IIIII-!1111o - International Nuclear Information System (INIS)

IIIII-!1111o' __+_-,i!_++_!i!!._ii!Ii_.

lUll+IIIIImIlllloil-__+_-_

Volume II

EARLY SITE PERMIT DEMONSTRATION PROGRAM

PLANT PARAMETERS ENVELOPES

COMPARISON WITH RANGES OF VALUES FORFOUR HYPOTHETICAL SITES

Rev.0September1992

EARLY SITE PERMIT DEMONSTRATION PROGRAM

PLANT PARAMETERS ENVELOPES

Volume II- Comparison With Rangesof Values for Four Hypothetical Sites

Table of Contents

1.0 METHODOLOGY

2.0 DEVELOPMENT OF HYPOTHETICAL SITE VALUES

2.1 General

2.2 Structures

2.3 Cooling and Plant Water Use

2.4 Climatology/Meteorology

3.0 COMPARISON OF HYPOTHETICAL SITE VALUES WITH PLANTPARAMETERS ENVELOPE VALUES

3.1 General

3.2 Structures

3.3 Cooling and Plant Water Use

3.4 Climatology/Meteorology

4.0 CONCLUSIONS

5.0 REFERENCES (Cited in Volume II)

Appendix F SERCH Summaries Re£arding Siting

=l,=_,,o,_o,P,j.Es_ i Rev. 2March 1993

List of Tables

Table Title

2-1 Hypothetical Site CharacteristicsComparison With PlantParameters Envelope Values

=l_.,o._,.,o_Es_ ii Rev. 2March 1993

1.0 METHODOLOGY

The purpose of this volume is to report the results of the comparison of the ALWR

plant parameters envelope with values of site characteristics developed for four

hypothetical sites that generally represent conditions encounteredwithin the United

States. This effort is not intended to identify or address the suitability of any

existing site, site area, or region in the UnitedStates. Also included in this volume

is Appendix F, SERCH Summaries Regarding Siting.

To assign values for site characteristics used in testing the plant parameters

envelope, the following methodologywas implemented:

a. Values for a selected (partial)list of site characteristics were developed

for four "hypothetical sites." The values assigned to the characteristics

of the four hypotheticalsites were considered to be generally represen-

tative of ranges of values within the contiguous United States.

b. Ranges of characteristic data for the site values were established

primarily on information from existing nuclear power plants. The

reasons for this were:

1. Significant information was readily available to the public in

various documents and summaries.

2. The site-related informationwas extensiveand of high quality and

had been thoroughly reviewed previously by various agencies.

c. Numerical data ranges for the four site values were developed by

specialists in the followingfields: geology,seismology,geotechnicaland

civil engineering, meteorology, hydrology,and environmental/licensing.

21890kReport\DernonPtj,ESP_ 1-1 Rev. 0September 1992

d. Extremes and outliers were excluded when developing the numerical

data ranges from which values for the site characteristics were

assigned.

e. The collection of characteristic values listed for each of the hypothetical

sites are not reflective of any single existingsite. The values assigned

to each site were developed from existing nuclear plant site data based

on groupings, averages, limits, etc., and modified or supplemented by

nonnuclear plant-related data. Statistical analyses of existing data

were not performed to develop the site values.

21890_eportkDi_on Prj.E6P_ 1-2 Rev. 0September 1992

2.0 DEVELOPMENT OF HYPOTHETICAL SITE VALUES

This section discusses developmentof the hypotheticalsite characteristicvalues.

The four hypothetical sites were designated Sites A through D. The hypothetical

site values and the corresponding plant parameters envelopevalues are shown in

Table 2-1.

The hypothetical site characteristics to be used for comparison with the plant

parameters envelopes have been separated into four groups: General, Structures,

Cooling and Plant Water Use, and Climatology/Meteorology.

2.1 GENERAL

2.1.1 SITE SIZE (TOTAL)

The assignmentof total site acreage for each of the four hypothetical sites is based

on the tabulation of 74 nuclear power plant sites in Table 2.1 of Reference 1,

Volume 1. After outliers of less than 100 acres and more than 20,000 acres were

discarded, the remaining sites were subjectively divided into four groups, resulting

in the hypothetical site acreage. Neither the unit sizes nor the number of units at

existing sites were factored into this determination. The site areas range from 250

acres to 2,500 acres. It should be noted that the 2,500 acres assigned to Site D

do not include the area required within the site boundary for a cooling pond or lake.

The smallest site of 250 acres was assigned to Site C. The location of this site at

the shoreline allowed the exclusion area to extend over the water and, thus,

minimize the size of the site.

21890\Repod\DemonPrj.ESF_ 2-1 Rev. 0September 1992

Table 2-1HYPOTHETICAL SITE CHARACTERISTICS COMPARISON WITH PLANT PARAMETERS ENVELOPE VALUES

Site Characteristics Unit Typical/HypotheticalSite CharacteristicsValues PlantParametersEnvelope Plant Comments&Parameters Text Sections

SITE A SITE B SITE C SITE D Evolutio- Passive Generic Envelopenary Plant Plant SectionsPlant

GENERAL:

Site Size TOTAL acre 1,500 750 250 2,500+ 500 500 500 For 0.5 mile 2.1.1& 3.1.1Cooiing Pond radiusor Lake ExclusionZone

Require_nt SiteC, 250acresdoes

PlantFacility not include256 130 256 Area w/o Pond waterbody

areathatprovides

Plant Facility additionalArea with Pond areato meet

3,240 1,420 3,240 exclusionzonerequirement

MinimumDistanceto ft 4,000 2,500 2,000 3,500 2,640 2,640 2,640 9.6.6 2.1.2& 3.1.2Boundary

S_S: 2.2 & 3.2

BearingMaterialType Hard Rock Soft/Weathered Very DenseSoilsRock Dense/Firm 2.2.1& 3.2.1

Soi ls

MinimumBearingCapacity ksf i00+ 30 15 I0 15 15 15 1.9.2 2.2.2& 3.2.1(Static)

For EmbeddedStructures

MinimumShearWave fps I0,000 6,000 3,000 1,000 1,000 1,000 1,000 !.9.3 2.2.3& 3.2.2Velocity (At BearingLevel)

Depth to First Suitable ft 20 80 40 80 75 60 75 I.I 2.2.4 & 3.2.3BearincILayer

Rev. 0

September 1992

Table 2-1 (Sheet 2)

Site Characteristics Unit Typical/HypotheticalSite CharacteristicsValues Plant ParametersEnvelope Plant ConTnents&.... Parameters Text Sections

SITE A SITE B SITE C SITE D Evolutio- Passive Generic Envelopenary P1ant P1ant SectionsPlant

MaximumFloodLevel ft 20 10 15 5 i I I 1.8.1 2.2.5& 3.2.4Below Plant Grade

J

th to Groundwater ft 10 20 5 40 2 2 2 1.8.2 2.2.6& 3.2.5ow Plant Grade

Site-SpecificSSE-PGA "g" 0.15 0.25 0.20 0.15 0.3 0.3 0.3 1.5.2 2.2.7& 3.2.6

Liquefaction Potential l_one None None None None None None i.9.I 2.2.8& 3.2.7of Site-Specific SSE

COO[I[; AID PLAIIIWAIER 2.3 & 3.3USE:

CoolingWater Source Large River Stall River Estuary/ Reservoir/Ocean Lake 2.3.I

On-SiteRiver Flow, cfs 423,000s 17.650Average gpm 190 x I0 7._)x I06 - -

PotentialAvailable cfs 455 180 N/A 240 46.7 24.3 46.7 CoolingWater 2.3.2& 3.3.1MakeupWater - Monthly gpm 205,000 80,000 108,000 20,950 10,900 20,950 MakeupReqt.Average NaturalDraft

C.T.& UHSPond

Source Water °F 82 88 79 85 100 9I 91 2.6 2.3.3 & 3.3.2Temperature,maximum

10 Year-7b'y , Low cfs 96,000 4.000 - - 32.3 - 32.3 3.7.6,For UHS 2.3.4& 3.3.3River Flow gpm 43 x I0s I_ x I08 14,500 - 14,500 Once Through

! Con1incj

Net ConsumptiveWater cfs 150 60 N/A 80 34.3 16.5 34.3 2.8.9 2.3.5& 3.3.4Use, EvaporationLoss, gpm 67,000 27,000 36,000 15,400 2,400 15,400Average

Rev.0September 1992

Table 2-1 (Sheet 3)

Site Characteristics Unit Typical/HypotheticalSite ChAracteristicsValues Plant ParametersEnvelope Plant Comments&Parameters Text Sections

SITE A SITE B SITE C SITE D Evolutio- Passive Generic Envelopenary PIant Plant SectionsPlant

2.4, 3.4CLI_TOLOGY/METEOROLOGY

Ref. 4_ 5

General Description of Contir_ntal Continental Maritime Continental 2.4.1CIimate Low- Hid-Latitude High-

Latirude Latitude

__: 2.4.2& 3.4.1

ExtremeMaximum °F 112 115 108 114 115 115 115 2.1.5

ExtremeMinimum °F -13 -35 -18 -50 -40 -40 -40 2.1.7 VaIuesbelow-40 °F arefor extremel ocat ions

_TII_[ROISTURE 2.4.3 & 3.4.2CIITENI:

WINTERDESIGNDRY BULB °F 10 -15 -11 -31 -10 -I0 -I0 2.1.3 2.4.3& 3.4.2

TEMPERATURE (I_.)

SUI_R DESIGNDRY/WET °F 2.4.3& 3.4.2BULC TEMPS.:

- MaximumCoincident °F 102 / 76 102 / 72 97 / 76 99 / 71 100 / 77 100 / 77 100/77 2.1.1Dry/WetBu]b

- MaximumWet Bulb °F 81 81 8I 2.1.6Temp.Noncoincident

SummerDesign Wet Bulb °F 81 81 82 77 80 80 80 2.1.2Temp.(P,)-Noncoinci dent ==

Rev. 0

September 1992

Table 2-1 (Sheet 4)

Site Characteristics Unit Typical/Hypothetical Site Characteristics Values Plant Parameters Envelope Plant Comments&j Parameters Text Sections

SITE A SITE B SITE C SITE D EvoIutio- I Pass_,ve Generic Envelope

nary J PIant P1ant SectionsPlant

PHP PRECIPITATIOII: 2.4.4& 3.4.3

Ref.6

1 - Hour PMP inch 1g.4 18 ]9.4 19.4 Ig.4 19.4 19.4 !.4.1

5 - MinutePMP inch 6.2 6.1 6.2 6.2 6.2 6.2 6.2 1.4.1

GI_31JNDSHOW LOAD: 2.4.5& 3.4.4

50 - Year psf 15 40 70 100 Refs.7 & 11

ValuesaboveRI30FDESIGNSNOW LI]_LD: psf 50 50 50 1.4.2 50 psf for

Roof Design groundsnowSnow Load loadsare for

extremeNEJ U.S.,

northernGreatLakes,northernportionofAppaIachi anRange, andsomesectionsof mountainranges _n thewestern U.S.

BASICWIND SPEED: 2.4.6& 3.4.5

50 - Year mph gO 90 110 90 IIO IiO I10 !,!2.1 Refs.7 & 11

Rev. 0

September 1992

Table 2-1 (Sheet 5)

Site Characteristics Unit Typical/HypotheticalSite CharacteristicsValues PlantParametersEnvelope Plant Comments&Parameters TextSections

SITE A SITE B SITE C SITE D Evolutio- Passive Generic Envelopenary Plant P1ant SectionsPIant

l

DESIGNBASIS TORNADO: Assessmenton selectedvalues,in Vol. IT, Chapt.I, 2.4.7& 3.4.6App. B_ Sec. 2.1.2of URD. Refs.8 & g

MaximumWind Speed mph 360 360 360 360 300 260 260 1.11.4 2.4.7 & 3.4.6

MaximumRotationalSpeed mph 290 290 290 290 240 240 240 1.11.2 2.4.7& 3.4.6

TranslationalSpeed: 2.4.7& 3.4.6- Maximum n_)h 70 70 70 70 60 57 57 1.11.3- Minin_ n_)h 5 5 5 5

Radiusof Max. ft 150 150 150 150 150 453 150 1.11.6 2.4.7& 3.4.6RotationalSpeed

MaximumPressureDrop psi 3.0 3.0 3.0 3.0 2.0 1.46 1.46 1.11.1 2.4.7& 3.4.6

Rate of PressureDro_ _si/sec 2.0 2.0 2.0 2.0 1.2 0.27 0.27 1.11.7 2.4.7& 3.4.6

ATMOSPHERICDISPERSION 2.4.8& 3.4.7(CHI/Q):

AnnualAverage x/Q sec/m3 8.0E-06 8.0E-06 8.0E-06 8.0E-06 7.2E-05 2.0E-05 7.2E-05@ 9.2 2.4.8& 3.4.7@ 0.5 0.5 milesmiles

Accidentx/O 2.4.8& 3.4.7

0-2 HR sec/m3 1.2E-03 1.2E-03 1.2E-03 1.2E-03 1.0E-03 1.0E-O] 1.0E-03 9.1.1 2.4.8& 3.4.7

0-8 HR sec/m3 3.2E-04 3.2E-04 3.2E-Oa 3.2E-04 1.35E-04 1.35E-04 1.35E-04 9.1.2 2.4.8& 3.4.7

8-24 HR sec/m3 1.9E-04 1.9E-04 1.9E-04 1.9E-04 1.0E-04 1.0E-04 1.0E-04 9.1.5 2.4.8& 3.4.7

I-4 DAY sec/_ 4.0E-OS 4.0E-05 4.0E-05 4.0E-05 5.4E-05 5.4E-05 5.4E-05 9.1.3 2.4.8& 3.4.7 H

Rev. O

September 1992

Table 2-1 (Sheet 6)

Site Characteristics Unit Typical/HypotheticalSite CharacteristicsValues PlantParametersEnvelope Plant Comments&Parameters Text Sections

SITE A SITE B SITE C SITE D Evolutio- Passive Generic Envelopenary Plant Plant SectionsPlant

4-30 DAY sec/m3 2.5E-05 2.5E-05 2.5E-05 2.5E-05 2.2E-05 2.2E-05 2.2E-05 9.1.4 2.4.8& 3.4.7

Rev. 0September 1992

2.1.2 MINIMUM DISTANCE TO BOUNDARY

Based on the total site acreage, the minimum distance to the site boundary was

determined by arbitrarily assigning a simple geometric shape to each site with the

plant assumed to be at the center of that shape. The shapes assigned were a

square, an isosceles triangle, a circle, and an ellipse for Sites A through D,

respectively. The estimated minimum distances vary from 2,000 to 4,000 feet.

2.2 STRUCTURES

In general, placement of plant structures on a site requires horizontal and vertical

layouts that meetsafety requirements and take into account cost (construction and

operation/maintenance) and schedule (construction) considerations. After safety

requirements are defined, the relative merits of a variety of factors are evaluated.

These factors include plant location within the site property, the site yard grade

elevation, and the foundation excavation bottom elevation on a suitable bearing

layer. The standardized design fixes the vertical distance between the yard grade

and the foundation excavation bottom (embedment) for structures, the maximum

flood level, and the maximum level of groundwater. However, factors such as

setting of the yard grade with respect to earthwork requirements, depth to a

suitable foundation bearing layer, and elevation difference with the cooling water

source (pumping costs) are site-specific and are not addressed in the plant

parameters envelopes.

2.2.1 BEARING MATERIAL TYPE

The four bearing materials selected for the hypothetical sites represent the range

of typical geologic materials, in terms of load bearing properties, that would be

expected to be generally capable of supporting heavy plant structures. Each site

was assumed to have different subsurface conditions. The conditions assumed

21890\Report\DemonPrj.ESP\ 2-2 Rev. 0September 1992

are hard rock (Site A), soft/weathered rock (Site B), very dense/firm soils (Site C),

and dense granular soils (Site D).

2.2.2 MINIMUM BEARING CAPACITY (STATIC)

Soil and rock bearing capacity values for the four hypothetical sites were assigned

based on geotechnical experience with respect to the types of bearing matarials

selected. No details with respect to stratigraphic layering _ the bearing materials

were generated because the specifics of dynamic soil-structure interaction were

addressed on parametric bases in the standardized plant analyses. The allowable

bearing capacity values of 10 to 100 ksf shown in Table 2-1 are assumed to

include layering and considerations related to differential and total settlement of

structures duringand after construction. Inevaluating a specific site, estimates are

made of the elastic rebound or other types of heave of the excavation bottom due

to unloading of the overburden and subsequent settlement due to re-loading as

construction of the structures proceeds. Dynamic bearing capacities and the

effects of temporary lowering of the groundwater level as a result of construction

dewatering are also taken into account. Thus, the total process is very site

specific.

2.2.3 MINIMUM SHEAR WAVE VELOCITY (AT BEARING LAYER)

The shear wave velocity at the bearing layer for the different types of bearing

materials was assigned based on information available for these materials. The

values range from 1,000 to 10,000 feet per second (fps). The highest value

assigned was for hard rock and the lowest value was for dense soils.

21890_,_,_,_onP,j.ES_ 2--3 Rev. 0September 1992

2.2.4 DEPTH TO FIRST SUITABLE BEARING LAYER

This characteristic addresses the depth of excavation required to expose the first

soil or rock layer that would safeiy support the plant structures. Based on existing

nuclear power plant site experience,this depth generally varies from 20 to 80 feet,

depending on site geologic conditions and the plant foundation design. Fixed

embedment of a standard plant could result in the over-excavation of this suitable

bearing layer in order to reach the design depth for some structure foundations.

2.2.5 MAXIMUM FLOOD LEVEL BELOW PLANT GRADE

The plant yard grade is located above the maximum flood level to minimize plant

site flooding, facilitate the construction of penetrations, and achieve site drainage,

among other site-specific considerations. When considering the "dry yard" type

sites (Reference 2), most of the existing plants have yard grades that are in the

range of 5 to 20 feet above the maximum flood level.

2.2.6 DEPTH TO GROUNDWATER BELOW PLANT GRADE

Because plant yard grades are set considering site-specific groundwater levels,

most of the existing plant yard grades are 5 to 40 feet or more above the

regional/local high groundwater level. High groundwater levels impact flood

protection, wall design, penetrations and ducts, waterproofing, design hydrostatic

loads on buried structures, construction and operation dewatering, and other site

considerations.

21B_o_,_,_oop,j._s_ 2-4 Rev. 0September 1992

2.2.7 SITE-SPECIFIC SSE-PGA

Safe shutdown earthquake peak ground accelerations (SSE-PGA) of 0.15g to

0.25gwere subjectively assigned to thefour hypotheticalsites based onexperience

with existing plants.

2.2.8 LIQUEFACTION POTENTIAL OF SITE-SPECIFIC SSE

None of the hypothetical sites were considered to have subsurface conditions

favorable for the development of liquefaction during an SSE. Therefore, lique-

faction potential at all four sites was established as nonexistent for their respective

site-specific SSE.

2.3 COOLING AND PLANT WATER USE

2.3.1 COOLING WATER SOURCE

In assessing coolingwater aspectsfor the hypotheticalsites, Reference3 provided

data needed to establish the water sources. The following indicates, in percent of

the total number of plants, the water sources used for 108 nuclear power plants

that were operating, under construction, or proposed through about 1985:

Rivers - 48%

Oceans and estuaries - 20%

Lakes - 29%

Other- 3%

Based on this breakdown, Sites A and B were established as river sites, Site C as

an ocean or estuary site, and Site D as a reservoir/lake site, As a matter of

descriptive convenience, the two river sites were separated into a "large river site"

2,,,o,,=,_,,o,P,iEs_ 2-5 Rev. 0September 1992

(Site A) and a "small riversite" (Site B) based on the range of river flows. The

average riverflowsfor a largeriverand small riverwereset at 190 x 106gpm and

7.9 x 106gpm, respectively.

2.3.2 POTENTIAL AVAILABLE MAKEUP WATER - MONTHLY AVERAGE

The plant makeup water requirement variesdepending on the plant coolingwater

system selected for condenser cooling. The monthly average available makeup

water values for the four siteswere developed based on experience and judgment,

including experience with nuclear power plants located on rivers and lakes

throughout the United States. This site characteristic is not applicable to ocean

sites, which have once-through cooling. The monthly average values of potential

makeup water range from about 80,000 gpm for Site B to about 205,000 gpm for

Site A.

2.3.3 SOURCE WATER MAXIMUM TEMPERATURE

Maximum temperature rangesfor the coolingwater sources were developed based

on experience and judgment supplemented by available data contained in

References 1 and 2 for 66 nuclear plant sites. The source water maximum

temperature varies from 82 °F to 88 °F.

2.3.4 LOW RIVER FLOW

The average and 10-year 7-day low river flow values for the large river (Site A)

and the small river (Site B)were generated based on data for 25 nuclear plant sites

contained in Reference 1, supplemented by experience on nuclear and nonnuclear

projects.

21.o_._._,,.onP,j_8_ 2-6 Rev. 0September 1992

2.3.5 NET CONSUMPTIVE WATER USE, EVAPORATION LOSS

The numerical values for netconsumptivewater use at Sites A and B were based

on criteria limiting the consumptive water use to between 5 and 7 percent of the

average river flowsduring low flow seasons, excludingextreme low flows. This was

based on data given in Reference 3. The net consumptive water use varies from

27,000 gpm to 67,000 gpm, depending on the cooling system, source, and quality

of the makeup water and unit size. This site characteristic is not applicable to

ocean sites.

2.4 CLIMATOLOGY/METEOROLOGY

2.4.1 GENERAL DESCRIPTION OF CLIMATE

The climatic regions in which the four hypothetical sites may be located are

distinguished by subjective descriptions. Sites A, B, and D are broadly depicted,

respectively, as continental climatic types in low-latitude, mid-latitude, and high-

latitude regions within the contiguous United States. Site C is depicted as a

maritime climate ranging regionally from low to high latitudes along coastlines.

Climatological and meteorological conditions can be quite variable, even over

relatively small distances. Therefore, representative climatological and meteoro-

logical characteristics of the sites can most logically be defined using a range of

values, rather than a single value. However, only the applicable limiting value of

the range is provided in Table 2-1. These values were determined from data

summaries for locationseast of the 105ttl meridian because topographic influences

are significant farther to the west.

September 1992

2.4.2 TEMPERATURE

Extremetemperature values were determined from the summary plots of long-term

climatic data in Reference4 and represent the highest and lowest values on record

within the data summary period.

2.4.3 TEMPERATURE AND MOISTURE CONTENT

The site characteristics, coincident and noncoincident dry bulb and wet bulb

temperatures, are grouped under this headingbecause they are all used to support

cooling tower and HVAC design. These values were determined from statistical

data summaries in Reference 5.

The winter design dry bulb temperatures represent those values that are not

exceeded 1 percent of the time during the coldest three consecutive months (i.e.,

standardized as December, January, and February in the contiguous United

States).

Maximum coincidentdesigndry and wet bulb temperatures represent those dry

bulb values that are exceeded 1 percent of the time during the four warmest

consecutive months (i.e., standardized as June through September in the

contiguous United States). The mean coincident wet bulb temperatures are the

average of those values that occur coincidentally with the respective 1 percent

summer design dry bulb temperatures.

The maximum noncoincident summer design wet bulb temperature values

represent those values that are exceeded 1 percent of the time during the four

warmest consecutive months.

2,,_o_R,,o,_,,,o°p,iEs_ 2-8 Rev. 0September1992

The maximum wet bulb temperature has not been included because this value is

not routinelycollected by either private orgovernmental (National Weather Ser_rice)

data collection services and is not included in the engineeringdesign weather data

summaries referenced previously. Comparison of the summer design wet bulb

temperatures (1 percent exceedance, noncoincident) and maximum coincident dry

bulb - wet bulb temperature with the associated plant parameters envelope

provides an appropriate alternate measure of plant-site design compatibility.

2.4.4 PROBABLE MAXIMUM PRECIPITATION

Probable maximum precipitation (PMP) for durations of 1 hour and 5 minutes was

estimated from summary plots in Reference 6.

2.4.5 GROUND SNOW LOAD

Ground snow load values were estimated from summary plots (based on 50-year

return periods) in Reference 7. For the roof design load applications, the ground

snow load should be adjusted using values called importance factors, exposure

factor and thermal factor. Specific values for these factors depend on the category

of the structure being designed (safety or nonsafety related), the corresponding

recurrence interval (e.g., 100-year return period for safety-related structures), and

roof shape and design details as discussed in Reference 7.

2.4.6 BASIC WIND SPEED

Basic wind speed values (50-year return period) were estimated from summary

plots in Reference 7. For design basis applications, the basic wind speed is

required to be adjusted by values called importance factors. Specific values for

these importance factors depend on the category of the structure being designed

(safety or nonsafety related),the corresponding recurrence interval (e.g., 100-year

_1a_o_,,o,,_,,,,o,p,jEs_ 2-9 Rev. 0September 1992

return for safety-related structures), and the locationof the facility (i.e.,whether it

is within or beyond 100 miles of the hurricane-prone zone along coastlines).

The wind speed and precipitation discussions individually address the pertinent

plant-site interface features associated with hurricanes. Consequently,hurricanes

are not addressed as a separate and individual site parameter.

2.4.7 DESIGN BASIS TORNADO

The designbasis tornado(DBT)valuesassignedto thehypotheticalsites represent

the most stringent values specified in NRC Regulatory Guide 1.76 (i.e., for Region

I- all areas east of the 105th meridian) (References 8 and 9). The regulatory

guide allows the use of lessstringent values,although they must bedetermined by

a probabilistic analysis for a specific site region.

This methodology does not incorporate the recent EPRI settlement with the NRC

on tornado wind speed where the NRC has indicated that it will develop interim

regulatory guidance which merges Regions I and ii into a single region that retains

the characteristics of RegionII (Reference 10).

Regulatory guidance was used in this situation in lieu of historical tornadowind

speedandpressuredropdata becausethesedata are unavailable, The transient,

spatiallylimited,and highlydestructivenatureof tornadoeshas made collectionof

such data a technologicalchallengeyet to be adequatelymet.

2.4.8 ANNUAL AVERAGE AND ACCIDENT ATMOSPHERIC DISPERSION

COEFFICIENTS- X/Q VALUES

]'he development of annual average and accident atmospheric dispersion

coefficients (x/Q values) requiresthe collectionand reduction of site-specific

_,e_o_...o,,_o°.,jE._ 2-10 Rev. 0September 1992

meteorological data, identification of the release characteristics ilocation,

orientation,flowrate,temperature),and theapplicationof mathematicaldispersion

models(e.g., PAVAN, Reg. Guide 1.145;XOQDOQ, Reg. Guide 1.111). In lieuof

these complex,.!te-specificanalyses,informationprovided in FSARs for nuclear

power plant fucilities was used to identify typical site accident x/Q values.

Specifically,annualand accidentx/Q values from PWR and BWR FSARs were

collectedand tabulated (Table 2-1) to facilitate review and comparison. The

maximumx/Q values foreach applicableperiod (annualaverage,0-2 hours,2-8

hours, 8-24 hours, 1-4 days, 4-30 days) were selected for this tabulation.

Additionaljudgmentswere made inthosesituationswhere the reporteddispersion

coefficientswerebiased byunusualplantdesignfeaturesorby an exclusionor low

populationzone distancethat varied significantlyfrom the distancesassumed in

the plant parameterx/Q values.

Although these facilities could be arranged to fit the broad definitions used to

organizesite parametersintofour typicalregions,the site-specific nature of the

analysesused to generatethesevalueswouldinvalidateany attemptto regionally

evaluateand categorizethesecoefficients. Consequently,onlyone set of annual

averageand accidentx/Q valueswas developedfor the four hypotheticalsites.

_,a=o,f_,,_,._o,,.,,_sr_ 2-11 Rev, 0September t992

3.0 COMPARISON OF HYPOTHETICAL SITE VALUES WITHPLANT PARAMETERS ENVELOPE VALUES

A comparisonof the site characteristicvalues for the four hypotheticalsites in

Table 2-1 was made with the correspondingplant parametersenvelopevaluesto

determine their compatibility. Based on the values containedin Table 2-1, the

majorityof the plantparametersenvelopevaluesestablishedfor the evolutionary

plant, passive plant, and genericplant compare well with the range of values

encompassedby SitesA throughD. The followingsectionsdiscuss the resultsof

the evaluation of the Individualsitecharacteristicswith respect to the applicable

plant parametersenveloperequirements.

3.1 GENERAL

3.1.1 SITE SIZE (TOTAL)

The total site acreages for Sites A, B, _nd O exceed the plant parameters

envelopesvalues. The plantparametersenvelopevaluesare basedona circular

site that achievesthe 0.5-mile exclusionzone requirementby assumingthat the

plant is located in the center of the circle.

Site C does not appear to meet the plant parametersenvelopevalue for total site

acreage. This is because the site is assumed to be located at the edge of an

estuary,or ocean. The totalacreagerequiredforsiteslocatedadjacentto a river,

lake, estuary,or oceancan be minimizedif the waterbodyis includedwithinthe

required0.5-mile exclusionzone. Therefore,because Site C was assigned a

circularshape resultingin halfof theexclusionarea extendingoverthewater, only

250 acres, or half of the minimum500 acres, is requiredfor the landwardportion

of the site.

_,a=_..o,,_o,p,j_s_ 3-1 Rev. 0September 1992

3.1.2 MINIMUM DISTANCE TO BOUNDARY

The 0.5-mile exclusion zone incorporated into the plant parameters envelope

addressinga minimumdistance to the site boundarycomparesfavorablyto the

valuesgivenforSitesA andD. SitesBandC are marginallyacceptable,indicating

the need forearly verificationduringsitingconsiderations.

3.2 S.STRUCTURES

3.2.1 MINIMUM BEARING CAPACITY (STATIC)

The plant parametersenvelope of 15 ksf establishedfor the bearingcapacity of

embedded structures may be high for some soil conditions. Sites with poor

foundationconditionswill need to be determinedbased on site-specific field

investigations.

Bearingcapacityis rarelya problemfordeeplyembeddedstructureswithstructural

mat soil/rockfoundations.Formanyrocksites,allowablebearingcapacitymay be

related to the presence of features or propertiessuch as joints, voids, swell,

potential,etc. Eventhough"minimum"bearingcapacitymay be a goodapproach

to establishingsite parameters for a standard design, the comparisonof an

envelopevalueof bearingcapacityto actualsite valuesrequiresfurtherexplana-

tion.

As discussed in Subsection2.2.2, in assessingan actual site, a 15 ksf bearing

capacity(static)envelopewould needto be supplementedwithconsiderationsof

settlements(total, differential),excavation-relatedheave or rebound,dynamic

effects,and construction-relatedfactors, Eventhough,for mostsites,thisshould

notbe a problem,for somesites this may be unacceptable. Improvementsmay

be possiblebut costly. For somesoft, fine-tnrainedsoils (silts,clays),even with

_,,=,,,_,,,or,,j_s_ 3-2 Rev, 0September 1992

a bearingcapacitydemonstratedto beacceptable, otherfactorssuchas excessive

settlements may eliminate a site from consideration. For shallow foundations, the

15 ksf minimum bearing capacity requirement could be problematic. Thus, the

minimum static bearing capacity demand of 15 ksf, by itself, may be high,

particularly for foundations with shallow embedment.

3.2.2 MINIMUM SHEAR WAVE VELOCITY (AT BEARING LEVEL)

The minimumshear wave velocityof 1,000fps for the plant parameters envelope

compares well with the values assigned to the hypothetical sites.

3.2.3 DEPTH TO FIRST SUITABLE BEARING LAYER

The depth to the first suitable bearing layer is site dependent. At those sites where

the layer is encountered significantly above the maximum excavation depth

required due to plant embedment, additional cost will be incurred for overexcava-

tion, especially if the material is hard rock.

3.2.4 MAXIMUM FLOOD LEVEL BELOW PLANT GRADE

As discussed in Subsection 2.2.5, the selection of yard grade for a specific site

includes a consideration of flood levels. In this sense, the plant parameters

envelope (same as URD values) are conservative. A review of 85 existing nuclear

power plant sites that are "dry yard" sites (for flooding) shows that about a third of

the sites have flood levels 5 feet or less below grade, while another third of these

sites have maximum flood levels that are greater than 15 feet below yard grade.

The design flood level of 1 foot below grade is conservative and may result in

higher costs than needed at those sites where the maximum flood level is

significantly lower.

2_8_.._,_,,_oop,jEs_ 3-3 Rev. 0September 1992

3.2.5 DEPTH TO GROUNDWATER BELOW PLANT GRADE

As discussed in Subsection 2.2.6, the selection of yard grade for a specific site

includes a consideration of groundwater levels. In this sense, the plant design

groundwater level of 2 feet below grade may be conservative with regard to design

of structure exterior walls for hydrostatic loads.

3.2.6 SITE-SPECIFIC SSE-PGA

The SSEs assigned to Sites A through D meet the plant parameters envelope

SSE-PGA value of 0.3g. However, based on this value for the standard plant

design, some parts of the United States, where site-specific SSEs are likely to

exceed 0.3g,will be excluded from siting considerations. Those areas of the United

States that could be expectedto havesite-specific SSEs greater than 0.3g include:

the west coast; the area around New Madrid, Missouri; the Charleston, South

Carolina area; and parts of the northeastern United States. Thus, there is a need

to determine site-specific SSEs for candidate sites early in the site selection

process because the inability to meet this value would be considered unaccept-

able.

