing Parameters

4.5.2.2 Vermont Yankee Liquid Radwaste System Tank Capacities

4.5.2.3 Vermont Yankee Liquid Effluents

4.5.2.4 Activity Input to Liquid Radwaste System (Ci/yr)

4.5.2.5 Radionuclide Discharge Concentrations

4.6.1.1 Solid Radwaste Annual Disposal History

4.7.1.1 Process Radiation Monitoring Systems Characteristics

4.7.1.2 Process Radiation Monitoring System Environmental and Power Supply Design Conditions

4.7.1.3 Plant Stack Radiation Monitoring System Characteristics

4.7.2.1 Area Radiation Monitoring System Environmental and Power Supply Design Conditions

4.7.2.2 Locations of Area Radiation Monitors

4.7.2.3 Reactor Building Area Airborne Radiation Monitoring System


RADIOACTIVE WASTE MANAGEMENT LIST OF FIGURES Figure No. Title 4.5.2-8 Radwaste Area - Plan View

4.7.2-2 Reactor Building Area Airborne Radiation Monitoring System


4.1 SOURCE TERMS

In the permanently defueled condition VYNPS will no longer produce fission,

corrosion, or activation products from operation. The radioactive inventory

that remains is primarily attributable to activated reactor components and

structural materials and residual radioactivity. The accumulation of small

amounts of solid waste may easily be controlled. Any future planned liquid

effluent releases will be evaluated prior to release, and appropriate controls

will be established. The Offsite Dose Calculation Manual ensures that VYNPS

complies with 10 CFR 50, Appendix I.

4.2 RADIATION SHIELDING

4.2.1 Objective

Radiation shielding is utilized as appropriate to limit radiation damage to

equipment and associated structures and minimize exposure of station personnel

to radiation.

4.2.2 Design Basis

Radiation shielding was provided to restrict radiation emanating from various

sources throughout the plant. Since VYNPS is permanently defueled, many

installed components are no longer required to safely store irradiated fuel.

However, many of these components continue to contain radioactive material or

remain radioactive. Shielding that was originally designed to shield these

components while they supported reactor operation continues to provide

shielding from residual radioactivity in the permanently shut down condition.

Shielding is provided to maintain personnel exposures below the limits

specified in 10CFR20. Compliance with these regulations is achieved through

shielding design based upon generalized occupancy requirements in various areas

of the station, and upon administrative radiological protection procedures.

Continuous occupancy areas outside the controlled access area, designated

Zone I, are designed to a radiation level of 0.5 mrem/hr, while those inside

the controlled access area, designated Zone II, are designed to a level of

1 mrem/hr.

Within the controlled access boundary are areas, designated Zone III, which

will allow up to 10 hours per week occupancy and are designed for 6 mrem/hr.

Controlled areas that are designed for 100 mrem/hr allowing occupancy up to

5 hours per week are designated Zone IV.


Section 6.5 of the Vermont Yankee Permanently Defueled Technical Specifications

describes the radiation protection controls for all radiation areas with dose

rates exceeding 100 mrem/hr.

Select areas will be equipped with local area monitoring devices. The Area

Radiation Monitoring System detects, measures, and records the general

radiation levels in areas where personnel may be required to work. The system

will actuate alarms if radiation exceeds preset levels.

4.2.3 Description

4.2.3.1 Materials Description

The shielding materials used are primarily concrete, water, and steel. High

density concrete, lead, and neutron-absorbing materials are used as

alternatives in special applications.

4.2.3.2 Reactor Building

The design dose rate in most areas outside the drywell in the Reactor Building

is 1 mrem/hr. The drywell and its internal structure are shielded so that most

areas outside it are accessible.

4.2.3.3 Main Control Room

The shielding of the Main Control Room consists of poured-in-place reinforced

concrete. Side walls and roof are 2 feet thick and 1 foot, 8 inches thick,

respectively.

The Main Control Room is shielded so that no individual exposure will exceed

the limits set forth in Criterion 19, Appendix A of 10CFR Part 50.

4.2.4 Surveillance and Testing

Appropriate surveillances will be conducted by trained facility personnel.

These surveys provide continuing assurance that changes which might occur and

produce significantly different radiation fields are located and appropriately

posted.


4.3 HEALTH PHYSICS INSTRUMENTATION

4.3.1 Objective

The health physics instrumentation system is a supplemental system which

provides a flexible radiation detection capability throughout the facility. It

is intended to supplement the facility process and area radiation monitoring

systems in assuring that the facility is within design limits and to supply the

required radiation control information.

4.3.2 Description

Health physics instrumentation consists of both portable and fixed equipment.

Portable Instrumentation

Portable health physics instrumentation consists of the following types of

equipment:

1. Alpha survey meters, which contain a thin "window" and an alpha sensitive

detecting element that permits the location and measuring of low levels of

alpha radiation contamination.

2. Beta-Gamma survey meters, which contain a thin windowed Geiger-Mueller tube

or ionization chamber, and are used for detecting low levels of surface

contamination or for making direct radiation surveys.

3. Neutron survey meters, which contain a thermal neutron sensitive BF3 tube

or tissue equivalent proportional counter. These meters are used for

locating possible shielding voids, streaming paths, etc., in the reactor

building.

4. Beta-Gamma and neutron dose rate meters are used for determining stay times

for radiation workers and for posting radiation area warning signs.

High range beta-gamma meters provide dose rate information during any event

involving high levels of radiation. Neutron dose rate meters respond to

and provide an indication of the entire spectrum of neutrons encountered

around a nuclear reactor.


5. Emergency Kits - Emergency kits are used by the on-site mobile team during

any event involving a possible release of radioactive materials. Each kit

contains a beta-gamma survey meter and an air particulate sampler, plus any

other equipment normally used by a particular survey team.

6. Air particulate samplers, which are air pumps which pull a known flow rate

of air through filters for the purpose of sampling the atmosphere for

radioactive particulates and radioiodines. These samplers are mobile and

may be used at most parts of the plant.

7. Approved dosimeters are used in evaluating the exposure to personnel

working at the site.

Fixed and Laboratory Instrumentation

In addition to the portable health physics instrumentation available, there are

a number of fixed and laboratory instruments which are used to assess or

control the spread of radioactivity throughout the facility.

1. Gamma or beta sensitive portal monitors are located in the guardhouse and

several entrances to the controlled areas and monitor all outgoing

personnel for radioactive contamination.

2. Personal friskers are located at key places within the facility, and are

used by facility personnel to detect surface contamination on clothing,

skin, etc.

3. Dosimeter readers, which contain the equipment for measuring the dose

received by personal dosimeters. These instruments are located in an

off-site dosimeter processing facility under contract with Vermont Yankee.

4. Multi-channel gamma spectrometer, which consists of a NaI, GeLi, or HpGe

crystal, and analyzer circuits necessary for the identification of

individual isotopes by gamma ray energy.

5. Laboratory alpha and beta-gamma counters, which are used for measuring low

levels of radioactivity in specially prepared samples such as smears, air

particulate sample filters, etc.


6. Body-burden counters, which are used to assess internal contamination from

both natural sources and from inhaled/absorbed radioactive gases or

particulates.

4.4 RADIATION PROTECTION

4.4.1 Health Physics

All employees of Vermont Yankee are given training in radiological safety and

in the requirements for working in the plant.

Administrative controls are established to assure that all procedures and

requirements relating to radiation protection are followed by all station

personnel. These procedures include a radiation work permit system. All work

on systems or in locations where exposure to radiation or radioactive materials

is expected to approach prescribed limits, requires an appropriate radiation

work permit before work can begin. The radiological hazards associated with

the job are determined and evaluated prior to issuing the permit.

4.4.1.1 Personnel Monitoring Systems

Personnel monitoring equipment is assigned to Vermont Yankee personnel by the

Radiation Protection Department. Personnel monitoring equipment is also

available on a day-to-day basis for visitors not assigned to the station that

enter radiation control areas. Records of radiation exposure history and

current occupational exposure are maintained by the Radiation Protection

Department for each individual issued personnel monitoring equipment.

4.4.1.2 Personnel Protective Equipment

Special protective clothing and respiratory equipment are furnished and worn as

necessary to protect personnel from radioactive contamination.

4.4.1.3 Change Area and Shower Facilities

A change area is provided where personnel may obtain clean protective clothing

required for station work. Temporary change areas are provided when required.

Decontamination shower facilities are maintained on-site to assist in timely

personnel decontamination. Monitoring equipment is used to assess the

effectiveness of personnel decontamination efforts.


4.4.1.4 Access Control

To prevent inadvertent access to high radiation areas, warning signs, audible

and visual indicators, barricades and locked doors are used as necessary.

Procedures are also written to control access to high radiation areas.

4.4.1.5 Laboratory Facilities

The facility includes a laboratory with adequate facilities and equipment for

detecting, analyzing, and measuring radioactivity and for evaluating any

radiological problem that may be anticipated. Counting equipment, such as a

multichannel analyzer, liquid scintillation, G-M and proportional counters, and

scalars, are provided in an appropriately designed counting room.

Environmental sample analyses are conducted by outside laboratories.

4.4.1.6 Bioassay Program

In vivo bioassay counting equipment is available for quantitative and

qualitative analysis of possible internal deposition of radioactive

contaminants. Consulting laboratory services are used as backup and support

for this program. Appropriate bioassay (urine and fecal) samples are

collected, as necessary, from personnel who work in control areas as an aid in

the evaluation of internal exposure.

4.4.2 Radioactive Materials Safety Program

All Vermont Yankee personnel who work in controlled areas are given training

in radiological safety. Training Program content is specified in appropriate

training department procedures.

Additionally, those personnel in the Radiation Protection Department whose job

entails the handling of sealed and unsealed sources are given departmental

training.

Other departmental procedures detail methods of leak testing sealed sources and

receipt, handling, and storage of radioactive materials. A general calibration

procedure outlines specific techniques for the safe and expeditious handling of

all calibration sources.


Accountability of sources is maintained in inventory records that are updated

semi-annually. Accessibility control is achieved through locked storage,

securing the source in place to prevent unauthorized removal, or continuous

surveillance by authorized personnel.

Accountability of sources that are exempt from leak testing required by the

TRM, but exceed the limits for licensable quantities of radioactive material

specified in Title 10, Code of Federal Regulations, is maintained in inventory

records that are updated annually. All sources of licensable quantity that are

not in use are kept in suitably shielded containers when it is necessary to

minimize personal radiation exposure. All sources of licensable quantity are

kept under the control of authorized personnel when in use.

This system of procedures, training, access control, and accountability is

periodically audited by the Vermont Yankee Quality Assurance Department and/or

one or more contracted service organization(s), collectively defined as the

Quality Assurance Department, as its authorized agent for provision of certain

quality assurance and related support services. Through this mechanism,

compliance with applicable regulations is assured.

4.4.2.1 Facilities and Equipment

Station laboratory facilities and monitoring equipment are discussed in DSAR

Sections 4.3 and 4.4.1.5.

4.4.2.2 Personnel and Procedures

Implementation of the Vermont Yankee radiation protection program, including

source, special, and byproduct material safety, is accomplished by Radiation

Protection Department personnel. The qualifications of these personnel in

radioactive materials safety stem from formal and informal training and from

applied experience in the radiation protection field. Specific training of

Radiation Protection personnel in the safe handling of radioactive materials is

covered by a department training program.


4.4.2.3 Required Materials

All byproduct, source, and special nuclear materials used as reactor fuel,

sealed neutron source for reactor startups, sealed sources for calibration of

reactor instruments, and radioactive monitoring equipment and fission detectors

are possessed in the amounts required for relevant use. All byproduct material

consisting of mixed fission products and corrosion products in the form of

contamination affixed to equipment used for reactor system repair, maintenance,

testing, and/or surveillance may be received, possessed or used in amounts as

required without restriction to chemical or physical form.

With the permanent defueled condition of Vermont Yankee, fission, corrosion,

and activation products from operation are no longer produced. The radioactive

inventory that remains is primarily attributable to sealed radioactive sources,

activated reactor components, nuclear instrumentation, structural materials and

residual radioactivity. The accumulation of small amounts of solid waste as

contaminated materials may easily be controlled.

4.5 LIQUID WASTE MANAGEMENT SYSTEMS

4.5.1 Equipment and Floor Drainage Systems

4.5.1.1 Objective

The objective of the various equipment and floor drainage systems is to remove

all waste fluids from their points of origin in a controlled effective manner

and to deliver them to a suitable disposal system. Radioactive drain

collection is arranged to minimize radioactive exposure to operating personnel

and to prevent uncontrolled leakages to the environs.


Equipment and floor drainage systems shall operate satisfactorily and create no

danger to the health and safety of the general public. These systems shall be

designed and installed to guard against fouling, deposit of solids, and

clogging. Sumps and pumps shall be provided to preclude leakage accumulation

from preventing operation of required equipment. Nonradioactive drainage

systems shall be arranged to assure that no infiltration of radioactive waste

will occur.


Fluids from radioactive and potentially radioactive drains will be collected,

sampled, treated, stored, and/or analyzed prior to disposal in accordance with

10CFR20. Nonradioactive equipment and floor drains empty into the Storm Sewer

System and then discharge into the Circulating Water System piping at the

discharge structure or directly to the Connecticut River at the North Storm

Drain Outfall.

4.5.1.3 Description

4.5.1.3.1 General

The six basic drainage systems are:

1. Radioactive equipment drainage systems

2. Radioactive floor drainage systems

3. Radioactive liquid chemical drainage systems

4. Oil drainage systems

5. Nonradioactive water drainage systems

6. Sanitary drainage systems

The first four systems handle fluid wastes which are radioactive or potentially

radioactive. The last two systems handle fluid wastes originating in areas

which are not radioactive or potentially radioactive. Radioactive wastes are

pumped or drained to the Radwaste System. Nonradioactive wastes are drained to

either the Storm Sewer Drainage System or Sanitary Disposal System.

Radioactive drainage piping is sloped 1/4 inch per foot, and concrete floors

are pitched a minimum of 1/8 inch per foot wherever possible to remove

radioactive wastes as quickly as possible.

Accessible cleanouts are provided at each horizontal change of direction

greater than 45 degrees. Base cleanouts are provided at the base of each stack

approximately 12 inches above the finished floor. In the event a drainage line

becomes stopped or clogged, it can be quickly cleaned out.

The chemical waste sump and equipment and floor drain sumps in the Radwaste

Building and Reactor Building are routed to the Torus. Torus water is

processed through the Torus-as-CST System and is normally used to control spent

fuel pool inventory. Water stored in the torus may be disposed offsite or

discharged to the environs in accordance with applicable permits and regulatory

approvals.


Equipment and floor drain sumps in the Turbine Building are routed to a batch

tank. Tank contents are sampled prior to being transferred, disposed of (via

offsite shipments), or discharged to the environs in accordance with applicable

permits and regulatory approvals.

With the exception of that in the nonradioactive waste drainage systems and

sanitary drainage systems, all drainage piping is carbon steel Schedule 80,

except oil drainage piping, which is carbon steel Schedule 40, and a portion of

condensate drainage piping from the drywell cooling units, which is type 304

stainless steel, Schedule 40. Material used is ASTM A-106, Grade "B". Joints

are welded construction without backing rings. Concrete embedded piping

conforms to the USAS Code B31.1, Sections 1 and 6 for pressure piping. A

portion of condensate drainage piping from the drywell cooling unit is

stainless steel Schedule 80, ASTM A358, Grade TP304.

The Chemistry Laboratory and health physics detergent waste drains are a common

above ground polypropelene lined carbon steel flanged pipe.

Above ground drainage piping used for the Sanitary and Nonradioactive Water

Drainage System is galvanized steel Schedule 40 and galvanized cast iron

drainage fittings. Piping and fittings installed below ground are extra heavy

cast iron.

Vent piping installed above ground is galvanized steel Schedule 40 with

galvanized malleable iron fittings. Piping and fittings installed below ground

are extra heavy cast iron.

All fixtures in the health physics work area, the chemical laboratory, and

fixtures discharging into the Sanitary Drainage System are vented. Each

fixture trap is protected against siphonage and back pressure. The individual

vents collect in a main vent header and terminate full size above the roof.

The radioactive equipment drainage systems receive clean radioactive waste

which is processed and reused. These radioactive systems receive equipment

leak-offs and drains only from equipment handling radioactive liquids.

Radioactive liquids are routed to an equipment drain sump in a closed system

and then pumped to the Radwaste System waste collector tank for future

filtering and demineralization before returning to the Condensate System.

Radioactive or potentially radioactive floor drains are routed to a floor drain

sump in open-ended lines and then pumped to the Radwaste System floor drain

collector from where they are processed and reused. Nonradioactive floor

drains are routed directly to the Storm Sewer System.


Radioactive equipment drains are connected directly to the component serviced

to preclude the possibility of spillage.

4.5.1.3.2 Radioactive Equipment Drainage Systems

Drywell Equipment Drainage Systems

Equipment drains are provided for various components in the drywell and these

lines are run directly to a 500-gallon equipment drain sump. The sump is

provided with two 50 gpm pumps and a number of level switches. A sump pump

will start automatically upon the liquid reaching a pre-set high level and will

trip automatically upon the liquid being lowered to a pre-set low level. A

second sump pump starts and an alarm sounds in the Control Room upon the liquid

reaching a high-high level. Two sump pumps are provided to improve

reliability. An alternator is provided to ensure equal wear on each pump.

Remote operating capability is provided for this system. The common discharge

pipe from the two sump pumps runs through a containment penetration and has two

air-operated valves outside the containment wall. A relief valve provides

overpressure protection of the penetration and connected piping.

Reactor Building Equipment Drainage System

Various equipment drainage in the Reactor Building is piped directly to one of

the two 1000-gallon equipment drain sumps. Each sump (one on the north side

and one on the south side of the Reactor Building) is provided with two 50 gpm

pumps, which discharge to the waste collector tank in the Radwaste Building.

Each sump is provided with level switches used for automatic pump control and

sump high-high level alarm. Pump control switches are located on the radwaste

control panel.

Turbine Building Equipment Drain System

One 1000-gallon sump, located in the feedwater heater area of the Turbine

Building, is provided to collect various equipment drainage. The sump contains

two 50 gpm pumps that discharge to the waste collector tank in the Radwaste

Building.

Sump level switches are used to operate the pumps automatically and provide a

high-high level alarm.


Radwaste Building Equipment Drain System

Various radwaste pumps seal leakage and radwaste tanks, drains, and overflows

are piped directly to one 1000-gallon sump. One 50 gpm sump pump is controlled

automatically by sump level switches and discharges to the waste collector

tank. A sump high level alarm annunciates on the radwaste control panel and in

the Control Room.

4.5.1.3.3 Radioactive Floor Drainage Systems

Drywell Floor Drainage System

The Drywell Floor Drain System collects and disposes of leakage from various

systems and components. Remote operating capability is provided for this

system.

Reactor Building-Floor Drainage System

The Reactor Building Floor Drainage System collects drainage into two

1000-gallon sumps. Each sump (one on the north side and one on the south side

of the Reactor Building) is equipped with two 50 gpm sump pumps which discharge

to the floor drain collector tank.

Turbine Building Floor Drainage System

The Turbine Building Floor Drainage System consists of two 1000-gallon sumps,

each provided with two 50 gpm sump pumps which discharge to the floor drain

collector tank. One sump is located in the condenser area and the other is in

the condensate pump area.

Radwaste Building Floor Drainage System

The Radwaste Building contains one 1000-gallon sump utilized to collect various

floor and tank overflow drainage. One 50 gpm sump pump is provided which

discharges to the floor drain collecting tank.


4.5.1.3.4 Radioactive Liquid Chemical Drainage Systems

The Radioactive Liquid Chemical Drainage System consists of radioactive filter

sludge piping, radioactive chemical drain piping, and radioactive detergent

waste piping. The system handles drainage of radioactive contaminants and

foreign matter such as sludge, detergents, or chemicals from equipment. The

various systems begin with floor drains, direct connection equipment drains,

gutter drains, shower drains, service sinks, and laboratory benches from the

Reactor, Turbine, and Radwaste Buildings. Drainage is collected in waste lines

and discharged into various items of storage and treatment equipment in the

Radwaste Building.

Special showers are provided in the health-physics work area for personnel

decontamination purposes. The drainage from the fixtures in this area is

collected in waste lines and discharged directly into the chemical waste tank

in the Radwaste Building.

4.5.1.3.5 Oil Drainage Systems

Oil drain systems outside the restricted are not considered radioactive. Oil

drain systems within the restricted area are treated as potentially

contaminated. Drainage from systems and equipment using oil is either

collected in sumps or drains to oil separator manholes. Separated oil is

retained while the oil-free water drains into the Storm Sewer System.

Two oil sumps are provided. One is located beneath the floor of the Reactor

Building, and the second is located in the northwest corner of the Turbine

Building.

Oil drainage not routed and collected in a sump is collected in branch lines

which empty into main lines and discharge directly into oil separator manholes

outside the Turbine Building and Control Room Building. The oil separator

manholes function to separate and retain the oil while discharging oil-free

water into the Storm Water Sewer System which drains either to the discharge

structure or the North Storm Drain Outfall to the Connecticut River.

Oil drainage systems in specific areas, which could have propagated a fire,

have been modified. To ensure that spilled fluid is contained within the

respective berm areas, various transformer oil drains have been permanently

plugged.

The oil collected in the oil sumps will be pumped to suitable containers for

disposal using a portable pump. Grab samples can be taken at this time for

radioactive analysis


4.5.1.3.6 Nonradioactive Water Drainage System

The Storm and Nonradioactive Water Drainage System receives rain water, clear

liquid wastes not hotter than 140°F, and drainage from equipment which is

nonradioactive. This drainage is routed separately to the Storm Sewer System.

Heating, ventilation and air conditioning equipment in the Reactor and Turbine

Buildings was considered nonradioactive in the original plant design. Low

levels of tritium have been found in the various drains associated with this

equipment even though modifications to alleviate the condition have been

performed. The levels of contamination have been evaluated and found to be

acceptable for continued discharge to the storm drain system. The condition is

monitored through a surveillance program and reported in the “Annual

Radiological Environmental Operating Report”. Funnel type equipment drains and

floor drains serving this equipment are collected in branch lines, empty into

main drain lines, and discharge into the Storm Sewer System. A separate

Nonradioactive Water Drainage System is provided in the Turbine Building for

certain items of equipment. Air handling equipment, certain water pumps on the

basement floor and miscellaneous equipment on the ground floor were considered

nonradioactive in the original plant design. Low levels of activity have been

found in the turbine building clean sump associated with this equipment. The

levels of contamination have been evaluated and found to be acceptable for

continued discharge to the discharge structure. The condition is monitored

through a surveillance program and reported in the “Annual Radiological

Environmental Operating Report”. Funnel type equipment drains and floor drains

in these areas are collected in branch lines, empty into main drain lines, and

discharge into the clean equipment and floor drain sump, located below the

basement floor. To improve administrative control over the sources of

radioactive liquid entering this sump, floor drains, which are aligned to this

sump, have been permanently plugged or fitted with removable plugs. In

addition, the drain header from floor and equipment drains in the vicinity of

the demineralized water transfer pumps and the station air compressor receiver

tanks has been cut, capped and valved to allow sampling prior to release.

Other equipment drains are either permanently plugged or go directly to the

sump. Sump pumps are provided to transfer the discharge from the Turbine

Building to the service water discharge.

In addition to the low levels of tritium discussed above, surface run-off from

within the Protected Area carries low levels of particulate activity to the

Storm Sewer System. The low levels of contamination in the Storm Sewer System

have been evaluated to ensure that the calculated maximum release is a

significant percentage less than the total body and critical organ doses

allowed under the routine effluent ALARA objectives of 10CFR50, Appendix I.


Acid resistant drains and piping are provided in areas where highly

concentrated acids are present. Duriron floor drains and piping are provided

to serve the Turbine Building water treatment area. Duriron floor drains and

piping, plus Duriron funnel type equipment drain and piping, are also provided

to serve the Battery Room in the Control Building.

4.5.1.3.7 Sanitary Drainage Systems

The Sanitary Drainage System is provided for the convenience, health, and

safety of facility personnel. This system receives the domestic sewage from

various fixtures that are water supplied and discharges liquid wastes.

Except for fixtures in the health-physics work area, all water closets,

urinals, lavatories, drinking water coolers, service sinks, kitchen units, and

showers in the Turbine and Control Buildings discharge into the Sanitary

Drainage System. Each fixture is trapped and vented, then collected in branch

lines, emptied into main soil lines, and discharged by gravity into the

Sanitary Disposal System.


The Equipment and Floor Drainage Systems are normally in operation during all

modes of facility operation. Satisfactory operation is demonstrated

continuously without the need for special testing or inspection.

4.5.2 Liquid Radwaste System

4.5.2.1 Objective

Liquid radwaste is collected and processed as required for reuse. Any liquid

waste which would not be suitable for reuse could be diluted and discharged

from the facility.


Liquid radwaste is collected and processed to assure that the release of liquid

radwaste is kept as low as reasonably achievable and is within the annual dose

limits specified in 10CFR20.1301.

Liquid radwaste shall be contained to prevent the inadvertent release of

significant quantities of liquid radioactive material to unrestricted areas so

that resulting radiation exposures are within the limits of 10CFR20.1301.


4.5.2.3 Description

The Liquid Radwaste System collects, processes, stores, and disposes of all

radioactive liquid wastes. The radwaste facility is located in the Radwaste

Building, with the exception of the waste sample tanks, floor drain sample

tank, and waste surge tank, all located outdoors at grade level (see Drawings

G-191151 and G-191152).

Included in the Liquid Radwaste System are the following:

1. Floor and Equipment Drain System for handling potentially radioactive

wastes.

2. Tanks, piping, pumps, process equipment, instrumentation, and auxiliaries

necessary to collect, process, store and dispose of potentially radioactive

wastes.

Equipment is selected, arranged, and shielded to permit operation, inspection,

and maintenance with acceptable personnel exposures. For example, sumps,

pumps, valves, and instruments are located in controlled access areas. Tanks

and processing equipment which can contain large quantities of liquid radwastes

are shielded. The Radwaste System equipment, equipment arrangement,

capacities, flow paths, and flow rates are shown in Drawings 5920-644, Sh.2 and

G-191177, Shs. 1 through 4. Operation of the Waste System is essentially

manual start-automatic stop.

This is a batch-type system wherein the wastes are separately collected and

processed based on the most efficient methods. Cross-connections between

subsystems provide additional flexibility for processing of the wastes by

alternate methods. Treated wastes can be: (a) returned to the system for

reuse, (b) diluted and discharged from the facility, or (c) if not suitable for

either reuse or discharge, they receive additional processing. The liquid

radwastes are classified, collected, and treated as high purity, low purity,

chemical or detergent wastes. The terms "high" purity and "low" purity refer

to conductivity and not radioactivity.


The Liquid Radwaste System has been or is in the process of being abandoned and

it is no longer utilized. However, it can be restored to operational status if

desired. The chemical waste sump and sumps (equipment and floor) in the

Radwaste Building and Reactor Building are routed to the Torus. The

Torus-as-CST water treatment system, installed after permanent plant shutdown,

recirculates and cleans Torus water. The system provides for suitable Spent

Fuel Pool water makeup and letdown. The system contains parallel paths of

pumps, filters and demineralizers. Suction is taken from the Torus and

discharge is either to the SFP or back to the Torus.

The Liquid Radwaste System is operated such that any liquid radioactive waste

releases would be minimized. Successful processing of all liquid wastes to

maintain a low release system calls for special plant controls. Detergent and

soap used to clean areas and equipment are kept to a minimum. The majority of

chemical wastes are neutralized and metered slowly into higher purity water for

processing. Low purity water is filtered, and combined with higher purity

water for reprocessing. The combination of low purity, chemical, and detergent

waste into higher purity waste streams allows for reprocessing and plant reuse

and reduces the need to discharge any fraction of the waste stream. Liquid

waste could be discharged from the waste sample tank to the environment through

approved discharge pathways. The maximum concentration of tritium and

dissolved noble gases at the point of discharge will not exceed applicable

limits.

The processing equipment is located within a concrete building to provide

secondary enclosures for the wastes in the event of leaks or overflows. Tanks

and equipment which contain wastes with radioactive concentrations are

shielded. Except where flanges are required for maintenance, all pipe

connections are welded to reduce the probability of leaks. Chemistry

lab/detergent waste piping is lined with plastic and cannot be welded. As a

result, these lines are flanged, and the flanges, located in the switchgear

room, are fitted with Vue-Guards to reveal and collect any leakage to minimize

the potential for flooding. Process lines which penetrate shield walls are

routed to prevent a direct radiation path from the tanks or equipment for which

shielding is required. Control of the Waste System is from a local panel in

the Radwaste Building Control Room.

Therefore, because the radioactivity concentrations in the liquid radwaste

effluent do not exceed the guideline limits of 10CFR20, the Liquid Radwaste

System fulfills the design basis.


4.5.2.3.1 High Purity Wastes

High purity (low conductivity) liquid wastes are collected in the waste

collector tank.

The high purity wastes are processed by filtration and ion exchange through the

waste collector filter or fuel pool and waste demineralizers as required.

After processing, the liquid is pumped to the waste sample tank where it is

sampled and either recycled for additional processing or transferred to the

condensate storage tank for spent fuel pool inventory makeup.

Should discharge be necessary, wastes would be sampled on a batch basis.

Samples from the waste sample tanks are analyzed for water quality and

radioactivity. If high purity requirements are met, the contents are

transferred to the condensate storage tank.

If high purity requirements are not met, the liquid wastes are recycled through

the Radwaste System or could be discharged. The high purity requirements are

specified in plant procedures.

Table 4.5.2.5 lists the radionuclide discharge concentrations at original 100%

power operation, assuming an 80% plant capacity factor and a dilution flow of

20,000 gallons per minute. The total annual release values are based on output

from the BWR Gale Code (NUREG-0016). The information in Table 4.5.2.5 is

historical and is being retained to provide bounding values.

It can be seen from Column 5 (Fraction of ECLw) that the concentration for each

radionuclide at the point of discharge is several orders of magnitude below

limits established in 10CFR20 for release of effluents to unrestricted areas.

The design of the Radwaste Treatment System is therefore consistent with the

policy that any radioactive effluents would be reduced to the lowest reasonably

achievable level.

Liquid effluents discharged from the plant enter a 30-inch dilution water line

which terminates in a diffuser at the head end of the aerating apron at the

discharge structure. The effluent enters the Vernon Pond at the downstream end

of the aerating apron.


4.5.2.3.2 Low Purity Wastes

Low purity (high conductivity) liquid wastes which are collected in the floor

drain collector tank are from the following sources:

1. Drywell floor drains

2. Reactor Building floor drains

3. Radwaste Building floor drains

4. Turbine Building floor drains

These wastes generally have low concentrations of radioactive impurities, and

processing consists of filtration and a combination with the high purity waste

in the waste collector tank, with subsequent processing, as high purity waste.

Operation of the Liquid Radwaste System is such that all liquid wastes will be

processed and reused without having to discharge for total system volume

control, or water purity constraints. For the purpose of analyzing future

radiological impacts during the plant's life, it is assumed that 1% of the

combined processed stream treated each year would be discharged from the

facility. Table 4.5.2.1 indicates the radioactive liquid waste sources, flow

rates, expected activities, holdup times, decontamination factors, and assumed

fraction of waste discharged from the Liquid Waste Processing System at

original 100% power operation. The information in Table 4.5.2.1 is historical

and is being retained to provide bounding values. Table 4.5.2.2 lists the

capacity of all major tanks in the Liquid Radwaste System. The plant operating

parameters and design information provides the necessary inputs for the

calculation of potential radioactive source terms by the Nuclear Regulatory

Commission's BWR Gale Computer Code (NUREG-0016). Table 4.5.2.3 lists the

calculated liquid source terms for Vermont Yankee at original 100% power

operation. The information in Table 4.5.2.3 is historical and is being

retained to provide bounding values. Based on processing parameters in Tables

4.5.2.1 and 4.5.2.2, Table 4.5.2.4 lists the activity input to the Liquid

Radwaste System at original 100% power operation for all major nuclides. The

information in Table 4.5.2.4 is historical and is being retained to provide

bounding values. The radioactivities listed represent activities prior to

treatment and will be reduced significantly due to decontamination and isotopic

decay while passing through treatment systems.


4.5.2.3.3 Chemical Wastes

Chemical wastes are collected in the chemical waste tank and are from the

following sources:

1. Chemical lab waste

2. Laboratory drains

3. Sample sinks

When the chemical concentrations are low enough, these wastes may be

neutralized and processed by filtration and dilution in the same manner and

with the same equipment as the low purity wastes. When the chemical

concentrations are too high, these wastes may receive additional processing.

4.5.2.3.4 Detergent Wastes

Detergent wastes are collected in the detergent waste tank. These wastes are

primarily from radioactive decontamination solutions which contain detergents.

Detergent wastes are of low radioactivity concentration (<10-5 µCi/cc). Because

detergents will foul ion exchange resins, their use is minimized in the plant.

For initial cleanings, little or no detergent is used. The facility uses an

off-site cleaning laundry, thus minimizing the quantity of waste generated.

Detergent wastes are normally dumped to the floor drain collector tank for

processing with low purity waste.

4.5.2.4 Evaluation

The Radwaste Building is classified as a Class II seismic design structure, and

the Waste System is classified as Class II seismic design equipment, since

failure of the structure and/or the equipment will not cause a significant

release of radioactivity.


With the exception of three 10,000 gallon sample tanks and a 35,000-gallon

waste surge tank, the Radwaste System processing equipment and storage tanks

are located in the Radwaste Building. Failure of the building could be

postulated and the failure could conceivably result in damage to storage tanks

within the building. If the contents of all the tanks within the building were

released, and this is extremely unlikely because of the compartment-like

arrangement and the arrangement of shield walls, the liquid waste would

ultimately accumulate in the basement of the building. Considering the volume

with all tanks two-thirds full, and the existing basement floor space, the

accumulation would amount to approximately 18 inches of water. Considering the

low driving head of the liquid waste in the basement of the building and the

distance to the river, it is very unlikely that entrained activity would find a

leakage path to the river. It is possible that some seepage may occur through

the building foundation, but such seepage would be expected to be small in

quantity and would tend to be absorbed in the soil surrounding the foundation.

If the seepage persisted over a long period of time, the soil surrounding the

foundation would not only act as a liquid absorber, but also as a filter.

The outside storage tanks located within approximately 1.5 foot high concrete

dikes, Figure 4.5.2-8, provide a less remote potential for off-site discharge

of activity. Sumps are provided within the diked area to provide for draining

any leakage or rainwater. Although it is virtually impossible to postulate a

condition which would result in the complete discharge of the contents of the

four outside tanks into the river, the consequences of such an occurrence have

been analyzed.

The maximum gross radioactivity in the four outside tanks is limited to

3.2 curies on the basis of an accidental spill from all tanks due to a seismic

event great enough to damage them. Assuming a low river flow of 108 ft3/sec, a

one day period over which the radioactive liquid wastes are diluted in the

river, and consumption of the water by individuals at standard man consumption

rate (3,000 ml/day), the single intake by an individual would not exceed

one-third the yearly intake allowable by 10CFR20 for unidentified radioisotopes

(1 x 10-6 Ci/ml).

Radwaste liquids are processed on a batch basis. The design of the system

precludes direct discharge of either unprocessed or processed liquids without

first holding them up in the sample tanks where the liquid is analyzed for

activity levels. Procedural controls would be implemented to ensure that the

activity of processed liquid, after dilution, will not exceed the guideline

limits of 10CFR20, prior to liquids being released to the river.


In order to release liquid from the sample tanks to the river, the sample pumps

must be started, valves opened, and the flow controller positioned. In

addition, a dilution water pump must be put into operation prior to discharge

of the processed liquid. An interlock precludes discharge of processed liquid

to the river when dilution water is unavailable.

The process radiation monitor in the discharge line from the sample tanks to

the river is provided to back up the administrative control provided by sample

tank liquid analysis. It provides a warning to the appropriate facility

personnel that the activity of the processed liquid is approaching ten times

the annual average concentration values of Appendix B, Table 2, Column 2, of

10CFR20.1001-2402. When appropriate, facility personnel could take action to

reduce processed liquid flow or terminating flow entirely to assure that the

releases do not exceed the limits for which the facility is licensed.

Sufficient administrative and design control is provided to prevent accidental

releases of liquid effluents from the Radwaste System.

Therefore, the design basis is considered met.


The Liquid Radwaste System is normally operating on an "as-required" basis

thereby demonstrating the ability to perform its function without special

testing.


TABLE 4.5.2.1

Vermont Yankee Radioactive Liquid Waste Processing Parameters

Waste Stream Input Sources Input Flow

Rates (gpd)

Fraction of Primary Coolant Activity (pca)

Holdup Time (Days) Available Process Decontamination Factors

Assumed

Fraction of Waste Stream Discharged

Collection Process Discharge Nuclide 1st Demin(a)

2nd Demin(b) Total DF

High Purity Waste

1) Drywell Equip. Drains

3400 1.0 I = 10 10 102

2) Reactor Bldg. Equip. Drains

3720 0.01

3) Radwaste Bldg. Equip. Drains

1060 0.01 0.75 0.15 1.11 Cs, Rb = 2 10 20 0.01

4) Turbine Bldg. Equip. Drains

2960 0.01

5) Condensate Phase Sep.

8100 2 x 10-6

6) Cleanup Phase Sep.

640 0.002 Other Nuclides =

10 10 102

7) Resin Rinse 5000 0.002

Low Purity Waste

1) Drywell Floor Drains

700 1.0 I = 10 10 102

2) Reactor Bldg. Floor Drains

2000 0.01 1.39 0.15 1.11 Cs, Rb = 2 10 20 0.01

3) Radwaste Bldg. Floor Drains

1000 0.01 Other Nuclides =

10 10 102

4) Turbine Bldg. Floor Drains

2000 0.01

Chemical Waste

1) Chem. Lab. Waste 100 0.02 I = 10 10 102

2) Lab. Drains 500 0.02 2.67 0.15 1.11 Cs, Rb = 2 10 20 0.01

3) Personal Shower and Decon. Drains

900 1.4x10-4 0.44 Other Nuclides =

10 10 102

Detergent Waste

1) Decon. Drains 900 1.4x10-4 0.44 0.15 1.11 I = 10 10 102 0.01

Cs, Rb = 2 10 20

Other Nuclides =

10 10 102

(a) Fuel Pool (Powdered Resin) Filter-Demineralizer

(b) Radwaste Deep Bed Demineralizer

NOTE: Data contained in this table is based on original 100% power operation, and is retained as historical information.


TABLE 4.5.2.2

Vermont Yankee Liquid Radwaste System Tank Capacities

Tank Capacity Per Tank (Gal.)

Waste Collection Tank (1) 25,000 Waste Surge Tank (1) 35,000

Floor Drain Collection Tank (1) 25,000 Chemical Waste Tank (1) 4,000 Detergent Waste Tank (1) 1,000

Floor Drain Sample Tank (1) 10,000 Waste Sample Tank (2) 10,000

TABLE 4.5.2.3

Vermont Yankee Liquid Effluents

Assumption of Concentration Annual Releases to Discharge Canal in Primary Adjusted Detergent Total

Half-Life Coolant High Purity Low Purity Chemical Total Lws. Total Wastes

Nuclide (Days) (Micro Ci/ml) (Curies) (Curies) (Curies) (Curies) (Ci/Yr) (Ci/Yr) (Ci/Yr)


4.0-29 of 53

CORROSION AND ACTIVATION PRODUCTS Na24 6.27E-01 8.72E-03 .00130 .00021 .00000 .00151 .01008 .00000 .01000

P32 1.43E+01 2.12E-04 .00010 .00002 .00000 .00012 .00078 .00000 .00078

Cr51 2.78E+01 5.31E-03 .00249 .00052 .00001 .00302 .02014 .00000 .02000

Mn54 3.03E+02 6.38E-05 .00003 .00001 .00000 .00004 .00025 .00000 .00025

Mn56 1.08E-01 3.90E-02 .00004 .00000 .00000 .00004 .00030 .00000 .00030

Fe55 9.49E+02 1.06E-03 .00051 .00011 .00000 .00062 .00415 .00000 .00420

Fe59 4.51E+01 3.19E-05 .00002 .00000 .00000 .00002 .00012 .00000 .00012

Co58 7.10E+01 2.13E-04 .00010 .00002 .00000 .00012 .00082 .00000 .00082

Co60 1.92E+03 4.25E-04 .00020 .00004 .00000 .00025 .00166 .00000 .00170

Cu64 5.35E-01 2.87E-02 .00352 .00054 .00001 .00407 .02713 .00000 .02700

Zn65 2.45E+02 2.13E-04 .00010 .00002 .00000 .00012 .00083 .00000 .00083

Zn69m 5.73E-01 1.92E-03 .00026 .00004 .00000 .00030 .00199 .00000 .00200

Zn69 3.95E-02 .0 .00028 .00004 .00000 .00032 .00214 .00000 .00210

W187 9.95E-01 3.00E-04 .00007 .00001 .00000 .00008 .00054 .00000 .00054

Np239 2.35E+00 7.24E-03 .00254 .00049 .00001 .00304 .02026 .00000 .02000

FISSION PRODUCTS

Br83 1.00E-01 2.13E-03 .00000 .00000 .00000 .00000 .00001 .00000 .00001

Sr89 5.21E+01 1.06E-04 .00005 .00001 .00000 .00006 .00041 .00000 .00041

Sr90 1.03E+04 6.38E-06 .00000 .00000 .00000 .00000 .00002 .00000 .00003

Sr91 4.03E-01 3.72E-03 .00030 .00004 .00000 .00034 .00227 .00000 .00230

Y91m 3.47E-02 .0 .00019 .00003 .00000 .00022 .00146 .00000 .00150

Y91 5.90E+01 4.25E-05 .00003 .00001 .00000 .00004 .00025 .00000 .00025

Sr92 1.13E-01 7.86E-03 .00001 .00000 .00000 .00001 .00008 .00000 .00008

Y92 1.47E-01 4.90E-03 .00011 .00001 .00000 .00012 .00082 .00000 .00082

Y93 4.24E-01 3.74E-03 .00033 .00005 .00000 .00037 .00249 .00000 .00250

Zr95 6.52E+01 7.44E-06 .00000 .00000 .00000 .00000 .00003 .00000 .00003

Nb95 3.50E+01 7.43E-06 .00000 .00000 .00000 .00000 .00003 .00000 .00003

Mo99 2.80E+00 2.08E-03 .00077 .00015 .00000 .00092 .00613 .00000 .00610

Tc99m 2.50E-01 1.76E-02 .00117 .00020 .00000 .00138 .00919 .00000 .00920

Ru103 3.95E+01 2.12E-05 .00001 .00000 .00000 .00001 .00008 .00000 .00008

Rh103m 3.95E-02 .0 .00001 .00000 .00000 .00001 .00008 .00000 .00008

Ru105 1.85E-01 1.69E-03 .00002 .00000 .00000 .00002 .00014 .00000 .00014

Rh105m 5.21E-04 .0 .00002 .00000 .00000 .00002 .00014 .00000 .00014

Rh105 1.50E+00 .0 .00007 .00001 .00000 .00008 .00053 .00000 .00053

Ru106 3.66E+02 3.19E-06 .00000 .00000 .00000 .00000 .00001 .00000 .00001

Rh106 3.47E-04 .0 .00000 .00000 .00000 .00000 .00001 .00000 .00001

Te129m 3.40E+01 4.25E-05 .00002 .00000 .00000 .00002 .00016 .00000 .00016

TABLE 4.5.2.3 (Continued) Vermont Yankee Liquid Effluents Concentration Annual Releases to Discharge Canal in Primary Adjusted Detergent Total Half-Life Coolant High Purity Low Purity Chemical Total Lws. Total Wastes Nuclide (Days) (Micro Ci/ml) (Curies) (Curies) (Curies) (Curies) (Ci/Yr) (Ci/Yr) (Ci/Yr)


Te129 4.80E-02 .0 .00001 .00000 .00000 .00002 .00010 .00000 .00010

Te131m 1.25E+00 1.01E-04 .00003 .00000 .00000 .00003 .00021 .00000 .00021

Te131 1.74E-02 .0 .00000 .00000 .00000 .00001 .00004 .00000 .00004

I131 8.05E+00 5.24E-03 .00230 .00048 .00001 .00278 .01858 .00000 .01900

Te132 3.25E+00 1.04E-05 .00000 .00000 .00000 .00000 .00003 .00000 .00003

I132 9.58E-02 2.12E-02 .00002 .00000 .00000 .00002 .00011 .00000 .00011

I133 8.75E-01 1.90E-02 .00392 .00067 .00001 .00460 .03068 .00000 .03100

Cs134 7.50E+02 3.19E-05 .00008 .00002 .00000 .00009 .00062 .00000 .00062

I135 2.80E-01 1.65E-02 .00062 .00008 .00000 .00070 .00468 .00000 .00470

Cs136 1.80E+01 2.11E-05 .00005 .00001 .00000 .00006 .00039 .00000 .00039

Cs137 1.10E+04 1.45E-05 .00018 .00004 .00000 .00022 .00145 .00000 .00150

Ba137m 1.77E-03 .0 .00017 .00004 .00000 .00020 .00136 .00000 .00140

Ba140 1.28E+01 4.23E-04 .00019 .00004 .00000 .00023 .00155 .00000 .00160

Tritium Release 4 Curies per year

NOTE: Data contained in this table is based on original 100% power operation, and is retained as historical information.

La140 1.67E+00 .0 .00007 .00002 .00000 .00009 .00059 .00000 .00059

La141 1.62E-01 .0 .00000 .00000 .00000 .00000 .00003 .00000 .00003

Ce141 3.25E+01 3.18E-05 .00002 .00000 .00000 .00002 .00013 .00000 .00013

Ce143 1.38E+00 3.05E-05 .00001 .00000 .00000 .00001 .00007 .00000 .00007

Pr143 1.57E+01 4.23E-05 .00002 .00000 .00000 .00002 .00016 .00000 .00016

Ce144 2.84E+02 3.19E-06 .00000 .00000 .00000 .00000 .00001 .00000 .00001

Pr144 1.20E-02 .0 .00000 .00000 .00000 .00000 .00001 .00000 .00001

Nd147 1.11E+01 3.17E-06 .00000 .00000 .00000 .00000 .00001 .00000 .00001

All

Others 1.86E-01 .00000 .00000 .00000 .00001 .00004 0.0 .00004

Total

(Except

Tritium)

3.86E-01 02235 .00403 . .00005 02644 .17644 .00000 .18000


TABLE 4.5.2.4

Activity Input to Liquid Radwaste System (Ci/yr)

Nuclide

ci/ml

PCA

Drywell

Equip.

Drains

Reactor

Bldg.

Equip.

Drain

Radwaste

Bldg.

Equip.

Drain

Turbine

Bldg

Equip.

Drain

Condensate

Phase Sep.

Cleanup

Phase

Sep.

Resin

Rinse

Drywell

Floor

Drains

Reactor

Bldg.

Floor

Drains

Radwaste

Bldg.

Floor

Drains

Turbine

Bldg.

Floor

Drains

Chem

Lab.

Waste

Lab.

Drains

Personal

Shower

and

Decon.

Drains

Na24 8.72E-03 3.3E+01 3.6E-01 1.02E-01 2.8E-01 1.6E-04 1.2E-02 9.6E-02 6.7E+00 1.9E-01 9.6E-02 1.9E-01 1.9E-02 9.6E-02 1.2E-03

P32 2.12E-04 7.9E-01 8.7E-03 2.5E-03 6.9E-03 3.8E-06 3.0E-04 2.3E-03 1.6E-01 4.7E-03 2.3E-03 4.7E-03 4.7E-04 2.3E-03 2.9E-05

Cr51 5.31E-03 2.0E+01 2.2E-01 6.2E-02 1.7E-01 9.6E-05 7.5E-03 5.9E-02 4.1E+00 1.2E-01 5.9E-02 1.2E-01 1.2E-02 5.9E-02 7.4E-04

Mn54 6.38E-05 2.4E-01 2.6E-03 7.5E-04 2.1E-03 1.1E-06 9.0E-05 7.0E-04 4.9E-02 1.4E-03 7.3E-04 1.4E-03 1.4E-04 7.0E-04 8.9E-06

Mn56 3.90E-02 1.47E+02 1.6E+00 4.6E-01 1.3E+00 7.0E-04 5.5E-02 4.3E-01 3.0E+01 8.6E-01 4.3E-01 8.6E-01 8.6E-02 4.3E-01 5.4E-03

Fe55 1.06E-03 3.9E+00 4.4E-02 1.2E-02 3.5E-02 1.9E-05 1.5E-03 1.2E-02 8.2E-01 2.3E-02 1.2E-02 2.3E-02 2.3E-02 1.2E-02 1.5E-04

Fe59 3.19E-05 1.2E-01 1.3E-03 3.7E-04 1.0E-03 5.7E-07 4.5E-05 3.5E-04 2.5E-02 7.0E-04 3.5E-04 7.0E-04 7.0E-05 3.5E-04 4.4E-06

Co58 2.13E-04 8.0E-01 8.8E-03 2.5E-03 7.0E-03 3.8E-06 3.0E-04 2.4E-03 1.6E-01 4.7E-03 2.4E-03 4.7E-03 4.7E-04 2.4E-03 3.0E-05

Co60 4.25E-04 1.6E+00 1.7E-02 5.0E-03 1.4E-02 7.7E-06 6.0E-04 4.7E-03 3.3E-01 9.4E-03 4.7E-03 9.4E-03 9.4E-04 4.7E-03 5.9E-05

Cu64 2.87E-02 1.1E+02 1.2E+00 3.4E-01 9.4E-01 5.2E-04 4.0E-02 3.2E-01 2.2E+01 6.3E-01 3.2E-01 6.3E-01 6.3E-02 3.2E-01 4.0E-03

Zn65 2.13E-04 8.0E-01 8.8E-03 2.5E-03 7.0E-03 3.8E-06 3.0E-04 2.4E-03 1.6E-01 4.7E-03 2.4E-03 4.7E-03 4.7E-01 2.4E-03 3.0E-05

Zn69m 1.92E-03 7.2E+00 7.9E-02 2.2E-02 6.3E-02 3.5E-05 2.7E-03 2.1E-02 1.5E+00 4.2E-02 2.1E-02 4.2E-02 4.2E-03 2.1E-02 2.7E-04

W198 3.00E-04 1.1E+00 1.2E-02 2.5E-03 9.8E-03 5.4E-06 4.2E-04 3.3E-03 2.3E-01 6.6E-03 3.3E-03 6.6E-03 6.6E-04 3.3E-03 4.2E-05

Np239 7.24E-03 2.7E+01 3.0E-01 8.5E-02 2.4E-01 1.3E-04 1.0E-02 8.0E-02 5.6E+00 1.6E-01 8.0E-02 1.6E-01 1.6E-02 8.0E-02 1.0E-03

Br83 2.13E-03 8.0E+00 8.8E-02 2.5E-02 7.0E-02 3.8E-05 3.0E-03 2.4E-02 1.6E+00 4.7E-02 2.4E-02 4.7E-02 4.7E-03 2.4E-02 3.0E-04

Sr89 1.06E-04 3.9E+00 4.4E-03 1.2E-03 3.5E-03 1.9E-06 1.5E-04 1.2E-03 8.2E-02 2.3E-03 1.2E-03 2.3E-03 2.3E-04 1.2E-03 1.5E-05

Sr90 6.38E-06 2.4E-02 2.6E-04 7.5E-05 2.1E-04 1.1E-07 9.0E-06 7.0E-05 4.9E-03 1.4E-04 7.0E-05 1.4E-04 1.4E-05 7.0E-05 8.9E-07

Sr91 3.72E-03 1.4E+01 1.5E-01 4.4E-02 1.2E-01 6.7E-05 5.2E-03 4.1E-02 2.9E+00 8.2E-02 4.1E-02 8.2E-02 8.2E-03 4.1E-02 5.2E-04

Y91 4.25E-05 1.6E-01 1.7E-03 5.0E-04 1.4E-03 7.7E-07 6.0E-05 4.7E-04 3.3E-02 9.4E-04 4.7E-04 9.4E-04 9.4E-05 4.7E-04 5.9E-03

Sr92 7.86E-03 2.9E+01 3.2E-01 9.2E-02 2.6E-01 1.4E-04 1.1E-02 8.7E-02 6.1E+00 1.7E-01 8.7E-02 1.7E-01 1.7E-02 8.7E-02 1.1E-03

Y92 4.90E-03 1.8E+01 2.0E-01 5.7E-02 1.6E-01 8.8E-05 6.9E-03 5.4E-02 3.8E+00 1.1E-01 5.4E-02 1.1E-01 1.1E-02 5.4E-02 6.8E-04

Y93 3.74E-03 1.4E+01 1.5E-01 4.4E-02 1.2E-01 6.7E-05 5.3E-03 4.1E-02 2.9E+00 8.3E-02 4.1E-02 8.3E-02 8.3E-03 4.1E-02 5.2E-04

Zr95 7.44E-06 2.8E-02 3.1E-04 8.7E-05 2.4E-04 1.3E-07 1.0E-05 8.2E-05 5.8E-03 1.6E-04 8.2E-05 1.6E-04 1.6E-05 8.2E-05 1.0E-06

Nb95 7.43E-06 2.8E-01 3.1E-04 8.7E-05 2.4E-04 1.3E-07 1.0E-05 8.2E-05 5.8E-03 1.6E-04 8.2E-05 1.6E-04 1.6E-05 8.2E-05 1.0E-06

Mo99 2.08E-03 7.8E+00 8.6E-02 2.4E-02 6.8E-02 3.7E-01 2.9E-03 2.3E-02 1.6E+00 4.6E-02 2.3E-02 4.6E-02 4.6E-03 2.3E-02 2.9E-04

Tc99m 1.76E-02 6.6E+01 7.2E-01 2.1E-01 5.8E-01 3.2E-04 2.5E-02 1.9E-01 1.4E+01 3.9E-01 1.9E-01 3.9E-01 3.9E-02 1.9E-01 2.4E-03

Ru103 2.12E-05 8.0E-02 8.7E-04 2.5E-04 6.9E-04 3.8E-07 3.0E-05 2.3E-04 1.6E-02 4.7E-04 2.3E-04 4.7E-04 4.7E-05 2.3E-04 2.9E-06

Ru105 1.69E-03 6.3E+00 6.9E-02 2.0E-02 5.5E-02 3.0E-05 2.4E-03 1.9E-02 1.3E+00 3.7E-02 1.9E-02 3.7E-02 3.7E-03 1.9E-02 2.3E-04

Ru106 3.19E-06 1.2E-02 1.3E-04 3.7E-05 1.0E-04 5.7E-08 4.5E-06 3.5E-05 2.5E-03 7.0E-05 3.5E-05 7.0E-05 7.0E-06 3.5E-05 4.4E-07


TABLE 4.5.2.4

(Continued)

Activity Input to Liquid Radwaste System (Ci/yr)

Nuclide

ci/ml

PCA

Drywell

Equip.