The NRC-Reactor Site Criteria sets forth the procedure for estimation of the SSE.

Where there is sufficient geologic and tectonic information, the SSE is determined

by evaluating the seismic capacities of capable faults. In the western mountain

region and, particularly, in the Pacific region, this is usually the case. Therefore,

the actual SSE acceleration would depend very heavilyon local geologic structures

in the site proximity.

In other areas, tectonic features are not as well defined, and the SSE is determined

by considering the historical seismicity in relevant tectonic provinces. The

maximum historical intensity in the tectonic provinces surrounding a site is used in

2,8_R_,_.-_oop,jEsF_ 3--4 Rev. 0September 1992

assessment of the SSE acceleration. The tectonic provinces are selected to be

large enough so that, even in the limited historical period available, an event of a

size similar to the SSE has most likely occurred.

3.2.7 LIQUEFACTION POTENTIAL OF SITE-SPECIFIC SSE

In the plant parameters envelopes, no potential for liquefaction is considered

mandatory for all candidate sites. Thus, there is a need to determine a site's

liquefaction potential early in the site selection process because the inability to

meet this requirement would be considered unacceptable.

3.3 COOLING AND PLANT WATER USE

The water use values for the ALWR plants are typically lower than the water

available for existing plants.

3.3.1 POTENTIAL AVAILABLE MAKEUP WATER - MONTHLY AVERAGE

The average makeupwater value from the plant parameters envelope is lower than

the average monthly flow at the hypothetical sites. This does not present any

limitation on plants to be located at the hypothetical sites.

3.3.2 SOURCE WATER MAXIMUM TEMPERATURE

The plant parameters envelope values for source water temperature are equal to

or higher than those of the hypothetical sites. Thus, they do not present any

limitation on the hypothetical sites.

21890\Re_:)ort_)ernonPrj,ESP_ 3-5 Rev. 0September 1992

3.3.3 LOW RIVER FLOW

With the assumed plant makeup water withdrawal limited to 5 percent of the

10-year, 7-day low flow of the river, the required river 10-year, 7-day flow should

be 218,000 gpm to 419,000 gpm. If the river flows are lower, on-site water

impoundment will be required to supplement plant makeup during low flow periods.

In this assessment, the UHS has been assumed to be an on-site pond with

30-day consumptive use water storage and other allowances as required.

In emergency situations, where the river is to be used as the UHSwith 100 percent

withdrawal, the 10-year, 7-day low river flow needs to be higher than 14,500gpm.

By limiting the withdrawal to 5 percent of river flow, the river flow required could be

as high as 2.9 x 105gpm.

Thus, the cooling and plant water availability are site-specific items and, therefore,

need to be reviewed considering the source of water and regulatory requirements.

3.3.4 NET CONSUMPTIVE WATER USE, EVAPORATION LOSS

The net evaporation loss fr,r plant cooling is lower than the values for the

hypothetical sites. This does not result in any limitation on locating plants at the

hypothetical sites.

3.4 CLIMATOLOGY/METEOROLOGY

3.4.1 TEMPERATURE

The maximum dry bulb plant parameters envelope values associated with the

evolutionary, passive, and generic plants bound therange of maximumtemperature

values listed for Sites A through D in Table 2-1. The site characteristic values are

2,,_,_,_onP,j_s_ 3-6 Rev. 0September 1992

absolute extremes, not the historical limit excluding peaks less than 2 hours as

described in the plant parameter URD references.

Similarly, the minimumdry bulb plant parameters envelope values bound the range

of minimum temperature values listed for the sites, except for the high latitude

Site D. The site characteristic values are absolute extremes, not the historical limit

excluding minimum values for periods of less than 2 hours as described in the

plant parameter URD references.

3.4.2 TEMPERATURE AND MOISTURE CONTENT

The minimum (dry bulb) temperature requirementof the plant parameters envelope

which is not exceeded 1 percent of the time bounds the range of minimum winter

design dry bulb temperatures listed for Site A. Based on the cited reference for the

hypothetical site data, minimumwinter designdry bulb temperatures lower than the

plant parameter..':._nvelope values are found within the areas where Sites B, C, and

D might be located.

The maximum coincident dry and wet bulb temperatures associated with the

evolutionary, passive, and generic plants bound the corresponding range of values

for Sites C and D. However, based on the cited reference for the hypothetical site

data, maximum coincident dry bulb temperatures higherthan the plant parameters

envelope are found within the areas where Sites A and B might be located, while

coincident wet bulb temperatures are lower for all of the hypothetical sites.

The summer design 1 percent exceedance (noncoincident)wet bulb temperatures

associated with the evolutionary, passive, and generic plants bound the corre-

sponding range of values for Site D. However,based on the cited reference for the

hypothetical site data, maximumsummer designwet bulb temperatures higher than

218_o_R_._..o,,P,jEs_ 3-7 Rev. 0September 1992

the plant parametersenvelopeare found withintheareas where SitesA, B, and C

mightbe located.

No comparisonwith the plant parameter maximum wet bulb temperaturecan be

made because relatedhistoricalinformationis unavailable.

3.4.3 PROBABLE MAXIMUM PRECIPITATION

The maximum 5-minute and 1-hour rainfall rate values associated with the

evolutionary, passive, and generic plants bound the corresponding precipitation

values for the hypothetical sites.

3.4.4 GROUND SNOW LOAD

Overall, the maximum snow load for roof design for th,'_evolutionary,passive, and

generic plants may not boundvalues for sites in some northern areas. Roofdesign

snow loads greater than 50 psf (plant parameters envelope) occur in the extreme

northeastern portions of the United States, the northern Great Lakes region, the

northern portion of the Appalachian Range, and in some sections of mountain

ranges in the western United States. The roof designs with scuppers and low or

no p_rapets can limit the amount of snow that can build up on the roof and the

snow loading. The parameter is site specific and needs to be evaluated for the

plant location and meteorological conditions.

The URD values of 50 psf for roof snow load could be reviewed to assess whether

the current roof designs that include seismic and tornado requirements result in

capabilities for the roofs to accommodate greater snow loading.

218gO\ReportkDemonPrj,ESP_ 3-8 Rev. 0September 1992

3.4.5 BASIC WIND SPEED

The term "basic wind speed" is associated with a 50-year return period.

Importancefactorsare usedtoadjustthebasicwindspeedbasedonthestructural

category. Extreme plant parameters envelope wind speed values bound the

correspondingrangeof basic windspeedvalues listedfor the hypotheticalsites.

3.4.6 DESIGN BASIS TORNADO

Because the hypotheticalsite's designbasis tornado(DBT) parameters values

were based on Reg. Guide 1.76 (Reference8), the DBT parametersassociated

withthe evolutionary,passive,andgenericplantsdo notboundthe corresponding

values for any of the hypotheticalsites. This comparisonwas based on the

minimumDBT values, not maximum,of the three planttypes. In thiscase, the

minimumvalueswere used becausethey representthe limitingsituation,i.e., the

genericplant'sabilityto withstandtheseconditions.An assessmentof the status

of the regulatorycriteriafor the DBT is presentedin Reference12.

3.4.7 ANNUAL AVERAGE AND ACCIDENT ATMOSPHERIC DISPERSION

COEFFICIENTS- x/Q VALUES

Given the conservativemethodologyused to develop these x/Q values and their

site-specific origins, the hypothetical site annual average and accident x/Q values

should not be considered thresholds in comparison to the corresponding plant

parameters envelope. Rather, these typical x/Q values should be considered

conservative rough estimates, useful only to broadly measure agreementwith the

plant parameters envelope dispersion coefficients.

Such a broad-based comparison of the hypotheticalsite and plant parameter

generic and individual plant parameter accident x/Q values shows that, while some

2_8_,,_,,,D.,,,o,,P,j,_s_ 3-9 Rev. 0September 1092

of the accident x/Q values exceed the corresponding plant parameters, these

exceedancesare not large enoughto be clearly indicativeof an ALWR design

issue. The annualaverage(routine)xlQ valuesassociatedwiththe evolutionary,

passive, and generic plants boundthe correspondingvalues compiled for the

hypotheticalsites.

218QO_R_r,_,,,o,P,j._S_ 3--10 Rev. 0September 1992

4.0 CONCLUSIONS

Basedon the evaluationspresentedin Section3.0, the plant parametersenvelope

values for the ALWR designsdo not raise insurmountablesiting issues at the

hypotheticalsites.

The design basis tornadoparameters are an open issueand are presentlybeing

assessed as discussedIn Reference12, Section2.1.2.

,,a_o_,_..onF',,_8_ 4-1 Rev. 0September 1992

5.0 REFERENCES (Cited In Volume II)

1. NUREG-1437, 1991. "Generic EnvironmentalImpactStatementfor License

Renewalof Nuclear Plants"- 2 volumes.Main Repc,rt Vol. 1, Appendices

Vol. 2, Draft Report for Comment,USNRC - Office of Nuclear Regulatory

Research,August,

2. ORNL-NSIC-55, "DesignData and Safety Featuresof CommercialNuclear

Power Plants"- 5 volumes. NuclearSafety InformationCenter, Oak Ridge

NationalLaboratory, Vol. I Dec. '73, Vol. II Jan '72, Vol, IIi Apr. '74, Vol, IV

Mar '75, Vol. V Jun '76.

3. Giusti, E. V., Meyer, E. L., 1977. Water Consumption by Nuclear Power

Plants and Some Hydro/ogica/ /mp/ications,GeologicalSurveyCircular745.

4. U.S, Departmentof Commerce (DOC), 1968. C/imatic At/as of the United

States, EnvironmentalScienceServicesand Administration,Environmental

Data Service,Washington,DC, June.

5. U.S. Departmentof Defense (DOD), 1978. "Facility Design & Planning,

WeatherData", Departmentof Air Force Manual (AFM 88-29), Department

of ArmyTech. Manual (TM6-785), Dept. of Navy Manual (NAVFAC P-89),

Washington,DC, July.

6. a. NOAA Hydrometeoro/ogy Report No. 51, 1978. "Probable Maximum

Precipitation(PMP) Estimates"- United States East of the 105th

Meridian,U.S. Dept.of theArmy,Corpsof Engineers,Washington,DC,

June.

_1,=,,_,_,_o,P,jEs_ 5-1 Rev. 0September 1992

b. NOAA Hydrometeorology Report No. 52, 1982. "Application of

ProbableMaximumPrecipitationEstimates"- UnitedStatesEastofthe

105th Meridian, U.S. Departmentof the Army, Corps of Engineers,

Washington,DC, August.

7. American National Standards Institute (ANSI), 1982. American National

Standard Minimum DesignLoads for Buildings and Other Structures, ANSI

A58.1-1982 (Revisionof A58.1-1972), New York, NY, March.

8. U,S. NuclearRegulatoryCommission,i974. Regulatory Guide 1.76,Design

Basis Tornado for Nuclear Power Plants, Washington,DC, April.

9, Markee, E. H., Jr., et al., 1974. Technical Basis for interim Regional

Tornado Criteria, WASH-1300, U.S. AtomicEnergyCommission,May.

10, Ehlert, G, W, to Fox, ,I. and J, Baechler,MemorandumSubject:Telephone

ConversationwithJ. Lee of NRC, November25, i991.

11. U.S. Nuclear Regulatory Commission, 1978. Regulatory Guide 1.70,

Standard Format and Content of Safety Analysis Reports for Nuclear Power

Plants, LWR Edition,Revision3, Washington,DC, November.

12. Utility Requirement Documents, Volumes II/111,Chapter 1, Appendix B,

Section2.1.2.

_t.o_,,=,_,.o,P,jEsP, 5-2 Rev. 0September 1992

Appendix F

SERCH SUMMARIES REGARDING SITING

This appendixinclude=a compilationof SERCH materialthat is applicableto the

EarlySite PermitDemonstrationProgram.

The material Included ts drawn from the SERCH index, which t= a cumulative

Ilsttngfrom 1982 of the followingtypesof documents:

• Meeting Summary (MS) - Overviewof NRC meeting=(Commission,

Staff,ACRS), as well as selectedindustrymeeting=

• Report Summary (RS) . Review of NRC reports (NUREGs and

NUREG/CRa), a= well as selected Industryreports (e,g,, NUMARC,

INPO),

• RegulatoryCriteria (RC). Briefdescriptionof proposed/finalchanges

to 10CFR, as wellas otherfederalregulations,RegulatoryGuides,and

other NRC GenericCommunications,

Also Included tn this appendixis a listingof documentsdrawn from REGIS,

REGIS, REGulatory InformationSystem, is a computerizedhistoricalcomptiation

of much of the matedal used by licensingpersonnel, It includesa ii=ttngof all

InformationNotice=,Circulars,Bulletin=,GenericLetters,RegulatoryGuide=,and

StandardReviewPlans,as wellasmostNUREGandNUREG/CR reports,SECY=,

AEOD reports, and other miscellaneousdocuments,

t,=_,__j=s_ F.1 Rev, 2March 1993

AllSERCH and REGIS materialis indexedby keywordwhichis filed tn numerical

orderand easily retrieved. The followingkeywordscorrespondto Issues relating

to the early site permttprocess:

• AdvancedReactors

• _i¢tng

• SourceTerms

• Standardization

A itstingof the SERCH and REGIS documentsfiledby thesekeywordsis provided

tnthisappendix. Alsoincludedts a selectionof SERCH summariesfrom1988 to

presentwhichdiscussIssuespertinentto the eadysitepermitprocess.Fora copy

of SERCH documentslistedInthe index,butnotIncludedinthe appendix,please

contactBechtelPowerCorporation,NuclearChief Engineer, (301) 417-3099

,,,,ov_,_,,,,,_==_ F-2 Rev. 2March 1993

LIST OF DOCUMENTS CONTAINED IN

APPENDIX F

Excerptsfrom SERCH Index

Regulatory InformationService(REGIS):

SERCH Document RC.88.45

SERCH Document RC-89-22

SERCH Document MS.89.36

SERCH Document MS-90.01




SERCH Document MS-9t.43





SERCH Document MS-92-26




,,,,w_=,c,.,,=_j,,_ F-3 Rev. 2March 1993

ACCZD_ AH_YS IS

K[YI_OP,J)/T ITIJ_ ]U_CUIATORY IERCHCRZTF_ZA It__cm

JuZy 199i

S_ary of 8 Workshop on Severe Accident NIJ_.I_G/_ll-5780 LS-91-0/.7Nanajenent for JgRs, November 1991

Sua_ary Of York|hop on Severe Accident _G/CR-57111 IU;.91-0S0Hans|chertS tot Fw_s, November 1991

Hath Stems Xeoletion Valve Concerns BH-V2-N/*(Z.H), December 196/,

Proposed Cancellation of Isolation Con- D-Y3-N3denser Vent8 (Z.K), Sept. 1985

AJ_V.AJqCEDR£ACTORSllJtlOlJogOllJlllOll6Qlpl

ACRS Subcon_ittee on Advanced Reactors, KS-85-011February 1985

ACRS Fu_l Convnittee Heetin 6 on Advanced 1qS.66-102Lisht Vater Reactor Deji|nl, Saps. 1986

Co_nission mriefin 6 by General Electric HS-86-103Company on T_eir Advanced BoiZ/ns WaterReactor, Sepia=bar 1986

Conu_iselon Briefin_ on Advanced Reactor HS-86-106Desisns, October 1986

ACRS Full Con_ittee Heetin 8 on Improved MS-86-129Lisht Water Reactors, December 1966

NI_C Co.isaiah Briefin s on Advanced HS-87-012Reactor=, February 1987

ACRS FuZ1 Committee Heecin I on Improved HS-67-025Li&ht Water Reactor DesiJns, April/1987

ACRS Subcommittee Heetin& on Advanced HS-88-001Reactor Desl&ns

ACRS Sull Con_ittee Heetin& on GE's HS-66-003AB_ Desl&n ScheduleJanuary 1988

NRC Staff Heetin& rich £PRI on Advanced HS-88-005L_JI_Utility Requirements Document,Chapter S: £ngineered SafeguardsSystemsJanuary 1988

Rev,0s September1992

ADVANCED REACTORS

KEYWORD/TITLE REGULATORY $ERC'_CRITERIA EEFERENCE

Commission Briefing on Certification of MS-88-008ABWR Design Certification Status,January 1988

ACRS Full Committee Meeting on Advanced MS-88-037Depar_enn of Energy Reactors, Apr. 1988

ACRS Subcommittee Meeting on Advanced MS-$8-064Boiling Water Reactors, June 1988

Commission Briefing by EPRI on the MS-88-073Advanced Light Water Reactor Program,June 1988

ACRS Subcommittee on Advanced Reactor MS-$8-078

Designs Briefing by the Staff on the

Review of the Modularized High

Temperature Gas-Cooled Reactor, June '88

ACRS Subcommittee Meeting on Advanced MS-88-109

Reactors, August 1988

ACRS Subcommittee on Improved Light M3-88-I13

Water Reactors Meeting to Discuss the

EPRI Requirements Document, August 1988

Commission Briefing on Key Licensing MS-88-I16Issues Associated with DOE Advanced

Reactor Design, August 1988

ACRS Full Commltree on the Modular High MS-88-I17

Temperature Gas-Cooled Reactor.

August 1988

Commission Briefing on the Treatment of MS-88-134Severe Accidents in Future LWRs,

September 1988

ACRS Subcommittee Meeting on PRISM MS-88-141

Design, October 1988

ACRS Full Committee Meeting on PFISM, MS-88-157November 1988

ACRS Subcommittee Meeting on Advanced MS-88-158BWRs, November 1988

NRC Workshop on Severe Accident Policy MS-88-167

for Lt/Rs of an Evolutionary Design,December 1988

Rev.06 September 1992

ADVANCED REACTORS

KEYWORD/TITLE RE_TORY SERCH; CRITERIA REFERENCE

ACRS Full Committee MeeClng to Discuss MS-88-170

the SAPR Design, December 1988

ACRS Subcommittee on Improved _ 10 CFR 52 MS-$9-004Meeting with the Staff to Discuss theStandardization Rule, January 1989

Commission Briefing on the Current MS.89-019Status of the CE A_WR, January 1989

e

ACRS Joint Subcommittee Meeting on 10 CFR 52 !_-$9-073Containment Systems and Structural

Engineering to Discuss Future Plant

Designs, April 1989

ACRS Full Committee Meeting on Advanced MS-89-105

BWR Severe Accident Design, June 1989

Commission Briefing on Severe Accidents MS-89-I08

Design Features for the ABWR, June 1989

NRC Staff Meeting with the EPRI ALWR MS-89-I15

Utility Steering Group on SevereAccident Issues, June 1989

Commission Briefing on Integration of MS-89-127

Policy Statements for Severe Accidents,

Advanced Reactors, Safety Goals, and

Standardization. July 1989

NRC Commission Briefing on EPRI Design MS 39-131

Requirements Document of Advanced

Light Water Reactors, August 1989

ACRS Full Committee Meeting on Advanced MS-89-136Boiling Water Reactors (ABWR),

August 1989

ACRS Full Committee Meeting on EPRI MS-89-155

Design Requirements Document for

Advanced Light Water Reactors Program,

September 1989

Commission Briefing on EPRI Design MS-89-161

Requirements Document for Advanced Light

Water Reactors, September 1989

ACRS Subcommittee on Advanced MS-89-162Pressurized Water Reactors Presentation

on SP/90 Severe Accident Issues,September 1989


7

/_V_C£DKE_CTOKS

KEYWOP,D/'r ITLE REGULATORY $ERCHCRITERIA REFERENCE )

ACRS Full Committee Meetln& to Discuss MS-89-166

CANDU-30 Ocuober 1989

ACRS Advanced Boiling Water Reactor MS-89-171

Subcommittee Meecin g on the GEAB_K,October 1989

Commission Briefin B on CE, CE, and MS-89-173Westinghouse Advanced Reactors,November 1989

ACRS Subcommittee Meetin_ on AP_rR MS-89-174Westinghouse SP/90, November 1989

ACRS Full Committee Meeting on Advanced MS-89-181Reactors, November 1989

ACRS Full Committee Meeting on GE AB_/R MS-89-182

Design, November 1989

Commission Briefing on DOE Views on HS-89-186

Advanced Light Water Reactor Designs andCertification, November 1989

ACRS Joint Subcommittee on Containment- MS-89-198

Structural Engineering Meeting to

Discuss Containment Design Criteria forFuture Plants, December 1989

ACRS Full Committee Meeting on MS-90-014

SECY.90-16, Evolutionary Light WaterReactor Certification Issues,

February 19900

ACRS Joint Subcommittees on Extreme MS-90-031External Phenomena and Severe Accidents

Meeting on the External Events (IPEEE).March 1990

ACRS Full Committee Meeting to Discuss MS-90-037

Evolution AL_Rs, April 1990

Commission Briefing on the AL_;R Review MS-90-047

Program, April and May 1990

ACRS Subcommittee on _mproved L'JRs MS-90-052

Meeting on the Progress and Status of

Development of Passive L_;Rs - May 1990

Commission Briefing by the Advanced L_R MS-90-059

Utillty/EPRI Steering Group on

Resolution of Evolutionary Plant Key _ev. 0

8 September 1992

ADVANCED REACTORS

KEYWORD/TITLE REGUlaTORY $ERCHCRITERIA REFERENCE

Regulatory Issues, June, 1990

ACRS Full Committee Meeting Co Discuss MS-90-06201-84, June 1990

ACRS Full Committee Meeting Co Discuss MS-90-062Certification of ELWRs, June 1990

ACRS Subcommlccee on Improved Light MS-90-077Nater Reactors Meeting on the AdvancedLight Water Reactor Program, July 1990

ACRS Full Committee Meeting ¢o Discuss MS-90-078Advanced Light Water Reactor Program,July 1990

ACRS Full Committee Meeting co Discuss MS-90-078Requirements for an Essentially CompleteDesign, July 1990

Commission Briefing by NUMARCon 10 CPR 52 MS-90-082Essentially Complete Design for Part 52Submittals, July 1990

Commission Briefing by the NRC Staff on 10 CFR 52 MS-90-083Essentially Complete Design for Paru 52Submittals, July 1990

ACRS Full Committee Meeting: Proposed MS-90-090Schedule for Review of Evolutionary andAdvanced Nuclear Power Plant Designs,July 1991

ACRS Improved L_TRsSubcommittee Meeting, MS-90-092August 1990

ACRS Full Committee Meeclng on MS-90-095Essentially Complete Design forStandardized Plants, August, 1990

ACRS Full Committee Meeting on the MS-90-096Advanced Light Water Reactor Program,August 1990

Proposed Source Term Methodology for MS-90-101Passive Plant Designs, August 1990

NUHARC and NRC Meeting to Discuss Level I0 CFR 52 HS-90-I03of Detail in CFR 52 Design Certificationand Combined Operating LicenseApplications, August 1990

Rev. 0September1992

9

ADVANCED REACTORS

KEYWORD/T ITIE REGULATORY $ERCH

i CRITERIA REFERENCE I

NRC Staff Meeting with EPRI to Discuss MS-90-I04External Events PRA and Seismic Issues

For Advanced LWRs, August 1990

ACRS Subcommittee Meeting on Advanced 10 CFR 52 MS-90-113Pressurized Water Reactors, Sept. 1990

ACRS Subcommittee on Advanced HS-90-114

Pressurized Reactors Meeting with ABB/CETo Discuss the System 80+ EvolutionaryReactor, September 1990

ACRS Full Committee Meeting to Discuss MS-90-121Advanced Reactors, October 1990


the Westinghouse Standardized PlantSP/90, October 1990

ACRS Subcommittee Meeting on Improved I0 CFR 52 MS-90-130

Light Water Reactors, October 1990

ACRS Subcomlttee Meeting on Advanced MS-90-131

Boiling Water Reactors, October 1990

ACRS Subcommittee on Advanced P_/Rs re MS-90-135

Ssytem 80+ and Licensing Review BasisDocument, November 1990

ACRS Full Committee Meeting to Discuss MS-90-143Advanced Reactors, November 1990

Joint ACRS Subcommittee on Containment MS-90-148

Systems and Structures Engineering

Meeting on Contalnment Design Criteriafor Future Plants, December 1990

ACRS Subconu_tttee Briefing on MS-90-149SECY-90-377, December 1990

Commission Briefing on Standardization 10 CFR 52 MS-90-151and Part 52 Licensing, December 1990

ACRS Full Committee Meeting to Discuss MS-90-1_5Certification of S_andardlzed Plant

Design, December 1990

ACRS Full Committee Meeting on the MS-91-OI

General Electric SB_R, July 1991


EPRI's Requirements Document for AL_,s. Rev.0lO September 1992

ADVANCEDREACTORS

K.._"_1gORD/TITIE REGULATORY BERTHi CRITERIA REFERENCE

February 1991

ACRS Full Committee Meeting to Discuss MS-91-020Contair_ent Design Criteria for FutureL_s, February 1991

ACRS I=proved LiBht Water Reactors MS-91-051Subcommittee Meeting to Review theDraft SERs on Chapters 6-11 of EPRI'sAL_;RRequirements Document, April 1991

ACRS Full Committee Meeting to Discuss MS-91-053Certification Issues for LWRs,April 1991

ACRS Full Committee Meeting with the MS-91-065Director of the Office of Nuclear

Reactor Regulation, May 1991

ACRS Improved Light Water Reactor Sub- MS-91-074committee Meeting to Discuss the DraftSERs for Chapters 7, 11-13 of the EPRI

O AL_ Requirements Document, May lO91ACRS Subcommittee on Advanced Bollin_ MS-91-075Reactors, May 1991

ACRS Joint Subcommittees on Plant MS-91-076

Operations and PRA Meeting on ShutdovnRisk, June 1991

ACRS Full Com_ittee/Con_ission Briefing MS-91-078re Westinghouse AP-600 NPP; PolicyIssues Associated with the EPRI Require-ments Document, June 1991

Commission Briefing on the Progress of MS-91-080Design Certification Review andImplementation, June 1991

Nuclear Safety Research Review Committee MS-91-082Meeting to Discuss an Overview of SixAdvanced Reactor Types, June 1991

ACRS Full Committee Meeting: AB_TRs, MS-91-090July 1991

NRC Staff Meeting on AL_ Program on MS-91-093Source Term Changes, July 1991

NRC Staff Meeting wich EPRI to Discuss MS-91-I02Issues Associated with Staff Review of Rev.0

11 September1992

ADVANCEDIUL4CTO&S

KEY_ORD/TXTL£ _6"q_TORY |riCHCRZTE/IXA I_PgRLJ_CE

the Evolutionary and Passive Require-ments Docm_ents, July 1991

ACRS Subcommittee on AC/DC Power $yatems HS-91-I05.Reliability to Discuss the N+2 Concept,July 1991

AC&S Advanced Reactor Delitnl Sub- MS-91-106committee Neetin B on the PRXSN and MHTGRDesil_nS, August 1991

ACRS Full Co._,ittee Meeting: Systems MS-91-107Reliability - N+2, AueuJt 1991

Commission Briefins: Pro&ram for 10 CFR 52 MS-91-110Inspection, Testa, Analysis, andAcceptance Criteria (ZTAAC) for AdvancedReactors, August 1991

ACRS Advanced Pressurized Water Reactors MS-91-116Subcommittee: ABB/CE System 80+,September 1991

ACRS Advanced Boilln& Water Reactors HS-91-123Subcommittee: Review the Draft SERS forChapters 1-6 of the OE AB_R StandardSafety Analysis Report, September 1991

Commission Briefing: Implementation of 10 CPR 52 MS-91-13610 CPR Part 52 and Possible OptionsAvailable, October 1991

l_C Staff: General Electric and NIJ]_RC 10 CFR 52 MS-91-137to Discuss GE Pilot ITAAC Submittal,October 1991

ACRS Subcon_ittee on Severe Accidents: MS-91-141Discuss the Severe Accident ResearchProgram, October 1991

ACRS Advanced Boiling Water Reactor MS-91-143Subcommittee: Review the Draft SERs forChapters 3, 9, 10, ll, and 13 of the GEABWRStandard Safety Analysis Report,October 1991

Commission Briefing: NRC Staff on the MS-91-144Status of Advanced Light Water ReactorReviews, October 1991

NRC Conference on Advanced Light Water MS-91-147

Reactors, November 1991 Rev.0

12 September1992

ADV_C£D I_mCTORS

; KEY_OP_/TITL£ I_G_jI.qTOR¥ |ERCI4CRXT_IA REFERENCE

ACRS Full Committee: GE _IR Reactor MS-91-151

Water Cleanup System, November 1991

ACRS Full Committee: Standard De|llrn NS-91-151Certification and Level of Design Decal1November 1991

ACRS Full Committee: Vendor Test Prosram 1_-91-151to Support Desi|n Certification ofPassive LI;RaNovember 1991

ACRS eatery Philosophy Technolo|y and MS-91-158Criteria Subcommittee: Evolutionary LWP.sand Safety Goai Policy Implementation,December 1991

Joint Heetlns of Computers in Nuclear HS-91-159Power Plant Operations and Advanced I_Subcommittee: Dlsital ComputerExperiences in _estlnshouse and ABB-CEPlants, December 1991

Improved Lilhc Water Reactors Sub- MS-91-161committee: Discuss £PR! AL_R Require-mencs Document and Deslgn AcceptanceCrlcerla for AL_'R,December 1991

ACRS Full Committee: Personnel and MS-91-163Control Rooms for Advanced Reactors,December 1991

Commission Briefinl rich the ACRS, HS-91-164December 1991

ACRS Full Committee: Wescinshouse AP600 MS-92-005Test Programs and Design Acceptance,January 1992

ACRS Improved Lisht _acer Reactor MS-92-011Subcommittee: Draft SER for Chapter 10of the £PRI AL_R Requirements Document,January 1992

Senate Hearing: S.1220 Licensin_ lO CFR 52 MS-92-013Provisions, January 1992

NRC Staff: CE and EPRI Discussion of HS-92-015Testing Issues for Sgl0R, January 1992

ACRS Full Committee: Incesral Systems Rev. 0 Hs.92.022

13 September1992

A/)VANC_DFJLqCTORSf

K£WORD/TITL£ ItZCUI_TOP.¥ |FJLCd| CRIT£RIA R£F£R[NC[ t

TesCin| for the AP600 Nuclear Plant,February 1992

ACRSFull Couiccee: Dasl|n Acceptance i0 CFR 52 NS-92-023Criteria, February 1992

ACRS Advanced moilin| Water _eactors ILq-92-033Subcommittee Heating co Raviev Two DraEt|Lqa of the GE AJ_t_l_Standard SafetyA_alysis hport, February 1992

Comisston _rtefin I rich the ACRS Full 10 CI_R50 KS.92-O40Comities re Desi|n Acceptance Criteria, 10 C1_:200Proposed Revisions to 10 CFR 50 & 100,and Status of Advanced Reactors,Harch 1992

ACRSFull Committee Heetint on lnteirai HS.92-O_1Systems Testin| for Westinghouse AP600Safety Goal Zmplsmsncation_,Harch 1992

ACRSJoint Subco_ittesa on Computers in 1U.9_-043Nuclear Povsr Plant OperationsInstrumentation and Control Systems andHu;nan Factors, Hsrch 1992

Com_ission _riefin& on Requirements for MS-92-0_4Integr.! System T.stings of the AP600,March 1992

ACRSThermel.Hydraultc Phenomena MS.92-0_6Subcon_mitcee on oh, Nead for InceBralS),stem Testin_ of the Westinghouse AP600Nuclear Plant, Harch 1992

h"RCStaff Heecin_ rich EPRI to Discuss HS-92-053the Development of Reliability hsedTechnical Specifications for PlantDesigns that Employ Passive SafetySystems, Herch 1992

Comtnission grleflns: ABB/C£ Nuclear HS-92-065Pover'e St.Cue of System 80+ Applicationfor Design Certification, April 1992

ACRSFull Committee: Discuss CE's A_WR HS-92-066Design Certification, April !992

Commission Eriefin&: Status of Advanced HS-92-072Reactor Reviews, April 1992

Rev,0September1992

AL_oeoo_o.o.omeo oo.. o.*....o

ACRS Full Committee Heetin8 on Proposed 10 Cl_ 20 HS-83-113Revisions to 10 Cl_ Part 20, November1983

ACRS Subconu_ittee on Reactor Radio- H8-83-122lo&ical Effectl/_as_e Hanaaement/AirSystems, December 1983

ACRS Subconv_ittee on Reactor Red. HS-84-010Effects, January 198t,

An Approach to Assure AUkl_ Compliance, NUR[G/CR-2595 BS-83-092June 1983

Optimization of the Control of N1JREG/CR-$038 RS-88-0_1Concaimination at Nuclear Pover Plants,

Hay 1988

ALTEP_NATESHUTDO_ CAPABXLITY*************************

NRC Staff Heetin& with Oconee & HcCuire 145-83-057on Standby Shutdown Facility, June 1983 RBV.0