Drains

Reactor

Bldg.

Equip.

Drain

Radwaste

Bldg.

Equip.

Drain

Turbine

Bldg.

Equip

Drain

Condensate

Phase Sep.

Cleanup

Phase

Sep.

Resin

Rinse

Drywell

Floor

Drains

Reactor

Bldg.

Floor

Drains

Radwaste

Bldg.

Floor

Drains

Turbine

Bldg.

Floor

Drains

Chem.

Lab.

Waste

Lab.

Drains

Personal

Shower

and

Decon.

Drains

Te129m 4.25E-05 1.6E-01 1.7E-03 5.0E-04 1.4E-03 7.7E-07 6.0E-06 4.7E-04 3.3E-02 9.4E-04 4.7E-04 9.4E-04 9.4E-05 4.7E-04 5.9E-06

Te131m 1.01E-04 3.8E-01 4.2E-03 1.2E-03 3.3E-03 1.8E-06 1.4E-04 1.1E-03 7.8E-02 2.2E-03 1.1E-03 2.2E-03 2.2E-04 1.1E-03 1.4E-05

I131 5.24E-03 2.0E+00 2.2E-01 6.1E-02 1.7E-01 9.4E-05 7.4E-03 5.8E-02 4.1E+00 1.2E-01 5.8E-02 1.2E-01 1.2E-02 5.8E-02 7.3E-04

Te132 1.04E-05 3.9E-02 4.3E-04 1.2E-04 3.4E-04 1.9E-07 1.5E-05 1.1E-04 8.0E-03 2.3E-04 1.1E-04 2.3E-04 2.3E-05 1.1E-04 1.4E-06



Cs134 3.19E-05 1.2E-01 1.3E-03 3.7E-04 1.0E-03 5.7E-07 4.5E-05 3.5E-04 2.5E-02 7.0E-04 3.5E-04 7.0E-04 7.0E-05 3.5E-04 4.4E-06




Ba140 4.23E-04 1.59E+00 1.7E-02 4.9E-03 1.4E-02 7.6E-06 6.0E-04 4.7E-03 3.3E-01 9.3E-03 4.7E-03 9.3E-03 9.3E-04 4.7E-03 5.9E-05

Ce141 3.18E-05 1.2E-01 1.3E-03 3.7E-04 1.0E-03 5.7E-07 4.5E-05 3.5E-04 2.5E-02 7.0E-04 3.5E-04 7.0E-04 7.0E-05 3.5E-04 4.4E-06


Pr143 4.23E-05 1.6E-01 1.7E-03 4.9E-04 1.4E-03 7.6E-07 6.0E-05 4.7E-04 2.3E-02 9.3E-04 4.7E-04 9.3E-04 9.3E-05 4.7E-04 5.9E-06


Nd147 3.17E-06 1.2E-02 1.3E-04 3.7E-05 1.0E-04 5.7E-08 4.5E-06 3.5E-05 2.5E-03 7.0E-05 3.5E-05 7.0E-05 7.0E-06 3.5E-05 4.4E-07

All

Others 1.86E-01 7.0E+02 7.6E+00 2.2E+00 6.1E+00 3.3E-03 2.6E-01 2.1E+00 1.4E+02 4.1E+00 2.1E+00 4.1E+00 4.1E-01 2.1E+00 2.6E-02

NOTE: Data contained in this table is based on original 100% power operation, and is retained as historical information


4.0-33 of 53

TABLE 4.5.2.5

Radionuclide Discharge Concentrations Nuclide

Total Annual

Release (Ci/Yr)

Discharge Concentration

(Ci/ml) ECLw

(Ci/ml) Fraction of

ECL

Na24 1.0x10-2 3.1x10-10 5x10-5 6.2x10-6

P32 7.8x10-4 2.4x10-11 9x10-6 2.7x10-7

Cr51 2.0x10-2 6.3x10-10 5x10-4 1.3x10-6

Mn54 2.5x10-4 7.8x10-12 3x10-5 2.6x10-7

Mn56 3.0x10-4 9.4x10-12 7x10-5 1.3x10-7

Fe55 4.2x10-3 1.3x10-10 1x10-4 1.3x10-6

Fe59 1.2x10-4 3.8x10-12 1x10-5 3.8x10-7

Co58 8.2x10-4 2.6x10-11 2x10-5 1.3x10-6

Co60 1.7x10-3 5.3x10-11 3x10-6 1.8x10-5

Cu64 2.7x10-2 8.5x10-10 2x10-4 4.3x10-6

Zn65 8.3x10-4 2.6x10-11 5x10-6 5.2x10-6

Zn69m 2.0x10-3 6.3x10-11 6x10-5 1.1x10-6

Np239 2.0x10-2 6.3x10-10 2x10-5 3.2x10-5

Br83 1.0x10-5 3.1x10-13 9x10-4 3.4x10-10

Sr89 4.1x10-4 1.3x10-11 8x10-6 1.6x10-6

Sr90 3.0x10-5 9.4x10-13 5x10-7 1.9x10-6

Sr91 2.3x10-3 7.2x10-11 2x10-5 3.6x10-6

Y91 2.5x10-4 7.8x10-12 8x10-6 9.8x10-7

Sr92 8.0x10-5 2.5x10-12 4x10-5 6.3x10-8

Y92 8.2x10-4 2.6x10-11 4x10-5 6.5x10-7

Y93 2.5x10-3 7.8x10-11 2x10-5 3.9x10-6

Zr95 3.0x10-5 9.4x10-13 2x10-5 4.7x10-8

Nb95 3.0x10-5 9.4x10-13 3x10-5 3.1x10-8

Mo99 6.1x10-3 1.9x10-10 2x10-5 9.5x10-6

Tc99m 9.2x10-3 2.9x10-10 1x10-3 2.9x10-7

Ru103 8.0x10-5 2.5x10-12 3x10-5 8.3x10-8

Ru105 1.4x10-4 4.4x10-12 7x10-5 6.3x10-8

Ru106 1.0x10-5 3.1x10-13 3x10-6 1.0x10-7

Te129m 1.6x10-4 5.0x10-12 7x10-6 7.1x10-7

Te131m 2.1x10-4 6.6x10-12 8x10-6 8.3x10-7

I131 1.9x10-2 6.0x10-10 1x10-6 6.0x10-4

Te132 3.0x10-5 9.4x10-13 9x10-6 1.0x10-7

I132 1.1x10-4 3.4x10-12 1x10-4 3.4x10-8

I133 3.1x10-2 9.7x10-10 7x10-6 1.4x10-4

Cs134 6.2x10-4 1.9x10-11 9x10-7 2.1x10-5

I135 4.7x10-3 1.5x10-10 3x10-5 5.0x10-6

Cs136 3.9x10-4 1.2x10-11 6x10-6 2.0x10-6

Cs137 1.5x10-3 4.7x10-11 1x10-6 4.7x10-5

Ba140 1.6x10-3 5.0x10-11 8x10-6 6.3x10-6

Ce141 1.3x10-4 4.1x10-12 3x10-5 1.4x10-7

Ce143 7.0x10-5 2.2x10-12 2x10-5 1.1x10-7

Pr143 1.6x10-4 5.0x10-12 2x10-5 2.5x10-7

Ce144 1.0x10-5 3.1x10-13 3x10-6 1.0x10-7

Nd147 1.0x10-5 3.1x10-13 2x10-5 1.6x10-8

All others 4.0x10-5 1.3x10-12 1x10-6 1.3x10-6

Tritium 4.0 1.3x10-7 1x10-3 1.3x10-4

NOTE: Data contained in this table is based on original 100% power operation, and is retained as historical

information.


Vermont Yankee


Radwaste Area – Plan View

Figure 4.5.2-8


4.0-35 of 53

4.6 SOLID WASTE MANAGEMENT

4.6.1 Solid Radwaste System

4.6.1.1 Objective

The Solid Radwaste System collects and processes radioactive solid wastes for

possible temporary on-site storage and off-site shipment for permanent disposal.


The Solid Radwaste System shall be designed to package radioactive solid wastes

for ultimate off-site shipment for disposal in accordance with applicable

published regulations.

4.6.1.3 Description

4.6.1.3.1 General

The Solid Radwaste System is a contiguous part of the Liquid Radwaste System and

is an integral part of the Radwaste Building. The system processes wet and dry

solid wastes. Because of physical differences and differences in radioactivity

or contamination levels, various methods are employed for processing and

packaging the solid radwaste. Wet solid wastes are packaged in appropriate

liners or high integrity containers for transportation within licensed shipping

casks. Dry active waste is collected in general design packages for shipment to

a licensed disposal site or a licensed processing facility for volume reduction.

Each type of waste is kept segregated to reduce shielding requirements for

storage.

Table 4.6.1 shows a history of both the wet and dry waste volumes and activity

levels that have been processed for off-site disposal. Subsequent to 1992, this

data is contained in the Radioactive Effluent Release Report.

4.6.1.3.2 Wet Wastes

Wet wastes consist of spent demineralizer resins and filter sludge. These are

pumped from the phase separators or waste sludge tanks as a slurry to disposable

liners preplaced within the licensed transportation casks. The slurry is then

dewatered from within the liner using a remote dewatering system located in the

Cask Room. The Dewatering System is kept in continuous operation as long as the

cask liner is being filled. When the cask liner is full, a high-level trip

recirculates the resin slurry to either the waste collector tank or to one of

the condensate phase separators.


4.0-36 of 53

The Dewatering System level instruments indicate in the Radwaste Building

Control Room.

The Dewatering System is accessible for cleaning and maintenance when not being

operated. The Dewatering System and its associated controls are arranged for

remote operation, which is manually initiated.

When feed to the Dewatering System is stopped, the feed piping is flushed in

accordance with plant procedures. External water connections are provided for

cleaning and decontamination.

The radioactive wet wastes are transported in licensed steel/lead casks. The

casks contain disposable steel liners or high integrity containers. The casks

are placed on trolleys and rolled on tracks below the Dewatering System fill

head. The solid wastes are processed through the fill head into the cask liner.

After filling, the liner is closed and the cask is rolled to a decontamination

area in the Radwaste Building where the cask is wiped or washed down to remove

surface contamination. The cask is lifted to a truck for transportation to the

on-site waste storage area or off-site to a waste disposal site. Design and use

of the cask are in accordance with 10CFR71 and 49CFR170-178 regulations of the

Department of Transportation. All resin shipments are via sole-use vehicles.

There are associated high and high-high level alarms which initiate the

following:

1. High level - reposition the three-way V20-422 valve to recirculate resin

slurry.

2. High-high level - cessation of feed.

Spent resins from the various filter systems are flushed to the Radwaste

Processing System and normally combined for dewatering through the Dewatering

System. The moisture content of the processed spent resins is less than 1% by

weight.

The principal gamma-emitting radionuclides normally found in the spent resins

include Manganese-54, Cobalt-58 and 60, Cesium-134 and 137, and Zinc-65. The

volume of spent resin and filter sludge is provided in the yearly Radioactive

Effluent Release Report.


4.0-37 of 53

4.6.1.3.3 Dry Wastes

Dry wastes consist of air filters, miscellaneous paper, rags, shoe covers, etc.,

from contaminated areas; contaminated clothing, tools, and equipment parts,

which cannot be effectively decontaminated; solid laboratory wastes; used

reactor equipment such as poison curtains, spent control rod blades, fuel

channels and in-core ion chambers; and large pieces of equipment.

The disposition of a particular item of waste is determined by its radiation

level and type, and the availability of disposal space. Because of high

activation and contamination level, used reactor equipment is stored in the fuel

storage pool for sufficient time to obtain optimum radioactive decay before

removal and final disposal. Most solid radwaste such as contaminated clothing,

rags, and paper can be handled manually because of low radioactivity or

contamination levels.

Dry Active Waste (DAW) is collected into shipping containers to be sent to an

off-site disposal site or an off-site waste processor for volume reduction.

Table 4.6.1 indicates the volume of compacted dry waste that has been shipped

for disposal between 1985 and 1992. This includes material sent directly from

Vermont Yankee and from various vendors after processing. The dry compacted

waste comprises about 65% of volume of total waste during the time period. The

volume for years subsequent to 1992 is contained in the Radioactive Effluent

Release Report.

The principal radionuclides in the dry active waste are Cesium-134, Cesium-137,

Cobalt-60, Iron-55, Manganese-54, and Zinc-65. Other nuclides which are

generally detected in the waste include Chromium-51, Cobalt-58,

Barium-Lanthanum-140, Cerium-141, Iron-59, Antomony-124, and Zirconium-95. The

ratios of these isotopes vary. Samples are drawn periodically to determine

current ratios and identify trends.


The Solid Radwaste System is normally operated on a regular basis thereby

demonstrating functionality without special testing.


4.0-38 of 53

TABLE 4.6.1.1

Solid Radwaste Annual Disposal History

Spent Resin and Filter Sludge Dry Processed Trash

Year Volume (ft3)

Activity (Ci)

Volume (ft3)

Activity (Ci)

1985 3,383 254 9,940 8.9

1986 1,836 196 9,018 16.6

1987 2,892 287 4,968 12.2

1988 2,655 417 3,467 7.7

1989* 171 2 0 0.0

1990* 0 0 0 0.0

1991 7,937 1,568 7,872 40.2

1992 2,391 476 3,238 51.7

Data for subsequent years is contained in the Radioactive Effluent Release Report pursuant to Technical Specifications.

* Vermont Yankee was denied access to disposal facilities.


4.0-39 of 53

4.7 EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING

4.7.1 Process Radiation Monitoring Instrumentation

A number of radiation monitors and monitoring systems are provided on process

liquid and ventilation lines that may serve as discharge routes for radioactive

materials. The monitors include the following:

Plant Stack Radiation Monitoring System

Process Liquid Radiation Monitoring System

Reactor Building Ventilation Radiation Monitoring System

These systems are described individually in the following paragraphs.

4.7.1.1 Plant Stack Radiation Monitoring System

4.7.1.1.1 Objective

The objective of the Plant Stack Radiation Monitoring System is to

representatively sample, monitor, indicate, and record the radioactivity level

of the station effluent gases being discharged from the plant stack and to alert

personnel in the event radiation levels approach or exceed pre-established

limits.


1. The Plant Stack Radiation Monitoring System shall provide a clear

indication to operations personnel of the current release level of

radioactive materials to the environs.

2. The Plant Stack Radiation Monitoring System shall record the rate of

release of radioactive materials to the environs so that determination of

the total amounts of activity release is possible.


The Plant Stack Radiation Monitoring System is shown on Drawing 5920-3994, and

specifications are given in Table 4.7.1.3. The system consists of two (2)

radiation monitors (Stack Gas I and Stack Gas II).

The primary channel provides for the continuous monitoring of radioactive gas in

the plant stack effluent. It also provides filter media to be analyzed in the

plant laboratory by gamma spectroscopy to evaluate long-lived isotopic

composition of particulates in plant stack effluents.


4.0-40 of 53

The primary stack monitoring channel consists of four (4) sampling chambers and

two (2) radiation monitors (Stack Gas I and II). The monitors observe the

radio-gas activity, and composites of long-lived particulates, iodine, and

tritium can be collected for laboratory analysis.

The sample flow is withdrawn from the plant stack through an isokinetic sample

probe located at elevation 464'-0", approximately 217 feet above the point where

the gases enter the stack. The sample train is branched prior to the point of

measurement. Branch I consists of one I-131 charcoal cartridge filter and one

radio-gas monitor with associated 8 cfm air pump and flow indicator. Branch II

is a duplicate of Branch I with the additional capability to sample gaseous

tritium. The fixed filters and tritium samplers can be changed on a routine

schedule. The plant radiochemistry laboratory analyzes filter media by gamma

spectroscopy to evaluate long-lived isotopic particulate and I-131 composition.

The tritium samplers are analyzed by liquid scintillation spectrometry.

Remote controls for pump motors are located in the station Main Control Room.

The sample flow is directed back to the stack at the completion of the

monitoring process.

All other monitoring equipment is located in an enclosure at the base of the

plant stack at grade level (elevation 250'-0"). Facilities for the collection

of air particulates and radio-gas grab samples are provided at elevation 462',

several feet downstream of the isokinetic sample probe. Facilities are also

available for sampling prior to the monitoring system at the base level at the

stack.

The monitors in the primary channel will indicate and alarm in the station Main

Control Room; no control action is provided by this system. Each monitor is

equipped with a sensor, a power supply, a logarithmic rate meter, and a trip

unit. The readout of each normal range monitor is continuously recorded in the

station Main Control Room.

Each trip unit has an adjustable trip and also signals loss of high voltage

power supply, low flow, and loss of input signal.


4.0-41 of 53


Each monitor is inspected according to surveillance procedures and is tested in

accordance with the Off-Site Dose Calculation Manual. Stack Gas I and II are

calibrated and functionally tested per the Off-Site Dose Calculation Manual.

4.7.1.2 Process Liquid Radiation Monitoring System

4.7.1.2.1 Objective

On process streams that normally discharge to the environs, process liquid

radiation monitors are provided to indicate when pre-established limits for the

normal release of radioactive material to the environs are exceeded.

On process streams that do not discharge to the environs, process liquid

monitors are provided to indicate process system malfunctions by detecting the

accumulation of radioactive material in a normally uncontaminated system.


Process liquid radiation monitors located in streams that normally discharge to

the environs shall provide a clear indication whenever the radioactivity level

in the stream reaches or exceeds pre-established limits for the discharge of

radioactive material to the environs.

Process liquid radiation monitors located in streams that do not discharge to

the environs shall provide a clear indication whenever the radioactivity level

in the stream reaches or exceeds a pre-established limit.


The processes being monitored are given in Table 4.7.1.1 and monitor the

discharge from the Liquid Radwaste and service water systems. Instrumentation

is connected to the ±24 V dc system.

Each channel has a scintillation detector, a radiation monitor, and strip chart

recorder. A representative sample may be continuously extracted from either of

two possible points of discharge and monitored for radioactivity. A radwaste

system recorder is located in the Radwaste Building Control Room. All monitors

and the other recorders are located in the Main Control Room.


4.0-42 of 53

Each channel has an upscale trip to indicate high radiation level and one

downscale trip to indicate instrument trouble. The trips give an alarm but no

control action.

The Liquid Radwaste System provides for collection of waste liquids through

various drainage systems. Because of high conductivity, some of the waste

liquids may not be economically purified by demineralization. Consequently,

some liquid containing radioactivity may eventually be discharged from the

system. The process liquid monitoring channel on the Liquid Radwaste System

discharge indicates discharge radiation levels.

The Service Water System serves as the heat sink for the Standby Fuel Pool

Cooling System. The water circulated through the heat exchangers by the Standby

Fuel Pool Cooling System will be spent fuel pool water, which may have a

significant activity level. Changes in the normal radiation level in the

service water discharge could indicate leakage in the Standby Fuel Pool Cooling

heat exchangers.

The environmental and power supply design conditions are given in Table 4.7.1.2.

The process liquid radiation monitors for radwaste and service water discharges

have radiation detection and monitoring characteristics sufficient to inform

facility personnel whenever radiation levels in the discharges rise above preset

limits.


All alarm trip circuits can be tested by using test signals or portable gamma

sources.

Surveillances are performed as required by the Off-Site Dose Calculation Manual.

4.7.1.3 Reactor Building Ventilation Radiation Monitoring System

4.7.1.3.1 Objective

The objective of the Reactor Building Ventilation Radiation Monitoring System is

to indicate whenever abnormal amounts of radioactive material exist in the

Reactor Building ventilation exhaust.


4.0-43 of 53


The Reactor Building Ventilation Radiation Monitoring System shall provide a

clear indication to facility personnel whenever abnormal amounts of

radioactivity exist in the Reactor Building ventilation exhaust.


The Reactor Building Ventilation Radiation Monitoring System is shown on Drawing

5920-00526, and characteristics are given in Table 4.7.1.1. The system consists

of two sets of exhaust system monitors with one set of detectors located in the

refuel floor zone at one half the distance between the centerline of the reactor

vessel and centerline of the fuel pool, near the wall and 10 feet above the

refuel floor. One detector is located on one side of the refuel pool and the

other on the opposite side. The other set of detectors is located in contact

with the Reactor Building exhaust duct, upstream of the exhaust ventilation

isolation valve on elevation 280 of the Reactor Building.

Each set includes two individual channels. Each channel includes a

Geiger-Muller type detector and a combined indicator and trip unit. Both

channels share a two-pen strip chart recorder. All equipment is located in the

Main Control Room except the detectors.

Power for this system is from 120 V ac buses.


4.0-44 of 53

TABLE 4.7.1.1 PROCESS RADIATION MONITORING SYSTEMS CHARACTERISTICS

Monitoring System

Instrument Range (1)

Instrument Scale

Upscale Trips Per Channel

Downscale Trips Per Channel

Liquid Process (17-351)

10-1 to 106 counts per second (2)

7 Decade Log 1 1

Reactor Building Ventilation Exhaust (17-452A, B)

0.1 mR/hr to 1 R/hr

4 Decade Log

Reactor Building Refuel Floor (17-453A, B)

1 to 104 hr

mR

4 Decade Log

(1) Range of measurements is dependent on items such as the source geometry, background radiation,

shielding, energy levels, and method of sampling. (2) Readout is dependent upon the pulse height discriminator setting.


4.0-45 of 53

TABLE 4.7.1.2

PROCESS RADIATION MONITORING SYSTEM ENVIRONMENTAL AND

POWER SUPPLY DESIGN CONDITIONS

Sensor Location Main Control Room

Parameter Design

Requirements Range Design

Requirements Range

Temperature 25°C 0°C to 60°C 25°C 5° to +50°C

Relative Humidity

50% 20 to 98% 50% 20 to 90%

Power, AC 115 V 60 Hz

±10% ±5%

115 V 60 Hz

±10% ±5%

Power, DC +24 V dc -24 V dc

+22 to +29 V dc -22 to -29 V dc

+24 V dc -24 V dc

+22 to +29 V dc -22 to -29 V dc


4.0-46 of 53

TABLE 4.7.1.3

PLANT STACK RADIATION MONITORING SYSTEM CHARACTERISTICS

Monitor Type Instrument Range Instrument Scale

Type Detector Remarks

Radio-Gas Monitors I & II (17-156, 157)

10 to 107 cpm 6 Decade Digital

Beta Scintillation


4.0-47 of 53

4.7.2 Area Radiation Monitoring System

4.7.2.1 Objectives

The objectives of the Area Radiation Monitoring System are:

1. To warn of abnormal gamma radiation levels in areas where radioactive material

may be present, stored, handled or inadvertently introduced.

2. To warn facility personnel whenever abnormal concentrations of airborne

radioactive materials exist in the Reactor Building.


1. The Area Radiation Monitoring System shall provide facility personnel with a

record and an indication of gamma radiation levels at selected locations within

the various facility buildings and radioactive airborne concentrations within the

Reactor Building.

2. The Area Radiation Monitoring System shall provide local alarms where it is

necessary to warn personnel of substantial immediate changes in radiation levels.

4.7.2.3 Description

4.7.2.3.1 Monitors

1. Area Gamma Radiation Monitoring System

The Area Gamma Radiation Monitoring System is shown as a functional block

diagram on Drawing 5920-430, Sh.1. A typical channel consists of a combined

indicator and trip unit, a shared power supply, and computer points for selected

monitors. Some channels have, in addition, a local audio alarm auxiliary unit.

Each monitor has an upscale trip that indicates high radiation and a downscale

trip that may indicate instrument trouble. These trips sound alarms but cause

no control action. The system is powered from the 120 V ac instrument bus. The

trip circuits are set so that loss of power causes an alarm. The environmental

and power supply design conditions are given in Table 4.7.2.1.

2. Area Airborne Radiation Monitoring System

The Reactor Building Area Airborne Radiation Monitoring System is shown as a

functional block diagram in Figure 4.7.2-2. Applicable specifications are

provided in Table 4.7.2.3. The Reactor Building Area Airborne Radiation

Monitoring System is a two (2) channel system employing a continuous air

particulate monitor and an off-line radiogas monitor located within a single

enclosure.


4.0-48 of 53

The air particulate monitor consists of a continuous moving tape sampler with a

beta scintillation detector to provide for the continuous monitoring of air

particulates. Each off-line gas monitor contains a beta scintillation probe for

measurement of radiogas.

The sample for the Reactor Building Area Airborne Monitoring System is withdrawn

from the Reactor Building Exhaust Ventilation System through an isokinetic

sample probe located in the exhaust duct. An air pump with flow indicator and a

low flow alarm is used to obtain the required sample flow. The sample is

directed back to the ventilation duct at the completion of the monitoring

process.

The air particulate and radiogas monitor will indicate an alarm in the station

Main Control Room; no control action is provided. The monitor is equipped with

a sensor, a power supply, a logarithmic ratemeter and trip unit. Trip units

have adjustable trips which may be verified by the use of a remotely operated

radioactive check source mechanism; trip units also signal loss of high voltage

power supply, low sample flow and loss of signal input. The monitor readout is

continuously recorded on a recorder located in the station Main Control Room.

4.7.2.3.2 Locations

Work areas where gamma monitors will be located are tabulated in Table 4.7.2.2.

Annunciation and indication are provided in the Main Control Room.


Area Gamma Radiation Monitoring System

An internal trip test circuit, adjustable over the full range of the trip circuit,

is provided. The test signal is fed into the indicator and trip unit input so that

a meter reading is provided in addition to a real trip. All trip circuits are of

the latching type and must be manually reset at the front panel.

A portable calibration unit is also provided. This is a test unit designed for use

in the adjustment procedure for the area radiation monitor sensor and converter

unit. It provides five gamma radiation levels for calibration purposes. A cavity

in the calibration unit is designed to receive the sensor and converter unit.

Located on the back wall of the cylindrical lower half of the cavity is a window

through which radiation from the source emanates. A chart on each unit indicates

the radiation levels available from the unit for the various control settings.


4.0-49 of 53

Reactor Building Airborne Radiation Monitoring System

The Monitoring System includes a built-in check source for each detector. The check

source is operated from the Main Control Room. Alarm circuits can be tested by

using the built-in check source.


4.0-50 of 53

TABLE 4.7.2.1 AREA RADIATION MONITORING SYSTEM ENVIRONMENTAL AND POWER SUPPLY DESIGN CONDITIONS Sensor Location Control Room Design Design Parameter Requirements Range Requirements Range

Temperature 25°C 0° to 60°C 25°C 5° to 50°C Relative 50% 20 to 100% 50% 20 to 90% Humidity

Power 120 V ±10% 120 V ±10% 60 Hz ±5% 60 Hz ±5%


4.0-51 of 53

TABLE 4.7.2.2 LOCATIONS OF AREA RADIATION MONITORS Station Number Location 1 Reactor Building Supp. Chamber Catwalk 232' 2 Reactor Building North Pers. Access 252' 3 Reactor Building South R.R. Access 252' 4 Reactor Building TIP Room 252' 5 Reactor Building Reactor Pers. Acc. Hatch 252' 6 Reactor Building Elev. Ent. 280' 7 Reactor Building CRD Repair 252' 8 Reactor Building Elev. Ent. 303' 9 Reactor Building RCUW Sample Sink 303' 10 Reactor Building Elev. Ent. 318' 11 Reactor Building RCUW Panel 318' 12 Reactor Building Elev. Ent. 345' 14 Reactor Building West Refuel 345' 15 Reactor Building Spent Fuel Pool 345' 16 Reactor Building New Fuel Vault 345' 17 Radwaste Recirc. Pump Room 252' 18 Radwaste R.W. Oper. Area 252' 19 Radwaste Pump and Tank Area 230'


TABLE 4.7.2.3

REACTOR BUILDING AREA AIRBORNE RADIATION MONITORING SYSTEM

Monitor Type

Instrument Range

Instrument Scale

Type Detector

Remarks

Air Particulate Monitor

10 to 107 cpm 6 Decade Log Digital

Beta Scintillation

Tape transport mechanism - tape speed selectable .5, 1, 2, and 10 in/hr; alarm for broken tape and low flow

Radio-Gas Monitor

10 to 107 cpm 6 Decade Log Digital

Beta Scintillation


Vermont Yankee


Reactor Building Area Airborne

Radiation Monitoring System

Figure 4.7.2-2


CONDUCT OF OPERATIONS TABLE OF CONTENTS Section Title Page

5.1 ORGANIZATION AND RESPONSIBILITY ..................................... 2

5.2 TRAINING .............................................................. 2

5.2.1 Program Description (General) ............................... 2

5.2.2 General Employee Training ................................... 2

5.2.2.1 Access to Plant ................................. 2

5.2.3 Fire Brigade Training ....................................... 2

5.2.4 Operations Training ......................................... 2

5.2.5 Craft, Technician, and Engineering Staff Position (ESP) Training ..................................... 3

5.2.6 Training Records ............................................ 3

5.2.7 Training Program Approval and Evaluation .................... 3

5.2.8 Responsibility .............................................. 3

5.3 EMERGENCY PLAN ........................................................ 4

5.4 QUALITY ASSURANCE PROGRAM ............................................. 4

5.4.1 Scope ....................................................... 4

5.4.2 Responsibilities ............................................ 4

5.4.3 Implementation .............................................. 4

5.4.4 Management Evaluation ....................................... 5

5.5 REVIEW AND AUDIT OF OPERATIONS ........................................ 5

5.5.1 General ..................................................... 5

5.5.2 Independent Safety Review ................................... 5

5.5.3 Safety Review Committee ..................................... 5

5.6 TECHNICAL REQUIREMENTS MANUAL ......................................... 5


5.1 ORGANIZATION AND RESPONSIBILITY

The Vermont Yankee Nuclear Power Station organization, including the

responsibilities and duties of staff personnel, are detailed in the Vermont

Yankee Quality Assurance Program Manual.

5.2 TRAINING

5.2.1 Program Description (General)

The objective of the Training Program is to provide qualified personnel to

operate and maintain the permanently defueled facility in a safe manner,

including the storage and handling of irradiated fuel. All operations, craft,

technician, engineering staff, and general employee training requirements are

described in position-specific program descriptions or procedures. Training

programs are implemented and maintained using a Systems Approach to Training

(SAT), in accordance with 10CFR50.120, Training and Qualification of Nuclear

Power Plant Personnel, and ANSI/ANS 3.1, 1978, Selection, Qualification, and

Training of Personnel for Nuclear Power Plants.

5.2.2 General Employee Training

All persons permanently employed at the facility shall be trained in the

applicable following areas commensurate with their job duties:

1. Chemical and Hazardous Material Program

2. Radiological Health and Safety Program

3. Site Emergency Plans

4. Industrial Safety

5. Fire Protection

6. Security

7. Quality Assurance

8. Fitness for Duty

5.2.2.1 Access to Plant

Requirements to gain access to the facility protected area, including training

requirements, are contained in applicable facility procedures.

5.2.3 Fire Brigade Training

Fire brigade training for appropriate facility personnel meets the

requirements of NFPA 600, Standard on Industrial Fire Brigades.

5.2.4 Operations Training

The initial and continuing training programs for the personnel performing

operator functions, including certified fuel handler and shift manager, are

based on a SAT.


5.2.5 Craft, Technician, and Engineering Staff Position (ESP) Training

The initial and continuing training programs for the instrument control

technician, chemistry technician, radiation protection technician, plant

mechanic (electrical and mechanical maintenance), and engineering staff

positions are based on a SAT.

5.2.6 Training Records

Records of employee and contractor participation in, and completion of,

training activities are maintained in accordance with the VY records retention

policy.

5.2.7 Training Program Approval and Evaluation

The Vermont Yankee position-specific training program descriptions are

approved by appropriate Training Department and facility management, as

specified in applicable facility procedures. This ensures that the content and

the intent of the training programs provide the necessary training for

personnel associated with the safe storage and handling of irradiated fuel and

management of radioactive waste. Training processes are controlled and

maintained in accordance with applicable Training directives.

The effectiveness of training programs is evaluated by the performance of

employees in carrying out their assigned duties, by performance on facility

evaluations, and the employment of various types of feedback mechanisms. The

results of the evaluations are maintained in accordance with applicable

records retention requirements.

5.2.8 Responsibility

As delegated by the responsible manager, the Superintendent, Training is

responsible for the conduct and administration of the specified training

activities, including:

1. Initial and continuing training programs for the non-certified operator,

certified fuel handler and shift manager.

2. Fire brigade training.

3. Initial and continuing training programs for instrumentation and control,

maintenance, and engineering staff positions.

4. Initial and continuing training programs for chemistry and radiation

protection positions.

5. General employee training.


5.3 EMERGENCY PLAN

The emergency plan for the Vermont Yankee Nuclear Power station was originally

issued in accordance with NRC's regulations on April 1, 1981. Any information

regarding this plan should be obtained from the most current revision to that

document.

5.4 QUALITY ASSURANCE PROGRAM

5.4.1 Scope

This section establishes the criteria to be applied to systems requiring

Quality Assurance which prevent or mitigate the consequences of postulated

accidents which could cause undue risk to the health and safety of the public.

The structures, systems, components, and other items requiring quality

assurance are listed in the Vermont Yankee Safety Classification Program.

5.4.2 Responsibilities

1. Compliance with the requirements of the VY Quality Assurance Program Manual

(VYQAPM) based on the criteria of Title 10 of the Code of Federal

Regulations, Part 50, Appendix B, and as committed to within the VYQAPM,

shall be the responsibility of all personnel involved with activities

affecting operational safety. Vermont Yankee shall cross reference the

applicable criteria of 10CFR50 Appendix B in procedures that implement the

VYQAPM. The performance of quality-related activities shall be

accomplished with specified equipment under suitable environmental

conditions.

2. Individuals having direct responsibilities for establishment/distribution

control/implementation of the VYQAPM are delineated in the “Organization,"

section of the VYQAPM.

5.4.3 Implementation

Establishment of an effective Operational Quality Assurance Program is assured

through consideration of, and conformance with, the Regulatory Position in the

Regulatory Guides listed the VYQAPM. Implementation of this program is assured

through Quality Assurance procedures, derived from Quality Assurance policies,

goals, and objectives.


5.4.4 Management Evaluation

The Safety Review Committee (SRC) reports to the executive responsible for

oversight on those areas of responsibility specified in the Quality Assurance

Program Manual. The SRC conducts its function in accordance with a procedure

approved by the executive responsible for oversight. The SRC independently

monitors applicable programs and provides management with evaluations and

assessments related to the effectiveness of the nuclear program.

5.5 REVIEW AND AUDIT OF OPERATIONS

5.5.1 General

Two review bodies have been established to review operating procedures,

evaluate and process changes and assure compliance and safe operation.

5.5.2 Independent Safety Review

The responsibilities and authorities of the Independent Safety Review are

described in an approved Quality Assurance Program Manual implementing

procedure.

5.5.3 Safety Review Committee

An independent safety review of activities affecting nuclear safety is

performed by the Safety Review Committee in accordance with an approved

Quality Assurance Program Manual implementing procedure.

5.6 TECHNICAL REQUIREMENTS MANUAL

Requirements pertinent to the permanently defueled state which have been

relocated out of Technical Specifications, as well as any other items deemed

appropriate by facility management, which do not meet the Technical

Specification screening criteria provided in 10CFR50.36(c)(2)(ii), are located

in the Technical Requirements Manual (TRM). Changes to the TRM are evaluated

per the requirements of 10CFR50.59.


SAFETY ANALYSIS

TABLE OF CONTENTS

Section Title Page

6.1 INTRODUCTION .......................................................... 4

6.2 ACCEPTANCE CRITERIA ................................................... 5

6.2.1 DBA Acceptance Criteria ...................................... 5

6.2.2 Site Event Acceptance Criteria ............................... 6

6.3 ACCIDENTS EVALUATED ................................................... 7

6.3.1 Fuel Handling Accident ....................................... 7

6.3.1.1 Analytical Methodology .......................... 7

6.3.1.2 Assembly Drop in SFP with Open Containment Scenario ............................ 7

6.3.1.3 Assembly Drop in SFP with Closed Containment Scenario ............................ 9

6.3.1.4 Software ........................................ 9

6.3.1.5 Assumptions ..................................... 9

6.3.1.6 Inputs ......................................... 10

6.3.1.7 Impact of Water Depth on Iodine Decontamination Factor ......................... 10

6.3.1.8 Fuel Damage from Assembly Drop onto SFP Fuel Racks ..................................... 12

6.3.1.9 Radiological Consequences/Results 13

6.4 SITE EVENTS EVALUATED ................................................ 25

6.4.1 High Integrity Container (HIC) Drop Event ................... 25

6.4.1.1 Analytical Methodology ......................... 25

6.4.1.2 Assumptions .................................... 25

6.4.1.3 Inputs ......................................... 26

6.4.1.4 Radiological Consequences/Results .............. 27

6.5 REFERENCES ........................................................... 29

6.6 APPENDICES ........................................................... 32


SAFETY ANALYSIS

LIST OF TABLES Table No. Title 6.3.1 FHA Scenarios Analyzed

6.3.2 Input Conditions for FHA

6.3.3 Undecayed Core Inventory for Radionuclides Important in the

Radiological Evaluation of DBAs

6.3.4 Undecayed Gap Activity Available for Release from Fuel Assembly

Drop in the SFP

6.3.5 Typical Iodine Decontamination Factors and Iodine Speciation vs

Water Depth above Dropped Assembly

6.3.6 Atmospheric Dispersion Factors for the Postulated FHA

6.3.7 EAB TEDE Dose vs Long Decay Time

6.4.1 HIC Drop Source Term Release Activity


SAFETY ANALYSIS

LIST OF FIGURES Reference Figure No. Drawing No. Title 6.3-1 VY FHA – EAB TEDE Dose vs Decay Time

6.3-2 VY FHA – MCR TEDE Dose vs Decay Time

A-1 Dimensions of SFP, Fuel Rack and Fuel

Handling Equipment


6.1 Introduction

In January of 2015, the licensee certified to the NRC that Vermont Yankee had

both permanently ceased operations (final shutdown 12/29/14) and that all fuel

had been removed from the reactor vessel and placed in the spent fuel pool

(SFP) (Reference 6.5-1). Since Vermont Yankee will never again enter any

operational mode, reactor related accidents are no longer a possibility.

This chapter discusses: (a) a postulated fuel handling accident (FHA)

associated with fuel movement until the fuel has been transferred to the

Independent Spent Fuel Storage Installation (ISFSI); and (b) the postulated

drop of a high integrity container (HIC) containing radioactive resins

Bounding conditions, conservatism in equipment design, conformance to high

standards of material and construction, the control of loads and strict

administrative controls over facility operations all serve to assure the

integrity of the fuel while in the spent fuel pool and during fuel transfer to

the Independent Spent Fuel Storage Installation (ISFSI).

Accidents involving fuel and the Holtec International HI-STORM system storage

casks are discussed in the HI-STORM FSAR (Reference 6.5-2).

For site events, a drop and fire of a High Integrity Container (HIC) containing

resins was evaluated.

New hazards, new initiators or new accidents that may challenge offsite

guideline exposures, may be introduced as a result of certain decommissioning

activities. These issues will be evaluated when the scope and type of

decommissioning activities are finalized.


6.2 Acceptance Criteria

6.2.1 DBA Acceptance Criteria

The radiological release acceptance criteria associated with the Alternative

Source Term (AST) methodology are identified in 10CFR50.67 (Reference 6.5-3)

and dose levels are not to exceed:

Exclusion Area Boundary (EAB): 25 rem TEDE

Low Population Zone (LPZ): 25 rem TEDE

Control Room (CR): 5.0 rem TEDE

These criteria, however, are for evaluating potential reactor accidents of

exceedingly low occurrence probability and low risk of public exposure to

radiation. For events with higher probability of occurrence, such as a FHA,

the acceptance criteria for the offsite receptors are more stringent, while

that for the control room operators remains the same. The applicable AST

criteria for an FHA are identified in Regulatory Guide 1.183 (Reference 6.5-4)

and 10CFR50.67 and dose levels are not to exceed:

Exclusion Area Boundary (EAB): 6.3 rem TEDE

Low Population Zone (LPZ): 6.3 rem TEDE

Control Room (CR): 5.0 rem TEDE

The EAB and LPZ criteria are referred to as being "well within" the regulatory

limits (i.e., 25%).

The LPZ doses are bounded by the dose at the EAB, since the LPZ is farther

away.

Additional acceptance criteria are as follows:

• The required decay time that would preclude Evacuation as a protective

action following an FHA. The limit for such an action is 1 rem TEDE (EPA

400-R-92-001, Table 2-1 (Reference 6.5-5)).

• The required decay time that would reduce the TEDE dose from gaseous

releases of iodines and particulates to unrestricted areas to "well

within" (i.e., 25 %) of the 10CFR50 Appendix I (Reference 6.5-6) annual

dose limit of 15 mrem (or 3.75 mrem).


The latter acceptance criterion was selected as a suitable basis no longer

requiring the Standby Gas Treatment System (SGTS). The "well within" limit was

selected so as to accommodate other potential releases from the plant. It is

noted that, by definition, this criterion excludes the noble gas dose. There

exists a separate criterion applicable to the nobles, which however is of no

interest in the present application since the noble gas release is not impacted

by the SGTS filtration.

6.2.2 Site Event Acceptance Criteria

The HIC drop acceptance criteria are based on 10% of the 10CFR100 dose

acceptance criteria.

10CFR100 Acceptance Criteria (1)

(rem)

10% of 10CFR100 Acceptance Criteria

(rem)

EAB

and

LPZ

25 (whole body)

2.5 (whole body)

300 (thyroid)

30 (thyroid, critical organ)

(1) EAB and LPZ dose acceptance criteria from 10CFR100.11


6.3 Accidents Evaluated

6.3.1 Fuel Handling Accident

Two bounding scenarios of the FHA are considered and the scenario objectives

are summarized in Table 6.3.1.

The first scenario is a drop in the SFP with an open containment (no filtration

by SGTS) with an instantaneous release (ground level) to determine the required

decay time prior to fuel movement that would result in the EAB dose, and main

control room (MCR) dose within 10CFR50.67 and R.G. 1.183 limits and the EAB

TEDE dose under the EPA PAG limit of 1 rem for evacuation.

The second scenario is with a drop in the SFP, with a closed containment (not

under a negative pressure and not ‘air tight’) and an elevated (stack release)

instantaneous release with and without filtration by the SGTS, to determine the

required decay time prior to fuel movement that would result in 25% of the

10CFR50, Appendix I TEDE annual dose limit of 15 mrem at the EAB. The 10CFR50,

Appendix I dose of 15 mrem from gaseous effluents coincides with maintaining

dose as low as reasonably achievable. A revision to the Technical

Specifications proposed by BVY 13-097 (Reference 6.5-7) and approved by NVY 15-

013 (Reference 6.5-8) contains the following commitment: "During fuel

handling/core alterations, ventilation system and radiation monitor

availability (as defined in NUMARC 91-06, Reference 6.5-9) will be assessed,

with respect to filtration and monitoring of releases from the fuel. The goal

of maintaining ventilation system and radiation monitor availability is to

reduce doses even further below that provided by the natural decay." This FHA

scenario is the analysis which demonstrates that SGTS operation is not required

to maintain dose as low as reasonably achievable consistent with 10CFR50,

Appendix I, thus fulfilling this VY commitment.

6.3.1.1 Analytical Methodology

The postulated FHA scenarios were based on the Alternative Source Term (AST)

Methodology in RG 1.183, Appendix B. Two main configurations of the reactor

building during fuel movement were considered, one with an open containment and

the other with a closed containment.

6.3.1.2 Assembly Drop in SFP with Open Containment Scenario

(a) The reactor operated at full power (1950 MWt) for an extended period of

time until permanent shutdown and full core offload is completed with all

fuel in the SFP.


(b) Fuel moves are in progress and an FHA takes place in the SFP with an

assembly falling onto a fuel rack at various assumed decay times after

reactor shutdown, from 1 to 19 days.

(c) The accident leads to the damage of 98 fuel rods (a bounding value from

Table 6.3.2, Item #D1), addressing both GE14 and GNF2 10x10 assemblies

removed from the final core offload. All failed rods are peak powered,

with a radial peaking factor of 1.65 (from Table 6.3.2, Item #A3).

(d) All activity within the gaps of the failed fuel rods is released to the

refueling cavity pool. The released activity corresponds to 8% of the

entire inventory of I-131 in the rods (i.e., within the fuel matrix and

gaps), 10% of the Kr-85, 5% of the remaining halogens and noble gases,

and 12% of the alkalis (Cs and Rb), from Table 6.3.2, Item #A4. The

undecayed activity released from the damaged fuel rods is presented in

Table 6.3.4; decay correction from plant shutdown to the time of the

postulated FHA is properly accounted for by the radiological software

used in the analysis (ELISA-2).

(e) All the noble gases and a fraction of the halogens (see Item (g) below)

escape from the pool and are released to the refueling level. All the

alkalis are retained by the pool. The halogen speciation above the pool

is not pertinent in this scenario since there is no pre- or post-release

filtration of radioactivity.

(f) The water depth above the dropped assembly in the SFP is 20.5 feet

(versus the 23-foot requirement in Regulatory Guide 1.183 for full credit

of the decontamination factor (DF) of 200 for iodine retention by the

pool water), leading to a corresponding decrease in the DF to 125.4 (from

Table 6.3.5).

(g) The radioactivity which becomes airborne above the SFP is instantly

released to the environment without holdup (a conservative assumption

which accommodates the 2-hr release requirement in Regulatory Guide

1.183, Appendix B, Sec. 5.3).

(h) The reactor building is open during the refueling operations, with all

post-FHA releases to the environment assumed to be at ground level, via

the RB blowout panels.

(i) Transport of the released radioactivity to the receptors of interest is

dictated by the applicable atmospheric dispersion factors in Table 6.3.6.

(j) The MCR ventilation configuration is in the normal operating mode during

the entire exposure interval (30 days), with an intake flow of 3700 cfm,

unfiltered.

(k) Breathing rates and MCR occupancy factors are as given in Table 6.3.2,

Items #F2 and #F3.

(l) The control room operator dose point is at the base of a hemispherical

cloud having a volume equal to the free air volume of the control room.

Finite-cloud correction to the submersion dose was based on the

Murphy/Campe equation in Reg. Guide 1.183 (Sec. 4.2.7).


6.3.1.3 Assembly Drop in SFP with Closed Containment Scenario

The scenario is the same as an assembly drop in the SFP with an open

containment, except with the following differences:

(a) The radionuclide list is as given in Table 6.3.4, though without the

noble gases. The acceptance criterion for no longer requiring the SGTS

filtration was selected to be 25% of the 10CFR50, Appendix I annual TEDE

dose limit of 15 mrem (or. 3.75 mrem) from gaseous releases of iodines

and particulates to unrestricted areas. As such, this criterion excludes

the noble gas dose.

(b) The reactor building is closed during fuel moves, such that all releases

to the environment would be via the main stack, with and without credit

for filtration by the SGTS system. In-transit decay and plateout were not

credited.

(c) The decay times prior to fuel movement are fairly long in this scenario

because the interest is in determining the time beyond which there will

be no dose-wise beneficial purpose for maintaining the SGTS operational

after permanent plant shutdown.

(d) The EAB dose consequences were evaluated without SGTS filtration, but

also with SGTS filtration for informational purposes. The filtration

efficiencies for the latter case are as given in Table 6.3.2, Item #D7.

(e) The MCR is of no interest in this scenario and was therefore excluded

from the analysis.

6.3.1.4 Software

Computation of the EAB and MCR radiological consequences for the postulated FHA

were based on the ELISA-2 computer code, Version 2.4 (Reference 6.5-10) for all

analyzed scenarios.

The dose conversions in ELISA-2 are from Federal Guidance Reports 11 (Reference

6.5-11) and 12 (Reference 6.5-12). Dose rates and cumulative doses are

computed for each organ, TEDE, skin and air. Of these, only the TEDE doses are

presented for comparison to the TEDE regulatory limits.

ELISA-2 was designed to handle the pre-FHA decay correction and the time-

release from the RB for all scenarios. Its built-in logic accounts for the

time-dependent generation and release of noble gases from the decay of halogens

retained by the pool water, and also from the halogens on the SGTS exhaust

filtration system when credited. These releases extend beyond the end of the

2-hr release from the RB.

6.3.1.5 Assumptions


Release Rate from Reactor Building

In accordance with RG 1.183, Appendix B, Sec. 4.1 for an FHA in the SFP, the

radioactive material that escapes the water pool is released to the environment

over a 2-hour interval. Analytically, this is conservatively accommodated by

assuming an instantaneous release to the environment, in all accident scenarios

and for all receptors (EAB and MCR). Credit for in-transit decay or plateout

was not taken.

Fuel Rod Gap Activity

All activity within the gaps of the failed fuel rods is released to the

refueling cavity pool. The released activity corresponds to 8% of the entire

inventory of I-131 in the rods (i.e., within the fuel matrix and gaps), 10% of

the Kr-85, 5% of the remaining halogens and noble gases, and 12% of the alkalis

(Cs and Rb), from Table 6.3.2 (Item #A4), Fuel Rod Gap Fractions. The

undecayed activity released from the damaged fuel rods is presented in Table

6.3.4.

MCR Finite Cloud Correction

Doses to MCR personnel due to the external gamma radiation from airborne

radioactivity within the MCR were adjusted using the Murphy/Campe finite-cloud

correction model in R.G. 1.183, Section 4.2.7.

Modeling Simplifications

There are no modeling simplifications.

Number of Failed Fuel Rods

Assumptions associated with determining the number of failed fuel rods due to a

dropped fuel assembly are provided in Section 6.6, Appendix A.

6.3.1.6 Inputs

Inputs for the analysis are identified in Table 6.3.2.

6.3.1.7 Impact of Water Depth on Iodine Decontamination Factor

According to RG 1.183, Appendix B, Section 2, Water Depth, if the water depth

above the damaged dropped assembly in a fuel handling accident is 23 feet or

greater, an overall decontamination factor of 200 may be credited for the

expected iodine retention by the pool water. As clarified in RIS 2006-004,

Item 8 (Reference 6.5-13) and in the proposed revision to RG 1.183, with an

iodine speciation consisting of 99.85% elemental (including CsI) and 0.15%

organic in the fuel-rod gaps, the overall DF of 200 is achieved when the DF for

the elemental iodines is 285, with the ensuing iodine composition in air above

the pool being 70% elemental and 30% organic.