15 September1992

Potential H_an Factors Deficiencies in INILEG/CR.3696 U-8_-055the Desi|n of Local Control Stations andOperator Interfaces in Nuclear FoyerPlants, May 1984

Human Factor| Reviev of Remote Shutdovn |_-Vl-N4Capability (l,C), December 1983

Deletion of Source Ran|e Hone:or i |N-V2-N3Requirement for Alternate Shucdo_m (l.O)September 198_

AT_JSIm o o a o o am b I O • o o 0 m I o e elm o qJ *m o o

Deviations From Vestin|houae AMSAC 10 CFR S0. 62 LA.88-002Desi|n By Duke Foyer For HcGuire andCatawba

Vendor Interface Inspec:ion 1.4-88-037

Desil[nDeficiency and Procedures LAogl-020Deficiency Cause Spurious AT_TSSystemActuation Resultln& in Manual Trip

_C Staff HeetinB on ATlaS, Hay 1983 10 CPa 50 L_o83-OOa

ACR$ Subcommicce Meecin B on th* AT_S 10 Cl_ 50 I_-83-00SRulemakinB, May 1983

Staff Action in Response to Salem i ATI_S US.83-006Events, July 1983

Approval of the AT_S Rule, June 1984 I0 Cl_ 50. 62 I_-8a-008

ACRS Subcommiccee on Anticipated HS-82-027Transients _/ithout Scram (A_S),October 1982

ACRS Full Conu_iccee Meetin& on AT'J5 H5-82-029(ACRS Opinion), November 1982

ACRS Full Committee Meeting on Salem HS-83-034AT'SS, April 1983

Commission _riefin& on Salem Restart, HS-83-038April 1983

Commission Briefin& on Evaluation of _REC-!O00 MS-83-045

Implications of t:he Salem Events NUREG'O97_eV, 0

16 September 1992

_OI_/TITL[ UG_J_.ATOR¥ ll;RCXCAITnXA Ulql:I,AqCE

Light Co, (CP&L) to Discuss SingleFailure Criterion for ElectricalSystem, November 1990

Clarification of Application of Single lie.W-N1Failure Criterion l.F

SITINGa imeoes oooomQoeo QoeQmoamoo

ACRSSub©anal tie, on Reactor 10 Cle_ 20 MS-I2-0ZORadiOlOliCal Effects and SiteEvaluation, June 1982

Commission Briefing on the Statue of 10 CFR:I00 1U.90.001[floral to Develop an Updated SourceTern, January 1990

Commission Briefing on Source Term 10 C1_ 50 l(S.90-12aUpdate end Decouplin I Siting Froe Dlliln 10 ¢1r1:100October 1990 RG4,.7

b'P.CStaff Meetins rich RUI_.ACon 10 CFR:IO0 HS.91-031 9Potential Revisions to 10 Clq_Part 100.Appendix A, March 1991

I_C Staff Meecln| rich b_7_J_C to Discuss 10 C1_ 52 HS.91-03610 Cl_ Part 52 Issues, Hatch 1991

Commission Briefing on Status of Staff MS.91-0t,3Efforts to Update Slclng/Deslgn SourceTerm (TID-14844), _arch 1991

I_C Staff Heecin t rich KUKARCon the 10 Ctq_:100 HS.91-056Revision of IOCI_ Part 100, Appendix A,April 1991

b'RC Staff HeetinS rich Industry 10 CFR 52 MS-91-08SRepresentatives co Propose a CooperativeApproach Toward Early satin|, June 1991

ACRSExtreme External PhenomenaSub. 10 CFR:IO0 HS-91-166¢onvnittee: 10 CFP.Part 100. Appendix ARevisions, December 1991

ACRSJoint Subco,gnitceea on Safety 10 CFR:100 HS.92-004Philosophy. TechnoloBy end Criteria, I0 CFR $0Severe Accidents, Rel;ulatoryPoliciesand Practices: Part 100, Source Term,and Large Release Definition,

January 1992 Rov.0302 September1992

SITING

K_rYWOR.D/TIT_ RZG'ULATORY IIEP,CH, CRIT£RIA P,£Fr.,RENCE

ACRS Full Committee: Proposed Part 100 10 C1_:100 iqS.92.006Revision, Updated Source Term, and 10 CFR 50Xeric Release Definicl0n, Jan_ut_ 1992

ACRS lrull Committee: Proposed Revision 10 Cl_:Z00 KS-92-023of 10 CFP, Part 100, Appendix A, 'Seismicand Geologic Sitin| Cricerie tot NuclearPower Plants, Febru,r_ zgg2

^CRS Subcouittse on Extreme External 10 CFR:100 1qS.92.026Phenomena: 10 Clr_ 100, Appendix A,"Seismic and Geologic Sttln| Criteriafor Nucle,r Power Pla_ts," February 1992

Com_.ission Briefln$ vlth the ACRS lrull 10 Clq_ 50 NS.92-0t_OConu_,itcee re Design AccePtance Criteria, 10 CFR:IO0Proposed Revisions Co 10 CF_ 50 & 100,*nd Status of Advanced Reactors,Hatch 1992

l_C Stsff: h'U_C', Discussion of the 10 CI_:X00 145.92.078Proposed 10 CFI_ Part 100, Appsndlx AP,evision, April 1992

o l* e o IJ . e m • _, q* • I m l* @ o e G* e m _ 0 e •

Catawba Snubber l_eductlon Program I,A.88-02S

NEC Staff Heecinl_ rich Florida Power 10 CF_ 50:Ap.A MS-85.092Corporstion on GDC.t, Exemption for RCPSupports, August 19B5

ACES Subcon_mittee on Rsecto:' Operitions, 145.86-061June 1986

ACRS Full Committee: _r&e Bore Snubbers MS.91.13t4(GI-113), October 1991

ACR5 Subcommittee on Structural HS-91-135

Englneering: Generic Issue !13, "DynamicQualification Testing of Lmrge BoreHydraullc Snubbers," October 1991

Alternative Requirements for Snubber GL 90-09 RC-90-051Visual Inspection Intervals andCQrz'eccive Actions (Generic Letter 90-09Decernl_er 11, 1990)

Agtng and Service of Ideer Hydraulic and NIJREG/CR-I,3_tBV' 0S-86-025

3o3 September1992

S0'JRC[Tr._S_omm_ooolmm.ommommmoeomel

Co.lesion _riefin s on the Severe HS.83-006Accident Policy Statement, January 1983

ACRS subco_iccee Heecin& on Class 9 NUI_EG-0956 HS-83-037AccidenCe, April 1983

ACRSFul_ Committee Meecin$ on Severe MS-83-047Accident Research, May 1983

Co_isJion Briefin s on Severe Accident MS-83-071ResearchPlan, July 1983

ACRSFul! Committee Meetinl on the NUREG.0900 HS-83-090SevereAccident Research Program,September 1983

Seventh _ecer Reactor Safecy Research NUREG-CP.O047 HS-83-109_eeCinBonP_ Decay Heac Removal,Occober 1983

ACRSSubcommittee on Claisi9 Accidence HS-84-006January 1984

ACRS Full Committee HeetinB on Hey HS-84-O09Source Term Affects the Emer&encyPlanning Zone, January 198_

Congntcsion Briefin& on the ANS Source HS-84-173Term Report,November 198_

Commission Briefing on the American HS-85-021Rev, 0

3o_ September 1992

SOURCE TER_

KEY'_ORD/TI TIE I_E_TORY $ ERCH; CRIT£RIA REFERENCE

Physical Society Source Term Report,February 1985

Commission Briefing on Status of Source MS-85-032

Term Study, March 1985

Commission Brlefin E on Source Term and MS-85-O41IDCOR Severe Accident Eval., April 1985

Commission and ACRS Full Committee MS-85-044Meeting on Proposed Plant-SpecificBackfits Rule, Aprll 1985

Commission Briefing on Status of Source MS-85-084

Term Study, July 1985

ACRE Full Committee Meeting, Oct. 1985 MS-85-121

Commission Briefing on Quarterly Source MS-85-129Terms, November 1985

ACRS Subcommittee on Class 9 Accidents MS-86-030 "

February 1986

Commission Quarterly Source Term MS-86-O41

Briefing, March 1986

ACRE Subcommittee on Severe (Class 9) NUREG-O956 MS-86-060

Accidents and Nuclear Plant Chemistry

Meeting to Discuss the Final Draft ofNUREG-0956, June 1986

Commission Briefing on the Status of NUREG-O956 MS-86-076

NUREG-0956, July 1986

Commission Quarterly Source Term MS-86-131

Briefing, December 1986

NRC Staff Meeting on NUREG-II50, Risk NUREG-I150 MS-87-027

Analysis Methods, April 1987

NRC Workshop on B_/RMark Z Containment MS-88-018

Issues, February 1988

Commission Briefing on the Status of 10 CFR:I00 MS-90-001

Efforts to Develop an Updated Source

Term, January 1990

ACRE Full Committee Meeting to Discuss MS-90-053

Source Term Update, May 1990

ACNN Heetlng on 1-129 Source Term Rev. 0_S'90"073

305 September 1992

SOURCE TERMS

KnOR/rITLE REGULATORY SERCHCRITERL

Methodology for LL_ Sites and the BEIR VReport, June 1990

Proposed Source Term Methodology for MS-90-101Passive Plant Designs, August 1990

Commission Briefing on Source Term 10 CFR 50 MS.90-124Update and Decoupling Siting From Design 10 CFR:100October 1990 RG 4.7

Commission Briefing on Status of Staff MS-91-043Efforts to Update Siting/Design SourceTerm (TXD-148_4), March 1991

NRC Staff Meeting on ALWR Program on MS-91-093

Source Term Changes, July 1991

NRC Staff Meeting: Report the Status of MS-91-I12

the Source Term Update, August 1991

ACRS Full Co_ittee: Site Characteristic MS-91-I17

for Part I00 Rule, September 1991

NRC Staff: Status of Work on Development MS-91-125

of Updated Documentation on LWR FissionProduct Release into Containment,

September 1991

ACRS Joint Subcommlttees on Safety 10 CFR:IO0 MS-92-OO&

Philosophy, Technology and Criteria, I0 CFR 50

Severe Accidents, Regulatory Policiesand Practices: Part 100, Source Term,

and Large Release Definition,

January 1992

ACRS Full Co._Ittee: Proposed Part I00 I0 CFR:IO0 MS-92-006

Revision, Updated Source Term, and 10 CFR 50

Large Release Definition, January 1992

Commission Briefing: Proposed Updated 10 CFR 50 MS-92-073Source Term (TID-14844), April 1992 10 CFR:100

Iodine Behavior in Steam Cenerator Tube NUREG/CR-2683 RS-B2-OI5

Rupture Accidents, April 1982

Interim Source Term Assumptions for NUREG/CR.2629 RS-82-020

Emergency Planning and Equipment

Qualification, June 1982

Iodine Transport Predicted for a NUREG/CR-2659 RS-83-038Postulated Steam Line Break with

Concurrent Ruptures of Steam Generator Rev. 0

306 September1992

SPECIAL NUCLEARMATERIALS

KEY1gORD/TITLE REGULATORY SERCHCRITERIA REFERENCE b

4/24/92)

SPENT FUEL_emoomomoommo_mmm

Commission Briefing on Nuclear Waste MSm$3°025Policy Act of 1982, March 1983

NRC Staff Meeting with Duke Power MS-84-014Company on Spent Fuel Pool RerackingJanuary 1984

Co_Isslon Meeting on High Level Rad- MS-84mI76waste Disposal, November 1984

NRC Staff Meeting wlth PASh'Yon the MS-85-069Licensabllity of Storage of ConsolidatedFuel at Cinna, June 1985

ACRS Full Committee Meeting, Oct. 1985 MS-85-120

ACRS Full Con_ittee Meeting, Nov. 1985 MS-85-136|

ACRS Subcommittee on Spent Fuel Storage I0 CFR 72 MS-86o032Faclllcy Design. March, 1986

ACRS Subcommittee Meeting on Spent Fuel 10 CFR 72 MS-86-122Storage, November 1986

NRC Staff Meeting wlth Vermont Yankee MSm87-053Nuclear Power Corporation on SpentFuel Pool Expansion, July 1987

NRC Staff Meeting wlch FP&L on St. Lucie MS-87-069Spent Fuel Pool Rerack, September 1987

ACRS Full Committee Meeting on MS-88-014Nuclear Waste Policy ActFebruary 1988

NRC Staff Meeting with Northern States MS-88-017Power on Prairie Island's Fuel

Consolidation Demonstration Program,February 1988

NRC Staff Meeting wlch Southern Califor- MS-88-122nla Edison on Spent Fuel Transhipment,August 1988

Commission Briefing on Cask Designs for MS-88-150Shipping and Scoring Nuclear Materials

Rev. 0308 September1992 ,

SOURCE TERMS,

KEYWORD/'rlTLE RE_TORY SERCH; CRITERIA REFERENCE

Tubes, February 1983

Dose Projection Considerations for Emg NUREG/CR-3011 RS-83-053Conditions, May 1983

Radionuclide Release Under Specific L_TR RS-83-079Accident Conditions, BMI-2104, July 1983

Radionuclide Release Under Specific LLIR RS-83-079Accident Conditions (BMI Drafc),July '83

Fission Product Removal in Engineered NUREG/CR-3727 P,S-84-052Safety Feature Systems, April 1984

Noble Gas Iodine, end Cesium Transport NUREG/CR-3617 RS-84-094in a Postulated Loss of Decay HeatRemoval Act st Browns Ferry, August 1984

Reassessment of the Technical Basis for NUREG-0956 RS-85-039Estimating Source Terms (Draft forComment), July 1985

Radionuclide Release Calculations for NUREG/CR-&624 RS-86-061Selected Severe Accident Scenarios,July 1986

Reassessment of the Technical Bases for NUREG-0956 RS-86-063

Estimating Source Terms, July 1986

Fission Product Release Characteristics NUREG/CR-_881 RS-88-034Into Containment Under Design Basis andSevere Accident Conditions, March 1988

Source Term Estimation During Accident NUREG-1228 RS-89-002Response to Severe Nuclear Power PlantAccidents, October 1988

Radionuclide Release Calculations for NUREG/CR-4624 RS-90-062Selected Severe Accident Scenarios,Supplemental Calculations (Volume 6),August 1990

Iodine Chemical Forms in Light Water NUREG/CR-5732 RS-91-033Reactor (L_rRs) Severe Accidents, DraftReport, July 1991

SPECIAL NUCLEAR MATERIALS

Proposed Rule on Receipt of Byproduct iO CFR 50. 54 RC-92-024

and Special Nuclear Material (57FR15034, RBV. 0

307 September1992

SSER0

KEYVORD/TITLE REGULATORY SFACHCRITERIA REFERENCE

SSER 5 for Seabrook, July 1986 NI_G-0896 P_-86-073

SSER 3 for Nine Mile Poinu Unit 2, NUREG-1047 RS-86-073July 1986

Supplement 10 to the Safety Evaluation NUREG-0887 RS-86-079Report Related co the Operation of PerrySeptember 1986

SSER 1 Related to the Operation of South NUREG-0781 P.S-86-081Texas Project, Units 1 and 2,September 1986

SSER 4 Related to the Operation of Nine NUREG-1047 RS-86-081Mlle Point Nuclear Station, Unit 2;September 1986

SSER 7 Related to the Operation of NUREO-O853 RS-86-081Clinton Power Station, Unit l;

September 1986

STANDARDIZATIONooommmomoeomoomelaomoDooo

SECY-90-377. "Requirements for DesiKn IO CFR 52 LA-90-044Certification Under I0 CFR 52"

NRC Staff Meeting wlch EPRI on Standard- MS-84-024izaclon, February 1984

ACRS Subcommlctee on Class-9 Accidents, NUREG-I070 MS-84-087

May 1984

ACRS Full Co.u_icteeMeecln_ on AI_, MS-84-I04July 1984

ACRS Subconvnltceeon Safety Philosophy MS-84-129Technology, and Criteria, September 1984

ACRS Full Committee on Future Standard- MS-84-133

ized Designs and Safety Issues,September 1984

Co.u_tssion Briefin_ on Standardization MS-85-016February 1985

Co.u_lssion Briefing on Standardization MS-85-146December 1985

ACRS Subcommittee on S_andard Plant MS-86-001

Design, January 1986 Rev.0317 September 1992

STANDARDIZATION

K_ORD/TITLE REGULATORY |ERCHCRITERIA REFERENCE )

ACRS Subcommittee on Standard Plant MS-86-034

Design. March 1986

ACRS Subcommittee on Improved _ 1_-86-078Designs, July 1986

ACRS Full Co_ictee MeecinS on Proposed MS-86-079Revision Co the Standardization PolicyStatement, July 1986

ACRS Full Committee Meetinj on Standard- KS.86-089lzed Nuclear Plants, Aususc 1986

ACRS Full Committee Meecins on the MS-86-102Standardization Policy Statement,September 1986

Commission Briefing by General Electric MS-86-103Company on Their Advanced Boiling WaterReactor, September 1986

ACRS Subcommittee on Standardization of NUREG-1225 MS-86-107

Nuclear Facilities Meeting, October 1988 I

Commission Briefing on Advanced Reactor MS-86-I08Designs, October 1986

ACRS Full Committee MeetlnS rich the MS-86-129Commissioners on Nuclear Power PlantStandardization, December 1986

ACRS Subcommittee Meeting on Standard- MS-87-080ization of Nuclear Facilities, Oct. 1987

ACRS Full Committee Meeting on MS-87-091Westinghouse Advanced P_TR,November 1987

Commission Briefing by Combustion MS-87-O98Engineering on New Standardized Plants,November 1987

Commission Briefing on New Westinghouse MS-87-105JStandardized Plants, December 1987

ACRS Full Committee Meeting on EPRI's I_JREG-1197 MS-87-108Advanced Light Water Reactor Program,December 1987

Commission Briefing on CertifiCation of MS-88-008AB_,,'RDesign Certification Status,January 1988

Rev. 0318 September1992

STANDARDIZATION

KEYVORD/TIT_ REGU_TOR¥ |ERCHCRITERIA REFERENCE

ACRS Full Co_ittee Meeing on MS-88-013

ACRS Subco-._ittee Meetin| on the MS-88-061Proposed Standardization Rule, May 1988

ACRS Subcommittee Meetinl on Advanced MS-88.064goalie& rater Reactors, june 1988

ACRS Full Comities Meeting on KS-88-068Standardization, June 1988

Commission Briefing by EPR! on the MS-88.073Advanced Light Vater Reactor Program,June 1988

Co_ission Briefing on Standardization. 10 CFR 52 MS-88.079and Licensing Reform Proposed Rulemakin_June 1988

ACR$ Subcommittee on Improved LIJRs 10 CFR 52 MS-89-004Meetin& with the Staff to Discuss theStandardization Rule, January 1989

ACRS Full Committee on the Proposed 10 CFR 52 ITS-89-01310 CFR 52 Rule, January 1989

ACRS Full Committee Meetlng on the I0 CFR 52 MS-89-034Standardization and Licenstn& ReformRule, February 1989

Commission Briefin& on the Current I0 CFR 52 M5-89-036Status of the Final Rule on Standard-

ization and Licensin B Reform,February 1989

NU_L_RCand NRC Meetin& to Discuss Level 10 CFR 52 MS-90-103of Detail in CFR 52 Design Certificationand Combined Operatin B LicenseApplications, August 1990

NRC Commissioners' Technical Staff MS-90-123Meeting with KUHARCon Level of DetailFor Design Certification, October 1990

ACRS Full Committee Meetin& to Discuss MS-90-142level of Design Detail for StandarizedNuclear Power Plants, November 1990

Commission Briefing by NUMARCon the 10 CFR 52 MS-90-158Level of Design Detail per Part 52.December 1990

Rev.0319 September1992

STANDARDZZATION

Itr OP)/TI TLE l,J b'trtATOR¥ SZR, CRITDtZA PJEFERI3C[ Jt

NRC Staff NeetinS with NURARCto Discuss 10 CFR 52 1qS-91.03610 CFR Pert 52 Issues, March 1991

NRC Staff HeeCin| with NiJ_AC on 10 CFR 10 CFR 52 P_-91.108Pert 52 Related Issues, aU|Ult 1991

Commission Brlefinl: Prolram for 10 CFR S2 1qS-91.110lnepeccion, Tests, _alysLe, andacceptance Criteria (ITAAC) for advancedReactors, Aul_St 1991

Commission Brieflnl: ImpXementation of 10 CFR 52 ITS-91-13610 CFR Part 52 and Posmible Optionmavailable, October 1991

I_C Staff: General Electric and N1J_C 10 CFR 52 P,s-gx-137to Discuss G[ Pilot ITAA¢ Submittal,October 1991

Improved Li|hc Water Reactors Sub- MS-g1-161committee: D_scuss EP_I ALtrR Require-ments Document end Design AcceptanceCriteria for AL'*'P,,December 1991

ACRS Full Committee: Desi&n Acceptance MS-91-163Criteria, December 1991

Com_ission Brieftns with the ACRS, MS-91-164December 1991

ACRS Full Com=itcee: _e|tinBhouse aP600 MS-92-005Test Prosrams and DesiBn Acceptance,January 1992

Senate HearinB: S.1220 LicenslnB I0 CPR 52 MS-92-013Provisions, January 1992

ACRS Full Committee: DesiBn acceptance 10 CFR 52 MS-92-023Criteria, February 1992

Commission Brlefin_ with the ACES Full 10 CFR 50 HS-92-040Committee re Deslsn Acceptance Criteria, 10 CFR:100Proposed Revisions to 10 CFR 50 & 100,and Status of Advanced Reactors,March 1992

Proposed Policy for Regulation of 10 CFR 50 RC-85-017Advanced Nuclear Power Plants, Mar 1985

)Revised Policy Statement on Standard- lO CFR $0 RC-87-044ization (52FR34884, September 15, 1987)

Rev,0320 September1992

STANDAADXZATION

KZYWORD/TIT_ RI_CUI.qTORY |[RCH| CRITERIA ItEFERENCE

Proposed Rule on Early Site Permits, 10 CFR 52 RC-88-045Standard Desiin Certifications; andCombined Licenses for Nuclear PowerReactors (531_32060, 8/23/88)

Final Fule on Early Site Permits, 10 Cl_ 2 RC-89-022Standard Desil_ Certifications, and 10 CFR 50Combined Licenses for Nuclear Power 10 Clq_ 51Pl,nts (54FR15372, April 18, 1989) 10 CFR 52

10 Cir.: 170

STATION BLACKOUToaQooomomommoooooomoo_omo

Safety Evaluation Report on the Byron/, LA-90-035Brsidvood Statio_ Blackout Submittal

Sa/ecy Evaluation Report on the Crystal LA-90.038River 3 Station _lackouc Submittal

Safety [vsluscio_ Report (8[R) on the LA-90-O&OMaine Yankee Station Submittal

ACRS Reco_endscion on Station Blackout 10 CFR SO, 63 Lm-87-003

Approval of Station Blackout Rule, 10 CFR 50 I.B-88-001May 1988

ACRS Subco_ittes MeecinB on AC/DC Power HS-82-0_9Systems Reliability, September 1982

ACRS Subcommittee Meecin S on AC/DC Power MS-83-043Systems Reliability, May 1983

ACRS Subconv_icte, on AC/DC Reliability, MS-83-065July 1983 !

ACRS Subcommittee on Electrical Systems, MS-85-024February 1985

ACRS Full Committee Meeting, March 1985 MS-85-027

NRC Staff Meeting with BGb£, APbL on MS-85-031RCP Seal Failure, March 1985

NRC Sca/f MeeCin 8 with NUGSBOon MS-85.063Proposal co Resolve A-44, Hay 1985

Commission Briefin B on Station Blackout, MS-8S-104

September 1985 Rev.0321 September1992

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SITING SECY-83-SO& 831209 GENERALGUIDELINES FOR CONCURRENCEOF THE RECONNENDATIONOF SITES FORNUCLEARUASTE REPOSiTORIES

SITING SECY-84-O02 840104 DOEIS GENERALGUIDELINES FOR RECOIqqENDATIONOF SITES FOR NUCLEARHASTEDEPOSITORIES

SITING SECY-QO-341 901000 STAFF STUDY ON SOURCE TERM UI_ATE AND DECQUPLING SITING FRONDESIGNSITING SECY-91-041 910200 EARLY SITE PERMIT REVIEU READINESSSITING SRP- 2.1.1 (2) 810700 SITE LOCATION AND DESCRIPTIONSITING SRP- 2.1.2 (2) 810700 EXCLUSION AREA AUTHORITY AND CONTROLSITING SRP- 2.1.3 (2) 810700 POPULATION DISTRIBUTIONSNUBBERS GENLTR-8_'013 840503 TECH SPECS FOR SNUBBERSSNUBBERS GENLTR-90-O09 901211 ALTERNATIVE REQUIREMENTSFOR SNUBBERINSPECTION INTERVALS AliO CORRECTIVE

ACT! ONS

SNUBBERS 186-78-010 780627 BERGEN-PATERSONHYDRAULIC SHOCK SUPPRESSORACCUI4)LATONSPRING COILSSNUBBERS IE9-81-001 810127' SURVEILLANCE OF MECHANICALSNUBBERSSNUBBERS 1EC-76-05 761008 HYDRAULIC SHOCK AND SUAY SUPPRESSORSSNUBBERS IEC-78-07 780531 DAMAGEDCOMPONENTSOF A BERGEN-PATERSONSERIES 2S000 HYDRAULIC TEST STANDSNUBBERS IEC-79-25 791220 SHOCKARRESTORSTRUT ASSEIqBLT INTERFACESNUBBERS 1EC-79-2_3A 800131 SHOCKARRESTORSTRUT ASSEI_LY INTERFACESNUBBERS IEN-79-001 790202 BERGEN-PATERSOllHYDRAULIC SHOCKAND SWAYARRESTORS(HSSA)SNUBBERS |EN-OO-O_2 80112& EFFECT OF RADIATION ON HYDRAULIC SNUBBERFLUIDSNUBBERS |EN-82-012 820&21 SURVEILLANCE OF HYDRAULIC SNUBBERSSNUBBERS IEN-83-013 830321 DESIGN MISAPPLICATION OF BERGEN-PATTERSONSTANDAIIOSTRUT RESTRAINT CLAMPSNUBBERS 1EN-83-020 830k13 ITT GRINNELL FIGURE 306/307 MECHANICAL SNUBBERATTACHMENTINTERFACE

cjr) _[_ SNUBBERS IEN-83-047 830712 FAILURE OF HYDRAULIC SNUBBERSAS A RESULT OF CONTAMINATEDHYDRAULIC FLUID

-_ _<_ SNUBBERS IEN-8.3-080 831123 USE OF SPECIALIZED "STIFF" PiPE CLAHPS• SNUBBERS 1EN-04-067 _40817 RECENT SNUBBERINSERVICE TESTING VlTII NIGH FAILURE RATES(I)C:3 SNUBBERS 1EN-84-073 0_091& DOUNRATINGOF SELF-ALIGNING BAIL BUSHINGS USED IN ._IUBBERS

CT SNUBBERS IEN-BB-102 861215 REPEATEDMULTIPLE FAILURES OF STEAM GENERATORHYDRAULIC glIJOB4_S DUE TO(I) CONTROl.VALVE SENSITIVITY"-_ SNUBBERS UEG-04_7 7'80600 OPERATING EXPERIENCE WITH SIAIBBERS

(4:) SNUBBERS NUREG/CR-2032 8117300 SINGLE VS. DUAL SNUBBER INSTALLATIONSCC) SNUBBERS NUREG/CR- 217'5 8107'00 SNUBBERSENS! T! Vl TY STUDYr_) SNUBBERS NUREGICR-4006 850800 CLOSEOUTOF IE BULLETIN 81-01: SURVEILLANCEOF MECHANICALSII_llERS

liD. 27905106192

BECHTELPOUER CORPORATIONREGutmtory Information Service (REGIS)

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INDEX BY KEYMOR9

KEYMORD DOCUMENT DATE TI TLE

SNUBBERS NUREG/CR-&263 850500 RELIABILITY ANALYSIS OF STIFF VERSUS FLEXIBLE PiPiNG FINAL PROJECT REPORTSNUBBERS NUREG/CR-&279 860200 AGING ANO SERVICE UEAR OF HYDRAULIC ANO NECHANICAL SNUBBERSUSED ON

SAFETY-RELATED PIPING

SNUBBERS NUREG/CR-5_6 900100 BASIS FOR SNUBBERAGING RESEARCH: NUCLEARPLANT AGING RESEARCHPROGRAMSOURCE TERMS ANS-O5.1)4 820000 METHCO FOR CALCULATING THE FRACTIONAL RELEASEOF VOLATILE FISSION PRODUCTS

FROMOXIDE FUEL

SOURCE TERNS ANS-18.01/N237' 8&O000 RADIOACTIVE SOURCE TERN FOR IIORFlALOPERATION OF LIGHT MATER REACTORS(REVISION OF N237-1976)

SOURCE TERMS NUREG-0155 761209 FISSION PRODUCTGAMMARAY Ale PHOTONEUTRONSPECTRA AND ENERGY INTEGRATEDSOURCES INFORMAL REPORT

SOURCE TERMS NU1REG-0384 771200 PROCEDURESSOURCE TERM NEASUREMENTPR(X;RAMSOURCE TERMS NUREG-O41B 771200 FISSION GAS RELEASE FROM FUEL AT HIGN BURNUPSOURCE TERMS NUREG-07'71 810600 REGULATORYINPACT OF NUCLEARREACTORACCIDENT SOLII_3ETERN ASSUNPTIONSSOURCE TERMS NUREG-OTTJ 821100 THE DEVELOPMENTOF SEVERE REACTORACCIDENT SOURCETERNS: 19S7 - 1981SOURCE TERMS NUREG-0850 811100 PRELIMINARY ASSESSMENTOF COREMELT ACCIDENTS AT THE ZION AND lliOIAN POINT

NUCLEARPOUER PLANTS AND STRATEGIES FOR MITIGATING THEIR EFFECTS

SOURCE TERMS NUREG-0956 850524 REASSESSMENTOF THE TECHNICAL BASES FOR ESTIMATING SOLmCE TERNS, DRAFTSOURCE TERMS NUREG-OQ_6 860700 REASSESSMEHTOF THE TECHNICAL BASES FOR ESTIMATING SOURCETERMSSOURCE TERMS NUREG-1150 870200 REACTORRISK REFERENCEDOCUMENT, V(X.. 1SOURCE TERMS NUREG-1150 871)21)0 REACTORRISK REFERENCED(X3JNENT, VOL. 2 (APPENOICES A TBRU I)SOURCE TERMS NUREG-1150 870200 REACTORRISK REFERENCEDOCUMENT, VOL. ] (APPENDICES J TFlRUO)SOURCE TERMS NUREG-1189 860301) ASSESSMENTOF THE PUBLIC HEALTH IMPACT FROM THE ACCIDENTAL RELEASE OF UF6

AT THE SEOUOYAHFUELS CORPORATIONFACILITY AT GORE, OKLAIIONA, VOLS. I & 2

SOURCE TERMS NUREG-1228 881000 SOURCE TERM ESTIMATION DURING INCIDENT RESPONSETO SEVERE NUCLEARPOMERPLANT ACCIDENTS

SOURCE TERMS NUREG/CR-O0]6 780900 RADIOACTIVE GASEOUSEFFLUENT SOURCETERNS FOR POSTULATEDACCIDENTCONOITiONS OF LIGHT MATER COOLEDNUCLEARPOldERPLANTS

SOURCE TERMS NUREG/CR-0160 780800 EVALUATION OF FISSION-PRODUCT AFTERHEATSOURCE TERMS NUREGICR-9162 780800 BELAYEO BETA AMO GANqA RAY PRODUCTIONOUE TO TNERNAL NEUTRONFISSION OF

235U SPECIAL DISTRIBUTIONS FOR TINES AFTER FISSION BETMEEli2 AND 14,000SECONDS - TABULARAND GRAPHICAL DATA

SOURCE TERMS IiU_EG/CH-017'1 780700 FISSION-PROOUCT ENERGYRELEASE FOR TINES FOLLOMING THERNAL-REUTRONFISSIONOF 241 PU BETUEEN 2 AND l&,O00 SECONOS

SOURCE TERMS NUREG/CR-018.3 780500 DEFINITION OF LOSS-OF-COOLANTACCIDENT RADIATION SOURCE: StJlelARYANOCONCLUSIONS

C/') _CJ SOURCE TERMS NUREG/CR-0682 790S00 FISSION PROOUCTBEHAVIOR IN LURS

_ SOURCE TERMS NUREG/CR-0697' 790000 FISSION PRODUCTTRANSPORTANALYSIS (QUARTERLY PROGRESSREPORT)• SOURCE TERNS NUREGICR-0698 790000 FISSSiON PRODUCTTRANSPORTANALYSIS(2)3 C_ SOURCE TERMS UEG/CR-0715 790200 IN PLANT SOURCETERN NEASUREMEHTSAT ZION STATION[3" SOURCE TERMS NUREGICR-07'22 800200 FISSION PRODUCTRELEASE FROMHIGHLY IRRADIATED LHR FUEL(D SOURCE TERNS NUREG/CR-0917' 790800 URTERLY PROGRESSREPORTON FISSON PRODUCTBEHAVIOR IN LMAS FOR THE _RIOD-I JANUARY TO NARCH 197'9--.Ik