For water depths less than 23 feet, RG 1.183 recommends the use of the Burley

model (Reference 6.5-14). According to this model, the DF for a reduced water

depth is determined through use of the following formula (Reference 6.5-14, pg

26):

DFinorg = exp {(6/db)*(keff Hb / vb)} (Eq. 1)

where

DFinorg = elemental iodine pool retention factor (or DF)

db = bubble diameter (cm)

keff = effective mass transfer coefficient (cm/sec)

Hb = bubble rise height (cm, water depth above dropped

assembly)

vb = bubble rise velocity (cm/sec)

This equation can be rewritten by combining all of the independent

variables (excluding Hb), as:

DFinorg = exp (K*Hb) (Eq. 2)

where

K = {(6/db)*(keff / vb)} (Eq. 3)

The new variable K can now be back-calculated by using the

recommended values given above for DFinorg and Hb, namely 285 and 23

ft (or 701.0 cm), respectively, and is as follows:

K = loge(285) / 701.0 = 0.008063 (Eq. 4)

The applicable DF equation for reduced SFP water then becomes:

DFinorg = exp (0.008063*Hb) (Eq. 5)

For the iodine speciation given above, namely 99.85% elemental

(including CsI) and 0.15% organic, and no organic iodine retention

by the pool water (i.e., DForg = 1), the overall (total iodine) DF

is given by:

DFtotal = [(0.9985/DFinorg) + (0.0015/1)]-1 (Eq. 6)

Proceeding further, the above-water (i.e., airborne) iodine

speciation formulas (in percent) are given by:

Sinorg = 99.85* (DFtotal / DFinorg) (Eq. 7)

and

Sorg = 0.15 * DF total (Eq. 8)


Typical DF values and iodine speciation versus SFP water depth are presented in

Table 6.3.5. The iodine speciation is of no interest since (a) there is no

filtration credit in the open containment scenario, and (b) all iodine species

were subjected to the same SGTS filtration efficiency in the closed containment

scenario.

6.3.1.8 Fuel Damage from Assembly Drop onto SFP Fuel Racks

Fuel pin damage due to a drop of a fuel assembly onto a spent fuel rack within

the SFP was evaluated. A drop in the SFP is limited to a drop distance of 3

feet, rounded up from 1.9 feet for conservatism. The General Electric Standard

Application for Reactor Fuel, GESTAR II (Reference 6.5-15) is utilized to

determine the number of damaged fuel pins resulting from the drop. This

analysis is presented in Section 6.6, Appendix A.


6.3.1.9 Radiological Consequences/Results

The radiological consequences of the FHA scenarios are shown below.

Scenario Limiting

Dose

Acceptance

Criteria

Open Containment

Ground Level

Instantaneous

Release

No SGTS

Filtration

Closed Containment

Elevated Release

(Main Stack)

Instantaneous

Release

No SGTS Filtration

Required

Decay Time to

Meet Most

Restrictive

Acceptance

Criteria

(Section 6.2.1)

15 days 50 days

Dose

TEDE

MCR 5 rem

(10CFR50.67)

< 5 rem NA

EAB 6.3 rem

(R.G. 1.183)

< 6.3 rem NA

EAB EPA PAG

1 rem

(initiation of

evacuation)

< 1 rem

NA

EAB 10CFR50

Appendix I

15 mrem

annual limit

NA 3.75 mrem

(25% of the limit)

Decay Time and Dose Details Figure 6.3-1

Figure 6.3.2

Table 6.3.7

Note that doses at the EAB bound the corresponding dose at the Low

Population Zone (LPZ), as the LPZ is farther away from the station.

The conclusion is that if there is a FHA 50 days following cessation of

power operations, the benefits of SGTS are minimal and resultant dose

without SGTS operation at the EAB is considered to be maintained as low

as reasonably achievable.


Table 6.3.1

FHA Scenarios Analyzed

Scenario Open containment Closed containment

Objective To determine the required

decay time prior to fuel

movement in the SFP that

would meet the following:

(a) 90% of the 10CFR

50.67 dose acceptance

criteria, and

(b) an EAB TEDE dose less

than the PAG limit of 1

rem for evacuation.

To provide basis for no longer

requiring the SGTS in support of a

VY TS commitment (contained within

Reference 6.5-7) with respect to

dose minimization following an FHA

Containment Building

Configuration

Open containment, with

instantaneous atmospheric

release via the blow-out

panels

(ground-level release)

Closed containment with

instantaneous atmospheric releases

via main stack (with and without

SGTS iodine and particulate

filtration)

Location of Assembly

Drop

SFP

(3 foot drop onto fuel

racks)

SFP

(3 foot drop onto fuel racks)

Water Depth above

Dropped Assembly

Credited for Iodine

Retention

20.5 feet

[DF = 125.4]

20.5 feet

[DF = 125.4]

Fuel Damage To be determined in

present calculation based

on both GE14 and GNF2

assembly drops

To be determined in present

calculation based on both GE14 and

GNF2 assembly drops

Pre FHA Decay Time

from Reactor

Shutdown

Required decay time to be

determined in present

calculation to meet the

dose consequence

objectives

Required decay time to meet 25% of

the 10 CFR 50 Appendix I TEDE

annual dose limit of 15 mrem at

the EAB from iodines and

particulates


Table 6.3.2

INPUT CONDITIONS FOR FHA

Item No. DESCRIPTION VALUE REFERENCE

FHA Source TermA1 Power level for DBA analysis

[Includes 2 % measurement uncertainty] 1950 MWt

6.5‐16 (VYC‐2299) A2 Number of assemblies in core 368

A3 Maximum allowed radial peaking factor(a) 1.65A4 Fuel rod gap fractions (AST Methodology)

I‐131 Kr‐85 Other noble gases Other halogens Alkali metals (Cs and Rb)

0.08 0.10 0.05 0.05 0.12

6.5‐4 (Reg. Guide 1.183, Table 3)

A5 Undecayed core inventory for radionuclides important in the evaluation of DBAs

Table 6.3.3 6.5‐17 (VYC‐2260, Table 4.5)

A6 Post‐shutdown decay time prior to postulated accident

Various Assumed values for sensitivity analyses

Variables for Fuel Damage Calculation for FHA in Spent Fuel Pool B1 Number of fuel rods in 10x10 assemblies

(GE14 and GNF2) 92

6.5‐18 (VYC‐2206, Sec. 1)

B2 Assembly drop height above fuel racks

Bounding value 22.74"(1.9 ft)

6.5‐19 (App. A, pg 17 of 20)

Used in analysis(c)36" (3 ft)

Conservatively assumed value

B3 Wet weight of fuel assembly and channel

GE14 569.5 lbm 6.5‐18 (VYC‐2206, pg 12 of 31)

GNF2 580.0 lbm 6.5‐19 (App. A, pg 8 of 20)

B4 Weight of mast and grapple

GE14 619 lbm 6.5‐19 (App. A, pg 9 of 20) GNF2 619 lbm

B5 Percent energy for clad deformation

GE14 51% 6.5‐15 (GESTAR II , Sec. 5.3.1 (also applied to GNF2 assemblies)

GNF2 51%

B6 10x10 rod compression failure (energy required to damage stationary fuel rods)

GE14 167 ft‐lb

6.5‐19 (App. A., page 10 of 20) GNF2 157 ft‐lb


Table 6.3.2 (Continued)


DESCRIPTION VALUE REFERENCE Atmospheric Release Resulting from Postulated FHA in Spent Fuel Pool

D1 Number of damaged fuel assemblies

GE14 97 6.5‐19 (VYC‐3187) GNF2 98

Used in analysis 98 Bounds both GE14 and GNF2 assembly types

D2 Water depth above dropped assembly (resting on top of fuel racks)

Minimum value 20.67 ft 6.5‐19 (App. A, pg 2 of 19)

Used in analysis 20.5 ft Conservative

D3 Undecayed gap inventory available for release from 98 damaged fuel rods

See Table 6.3.4

D4 Overall pool DF for given water depth

Noble gases 1 6.5‐4 (RG 1.183)

Halogens 125.4 See Table 6.3.5Alkalis Infinite

6.5‐4 (RG 1.183)

D5 Percent of damaged‐fuel rod gap activity release

100 %

D6

Reactor building configuration during refueling operations

Closed Containment w/wo SGTS

Instantaneous Stack Release

See Table 6.3.1 Open

Containment No SGTS

Ground Level Release

D7

Potential release point to the atmosphere (see Table 6.3.6 for the atmospheric dispersion factors)

Closed Containment w/wo SGTS

Stack Release See Table 6.3.1

Open Containment No SGTS

Instantaneous RB blowout panel release

6.5‐20 (VYC‐2275)

D8 SGTS filtration efficiency, all halogens and particulates

95% 6.5‐21 (VYC‐2302, Page 11 of 59)

D9 Release duration to atmosphere Instantaneous

Meets the RG 1.183 requirements


Table 6.3.2 (Continued)


DESCRIPTION VALUE REFERENCE Control Room Characteristics

E1 Control room free air volume 41534 ft3

6.5‐16 (VYC‐2299)

E2 MCR HVAC nominal unfiltered intake flow for accident duration and all FHA scenarios (assumed to include fresh air and air from surrounding areas as a result of ingress, egress and inleakage)

3700 cfm

DESCRIPTION VALUE REFERENCE / COMMENTS Other Variables

F1 Atmospheric dispersion factors from release point to locations of interest

See Table 6.3.6

6.5‐16 and 6.5‐20 (VYC‐2299 and VYC‐2275, Section 6)

F2

Breathing rates

Control Room

0 ‐ 720 hrs

3.5E‐04 m3/sec

6.5‐4(RG 1.183, pg 1.183‐18)

EAB 0 ‐ 2 hr

3.5E‐04 m3/sec

6.5‐4(RG 1.183, pg 1.183‐16)

F3

Control room occupancy factors

0 ‐ 24 hrs

1.0

6.5‐4 (RG 1.183, pg 1.183‐18)

24 ‐ 96 hrs

0.6

96 ‐ 720 hrs

0.4

F4 Exposure Intervals(b)

Control room 30 days 6.5‐4(RG 1.183, Sections 4.1.3, 4.1.5 and 4.2.6)

EAB 2 hrs

F5 Regulatory dose limits Control room TEDE 5 rem

6.5‐4(RG 1.183, pg 1.183‐19 and 10CFR50.67, Sec. (b)(2)(iii))

EAB TEDE 6.3 rem 6.5‐4(RG 1.183, Table 6)

LPZ TEDE 6.3 rem

F6 PAG Evacuation dose limit (EAB TEDE)

1 rem 6.5‐5(EPA 400‐R‐92‐001)

(a) In line with RG 1.183, Sec. 3.1, the radial peaking factor is applied to the average fuel-

assembly inventory based on the core inventory in Table 6.3.3. This is a conservative

approach and bounds any potential variations in the FHA source term resulting from

variations in the EFPDs and burnup in any given cycle.

(b) Even though all radioactivity is released to the atmosphere within 2 hours following a

design-basis FHA, the exposure intervals for the CR personnel was assumed to be 30 days.

This provides adequate time for cleanup of the airborne radioactivity still present within

the CR after termination of the 2-hr release, and also accounts for the delayed release of

noble-gas decay products from the refueling pool water produced upon decay of halogens

retained therein.

(c) The assembly drop height within the SFP was conservatively increased to 3 ft to account for

the difference in elevations between the top of the racks and the top of an assembly within

the racks, as well as any other dimensional uncertainties.


Table 6.3.3

Undecayed Core Inventory for Radionuclides Important in the Radiological Evaluation of DBAs

(From VYC-2260, Table 4.5, based on 1950 MWt, an enrichment range from 3.0 to 4.65 wt % U-235, and

core-average burnup from 5 to 58 GWD/MTU)

Nuclide Core Ci Nuclide Core Ci

Br‐83 8.267E+06 I‐132 7.900E+07

Kr‐83m 8.265E+06 Te‐133 6.602E+07

Br‐85 1.874E+07 Te‐133m 4.493E+07

Kr‐85 9.852E+05 I‐133 1.130E+08

Kr‐85m 1.894E+07 Xe‐133 1.128E+08

Rb‐86 2.496E+05 Xe‐133m 3.428E+06

Kr‐87 3.788E+07 Te‐134 1.036E+08

Kr‐88 5.355E+07 I‐134 1.254E+08

Kr‐89 6.755E+07 Cs‐134 2.971E+07

Sr‐89 6.724E+07 I‐135 1.051E+08

Sr‐90 7.999E+06 Xe‐135 4.540E+07

Y‐90 8.363E+06 Xe‐135m 2.232E+07

Sr‐91 8.684E+07 Cs‐136 7.602E+06

Y‐91 8.270E+07 Xe‐137 9.893E+07

Sr‐92 8.987E+07 Cs‐137 1.186E+07

Y‐92 9.008E+07 Ba‐137m 1.124E+07

Y‐93 9.857E+07 Xe‐138 9.851E+07

Zr‐95 9.645E+07 Ba‐139 1.043E+08

Nb‐95 9.673E+07 Ba‐140 1.004E+08

Zr‐97 9.596E+07 La‐140 1.009E+08

Mo‐99 1.034E+08 La‐141 9.573E+07

Tc‐99m 9.051E+07 Ce‐141 9.255E+07

Ru‐103 9.889E+07 La‐142 9.387E+07

Ru‐105 7.844E+07 Ce‐143 9.228E+07

Rh‐105 7.183E+07 Pr‐143 9.181E+07

Ru‐106 5.554E+07 Ce‐144 7.268E+07

Sb‐127 7.194E+06 Nd‐147 3.736E+07

Te‐127 7.151E+06 Np‐239 1.496E+09

Te‐127m 9.705E+05 Pu‐238 7.668E+05

Sb‐129 1.976E+07 Pu‐239 2.864E+04

Te‐129 1.947E+07 Pu‐240 6.061E+04

Te‐129m 2.890E+06 Pu‐241 1.281E+07

Te‐131m 8.405E+06 Am‐241 1.702E+04

I‐131 5.564E+07 Cm‐242 6.669E+06

Xe‐131m 6.192E+05 Cm‐244 2.358E+06

Te‐132 7.739E+07


Table 6.3.4

Undecayed Gap Activity Available for Release from Fuel Assembly Drop in

the SFP

Nuclide

Damaged Fuel‐Rod Gap Source Term for FHA

(Ci Available for Release from 98 Damaged Fuel Rods)

Assembly Drop in SFP

Kr‐83m 1.977E+03

Kr‐85 4.713E+02

Kr‐85m 4.530E+03

Kr‐87 9.060E+03

Kr‐88 1.281E+04

Kr‐89 1.616E+04

Xe‐131m 1.481E+02

Xe‐133 2.698E+04

Xe‐133m 8.199E+02

Xe‐135 1.086E+04

Xe‐135m 5.338E+03

Xe‐137 2.366E+04

Xe‐138 2.356E+04

Br‐83 1.977E+03

Br‐85 4.482E+03

I‐131 2.129E+04

I‐132 1.889E+04

I‐133 2.703E+04

I‐134 2.999E+04

I‐135 2.514E+04

Rb‐86 1.433E+02

Cs‐134 1.705E+04

Cs‐136 4.364E+03

Cs‐137 6.808E+03

Te‐131m 3.216E+03

Te‐132 1.851E+04

Te‐133 1.579E+04

Te‐133m 1.075E+04


Table 6.3.5

Typical Iodine Decontamination Factors and Iodine Speciation

Versus Water Depth above Dropped Assembly

SFP Water Depth (Hb) above Dropped Assembly

Iodine Decontamination Factor Iodine Speciation above Pool Water

(%)

(ft) (cm) Inorganic

(DFinorg, Eq. 5) Total

(DFtotal, Eq. 6) Inorganic (Sinorg,

Eq. 7) Organic

(Sorg, Eq. 8)

23 701.0 285.0 199.9 70.0 30.0

22.5 685.8 252.0 183.1 72.5 27.5

22 670.6 222.9 167.2 74.9 25.1

21.5 655.3 197.1 152.3 77.2 22.8

21 640.1 174.3 138.4 79.2 20.8

20.5 624.8 154.2 125.4 81.2 18.8

20 609.6 136.3 113.3 83.0 17.0

19 579.1 106.6 92.1 86.2 13.8

18 548.6 83.4 74.2 88.9 11.1

17 518.2 65.2 59.5 91.1 8.9

16 487.7 51.0 47.5 92.9 7.1

15 457.2 39.9 37.7 94.3 5.7

14 426.7 31.2 29.9 95.5 4.5

13 396.2 24.4 23.6 96.5 3.5

12 365.8 19.1 18.6 97.2 2.8

11 335.3 14.9 14.6 97.8 2.2

10 304.8 11.7 11.5 98.3 1.7

0 0.0 1.0 1.0 99.85 0.15


Table 6.3.6

Atmospheric Dispersion Factors for the Postulated FHA

(From VYC-2299 and VYC-2275, Section 6)

Scenario Release Point

Receptor Point

Post‐FHA Interval(a)

χ/Q(b) (sec/m3)

Control RoomFresh Air Intake

Instantaneous release

6.04E‐05

FHA Spent Fuel Pool

Ground Level Release(RB blowout panel) Open Containment

EAB Instantaneous

release 1.69E‐03

Control Room Fresh Air Intake

0 ‐ 2 hrs 5.89E‐03

2 ‐ 8 hrs 1.53E‐03

8 ‐ 24 hrs 6.41E‐04

24 ‐ 96 hrs 6.64E‐04

96 ‐ 720 hrs 5.10E‐04

FHA Spent Fuel Pool

Elevated Release(Main stack)

Closed Containment (w/wo SGTS)

EAB Instantaneous

release 1.35E‐04


Table 6.3.7

EAB TEDE Dose vs. Long Decay Time

(Assembly Drop in SFP with Closed Containment)

(Instantaneous elevated release with closed containment, with and without SGTS

filtration)

Post‐FHA Time (days)

EAB TEDE Dose (rem)

Without SGTS Filtration With SGTS Filtration

30 2.055E‐02 1.028E‐03

40 8.672E‐03 4.336E‐04

50 3.662E‐03 1.831E‐04

60 1.546E‐03 7.731E‐05


Figure 6.3-1

VY FHA - EAB TEDE Dose vs. Decay Time

(Assembly Drop in SFP with Open Containment)


Figure 6.3-2

VY FHA - MCR TEDE Dose vs. Decay Time

(Assembly Drop in SFP with Open Containment)


6.4 Site Events Evaluated

6.4.1 High Integrity Container (HIC) Drop Event

The drop of a HIC containing reactor water cleanup (RWCU) resins was evaluated

as taking place during normal operation of the plant, and the results are

reported in this section. Although these types of resins are no longer

expected to be on site after a period of time subsequent to cessation of power

operations (they will no longer be generated), the source term from these

resins is expected to bound source terms from other items (spent fuel pool

demineralizer resins, filter cartridges, etc.) that may be placed in containers

and moved subsequent to permanent shutdown.

6.4.1.1 Analytical Methodology

The list of radionuclides released into the cloud following the postulated

resin fire is provided in Table 6.4.1. The basis for this table is provided in

Section 6.4.1.2. The release was assumed to be instantaneous. Radiation doses

were calculated to the total body due to cloud submersion and a 2-hr direct

shine dose from standing on contaminated ground, and to the thyroid and

identified critical organ (lung) based on the inhalation pathway.

The whole body and organ doses were based on the standard equations for

instantaneous releases and the applicable dose conversion factors. The DCFs

were extracted from NUREG/CR-1918 (ORNL/NUREG-79) (Reference 6.5-23) for the

air submersion pathway, Regulator Guide 1.109 (Reference 6.5-24) for the

inhalation pathway and all nuclides except I-129, ICRP-30 (Reference 6.5-25)

for the inhalation pathway and I-129, and Regulator Guide 1.109 (Reference 6.5-

26) for the contaminated ground-shine pathway.

With respect to the whole body dose from ground deposition, the analysis was

based on assuming uniform dispersion of the released activity from Table 6.4.1

over the deposition area, and a 2-hr radiation exposure interval. The

deposition area (about 1400 m2) was conservatively assumed to encompass the

distance between the reactor building and the closest receptor at the site

boundary and a 2-sigma plume width for the assumed prevailing atmospheric

stability (F) at the time of the postulated incident.

6.4.1.2 Assumptions

Sandia National Laboratory has conservatively estimated, for a severity

Category 3 transportation accident (which includes 99% of urban and 94% of

rural accidents), no more than 1% (0.01) of any package contents would be

released. For the purposes of the analysis, it was assumed that 0.5% of the

released activity becomes aerosolized as a result of the fire.


A HIC of 150 feet3 capacity contains dewatered reactor water cleanup (RWCU)

resins at a density of 0.8 (g/cc), and contains all radionuclides typically

found in nuclear power plant radwaste. Each radionuclide inventory in the HIC

is at the Department of Transportation (DOT) limit for Low Specific Activity

(LSA) material, except for I-129, which is assumed to be at the 10CFR61 limit

for disposal. A source term of RWCU resins is considered to be the most

limiting from a radiological perspective.

The assumed liner drop occurs 250 meters from the site boundary (EAB). This is

based on original analysis performed for a drop of a HIC at the corner of the

waste storage pad (corner closest to the site boundary), built for

prefabricated concrete storage modules. This is a conservative assumption

because the radwaste loading area is farther away from the closest site

boundary than the 250 meters in the original HIC drop analysis.

Conservative dispersion conditions are assumed for a ‘puff release’ under

Stability Class F and a wind-speed of 1 meter/second. The puff is assumed to

travel along the ground in the direction of the nearest site boundary, at

ground level.

The dose acceptance criteria were set equal to "a small fraction" of the 10 CFR

100 dose limits of 25 rem whole body and 300 rem thyroid (i.e., to 10% of these

values, or 2.5 rem whole body and 30 rem thyroid). Because of the nature of

the source term (which consists mostly of long-lived radionuclides), the

thyroid limit of 30 rem was also applied to the critical organ (identified to

be the lung in this case).

Other assumptions are contained in the footnotes in Table 6.4.1.

6.4.1.3 Inputs

The source term for the dropped container containing RWCU dewatered resins is

provided in Table 6.4.1.

The atmospheric dispersion factor is based on a conservative downwind distance

of 250 meters (to the closest site boundary from the reactor building, and is

determined to be 0.079 sec/m3.

The breathing rate for the organ dose is 8000 m3/yr (2.537E-04 m3/sec), from RG

1.109.


6.4.1.4 Radiological Consequences/Results

10% of 10CFR100 Dose Acceptance Criteria

(rem)

Calculated Dose (rem)

EAB (2 hours)

2.5 rem (whole body) 6.52E‐03 (a) 9.59E‐03 (b) 16.1E‐03 (c)

30 rem (thyroid, also applied to

the critical organ) 2.03E‐03 (thyroid)

4.58 (lung)

(a) Dose from standing on contaminated ground (2‐hr exposure) (b) Dose from cloud passage overhead due to resin fire and aerosol release (c) Sum of ground plane external plus airborne from cloud


Table 6.4.1 HIC Drop Source Term Release Activity

Nuclide 1

A2 Values

2 (Ci)

LSA Limit 3

(mCi/gm)

Total Activity 4

(Ci)

Liner Drop

ReleaseActivity 5

(Ci) Cr‐51 600 0.3 1020 0.051

Mn‐54 20 0.3 1020 0.051

Fe‐55 1000 0.3 1020 0.051

Co‐58 20 0.3 1020 0.051

Co‐60 7 0.3 1020 0.051

Fe‐59 10 0.3 1020 0.051

Ni‐59 900 0.3 1020 0.051

Ni‐63 100 0.3 1020 0.051

Sb‐124 5 0.3 1020 0.051

Zn‐65 30 0.3 1020 0.051

Ag‐110m 7 0.3 1020 0.051

Sr‐89 10 0.3 1020 0.051

Sr‐90 0.4 0.005 17 0.00085

Zr‐95 20 0.3 1020 0.051

NB‐95 20 0.3 1020 0.051

Tc‐99 25 0.3 1020 0.051

I‐129 6 2 NA 0.34 0.000017

Cs‐134 10 0.3 1020 0.051

Cs‐137 10 0.3 1020 0.051

Ce‐141 25 0.3 1020 0.051

Ce‐144 7 0.3 1020 0.051

Pu‐238 0.003 0.0001 0.34 0.000017 Pu‐

239/240 0.002 0.0001 0.34 0.000017

Am‐241 0.003 0.0001 0.34 0.000017

Cm‐242 0.2 0.005 17 0.00085 Cm‐

243/244 0.01 0.0001 0.34 0.000017

19415.7 0.970785

Footnotes: 1‐ Nuclide Listing: A listing of radionuclides that typically are determined by laboratory analysis to be present in

RWCU resin. Short lived gaseous and volatile radionuclides are not detected in typical radwaste streams. 2 ‐ A2: For informational purposes, quantities of normal form (not special form) radionuclides, expressed in

curies, permitted by DOT to be contained in a Type A disposal package. Refer to 49CFR173.435 for listing. 3 ‐ LSA Limit: DOT determined Low Specific Activity concentration limit, expressed in units of millicuries per gram

of material. Under regulations dated January 1989, LSA is a function of the tabulated A2 variable above. Refer to 49CFR173.403(n)(4) for the relationship.

4 ‐ Total Activity: Because concentration and distribution of radionuclides in waste are expected to vary over time, it is assumed for purposes of this radiological accident analysis that all radionuclides are at their upper limit. In reality, a small number of radionuclides might be expected to approach a limiting condition while the majority would be at some lower level. Total activity is based on the following: A) 150 ft3 (4.25 m3) liner waste, density of 50 lb/ft3 = 4.248E+06 cc @ 0.8 gm/cc giving 3.40E+06 gm. B) Each nuclide is at the LSA limit.

5 ‐ Release Activity: The quantity of each nuclide assumed to be released from the waste liner to form the source term. The release activity is based on: A) Liner drop incident results in liner failure and release of 1% total contents. B) Of the 1% material released, 0.5% is aerosolized to form a "release cloud" source term. The release fraction is 0.01 and the aerosol fraction is 0.00005 of the total HIC activity).

6 ‐ I‐129 is limited by 10CFR61 burial requirements rather than DOT. The class C disposal limit for I‐129, as listed in 10CFR61.55, Table 1, is 0.08 Ci/m3 (or μCi/cc).


6.5 References

1. BVY 15-001, “Certifications of Permanent Cessation of Power Operations

and Permanent Removal of Fuel from the Reactor Vessel, Vermont Yankee

Nuclear Power Station”, January 12, 2015.

2. Holtec International Final Safety Analysis Report for the Hi-Storm 100

Cask System, Revision 4.

3. Code of Federal Regulations Title 10 Part 50.67 (10CFR50.67), Accident

Source Term

4. Regulatory Guide 1.183, Alternative Radiological Source Terms for

Evaluating Design Basis Accidents at Nuclear Power Reactors, Rev. 0, July

2000

5. EPA 400-R-92-001, Manual of Protective Action Guides and Protective

Actions for Nuclear Incidents (1991)

6. 10CFR50, Appendix I, Numerical Guides for Design Objectives and Limiting

Conditions for Operation to Meet the Criterion “As Low as is Reasonably

Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power

Reactor Effluents

7. BVY 13-097, "Technical Specifications Proposed Change No. 306 – Eliminate

Certain ESF Requirements during Movement of Irradiated Fuel", Nov. 14,

2013.