(D SOURCE TERNS NURE_,/CR-1061 791109 FISSION PRODUCTBEHAVIOR IN LVRSf.C) SOURCE TERMS NUREG/CR-1172 800100 DELAYED BETA- AM) GANqA-RAY PRODUCTIONDUE TO TRERNAL-HEUTIK)NFISSION OFI_0 239mJ

No. 28005106192

BECHTEL POKIERCORPORATIONREGu|ntory lnformetion Service (REGIS)

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INDEX BY _EYVOND

KEYMORD DOCUMENT DATE T I TLE

SOURCE TERMS lAIREG/CR-1213 800100 ANS 5.&. A CO_ER SUBROUTINE FOR PREDICTING FISSION GAS RELEASESOURCE TERMS NUREG/CR-1237 800600 BEST-ESTIMATE LOCA RADIATION SIGNATURESOURCE TERMS NUREG/CR-12B8 820300 FISSION PRODUCI SOURCETERMS FOR THE Ll_ LO_-OF-COOLANT ACCIDENTSOURCE TERMS NUREGICR-1386 801100 FISSION PROOUCTRELEASE FORMHIGHLY IRRADIATED LI_I FUEL HEATED TO 13000 TO

16000 IN STEAM ISOURCE TERMS NUREG/CR-1629 8OOQOO IN-PLANT SOURCE TERM MEASUREMENTSAT TU_Y POINT STATION - UNITS 3 AND 4SOURCE TERMS NUREG/CR-1820 820400 STATUS REPORT ON THE FISSION-PROI_T RESEARCHSOURCE TERMS NUREG/CR-182& 801200 ASSESSMENTOF FRAPCON-1 BE/EN CALCULATED FISSION GAS RELEASE IN RISO FUEL

RODSSOURCE TERMS NUREG/CR-185T 810500 THE STATE OF FISSION PRODUCTSIN DEBRIS BED EXI:_c'ltlNENTSSOURCE TERMS NUREG/CR-1992 810800 IN-PLANT SOURCETERM MEASUREMENTSAT FOUR PURLSSOURCE TERMS NUREG/CR-2348 811000 IN-PLANT SOURCETERM MEASUREMENTSAT RANCHOSECO STATIONSOURCE TERMS NUREG/CR-2SOT 820100 BACKGROUNDAND DERIVATION OF ANS-S.& STANDARDFISSION _ RELEASE MODELSOURCE TFPMS NUREG/CR-2629 820600 INTERIM SOURCE TERN ASSUMPTIONS FOR EMERGENCYPLANNING AND EQUIPMENT

QUALIFICATIONSOURCE TERMS NUREG/CR-2659 830200 IODINE TRANSPORTPREDICTED FOR A POSTULATEDSTEAM LINE BREAK I/ITII

CONCURRENTRUPTURESOF STEAM GENERATORTUBES

SOURCETERMS NUREG/CR-2683 820&00 IODINE BEHAVIOR IN STEAMGENERATONTUBE RUPTUREACCIDENTSSOURCE TERMS NUREG/CR-290T 860200 RADIOACTIVE MATERIALS RELEASED FROMNUCLEARPOldERPLANTS, VOL. 3SOURCE TERMS NUREGICR-3011 830500 DOSE PROJECTION CONSIDERATIONS FOR ENG CONDITIONSSOURCE TERMS NUREG/CR-361T 840800 NOBLE GAS IODINE, AND CESIUM TRANSI_T IN A POSTULATEDLOSS OF DECAY HEAT

REMOVALACC AT BROblMSFERRYSOURCE TERMS NUREG/CR-3Y27 8&1)400 FISSION PRODUCTREMOVALIN ENGINEERED SAFETY FEATURE SYSTEMSSOURCETERMS NUREGICR-3Y87' 840800 EFFECTIVENESS OF ENGINEERED SAFETY FEATURE(ESF) STSTENS IN RETAINING

FISSION PRODUCTSSOURCETERMS NUREG/CR-4037' 850700 DATA SU_RY REPORT FOR FISSION P_T RELEASE TEST NI-5SOURCE TERMS NUREG/CR-&081 850800 ABSORPTION OF GASEOUS IODIME BT UATER DROPLETSSOURCE TERMS UEGICR-&08S 8S07_)Q USERS MANUAL FOR CONTAIN 1.0.A COMPUTERCODE FOR SEVERE REACTORACCIDENT

CONTAINMENT ANALYSISSOURCE TERMS IP.JREGICR-&130 8S0900 ICEDF:A CODE FOR AEROSOLPARTICLE CAPTURE IN ICE CONPARINENTSSOURCE TERMS MUREGICR-4255 851200 AEROSOLRELEASE AND TRANSPORTPROGRANSEIqlANNIJALPROGRESSREPORTSSOURCE TERMS UEG/CR-&327' 851100 ORGANIC IODIDE FORMATION FOLLOUING NUCLEARREACTORACCIDENTSSOURCE TERMS NUREG/CR-439T 850900 IN-PLANT SOURCETERM I_,:ASURENEMTSAT PRAIRIE ISILANONUCLEARGENERATING

STATI ONSOURCE TERMS NUREG/CR-4AS$ 860100 LIGNT-UAT£R-REACTOR SAFETY FUEL SYSTElqSRESEARCMPlKIGRAqs, QUARTERLY

(,/') _]0 REPORT, VOL. 1(i) C_ SOURCE TERMS NUREGICR-4J_99 861000 CONTAIN CODECALCULATIONS OF THE EFFECTS ON THE SOURCE TERIqOF CSi TO I/2"C3 < commRsuoNOUeTOSEVERER_eOGENmmNGP,l- .

(i) SOURCE TERMS NUREGICII-&587' 860700 SOURCE TERN CODE PACICAG_:AUSER'S GUIDE (NOD 1)3 0 SOURCE TERMS NUREGICR-&62& 860700 RADIONUCLIDE RELEASECALCULATIONS FOR SELECTED SEVERE ACCIDENT SCEIIAR|OS,CT VOLS. 1 -S(D-_ SOURCE TERMS NUREGICR'&629 860700 INDEPENDENTVERIFICATION OF RADIONUCLIDE RELEASE CALCULATIOIIS FOR SELECTEO.--L ACCIDENT SCENARIOSCO SOURCE TERMS UEGICR'&656 860700 VERIFICATION TEST CALCULATIONS FOR THE SOURCETERN CODE PACKAGE

Page No. 28105106192

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INDEX BY KEYUORD

KEYbR)RD DOCUMENT DATE T! TLE

SOURCE TERMS NUREGICR-/d_88 860600 QUANTIFICATION ANO UNCERTAINTY ANALYSIS Of SOURCE TERMS FOR SEVEREACCIDENTS IN LIGHT DATER REACTORS(QUASAR)

SOURCE TERMS NUREGICR-4688 871000 QUANTIFICATION ANO UNCERTAINTY ANALYSIS Of SOURCE TERMS FOR SEVEREACCIDENTS IN LIGHT WATERREACTORS(QUASAR), VOL. 2

SOURCE TERMS NUREG/CR-469T 861000 CHEMISTRY ANO TRANSPORTOF IODINE IN CONTAINMENTSOURCE TERMS NUREG/CR-4722 871000 SOURCE TERM ESTIMATION USING MENU-TACTSOURCE TERMS NUREG/CR-4757' 910200 LINE-LOSS DETERMINATION FOR AIR SAMPLERSYSTEIqSSOURCE TERMS NUREG/CR*&77Q 8705Q0 HEM DATA FOR AEROSOLSGENERATEDBY RELEASES OF PRESSURIZED POUDERSANO

SOLUTIONS IN STATIC A

SOURCE TERMS NUREG/CR-A786 870900 TRANSPORTBEHAVIOR OF IODINE IN EFFLUENT RADIOACTIVITY MONITORING SYSTEMSSOURCE TERMS NUREG/CR-&881 880300 FISSION PRODUCTRELEASE CHARACTERISTICS INTO CONTAINMENTUNDERDESIGN BASIS

AND SEVERE ACCIDENT CONDITIONSSOURCE TERMS NUREGICR-&883 87'0400 REVIEU OF RESEARCHON UNCERTAINTIES IN ESTIMATES OF SOURCE TERNS FROM

SEVERE ACCIDENTS IN NUCLEARPOMERPLANTS

SOURCETERMS NUREG/CR-&Bt.9 890700 SOURCE TERM CALCULATIONS FOR ASSESSING DOSE TO EQUIPMENTSOURCE TERMS NUREG/CR-517_ 89071)0 BETA ANO GAMMADOSE CALCULATIONS FOR PVR ANO BtdlRCONTAINMENTSSOURCE TERMS NUREG/CR'S252 890500 AEROSOLSAMPLING AND TRANSPORTEFFICIENCY CALCULATION (ASTEC) AND

APPLICATION TO SURTSEY/1)CHAEROSOLSAMPLING SYSTEN

SOURCE TERMS NUREG,'CR-S2S3 9OOSO0 PARTITION: A PROGRAMFOR DEFINING THE SOURCETERNICONSEQIJENOEANALYSISINTERFACE IN THE MUREG-1150 PROBABILISTIC RISIC ASSESSMENT

SOURCE TERMS NUREGICR-S_9 900600 DETERMINATION OF THE NEUTRONANO GNqqA FLUX DISTRIBUTION IN THE PRESSUREVESSEL AND CAVITY OF A BOILING DATER REACTOR

SOURCE TERMS NUREGICR-S681 910500 LOM-LEVEL WASTESOURCE TERM NOOEL DEVELOPMENTAND TESTINGSOURCE TERMS NUREG/CR-ST32 910700 IODINE CHEMICAL FORMSIN LIJR SEVERE ACCIDENTS (DRAFT REPORT FOR COMMENT)SOURCE TERMS NUREG/CR-S?'47' 920100 ESTIMATE OF RADIONUCLIDE RELEASECHARACTERISTICS INTO CONTAINMENTUNDER

SEVERE ACCIDENT CONOITIONS [DRAFT REPORTFOR COMMENT]

SOURCE TERNS NUREG/CR-ST'BS 911000 SPARC-90: A.CODE FOR CALCULATING FISSION PRODUCTCAPTURE IN SUPPRESSIONPOOLS

SOURCE TERMS NUREG/CR-S773 911000 SELECTION OF MODELS TO CALCULATE THE LLW SOURCE TERNSOURCE TERNS SECY-83-O&4 830131 SUBMITS ALTERNATIVES FOR SCOPE OF GUIDELINES FOR POST-ACCIDENT RECOVERYIN

EVENT OF NUCLEAR-RELATEDACCIDENTS

SOURCE TERMS SECY-83-219 830606 STATUS REPT ON LUR ACCIDENT SOURCE TERM NEASSESSNENTSOURCE TERNS SECY-86-076 860228 IMPLEMENTATION PLAN FOR SEVERE ACCIDENT POLICY STATEMENTAND REGULATORYUSE

OF NEW SOURCE-TERMINFORMATION

SOURCE TERMS SECY-86-096 860324 STATUS REPORTON SEVERE ACCIDENT SOURCE TERM REASSESSMEIITC/) _D SOURCE TERMS SECT-86-369 861212 PLAN TO ADDRESS SOURCETERM TECHNICAL UNCERTAINTY AREAS(D (D

"0 <: SOURCE TERNS SECY-BQ-341 891100 UPOATED LIGHT MATER REACTOR(LUR) SOtJRC3ETERM METHODOLOGYANO POTENTIAL" REGULATORYAPPLICATIONS

::_ O 900500 USE OF PRA-BASEO SOURCETERM METHODSSOURCE TERMS SECT-90-1T5

(3" SOURCE TERMS SECY-90-307' 900830 IMPACTS OF SOURCE TERN TINING ON NRC REGULATORYPOSITIONS. NEGATIVE CONSENT(I) SOURCE TERMS SECY'90"3&I 901000 STAFF STUDY ON SOURCE TERM UPDATE ANO DECOUPLING SITING FROMDESIGN.-q

SOURCE TERMS SECY-90-&05 901200 FORHULATION OF A LARGE RELEASE DEFINITION AND SUPPORTING RATIONALE.--k

CO SOURCE TERMS SECY-92-127' 920400 REVISED ACCIDENT SOURCE TERMS FOR LIGHT-DATER IKJCLEARPOldERPLANTSCO SOURCE TERMS SERCHTOP-OSB 860700 SOURCE TERN RE-EVALUATIONro

No. 28205106192

BECHTEL POq_R CORPORATIONREGulatory Infornmtion Service (REGIS)

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INOEX 91' L'EYIa3RD

I(1EYWORD DOCUMENT DATE T! TLEsss _-__;z _--sss,zsssmssssssms| u: _-,__,e-,ss,__;,_zsfl _s| n_-st - iiii._..1._ _ _ii.,_ --,ti.i m,l_i.lii.,,,,ili..i.i_ _... ,, i_i _, _ - m--_ i_

SOURCE TERMS SERCRTOP-080 830990 LEAKAGE FROMESF COMPONENTSOUTSIOE CONTAINNENTSOURCE TERMS SRP-11.1 (2) 810700 SOURCE TERMSSOURCE TERMS SRP-12.2 (2) 810700 RADIATION SOURCESSPECi AL NUCLEARMATERIALS I EN-89- 02& 890306 NUCLEARCRI TICALi TY SAFETYSPECIAL NUCLEARMATERIALS NUREG-0290 Tr0600 A STUDY Of NUCLEARMATERIAL ACCOUNTINGFINAL REPORTJULY 1, 1976 TO APRIL

1, 1977 VOLUIqE1, 2, & 3SPECIAL NUCLEARMATERIALS MUREG-0350 770700 REPORTON STRATEGIC SPECIAL NUCLEARIqATERIAL IlIVENTORY DIFFENIENCESSPECIAL NUCLEARMATERIALS NUREG-04$O 780A00 REPORTOF THE MATERIAL CONTROLACCOUNTINGTASK FORCE: gJIqlqART, VIOL. 1,

VOL. 2, V(X.. 3, & VOL.4SPECIAL NUCLEARMATERIALS HUREGICR-O033 781_00 PROCEOURESFOR ROUNOINGMEA_NT RESULTS IN IKICLEARIqMERIALS CONTROL

AND ACCOUNTING

SPECIAL NUCLEAR MATERIALS MUREG/CR-OTT5 801100 TRAINING ANO QUALIFYING PERSONNELFOR PERFORNINGMEASUREMENTSFOR THECONTROLANt) ACCOUNTINGOF SPECIAL IBJCLEARMAIERIAL

SPECIAL NUCLEARMATERIALS NUREGICR-1260 800600 EVALUATION OF DOCKET FILES FOR TERNIIIATED SPECIAL IIUCLEAR MATERIAL LICENSESSPECIAL NUCLEAR IqATERIALS NUREGICR-1670 801000 VOLUIq£ 1: THE USE OF PROCESSIqONITORING DATA FOR IKICLEARMATERIAL,

ACCOUNTING sUIqlqAR1,REPORT

SPECIAL NUCLEARNATER|ALS NUREG/CR-18&7' 801100 STANDARDCONTAIHEkS FOR SNN STORAGE, TRANSFERAND IqEASURElqL_: FINALREPORTAPRIL 1978 TO AUGUST 1980

SPENT FUEL ANS-02.19 810000 GUIDELINES FOR ESTABLISHING SITE-RELATEO P/UlANETERS FOR SITE SELECTION MIDDESIGN OF All INDEPENDENTSPENT FILEL STORAGEINSTALLATION (MATER POOL TYPE)

SPENT FUEL ANS-OS.01/N16.1 830000 NUCLEAR CRITICALITY SAFETY IN OPERATIONS WITH FISSIONABLE NATERIALS OUTSIOEREACTORS

SPENT FUEL ANS-08.O7 87'0000 GUIDE FOR NUCLEARCRITICALITY SAFETY IN THE STORAGEOF FISSILE MATERIALSSPENT _FUEL ANS-08.17 840000 CRITICALITY SAFETY CRITERIA FOR THE _LING, $10RAGE R T_TATION OF

Lb_ FUEL OUTSIDE REACTORS

SP;-NT FUEL ANS-08.19 840000 JUJiIHISTRATI_E PRACTICES FOR NUCLEARCRITICALITY SAFETYSPENT FUEL ANS-57.01 800000 DESIGN REOUIRENENTS FOR LIGHT MATER REACTORFUEL IMli)LIN(; STSTENSSPENT FUEL ARS-ST.O2 830000 DESIGN REQUIREIqENTSFOR LIGHT MATER REACTORSPENT FUEL FACILITIES AT

NUCLEAR POkeR STATIONS

SPENT FUEL ANS-57',03 830000 DESIGN REQUIREMENTSFOR _ FUEL STOR/_ FACILITIES AT LIGHT MATER REACTORPLANTS

SPENT FUEL ANS-57'.07' 880000 OESIGN CRITERIA FOR AN INOEPENOENT SPENT FUEL STORAGE INSTALLATIONSPENT FUEL ANS-S7.09 84,0000 DESIGN CRITERIA FOR AN INOEPENOENT SPENT FUEL STORAGEINSTALLATION (DRY

STORAGETYPE)

SPENT FUEL ANS-57.10 871200 DESIGN CRITERIA FOR CONSOtlDATIOII Of LMA SPENT FUEL

Cf) ::][J SPENT FUEL GAO/RCED-88-079 880200 INFORMATION ON THE RERACI(ING OF THE DIABLO CANTONSPENT FUEL STORAGEPOOLS(I) (!) SPENT FUEL IEB-84-003 8_0824 REFUELING CAVITY MATER SEAL

"C) < SPENT FUEL IEN-SJ-029 830506 FUEL IlINOING CAUSED BY FUEL RACK DEFORNATIOII0 SPENT FUEL IEH-S&-093 841217' POTENTIAL FOR LOSS OF MATER FRON THE REFUELING CAVITY

3 SPENT FUEL |EN-87-013 871)224 POTENTIAL FOR HIGH RADIATION FIELDS FOLLOilING LOSS Of MATER FRoIq FUEL POOLCTCD SPENT FUEL IEN-8T-043 870908 GAPS IN NEUTRON-ABSORBINGMATERIAL Itl HIGH-DENSITY SPENT FUEL STORAGERACKS"_ SPENT FUEL IEN-88-065 880818 I_TEHT DRAINAGES Of SPENT FUEL POOLS._L SPENT FUEL IEH-88-092 881122 POTENTIAL FOR SPEHT FUEL POOL DRAINOOUNr_) 911129 POTENTIAL FOR SPENT FUEL PQOL ORAINOOMNU) SPENT FUEL IEN-88-092, $1r_

Page No. 28505/06192

BECHTELPOUER CORPORATIONREGulatory Information Service (REGIS)

INDEX BY [EYW(Xm

ICEYI/ORD DCK3JIqENT DATE T ! TLE

SPENT FUEL RG- 3.060 870300 DESIGN OF AN INDEPENDENT SPENT FUEL STORAGEINSTALLATION (DRY STORAGE)SPENT FUEL RG- 3.062 890100 STANDARDFORIqATANO CONTENT FOR TIlE SAFETY ANALYSIS IIIEP(_T FOR ONSITE

STORAGEOF SPENT FUEL STORAGE r.AS_SSPENT FUEL SECY-B3-T12 830323 P_mSED mEWIOCFRS3 ESTABLISHING PROCEDIJ_S & I_IIERIA ENRSLING IIRC TO

OETERMINE UHETHER ADEQUATESPENT FUEL STORAGECAPACITY CiimlOT RtEA_T BEPROVIDED AT INDIVIDUAL REACTORSITE

SPENT FUEL SECY-S&-&33 850116 FINAL RULE 10 CFR PART 53, " CRITERIA ARO _S FOR DETERRI11111GTIlEADEQUACYOF AVAILABLE SPENT iflJCLEARFUEL STORAGECAPACITYu

SPENT FUEL SECT-90-058 900221 PROGRESSm_E BY THE US DEPANTIgENTOF EHENGTANO THE INDUSTRT TO DEVELOPCASK DESIGNS TO ACHIEVE COMPATIBILITY FOIl DRY STORAGEAND TRANSPORTATIONPURPOSES

SPENT FUEL SECY-90-285 900813 ISSUANCEOF CERTIFICAI_ES Of_ colqPLIA11CEFOR OilY SPENT FUEL STQIIJUGECASk'SSPENT FUEL SECT-q1-043 9'10200 U.S. DEPARTNENTOF ENERGY (DOE) AND INDUS_Y IqlOGRESS !11 DEVELOPING CAS_

DESIGNS TO ACHIEVE CQMPATIIlILITT FOR _ STORAGEAND TRANgq3RTATIORPURPOSES

SPENT FUEL SECT-91-324 9110(0 PROPOS_ LICENSE, UND_ 10 CFR PART 72, FOR FRY STORAGEOF SPTNT FUEL ATPUBLIC SERVICE CONPANVOf COLOIU_O'S FONT ST. VILI_111IRN_EAN GEII_-AT111GSTATION SITE

SPENT FUEL SECY-92-07_ 920300 STATUS OF THE DEPARTNENTOF EI_RGY'S OFF-SITE FUELS PQLII_r, AS IT IIIELATESTO RECEIVING AND REPROCESSINGNIGN ENRICNED UItA111UNSPENT FUEL FRON FOREIGNRESEARCRREACTORS

SPENT FUEL SERCHTOP'OgA 870&00 NOq-CATEGORYI SPENT FUEL POOl. COOLINGSPENT FUEL SERCHTOP-09C 860600 HIGH DENSITY SPENT FUEL STORAGESPENT FUEL SERCHTOP-O_ 860500 LICENSING OF INOEPENOENT SPENT FUEL STORAGE I11STALLAT1011SSPENT FUEL SRP" 9.1.2 (3) 81071)0 SPENT FUEL STORAGESPENT FUEL SRP" 9.1.3 (1) 810700 SPENT FUEL POQL COOLING AND CLEANUP SYSTEIISTAUr_AROIZATION NUREG-O|02 7'60800 INTERFACES FOR STANDARDDESIGNSSTANDARDIZATION NUREG-O&27 780600 REVIEW OF THE CoMIqlSSiOli PROGRNqFOR STANDARglZATION Of IWCLEAR POMER

PLANTS ANO RECOMIqENOATIONSTO llqPIIOVE ST_UiD_UIDIZATiOII CONCEPTS

STANDARDIZATION SECT'82"128 820325 PROPOSEDLEGISLATION: NLI_EAR STANDRI_IZATIOR ACT OF 19_.STANDARDi ZATI ON SECT"85 - 382 851 _q_ STANDAROI ZATION POLI CY STATENENTSTANDARO|ZATIC]_I SECY-87'-010 870113 FEE OPTIONS Fgll STANDARIZEO REACTORDESIGNSSTANDARDIZATION SECY-88-169 880620 RULEMAKINGON STANDARDIZATION AND LICIEIi$1NG REFOIOISTANDARDIZATION SECT-89-036 890200 RUILEIqAKINGON EARLY SITE PIERNITSo DESIGN CERTIFICATIONS, AM) C011BIHED

L!CENES

C/3 _]_ STANDARDIZATION SECY-91-4_I 911200 NRC'S ABILITY TO ISSUE A COIqSINED LICIE11_ (COL) TO A CONS_TIUIqCD G) STANDARDIZATION SERCHTOP-11A BSIO00 STANORROIZATIONSTATUS"C3"_ " STATION BLACKOUT GENLTR-81-O0_ 810225 EIqERGENCTPROCEINJRESAND TRAINING FOR STATION 8LACIOgUTEVENTS(I)3 0 STATION BLACKOUT GE11LTR-91-O07 910502 GI-Z3, nll_qCTOR CIX)LANT PUNP SEAL FAILIJIIES, u ANO ITS POTENTIAL IRPACT ONC_" STATI ON 11LACIC_r_CD STATION BLAC3(OUT IEN-8_-07'6 841019 LOSS OF ALL AC POMER-_ STATION 8LAC3(OUT 1EN-85-077 850920 POSSIBLE LOSS OF ElqERGENCYNOTIFICATION SYSTEN DUE TO LOSS OF AC POIIEII

rdD STATION BLACKI3UT UEG-1032 850131 EVALUATION OF STATION BLAC1CI3UTACCIDENTS AT IU_EAR POUER PLANTS, TECIINICALrd_ FINDINGS RELATED TO UNRESOLVEDSAFETY ISSUE A-/_, OtAFTr_3

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SERVICE FOR EVALUATING REGULATORY CHANGESSUBJECT: Proposed RLIle oll Elrly Site 9orNi_l; St:llldlrd DoJt_11

• Certi_ications; and Combined Licenses for Nuclear PowerReactors

DATE: August: 23, 1988

NUMBER: RC. 88 -45 KEY SUBJECT: S_andardization

The Comm£siion is considerin 8 addin 8 I new pare Co i_l resullCions (10 CFRPare 52) which would provide for issuance of early lice permits, s_andarddesign certifications, and combined construction permits and conditionalopera_ing licenses for nuclear power plan_s. The proposed rule se_s ou__he review procedures and licensin_ requiremen_s _ha_ would apply _oapplications for _hese new licenses and certifications.

There are twelve questions included in _he proposed rule package on which_he Commission would like _o receive commence. The commen_ period expiresOctober 24, 1988.

A de_ailed analysis of _he rule is available _o in_eros_ed individuals.The analysis and/or the actual rule may be requested from the SERCH S_aff.

BECHTEL POWER CORPORATION

_ 15740 Shady Grove Road, Gaithersburg, Maryland 20877.1454(301) 258-3099 Rev. 0

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SUB3ECT: Final Rule on Early Site Permits, Standard Design Certiftcatiorm,and Combined Licenses for Nuclear Power Reactorm - 10 CFR Parts 2,50, 51, 52, and 170

DATE: April 18, 1989

NUMBER: RC-89-22 KEY SUBJECT: Standardization

On April 18, 1989, the NRC posted notice (5_FR15372) of m final rule on early

site permits, standard design certifications, and combined licenses fornuclear power reactors. The new part facilitates the early resolution ofsafety issues by providing for pre-construction permit approval of power plantsites, Commission certification of standardized designs, and the issuance oflicenses which combine permission to construct a plant with permission to

operate it once construction of it has been successfully completed. The ruleis intended to settle contentious issues (e.g., emergency planning and the

safety and adequacy of proposed designs) prior to construction through theissuance of a combined construction permit and conditional operating license.

The final rule addressed public comment on several issue_. With regard todesign certification applications, the final rule is clearer than the proposedrule was in identifying those designs which cannot be certified without aprogram of testing (see SERCH MS-89-04). The rule distinguishes between alladvanced designs, both passive light-water and non-light-water, andevolutionary light-water designs. Some testing may be required of alladvanced designs. Since passive light-water designs have features which arenot present on plants licensed and operated in the United States, the rulerequires that the maturity of these designs be demonstrated through acombination of experience, appropriate tests or analyses, but most likely notthrough prototype testing. Full-scale prototype testing is likely to berequired for certification of advanced non-light-water designs since theserevolutionary designs use innovative means which have not yet been licensed inthe United States to accomplish their safety functions (e.g., passive decayheat removal and reactivity control).

The Commission responded to comments and clarified several issues regardingthe requirements for the contents of applications (10 CFR Part 50._7). Thefinal rule, like the proposed rule, requires applicants to propose technicalresolutions of Unresolved Safety Issues and high- and medium-priority CenericSafety Issues. Applicants must show either that a particular issue is notrelevant to the design or that the applicant has in hand a design-specificresolution of the issue. The application also must demonstrate compliance

with any technically relevant portions of the TMI requirements set forth in 10CFR 50.3_(f). The inclusion of this severe accident requirement for future

plants in the standardization rule is, at least for the present, taking the


i_ 5740 Shady Grove Road, Gaithersburg, Maryland 20877-1454(301) 258-3099

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place of a separate severe accident rulemaking for future plants (|ee SF_RCHMS.89.73).

As to the level of design information which an application muit contain, the

final rule is more mtringent about the completenesm of the propomed designthan the proposed rule was (see S_RCH MS-89-04). Applications forcertification of evolutionary designs must provide an essentially completenuclear power plant design: a design that includes all structures, systems,and components which can affect the safe operation of the plant except forsite-specific elements such as the service water intaks structure and the

ultimate heat sink. In addition, an essentially complete design is a designthat has been finalized to the point that procurement specifications andconstruction and installation specifications can be completed and madeavailable for audit if it is determined that they are required for Conuaissionreview.

One of the mos_ contentious issues that received public ooau_ent was whetherthe final rule would provide the opportunity for a post-construction hearingand to what extent the Commission would limit the hearing's scope. The finalrule provides an opportunity for a limited public hearing after completion ofconstruction. The Conuaission will confine the hearin B to the single issuewhich could not have been litigated earlier -- whether the design, as built,complies with the terms of the license and the NRC rules and regulations. The

Con_ntssion does not plan to seek additional congressional authority vialegislation in order to eliminate the post.construction hearing entirely,

Copies of the rule may be requested from the SERCH staff.

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KEETINC: ConliJsAon Briefin| on the Current Status of the Final Rule onStandardization and Licensing Reform.

DATe-: February 22, 1989

_EP.: HS-89-36 KEY SUBJECT: Standardization

Licensing

The ComniJsion vas briefed by William Parler, General Counsel, on the proposedfinal rule on Standardization and Licensin| Reform. i/r. Parler provided anoverview of the procedural aspects.

M_. Parler stated that the proposed rule could increase safety, providegreater reliability and shorten constr_ctlon periods. These benefits would bea result of the removal of delays, uncercalncy and llcenslng costs. Theproposed rule (SECY-89-36) addresses some of the concerns of the NRC and thepublic, The primary general coBents received on the proposed rule were:

o There was a broad agreement chat standardization and earlyresolution of 1tcenslng issues are desirable.

• There should be an opportunity for having a hearing prior cooperation,

o The I_C agreed with the ACRS in that the rule should include aprovision that requires licensees to provide specific details ofmaterials and equipment (as in procurement and installationspecifications) when submitting a design for review.

There were several comments received in support of the early site permit(ESP). The flnallty of ESPs depends on the extent to which assumptions playeda role in slte specifics. For example, if the licensee has numerousassumptions in the ESP package, the finality of the permit is lessened becauseIt will be subject to questioning. Prior to construction, the Commissionwould evaluate the permit to see that all its terms have been met. Requestsfor variances from the permit will be permitted. However, these requests maybe subject to litigatLon.

In the area of combined licenses, the proposed final rule is more precise onwhat issues can be raised in a hearing followins construction. The Commissionwill make the final decision as to whether a hearing is necessary and whatprocedures will be followed.

In order Co preserve finality in the proposed rule, the following provisionswere made: I) amendments requested by the holder will be permitted after


1_ 15740 Shady Grove Road, Galthersburg, Maryland 20877.1454(301) 258.3099E'.:.-....."....................., . ....... . ,.,... ._..,...,.. -",-',':',........,..,......_"......."-........,v'-'_...... '."_........Rev.O-.---,.-.--,,,,..... •............

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rulemakinB, 2) no backfic Co a certified desia_n will be imposed unless ic isnecessary co assure compliance, and 3) chanses vii1 noc be made to a certifieddeslsn unless an exemption is 6ranted for compliance with h3tC resulacion.

In closing, Chairman Zech commended Hr. Parler and Hr. Scello for their Jointefforts, However, he commenced chac the rule could not be finalized until theScarf addresmes the concern as co what level of detail should be required.

Copies of SEC?-89-36 and the meecin_ handouts are available through the SERCHScarf.

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_ETING: Commission Briefing on the Status of Efforts to Develop an Updated Source Term

DATE: January 9, 1990

N'UMBER: MS-90.01 KEY S_CT: Source TermlSiting

Members of the NRC Staff including James Taylor, Executive Dlrector for Operations, ThomasMurley, Director of the Office of Nuclear Reactor Regulation (NIt.R), Eric Beckjord, Director ofthe Office of Nuclear Regulatory Research (RF_), Themis Speis, Deputy Director for Genetic IssueResolution, RF_, and Leonard Softer, RE.S, briefed the Commissioners on the status of effortsto develop an updated source term methodology. More specifically, the possibility of decouplingsiting from plant design was discussed for future plants as well as the identification of otherregulatory applications for the updated source term.

James Taylor opened the meeting by noting that a 3- to (>-monthStaff study is currently under wayto examine the implications of decoupllng plant siting from source term considerations. Accordingto Murley, decoupling here would mean not having to do dose calculations for siting purposes.Although the Staff is leaning toward decoupling, it is not sure how this should be achieved and the

_, siting study is intended to investigate/identify the options to be recommended to the Commission.

The Staff presentation, provided by Speis and Softer, covered: current practice, i.e., TID.14844,"Calculationof Distance Factors for Power and Test Reactor Sites," and its relationship to designbasis and to containment performance; research insightswith respect to accident source terms',theissues pertaining to the updating of TID.14844; applications of the updated source term such assiting, plant design, and emergency planning; and future Staff actions. These issues are alsodiscussed in SECY-89-341, "Updated Light Water Reactor (LWR) Source Term Methodology andPotential Regulatory Applications," which was made available to the public during this meeting.