8. NVY 15-013, Vermont Yankee Nuclear Power Station – Issuance of Amendment

to Renewed Facility Operating License RE: Eliminate Operability

Requirements for Secondary Containment When Handling Sufficiently Decayed

Irradiated Fuel or a Fuel Cask (TAC No. MF3086), dated February 12, 2015.

9. NUMARC 93-01, “Industry Guidelines for Monitoring the Effectiveness of

Maintenance at Nuclear Power Plants”

10. AREVA Document 32-9053350-001, “ELISA-2 - A Software Package for the

Radiological Evaluation of Licensing and Severe Accidents at Light-Water

Nuclear Power Plants Based on the Classical and Alternative-Source-Term

Methodologies” (Aug. 2008) [See also AREVA Document 2A4.26-2A4-ELISA2-

2.4_Users_Manual-000, “ELISA-2 Version 2.4 User’s Manual – Revision 2”.]

11. EPA 520/1-88-020, Federal Guidance Report No. 11, "Limiting Values of

Radionuclide Intake and Air Concentration, and Dose Conversion Factors for


Inhalation, Submersion, and Ingestion" (ORNL, September 1988)

12. EPA 402-R-93-081, Federal Guidance Report No. 12, "External Exposure to

Radionuclides in Air, Water, and Soil" (ORNL, September 1993)

13. US NRC Regulatory Issue Summary (RIS) 2006-04, "Experience with

Implementation of Alternative Source Terms", March 2006.

14. G. Burley, "Evaluation of Fission Product Release and Transport for a

Fuel Handling Accident," U.S. NRC Technical Paper (October 1971,

Accession Number: 8402080322 in ADAMS or PARS).

15. General Electric Standard Application for Reactor Fuel, GESTAR II

(Supplement for the Unite States) Licensing Topical Report, pages US-25

through US-28, NEDE-24011-P-A-14-US, Class III14, June 2000.

16. ENTERGY Calculation VYC-2299, “Radiological AST Fuel Handling Accident

Analysis [PSAT 3019CF.QA.05, Rev. 0]” (Jun. 2003)

17. ENTERGY Calculation VYC-2260, “Bounding Core Inventories of Actinides and

Fission Products for Design-Basis Applications at 1950 MWt” (Rev. 0, Feb.

2003)

18. ENTERGY Calculation VYC-2206, “Determination of Number of Damaged Fuel

Rods due to Refueling Accident”, Rev. 0

19. ENTERGY Calculation VYC-3187, “Fuel Handling Accident Supplemental

Analysis (Specific to the Spent Fuel Pool”, Rev. 0

20. ENTERGY Calculation VYC-2275, “Control Room Air Intake X/Q Due to Release

from Reactor Building Blowout Panel Using Arcon96 Methodology” (Rev. 0,

April 2003)

21. ENTERGY Calculation VYC-2302, "Radiological AST LOCA Analysis" [PSAT

3019CF.QA.08, Rev. 2]

22. ASME Steam Tables, Sixth Edition

23. NUREG/CR-1918 (ORNL/NUREG-79), Dose Rate Conversion Factors for External

Exposure to Photons and Electrons (August 1981)

24. Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine

Releases of Reactor Effluents for the Purpose of Evaluating Compliance

with 10 CFR Part 50, Appendix I (Revision 1, October 1977), Table E-7,

Inhalation Dose Factors for Adults, Thyroid and Lung


25. ICRP-30, Limits for Intake of Radionuclides by Workers, Supplement 1, pg

202

26. Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine

Releases of Reactor Effluents for the Purpose of Evaluating Compliance

with 10 CFR Part 50, Appendix I (Revision 1, October 1977), Table E-6,

External Dose Factors for Standing on Contaminated Ground, Total Body


6.6 Appendices

Appendix A - Fuel Damage from Assembly Drop onto SFP Fuel Racks

BACKGROUND

Damage due to a drop of a fuel assembly within the SFP was evaluated using the

GESTAR II method (Reference 6.4-15). In GESTAR II Section S.2.2.3.5, GE presents

a methodology for estimating the number of fuel rods that fail as a result of a

fuel handling accident where one fuel bundle is dropped onto a grouping of

additional fuel bundles. This methodology was used to assess the fuel rod damage

impacts of a FHA in the spent fuel pool where one fuel bundle is dropped onto a

second fuel bundle contained within the SPF racks.

Figure A-1 shows the pertinent dimensions for the SFP. Critical to the analysis is

the maximum drop distance for an assembly, 22.74 inches (1.9 feet), increased to

3 ft to account for the difference in elevations between the top of the racks and

the top of assembly within the racks, as well as any other dimensional uncertainties


Figure A‐1

Dimensions of SFP, Fuel Rack and Fuel Handling Equipment

METHODOLGY

Key inputs to the analysis are as shown Table 6.3.2.

The basic steps involved in the GESTAR II evaluation methodology are described in

the following paragraphs. Because this methodology is an approximation of the

behavior of an extremely complex series of mechanical interactions, a commentary

describing the explicit and implicit assumptions involved in each step is provided

with description of how each step is implemented.


1. Calculate the available impact energy as a result of the initial drop.

The GESTAR II methodology is a simplified energy balance approach which

examines the kinetic energy of the falling fuel bundle and associated fuel

handling equipment, and from that calculates the number of failed fuel rods

based on an established fuel rod failure energy threshold. The kinetic

impact energy is taken to be equal to the initial potential energy of the

dropped bundle system, that is, the weight of the fuel bundle (with

associated fuel handling equipment weights) multiplied by the drop height.

This assumption avoids the complexities in establishing an expected impact

velocity for the fuel bundle falling through a fluid exerting a significant

drag force. This assumption will result in a conservatively high estimated

impact energy.

2. Calculate the energy available to fail rods.

The calculated impact energy is apportioned amongst the various components

involved in the impact event. One half of the energy is assumed to be

absorbed by the dropping fuel bundle and one half is assumed to be absorbed

by the stationary fuel bundle. Within the stationary fuel bundle, the impact

energy is further divided between the cladding and the remaining structural

(non-fuel pellet) material of the fuel bundle. This division is made by

using the ratio of the mass of the fuel cladding to the mass of the

remaining fuel bundle structure. For the GE fuel designs evaluated in the

GESTAR II analysis, the maximum ratio is 0.510.

3. Calculate the energy required to fail one fuel rod.

In the GESTAR II evaluation, detailed for GE-13 fuel, each fuel rod in the

stationary fuel bundles is expected to fail upon absorbing 200 ft-lb of

energy based on a failure criteria of 1% uniform plastic deformation. In

that evaluation, every rod in the dropped bundle is assumed to fail due to

excessive bending moments imposed on the lower tie plate. In VYC-2206, the

failure threshold for the stationary bundles was calculated to be 167 ft-lb

for the GE14 fuel design using a simple scaling of clad cross-sectional

area. Though no details of the derivation of this threshold are provided in

the GESTAR II documentation, it is anticipated that sufficient conservatism

is built in to use this value, particularly given the assumption of complete

fuel rod failure in the dropped bundle and the conservative apportionment of

impact energies amongst the fuel rods described below.


4. Calculate the number of failed fuel rods as a result of the initial drop.

The number of fuel rods expected to fail as a result of the initial drop are

calculated simply as the energy available to fail fuel rods divided by the

energy required to fail one fuel rod. This method is conservative in that it

assumes that no deformation energy is “wasted” by deforming additional fuel

rods, though not to the point of failure. In other words, no other fuel rods

provide any support for the fuel rods that fail.

5. Calculate the Number of additional fuel rod failures due to the secondary, tip-over impact.

Added to the number of fuel rods failed as a result of the initial impact is

the number of rods expected to fail as the dropped fuel bundle tips over and

impacts additional fuel bundles in the core. Since the analysis considers a

drop within the spent fuel pool, and the tops of the fuel bundles being

stored are below the tops of the spent fuel racks, no additional bundle

impacts result from the tip-over portion of the accident.

ASSUMPTIONS

All rods in the dropped bundle are conservatively assumed to fail. The

number of failed rods for impacted bundles are established from an estimate

of "energy required to fail a fuel rod," given the size (thickness and

diameter) and thus the strength of the GE14 rod cladding, which is slightly

thinner than that for the GE-13 fuel (Reference 6.4-18, VYC-2206, Attachment

2). A strength for the GE14 10x10 using a ratio of cross-sectional areas

for GE-9 8x8 and GE-13 9x9 fuel arrays is determined (Reference 6.4-18, VYC-

2206. Attachment 2). Strength (necessary compression failure energy) for

the GE14 10x10 bundles was estimated to be 170.3 ft-lbs. As a conservative

estimate, the value of 167 ft-lb was used.

Fuel rods in an assembly are assumed to fail by 1% strain in compression

(Reference 6.4-15). It is expected that a GE14 fuel rod will absorb 167 ft-

lbs of energy in this failure mode. Since the partial length rods are not

attached to, and do not contact the upper tie plate, less energy should be

transferred to them if the fuel assembly is impacted. Therefore, assuming

that the part length rods fail at these force levels (if they do fail) is

conservative.

All fuel assemblies employed in the analysis were assumed to be discharged at

the same time so as to maximize the released radioactivity, as the noble gases

and halogens in older assemblies have already decayed to insignificant levels.


The GESTAR II (Reference 6.4-15) analysis considers dropping a fuel bundle

drop, including a grapple mast and head. The added weight of the grapple

hook assembly and the fuel handling mast are added to the weight of the

dropped fuel bundle.

All rods in the dropped assembly are conservatively assumed to fail. The

GESTAR II (Reference 6.4-15) makes this assumption.

Toppling of the dropped assembly is assumed to result in no further damage

to assemblies in the rack due to the configuration of the rack, which

expends above the top of the assemblies themselves.

Conservatism relative to the results of impact is provided by assembly features that

should absorb some impact energy generated by a dropped assembly:

The raised rectangular lifting-handle assembly at the top of the assembly,

which may bend or fracture, dissipating impact energy,

Expansion springs within the assembly

The semi-circular insertion guide (assembly base extension) that stands out at

the bottom of the assembly and that are used to guide it into the centering

socket at the base of the fuel rack slot.

RESULTS

Results are summarized below for the failure of fuel rods in the impacted GE14 and

GNF2 10x10 assemblies. Most entries in this table are from Table 6.3.2;

calculated values include their derivation basis, in parentheses. All 92 rods in

the dropped assembly are assumed to fail. Impact of the dropped assembly in the

fuel rack results in an additional 5 GE14 failed rods (for a total of 97 failed

rods due to the accident), and in 6 GNF2 failed rods (for a total of 98). The

number of failed fuel rods for the drop in the SFP will be conservatively based on

98 assemblies, thus representative for both fuel assembly types.


Description Assembly Type

GE14 GNF2

Maximum drop height (ft) 3 3

Weight of 10x10 fuel assembly and channel

submerged in water (lb) 569.5 580

Dry weight mast & grapple (lb)(a) 707.6 707.6

Weight of fuel assembly/channel submerged in

water + mast/grapple dry weight (lb)

1277.1

(569.5 +

707.6)

1287.6

(580.0

+707.6)

Available Impact Energy (wt x height, ft-lbs) 3831.3

(3 * 1277.1)

3862.8

(3 * 1287.6)

Energy absorbed by dropped fuel assembly (ft-

lbs)

1915.7

(3831.3 / 2)

1931.4

(3862.8 / 2)

Energy absorbed by stationary fuel bundle (ft-

lbs) 1915.7 1931.4

Percent energy for stationary clad deformation

(GESTAR II) 0.51 0.51

Energy to stationary fuel bundle for clad

deformation (ft-lbs)

977.0

(1915.7 *

0.51)

985.0

(1931.4 *

0.51)

Failed rods in dropped assembly (assumed all) 92 92

Energy required to damage stationary fuel rods 167 157

1st impact damaged stationary fuel rods

(analytical value)

5.9

(977.0 / 167)

6.3

(985.0 / 157)

Total number failed rods (Note: Damaged

stationary rod numbers were rounded down to

whole rods since there can be no partial rod

damage.)

97

(92+5)

98

(92+6)

(a) The mast and grapple were conservatively assigned their dry weight since the

SFP geometry permits only their partial submersion during the assembly drop.

The dry weight of 707.6 lbs. was calculated based on a wet weight of 619

lbs. (Table 6.3-2, Item #B4), a density of 7.85 g/cc for the mast and

grapple (i.e., that of iron) and a density of 0.983 g/cc for water at 140 oF

from ASME Steam Tables, Sixth Edition (Reference 6.4-22): 619 * 7.85 /

(7.85 – 0.983) = 707.6 lb.


AGING MANAGEMENT

TABLE OF CONTENTS Section Title Page

7.1 SUPPLEMENT FOR RENEWED OPERATING LICENSE ............................. 2

7.2 AGING MANAGEMENT PROGRAMS AND ACTIVITIES .............................. 2

7.2.1 Deleted ..................................................... 2

7.2.2 Diesel Fuel Monitoring Program .............................. 3

7.2.3 Fire Protection Program ..................................... 3

7.2.4 Fire Water System Program ................................... 3

7.2.5 Instrument Air Quality Program .............................. 4

7.2.6 Non-EQ Inaccessible Medium-Voltage Cable Program ............................................................ 4

7.2.7 Oil Analysis Program ........................................ 5

7.2.8 Periodic Surveillance and Preventive Maintenance Program ..................................................... 5

7.2.9 Service Water Integrity Program ............................. 5

7.2.10 Structures Monitoring – Masonry Wall Program. ................ 6

7.2.11 Structures Monitoring – Structures Monitoring Program ..................................................... 6

7.2.12 System Walkdown Program ..................................... 6

7.2.13 Water Chemistry Control – Auxiliary Systems Program ..................................................... 6

7.2.14 Water Chemistry Control – BWR Program ....................... 7

7.2.15 Deleted ..................................................... 7

7.2.16 Bolting Integrity Program ................................... 7

7.2.17 Deleted ..................................................... 7

7.2.18 Deleted ..................................................... 7

7.2.19 Neutron Absorber Monitoring Program ......................... 7

7.3 REFERENCES ........................................................... 8

7.4 LIST OF LICENSE RENEWAL COMMITMENTS .................................. 9


7.1 SUPPLEMENT FOR RENEWED OPERATING LICENSE

The Vermont Yankee Nuclear Power Station (VYNPS) license renewal application

(LRA) (Reference 7.3.1) and information in subsequent related correspondence

provided sufficient basis for the NRC to make the findings required by 10 CFR

54.29 (Final Safety Evaluation Report) (References 2, 3 and 4). As required by

10 CFR 54.21(d), this DSAR supplement contains a summary description of the

remaining programs and activities for managing the effects of aging.

7.2 AGING MANAGEMENT PROGRAMS AND ACTIVITIES

The integrated plant assessment for license renewal identified aging

management programs necessary to provide reasonable assurance that components

within the scope of license renewal will continue to perform their intended

functions consistent with the current licensing basis (CLB). This section

describes the aging management programs and activities that will be required

during the period of wet fuel storage.

VYNPS quality assurance (QA) procedures, review and approval processes, and

administrative controls are implemented in accordance with the requirements of

10 CFR 50, Appendix B. The Quality Assurance Program applies to safety-related

structures and components. Corrective actions and administrative (document)

control for both safety-related and non-safety related structures and

components are accomplished per the existing VYNPS corrective action program

and document control program and are applicable to all aging management

programs and activities that will be required during the period of wet fuel

storage. The confirmation process is part of the corrective action program and

includes reviews to assure that proposed actions are adequate, tracking and

reporting of open corrective actions, and review of corrective action

effectiveness. Any follow-up inspection required by the confirmation process

is documented in accordance with the corrective action program.

The corrective action, confirmatory process, and administrative controls of

the (10 CFR Part 50, Appendix B) Quality Assurance Program are applicable to

all aging management programs and activities that will be required during the

period of wet fuel storage.

7.2.1 Deleted


7.2.2 Diesel Fuel Monitoring Program

The Diesel Fuel Monitoring Program entails sampling to ensure that adequate

diesel fuel quality is maintained to prevent plugging of filters, fouling of

injectors, and corrosion of fuel systems. Exposure to fuel oil contaminants

such as water and microbiological organisms is minimized by periodic draining

and cleaning of tanks and by verifying the quality of new oil before its

introduction into storage tanks.

7.2.3 Fire Protection Program

The Fire Protection Program includes a fire barrier inspection and a diesel-

driven fire pump inspection. The fire barrier inspection requires periodic

visual inspection of fire barrier penetration seals, fire barrier walls,

ceilings, and floors, and periodic visual inspection and functional tests of

fire rated doors to ensure that their functionality is maintained. The diesel-

driven fire pump inspection requires that the pump be periodically tested to

ensure that the fuel supply line can perform its intended function.

Corrective actions, confirmation process, and administrative controls in

accordance with the requirements of 10 CFR Part 50 Appendix B are applied to

the Fire Protection Program.

7.2.4 Fire Water System Program

The Fire Water System Program applies to water-based fire protection systems

that consist of sprinklers, nozzles, fittings, valves, hydrants, standpipe

hose connections, standpipes, and aboveground and underground piping and

components that are tested in accordance with applicable National Fire

Protection Association (NFPA) codes and standards. Such testing assures

functionality of systems. Also, many of these systems are normally maintained

at required operating pressure and monitored such that leakage resulting in

loss of system pressure is immediately detected and corrective actions

initiated.

In addition, wall thickness evaluations of fire protection piping are

periodically performed on system components using non-intrusive techniques

(e.g. volumetric testing) to identify evidence of loss of material due to

corrosion.

A sample of sprinkler heads will be inspected using the guidance of NFPA 25

(2002 Edition) Section 5.3.1.1.1, which states, “Where sprinklers have been in

place for 50 years, they shall be replaced or representative samples from one

or more sample areas shall be submitted to a recognized testing laboratory for

field service testing.” This sampling will be repeated every 10 years after

initial field service testing.


7.2.5 Instrument Air Quality Program

The Instrument Air Quality Program ensures that instrument air supplied to

components is maintained free of water and significant contaminants, thereby

preserving an environment that is not conducive to loss of material. Dew point

and hydrocarbon concentration are periodically checked to verify the instrument

air quality is maintained.

7.2.6 Non-EQ Inaccessible Medium-Voltage Cable Program

In the Non-EQ Inaccessible Medium-Voltage Cable Program, medium-voltage cables

with a license renewal intended function that are exposed to significant

moisture and voltage are tested at least once every six years to provide an

indication of the condition of the conductor insulation. The specific test

performed is a proven test for detecting deterioration of the insulation

system due to wetting, such as power factor, partial discharge, polarization

index, or other testing that is state-of-the-art at the time the test is

performed. Significant moisture is defined as periodic exposures that last

more than a few days.

Inspections for water collection in cable manholes containing inaccessible

low-voltage and medium-voltage cables with a license renewal intended function

will occur at least once every year. Additional condition-based inspections of

these manholes will be performed based on: a) potentially high water table

conditions, as indicated by high river level, and b) after periods of heavy

rain. The inspection results are expected to indicate whether the inspection

frequency should be modified. The manhole inspection will include direct

observation that cables are not wetted or submerged, that cables/splices and

cable support structures are intact, and that dewatering/drainage systems

(i.e. sump pumps), if installed, and associated alarms operate properly.

Inaccessible low-voltage cables (cables with operating voltage from 400 V to 2

kV) with a license renewal intended function are included in this program.

Inaccessible low-voltage cables will be tested for degradation of the cable

insulation prior to the period of extended operation and at least once every

six years thereafter. A proven, commercially available test will be used for

detecting deterioration of the insulation system for inaccessible low-voltage

cables potentially exposed to significant moisture. Failure of the cable test

results and manhole inspections to meet the acceptance criteria will require

corrective actions. The corrective actions will address modifying the cable

test frequency and the manhole inspection frequency.


7.2.7 Oil Analysis Program

The Oil Analysis Program maintains oil systems free of contaminants (primarily

water and particulates) thereby preserving an environment that is not

conducive to loss of material, cracking, or fouling. Activities include

sampling and analysis of lubricating oil for detrimental contaminants, water,

and particulates.

Sampling frequencies are based on vendor recommendations, accessibility during

facility operation, equipment importance to facility operation, and previous

test results.

7.2.8 Periodic Surveillance and Preventive Maintenance Program

The Periodic Surveillance and Preventive Maintenance Program includes periodic

inspections and tests that manage aging effects not managed by other aging

management programs. The preventive maintenance and surveillance testing

activities are generally implemented through repetitive tasks or routine

monitoring of facility operations.

Periodic inspections using visual or other non-destructive examination

techniques verify that the following components are capable of performing

their intended function.

reactor building crane, rails, and girders

refueling platform carbon steel components

equipment lock sliding doors

yard concrete handholes and manholes

housings of control room HVAC package heating and cooling coils, control room chiller, and control room chilled water condensers

control room ventilation fan duct flexible connections

instrument air supply systems

internal surfaces of carbon steel components in the potable water system containing untreated water

internal surfaces of carbon steel and copper alloy components in the radwaste system containing untreated water

7.2.9 Service Water Integrity Program

The Service Water Integrity Program ensures that the effects of aging on the

service water system (SWS) will be managed for the period of wet fuel storage.

The program includes opportunistic component inspections for erosion,

corrosion, and blockage to verify the heat transfer capability of the safety-

related and nonsafety-related heat exchangers cooled by SWS. Chemical

treatment and periodic cleaning are used to control or prevent fouling within

the SWS heat exchangers.


7.2.10 Structures Monitoring – Masonry Wall Program.

The objective of the Masonry Wall Program is to manage cracking so that the evaluation basis established for each masonry wall within the scope of license renewal remains valid through the period of wet fuel storage.

The program includes all masonry walls identified as performing intended functions in accordance with 10 CFR 54.4. Included walls are the 10 CFR 50.48 required walls and masonry walls in the reactor building, intake structure and control room building.

Masonry walls are visually examined at a frequency selected to ensure there is no loss of intended function between inspections.

7.2.11 Structures Monitoring – Structures Monitoring Program Structures monitoring is in accordance with 10 CFR 50.65 (Maintenance Rule) as

addressed in Regulatory Guide (RG) 1.160 and NUMARC 93-01. Periodic

inspections are used to monitor condition of structures and structural

components to ensure there is no loss of structure or structural component

intended function.

7.2.12 System Walkdown Program The System Walkdown Program entails inspections of external surfaces of

components subject to aging management review. The program is also credited

with managing loss of material from internal surfaces, for situations in which

internal and external material and environment combinations are the same such

that external surface condition is representative of internal surface

condition.

Surfaces that are not readily accessible, such as piping located in

underground vaults, are inspected at least once every 5 years. The inspection

frequencies provide reasonable assurance that the effects of aging will be

managed such that applicable components will perform their intended function

during the period of wet fuel storage.

7.2.13 Water Chemistry Control – Auxiliary Systems Program

The purpose of the Water Chemistry Control – Auxiliary Systems Program is to

manage aging effects for components exposed to treated water.


7.2.14 Water Chemistry Control – BWR Program

The objective of the Water Chemistry Control - BWR Program is to manage aging

effects caused by corrosion and cracking mechanisms. The program relies on

monitoring and control of water chemistry based on BWR Water Chemistry

Guidelines, 2008 Revision (BWRVIP-190). EPRI guidelines in BWRVIP-190 include

recommendations for controlling water chemistry in the torus, condensate

storage tank, demineralized water storage tanks, and spent fuel pool.

7.2.15 Deleted

7.2.16 Bolting Integrity Program

The Bolting Integrity Program relies on recommendations for a comprehensive

bolting integrity program, as delineated in NUREG-1339, and industry

recommendations, as delineated in the Electric Power Research Institute (EPRI)

NP-5769, with the exceptions noted in NUREG-1339 for safety-related bolting.

The program relies on industry recommendations for comprehensive bolting

maintenance, as delineated in EPRI TR-104213 for pressure retaining bolting

and structural bolting.

7.2.17 Deleted

7.2.18 Deleted

7.2.19 Neutron Absorber Monitoring Program

The Neutron Absorber Monitoring Program is a new program that will manage loss

of material and reduction of neutron absorption capacity of Boral neutron

absorption panels in the spent fuel racks. The loss of material and the

reduction of the neutron-absorbing capacity will be determined through coupon

testing, direct in situ testing or both. Such testing will include periodic

verification of boron loss through areal density measurement of coupons or

through direct in situ techniques, such as measurement of boron areal density,

measurement of geometric changes in the material (blistering, pitting and

bulging), and detection of gaps through blackness testing.


7.3 REFERENCES

1. VYNPS License Renewal Application

2. NUREG-1907, “Safety Evaluation Report Relating to the License

Renewal of Vermont Yankee Nuclear Power Station,” May 2003.

3. NUREG-1907, Supplement 1, “Safety Evaluation Report Related to the

License Renewal of Vermont Yankee Nuclear Power

Station,” September 2009.

4. NUREG-1907, Supplement 2, “Safety Evaluation Report Related to the

License Renewal of Vermont Yankee Nuclear Power Station,” April 2011.

VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST

VYNPS DSAR

Revision 1 7.0-9 of 13

7.4 LIST OF LICENSE RENEWAL COMMITMENTS

During the review of the VYNPS LRA by the staff of the US Nuclear Regulatory Commission (NRC), Entergy Nuclear

Operations, Inc. made commitments related to aging management programs (AMPs) to manage the aging effects of

structures and components prior to the period of extended operation. The following table lists these

commitments remaining applicable following permanent cessation of operations and certification of permanent

defueling. The implementation schedules and the sources for each commitment are also provided.

ITEM

COMMITMENT

IMPLEMENTATION

SCHEDULE

LRA Section

SOURCE

3

The Diesel Fuel Monitoring Program will be enhanced to ensure ultrasonic thickness measurement of the fuel oil storage tank bottom surface will be performed every 10 years during tank cleaning and inspection. Ultrasonic thickness measurement of the fire pump diesel storage (day) tank bottom will be performed every 10 years.

March 21, 2012

B.1.9

BVY 06-009

BVY 07-018

LBDCR# FCR 26/009

4

The Diesel Fuel Monitoring Program will be enhanced to specify that UT measurements of the fuel oil storage tank bottom surface will have acceptance criterion in accordance with American Petroleum Institute standard API 653 and UT measurements of the fire pump diesel storage (day) tank bottom surface will have acceptance criterion in accordance with Steel Tank Institute standard STI SP001.

March 21, 2012

B.1.9

BVY 06-009

BVY 07-018

BVY 10-069 BVY 11-007

8

Procedures will be enhanced to specify that fire damper frames in fire barriers will be inspected for corrosion. Acceptance criteria will be enhanced to verify no significant corrosion.