Present siting practice postulates an instantaneous TID.14844 release into the containment and,based on this release, calculates offsite doses for site suitability determinations according to Part100 siting requirements. It was noted that the TID-14844 release has also had wide regulatoryapplications beyond siting, such as control room habitability,equipment qualification, post.accidentsampling systems, and isolation valve closure time. TID.14844 is largely based on late 1950sexperimental results of heating UOa pellets. The present Staff methodology is described inRegulatory Guides 1.3 and 1.4, and SRP 15.6.5 appendices.

Major NRC research efforts have been under way since about 1981 to obtain a betterunderstanding of fission product quantity, form, transport and release mechanisms. Thereare manysource terms, depending upon plant design and accident sequence. There is consensus that currentStaff practice regarding the timing of fission product release is overly conservative (the TID assumes

_, _ BECHTEL POWER CORPORATION15740 Shady Grove Road, Gaithersburg, Maryland 20877-1454(301} 258-3099 RAvO

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that the source term appears in the containment instantaneously). The largest single factoraffecting the source term is considered to be containment integrity where delay in containmentfailure reduces the source term substantially.

With respect to the issue of design basis vs. severe accident source terms, the design basis accidentsource terms are releases not to the environment, but rather into the containment. These are used

in licensing to assess site suitability and to evaluate certain plant systems. On the other hand,severe accident source terms are releases to the environment. They were first studied in riskassessments (e.g., WASH-1400) when examining sequences involving core melt and containmentfailure. They are not used in individual plant licensing, but they have significant regulatoryapplications in areas such as emergency planning.

With this background, the Staff discussed the source term update considerations.

, Updating TID.14844: With respect to updating TID.14844, research insights confirm thatthe TID.14844 approach, while providing substantial plant mitigation capability, is notcompatible with a realistic understanding of the progression of severe accidents. That is,the "lid has both conservative and non.conservative aspects.

As part of the "riD update effort, fission product timing, iodine's chemical form, thequantities of fission products released, and the extent to which nuclides other than noblegasesandiodinearereleasedaresomeoftheareastobe investigated,(Speisindicatedthatone totwoyearsofresearchmay be neededtocome togripswiththeiodineissue.)TheSt_fffdoesnotbelievethatachangetoPart100isrequiredtoreviseTID.14844;however,revisionof RegulatoryGuides1.3and 1.4and associatedSRPs would be neededtoincorporatetherevisedinsights.CarrrecommendedthattheStaffnotwaitinimplementingthischange.The Staffrespondedthatitintendstoproceedwiththismodificationinparallelto the siting study.

Also to be addressed as part of the TID update effort are three key issues: (1) What typeof accidents should be used to define release into the containment? For example, shouldreactorpressurevesselfailurebe assumed?The Staffnotedthatthisisprimarilya policyissuethattheCommissionwouldneedtoaddress.(2)What shouldbe therelationshipbetween assumed accidentconditions,releaseintocontainmentand containment

design/performancecriteria?The StaffnotedthattheACRS willbediscussingthismatteron January11. (3)What shouldbetherelationshipoftheupdatedTID tositing?

o Regulatory applications of updated source terms: The Staff previously identified (in SECY.86.76) potential changes making use of updated source term information. The short.termchanges that had been identified in SECY-86-76 are completed. In addition, three majorareas for potential application are as follows:

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1. Siting: The updated TID.14844 could be used for siting purposes, as in currentpractice. However, some Staff concerns are that (1) the 'lid does not directlyaddress containment performance, and (2) a best-estimate evaluation of fissionproduct cleanup systems assuming a large in-containment release would permit muchsmaller exclusion low population zone areas than previously allowed. The Staffbelieves that decoupling siting from plant design and elimination of dose calculationsin siting may have merit.

2. Plant Desiii,. Up,iated source term insights could have significant impact uponcertain plant systems and design features, such as control room habitability systems,fission product cleanup =;ystems,isolation valve closure time, containment leak rate,equipment qualification. The Staff intends to examine the updated source terminsights and apply them as appropriate to these areas (an upcoming Staff paper willdiscuss the Staff's recommendations). Chairman Carr noted that he hopes theseinsights will be applied to current plants, not just future plants. Taylor respondedthat they would be.

3. Emergency Planning: With respect to updated source terms and emergency planningrequirements, the Staff commented that the sizes of present emergency planningzones (EPZ,s) are based on both policy and technical considerations includingconsequences of design basis and severe accidents, using results from WASH.1400,Updated source term information from NUREO-I150 indicates that, for earlycontainment failures including bypass that have the greatest impact on risk, revisedsource terms are somewhat lower than predicted byWASH-1400. However, changesto the existing EPZ requirements are not being recommended by the Staff, Whenquestioned by Chairman Cart and Commissioner Roberts why not, Taylor respondedthat the Staff does not want to take on this subject at this time, It would ratherfocus on other source term.related issues. Commissioners Curtiss and Remick were

interested in knowing if NUREG.1150 had been available at the time the EPZ wasbeing established, what EPZ would have been selected? Curtiss was also interestedin knowing if there was a technical basis for establishing the 10-mile EPZ and if sowhat it was. The Staff did not know and indicated that it would attempt to find outin the future,

The Staff concluded its presentation by noting that future Staff actions include the previouslymentioned siting study for examining decoup!ing of plant siting from plant design for futurereactors. This study, which will take about 6 months, will also assess issues and implications forcurrent plants. Recommendations will be made to the Commission in June 1990. The Staff alsointends to pursue potential source term modifications for future LWRs based upon updatedresearch insights. Short-term changes (e,g,, fission product timing) are to be completed within 6to 9 months. The Staff feels that it already has enough information to proceed with Regulatory

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Guide/SRP changes at this time. Longer-term changes (e.g., iodine chemical form) may !ake upto 2 years and require additional research.

During the meeting, Commissioners Curtiss and Remick noted that they are interested in the Staffsplans for possibly combining the design basis and severe accident source terms. When asked byCurtiss whether the Staff envisions setting siting criteria based solely on probabilistic risk

considerations, Murley responded that he would not feel comfortable with such an approach. Can"

agreed.

Carr raised an issue regarding the draft EPRI ALWR SER which says that a containment

performance objective is not necessary. "When did the Staff reach this conclusion?", he asked.Murley commented that this matter is still under discussion with EPRI. The EPRI guidelines forsevere accidents and core melt are more stringent than the safety goal but they do not include acontainment design objective/criterion. The Staff feels that there should be a containment designobjective and has established a 0.1 failure probability until an objective is developed for generic

application.

In their concluding remarks, Rernick and Cart encouraged the Staff's decoupling efforts. Carr alsoencouraged the Staff to accelerate implementation of the fission product release changes. He feelsthat the 1- to 2-year schedule for implementing iodine insights should be accelerated as much as

possible. Carr also emphasized the need for applying updated source term insights/conclusions tocurrent plants as they become available. Curtiss requested that the Staff provide the Commission(1) further amplification of the risk-based approach to the source term, that is, the advantages anddisadvantages of using PRAs; (2) a list of the purely risk-driven considerations; and (3) a descriptionof what iodine research still needs to be done.

Handouts from this meeting, including SECY-89-341, are available upon request from the SERCHStaff.


SERVICE FOR EVALUATING REGULATORY CHANGES®

MEETING: CommissionBriefingon SourceTerm Updateand I_upHng SitingfromDesign

DATE: October15,1990

NUM'BER: MS-90-124 KEY SUBJECT: SitingSourceTerms

The CommissionwasbriefedbytheStaffon thestatusofproposedchangestotheregulationsonsourcetermsandsiting.The draftchangesinTID-14844willbeavailableforreviewinJuly1991.The StaffwillissuetheproposedPart100rulechangetoaddsitecriteriainDecember1991.ThefinalrulewillbegiventotheCommissioninFebruary1993.Part100and Part50rulechangeson dosecalculationandplantdesigncriteriawillbe issuedpriortocompletionofpassiveLWRdesignreview.

ThomasKing(NRR) spokeon theStaff'sprogressontheproposedchanges.A StaffRequirementsMemorandum (SRM) datedJuly31,1989requestedthattheStaffprovidea paperon theextenttowhichthecurrentsourcetermcanbeupdatedorimprovedforfutureLWRs. An SRM dated

February13,1990directedtheStafftoproposechangestoregulatorypositionsforcurrentandfutureplantswherethecurrentunderstandingofsourcetermwouldpermit./

A sourcetermisthereleaseoffissionproductsintocontainmentthatispotentiallyavailableforreleasetotheenvironment.The sourcetermincludesthetiming,form,andquantityoffission

products.Designbasisaccidentsourceterms(TID-14844)areusedinlicensinginthreedistinctwa3_:

o For siting evaluations as required by 10 CFR 100.

o To defineradiologicalenvironmentconditionsforcertainplantsystems.

o To assess the effectiveness of plant mitigation features.

There are various reasons why the Staff is changing the siting requirements. TID-14844 did notoriginally give credit for fission product cleanup systems. As plant size increased, however, fissionproduct cleanup systems became a factor to keep exclusion areas from becoming too large.Another reason is that Part 100 contains no other siting criteria than population center distance.Furthermore, Part 100 does not address challenges to containment or containment failure. Also,the current practice does not include an updated understanding of source terms. Therefore, theStaff is proposing changes to the current practice for future LWRs that would:

. Give designers more flexa'bilityto develop designs with safety features that utilizerealistic source term assumptions.


_9801 Blvd., Gaithersburg, Maryland 20878-5356Washingtonian

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o More directly address acceptable site parameters and containment performance.

The overall plan for revision requires the Staff to develop a technical update of TID-14844. Workis underway to update the technical basis for the source term based upon current severe accidentresearch. The updated TID-14844 will reflect changes in:

o timing of fission product release (impact and status given in SECY-90-307);

o composition and magnitude of the release into containment; and

o iodine chemical form.

The draft of the updated TID report is scheduled for 3uly 1991. Existing plants could voluntarilypropose to use the updated TID on a case-by-case basis.

Decoupling of siting and design will require two steps. First, siting criteria will be taken fromRegulatory Guide (RG) 4.7, "General Site Suitability Criteria for Nuclear Power Stations," andadded to Part 100. Second, the engineered safety feature (ESF) design and performancerequirements will be specified in Part 50 based on updated radiological conditions and severeaccident insights. Also, the source term and dose calculation portion will be removed from Part100. Advanced notice of the proposed rulemaking is to be issued in March 1991.

Commissioners Curtiss and Remick suggested that, instead of retaining the source term portion inPart 100 and moving it later, the Staff should move the source term to Part $0 once the updatedTID is incorporated in Part 100. Remick asked if all of RG 4.7 was going to be added to Part 100.K/ng said that only the site conditions and safety parameters of RG 4.7 will be added. ChairmanCarr closed the meeting with emphasis that future designs should be based on good regulations andplanning and that the Staff should give further review on containment releases for severe accidents.

Copies of a 26-page handout on the Staff's presentation are available through the SERCH Staff.

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MEETING: NRC Staff Meeting With NUMARC on the Revision of 10 CFR Part 100,Appendix A

DATE: March 6, 1991

NUMBER: MS-91-31 KEY SUBJECT: SitingSeismic Design

NRC Staff Members Present Included:

Larry Shao, Director, Division of Engineering (DE), Office of Nuclear Regulatory Research (RES)Robert Bosnak, Deputy Director of DE, RESAndrew Murphy, Chief, Structural & Seismic Engineering Branch (SSEB), DE, RESRoger Kenneally, SSEB, RESRichard McMullen, SSEB, RESNilesh Choksiai, SSEB, RESJohn Chen, Severe Accident Issues Branch (SAIB), RESGoutam Bagchi, Chief, Structural & Geosciences Branch (ESGB), Division of Engineering

Technology (DET), Office of Nuclear Reactor Regulation (NRR)Phyllis Sobel, ESGB, NRRMark Hartzman, Mechanical Engineering Branch (EMEB), DET, NRRRon Ballard, Chief, Geosciences & Systems Performance Branch (HLGP), Office of Nuclear

Material Safety & Safeguards (NMSS)Patty Jehle, Office of the General Counsel (OGC)Mat Taylor, Office of the Executive Director for Operations (EDO)

Industry Members Present Included:

Ray Ng, NUMARCOrhan Gurbuz, NUMARCCarl Stepp, EPRIMike Hayner, Cleveland Electric Illuminating CompanyTom O'Hara, Yankee AtomicDennis Ostrom, Southern California Edison CompanyBob Whorton, South Carolina Electric & Gas

The meeting was heavily attended by members of DOE. Also present were representatives fromthe Defense Nuclear Facilities Safety Board, LLNL, several industry consultants, and NUMARC'sAd Hoc Advisory Committee.

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Larry Shao opened the meeting by noting that RES is in the process of drafting a revision toPart 1130,Appendix A, "Seismic and Geologic Siting Criteria for Nuclear Power Plants." TheAppendix was first issued in 1973 and the NRC wants to bring it up to date. C.onsistent with theCommission's directive that criteria be in place in time to support the review of an early siteapplication in FY 1993, RES is aiming to issue the revised Appendix A final rule (along with anysupporting regulatory guides) by October 1992. Working back from that date, he noted that thefirst draft would need to be completed (by RES) by May 1, 1991. The tentative schedule is torequest NRC Staff concurrence on the draft rule on August 1, and to issue the ta, "q rule for publiccomment by December 30, 1991.

This was the first public meeting held to kick off the lulemaking effort. The objective of themeeting was for the Structural & Seismic Engineering Brarich to identify some of the areas it thinkswould need to be revised and to hear what other areas NUMARC and any other industryrepresentatives want to add. Orhan Gurbuz commented that NUMARC was not ready to discussdetails in this meeting because it had not had enough time to review all the Appendix A technicalissues. He suggested that another meeting be held to pursue such discussions. Ray Ng added thatNUMARC's Seismic Working Group is scheduled to meet in April after which NUMARC will beable to offer some input.

Andy Murphy introduced the Staff presenters -- Roger Kenneally and Dick McMullen -- andpointed out that the Staff has already started working with its contractors on the revision ofAppendix A. Some of the Staffs preliminary objectives are to clarify the role of the OBE; to takeout some of the geological/seismic engineering items out of Appendix A and place them in Part 50;to clarify or possibly change some of the geological terms such as capable fault, tectonic province,and vibratory ground motion; to take out the seismic design guidance that is thought to be toospecific and to possibly place it in a regulatory guide (RG); to address the use of probabilistictechniques (the Staff intends to allow the use of probabilistic analysis); and to address the use ofthe LLNL and EPRI seismic hazard studies. According to Murphy, the revise:_ Appendix A isexpected to be shorter than it currently is.

NRC's Perspective on the Engineering Issues:

Roger Kenneally presented the Staff's thoughts on the engineering issues in Appendix A. He notedthat, as currently written, Appendix A:

o Defines the SSE and OBE--in Section III(c) and III(d).o Defines safety-related structures, systems, and components -- in Section Ill(c), VI(a)(1)(i-iii),

and Vl(a)(2).o Defines the minimum value of the SSE -- in Section V(a)(1)(v).o Establishes the OBE/SSE ratio -- in Section V(a)(2).o Requires plant shutdown if the OBE is exceeded -- in Section V(a)(2).

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o States that vibratory ground motion is defined by response spectra at the elevations of thefoundations of structures -- in Section VI(a)(1) and (2).

o Identifies acceptable analytical methods such as dynamic analysis, qualification test,equivalent static method; states that SSI effects and expected duration of motion must beconsidered; and allows strain limits in excess of yield .. in Section VI(a)(1) and (2).

o Requires that plant designs consider surface faulting -- in Section VI(b).o Requires that plants be designed for seismically.induced floods and water waves -. in

Section VI(c).o Identifies various soil considerations that must be incorporated in the plant's design -- in

Section V(d)(1-4).

Based on the Commission's Staff Requirements Memorandum regarding the revision of Part 100,the Staff is expected to focus predominantly on site selection and the establishment of the SSE, andto ensure that uncertainties are fully accounted for (without anticipation of what further researchmight show in the future). Also, the Staff is to move engineering design issues into Part 50. Atthis point, the Staff does not know what portions will be placed in Part 50 but Shao noted thatmost of the seismic design guidance will be removed from Appendix A. Kenneally added that thisdetermination will be made with OGC's input.

Kenneally continued his presentation by referring to three areas that the Staff is particularlyinterested in obtaining industry input on:

o Regarding the concept of the OBE, the Staff is exploring the following questions: Shouldthe OBE be retained as a design requirement? For example, should it be used in the designof all safety-related structures, systems, and components, or only on some of them (and ifso, which ones)? Should the OBE be an inspection only earthquake (i.e., where the utilitywould possibly reduce power and perform a walkdown of the plant to ensure no damagehas occurred)? If the OBE is redefined, how would hazard information -- both spectra andreturn period -- influence the numerical value assigned to the OBE? What would be thesignificance of removing the OBE from design considerations or changing the OBE/SSEratio? That is, how would such a change affect the seismic margin and seismic risk,displacement-related failure criteria, fatigue requirements, national standards (e.g., ASME,ACI), and load combination-related acceptance criteria.

o Regarding plant response to an earthquake, the Staff wants to establish an evaluationthreshold and to develop a threshold exceedance criterion. The Staff believes thatguidelines for orderly plant shutdown and inspection will be needed, and intends to useEPRI and ANS input on this matter.

o Regarding ground motion designation, the current position on specifying ground motion isbased on response spectra, foundation elevation, and a minimum ground motion level (O.lgat the foundation level). Although Standard Review Plan sections were updated as part

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of the resolution of USI A-40, the Staff wants to know whether there are any otheralternatives that need to be explored (such as uniform hazard spectra and high frequencyground motion).

Kenneally ended his presentation by pointing out that the Staff is not trying to reinvent the wheelin these areas but would like suggestions from the industry on how to address them.

NRC's Perspective on the Geosclences Issues:

Dick McMullen discussed the geosciences issues that the Staff is considering revising/addressinginAppendix A. The Staff has identified seven issues: seismic sources, potentially hazardous faults,characteristics of soils and rocks, vibratory ground motions, tsunami or water waves, man-inducedseismic or fault displacement hazards, and volcanic hazards, in addressing these issues, McMullennoted that the Appendix A revision should emphasize the importance of deterministic an_dprobabilistic studies, should ascertain that acceptable up-to-date methodologies are used for bothtypes of studies, and should provide guidance on how the two analyses will be merged. Each ofthe seven issues was further discussed as follows:

1. Designation of seismic sources: Seismic sources are referred to as tectonic provinces andtectonic structures in Appendix A and are based on surface geology, tectonic structure, andphysiography. In the eastern and central U.S., most seismogenic structures(earthquake-generating faults) are apparently not exposed at the ground surface. In thewestern U.S., they are not always exposed at ground surface. The Staff wants to expandthe present concept of seismic sources to include consideration of seismicity, paleoseismicity,and tectonic structure deep within the earth's crust. In this regard, the Staff is asking thefollowing question: is it feasible to develop a seismic source zone map?

2. Identification and assessment of..poten!ially hazardous faults: Appendix A refers to suchfaults as capable faults. The Staff wants to: expand the current concept of potentiallyhazardous faults to include consideration of seismic potential, blind faults, and associatednear surface deformation; address the distinction between tectonic and nonteetonic faults;and expand the investigation requirements to assess additional fault characteristics (especiallyin the western U.S.) such as paleoseismicity, slip rate, history of displacements, temporalclustering, sense of displacements, and down dip geometry.

3. Determinat.ion of the maximum _round motions and other significant levels of shakimz: Themain goal of Appendix A is the determination of the design earthquake (SSE). The Staffnow is wondering whether the "maximum" ground motion is an appropriate parameter forthis determination or whether the "85%" (the 85th percentile of the estimated groundmotion distribution curve) should be used. Basically, the Staff wants to broaden theregulation to take into account other characteristics of earthquakes than just magnitudeand intensity (e.g., corner frequency and stress drop); consider the effects of sense of

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displacement and the location of the site with respect to the causative fault on groundmotion; and provide guidance on how to go from the ground motion determined byinvestigations to the ground motion used in the design.

4. Determination of the characteristics of soils and rocks: This is done to determine how soilsand rocks affect ground motion propagation both on a regional and a site-specific level andto assess the stability of soils and rocks under dynamic loading. Other potential nonseismicground disruption hazards such as landslides or subsidence are also considered. McMullendid not offer specific insights on how the Staff may revise Appendix A with respect to thisissue.

5. Assessment of.t.sunamiand water wave hazard.: This pertains to coastal sites or sites locatednear large lakes. It includes sea or lake floor offsets, landslides, or subsea slidesaccompanied by or triggered by earthquakes. Specific insights on the Staff's thinking in thisarea were not provided.

6. Evaluation 9f.man-induced seismic. 0r fault displacemen!, hazards: This includes fluidwithdrawal from or injection into the subsurface, large excavation or mining, and reservoirimpoundment. McMullen did not expand on this issue.

7. Volcanic hazard assessment: The Staff thinks that a document should be written on thissubject. However, except for the possibility of seismicity and ground rupture related tovolcanic activity, the document would only be applicable to limited regions in the U.S. (forexample, the Pacific northwest or Alaska). In view of this, the Staff would like to knowwhether this issue needs to be addressed in Appendix A.

McMullen noted that the Staff has not decided yet where to go with these issues but is reviewingthem with LLNL.

Tom O'Hara (Yankee Atomic) asked where the tectonic province will be defined, in a RG or inAppendix A? Andy Murphy responded that it probably will be in Appendix A. Carl Stepp (EPRI)commented that part of the problem in deciding where certain things should be placed is notknowing how prescriptive Appendix A needs to be. He pointed out that such a decision woulddepend on the legal aspects of what constitutes an adequate or proper regulation. In his view, abalance between a highly prescriptive regulation and one that is more general needs to beestablished before the Staff and industry can decide where certain technical information needs togo,

Step0 also noted that one cannot compile all the new seismic and geological technology of the past15 years in RGs within 2 months. Bosnak agreed but Chokshi stated that all RGs do not need tobe available before Appendix A is revised. However, Murphy pointed out that the Staff is aiming

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tohavetheRG(s)onthedesignandgeologicalaspectsdraftedwithinthesameschedulesthatwerementionedearlierfortherule.

O'Haraaskedwhetherthereisa mechanismforobtainingcommentsfromindividuals.Murphysaidthatindividualsmay providetheirthoughts/commentsdirectlytotheOfficeofResearchnow,beforetheofficialpubliccommentperiodon thedraftAppendixA revisionbegins.

One oftherepresentativessuggestedthattheNRC conductacasehistory/lessonslearnedreviewtoeliminatepotentialmisinterpretationsintheregulation.Murphynotedthatsofarthishasnotbeendone.

Inresponsetoaquestionastowhether/howthenew AppendixA willbeappliedtoliccnscrenewalconcernsthatmay arise,Murphystatedthatifa problemisidentifiedatanexistingplantandtheutilitywishestousetheAppendixA revisiontoresolveit,allaspectsoftheAppendixA revisionwillneedtobe consideredinresolvingtheconcern,notjusttheone ortwo thathappentobcadvantageous.(He notedthatitwouldbeappliedinasimilarfashionasAppendixJrevisionshavebeen.)He addedthatinsucha case,he thinksthata probabilisticapproachwouldbe allowedinresolvingtheproblem.

Inresponsetoquestionson how many RG-stheStaffenvisionsrevisingororiginatingtosupporttheAppendixA revision,Murphy statedthathe didnotknow yet;itcouldpossiblybc twelve.He clarifiedthattheRGs areonlyintendedtoaddresswhatiscurrentlyinAppendixA --notallpossibleareasoravailabletechnologies(e.g.,dampingofpipingsystems)thatpertaintoAppendixAissues.MurphypointedoutthattheAppendixA revisionandthcRGs willbcbasedoncurrentlyavailableinformation.The Staffwillnotwait,forexample,toreconciletheEPRI and LLNLseismichazardcurves.

DennisOstromofSouthernCaliforniaEdisonsuggesteddeferringvarioustechnicalissuesforthcRGs. Mikc HaynerofClevclandElectricofferedthefollowingcommentsastowhathc thinksshouldbc addressedintheAppendixA revision:separationofthcOBE fromthedesigncriteriashouldbe pursued;thesignificanceofOBE cxcccdanceshouldbe clarified--possiblyreducingpoweranddoingan investigationinsteadofshuttingdown wouldbe more appropriate;theSSEshouldbercnamcdbecausesafeshutdownhasa badconnotation--designbasisearthquakewouldbc better',thctermcapablefaultshouldbe eliminatedand replacedby a more understandableterm',and theeasternandwesternU.S.scismicityissuesshouldbc addressed.

MurphyclosedthemeetingbyencouragingthesubmittalofNUMAJ_C orindividualcommentsassoonaspossibletosupporttheStaff'sMay IstdeadlinefordraftingtheAppendixA revision.

The followinghandoutsareavailableuponrequestfromtheSERCH Staff:

a) a 7-pagevugraphofKenncally'spresentation

b) a 9.pagevugraphofMcMullcn'spresentation

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NUMARC and industryrepresentativesmet with the Staff to discussissuesrelated to 10 CFRPart 2. Two issuesthat were discussedat the meeting included:(1) the relationshipbetweentheissuanceof the Final Design Approval (FDA) for a reactor designand the Inspection,Test,Analysis,and AcceptanceCriteria (ITAAC) processand (2) siting in relation to the NationalEnvironmentalPolicyAct (NEPA) SevereAccidentMitigationDesignAlternative(SAMDA) _ue.

Marc Rowden,who w= representing_C, stressedthe industry'spositionon the designcertificationprocess.NUMARC believesthat an FDA canbe _ue._lwith openitemsandwithoutthe ITAAC. NUMARC wan= to know why the Commission, in the Staff RequirementsMemorandum to SECY.90.377, stated that ITAAC is not necessary for safety but that both Tier 1(design information that is certified) and Tier 2 (design information that is not certified but isincluded in the certification application) information is needed for FDA. Tom Murley, Director,NRR, said that the Staff intends to seek further Commission guidance on this _ue, The onlyproblem thet the Staff has with leaving ITAAC out of the FDA process is that it would preferthat the design reviewers be the same people that develop ITAAC in order to ensure a uniformdesign review. If ITAAC is not fully developed, the Staff would at least require the NSSS vendor'sthoughts on ITAAC at the time of SER submittal. Rowden said that NUMARC has no problemwith that but is concerned that ITAAC will not be completed when the SER is done.

Martin Malsch, Deputy General Counsel for Licensing and Regulations, OGC, asked if the SERcan be issued if ITAAC are not defined. Rowden replied that the indust_, realizes that someITAAC determinations must be made at the FDA stage but they would like to decouple the two.Dennis Crutchfield, Director, Division of Reactor Projects - HI, IV, V & Special Projects, NRR,said that the Staff is afraid that there will be no feedback between the design and the actualhardware.

Rowdcn asked the Staff what is actually considered Tier 1 and Tier 2. Crutchfield replied that theStaff is making a determination on these items as the certification process is being developed.Rowden said that NUMARC would prefer to have the information as soon as possible so that itcan make a determination on what is necessary for ITAAC.

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,'ltOIi .NUMARC realizes that a design must be complete when it is submitted for design certification, butthere is a problem with retrofitting already.submitted designs to Part 52. Malsch said that thereis some difficulty with using Part 50 for already-submitted designs because it does not apply toadvanced designs and new standards are needed. Rowden asked Malsch if he is saying thatadditional rulemaking is needed for certification and Part 50 does not apply. Malsch replied bysaying that there is concern that if the Staff goes beyond Part 50 on an issue, then an intervenorwill take all issues beyond Part 50. He emphasizedthat, to alleviate this concern, a rulemakingonstandards will be done only at the beginning of the certification process.

Rowden voiced concern that the Staff is revie_ng issues one at a time that are interrelated.NUMARC wants to take care of them all at once. After the FDA is issued, the Staff should onlyrequire a change in the design if it is of safety significance, Dave Rehn, Duke Power Company,said that NUMARC's overall goal is to make a safety determination at the design certification stageso that safety issues are not brought up again at the combined operating license (COL) stage.

According to the Staff, a major portion of the balance of plant has an effect on safety; the Staffis still developing a definition of what exactly is safety.significant. Rowden's concern with thisprocess is that it may open up new issues at COL for hearing consideration. Crutchfield said thatthis is not the intent but it will depend on how detailed the design is.

The Staff will submit a program for ITAAC models to the Commission within a month. The Staffwill work with and get input from the Commission and NUMARC on developing these models.

'tSdng:

Rowden spoke on the NEPA treatment of SAMDA issues. The Court of Appeals decision in theLimerick case requires NEPA analysis of SAMDA issues at the facility licensing stage if there isno prior resolution through rulemaking. NUMARC feels that, under Part 52, an environmentalassessment is needed at the design certification stage but the NEPA environmental impactstatement should not be required. Generic NEPA coverage of SAMDA issues which are specificto a certified design prior to the first COL application for a plant of that design and subsequentpreclusion of facility.specific design changes would assure the siting integrity of a standardize_design.

NUMARC proposes that the NEPA _ues should be left to each individual applicant to decidewhen they should he addressed. NUMARC wants to separate NEPA issues from designcertification because they are not required if NUMARC's S.5 siting approach is used. The S-5approach, a Part 51 rulemakingspecifying environmental effects of SAMDAs on a design-specificbasis, will provide a separate consideration for SAMDAs that meets the Limerick decisionrequirements. Rowden said that NUMARC wants to be sure that the Staff is prepared to reviewthis approach.

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CrutchHeldsaidthatthisapproachisacceptableforLimerickandotherconstructionpermitsissuedinthepastbecausetheyusedalreadyprovendesignsbutthereisa differencebetweenthesedesignsandtheABWR becauseitisonlyonpaper.RowdensaidthatNUMARC wantstoatleastholdopen the possibilit_ythatthe NEPA issuewouldbe separate,but thatthisisapplication-specific.Maisch asked if SAMDA is left out of designcertificationthen is a clarifierneeded in sitingsections that these areaswillbe coveredin SAMDA. Ray Ng, NUMARC, repliedthat anythingwhich is not covered in designcertificationmust be coveredat the COL stage.

Therewere no handoutsd=trtbutedat this meeting.

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MEETING: CommissionBriefingon Statusof Staff Efforts to UpdateSiting/DesignSourceTerm(TID.14844)

DATE, March 28, 1991

NUMBER: MS.91.43 KEY SUBJECT: SitingSource Terms

Members of the NRC StaffincludingJames Taylor,ExecutiveDirectorfor Operations;Thomas Murley,Directorof theOfficeofNuclearReactorRegulation(NRR); Thomas King,DeputyDirectorof Divisionof SafetyIssueResolution(DSIR),Officeof NuclearRegulatoryResearch(RES);AshokThadani,DivisionDirectorofSystemsTechnology,NRR; LeonardSoffer,SectionLeaderof theSevereAccidentIssuesBranch(SAIB),NRR; and ThemisSpeis,RES,briefedtheCommissionerson thestatusofeffortstoupdatethesiting/designsourcetermwhichisdocumentedinTechnicalInformationDocument(TID).14844.Theyalsopresentedpreliminaryresults of studies of fission product iodine chemistry which the Staff, as stated by James Taylor,would like to see in the TID.14844 update. The Staff's overall plan, fission product timing, andEPRI A.LWR proposedsourcetermforevolutionarydesignswerealsodiscussed.SeeMS.90.124,"CommissionBriefingonSourceTerm UpdateandDecoupiingSitingfromDesign,"andMS.90.01,"CommissionBriefingon the Statusof Effortsto Developan UpdatedSourceTerm,"forbackgroundtothisissue.

AccordingtoThemisSpeis,theStaff'soverallplanhasbeen modifiedsincethelastmeeting,(MS.90.124):

o The Staffisproceedingper StaffRequirementsMemorandum (SRM) datedJanuary25,1991.

o Therewillbea singlerevisiontoI0CFR PartI00includingAppendixA,"SeismicandGeologicalSitingCriteriaforNuclearPowerPlants,"andaninterimrevisiontoPart50followedbya finalrevision.Changestonoteincludetheremovalofsourcetermand dosecalculationsandtheadditionofsitecriteriawhicharetobe based

on Reg.Guide 4.7and additionofexistingsourceterm(TID-14844)and dosecalculationstotheinterimrevisiontoPart50. TheseproposedrulechangesareexpectedtobeissuedforcommentbyDecember1991.

o A technicalupdateofTID-14844w_llbedevelopedwhichwillbeissuedforcommentby January1992.The updatedsourceterminsightswillbe made availablefor

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voluntaryusebyexistinglicensees,ThemisSpeisstatedthatreportsoncompositionandmagnitudeandiodinechemicalformwillsoonbeissued(nextcoupleofweeks)asdraftsforcomment,Some implicationsoftheupdatedTID are:

• considerationofadditionalnuclides(e.g,,cesium),

. effectson fissionproductcleanupsystems,and

. containmentleakrateandcontrolroom habitability,

o The final revisionof Part 50 will incorporate updatedsource term and severeaccident insights. This rule change is expected to be issued for comment bySeptember1992. Upon the Commission'sinquiry as to how eheStaff will applyPart SOregulationsto advancedreactors,Thomas Mur!ey expresseduncertainty;however,he indicatedthat Part 50 mayhavetwo sections,The Staff doesnot wantcurrent reactors to be able to take credit for regulationsapplyingto advancedreactors, CommissionerRemick inquiredas to whereevolutionaryreactorsfit inPart 50? A..shokThadani indicatedthat evolutionaryreactorsare somewhereinbetween, In responseto CommissionerCurtiss'inquiries,LeonardSofter indicatedthatSeptember1993will be thescheduh;ddatefor final changesto be incorporatedinto Part 50 for advancedlight water reactors(A.[..WR) and that the Commissionwill see both EPRI Requirements Documents for ALWRs prior to Part 50finalization, The Staff also mentionedthat SECY 90.016, "Evolutionary LWRCertificationIssuesand Their Relationshipto Current RegulatoryRequirements,"will be quantifiedin the final Part 50,

FissionProductTiming:

Accordingto Seller, theStaff is investigatingtwo aspectsof the timingof fissionproductreleases:gap activityreleasefromearly fuel rod failures,investigatedby INEL usingR,ELA.P/SCDAPcodesfor a rangeof LOCA events,and fuel activityreleasefrom fuel melting_ Fissionproduct timing,magnitude,and composition from fuel melting are being investigatedby using NUREG.1150accidentsequences. Fissionproducttiming preliminaryresultsinclude the following:

o Containment isolationvalveclosuretimerequirementsarecontrolledbygapactivityreleases,_ bylarge releasesassociatedwith core.melt.

o Gap activity releasesoccurquickly(secondsto minutes), Seller defined the timeto releaseas the intervalbetweentheonsetof theaccidentand the first fuel failure.He statedthat the releaseat one plantwasestimatedto take 25 seconds,whichisthequickestso far. CommissionerCurtissaskedthe Staff what do we know todayasa resultof thisnew information,Seller repliedthat theprincipalfindingwasthat

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gap activity releases take longer than initially thought (5-10 seconds). Due to thislatest information, Commissioner Curtiss recommended that the Staff encourageutilities to relax valve closure times, certainly for BWRs. Murley stated that theStaff is reluctant to do this yet as it has not thoroughly looked at the information;however, the Staff is not discouraging utilities from coming to the Staff andpresenting their analyses.

o Large releases associated with core-melt do not occur earlier than about 20 minutesafter initiation of accident and may be significantly longer.

o Releases include different species and forms of fission products than assumed inTID-14844.

Brookhaven National Laboratory (BNL) examined 36 severe accident sequences which wereanalyzed by the NRC's source term code package (STCP) and used in NUREG-1150. The draftreport is presently being reviewed. The preliminary results are as follows:

o Large fission product releases associated with core melt appear in containment nosooner than twenty minutes after the accident initiation, in agreement with the Staff'sfindings.

o Vessel failure typically occurs no sooner than about 40 minutes and 90 minutes forPWR and BWR sequences, respectively. During that time, fission products are beingreleased into containment.

The duration of the releases after vessel failure (if this occurs) is about 2 hours _nd 3 hours forPWR and BWR sequences, respectively. Composition and magnitude of typical (neither worst orleast accident) releases into containment before vessel failure:

Noble Gases 90%Iodine 20-30%Cesium 15-20%Tellurium 5-10%Barium, Strontium 2-3%Others, < 1%

Ex-vessel releases (if these occur and are not cooled and/or scrubbed) are predicted to add mostof the remaining iodine and cesium, substantial tellurium plus smaller quantities of the non-volatilesas well.

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Iodine Chemical Form:

Iodine chemical forms were also discussed. Oak Ridge National Laboratories (ORNL) examinediodine fission product chemistry. The study was based on quantitative results of seven severeaccident scenarios used in NUREG-1150. Both high and low pressure sequences were chosen forthe principal plant types; a single sequence was considered for the PWR ice condenser plant. Eachsequence was evaluated using the STCP. The study considered the chemical kinetics of 20 reactionsof iodine with water, hydrogen, and cesium. Once a chemical equilibrium was established, analysiswas done to determine the iodine compounds and forms present. The results are as follows:

o In most calculations, iodine was released from the reactor coolant system (RCS) intothe containment as cesium iodide (CsI) wi_h very small amounts of I or HI. ORNLresults indicate that the iodine enterini_ containment is at least 95% Csl, with nomore than 5 percent I and HI.

o Iodine entering containment dissolves in water pools or plates out on wet surfacesas I.

o Iodine behavior within containment depends upon time and pH of the watersolutions.

If pH is maintained at a value of seven or greater, then the amount of iodinein solution which converts to elemental and organic iodine later in theaccident sequence will be very low.

- If pH is not controlled, radiation levels in water pools are sufficient to reducepH and significant amounts of dissolved iodine will be re-evolved as elementaliodine.

Thadani stressed the importance of long-term pH control.

EPRI ALWR Proposed Source Term (Evolutionary_Designs):

Currently the Staff is reviewing EPRI's submittal to their ALWR proposed source term forevolutionary designs. The Staff noted that EPRI assumed that accident management was successful.Based upon recoverable in-vessel melt accident, EPRI proposed:

Release magnitude into containmentNoble gases 80%Iodine 22.5%Cesium 22.5%Tellurium 12%Ba, Sr, Ru 0.3%Remainder 0.003%

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The release is at a uniform rate for over a 30.minute period beginning at 60 minutes after theinitiating event. Results include chemical forms in containment:

Iodine chemical form

- 97% particulate- 2.85% elemental- 0.15% organic

Other chemical forms: 100% particulate

The Staff concluded its presentation by noting that it needs to look at accident management. Otherongoing considerations are as follows:

o Should one source term formulation be applied to all future light water reactordesigns or should different source terms be designated for major plant Froups, suchas PWRs and BWRs?

o Should different source terms be selected for differing applications (e.g., equipmentqualification vs. control room habitability)?

o Relatively low amounts of elemental and organic iodine expected within containment(assuming that pH is controlled) calls into question the need for high efficiencycharcoal adsorbers.

o What accidents should be considered? For example:

- To what extent should ex-vessel progression and reactions be considered?

- What credit should be given for plant design features (e.g., core debriscoolability)?

The Staff is going to concentrate on the first and last of these considerations. Chairman Carr• inquired of the last consideration's resolution date. The Staff indicated that it will be resolved by

1991 and issued for comment by January 1992.

A 24-page handout on the Staff's presentation is available through the SERCH Staff.

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SERVICE FOR EVALUATING REGULATORY CHANGESf

MEETING: NRC Staff Meeting With NUMARC on the Revision of 10 CFR Part 100,Appendix A

DATE: April 17, 1991

NUMBER: MS-91-56 KEY SUBJECT: SitingSeismic Design


Andrew Murphy, Chief, Structural & Seismic Engineering Branch (SSEB), Office of NuclearRegulatory Research (RES)

Nilesh Chokshi, SSEBRoger Keneally, SSEBGoutam Bagchi, Chief, Structural & Geosciences Branch (ESGB), Office of Nuclear Reactor

Regulation (NRR)Robert Rothman, ESGB, NRRJohn Chen, Severe Accident Issues Branch, RES

Industry Members Present Included:

Ray Ng, NUMARCOrhan Gurbuz, NUMARC

Carl Stepp, EPRIMike Hayner, Cleveland Electric Illuminating CompanyBert Swan, Geomatrix ConsultantsJohn Reed, Jack Benjamin and Associates, Inc.

On March 6, 1991, the NRC held a meeting to discuss its views on how Part 100, Appendix Ashould be revised (see SERCH MS-91-31). During this meeting, the NRC heard the industry'spreliminary views on the matter. According to NUMARC, formal comments will be submitted in3-4 weeks which will include comments on how Appendix A should be revised, what the industryenvisions for potential changes to 10 CFR Part 50, detailed outlines on the contents of two keyRGs, and a position paper on how OBE (operating basis earthquake) should be treated.

NUMARC Recommendations:

Gurbuz opened the industry presentations by noting that the NUMARC Seismic Issues WorkingGroup favors adopting Option C, "Simplify Appendix A and Issue Supplementary Regulatory

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Guides," which it had developed and issued three years ago. He added that NUM_RC'sunderstandingis that the NRC is already follo_ng this thini_ng. NU_P_RC recommends:

o that Appendix A be simplified to include only what needs to be done to establish siteseismic requirements, with the details of how these things are to be done in regulatoryguides.

o that the design-related requirements (including plant shutdown and restart criteria) be placedin 10 CFR 50.

o that NRC Staff endorse EPRI reports NP-5930, "A Criterion of Determining Exceedanceof the OBE," and NP-6695, "Guidelines for Nuclear Plant Response to an Earthquake."Ray Ng added that NUMARC hopes the Staff will endorse these reports, possibly in ageneric letter format, for use by current plants not just future plants. Goutam Bagchiindicated agreement with NUMARC's comments on this matter and noted that before theend of the calendar year, the NRC should be able to issue something on this.

o the elimination of the OBE as a design requirement.

o that the Appendix A language consist of non.prescriptive considerations which guide theNRC in its evaluation of a site, minimize ambiguous legal interpretations, avoid codifyingexpert opinion, and avoid codifying the state of the art in the rule.

The regulatory guides (RGs) that NUMARC thinks will be needed to support the proposed revisionto Appendix A are listed in Attachment 1. NUMARC acknowledged that it will be difficult togain consensus on the first two geosciences RG issues. It has developed detailed outlinespositionson these two RCrs; they will be submitted to the NRC in 3.-4 weeks with the other NUMARCcomments on the Appendix A revision.

Appendix A Seismic and Geologic Siting Criteria:

Carl Stepp pointed out that the keys to the simplification of Appendix A are to remove thetechnical statements that require interpretation (such as the term tectonic province), address onlysiting and seismic design basis determination requirements, move all seismic design requirements to10 CFR Pan 50, allow the use of probabilisticseismic hazard methods, structure the regulation tominimize requirements for unnecessary information, and remove the OBE as a design requirement.

He recommended that Appendix A should adopt the following definitions:

o Site Locality -- A radius less than 1 krn.

o Site Area -- A radius less than 8 kin.

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nnn::o Site Vicinity -- A radius less than 25 kin.

o _ -- A radius less than 150 kin. Murphy was uneasy with the use of specificnumbers (for this and the other site definitions listed above) as cutoffs for performingevaluations. The Staff also wanted to know where the radius begins; Carl Stepp respondedthat this has not been established yet, adding that the philosophy is what is important at thispoint.

o _Des_n Basis Ground Motion (DBGM) -- The free-field vibratory ground motion determinedat the site ground surface, considering the seismic sources in the site region and specificgeotechnical characteristics of the site subsurface materials. It is that ground motion forwhich there is reasonable assurance of a low likelihood of being exceeded.

Murphy questioned the use of the term "low likelihood," noting that it is vague and likelyto raise interpretation problems. Stepp responded that accozding to the EPRI technical andlegal counsel, the term is workable; however, the industry would be willing to consider othersuggestions such as possibly "low probability." Rothman also thought that the term is vagueand could result in litigation on how it should be interpreted. In his view, this is one ofthose terms that the Staff wants to stay away from.

With respect to the d(_jterminationof site ge0technical design basi.spar.amete_, Stepp pointed outthat the purpose of this determination is to develop the design basis parameters for the foundationand the foundation-structure system and to formulate the bases for establishing procedures tomitigate adverse impacts on construction and operation of a plant.

The scope of these determinations would include determining the foundation excavation and designbasis parameters, liquefaction and ground settlement under DBGM loading, slope stability underDBGM loading, subsurface cavities and joints, ground water conditions, slope stability andsettlement under static loading conditions, and the geometric distribution of foundation materialsand their dynamic material properties.

Regarding the determinatio.o of tectonic surface deformation., the purpose of this effort would beto determine if there is a need for and, if so, to develop the design basis parameters for the effectsof tectonic surface deformation. Bagchi asked whether Stepp is committing the industry todeveloping a guideline for surface deformation. Stepp said no and added that such guidance, whendeveloped, should be placed in a RG .- not codified in the regulation. Murphy agreed with thisphilosophy.

Stepp pointed out that the scope of this determination should include: 1) an assessment of thetectonic framework of the site region, 2) an assessment of the relationship between seismicity andgeologic structures within the site vicinity, and 3) the identification and assessment of the geologicstructures within the site area that have a potential for tectonic deformation that could adverselyaffect the design basis of a plant. The industry would like the NRC to allow the use of

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probabilistic approaches for this determination to assess, for example, the likelihood of adeformation occurring given that a structure capable of deforming is present.

For the determinatio_n of thg..design basis _round motion, the industry thinks that this effort shouldinclude an evaluation of: 1) the seismic sources within the site region (Stepp advocated notrequiring plants to do evaluations over large regions -- he thinks this evaluation should be limitedto a 150 km radius); 2) the earthquake potential of each seismic source identified in the site region;3) vibratory ground motion transmission in the site region; 4) the site area geology and site dynamicgeotechnical parameters; and 5) the vibratory ground motion at the site from all seismic sources inthe site region. The industry wants the NRC to permit the use of probabilistic approaches fordetermining the level of conservatism that should be applied to the selection/determination of theDBGM for a given site. Nilesh Chokshi asked whether plants to be sited in low seismic hazardregions should be required to go through such a detailed evaluation, Rothman seemed to thinkso. Stepp commented he had not thought about this but that it is worth considering how theDBGM evaluation for such plants could be reduced.

With respect to the de!ermination of design .requirements for other s,eismic and geoJogic hazards,Stepp noted that the purpose of such guidance would be to determine the need for and establishdesign basis parameters for other seismic and geologic hazards. The scope of the determinationwould include seismically.induced floods and water waves, volcanic hazards, man-induced seismic orgeologic hazards, and subsidence or collapse. Stepp noted that the approach for dealing with theseareas is pretty well established; therefore, EPRI does not expect to spend much time on them.

10 CFR Part 50:

Mike Hayner identified the Appendix A provisions the industry thinks should be moved to10 CFR 50. He pointed out that Part 50 should include only the provisions that address whatneeds to be done to achieve the primary objective, i.e,, the protection of public health and safety.The siting requirements are to remain in Appendix A while the design requirements should beplaced in Part 50, The industry wants to see the following provisions reflected in Part 50:

o The designbasisgroundmotion.-whichhe notedwould replacethesafeshutdownearthquake(SSE)terminology.

Chokshiaskedwhetherthischangeinterminologyisworththeeffort,notingthatcurrentplantsarelicensedtotheSSE and by revisingtheterminology,ineffect,we wouldbecreatingtwosetsofcriteria..one forcurrentplantsandone forfutureplants.Bagchiemphasizedthatitwould involvea majorefforttochangetheterminologyacrossalldocumentsforconsistency.Hayneragreedbutaddedthattheindustrythinksthatchangingtheterminologyisworththeeffort.

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o Use of a single earthquake level (the DBGM) for design as opposed to the current dualearthquake criteria (i.e., the OBE and SSE).

o Permissibility to design beyond yield strain provided that the safety functions are maintained.

o The requirement that seismic analysis must account for soil.structure interaction (SSI)effects.

Industry Position on OBE:

Regarding the need for the OBE earthquake as part of the design basis, John Reed pointed outthat seismic PRA indicates that the median plant capacity of eastern U.S. plants is 2 to 4 times theSSE. The other thing learned from PRA is that there is no proverbial failure "cliff"just above theSSE level. Most failure cases have been found to be outside the design process (e.g., anchoragedetails and interaction issues). From Reed's perspective it is not clear what safety purpose theOBE serves in the design process. Any additional margin provided by the OBE is unintentionaland varies component to component and location to location. In his view, :he DBGM providessufficient margin to satisfy safety concerns; fatigue testing requirements could be based on DBGMinput; and attention to detail is more important than designing for the OBE.

Bagchi agreed with Reed's argument on the preference of having one design basis earthquake. InBagchi's view, the cost of using the OBE is greater than the gain. However, in reference to Reed'sstatement that PRA indicates a plant capacity 2 to 4 times the SSE, Bagchi pointed out that if theOBE is omitted, this may not be true any more. He also noted that changes in the capacity ofequipment supports as a result of leak-before-break considerations would also reduce the 2 to 4times capacity Reed referred to. Reed responded that he has found that the OBE does not controlthe results of a PRA.

Chokshi and Bagchi seemed to be concerned with the impact omission of the OBE will have onequipment currently designed for OBE. Bagchi thinks that eliminating the OBE may require thedevelopment of new load combinations. He indicated that going beyond yield for passivecomponents would not be a problem but for active components (pumps, valve',) there may be aproblem. Reed and Hayner stated that even though equipment may be allowed to go beyond yield,it would still be required to function at the SSE level. Bagchi was not convin_-ed that allowingequipment to go beyond yield would provide assurance that it will remain funr;tional.

Reed concluded his presentation by outlining the following recommendations for revising theregulations:

o The OBE should be removed as a design requirement.

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o "['he regulations should be written to allow, for the purposes of plant shutdown, licenseesto: 1) determine an equivalent OBE basedon DBGM requirements(for obtaining aspectrallimit), or 2) designdirectlyfor currentOBE requirements.

o NRC shouldpermit the useof the EPRI plantresponseguidelinesand OBE exceedancecriterionin determiningresponseto an earthquake.

Closing Comments:

Murphy offered the followingcomments:

o Regardingthe new SSE definition(i.e., the DBGM), the Staff doesnot havea problemapplyingthisto newplants. However,Murphyindicatedthat he is not inclinedto supportapplyingthe newdefinition to currentplantsalthoughhe didacknowledgethat there aresomebenefitsto changingthe SSE terminology.

,-, Regardingthe levelof detailandcontent of AppendixA, Pan 50, and the RCrs,the Staffandindustryare thinkingalike.

o Regardingthe industry'sconcernon whetherall the RGswill be _ued for publiccomment,henotedthat it is NRC policyto do so. He addedthatNRC iscurrentlyscheduledto issuefor commentthe draft RC,-ssupportingthe AppendixA revisionin January1992.

o Regardingwhenthe applicablecod= andstandardswill be revisedforconsistencywith theAppendixA revision,the NRC doesnot havecontrolof theseschedules.The draft rulesandRGs will be forwardedto the appropriatecommitteeswhenthey are _ued for publiccomment.

o Regardingthe eliminationof the OBE as a designrequirement,the NRC is very muchinterestedin maintainingthe samemarginsfor future plantsas are availablefor currentplants;therefore,the Staff is inclinedto leavethe OBE requirementalone.

o Regardingthe RGs listedin Attachment1, he noted that the first three RGs under thegeosciencessectionare the mostimportantin the Staff'sview. He doesnot know at thispoint whether therewill be three separateRGs or one on thesetopicsbut the NRC doesintend to issueguidanceon them. Regardingthe next two RG topics, on volcanicandtsunamihazards,he indicatedthat they affectonlya smallpercentof the plantpopulation;therefore, he doesnot know whether guidanceis really neededfor these topics. Withrespectto the fourdesignRGs listed inAttachment1, Murphystatedthat guidancewill beWovidedin theseareas.

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Bagchi commented that he is in agreement philosophically with what NUMARC presented at thismeeting. He urged NUMARC to submit, for NRC review, the detailed outlines/positions it hasdeveloped on the first two geosciences RGs listed in Attachment 1 as soon as |mssible.

In response to Ray Ng's question as to whether NUMARC's current schedule for submittingcomments in 3-4 weeks would be too late, Murphy noted that it should be all right although theStaff would prefer to have them in before May I -- RES's deadline for having the first Appendix Adraft available for internal NRC review. He added that "the concrete is being poured as we speak."He therefore encouraged submitting any available information as soon as possible.

The following handouts are available upon request from the SERCH Staff:

a) @page overview vugraph

b) 15-page seismic & geologic siting criteria vugraph

c) 5.page seismic design aspects vugraph

d) 5.page RC-s overview vugraph

e) 10-page OBE position vugraph

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...... Ill Ill, ill IIIIII I I IIIIII I IIIIIIIII

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_' MS-91-56

Attachment !

Page l of 1t

NEWREGULATORYGUXD_

GEO$CIENCES-

1. DETERMINATIONOF DESIGNBASISGROUNDMOTION

2. DETERMINATIONOF POTENTIALFORTECTONICSURFACEDEFORMATION

3. DETERMINATIONOF SITE GEOTECHNICALDESIGNPARAMETERS

4. ASSESSMENTOF POTENTIALFORVOLCANICHAZARDSi

5. ASSESSMENTOF TSUNAMIANDWATERWAVEHAZARDS

DESIGN:

1. ACCEPTABLEHETHODSFORSEISMICDESIGN

Z. DESIGNFORSEISMICALLYINDUCEDFLOODSANDWATERWAVES

3. SOIL-STRUCTUREINTERACTIONANALYSIS

4. PLANTSHUTDOWNANDRESTARTCRITERIAi

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(___ SERVICE FOR EVALUATING REGULATORY CHANGES

MEETING: NRC Staff Meeting with Industry Representatives to Propose a CooperativeApproach Toward Early Siting

DATE: June 21, 1991

NUIvh3ER: MS-91- 85 KEY SUBJECT: Siting

NRC Staff Members ParticioatinR

James Partlow, Associate Director, NRRBarry Zalcman, Program Manager for Early Siting, NRR

Utility Representatives Participating

Tom Tipton, NUMARCLou Long, Southern Nuclear Operating CompanyMichael Wallace, Commonwealth Edison, Joint Contractors Organization, ChairmanJoe Santucci, EPRI

Representatives from the nuclear industry briefed the Staff on their proposed approach to obtaininga nuclear plant site permit through cooperation and interface with the Staff during the applicationprocess. The Staff accepted the proposal with some reservations and both sides agreed that thiswas a good first step.

Industry_Presentation

In April of 1990, the Department of Energy (DOE) issued a Request for Quotation (RFQ) for anearly site permit (ESP) to the nuclear power industry. The RFQ was a 50-50 cost-sharedemonstration or successful use of 10 CFR 52 to obtain an ESP. The DOE offered to fund

one-half of the $26 million project to attract a pilot site. Individual utilities were hesitant toparticipate due to an unpredictable climate, so the industry made a qualified response. Aconsortium of participants will divide up the responsibilities among themselves and reach the samegoal as called for by the RFQ: an issued ESP.

Tipton presented a brief background on the "Strategic Plan for Building New Nuclear PowerPlants," and outlined the 14 parts, termed "building blocks," that must be completed to achieve anew power plant in the United States. This meeting addressed the industry's approach towardscompletion of two: Building Block 5 -- Siting, and Building Block 2 -- predictable licensing andstable regulation.

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Key industry groups include EPRI which will be co.responsible with NUMARC for BuildingBlock 5 and will provide the technical research and development that may be required. NUMARChas sole responsibility for Building Block 2 and will provide the licensing function and theregulatory interface management.

Long described the Industry Siting Group (ISG) as being made up of a cross section of 12 utilities,two architect engineers and a pool of licensing and technical expertise. The ISG will provide:

o coordination of industry resources to accomplish siting;o guidance and direction for specific tasks within the program;o review and endorsement of program deliverables; ando interfaces with the Utility Steering Committee and the Standardization Oversight

Working Group to assure Advanced Light Water Reactor Design Certification, Firstof a Kind Engineering, and the Utility Requirements Documents remain consistentwith siting initiatives.

Wallace then explained how the three previously mentioned industry groups will all work to supportthe Joint Contractors Organization (JCO), which is a group of three utility-owned organizations(Southern Electric International (SEI), Commonwealth Research Corporation (CRC), and PublicService Electric and Gas (PSE&G)), that have come together as a service to the industry torespond to the DOE RFQ on siting. The JCO submitted a qualified response to the DOE'scontractor, Sandia National Laboratory, on April 12, 1991. The response calls for a 3-stepapproach which deviates from that suggested in the RFQ. It specifies that no site will be selecteduntil all criteria and processes are reviewed and discussed with the NRC. The three steps are asfollows:

STEP 1

o identify all siting related regulatory documentso identify submittal requirementso identify studies, tests, analysis requiredo produce a document that can be used by applicant and regulatorso resolve issues, develop methods

STEP 2

, share information; workshop formato identify site(s) to license

STEP 3

o license site(s)

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The final result of this approach is the same as that desired by the DOE RFQ; a licensed site.

The JCO has designated a program management structure to carry out the program's day-to-dayactivities. This structure has EPRI and NUMARC functioning as described previously and reportingtheir results .*oa Program Management Council made up of CRC, SEI, PSE&G, EPRI, andNUMARC. SEi has been tasked with contracts administration while the ISG will act as an advisorybody to the council. The council will be the sole point of contact with Sandia/DOE for ESP issues.

Santu_i described contract negotiations with the DOE as nearing completion and expected to befinalized by July 1, 1991. This will allow the resolution of issues and document development ofStep 1 to continue with a completed document scheduled for late 1992. Site identification, Step 2,should be progressing concurrently with a site o: sites chosen by early 1993. The program envisionssubmittal preparation to take place in 1993 and license approval, Step 3, to be granted before theend of 1994.

Wallace summarized by saying that this program was developed because no single utility wantedto be the "lightning rod" of the ESP process. The lack of siting activity over the last two decadeshas seen many changes in requirements as well as technological advances. This program's objectiveis to "achieve, through a joint effort, the goal of technically valid predictable licensing and stableregulation."

Staff Resoonse

Barry Zalcman expressed concern over the issue of two programs, the one described by the industryand the Staffs, being performed in parallel with no sharing of information possible. Zalcman wasalso less than optimistic towards the eventual meshing of the two paths to a common endpoint.The incongruity he envisioned would be caused by the required public hearings and public commentportion of the ESP process which he felt was not adequately included in the JCO timeline.

Partlow was more optimistic towards the possibilities, considering how early in the program thisdiscussion was taking place. He thought the Staff could offer assistance in defining regulations, butmust not lose sight of its role as regulators. Partlow recommended continued dialogue and said thatat this stage "this is the best we can do."

A 14.page viewgraph is available upon request from the SERCH Staff.

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SI::RVIC:E,FORI::VAL.UA'T'INGRI:::GUI_A'I"ORYCHANGI:::S

MEETING:. ACRS Joint Subcommittees on Safety Philosophy,Technology and Criteria;SevereAccidents; and Regulatory Policies and Practices Meeting on Part 100 Changes, theUpdated Source Term, and the Safety Goal Large Reiea._ Definition

DATE: January 743, 1992 APPLICABILITY: ALWR/BWRJPWR

NUMBER: MS-92-04 KEY SUBJECT: SitingSource Terms

Reliability

ACRS Subcommittee Members pr_ent:

David Ward, Chairman, Safety Philosophy, Technology and Criteria SubcommitteeWilliam Kerr, Chairman, Severe Accidents SubcommitteeHarold Lewis, Chairman, Regulatory Policies and Practices SubcommitteeCharles WylieCarlyle MichelsonChester SiessPaul ShewmonThomas KressPeter Davis, ACRS Consultant


Thomas King, Deputy Director, Division of Safety Issue Resolution (DSIR), Office of NuclearRegulatory Research (RE,S)

Charles Ader, Chief, Severe Accident Issues Branch (SAIB), DSIR, RESLen Softer, SAIBJohn Ridgely, SAIB

NRC Consultants:

Ed Beahm, ORNLHossein Nourbakhsh, BNL

In this meeting, the Staff and its contractors presented an overview of 1) the proposed(non-seismic) changes to 10 CFR Part 100, "Reactor Site Criteria," and associated Part 50 revisions,2) work on the updated source term which in approximately 3 years will be incorporated into10 CFR Pan 50, and 3) the status on the Staff's attempts at developing a large release definitionas requested by the Commission during the development of the Safety Goal Policy Statement. The

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NRC Staff is requesting ACRS written comments/endorsementon items 1 and 2. Item 3 is separatefrom the first two and was presented at this meeting for information only -- a formal ACRSresponse was not sought at this time.

Over_.ew:

The purposeoftheproposedPart100revisionistodecoupleplantdesignfromsiting.TherequirementsfordosecalculationscurrentlyinPartI00arebeingmovedtoPart50.ThisPart50changewillbereferredtoasinterimbecauseitwillreferencethecurrentsourceterm(TID-14844)criteria.Eventually,when theworkbeingdoneinsupportofdevelopinganupdatedsourcetermisfinished,Part50willbe revisedagain(referredtoasthefinalrevision)toreferencethenewsourceterm. The updatedsourceterm isintendedto reflectcurrentunderstandingofin-containmentf'_ionproductbehavior(intermsofquantity,form,andtiming)ofa releaseduringandaftercoremeltaccidents.

The revisedPart100willapplyto aH futurereactorsand thefinalPart50 revisionwillbeapplicabletoallfutureLWRs (evolutionaryandpassive).Applicantsforanew LWR betweennowandwhen thefinalPart50 goesintoeffect(e.g.,evolutionaryplants)willbe allowedtousetheTID o_thenew sourceterm;however,useofthenewsourcetermwillrequirepriordiscussionwiththeNRC Stafftoensureagreementoverthespecifics.Currentplantswillbeallowedtou_ethenew sourcetermon a voluntarybasis(i.e.,itwillnotbe backfittedbytheNRC).

As tothelargereleasedefinition,theStaffisexploringthesensitivityofthemagnitudetovariousparametersusinga setofrepresentativesitecharacteristicsderivedfromtheproposedPart100revision.Sucha releasemay beexpressedinIodine-131equivalence.

ProposedPart100and InterimPart50Chan2es:

Background:

The present Part 100 regulation, which was established in April 1962, requires that every reactorhave the following:

An exclusion area -. a zone immediately around the reactor which has no residents. Theradiusof this area is determined by dose calculations; therefore, it is not fixed. For existingplants, it varies from 0.2 miles to over 1 mile and is typically 0.5 miles.

A low population zone (LPZ) - a zone outside the exclusion area (typically 2-3 miles)which may contain residents but not a densely populated center. Protective measures mustbe feasible for the residents within the LPZ.

A _:mulation center distance -- the distance to the nearest densely populated center canbe no closer than one and one-third times the LPZ radius.

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The current regulation requires that for an instantaneous fission product release inside containment,the doses to hypothetical individuals at the exclusion area boundaryand at the LPZ outer radiusmust meet the values stated in Part 100 (25 rein whole body and 300 rein thyroid). Part 100 is veryflexible and has no numeric criteria for calculatingol_ite doses. It simply refers to TID-14844 asproviding an example calculation. However, the TID caiculational methodology has been madeobsolete by Regulatory Guides (RGs) 1.3 and 1.4.

Other siting criteria that were introduced after the issuance of Part 100 include: 1) thedevelopment of informal criteria in the 1960s -- subsequently included in RG 4.7 -- preventingplants from being built in densely populated areas',and 2) the establishment, shortly after the TMIaccident in 1979, of the 10-mile emergency planning zone (EPZ) in 10 CFR 50.47 and therequirement that emergency plans account for severe accident scenarios within this zone. (Withthe establishment of the 10-mile EPZ, the emergency planning aspects of the LPZ weresuperseded.)

In 1980, the Staff published an advancednotice of proposed rulemakingto revise the siting criteria;however, the Commission stopped this work a year later so that the safety goals could be definedfirst and used to assess the adequacy of the proposed Part 100 revision. In 1990, as a result ofincreased interest in proposed new designs, the Staff recommended to the Commission (inSECY-90-341) that plant design be decoupled from siting. In a January 25, 1991 StaffRequirements Memorandum (SRM), the Commission directed the Staff to proceed with suchrulemaking.

Elements of the ProposedRule:

The proposed Pan 100 consists of two sub-parts: Sub-part A is identical to the current rule andwill continue to be applicable to existing plants. Sub-part B will provide the new' rule thatdecouples siting from plant design, and will be applicable to future plants. Part 100, sub-part Bincludes the following key changes:

o The source terms and dose guidelines are deleted. This information is being moved to10 CFR 50.34 for use in designing the containment and other plant features -- it will nolonger be used for siting purposes.

o A 0.4-mile minimum exclusion area (from the reactor centerline) is established. However,this has been questioned by sotne members of the NRC Staff; therefore, NRC will beasking the public to comment on whether this value should be reduced for reactors smallerthan 3800 MWt.

¢) The LPZ is being deleted. (In the Staff's view, the intent of the LFZ will now be met bythe population density criteria noted in the following bullet.)

o The projectedpopulationdensityatinitialsiteapprovalcannotexceed500 personspersquaremileoutto30miles,and40yearsaftersiteapproval,itcannotexceed1000persons

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per square mile. (This requirement is being transferred from RG 4.7.) In response toACRS questions, Softer indicated that if an applicant decides to start construction 15-20years after site approval wasreceived, the Staff would probablycheck the population densityagain. Peter Davis wanted to know what the Staff would do if the 1000 persons per squaremile criterionwas exceeded duringthe life of the plant. Softer stated that the plant wouldnot be shut down and Charles Ader added that this question will have to be addressed atthat time. Softer remindedeveryone that this criterion is for siting purposes only andfor making decisions on the adequacy of operating plants.

o Physical characteristics that could pose a significant impediment to the development ofemergency plans are to be identified.

o No meteorology evaluation will need to be done for site suitability.

o A new section will require the evaluation of man.related hazards and their incorporationinto the plant's design basis when they are found to pose a significant risk. (The Staffnoted that these requirements are pretty much the same as current practice -- the differenceis that they are being codified.)

o A new section will require periodic reporting of changes in population and significantchanges in man.related activities :n the site's vicinity. Population updates are to besubmitted every 10 years and man-related updates every 5 years.