March 21, 2012

B.1.12.1

BVY 06-009


VYNPS DSAR


ITEM

COMMITMENT

IMPLEMENTATION

SCHEDULE

LRA Section

SOURCE

9

Procedures will be enhanced to state that the diesel engine sub-systems (including the fuel supply line) will be observed while the pump is running. Acceptance criteria will be enhanced to verify that the diesel engine did not exhibit signs of degradation while it was running; such as fuel oil, lube oil, coolant, or exhaust gas leakage.

March 21, 2012

B.1.12.1

BVY 06-009

10

Fire Water System Program procedures will be enhanced to specify that in accordance with NFPA 25 (2002 edition), Section 5.3.1.1.1, when sprinklers have been in place for 50 years a representative sample of sprinkler heads will be submitted to a recognized testing laboratory for field service testing. This sampling will be repeated every 10 years.

March 21, 2012

B.1.12.2

BVY 06-009

11

The Fire Water System Program will be enhanced to specify that wall thickness evaluations of fire protection piping will be performed on system components using non-intrusive techniques (e.g., volumetric testing) to identify evidence of loss of material due to corrosion. These inspections will be performed before the end of the current operating term and during the period of extended operation. Results of the initial evaluations will be used to determine the appropriate inspection interval to ensure aging effects are identified prior to loss of intended function.

March 21, 2012

B.1.12.2

BVY 06-009


VYNPS DSAR


ITEM

COMMITMENT

IMPLEMENTATION

SCHEDULE

LRA Section

SOURCE

13

Implement the Non-EQ Inaccessible Medium-Voltage Cable Program as described in LRA Section B.1.17.

Inspections for water accumulation in manholes containing inaccessible low-voltage and medium-voltage cables with a license renewal intended function will be performed at least once every year. Additional condition-based inspections of these manholes will be performed based on: a) potentially high water table conditions, as indicated by high river level, and b) after periods of heavy rain. The inspection results are expected to indicate whether the inspection frequency should be modified.

Inaccessible low-voltage cables (400 V to 2 kV) with a license renewal intended function are included in this program. Inaccessible low-voltage cables will be tested for degradation of the cable insulation prior to the period of extended operation and at least once every six years thereafter. A proven, commercially available test will be used for detecting deterioration due to wetting of the insulation system for inaccessible low-voltage cables.

March 21, 2012

B.1.17

BVY 06-009

BVY 10-050

BVY 10-058

17

Enhance the Periodic Surveillance and Preventive Maintenance Program to assure that the effects of aging will be managed as described in LRA Section B.1.22, with the exception of SSCs which have been abandoned.

March 21, 2012

B.1.22

BVY 06-009

20

Enhance the Structures Monitoring Program to specify that process facility crane rails and girders, condensate storage tank (CST) enclosure, CO2 tank enclosure, N2 tank enclosure and restraining wall, CST pipe trench, diesel generator cable trench, fuel oil pump house, service water pipe trench, man-way seals and gaskets, and hatch seals and gaskets are included in the program.

March 21, 2012

B.1.27.2

BVY 06-009


VYNPS DSAR


ITEM

COMMITMENT

IMPLEMENTATION

SCHEDULE

LRA Section

SOURCE

22

Guidance for performing structural examinations of elastomers (seals and gaskets) to identify cracking and change in material properties (cracking when manually flexed) will be enhanced in the Structures Monitoring Program procedure.

March 21, 2012

B.1.27.2

BVY 06-009

24

System walkdown guidance documents will be enhanced to perform periodic system engineer inspections of systems in scope and subject to aging management review for license renewal in accordance with 10 CFR 54.4 (a)(1) and (a)(3). Inspections shall include areas surrounding the subject systems to identify hazards to those systems. Inspections of nearby systems that could impact the subject system will include SSCs that are in scope and subject to aging management review for license renewal in accordance with 10 CFR 54.4 (a)(2).

March 21, 2012

B.1.28

BVY 06-009

28

Revise program procedures to indicate that the Instrument Air Program will maintain instrument air quality in accordance with ISA S7.3

March 21, 2012

B.1.16

BVY 06-009

30

Revise System Walkdown Program to specify CO2 system inspections every 6 months.

March 21, 2012

B.1.28

BVY 06-009

31

Revise Fire Water System Program to specify annual fire hydrant gasket inspections and flow tests.

March 21, 2012

B.1.12.2

BVY 06-009

33

Include within the Structures Monitoring Program provisions that will ensure an engineering evaluation is made on a periodic basis (at least once every five years) of groundwater samples to assess aggressiveness of groundwater to concrete. Samples will be monitored for sulfates, pH and chlorides.

March 21, 2012

B.1.27

BVY 06-009


VYNPS DSAR


ITEM

COMMITMENT

IMPLEMENTATION

SCHEDULE

LRA Section

SOURCE

34

Implement the Bolting Integrity Program.

Details are provided in a LRA Amendment 16, Attachment 2 and LRA Amendment 23, Attachment 5.

March 21, 2012

B.1.31

BVY 06-058

BVY 07-003

BVY 06-091

35

Provide within the System Walkdown Training Program a process to document biennial refresher training of Engineers to demonstrate inclusion of the methodology for aging management of plant equipment as described in EPRI Aging Assessment Field Guide or comparable instructional guide.

March 21, 2012

B.1.28

BVY 06-058

46 Enhance the Diesel Fuel Monitoring Program to specify that fuel oil in the fire pump diesel storage (day) tank will be analyzed according to ASTM D975 and for particulates per ASTM D2276.

March 21, 2012 B.1.9 BVY 07-018

BVY 10-069

BVY 11-007

47

Enhance the Diesel Fuel Monitoring Program to specify that fuel oil in the common portable fuel oil storage tank will be analyzed according to ASTM D975, per ASTM D2276 for particulates, and per ASTM D2709 for water and sediment.

March 21, 2012

B.1.9

BVY 07-018

BVY 10-069

BVY 11-007

52

Implement the Neutron Absorber Monitoring Program as described in LRA Section B.1.31.

Test one coupon prior to the PEO to measure B-10 areal density and assess the geometric and physical condition of the tested coupon. If coupons are not able to be retrieved and tested or if coupons cannot be demonstrated representative of the Boral in the Holtec racks, then perform neutron attenuation testing using in-situ methods, as described in BVY 11-010, (BADGER or blackness testing method) prior to the end of 2014.

March 21, 2012

B.1.31

BVY 10-052

BVY 10-058

BVY 11-013

VYNPS DSAR Revision 1 G.2-1 of 40

APPENDIX G.2 CURRENT ON-SITE METEOROLOGICAL PROGRAM TABLE OF CONTENTS Section Title Page

G.2.1 Introduction......................................................... 4

G.2.2 Description of the Monitoring Program................................ 4

G.2.3 Results.............................................................. 5

G.2.3.1 Wind Data ................................................. 5

G.2.3.2 Inversion Data ............................................ 6


CURRENT ON-SITE METEOROLOGICAL PROGRAM LIST OF TABLES Table No. Title

G.2.1 Meteorological Data Recovery Rates for 1980

G.2.2 Joint Frequency Distribution of Wind Speed, Wind Direction,

and Stability Class (Stability Based on 295-33 Foot Delta-

T)(35.0 FT Wind Data)

G.2.3 Joint Frequency Distribution of Wind Speed, Wind Direction,

and Stability Class (Stability Based on 295-33 Foot Delta-T)

(297 foot level FT Wind Data)

G.2.4 Wind Direction Persistence Summary (35 foot level)

G.2.5 Wind Direction Persistence Summary (297 foot level)

G.2.6 Inversion Persistence Summary (198-33 foot Delta T)

G.2.7 Inversion Persistence Summary (295-33 foot Delta T)


CURRENT ON-SITE METEOROLOGICAL PROGRAM LIST OF FIGURES Reference Figure No. Drawing No. Title G.2-1 Location of Primary and Backup

Meteorological Towers

G.2-2 "Spring Wind Rose (35 foot level)

March 1980 May 1980"

G.2-3 Summer Wind Rose (35 foot level) June 1980

August 1980

G.2-4 Autumn Wind Rose (35 foot level) September

1980 November 1980

G.2-5 Winter Wind Rose (35 foot level) January

1980 February 1980; December 1980

G.2-6 Annual Wind Rose (35 foot level) January

1980 December 1980

G.2-7 Spring Wind Rose (297 foot level) March

1980 May 1980

G.2-8 Summer Wind Rose (297 foot level) June

1980 August 1980

G.2-9 Autumn Wind Rose (297 foot level)

September 1980 November 1980

G.2-10 Winter Wind Rose (297 foot level) January

1980 February 1980; December 1980

G.2-11 Annual Wind Rose (297 foot level) January

1980 December 1980


G.2 CURRENT ON-SITE METEOROLOGICAL PROGRAM

G.2.1 Introduction

The On-Site Meteorological Data Collection Program was upgraded in early 1976

to meet the intent of Revision 0 of Regulatory Guide 1.23. This report

describes the current on-site monitoring program and presents wind and

stability data summaries for one full year of operation; January 1, 1980

through December 31, 1980. A discussion of the data summaries is included,

and a comparison is made between data collected by the initial monitoring

program (August 1967 - July 1968) and data collected by the current monitoring

program (January 1980 - December 1980). It is concluded that results from

both monitoring programs are compatible, and that both programs produced data

bases which are representative of site meteorology.

G.2.2 Description of the Monitoring Program

The current Meteorological Monitoring System includes both a primary and a

backup system. The primary system utilizes a guyed 305-foot tower located

on-site as shown in Figure G.2-1. The parameters measured on the tower

include the following:

Wind speed at the 35-foot and 297-foot levels (3-cup anemometer sensors)

Wind direction at the 35-foot and 297-foot levels (airfoil vane sensors)

Temperature at the 33-foot level (RTD located in a radiation-shielded

aspirator)

Delta-temperature between the 198-33 foot and between the 295-33 foot

levels (RTDs located in radiation-shielded aspirators)

In addition, both precipitation and barometric pressure are measured on the

ground.

The translator cards for the tower sensors are located in an instrument shed

near the base of the tower. The analog output is then digitally transmitted

to one of the plant process computer's remote data acquisition terminals. The

plant process computer periodically scans each parameter and then digitally

compiles and records the data as 15-minute averages. The 15-minute averages

are available for display in the Control Room. A digital recorder, located in

the relay house is also utilized as an auxiliary data logger. The entire

system is currently supplied by redundant power sources.

A 140-foot guyed tower used previously for meteorological monitoring was


reinstrumented during 1980 to serve as a backup tower. This tower's location

is also shown in Figure G.2-1. The parameters measured on the backup tower

include:

Wind speed at the 100-foot level (3-cup anemometer sensor)

Wind direction at the 100-foot level (airfoil vane sensors)

Delta-temperature between the 135-33 foot levels (RTDs located in

radiation-shielded aspirators)

The translator cards for the tower sensors are located in an instrument shed

near the base of the tower. The analog is then transmitted digitally to the

Control Room and sampled by the plant process computer. The signals are also

captured by a digital recorder, which is also utilized as an auxiliary data

logger.

G.2.3 Results

The primary meteorological system was the data source for the 1980 data

summaries which follow. The digital recording system was the principal data

collection mechanism. The data base consists of hourly data where the first

15-minute average collected each hour is used to represent the hour. The

analog strip chart recorders were utilized as backup data loggers for quality

control analysis, and data from the strip charts were used to fill in gaps in

the digital data base. The resulting data recovery rates, which are well

above the Regulatory Guide 1.23 goal of 90%, are presented in Table G.2.1.

G.2.3.1 Wind Data

Seasonal and annual wind roses for all stabilities combined from each tower

level are presented in Figures G.2-2 through G.2-11. Annual three-way joint

frequency summaries of wind speed, wind direction, and stability (stability

defined as a function of delta-temperature per Rev. 0 of Regulatory Guide

1.23) for both tower levels are also presented in Tables G.2.2 and G.2.3.

Comparison with the initial monitoring program results shows good agreement.

The seasonal and annual wind roses from each tower level continue to

illustrate the channelling effect of the Connecticut River Valley upon the

winds. Winds generally blow with the highest frequency from the NNW and SSE.

The annual average wind speeds for the current monitoring program were 6.2 mph

for the 35-foot level and 9.2 mph for the 297-foot level. These average wind

speeds compare well with the 140-foot level annual average wind speed of 7.5

mph for the initial monitoring program, if one considers the expected

variation of wind speed with height from the surface.


Wind direction persistence summaries for 1980 are presented in Tables G.2.4

and G.2.5. Summarizing the persistence information, 74.9% of the 35-foot and

69.5% of the 297-foot cases during 1980 were one-hour events, and only one of

the 35-foot and nine of the 297-foot persistence cases were more than 15 hours

long. This compares well with the initial monitoring program which found that

69.0% of its persistence cases were one-hour events and only five persistence

cases were more than 15 hours long.

G.2.3.2 Inversion Data

The annual stability class frequency distribution for the initial and current

monitoring programs compare as follows:

Stability Case Initial Program Current Program

140-5 Foot Delta-T

198-33 Foot Delta-T

(Ref. Table G.2.2)

295-33 Foot Delta-T

(Ref. Table G.2.3)

Unstable (A,B,C) 25.4% 14.9% 8.1%

Neutral (D) 25.1% 37.1% 43.0%

Stable (E,F,G) 49.5% 48.0% 48.9%

Differences in the above frequency distributions can be expected due to the

differences in measurement heights. Measurement heights are important because

the largest temperature gradients occur near the ground due to surface heating

during the day and radiative cooling at night.

Inversions, defined as a positive value of delta-temperature, occurred 33.6%

and 32.7% of the time during 1980 for the 198-33 foot and 295-33 foot

delta-temperature measurements, respectively. Inversions occurred at the site

nearly 39% of the time during the initial data collection period.

Tables G.2.6 and G.2.7 present the annual inversion persistence summaries for

1980. The longest inversion measured during the year lasted 38 hours.


TABLE G.2.1 Meteorological Data Recovery Rates for 1980

Parameter Possible Hours Usable Hours Recovery Rate

35-Foot Wind Speed 8784 8723 99.3%

297-Foot Wind Speed 8784 8721 99.3%

35-Foot Wind Direction 8784 8763 99.8%

297-Foot Wind Direction 8784 8573 97.6%

33-Foot Temperature 8784 8734 99.4%

198-33 Foot Delta-T 8784 8703 99.1%

295-33 Foot Delta-T 8784 8710 99.2%

Precipitation 8784 8474 96.5%

Solar Radiation 8784 8768 99.8%

Composite (35' WS, 35' WD, 198-33' DT)

8784

8659

98.6%

Composite (297' WS, 297' WD, 295-33' DT)

8784

8474

96.5%


Table G.2.2

Joint Frequency Distribution of Wind Speed Wind Direction, and Stability Class

(Stability Based on 198-33 Foot Delta-T) 35.0 FT WIND DATA STABILITY CLASS A CLASS FREQUENCY (PERCENT) = 6.14 WIND DIRECTION FROM

SPEED (MPH) N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW VRBL TOTAL

CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

2 .38 .02

7 1.32 .08

6 1.13 .07

7 1.32 .08

71.32.08

101.88.12

91.69.10

5.94.06

4.75.05

00.000.00

2 .38 .02

2.38.02

1.19.01

1.19.01

1.19.01

3.56.03

00.000.00

67 12.59

.77

4-7 (1) (2)

14 2.63 .16

10 1.88 .12

13 2.44 .15

19 3.57 .22

285.26.32

173.20.20

142.63.16

203.76.23

71.32.08

4.75.05

2 .38 .02

1.19.01

1.19.01

3.56.03

173.20.20

366.77.42

00.000.00

206 38.72 2.38

8-12 (1) (2)

16 3.01 .18

3 .56 .03

2 .38 .02

0 0.00 0.00

5.94.06

4.75.05

142.63.16

458.46.52

193.57.22

5.94.06

1 .19 .01

2.38.02

5.94.06

4.75.05

224.14.25

6211.65

.72

00.000.00

209 39.29 2.41

13-18 (1) (2)

4 .75 .05

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

1.19.01

00.000.00

00.000.00

1 .19 .01

1.19.01

3.56.03

1.19.01

112.07.13

275.08.31

00.000.00

49 9.21 .57

19-24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

1.19.01

00.000.00

00.000.00

00.000.00

1 .19 .01

GT 24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

ALL SPEEDS (1) (2)

36 6.77 .42

20 3.76 .23

21 3.95 .24

26 4.89 .30

407.52.46

315.83.36

376.95.43

7113.35

.82

305.64.35

91.69.10

6 1.13 .07

61.13.07

101.88.12

101.88.12

519.59.59

12824.061.48

00.000.00

532 100.00

6.14

(1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH)


TABLE G.2.2 35.0 FT WIND DATA STABILITY CLASS B CLASS FREQUENCY (PERCENT) = 4.09 WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

0 0.00 0.00

7 1.98 .08

2 .56 .02

4 1.13 .05

51.41.06

51.41.06

51.41.06

41.13.05

3.85.03

2.56.02

1 .28 .01

1.28.01

1.28.01

1.28.01

1.28.01

3.85.03

00.000.00

45 12.71

.52

4-7 (1) (2)

14 3.95 .16

5 1.41 .06

3 .85 .03

8 2.26 .09

195.37.22

82.26.09

113.11.13

92.54.10

51.41.06

1.28.01

3 .85 .03

2.56.02

1.28.01

61.69.07

123.39.14

257.06.29

00.000.00

132 37.29 1.52

8-12 (1) (2)

9 2.54 .10

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

1.28.01

1.28.01

2.56.02

71.98.08

102.82.12

00.000.00

4 1.13 .05

41.13.05

51.41.06

113.11.13

123.39.14

349.60.39

00.000.00

100 28.25 1.15

13-18 (1) (2)

2 .56 .02

1 .28 .01

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

2.56.02

3.85.03

00.000.00

2 .56 .02

1.28.01

113.11.13

92.54.10

123.39.14

287.91.32

00.000.00

71 20.06

.82

19-24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

2.56.02

1.28.01

1.28.01

2.56.02

00.000.00

6 1.69 .07

GT 24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

ALL SPEEDS (1) (2)

25 7.06 .29

13 3.67 .15

5 1.41 .06

12 3.39 .14

257.06.29

143.95.16

185.08.21

226.21.25

215.93.24

3.85.03

10 2.82 .12

82.26.09

205.65.23

287.91.32

3810.73

.44

9225.991.06

00.000.00

354 100.0

4.09



TABLE G.2.2 35.0 FT WIND DATA STABILITY CLASS C CLASS FREQUENCY (PERCENT) = 4.64 WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

5 1.24 .06

5 1.24 .06

2 .50 .02

3 .75 .03

3.75.03

61.49.07

71.74.08

61.49.07

2.50.02

1.25.01

2 .50 .02

3.75.03

1.25.01

00.000.00

51.24.06

41.00.05

00.000.00

55 13.68

.64

4-7 (1) (2)

18 4.48 .21

5 1.24 .06

6 1.49 .07

6 1.49 .07

122.99.14

102.49.12

61.49.07

2.50.02

3.75.03

3.75.03

3 .75 .03

41.00.05

1.25.01

71.74.08

122.99.14

245.97.28

00.000.00

122 30.35 1.41

8-12 (1) (2)

9 2.24 .10

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

2.50.02

00.000.00

3.75.03

204.98.23

143.48.16

41.00.05

9 2.24 .10

61.49.07

102.49.12

153.73.17

102.49.12

235.72.27

00.000.00

125 31.09 1.44

13-18 (1) (2)

7 1.74 .08

1 .25 .01

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

1.25.01

1.25.01

1.25.01

00.000.00

1 .25 .01

41.00.05

112.74.13

215.22.24

163.98.18

225.47.25

00.000.00

86 21.39

.99

19-24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

1.25.01

1.25.01

2.50.02

3.75.03

61.49.07

00.000.00

13 3.23 .15

GT 24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

1.25.01

00.000.00

00.000.00

1 .25 .01

ALL SPEEDS (1) (2)

39 9.70 .45

11 2.74 .13

8 1.99 .09

9 2.24 .10

174.23.20

163.98.18

174.23.20

297.21.33

204.98.23

81.99.09

15 3.73 .17

184.48.21

245.97.28

4511.19

.52

4711.69

.54

7919.65

.91

00.000.00

402 100.00

4.64



TABLE G.2.2

35.0 FT WIND DATA STABILITY CLASS D CLASS FREQUENCY (PERCENT) = 37.14 WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

39 1.21 .45

15 .47 .17

22 .68 .25

26 .81 .30

401.24.46

23.72.27

411.27.47

391.21.45

331.03.38

321.00.37

28 .87 .32

421.31.49

24.75.28

421.31.49

521.62.60

601.87.69

00.000.00

558 17.35 6.44

4-7 (1) (2)

106 3.30 1.22

42 1.31 .49

28 .87 .32

20 .62 .23

351.09.40

571.77.66

1013.141.17

1454.511.67

652.02.75

15.47.17

22 .68 .25

28.87.32

441.37.51

541.68.62

1123.481.29

2317.182.67

00.000.00

1105 34.36 12.76

8-12 (1) (2)

74 2.30 .85

16 .50 .18

14 .44 .16

9 .28 .10

12.37.14

25.78.29

21.65.24

922.861.06

802.49.92

16.50.18

20 .62 .23

29.90.33

1304.041.50

1304.041.50

1193.701.37

1705.291.96

00.000.00

957 29.76 11.05

13-18 (1) (2)

34 1.06 .39

1 .03 .01

3 .09 .03

0 0.00 0.00

00.000.00

1.03.01

1.03.01

7.22.08

14.44.16

1.03.01

6 .19 .07

8.25.09

892.771.03

1093.391.26

1073.331.24

1263.921.46

00.000.00

507 15.76 5.86

19-24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

1.03.01

2.06.02

00.000.00

0 0.00 0.00

3.09.03

12.37.14

13.40.15

28.87.32

24.75.28

00.000.00

83 2.58 .96

GT 24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

4.12.05

2.06.02

00.000.00

00.000.00

6 .19 .07

ALL SPEEDS (1) (2)

253 7.87 2.92

74 2.30 .85

67 2.08 .77

55 1.71 .64

872.711.00

1063.301.22

1645.101.89

2848.833.28

1946.032.24

641.99.74

76 2.36 .88

1103.421.27

2999.303.45

35210.954.07

42013.064.85

61119.007.06

00.000.00

3216 100.00 37.14



TABLE G.2.2

35.0 FT WIND DATA STABILITY CLASS E CLASS FREQUENCY (PERCENT) = 30.70 WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

50 1.88 .58

26 .98 .30

30 1.13 .35

22 .83 .25

311.17.36

351.32.40

421.58.49

592.22.68

1043.911.20

1304.891.50

143 5.38 1.65

1224.591.41

1505.641.73

1274.781.47

1324.971.52

993.721.14

00.000.00

1302 48.98 15.04

4-7 (1) (2)

38 1.43 .44

16 .60 .18

1 .04 .01

7 .26 .08

16.60.18

321.20.37

602.26.69

873.271.00

853.20.98

281.05.32

29 1.09 .33

421.58.49

792.97.91

913.421.05

1636.131.88

1666.251.92

00.000.00

940 35.36 10.86

8-12 (1) (2)

16 .60 .18

1 .04 .01

0 0.00 0.00

0 0.00 0.00

2.08.02

6.23.07

18.68.21

371.39.43

391.47.45

3.11.03

5 .19 .06

8.30.09

521.96.60

441.66.51

501.88.58

672.52.77

00.000.00

348 13.09 4.02

13-18 (1) (2)

3 .11 .03

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

1.04.01

6.23.07

7.26.08

00.000.00

0 0.00 0.00

1.04.01

4.15.05

7.26.08

17.64.20

17.64.20

00.000.00

63 2.37 .73

19-24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

2.08.02

1.04.01

00.000.00

0 0.00 0.00

00.000.00

1.04.01

00.000.00

1.04.01

00.000.00

00.000.00

5 .19 .06

GT 24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

ALL SPEEDS (1) (2)

107 4.03 1.24

43 1.62 .50

31 1.17 .36

29 1.09 .33

491.84.57

732.75.84

1214.551.40

1917.192.21

2368.882.73

1616.061.86

177 6.66 2.04

1736.512.00

28610.763.30

26910.123.11

36313.664.19

34913.134.03

00.000.00

2658 100.00 30.70



TABLE G.2.2 35.0 FT WIND DATA STABILITY CLASS F CLASS FREQUENCY (PERCENT) = 13.87 WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

14 1.17 .16

10 .83 .12

8 .67 .09

10 .83 .12

10.83.12

9.75.10

181.50.21

433.58.50

514.25.59

1169.661.34

197 16.40 2.28

15713.071.81

1099.081.26

685.66.79

715.91.82

393.25.45

00.000.00

930 77.44 10.74

4-7 (1) (2)

6 .50 .07

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

4.33.05

3.25.03

3.25.03

8.67.09

151.25.17

181.50.21

38 3.16 .44

322.66.37

191.58.22

201.67.23

473.91.54

373.08.43

00.000.00

250 20.82 2.89

8-12 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

2.17.02

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

2.17.02

2.17.02

1.08.01

5.42.06

8.67.09

00.000.00

20 1.67 .23

13-18 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

1.08.01

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

1 .08 .01

19-24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

GT 24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

ALL SPEEDS (1) (2)

20 1.67 .23

10 .83 .12

8 .67 .09

10 .83 .12

141.17.16

141.17.16

211.75.24

514.25.59

665.50.76

13411.161.55

235 19.57 2.71

19215.992.22

13010.821.50

897.411.03

12310.241.42

846.99.97

00.000.00

1201 100.00 13.87



TABLE G.2.2 35.0 FT WIND DATA STABILITY CLASS G CLASS FREQUENCY (PERCENT) = 3.42 WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

5 1.69 .06

1 .34 .01

7 2.36 .08

5 1.69 .06

51.69.06

2.68.02

113.72.13

51.69.06

268.78.30

3311.15

.38

39 13.18

.45

258.45.29

268.78.30

217.09.24

124.05.14

134.39.15

00.000.00

236 79.73 2.73

4-7 (1) (2)

1 .34 .01

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

1.34.01

00.000.00

2.68.02

00.000.00

2.68.02

113.72.13

14 4.73 .16

72.36.08

41.35.05

41.35.05

41.35.05

31.01.03

00.000.00

53 17.91

.61

8-12 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

2.68.02

51.69.06

00.000.00

7 2.36 .08

13-18 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

19-24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

GT 24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

ALL SPEEDS (1) (2)