The Staff is scheduled to take the proposed Part 100 and interim Part 50 revisions to CRGR inFebruary. The Commission will be briefed in March and the proposed rules will be issued forpublic comment in April. In the Staff's view, the proposed rule package does not represent asignificant change in the present siting practice; rather, the package codifies it.

The Staff added that it has reviewed the proposed siting requirements relative to the safety goal'squantitative health objectives (QHOs) and land contamination aspects. In response to questionsfrom Kerr and Siess as to the appropriateness of considering land contamination, King explainedthat the reason the Staff looked at land contamination was to see whether a large area of land(especially in potentially large urban areas) could be rendered unusable by an accidental release.That is, it wanted a feel for whether the 30-mile area for population density control is adequate.Based on NUREG-1150 sequences that have a 104 per reactor year or greater chance of occurringand other plant information, but using a reactor size of 3800 Mwt, the Staff found that there is avery low likelihood (about 10" per year) of having to condemn land beyond approximately 20 miles.As to the safety goal checL the Staff found that the prompt fatality QHO is met for a 0.4 mileexclusion area boundary and that sequences capable of leading to early fatality must be less likelythan about 3xlo s per reactor year. For the NUREG-1150 plants (i.e., using the existing sitingcriteria and reactor sizes), such sequences are less than 10"sper reactor year. The latent fatalityQHO is readily met. Based on these evaluations, the Staff concluded that it is reasonable to lookat population out to 30 miles for reactor siting.

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Commen_/__ns:

Siess pointed out that the Staff's focus in the proposed Pnrt 100 revision has been maddress/accountfor radiationdose implicationson siting decisions. However,he noted thereareother factorsthat impactsitingsuchas NationalEnvironmentalPolicyAct (NEPA)issues(10 CF'RPart51). Sofferrespondedthat the Staffhas not looked at decouplingsitingwith respect to theenvironmentalimpactstatement;however,Kingaddedthatthe Staff is thinkingof revisingPart51.Siess noted thathe would hate to pursueall the Part100sitingrequirementsand not be able toproceedwithsitingapprovaldue to NEPAconcerns. Softercommentedthat the Staff is tryingtoaddressthe nuclearsafetyaspectsof sitingwiththe proposedrevision. ButSiesspointed out thatresolvingthe safety issues may still be fruitlessbecause the site could still be rejected fornon.nuclear(e.g., NEPA) reasons.

In response to Ward'squestionon whetherthere is a risk.basedtechnicalbasis for the variouszones thathavebeen selected,Suffersaidthatearlyfatalitywouldnotbe expectedto occurbeyond10 miles and, with respect to land contamination,condemnationwould not occur beyondapproximately20-25 miles. Wardnoted that the siting criteria should reflect risk evaluationsapplicableto futureplantsandthe insightsfrompastsitingexperience. The Staffrespondedthatit hasnot performedexplicitriskanalysesfor futureplants;however,futureplantsareassumedtobe at least assafeas currentplants. The riskof currentplants,as reflectedbyNUREG-1150,meetthe earlyfatalityOHO byat leastan orderof magnitude,and the late fatalitycriterionbyseveralorders of magnitude. Kerr sarcasticallycommented that this statement, that NUREO-1150representsthe populationof U.S. plants,contradictsstatementsmade by other membersof theNRC Staff (who participatedin NUREO-1150)who havesaid that one could not drawgenericconclusionsfrom1150. KingandAderreiteratedthat 1150is viewedby the NRCas reflectingthepopulationof currentplants in termsof level of safety.

Siesswantedto knowhow populationdifferencesfromsite to site affectrisk. Kingstated thattherevisedPart100populationnumberswereappliedto the NUREO-1150plantsto see whethertheywouldmeet the safetygoal QHOs. All plantsmet them;however,the Staffhasnot calculatedhowmuchthe actualnumberof fatalitieswould increasefor a site with a higher populationdensitybecause the safety goalsonly addressriskto an individual,not to a group.

In responseto Siess'squestionon whetherthereis a limiton the numberof unitsthatcan be builton a site, Softerrepliedthereis no suchlimit;however,the Staff'ssitingcalculationsareonlybased

•on 3800 MWt per site. Staffcalculationsfor effects of landcontaminationwere done with thecomputercode MAACSwhichcalculatespromptfatalitiesaveragedover a population.

Wardand Lewis commenteOthat although the Staff has considered land contaminationandpopulationdensity for the proposedrule, the safety goal policystatementdoes not addresslandcontaminationor populationdensity. In theirview, this is somethingthe Commissionerswillhaveto addressone day. lOess summarizedthe thrustof the newsitingcriteriaby sayingthat plantsshouldnot be sited in highpopulationareaseven thoughthey meet the safetygoals. The Staff

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agreed with the characterization. Davis interjected that population density _ come into play inthe safety goal policy statement when doing evacuation modeling .- the higher the populationdensity, the longer it will take to evacuate.

Siess's concluding remarks were that he thinks that the siting "issue"will not be primarilyone ofpopulation density but of NEPA concerns/aspects. He commented that the land contaminationissue is not closed and suggested that ACRS look into this iuue further and provide its views tothe Comm_ion; Ward agreed.

Kerr thought the language in the proposed Part 100 is too general and that applicants will needa RG(s) for details. As an example, he referred to a statement that "the Staff will consider PRAinformation,*however,what informationand how the NRC will use this informationis not specifiedanywhere. According to Kerr, such statements need an exegesis. King commented that the Staffis proposing to revise RG 4.7, however, the revision has not been developed yet. There are noplans to revise or to originate any other RCrs.

_urc_. Term U_ate:(ref: MS-91-125, MS-91-112, MS-91.93, MS-91-43)

The source term is the release of f't_ion products _ the containment which are potentiallyavailable for release into the environment. It includes the timing, form, and quantity of f'_ionproducts. The current design basis accident (DBA) source term, TID.14844, is used three distinctways in licensing: 1) to conduct siting evaluations as required by 10 CFR Part 100; 2) to definethe radiologicalenvironment conditions for certain plant systems, and 3) to assess the effectivenessof plant mitigation features. It is used, for example, to establish the containment's allowable leakrate, containment isolation valve closure times, shielding requirements, and environmentalqualificationdoses.

Work currentlyunderwaytoupdatethetechnicalbasisforthesourceterm(TID-14844)isbaseduponcurrentlyavailablesevereaccidentresearchinsights,Changeswillbemade tothetimingoftherelease,thecompositionandmagnitudeofthereleaseintocontainment,and intheiodinechemicalform. The updatedsourcetermwillbe documentedina NUREG reportwhichwillreflectbothproductionand(toalimiteddegree)removaloff'_ionproductswithinthecontainmentfollowingcoremeltaccidents.(DuringthismeetingtheStaffdidnotdiscussmitigativeeffectsonthesourcetermbecauseworkintheseareasiscurrentlygoingon.)

A conceptualdepictionoftheupdatedsourcetermrelativetotheTID isprovidedinAttachmentI(thereasonfortheshadedareaisthatwhen theNRC developedthisschematic,itdidnotknowwhetherthe.;bsolutevaluesoftheupdatedsourcetermweregoingtofallbeloworabovetheTID-.theyhaveturnedouttobeabove).Attachment2 providesa breakdownoftheupdatedsourcetermthatisreleasedduringthevariousphases(thesum ofthereleasesduringeachphasecanbecomparedtotheTID --AttachmentI inMS-92-06providessucha comparison),The gapactivityisreleasedduringthetimeintervalwhenfuelpinfailureoccurs(andnogrossrelocationofthecorehasoccurredyet).The Staffnotedthat0.3.hourgapactivitydurationinAttachment2 isan

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averageand thatthis is basedon a rangeof 0.1 to 0.5 hours. However,based on morer_ntresults,the Staffexpects to changethisrangeto 0.5 to I hour. The earlyin-vesselreleasephase,whichends when the vessel fails,has a range of 0.7 to 3 hours;the ex.vesselphase lasts 2 to 5hours.

ThereleasesinAttachment2 reflecttheprobabilisticweightedmeanvaluesfromthe NUREG-!150sequences, That is, the NUREG-1150expertelicitationuncertaintyrangeswereappliedto BNL'sresults which deal with the quantity, type and timing of the source term (the NUREG/CRdocumentingBNL'sfindingsis to be issuedfor commentshortly). The numbersin the attachmentare boundingfor the variousaccidentsequences thatwereevaluated;that is, they are the highestreleasevaluesthatwouldbe seen as a resultof anyindividualsequence. In addition,these valuesreflectno in-containmentremovalmechanisms.Creditforsuch mechanismswillbe given;however,the Staff was not preparedto discusshow much because this informationis currentlybeingdeveloped.

The Staffexplainedthat the coolant activityrelease phase,which occurspriorto any fuel failure(gap activity), occurs within 10 to 30 _ for plants that do not take credit forleak-before.break(LBB). The Staffnoted thatLBB plantswouldbe able to receive muchmorerelaxationfor containmentisolationvalveclosurebecausethe coolant activityrelease phase wasfoundto beginat 10 minutesin these plants. The durationof the coolantactivityrelease phaseis basedon the large break LOC,_. The Staff indicatedthat althoughits analysesidentifiedtheminimumtime for initiationof coolantactivityreleases to be 10 seconds for non.LBB plants,itwouldconsiderfurtherrelaxationof the containmentisolationvalveclosurerequirementsbasedonplant.specificanalysiswhichtakes into account plant-specificdifferences.

As to the iodinechemicalformsenteringcontainmentfromthe reactorcoolantsystem,ORNLhasconcludedthat a total of 5% elementaliodine (12)and hydrogeniodide (HI) would be released(withnot less than I% as eitherI, or HI), and the remaining95% wouldbe cesium iodine(Csl).Once in the containment,there is muchpotentialfor revolatilizationof iodine(as moleculariodine)fromwaterpools. This willdependon whetherpH is controlled. If pH is maintainedat or above7, revolatilizationis negligible;if the pH is not controlled(i.e., drops below 7), revolatilizationwouldoccur and the amountsand rates woulddepend on evolutionof _olarity, pH, dose, andtemperature. ORNL is currentlyevaluatingthese factorsin moredetail.

Beahm,who providedthe ORNLpresentation,commentedthatif pH iscontrolled,hydrazinewouldprobably11_ be needed in the spraysbecause very little iodine is present in the containmentatmosphere(hydrazineconvertsiodineinto iodide). He also commentedthatby havingboricacidin the sprays,you use up mostof the sodiumhydroxideor hydrazineto neutralizethe boricacid.Germanplants,whichuse waterin the spraysinsteadof boric acid, have recognizedthis fact. OneStaffmemberpointedout that the SRP was revisedapproximatelytwo years ago and it already

. allowsfor the eliminationof additivesfromthe spraysfor iodineremoval. However,Kressclarifiedthat the SRP onlyallowedfor the eliminationof bases (i.e., sodiumhydroxide);boricacidwas notincludedbecauseof criticalitycontrolrequirements.


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MS.92-04Page 8

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The draft updated source term NUREO is scheduled to undergo CROR review in February. ACommissionbr/efing will be held in March and it is to be issued for public comment in April. Asnoted earlier, the updated source term will be adopted formallyafter this report has been finalizedand a final 10 CFR Pan 50 revision is _ued.

£PRI ProposedSourceTerm vs.NRC UlglatedSo_ Term:

The Staff summarizedthe differencesbetween the EPRI and NRC resultsas follows: EPRIproposesto usesimilargeneralreleasephases(earlyandlate in.vesael,endex.veuel) althoughthegapreleasesare includedwiththe earlyin.vea_elreleasesandthedurationof the earlyandlatein.vessel phases are longer than the N"RC's. There b good agreement on the release fractions ofnoble gases, iodine, and cesium In.vessel; however, there is less agreement on the release fractionsof non-volatiles in.vessel. Dave Leaver of Tenera presented a more detailed description of EPRrsproposed source term and identified some of the design features that will be used to reduce itsimpact (see handout i).

Comments/_est_els :

o Kerr said that he cannot tell whether the updated source term is a good one without alsoknowing _ it will be used/applied.

o The Staflrexpects to initiate the process for the final Pan 50 revision in approximately oneyear.

o In response to Ward'squestion on whether the current source term is more conservativethan the updated version, King commented that it is not more conservative in all areas.With respect to timing, the updated source term is more realistic; however, in terms ofquantities released, the updated source term identifies more isotopes as being released thanwhat the TID assumed.

o Kerr wanted to know whether use of the updated source term will reduce plant risk. K/ngthought that it would because it is more realistic. Kerrseemed to think that the Stal]' hasnot assessed the implicationsof not using the new, more scientifically.based source term (inthe interim, i.e., on plants that may need to undergo licensing before the final Pan 50 ruleis issued). Ward added that use of the new source term would lead to a more economicplant design for certain systems and he thought that the Staff has not acknowledged this.Lewis interjected that the only way the source term can afrect risk is by changing the plantdesign. The new source term would Iced to a better balance between prevention andmitigation but, for this to happen, the source term will have to be incorporated into theregulatory process.

o Kerr posed the following questions: if the new source term reveals that the cons_uencesof a large break LOCA would not be as severe as previouslythought, will the Stal_ allowthe EDGe to D91have to be turned on as quickly? and will it allow a plant to change its

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MS.92.04Page 9

I| "

emergency plan since the immediate consequences would not be as harsh as previouslythought? King responded yes to the first question but did not know what the Staff woulddo for the second. Fie referred ACRS to HRR for insight on where and how the newsource terrawill be applied; however, he cautioned that it may still be too early to pursuesuch discussion:,because N'RR has not made any decisions yet on what areas will be relaxed.Such information w/ll be available when the final Part 50 change is developed. Kerrindicated that if the source term reveals that LBLOCA is not an appreciable contributorto risk F.DOs _ be allowed to start say within 15-20 minutes rather than 20 seconds.

o Softercommentedthat in additionto determiningthe onsetof fuel pin failure,theStaff islookingintoother factorsthatwouldaffectcontainmentisolationtiming,e.g.,whethertherewill bedebrisflying in thecontainmentpost.LOCA andwhethersuchdebriswill impactonthe abilityof thepurgevalvesto function.Thisworkwill be completedbythe endof 1992.The draft NUREG on the onsetof fuel pin failurewill be issuedfor commentshortly.

o The Staff plans to brief the ACRS on plateout and other deposition factors that will affectthe source term values presented during this meeting later this month.

o In response to ACRS questions during his presentation, Dave Leaver clarified that the(EPRI) containment design will be based on design bzsis accidents (LOCA), not on severeaccidents;however,EPPdhascommittedto not exceedASME ServiceLevel C for any ofthe g'vereaccidents,

_Lj.r__Release Dennl!lon:(ref: MS.91-117, MS.91-21, MS-91.18)

In SECY.90.405, the Staff proposed to define a large release as a release of radioactivity into theenvironment of a magnitude equal to or greater than a yet-to._-determined amount (most likelyexpressed in curies rather than s fraction of core inventory) which results in one or more earlyfatalities o_tte. The Commission approved this proposed definition in March 1991 and directedthe Staff to userepresentativesitecharacteristicsin developingthe magnitudeof the release. Inaddition,theStaffwasto ensurethat theproposedPart 100sitingrequirementswouldbeconsistentwith this definition.

The Staff's calculation of the largerelease magnitude is based on the source term releases from theNUREO.1150 and LaSalle studies since together they represent the entire population of currentU.S. plants. These source terms were then applied to the MELCOR Accident Consequence CodeSystem (MACCS) to calculate the resulting prompt fatalities that would be expected to occur.Basedon thesecalculations,the sourceteam comingclosestto predictingone prompt fatalitywereselectedascandidatelargereleasevalues.

However,severalparametersaffect the magnitudeof the largereleasesuchas the size of theexclusion area, the populationdensity,meteorology,protectiveaction assumptions,the thermalenergyof the release(the releaseheight),and the timingof the release. In itsevaluations,the

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Staffused the p,oposed Part100exx:Imdonareaboundary(EAB) andpopulationdensity(i.e.,EA.Bequal to 0.4 miles and a populationdensityof 1000personsper squaremile). The meteorologyused was bs,sed upon datafrom29 weatherstationsaroundthe U.S. Threeemergencyplanningcases wereconsidered:

o mean (i.e., the warning time is sequence.dependent,the delay time is 2 hours, theevacuationspeed is 2.6 m/see,and the percentevacuated in 99.5%);

, o conservative(i.e., the warningtimeis sequence.dependent,the delaytime is 6 hours, theevacuationspeed is 1.1 m/se_,and only9S% of the populationis evacuated);and

o no emergencyplanning(i.e., no evacuation).

In responseto ACRS questions,Softerexplainedthat the 99.5% and9_% evacuationvaluesaresomewhatarbitrary.

Attachment3 identifiesthe Staff'slargereleaseresultsassuminga 3800 MWtplantwitha 0.4-mileexclusionarea,a 1000persons/sq,mile populationdensity,an 80thpercentilemeteorology,and noprotectiveactions, The valuesshownwouldlead to one promptfatality. In equivalentcuriesof1.131, Ihe ground release values equal 3x10' and the elevated release values eq,_,l 1x10'.Attachment4 prov/desthe largereleasevalueswhen meanemergencyplanningactionsare taken(these vah:eshavenot been convened to equivalentcuries of 1-131yet).

The Staffhasdecided todeferadoptingthesevaluesfor the largereleasemagnitudeuntilsensitivitystudiesare done to assess furtherthe impactof such factorsas meteorolo_, the energy of therelease,and a rangeof protectiveaction assumptions. The Staffwants to ensure that the largereleasemagnitudewill notbe overlyconservativerelativeto the promptfatalityOHO (whichcouldoccurbased on the factors noted above), therebycreatinga de facto safety goal. The ACRSseemed to agree with the Staff'sdecisionon this point. In fact, Wardnoted that the ACRShadstated in the pastthat it did not reallylikedefininga largereleaserelativeto a fatalitybecauseitwouldend up being muchmoreconservativethanthe safetygoal. Accordingto the Staff,it turnedout to be approx/matelythreeordersof magnitudemoreconservativethanthe OHOs. This iswhythe Staff needs to do the sensitivitystudies andrethinkthis definition,accordingto Ader. Kerrurged the Staff"to explore how the large release definition will be used from a regulatoryperspectivebeforeformalizingthe definition. The Staff agreed to do so.

The followinghandoutsare availableupon requestfrom the SF..RCHStaff':

a) I schematicof the variousPartI00 and Pan ;50updateactivities,by King

b) 18.pagevugraphon the proposedPan 100 cr/teria,by Softer

c) 14-pagevugraphon the large releasedefinition,byAder


[ _ Ii

MS.92.04. Page II

II

d) 1-page overview of the source term update, by K/rig

e) 35.page vugraph on iodine chemical forms in LWR severe accidents, by Beahm

f) 34.page vu_aph on the estimate of radionuclide release characteristics intocontainment under severe accident conditions, by Nourbakhsh

g) 6-page vugraph on fission product timing for gap activity (fuel pin failurecalculation), by R/dgely

h) 7.3.page vugraphon the update of TID.14844 source term, by Softer

i) 14.page vugraphon the ALWR requirements document source term, by Lcaver

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q

............... i __ i L i ..... 3 __ ....... i _ t - _ nil

,:ONCEPTOF REVISEDSOURCETERM

Tl:)-14844

• RELEASESAFTERVESSELFALURE

RELEA._S TO CONT_TBEFOREVESSELFAILURE

GAP ACTIVITY

o

o

Table 3.11 BWR Releases Into Containment"•, - I lill " II |1 I '11 I . = n

F,.x..Vcs,lGap Early ....... I Late In--Vessel

Release ln--Ve.s_ Limestone, I Basaltic"ii ii_l I [ ] i • i ii i i iiii

Duration (Hours) .3 1.5 3. 3. 10.ii ii i ii i i [

Noble Gases .05 .95 0 0 0I iii u i ili i I iiiill i

Iodine .05 .27 ...17 .17 .07ii i I I I IU i| II i I i i

Ccsium .05 .20 .45 .4:5 .03• III iii • El|i I i i i iiiii

TcUurium 0 .11 ,.18 ,37 .01i ii I L I I I I I ii

Strontium 0 .03 24 .22 0I I ii I I I I ii

•Barium 0 .03 .21 .17 0iiiiii I I I i ii I I I iii I ii

Ruthenium 0 .007 .004 .1104 0ii i II III iiiiiiiiiI I ii lli i i i ii

Cerium 0 .009 .01 .005 0iiL I I III II I• I I I II II

Lanthanum 0 _. .002 .01 .007 0• | IL l _M l j 'I II ,,-- I I I q IIII I III ,,s , ,,,

• This is a ratio of PWR(Oasaltic/l,imeslone) limes BWR(Limesloue)

• _ :_ince there is no information on BWRs wilhBasaltic o0morele.Ill I I llI I I II II El I II I I

are coremventoryr:

co m.

I_ l-tl r'l- Io

I_ IXj ,I_,,

o

l

/

Table 3.12 PWR Releases Into Contaimncm"

Ili III I Jl , _ , =L ,, lib

IP_.t--V,msrJ

Gap " Early ! LateIn--Vessel. • " Release, In-.v_l l.im__ Basaldc

D_a_timm(l.lourl)_ '_:'.:_--" " .3 .'/ '2,,, 2, 10._ J. III I I I II u

NobleGascs '""':'_'":";" • ..- .,.:.......-.-_" .05 .95 0 0 0I i I I II II I I IN I I

+" : " ": .I:. ':.:

Io_ . ::-....-..--. .05 .40 .29 . -.29 .07iiii I iiii i

Cesium -'_"+_"+.::+ .05 .310 .39 .39 .06i+ I _ I I illlll II llli

TcUu_um -., :..:--.'. - . 0 .U .29 .7,8 .025i ii I II I II J III II II tl Ill II

Sllroutium --. 0 .03 .12 .11 0!

I l Jl i ii ii Ill I

i Barium 0 .04 .10 .08 0I I I III I I I i l I

Ruthenium 0 .008 .004 .004 0l I I I I III i I

Cerium • 0 .01 .02 .01 0lill i I i III _ w _ iii iii i ii mill I

lanthanum 0 .002 .015 .01 0--- .. ,.I' I pill i i i

'" Valuesshownarchacdom _ cmc inventory

r.n_

""" _rt"

-, _.mtD O_l_a_) I-h ct !

1

PRELIMINARY LARGE RELEASE RESULTS

o PRELIMINARY RESULTS FOR 3800 Mwt REACTORUSING REPRESENTATIVE SITE PARAMETERS:

- GROUND LEVEL RELEASE- 100% NOBLE GASES,- 3% IODINES,- < 1% OTHERS.

-- ELEVATED RELEASE - 100% NOBLE GASES,- 9% IODINE, CESIUM,- 8% TELLURIUM,- < 5% OTHERS.

O32]0¢D (D _lrt

"O < Wrte,-,I- -

uCl3 o mn_

_I I

fJD l--hrl" I0

l, o

PRELIMINARY LARGE RELEASE RESULTS...... (CONTINUED) -- -

O PRELIMINARY RESULTS FOR 3800 MwT REACTORUSING REPRESENTATIVE SITE PARAMETERS WITHMEAN EMERGENCY PLANNING ASSUMPTIONS

- GROUND LEVEL RELEASE- 100% NOBLE GASES- 18% IODINES- 9% CESIUM- 27°/; TELLURIUM- <1% OTHERS

- ELEVATED RELEASE- 100% NOBLE GASES- 29% IODINES- 26% CESIUM

o,:, - 19% TELLURIUMCO (!)

=< - 13% STROI_,rrlUM _3 ° =,-,.

8% BARIUM __(!) -., _n3:::TO3

= - <1% OTHERS °_r,.D O_bJI'_ Htrt I

oI--" .¢_ .¢_

.......................................... i!_ i,._ 'i_'_:, _ i_'_

I_ilJilHPlI,:;,_ _-i_ -_i]]_,,.............._-----_i..._-i-_;_i-_ ],,I,,-- -i i....._1.............I-_ "-_-",r

SERVICE FOR EVALUATING REGULATORY CHANGES

MEETING: ACRS Full Committee Meeting

DATE: January 10, 1992 APPLICABILITY: ALWR/BWR/PWR

NUMBER: MS-92-06 KEY SUBJECT: SitingSource TermsReliabilityBackfit

The topics ACRS Full Committee heard presentations on and/or discussed included the following:

1) The proposed Part 100 siting revision, the source term update, and large releasedefinition.

2) International work underway with respect to advanced reactor safety features. (Thiswas a very philosophical presentationdiscussion by Charles Forsberg of ORNL.)

3) The adequacy of the NRC's metrication policy. (This was a brief discussion betweenthe ACRS members on the pros and cons of converting from English units to metricand the potential safety issues that are likely to arise during the transition. TheCommittee has written to the Commission in the past supporting conversion tometric units and acknowledging that safety issues could arise during the process.During this meeting, the ACRS agreed to have one of its subcommittees look moreclosely into the potential safety issues e.g., whether the ASME is planning toconvert the Code to metric and when.)

4) NRC's proposed final policy statement on integrated schedules.

Part !o0/Sources Term Update[14}reeRelea,se:

_RC Staff Participants:

Thomas King, Deputy Director, Division of Safety Issue Resolution (DSIR), Office of NuclearRegulatory Research (RES)

Charles Ader, Chief, Severe Accident Issues Branch (SAIB), DSIR, RESLen Softer, SAIB

i,_ BECHTEL POWER CORPORATION9801 Washlngtontan Blvd., Galthersburg, Maryland 20878-5356

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The ACRS Full C_mmittee heard a condensedversion of what was pr_ent_ at the JointSubcommittee meeting on these three subjects January 7.8 (see SERCH MS-92.04). Below aresome of the comments/clarifications offered during this discussion:

o Regarding the proposed Part 100 revision, Cation thinks that Part 100 should includeprovisions concerning meteorology. In his view, if the Staff plans to place standardizedcontainments on the sites, it will not bc able to account for meteorological problems thatexist in some areas of the country by modifying the containment design. The Staff notedthat it is aware of this limitation and is currently thinking of how to account formeteorology (e.g., how to account for unique inversion problems that may exist in a specificarea) under Part50. However, Siess disagreed with Catton's suggestions that containmentdesign modifications may bc needed in such cases, i.e., to account for variations in risk fromone site to another due to differences in meteorology. Siess pointed to the seismic hazardiss,)c. The seismic hazardvaries from site to site., however, the NRC is not going to placea different containment on each site to account for this difference. All (eastern) plants willsimply bc required to be designed to a 0.3g SSE. This recognizes that the seismic hazardrisk will continue to differ from site to site. Catton agreed that a similarapproach can beused to account for meteorological differences or unique meteorological problems.

o Attachment I provides a numerical comparison between the updated source term andTID.148,.14. This updated source term is a summation of the releases that would occurduring the various phases of an accident (see SERCH MS-92-04, Attachment 2).

Kerr pointed out that the comparison (Attachment 1) of the 0.25 "liD iodine release to theupdated source term iodine values is inaccurate because the 'riD value reflects plateoutconsiderations whereas the updated source term values do not. Softer agreed with thispoint but did not necessarily agree that the 0.25 should bc replaced with 0.50 as Kerrsuggested.

o The Staff also clarified that the updated source term values are mean values based on therelease ranges identified through expert elicitation for NUREG-1150. As a result, theupdated source term does not reflect the upper end of the risk range.

o In discussing the large release definition, Ader explained that the Staff used theconsequences portion of the MACCS Code to calculate the prompt fatalities that wouldoccur due to a release (as opposed to the in-containment portion of MACCS which wasused to calculate the updated source term). As to why no protective actions were assumedin the first calculation, Adcr said that the Code does not account for any evacuation forseven days and the Staff wanted to keep the model as simple as possible. However, basedon the results of this calculation, the Staff has decided to model protective action variablesas well as other variables such as meteorology (see MS-92-04) to ensure that conservatismsin the large release magnitude are minimized because NRC does not want the large releasemagnitude to become a de facto safety goal (for the prompt fatality quantitative healthobjective -. QHO).

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As part of the rcassessment effort, the Staff will coordinate with the safety goalimplementation group to iron out how the large release definition should/will be used onceit is developed. Several ACRS members thought the Staff should have already identifiedhow it intends to use the large release definition because without this information, onecannot judge whether the definition (i,e,, magnitude) is appropriate, Ader acknowledgedthis and added that this is why the Staff is only providing a status briefing at this time andnot requesting endorsement until later.

Carton wanted to know how the large release definition may be used for siting or fordecidingwhat type of plant to build. Ader responded that the safety goals are not intendedto be used on a single site or plant but rather in evaluating the adequacy of NRCregulations by seeing how the entire population of plants meet them.

Lewis, Ward, and Kerr agreed with the Staffs approach in attempting to define a largerelease in termsof an equivalent Iodine-131 curie level. They also share the Staff's concernwith respect to the level of conservatism, relative to the QHO, that the large releasedefinition should be allowed to have.

!nt_rated Schedpl_s:

Marylee Siosson of the Policy Development and Technical Support Branch within NRC's Officeof Nuclear Reactor Regulation (NRR) provided a brief overview of the proposed final policystatement on integrated schedules. In 1987, the NRC issued the proposed policy statement toimplement the use of integrated schedules on a voluntarybasis. This programwould have includedall the NRC-initiated items as well as any voluntary actions by the utility. Adoption of this programthrough a license condition was optional but greatly emphasized. Changes to the schedule wouldhave been permissible with prior NRC notification.

Most of the comments the NRC received opposed the proposed policy because it placed too muchemphasis on incorporating the programinto the license, the scope was viewed as being too limited,and some felt that such a programwas not necessary for utilities with good regulatory compliancehistory and not much of a backlog.

The next program the NRC pursued, in response to the regulatory impact survey, was theintegrated regulatory requirementsimplementation schedule (IRRIS). It was developed to managethe cumulative impact of implementing generic requirements; utilities would have been allowed totrack their own initiatives. The implementation schedules would have been proposed by the utilityon a safety priority basis. IRRIS would have placed a 90-day negative consent review limit on theNRC, and implementation of the programwould not have required a license condition. Despitethese provisions, utilities did not choose to adopt the program. In response to ACRS questions asto why, Slosson said that utilities felt that NRC had "missedthe boat"in suggesting this programbecause utilities were interested in being able to influence the front-end of the generic

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requirementsprocess(before thegenericrequirementsare everissued),notjust thebackend(theimplementationlogistics). In the latestregulatoryimpactsurveyreport (SECY.91-172), the NRCdecidedto not pursueIRRIS and insteadto finalizethe proposedpolicy[hotwasissuedin 1987.

The Staff hasdevelopeda proposedfinal policy paperwhichiscurrentlywith theEDO and, uponACRS feedback,it willbe forwardedto the Commission.The policy paperidentifiestwo options:

1. Issue the final policy statement with modifications, That is, the program would bevoluntary,wouldn._ require a licenseconditionfor implementation,wouldincorporatea90.day negativeconsentlimit on the NRC'mreviews,and wouldintegrate three levelsofitems: regulatory requirements (e.g., rules, orders or license conditions, technicalspecificationsandlicenseamendments),utilitycommitmentsto genericlettersandbulletins,and utility-initiated items. Under this program, utilities could not change theirimplementationschedulesfor the Level I itemswithout requestingan exemptionor licenseamendmcnt. For the Level 2 items, utilities would establishand revise schedularcommitmentsin accordancewithprioritymethodology.As to theLevel 3 items,the utilitywould establisha monetary thresholdfor decidingwhich items it wants to track in theschedule.

2. Issuethe 1987proposedpolicy (see SERCH RC-87.53) as final with no modifications.

The NRC Staff is recommendingthat the Commissionadopt Option I. The Staff addedthat itwants to finalize the policy to protect the plants that already have integratedschedulesand toattract other utilities aswell. Attachment2 identifies the plants that currently have integratedschedules.

The NRC Staff was advisedthat ACRS has no objectionsto the Staff's recommendationforadoptingOption 1. However, becauseACRS has made its views known on this matter in pastletters to the Commission,it decidedthat it did not need to write another letter now.