6 2.03 .07

1 .34 .01

7 2.36 .08

5 1.69 .06

62.03.07

2.68.02

134.39.15

51.69.06

289.46.32

4414.86

.51

53 17.91

.61

3210.81

.37

3010.14

.35

258.45.29

186.08.21

217.09.24

00.000.00

296 100.00

3.42



TABLE G.2.2 35.0 FT WIND DATA STABILITY CLASS ALL CLASS FREQUENCY (PERCENT) = 100.00 WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

115 1.33 1.33

71 .82 .82

77 .89 .89

77 .89 .89

1011.171.17

901.041.04

1331.541.54

1611.861.86

2232.582.58

3143.633.63

412 4.76 4.76

3524.074.07

3123.603.60

2603.003.00

2743.163.16

2212.552.55

00.000.00

3193 36.87 36.87

4-7 (1) (2)

197 2.28 2.28

78 .90 .90

51 .59 .59

60 .69 .69

1151.331.33

2271.471.47

1972.282.28

2713.133.13

1822.102.10

80.92.92

111 1.28 1.28

1161.341.34

1491.721.72

1852.142.14

3674.244.24

5226.036.03

00.000.00

2808 32.43 32.43

8-12 (1) (2)

124 1.43 1.43

20 .23 .23

16 .18 .18

9 .10 .10

22.25.25

38.44.44

58.67.67

2012.322.32

1621.871.87

28.32.32

39 .45 .45

51.59.59

2042.362.36

2052.372.37

2202.542.54

3694.264.26

00.000.00

1766 20.39 20.39

13-18 (1) (2)

50 .58 .58

3 .03 .03

3 .03 .03

0 0.00 0.00

00.000.00

1.01.01

3.03.03

17.20.20

25.29.29

1.01.01

10 .12 .12

16.18.18

1181.361.36

1471.701.70

1631.881.88

2202.542.54

00.000.00

777 8.97 8.97

19-24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

3.03.03

3.03.03

00.000.00

0 0.00 0.00

4.05.05

16.18.18

17.20.20

33.38.38

32.37.37

00.000.00

108 1.25 1.25

GT 24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

4.05.05

3.03.03

00.000.00

00.000.00

7 .08 .08

ALL SPEEDS (1) (2)

486 5.61 5.61

172 1.99 1.99

147 1.70 1.70

146 1.69 1.69

2382.752.75

2562.962.96

3914.524.52

6537.547.54

5956.876.87

4234.894.89

572 6.61 6.61

5396.226.22

7999.239.23

8189.459.45

106012.2412.24

136415.7515.75

00.000.00

8659 100.00 100.00



TABLE G.2.3

Joint Frequency Distribution of Wind Speed, Wind Direction, and Stability Class (Stability Based on 295-33 Foot Delta-T) 297.0 FT WIND DATA STABILITY CLASS A CLASS FREQUENCY (PERCENT) = .91 WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

0 0.00 0.00

0 0.00 0.00

2 2.60 .02

2 2.60 .02

00.000.00

00.000.00

33.90.04

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

11.30.01

00.000.00

11.30.01

00.000.00

9 11.69

.11

4-7 (1) (2)

2 2.60 .02

2 2.60 .02

0 0.00 0.00

0 0.00 0.00

11.30.01

11.30.01

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

22.60.02

56.49.06

00.000.00

13 16.88

.15

8-12 (1) (2)

1 1.30 .01

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

45.19.05

11.30.01

1215.58

.14

67.79.07

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

11.30.01

1012.99

.12

00.000.00

35 45.45

.41

13-18 (1) (2)

1 1.30 .01

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

11.30.01

67.79.07

33.90.04

00.000.00

0 0.00 0.00

00.000.00

00.000.00

22.60.02

11.30.01

45.19.05

00.000.00

18 23.38

.21

19-24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

22.60.02

00.000.00

2 2.60 .02

GT 24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

ALL SPEEDS (1) (2)

4 5.19 .05

2 2.60 .02

2 2.60 .02

2 2.60 .02

11.30.01

56.49.06

56.49.06

1823.38

.21

911.69

.11

00.000.00

0 0.00 0.00

00.000.00

00.000.00

33.90.04

45.19.05

2228.57

.26

00.000.00

77 100.00

.91



TABLE G.2.3

297.0 FT WIND DATA STABILITY CLASS B CLASS FREQUENCY (PERCENT) = 2.62 WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

1 .45 .01

1 .45 .01

1 .45 .01

2 .90 .02

1.45.01

1.45.01

2.90.02

2.90.02

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

1.45.01

00.000.00

00.000.00

00.000.00

12 5.41 .14

4-7 (1) (2)

12 5.41 .14

3 1.35 .04

1 .45 .01

2 .90 .02

83.60.09

83.60.09

41.80.05

2.90.02

1.45.01

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

94.05.11

00.000.00

50 22.52

.59

8-12 (1) (2)

8 3.60 .09

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

62.70.07

52.25.06

114.95.13

125.41.14

2.90.02

0 0.00 0.00

1.45.01

1.45.01

1.45.01

52.25.06

2310.36

.27

00.000.00

75 33.78

.89

13-18 (1) (2)

5 2.25 .06

3 1.35 .04

0 0.00 0.00

0 0.00 0.00

00.000.00

1.45.01

1.45.01

41.80.05

52.25.06

00.000.00

1 .45 .01

00.000.00

62.70.07

2.90.02

83.60.09

3013.51

.35

00.000.00

66 29.73

.78

19-24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

1.45.01

4.180.05

125.41.14

00.000.00

17 7.66 .20

GT 24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

2.90.02

00.000.00

2 .90 .02

ALL SPEEDS (1) (2)

26 11.71

.31

7 3.15 .08

2 .90 .02

4 1.80 .05

94.05.11

167.21.19

125.41.14

198.56.22

188.11.21

2.90.02

1 .45 .01

1.45.01

73.15.08

52.25.06

177.66.20

7634.23

.90

00.000.00

222 100.00

2.62



TABLE G.2.3

297.0 FT WIND DATA STABILITY CLASS C CLASS FREQUENCY (PERCENT) = 4.60 WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

8 2.05 .09

4 1.03 .05

1 .26 .05

2 .51 .02

3.77.04

41.03.05

41.03.05

41.03.05

00.000.00

1.26.01

0 0.00 0.00

1.26.01

2.51.02

00.000.00

2.51.02

3.77.04

00.000.00

39 10.00

.46

4-7 (1) (2)

13 3.33 .15

4 1.03 .05

6 1.54 .07

3 .77 .04

92.31.11

184.62.21

82.05.09

82.05.09

2.51.02

1.26.01

1 .26 .01

00.000.00

1.26.01

2.51.02

102.56.12

184.62.21

00.000.00

104 26.67 1.23

8-12 (1) (2)

9 2.31 .11

5 1.28 .06

0 0.00 0.00

0 0.00 0.00

1.26.01

51.28.06

41.03.05

215.38.25

41.03.05

41.03.05

0 0.00 0.00

00.000.00

1.26.01

92.31.11

123.08.14

287.18.33

00.000.00

103 26.41 1.22

13-18 (1) (2)

8 2.05 .09

1 .26 .01

0 0.00 0.00

0 0.00 0.00

00.000.00

1.26.01

00.000.00

51.28.06

41.03.05

1.26.01

0 0.00 0.00

1.26.01

82.05.09

112.82.13

92.31.11

379.49.44

00.000.00

86 22.05 1.01

19-24 (1) (2)

2 .51 .02

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

3.77.04

51.28.06

82.05.09

266.67.31

00.000.00

44 11.28

.52

GT 24 (1) (2)

2 .51 .02

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

1.26.01

1.26.01

2.51.02

82.05.09

00.000.00

14 3.59 .17

ALL SPEEDS (1) (2)

42 10.77

.50

14 3.59 .17

7 1.79 .08

5 1.28 .06

133.33.15

287.18.33

164.10.19

389.74.45

102.56.12

71.79.08

1 .26 .01

2.51.02

164.10.19

287.18.33

4311.03

.51

12030.771.42

00.000.00

390 100.00

4.60



TABLE G.2.3

297.0 FT WIND DATA STABILITY CLASS D CLASS FREQUENCY (PERCENT) = 43.03 WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

34 .93 .40

24 .66 .28

23 .63 .27

26 .71 .31

23.63.27

401.10.47

481.32.57

22.60.26

21.58.25

12.33.14

5 .14 .06

10.27.12

11.30.13

5.14.06

16.44.19

421.15.50

00.000.00

362 9.93 4.27

4-7 (1) (2)

59 1.62 .70

29 .80 .34

21 .58 .25

19 .52 .22

28.77.33

381.04.45

902.471.06

952.611.12

501.37.59

15.41.18

11 .30 .13

4.11.05

12.33.14

20.55.24

481.32.57

1544.221.82

00.000.00

693 19.01 8.18

8-12 (1) (2)

105 2.88 1.24

44 1.21 .52

24 .66 .28

13 .36 .15

15.41.18

18.49.21

411.12.48

1704.662.01

932.551.10

22.60.26

29 .80 .34

32.88.38

812.22.96

1193.261.40

681.87.80

2035.572.40

00.000.00

1077 29.54 12.71

13-18 (1) (2)

82 2.25 .97

11 .30 .13

9 .25 .11

10 .27 .12

8.22.09

12.33.14

8.22.09

36.99.42

802.19.94

14.38.17

17 .47 .20

16.44.19

772.11.91

1925.272.27

1253.431.48

2175.952.56

00.000.00

914 25.07 10.79

19-24 (1) (2)

46 1.26 .54

3 .08 .04

3 .08 .04

0 0.00 0.00

00.000.00

1.03.01

1.03.01

3.08.04

15.41.18

1.03.01

5 .14 .06

3.08.04

26.71.31

992.721.17

972.661.14

1484.061.75

00.000.00

451 12.37 5.32

GT 24 (1) (2)

25 .69 .30

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

1.03.01

00.000.00

00.000.00

0 0.00 0.00

5.14.06

9.25.11

18.49.21

34.93.40

571.56.67

00.000.00

149 4.09 1.76

ALL SPEEDS (1) (2)

351 9.63 4.14

111 3.04 1.31

80 2.19 .94

68 1.87 .80

742.03.87

1092.991.29

1885.162.22

3278.973.86

2597.103.06

641.76.76

67 1.84 .79

701.92.83

2165.922.55

45312.425.35

38810.644.58

82122.529.69

00.000.00

3646 100.00 43.03



TABLE G.2.3

297.0 FT WIND DATA STABILITY CLASS E CLASS FREQUENCY (PERCENT) = 34.06 WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

88 3.05 1.04

41 1.42 .48

48 1.66 .57

49 1.70 .58

682.36.80

903.121.06

1103.811.30

541.87.64

19.66.22

18.62.21

15 .52 .18

7.24.08

18.62.21

19.66.22

391.35.46

722.49.85

00.000.00

755 26.16 8.91

4-7 (1) (2)

103 3.57 1.22

15 .52 .18

5 .17 .06

16 .55 .19

17.59.20

521.80.61

1314.541.55

1254.331.48

411.42.48

23.80.27

15 .52 .18

17.59.20

25.87.30

371.28.44

772.67.91

2679.253.15

00.000.00

966 33.47 11.40

8-12 (1) (2)

70 2.43 .83

6 .21 .07

0 0.00 0.00

2 .07 .02

7.24.08

14.49.17

411.42.48

873.011.03

702.43.83

15.52.18

15 .52 .18

17.59.20

531.84.63

602.08.71

622.15.73

2588.943.04

00.000.00

777 26.92 9.17

13-18 (1) (2)

31 1.07 .37

4 .14 .05

0 0.00 0.00

0 0.00 0.00

3.10.04

9.31.11

10.35.12

26.90.31

441.52.52

4.14.05

3 .10 .04

3.10.04

26.90.31

441.52.52

301.04.35

752.60.89

00.000.00

312 10.81 3.68

19-24 (1) (2)

4 .14 .05

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

2.07.02

14.49.17

00.000.00

0 0.00 0.00

00.000.00

4.14.05

7.24.08

9.31.11

22.76.26

00.000.00

62 2.15 .73

GT 24 (1) (2)

1 .03 .01

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

1.03.01

6.21.07

00.000.00

0 0.00 0.00

00.000.00

1.03.01

1.03.01

1.03.01

3.10.04

00.000.00

14 .49 .17

ALL SPEEDS (1) (2)

297 10.29 3.50

66 2.29 .78

53 1.84 .63

67 2.32 .79

953.291.12

1655.721.95

29210.123.45

29510.223.48

1946.722.29

602.08.71

48 1.66 .57

441.52.52

1274.401.50

1685.821.98

2187.552.57

69724.158.23

00.000.00

2886 100.00 34.06



TABLE G.2.3

297.0 FT WIND DATA STABILITY CLASS F CLASS FREQUENCY (PERCENT) = 13.03

WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

48 4.35 .57

42 3.80 .50

22 1.99 .26

21 1.90 .25

302.72.35

403.62.47

403.62.47

302.72.35

221.99.26

131.18.15

13 1.18 .15

4.36.05

171.54.20

111.00.13

232.08.27

474.26.55

00.000.00

423 38.32 4.99

4-7 (1) (2)

41 3.71 .48

8 .72 .09

2 .18 .02

5 .45 .06

131.18.15

343.08.40

665.98.78

393.53.46

252.26.30

121.09.14

16 1.45 .19

191.72.22

161.45.19

292.63.34

373.35.44

928.331.09

00.000.00

454 41.12 5.36

8-12 (1) (2)

12 1.09 .14

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

1.09.01

2.18.02

10.91.12

151.36.18

121.09.14

5.45.06

5 .45 .06

6.54.07

221.99.26

111.00.13

292.63.34

756.79.89

00.000.00

205 18.57 2.42

13-18 (1) (2)

1 .09 .01

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

4.36.05

00.000.00

0 0.00 0.00

00.000.00

2.18.02

4.36.05

00.000.00

8.72.09

00.000.00

19 1.72 .22

19-24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

1.09.01

00.000.00

00.000.00

2.18.02

00.000.00

3 .27 .04

GT 24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

ALL SPEEDS (1) (2)

102 9.24 1.20

50 4.53 .59

24 2.17 .28

26 2.36 .31

443.99.52

766.88.90

11610.511.37

847.61.99

635.71.74

302.72.35

34 3.08 .40

292.63.34

585.25.68

554.98.65

898.061.05

22420.292.64

00.000.00

1104 100.00 13.03



TABLE G.2.3

297.0 FT WIND DATA STABILITY CLASS G CLASS FREQUENCY (PERCENT) = 1.76

WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

2 1.34 .02

1 .67 .01

3 2.01 .04

1 .67 .01

1.67.01

42.68.05

21.34.02

32.01.04

42.68.05

32.01.04

3 2.01 .04

00.000.00

21.34.02

21.34.02

1.67.01

21.34.02

00.000.00

34 22.82

.40

4-7 (1) (2)

3 2.01 .04

1 .67 .01

0 0.00 0.00

0 0.00 0.00

00.000.00

21.34.02

74.70.08

53.36.06

74.70.08

32.01.04

3 2.01 .04

85.37.09

64.03.07

85.37.09

128.05.14

106.71.12

00.000.00

75 50.34

.89

8-12 (1) (2)

1 .67 .01

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

21.34.02

21.34.02

32.01.04

00.000.00

3 2.01 .04

32.01.04

117.38.13

32.01.04

21.34.02

74.70.08

00.000.00

37 24.83

.44

13-18 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

1.67.01

00.000.00

00.000.00

1 .67 .01

00.000.00

00.000.00

00.000.00

00.000.00

1.67.01

00.000.00

3 2.01 .04

19-24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

GT 24 (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

ALL SPEEDS (1) (2)

6 4.03 .07

2 1.34 .02

3 2.01 .04

1 .67 .01

1.67.01

64.03.07

117.38.13

117.38.13

149.40.17

64.03.07

10 6.71 .12

117.38.13

1912.75

.22

138.72.15

1510.07

.18

2013.42

.24

00.000.00

149 100.00

1.76



TABLE G.2.3

297.0 FT WIND DATA STABILITY CLASS ALL CLASS FREQUENCY (PERCENT) = 100.00

WIND DIRECTION FROM


CALM (1) (2)

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

00.000.00

0 0.00 0.00

C-3 (1) (2)

181 2.14 2.14

113 1.33 1.33

100 1.18 1.18

103 1.22 1.22

1261.491.49

1792.112.11

2092.472.47

1151.361.36

66.78.78

47.55.55

36 .42 .42

22.26.26

50.59.59

39.46.46

81.96.96

1671.971.97

00.000.00

1634 19.28 19.28

4-7 (1) (2)

233 2.75 2.75

62 .73 .73

35 .41 .41

45 .53 .53

76.90.90

1531.811.81

3063.613.61

2743.233.23

1261.491.49

54.64.64

46 .54 .54

48.57.57

60.71.71

961.131.13

1862.192.19

5556.556.55

00.000.00

2355 27.79 27.79

8-12 (1) (2)

206 2.43 2.43

55 .65 .65

24 .28 .28

15 .18 .18

24.28.28

49.58.58

1041.231.23

3183.753.75

2002.362.36

48.57.57

52 .61 .61

59.70.70

1691.991.99

2032.402.40

1792.112.11

6047.137.13

00.000.00

2309 27.25 27.25

13-18 (1) (2)

128 1.51 1.51

19 .22 .22

9 .11 .11

10 .12 .12

11.13.13

23.27.27

20.24.24

78.92.92

1401.651.65

19.22.22

22 .26 .26

20.24.24

1191.401.40

2553.013.01

1732.042.04

3724.394.39

00.000.00

1418 16.73 16.73

19-24 (1) (2)

52 .61 .61

3 .04 .04

3 .04 .04

0 0.00 0.00

00.000.00

1.01.01

1.01.01

5.06.06

29.34.34

1.01.01

5 .06 .06

3.04.04

34.40.40

1121.321.32

1181.391.39

2122.502.50

00.000.00

579 6.83 6.83

GT 24 (1) (2)

28 .33 .33

0 0.00 0.00

0 0.00 0.00

0 0.00 0.00

00.000.00

00.000.00

00.000.00

2.02.02

6.07.07

00.000.00

0 0.00 0.00

5.06.06

11.13.13

20.24.24

37.44.44

70.83.83

00.000.00

179 2.11 2.11

ALL SPEEDS (1) (2)

828 9.77 9.77

252 2.97 2.97

171 2.02 2.02

173 2.04 2.04

2372.802.80

4054.784.78

6407.557.55

7929.359.35

5676.696.69

1691.991.99

181 1.90 1.90

1571.851.85

4435.235.23

7258.568.56

7749.139.13

198023.3723.37

00.000.00

8474 100.00 100.00



TABLE G.2.4

Wind Direction Persistence Summary (35-foot level)

WIND DIRECTION PERSISTENCE SUMMARY = NUMBER OF OBSERVATIONS AND PERCENT PROBABILITY DIRECTION PERSISTENCE (HOURS) DIRECTION 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 GT.24 TOTAL

N 254 73

61 91

18 96

8 99

2 99

1 99

2100

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

346

NNE 120 82

22 97

3 99

1 100

0 0

0 0

00

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

146

NE 113 90

8 96

2 98

1 98

1 99

0 99

1100

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

126

ENE 104 84

17 98

2 99

1 100

0 0

0 0

00

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

124

E 153 80

31 96

6 99

0 99

1 99

1 100

00

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

192

ESE 177 83

25 95

6 98

4 100

0 0

0 0

00

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

212

SE 234 78

43 93

15 98

5 99

2 100

0 0

00

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

299

SSE 273 67

83 87

26 94

9 96

6 97

5 99

5100

0100

0100

1100

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

408

S 340 79

46 90

19 95

10 97

7 99

4 100

0100

0100

2100

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

428

SSW 304 84

47 97

6 99

3 100

0 0

0 0

00

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

360

SW 347 78

75 95

14 98

4 99

1 100

1 100

0100

1100

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

443

WSW 357 82

68 97

9 99

2 100

1 100

1 100

00

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

438

W 439 76

94 92

26 97

12 99

3 99

1 99

3100

0100

1100

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

579

WNW 443 75

108 93

16 96

18 99

4 94

2 100

1100

1100

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

593

NW 500 71

121 89

39 94

21 97

12 99

2 99

294

099

2100

1100

0100

0100

0100

0100

1100

00

0 0

00

00

00

00

00

00

00

00

701

NNW 384 57

133 77

67 87

27 91

17 94

18 97

497

398

298

799

199

199

199

1100

2100

1100

0 0

00

00

00

00

00

00

00

00

669

TOTAL 4542 982 274 126 57 36 18 5 7 9 1 1 1 1 3 1 0 0 0 0 0 0 0 0 0 6064


TABLE G.2.5

Wind Direction Persistence Summary

(297-foot level)

WIND DIRECTION PERSISTENCE SUMMARY = NUMBER OF OBSERVATIONS AND PERCENT PROBABILITY

DIRECTION PERSISTENCE (HOURS) DIRECTION 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 GT.24 TOTAL

N 351 69

85 86

33 93

13 95

12 97

5 98

499

199

1100

0100

0100

0100

1100

0100

0100

1100

0 0

00

00

00

00

00

00

00

00

507

NNE 169 83

25 95

8 99

0 99

1 100

1 100

00

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

204

NE 109 82

16 94

6 98

1 99

0 99

0 99

099

1100

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

133

ENE 134 88

16 99

0 99

1 99

1 100

0 0

00

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

152

E 160 83

24 96

3 97

4 99

1 100

0 0

00

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

192

ESE 264 81

52 97

5 98

3 99

2 100

1 100

00

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

327

SE 317 72

79 89

31 96

6 98

7 99

1 100

1100

0100

1100

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

443

SSE 301 66

74 83

33 99

19 94

7 96

10 98

499

399

2100

0100

0100

0100

1100

00

00

00

0 0

00

00

00

00

00

00

00

00

454

S 205 65

60 84

17 89

13 93

10 96

2 97

498

299

199

199

1100

0100

1100

00

00

00

0 0

00

00

00

00

00

00

00

00

317

SSW 140 91

12 99

1 99

1 100

0 0

0 0

00

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

154

SW 119 85

17 97

4 100

0 0

0 0

0 0

00

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

140

WSW 98 80

19 95

3 98

2 99

0 99

1 100

00

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

123

W 220 73

49 89

20 96

6 98

2 98

2 99

3100

00

00

00

00

00

00

00

00

00

0 0

00

00

00

00

00

00

00

00

302

WNW 246 62

79 82

31 89

24 95

7 97

3 98

098

298

399

199

099

1100

0100

0100

0100

0100

1 100

00

00

00

00

00

00

00

00

398

NW 336 69

81 86

33 93

14 96

10 98

2 98

6100

0100

1100

0100

0100

0100

0100

1100

00

00

0 0

00

00

00

00

00

00

00

00

484

NNW 336 47

147 68

77 78

38 84

35 89

21 91

1093

1094

795

997

597

698

398

499

099

099

1 99

099

299

099

1100

0100

0100

0100

3*100

715

TOTAL 3505 835 305 145 95 49 32 19 16 11 6 7 6 5 0 1 2 0 2 0 1 0 0 0 3 5045

* Of these three occurrences, one lasted 26 hours, the second lasted 34 hours, and the third lasted 36 hours.

TABLE G.2.6 Inversion Persistence Summary (198-33 foot Delta-T)

THE LONGEST INVERSION LASTED 38 HOURS OF THE LONGEST INVERSIONS, NUMBER 1 STARTED 18 HOURS INTO DAY 327 THIRD COLUMN DEFINES THE PERCENT PROBABILITY THAT IF AN INVERSION OCCURS, ITS DURATION WILL BE LESS THAN THE NUMBER OF HOURS SPECIFIED


Duration (hours)

Number of Observations

Percent Probability

1 146 27.39

2 60 38.65

3 49 47.84

4 34 54.22

5 31 60.04

6 22 64.17

7 26 69.04

8 20 72.80

9 20 76.55

10 27 81.61

11 24 86.12

12 26 90.99

13 21 94.93

14 12 97.19

15 4 97.94

16 3 98.50

17 2 98.87

18 1 99.06

19 1 99.25

20 2 99.62

21 0 99.62

22 0 99.62

23 1 99.81

24 0 99.81

25 0 99.81

26 0 99.81




Duration (hours)


Percent Probability

27 0 99.81

28 0 99.81

29 0 99.81

30 0 99.81

31 0 99.81

32 0 99.81

33 0 99.81

34 0 99.81

35 0 99.81

36 0 99.81

37 0 99.81

38 1 100.00




Duration (hours)


Percent Probability

1 140 27.94

2 50 37.92

3 49 47.70

4 25 52.69

5 27 58.08

6 19 61.88

7 31 68.06

8 18 71.66

9 17 75.05

10 23 79.64

11 25 84.63

12 29 90.42

13 15 93.41

14 15 96.41

15 4 97.21

16 2 97.60

17 3 98.20

18 4 99.00

19 1 99.20

20 1 99.40

21 0 99.40

22 0 99.40

23 1 99.60

24 0 99.60

25 1 99.80

26 0 99.80

27 0 99.80



TABLE G.2.7 (Continue)

Inversion Persistence Summary (295-33 foot Delta-T)

Duration (hours)


Percent Probability

28 0 99.80

29 0 99.80

30 0 99.80

31 0 99.80

32 0 99.80

33 0 99.80

34 0 99.80

35 0 99.80

36 0 99.80

37 0 99.80

38 1 100.00


Vermont Yankee


Location of Primary and Backup Meteorological Towers

Figure G.2-1


Vermont Yankee


Spring Wind Rose (35-Foot Level)

March 1980 – May 1980 Figure G.2-2


Vermont Yankee


Summer Wind Rose (35-Foot Level)

June 1980 – August 1980 Figure G.2-3


Vermont Yankee


Autumn Wind Rose (35-Foot Level)

September 1980 – November 1980 Figure G.2-4


Vermont Yankee


Winter Wind Rose (35-Foot Level)

January 1980 – February 1980; December 1980 Figure G.2-5


Vermont Yankee


Annual Wind Rose (35-Foot Level)

January 1980 – December 1980 Figure G.2-6


Vermont Yankee


Spring Wind Rose (297-Foot Level)

March 1980 – May 1980 Figure G.2-7


Vermont Yankee


Summer Wind Rose (297-Foot Level)

June 1980 – August 1980 Figure G.2-8


Vermont Yankee


Autumn Wind Rose (297-Foot Level)

September 1980 – November 1980 Figure G.2-9


Vermont Yankee


Winter Wind Rose (297-Foot Level)

January 1980 – February 1980; December 1980 Figure G.2-10


Vermont Yankee


Annual Wind Rose (297-Foot Level)

January 1980 - December 1980 Figure G.2-11

Vermont Yankee - Defueled Safety Analysis Report, Revision 1

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Transcript of Vermont Yankee - Defueled Safety Analysis Report, Revision 1