Thc following handoutsare availableupon requestfrom the SERCH Staff:

a) 42-pagevugraphon Part100,updatedsourceterm,andlargereleasedefinition,byKing,Soffer,andAder

b) 52.pagevugraphon passiveandinherentsafetyinpowerreactors,byForsberg

c) 20.pagearticle,"AdvancedReactors,PassiveSafety,and AcceptanceofNuclearEnergy,"byC.W.ForsbergandA.M. Weinberg

c) 13-pagevugraphon integratedschedules,bySlosson

Rev.0September1992

-2

ISOTOPE R.G.1.3/1.4 UPDATEDBk'R UPDATEDPMR

NOBLEGASES 1.0 1.0 .0IODIHE 0.25 0.76 0.81CESIUM 0 0.73 0.80TELLURIUM 0 0.5 0.465STRONTIUM 0 0.27 0.15BARIUM 0 O.24 O.14RUTHENIUM 0 O.011 O.012CERIUM 0 0.019 0.03

P

o, LANTHANUM 0 0.012 0.017Q Q

_:_ ,v,,-3 UPDATEDBWRANDPHRRELEASEFRACTLONSAREFORL]]F..STONE ,,-"':' CONCRETENITHOUTHATEROVERLYINGCOREDEBRIS _-_"

• 0 :::I_

0

iiiiiiiiM II I I

II

,, MS-92-06, .Attacb_nent 2

Page i of I

PLANT8 WiTHINTEGRATED SCHEDULE8

1SAP PLANT8

HADDAM NECKMILLSTONE 1

INTEGRATED SCHEDULE PLANT8

DUANE ARNOLD, BiG ROCK POINT

OYSTER GREEKPILGRIM8AN ONOFRE 2,8TMI- ITURKEY POINT 8,4

PLANTS INDICATING INTEREST ININTEGRATED SCHEDULE8

COMANCHE PEAK80UTH TEXASTROJANCRYSTAL RIVERMILLSTONE 2,8SALEM

Rev.0September1992

II IIIIIII I I ....

MEETINO: ACRS Subcommittee on Extreme _ernal Phenomena M_ting on 10 CFPart 100, Appendix A, "_Lsmlc and Geologic Siting Criteria for Nuclear PowerPlant,i"

DATE: February 5, 1992 APPLICABILITY: ALWR

]_JMBER: MS._2.26 KEY S__: Siting

_RS_Members PresenLinc_ded: Industry Representatives Irl;ludM:

Chester Siess, Chairman Tom O'Hara, Yankee AtomicCarlyle Michebon John Sutton, Yankee AtomicCharl_ Wylie Ray Ng, NUMARCHarold Lewis

. JamesCarrollBill Ltndblad,ACRS Consultant

NRC StaffMembersPrfsenLlncJ_uded:

LarryShaG,Director, Divisionof Engineering(DE), office of NuclearRegulatoryResearch(RES)And),Murphy,Chief,Structural& SeismicEngineeringBranch(SSEB),DE, RESNilesh Chokshi,SSEBRoger Kenneally,SSEBOoutamBagchi,Chief, Structural& OeosciencesBranch(ESGB), Office of NuclearReactor

Regulation(NRR)Robert Rothman, ESGBPhyllisSobel, ESGB

This was a follow.up meeting to the December 10, 1991 meeting (see MS.91.166). The Staffpresented a brief update to what was heard in December, an overview of draft Regulatory Guide(RO) DO.1015, "SeismicSources,"whichoutline=the dualsitingapproach,andan exampleof howthisapproachwouldbe used. In addition,the SubcommitteeheardYankeeAtomic'=viewson theproposedsitingrequirements.

Updateof the 10 CFR 100,AppendixA, Packalle:

Roger KenneallyinformedtheSubcommitteethat theStaff has not madeany technicalchangestothe proposedAppendix A regulationssince the December meeting, However, it has made

@BECHTEL POWER CORPORATION9801 WashinglontanBtvcl.,Gaithorllburg,MaP/tanct20878.5356(301) 417.3099 Rev,0

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4 ' ml .... :substantive editoflal changes as a result of discussions with the ACRS, the Omce of the GeneralCounsel and NRR. The changes include the following:

• The Staff hu removed the 'zero period acceleration" dei]nltion and repla_ it with thephrue "peakground acceleration."

• A definition for 'structures, systems, and components required to withstand the effects ofthe safe shutdown earthquake ground motion and surface deformation" was added.

e The phrase "ground motion" wu added after referen_ to "operating buis earthquake."

With respectto the supportingregulatoryguides,the Staff hasmade two technicalchanges: )) itremovedthe exceptionto thedefinition of "feltearthquake"in EPRI NP.6695 andnow allowsforinstrumentactivationg.Lconsensusof thecontrolroomoperators(before,instrumentactivationconsensus of control room operators constituted a felt earthquake); and 2) it added that, after anearthquake, the response spectrum and cumulative absolute velocity (CAV) should be calculatedusing a calibration standard to demonstrate that the free.field instrumentation functioned properly.The Staff subsequently noted that one more change has been made: draft RG DG-1015 is beingrevised to require use of only median values when calculating controlling earthquakes via theprobabilisticapproach (rather than median, mean, and 85th percentile).

In discussing draft RO DO-1017, 'Pre.Earthquake Planning and Immediate Nuclear Power PlantOperator Post-Earthquake Actions,"Sie._ noted thatconcrete sometimes requires seals for flooding,etc. The post-earthquake walkdown, he said, should look for damage to such seals/constructionjoints even though one may not necessarily know if some of these seals have been damaged untilwater seepage has occurred. Michelson and Kenneally agreed and Miche!son added that each utilityshouldhavea log of all the sealsfor the operatorsto use followingan earthquaketo be able tocheck for possible damage.

Ooutam Bagchi added that, at the request of NUMARC, NRR is proposing to issue a generic letterto allow for the voluntary use of newer instrumentation (than is referenced in RG 1.12,"Instrumentationfor Earthquakes') in current plants. According to Bagchi, the incentive would bethat if a plant does not use the newer instrumentation, it would need to continue to comply withstricter shutdown requirements (see SERCH RC-92.08).

Draft RG DG.101_;/Dual Sltlnll Approach:

Andy Murphy provided an overview of this RG by giving an example of how the proposed dualsitingapproach(see Attachment1) wouldbe applied. The Staff clarifiedthat the deterministicanal).sis would consist of the same investigation required by the current Appendix A regulation.The new portion is the probabilisticanalysis. At this point Siess made it known that he does not

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, MS.92.26Page 3

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see why the Staff is requiring both analyses rather than just the deterministic approach which hasworked fine until now. Both Sie.u and Lewis asked the Staff to state the philosophy/buts forrequiring both approaches because, in their view, it appean u though the NRC Staff made thedecision to require both approaches tint and then looked for reasons to back up its decision.

According to Murphy, the probabilistic approach will lay to rest debates on the probability ofexceeding the SSE at a site and whether the scope of the geological investigation on which the SSEselection was based went far enough. Sieu wu not satisfied with this explanation and again _kedwhy the probabilistic approach is being _. This time, Murphy responded that the intent isto make future plants easier to license and noted that the Commission has said that new plantsmust be at least as safe as current plants. St_s then asked the Staff to explain how licensing wouldbe made easier with the use of the probabJlisticapproach. Lewis asked the Staff to identify whatgoes into the deterministic approach that is not available for the probabilisttcapproach. Murphyresponded that the databases used for both approaches are identical; however, the deterministicapproach does not address or account for outlier issues (uncertainty). There is nothing in thedeterministicapproachthatisnot usedin the probabilistlcapproach,accordingto Murphy.However,Rothmannotedthatinthedeterministicapproach,thedetailofthegeologicstudiesisgreater;theprobabilisticapproachdra_ froma frozendatabaseofgenericregionaldata.

Murphy againemphasizedthattheselectionof theSSE underthecurrentregulationisnottechnicallywrong ..theprobabilisticapproachwillsimplyputtheearthquakethatisselected(deterministically)intheproperperspectiverelativetootherlargerearthquakesthatcanoccurintheareawhichsomemay argueshouldbeselectedastheSSE. Thisiswhy DiabloCanyonendedup usinga prohabilisticapproach,accordingtoRothman.Chokshiaddedthat,in1986,therewasa 3.daysymposiumon AppendixA whichrecommendedusingprobabilistictechniquesto"laytorest"questionsthatarise,evenaftera plantislicensed,regardingthepossibilityof havingtoincreasetheSSE level.Siessconcludedfromthisthatusingthetwo approacheswillnotmakeplantssaferbuteasiertolicense.

At thispoint,Murphybeganhisexampleofhow thedualapproachwouldbe used:

Basedon thegeologicalinvestigationoutlinedinAppendixA, one wouldidentifytheseismicsourcesinthevicinityofa proposednuclearpowerplantsite.The expectedmaximum earthquakewouldbe determinedforeachseismicsourceand thisearthquakewouldbe assumedtooccurattheportionoftheseismicsourcethatisclosesttotheproposedplantsite.The spectrawouldbecalculated(usingRG 1.60orsomeotherbasis)fortheexpectedmaximum earthquakefromeachseismicsource.Eachofthesespectrawouldthenbecombinedintoa singleenvelopingspectrumwhichwouldbetheplant'sdesignspectrum(i.e,theplant'sSSE groundmotionspectrum),Allthis(Attachment2)wouldconstitutethedeterministicsideofthedualapproach,inresponsetoStaffcommentsthatthe probabilisticapproachwould helpin respondingto a challengeof thedeterministically.selectedearthquake,Siessobservedthatnothingwouldstopan intervenorfromgeneratinghisown probabilisticweightingfactorsforcalculatingthebiggestearthquakefromeach

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, MS-92-26

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_BRIB

seismic source. The Staff responded that, in such a case, one would have to do sensitivity analysesand ultimately rely on the judgement of the majority of experts.

Under the probabilistic approach (Attachment 3), one would a) define the seismic sources from theEPRI or LLNL seismic hazard study archives; b) characterize the recurrence parameters for therange of earthquakes from each of the seismic sources defined in "a";c) attenuate the earthquakeground motion intensities from each source to the plant site; and d) integrate all the attenuationcurves from all the seismic sources and distances to obtain the seismic hazard curve for the plantsite. According to Rothman, all four steps can be carded out in approximately a half an hour byrunning the latitude and longitude of the proposed site into either the EPRI or LLNL computercode. Step "d" identifies the peak ground acceleration (PGA) for the site which would then beconverted to the CPUHS, the consequence percentile uniform hazard spectra (50th percentilespectra for all return periods).

At this point, the Staff noted that if one enters a plot of the cumulative distn'bution of theprobability of exceeding the design basis earthquake for all current eastern U.S. plants (using themedian LLNL hazard estimates) at the 50% cumulative distril:,ution level, one identifies a targetprobability of exceedance of 1E-4. New plants will need to have an annual probability of exceedingtheir SSE that is lower than this value. The 1E-,4target probability is based on the LLNL hazardestimate model of all five of the LLNL ground motion experts. Entering a similar plot whichreflects the hazard estimates derived from using only four of the LLNL experts, yields a 3E-5probability of exceedance (the 5th ground motion expert substantially skewed the LLNL seismichazard results -- this is why the Staff decided to check both cases here) If the EPRI seismichazard data are used, the target probability of exceedance is also 3E-5. One would enter thesite-specific seismic hazard curve (plotted as annual probability of exceedance vs. spectral velocityaveraged at 5 a,ld 10 Hz) to determine the corresponding site-specific spectral velocity. This is aballpark value, not the SSE yet.

One would lhen deaggregate the site probabilistic seismic hazard curves by magnitude and distance(see Table 1 in Attachment 4). For each bin in Table 1 of Attachment 4, the seismic hazard curvewould be computed. For each of these curves, one would enter at the site-specific spectral velocity(obtained earlier by using the 1E-4 value) and would read off a corresponding probability ofexceedance. This value would then be filled into the appropriate bin as shown in Table 2 ofAttachment 4. (The numbers in this Attachment reflect the Vogtle plant.) The next step wouldconsist of calculating the magnitude and distance of the controlling earthquake(s) by combining(taking the weighted sum of) the relative contributions/probabilities of all the earthquakes in thebins in Table 2 of Attachment 4. The results of this step, for Vogtle and three other plants, areprovided in the table on Attachment 5. The Staff intends to develop this information for theremaining plant sites while the proposed Part 100 rule is out for public comment.

In response to Siess's question as to why one needs to deaggregate, why not just use the valueobtained from the 1E-4 probability of exceedance, Chokshi noted that this value is only one point

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on the spectral shape (at 5 and l0 Hz) and the deaggregation would provide a broader picture.Siess still questioned the need for doing all this work He noted that we have known there areuncertainties in what was done/used in the past. We now have a method for quantifying theuncertainties, he said, but why integrate it into the licensing basis? He pointed out that thenumbers in Table 2 of Attachment 4 are only medians anyway, so why bother with all this? InSiess's view, the Staff is making the siting process too complicated without a clear basis for why ithas to be done this way - "all we want to do here is come up with a number that can be used fordesign!" he exclaimed. Murphy again attempted to explain that the probabilistic process isattempting to come up with the same controlling earthquake(s) identified under the deterministicapproach but with an accountlng/identification of the uncertainties. Siess thinks the probabilisticapproach has a lot of curves which could be challenged by someone; therefore, he did not seemto think that licensing would necessarily be made easier by requiring use of the probabilisticapproach.

In looking at the Vogtle results in the table of Attachment 5, Siess asked the NRR Staff whetheran applicant coming in with this information would be approved. Bagchi responded no, that thereis insufficient information, at which point Murphy interjectedthat one would need to develop aspectral shape for each of these earthquakes, see how they compare, and envelope the two. Thegraph in Attachment 5 provides such a comparison for the Vogtle earthquakes. The "D" curve isthe plant's actual design SSE spectrum; the "R" curve is a reference spectrum (the RG 1.60spectrum anchored at 0.3g); "C" is the ground motion spectrum for the controlling earthquake usingthe LLNL 5-expert model for attenuation; and "S"is the ground motion spectrum for the magnitudeand distance earthquake of the plant's actual design SSE using the LLNL 5-expert model forattenuation (i.e., the plant's deterministically generated controlling earthquake).

Looldng at these curves (Attachment 5), Siess again asked NRR if it would approve _:.:eof themagnitude 5.3 at a 15 km distance earthquake for the design of Vogtle if it was an applicar_tunderthe new rule. Goutam replied that he would not; Rothman did not respond. From RES'sperspective, the probabilistically-generated controlling earthquake (5.8 at 32 kin) compared favorablyto the deterministically-generated design-basis earthquake (5.3 at 15 km); therefore, such anapplicant would be approved. At this point Siess commented that if NRC Staff cannot agreebetween themselves whether a plant would be licensed based on this information, how can it saythat the licensing process is being simplified by adding the probabilistic approach? He did not thinkthat an applicant will gain anything from doing the dual-approach and added that the Staff iskidding itself by thinking that it will. In his view, the weakness of draft RG DG-1015 is that itrecommends use of the probabilistic approach. He added that although the Staff explained howthe dual process would work, it did not discuss what it would do/require if the deterministic andprobabilistic methods yielded greatly differing results.

Siess closed this portion of the meeting by saying that he will suggest to the ACRS Full Committeethat its letter to the Commission recommend that the proposed rule be issued for comment. He,however, will add in the letter that he sees no good reason for having to use the dual approach.

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He also commented that as long as this "requirement" to use the dual approach is in a RG(DG-1015), there is some flexibility whereas if the requirement is in the rule, it would be fixed.He observed that the source term details are being added to the rule (Part 100) while the seismicdetails are being taken out and being placed in a RG. Someone should look into this inconsistency,he said.

Ray Ng was asked to provide the industry's position on the proposed rule. Ng commented that theindustry does not have a position on this rulemaking package yet. The value of using the dualapproach in the licensing process is currently being assessed by NUMARC. He urged the Staff,however, to develop a clear purpose for requiring use of the dual approach, otherwise it may endup complicating the licensing process rather than simplifying it.

Yankee Atomic's Views on Appendix A:

Due to time restrictions, the Yankee presentations were considerably shortened and there was nottime for discussion or questioning of Yankee's suggested approach. Tom O'Hara basically informedthe Subcommittee that use of the deterministic approach will not work any longer in the easternU.S. unless the NRC defines the seismic sources and the expected maximum earthquakes for eachof the sources. If it does not, each applicant will be open for intervention on this step of theprocess. He proposed doing all the geological/seismological site determinations determinlstically andthen using _ the probabilistic approach to derive the SSE. This alternative siting approach forthe eastern U.S. is outlined in the two vugraphs and paper made available by the Yankee Atomicrepresentatives.

The following handouts are available upon request form the SERCH Staff:

a) 5-page vugraph, "An Update of the Presentations and Regulatory Guides AssociatedWith the Revision of 10 CFR Part 100, Appendix A," by Kenneally

b) 27-page vugraph, "Draft Regulatory Guide on Seismic Sources," by Murphy

c) 44-page copy of draft RG DG-1015, "Seismic Sources"

d) 41-page vugraph, "Comments on Draft RG DG-1015, Seismic Sources," by O'Hara

e) 7-page Yankee Atomic vugraph, "Proposed Revisions to 10 CFR 100 Seismic SitingCriteria and Related Implementation Guidance"

f) 27-page paper, "A Basis for Standardized Seismic Design (SSD) for Nuclear PowerPlants," by Yankee Atomic and Northeast Utilities

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SITE

DETERMINISTIC

PROBABI USTIC ANALYSIS

ANALYSISi i

I Geological, Seismologicaland GeophysicalInvestigationsiii i • I i I i II

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SCHEMATIC DEFINITION OF CONTROLLING EARTHQUAKES

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MS-92-26. Attachment 3

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d) Seismichazard curve (c) Ground motion predictionmodel (attenuation)

FOUR STEPS INVOLVED IN A PROBABILISTICSEISMIC HAZARD ANALYSIS


MS-92-26 1

. Attachment 4Page I of I

TABLE 1

DEAGGREGATE THE PROBABILISTIC SEISMIC HAZARD CURVES |YMAGNITUDEAND DISTANCEGRIDS,I.e., DEVELOPTHE CONTRIBUTIONTOTHE SEISMIC HAZARD FORTHE BINSINDICATEDBELOW.

Magnitudesand Distan©aBinsUsedIn Example

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DE_GGREGATEDRESULTS

MS-92-26., Attachment 5

Page I of i

INITIAL RESULTSFORFOURTRIAL SITES

SITE PROEABILISTICANALYSIS DESIGN"' -.......... EARTHQUAKE

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/_j__> SERVICE FOI_ EVALUATING REGUI_ATOFIY +HAI_I+ES "

_O:. Commission Briefing with the ACRS Full Committee

DATE: March 5, 1992 APPLICABILITY: AL_WR/PWR

NUMBER: MS-92-40 KEY SUBJECT: Advanced ReactorsSitingStandardization

The ACRS met with the N_C Commissionen to discuss three items of mutual interest:

(I) Design Acceptance Criteria (DAC)

(2) Proposed Revisions to I0 CFR Parts 50 and I00 (Nonseism/c) Rule Changes andProposed Update toSource Term

(3) Status of Advanced Reactors

Design AcceptanceCHterla (DAC)

In a February 14, 1992 letter to the Commission, the ACRS stated its support of the DACapproachfor I/m/tedapplicationsand encouragedthe Staff'to continue development of the processwith appropriate interchange with the industry and vendors. The ACRS believes, however, thatcareful]ydefined limits to the scope and extent of design coverage should be placed on the use ofDAC and recommendsthat the use of DAC be limitedto the portion of each design feature whereeither the technology is stiUevolving or the required information is unavailablefor a good reason.In any case, DAC should only be used when it is possible to specify pract/cal and technicallyunambiguous mater/al.

Michelson and Wylie disagree with the ACRS's v/ew that it wfl/be feasible to use DAC for theperformance of traditionalsafety reviews involvingfire and internal flooding. Wylie believes thatwhen safety decisions are made it is necessary to have _ed buildinglayouts, equipment locations,and space allocations,to have performed interference checks, and to have identified potential pipebreak and flooding concerns. Both Wylie and Micheison feel that it is technlcaIIyfeasible to supplythis kind of informationbefore cert/flcation.

SeIin stated that the shortcom]ngsof safety assessment with DAC are understood but wants toknow if havingthe complete design would reallymakethe use of DAC any better. Wylie said thathe personallycannot see making safety determ/nationswith just DAC; the layout of major pipingand conduits are needed, not just genera] locations of majorequipment.

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MS.92.40

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lOIISelin believes the DAC concept is nne and it is the implementationof DAC that is the problem,Selin would like the ACRS to continue the DAC discussion and look at the costs involved to finda DAC level that is acceptable but not outrageous economically.

ProposedRevisionstoI0CFR Parts$0and i00(Nonselsmlc)RuleChangesandProposedUpdate to Sourt_ Term

TheStafrhasproposedtheseparationof Part 100,"ReactorSiteCriteria,"fromthoserequirementsfor plantdesignwhichmoreproperlybelongin Part 50. A two-stageprogramto accomplishthishasbeendiscussed.In Stage1, radiologicaldosecriteriaandreferenceto tlsslonproductreleasequantities,thesourceterm,will be movedto 10CFR 50.34. Also,Part i(30will be augmentedbyaddingto it certainquantitativecriteria now specifiedin RegulatoryGuide 4.'7,'Genera] SiteSuitabilityfor NuclearPowerStations.'

In Stage 2, further changes will be made in Part 50 to update source term requirements byincorporating technical information about severe accidents developed since Part 100 was issued in1962. De:ailsof all thechangesto Part50 havenotbeendevelopedbut a preliminarydescriptionof thesourcetermandits derivationwill be containedin a draft technicaldocumentscheduledforissuanceinApril1992,

The ACRS stated,in a January15, 1992letterto the Commission,its beliefthat the firststageoftheStaff'sproposalis reasonableandshouldproceed.RegardingStage2, developmentof thenewsource tcrm is proceedingalong the right line; however,beyond that, the ACRS has majorconcerns:

o There is no planfor a Stage2 upgradeof Part 100. Stage1 merelyprovidesamore logicalarrangement and more completelycodiDes technicalinformationon sitesuitability which was developed30 years ago. No attempt is being made tomodernizethe rule. For example,the basisfor key requirementssuchas the0,4.mileradiusfor anexclusionzone,the 10-mileradiusl'oranemergencyplanningzone,and the maximumpopulationdensityfor the low populationzone has notbeenreexaminedor justifiedwith up.to-dateinformation.

o There is no planto incorporatemeteorologicalrequirementsin Part 100.

o Even thoughthe sourcetermplaysan importantpan in thePart 50 requirementsfor containmentperformanceandshouldbe includedin the update,it is not themost/mportantpart. Far moresignificantto riskare thecharacterLsticswhichwillgovernwhethera containmentwill continueto performits functionor fail duringa severeaccidentandwhethermitigationsystemswill operateell'activelyor not.

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.nnnStatus of Advanced Reactor Rt_vtews

General ElectdcAdvanczd Bollln_Water Reactor (ABe)

The ACRS Subcommittee on ,t.BWPasand other ACRS Subcommittees have held 14 meetingsbeginningon October 31, 1989to dtsctmthe NRC Staff_'sdraftsafety evaluation reports (DSER_t)corresponding to specific chapten of the GE StandardSafety Analysis Report for the ABWR.

The ACRS review of the final safety evaluation report (_ER) related to the ABWR will beginafter receiving the FSER from the Staff. This b tentativelyscheduled to be provided to the ACRSduring August 1992.

ABB,CF_._vuem_80+_

The ACRS Subcommittee on Advanced PreuuHzed Water Reactors has held four meetings todiscuss the System 80+ design features and related issues such as the Licensing Review BasisDocument. The Subcommittee plan=to hold one or two meetings in the near future. Subsequentto these meetings, the Subcommitteewill recommend a course of action to the Full Committee.

Carrollstated that he personallybelieves CE b doing a betterjob than GE in supplyinginformationto the ACRS. _ is actually building the System 80+ planu in France, while GE is merely apartnerin the Japan ABWR venture. Furthermore,the design of the Japanese ABWRs is differentthan the proposed U.S. design.

Michclson said that CE is provingthat design questions can be answered but that GE does notsupply the Staff with enough information. There are over 300 open issues in the GE SafetyAnalysisReport (SAR), leading Micheison to conclude that there b not enough information in theGE SAR for the Staff to develop their FSER.

WestinghouseA.P600

The ACRS has heard presentationsby Westinghouse Electric Corporation on the duign details of' Ithe AP600 plant, but the majority of the ACRS effort hu focused on mtegra systems testing

requirements.

Carton summarized the positions of the groups involved in the testing issue. The Staff is in favorof the full-height, full.pre=suretesta, while Westinghouse insists that they can prove their designthroughtheir codes and low.preuure tests at Oregon State Univenity (OSU). The Staff feels thatWestinghouse hu a robust testing program, but the toob they are using are too weak, especiallyfor small.break LOCA and slow flow conditions. The Staff is also not convinced that the systemswill work at high pressures and thus considen the full.height, full.preuture test to be necessary,

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MS.92.aOPase 4

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Seltnquestionedwhether full.height, full.pressuretatinll would be usefulat thisearly stageofdeslsn.He wonderedif it wouldnot be betterto put off the twtinltuntila marketfor the AP600desillndevelops,Cartonsaidthat thb is alsoa pointunderconsideration,

A 47.paRehandoutof backlFoundinformationfor thismeeting(ACRS lettersto the_mmission)is availableuponrequestfrom the SERCH Staff,


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I

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I1SERVICE FOR EVALUATING REGULATORY CHANGES

MEETING: NRC Staff Meeting with NUM.A.RCon Proposed Revision to l0 CFR Part 100,Appendix A, "Seismic and Geologic Siting Criteria for Nuclear Power Plants"

DATE: April 23, 1992 APPLICABILITY: ALWR

NUMBER: MS.92-78 KEY SUBJECT: Siting

N]_C S_taftParticipants:

Robert Bosnak, Deputy Director, Division of Engineering (DE), Office of Nuclear RegulatoryResearch (RES)

Andrew Murphy, Chief, Structural& SeismicEngineeringBranch(SSEB), DE, RESNile.shChok.shi,SSEB, RESRoger Kenneally,SSEB, RESRichardMcMullen,SSEB, RF.SErnstZofflueh,SSEB, RESRobertRothrnan,Structural& Ge.o_jencesBranch(ESGB),NuclearReactorRegulation(NRR)PhyllisSobel,F.SOB,NRRBakrIbrahim,Oeo]o_ & Enl_ineeringBranch,OfficeofNuclearMaterialSafetyand Safeguar_

Indust_Pa_icipants:

Ray Ng,NUMARCJohnButler,NUMARCCarlStepp,EPRIDennis Ostrom,SouthernCalifornia EdbonDonald Moore, SouthernCompanyServicesBob Whorton, SouthCarolinaElectric & O_Tom O'Hara, Yankee AtomicMike Hayner,ClevelandElectric IlluminatingCo.Sam Stone,TVAW,U. Savage,Pacific Gas & Electric

This meetingwasheld st the requmt of NUMARC to providecommentson the proposeddraftrevisionof Part 100, AppendixA, whichwasdiscu_edwith the ACRS in February(see SERCHMS.92.26, MS-91.166,MS-91.56,andMS.91-31). Murphynotedthat theproposedrule iscurrentlyundergoingeditorialrevisionfor claritybasedon commentsfrom theCommitteeto ReviewGenericRequirements(CROR), Once thesechang= are made, it will go back to CRGR (in May). Inmid-June,a briefingwill be held to discussthe packagewith the Comm_ion. The packageis

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expected to be i_ued in late July/early August for a _-day public comment period. According tothe Staff, the rule should be ready to be/uued as final approx/mateIy one year from now.

Ray Ng informed the Staff that NUMARC has mtablished an ad hoc advisory committee, whichwu present at this meeting, to advise the NUMARC Seismic Issues Working Group on thisrulemaking. The industry has reviewed the rulemaking package that was placed in the publicdocument room (PDR) earlier this year and found it difficult to understand. As a result, theindustry hoped to gain clearer understanding of the rulemaking package during this meeting.

According to Ng, the language in proposed Part 100, Appendix B is of concern to NUM.ARC. Itis too rigid and would lock utilities into a fixed siting approach/methodology thereby not allowingfor the use of any future advances in science that may provide a more stable licensing pr_.Therefore, NUMARC would like to change the language of the rule before the package is issuedfor public comment. Formal NUMARC comments on this matter are to be submitted in early May.

Murphy responded to this request by noting that, in the absence of some extraordinaryo¢,currence,it is unlikely that the Staff would revise the package prior to its release for public comment.Murphy apologized for the PDR version, agreeing that it is not very clear technically but addingthat the updated version the Staff is currently reviewingwould not be placed in the PDR prior toissuance for public comment. He pointed out that the release of the package earlier this year hadto go all the way up to the Commission for approval.

At this point Murphy presented several slides summarizing the structure and scope of therulemaking package. According to the Staff, the current Appendix A is too detailed, inflexible, andlacks clarity in some sections leading to conflicting interpretations and requiring discussions andadjudication by the licensing panels. Furthermore, Appendix A was issued in 1973. It does notreflect advances in the sciences of seismology and geology, and does not reflect the evolution ofthe licensing and adjudicatory proe___sses.He emphasized that the Staff does not want the newAppendix A to be prescriptive.

The revision to Appendix A is intended to remove current source_ of misinterpretation, increasethe ease for updating the technical guidance, and provide stability in license reviews. The revisionwill be completed in time for anticipated early site reviews. At this point, Ng interjected that, inOctober, the industry will hold a workshop for parties interested in siting an advanced reactor. Itis expected that a site will be selected in early 1993, characterized in 1993, and an early siteapplication submitted in 1994.

Murphy continued his presentation by noting that the existing Part 100, Appendix A will remainin effect for existing plants. The revised Appendix A will be called Pan 100, Appendix B and itwill be supported by a new regulatory guide on the identification and characterization of seismicsources and the use of probabilisticacceptance criteria, a revision to SRP Section 2.5.2, "Vibratory

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Ground Motion," two new regulatory guides on pre-earthquake planning and restart criteriafollowing an earthquake, as well as the possible revision of several existing regulatory guides.

The Staff's current version of Part 100, Appendix B was characterized as follows: It retains thedeterministic geological investigations, it allows for the use of either the EPRI or LLNLprobabilistic seismic hazard studies to assess the proposed site, and it requires that both aprobabilistic and deterministic analysis be done and that each analysis be given equal weight.

Carl Stepp questioned the need to lock in the last requirement mentioned (use of probabilistic an__dddeterministic analyses for the determination of the controlling earthquake). Murphy responded thatboth are needed to address uncertainty. Chokshi added that the existing Appendix A does notallow for the use of probabilistic techniques in the licensing process -- the new rule will. Stepp wasnot convinced, noting that the language of the rule can be made more flexible by simply requiringthe use of state-of-the-art technology; the regulatory guide can then provide the specifics. This way,the rule would allow for the process to evolve if, in the future, new insights become available.According to Murphy, the Staff considered requiring either a probabilistic or deterministic analysisfor the selection of the controlling earthquake but rejected the idea. (He did not explain why.)

An industry representative commented that one of the questions that exists in reading the PDRpackage is whether the deterministic process that is being required is identical to one in the presentAppendix A rule or whether it has evolved. The Staff basically said that it is identical, i.e., theStaff is not changing the deterministic methodology.

Chokshi noted that a report containing four plant sites on which the new siting approach has beenapplied by the Staff will be issued along with the rulemaking package. The Staff will apply theapproach to the other existing nuclear plant sites during the public comment period and these willbe released as they are finished. Ng pointed out that the Staff is only applying the probabilisticside of the new siting approach but he also wanted to know about the deterministic side as well,whether this has changed over the years. He, Stepp, O'Hara and other industry representativesquestioned whether the deterministic approach for selecting seismic sources and dominantearthquake magnitudes from those sources is any clearer today than in the past when each site waslicensed. The Staff indicated that it did not know, prompting Stepp to point out that this is thecrux of the issue and why the industry wants the language of the rule to be made more flexible onthe use of the dual (probabilistic and deterministic) approach. In the industry's view, theprobabilistic approach will benchmark some of the uncertainty that has existed in the deterministicapproach but it will not reduce it. As a result, the legal arguments that have existed in the pastwill recur under the new regulation.

Stepp emphasized that the regulation should allow for the use of the deterministic and/orprobabilistic techniques depending on the tectonic province. It should not specify the requirementthat both be used because this will complicate the technical work that will need to be done and asa result will complicate the licensing process.

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At this point, the Staff went over its schematic of the dual (probabilistic and deterministic) sitingapproach (ref: Attachment 1 of SERCH MS-92-26). According to Murphy, the deterministicanalysis is _ similar to what has been done in the past. O'Hara observed that, unlike theprobabilistic approach, there is a lot of uncertainty in attempting to select the seismic sourcesthrough the deterministic approach. Murphy noted that the Staff recognizes this.

Stepp then questioned the comparison step of the siting approach (just after both analyses havebeen done and their results are to be compared). He wanted to know what NRC would expectsomeone to do if the results from the two analyses are very different. Stepp pointed out that theregulation simply says to use mean values; therefore, what would one need to do in this step?Murphy acknowledged that the guidance does not give much insight on this step. He agreed thatit says that use of mean values is acceptable bu..Jthe also noted that it does not say that other valueswould not be acceptable.

With respect to the dual-analysis approach, Murphy added that if the industry wants to combinethe probabilistic and deterministic techniques into one analysis, it would be acceptable.

Robert Bosnak concluded the meeting by stating that NRC is interested in stability in the regulatoryprocess. As a result, before the package goes back to CRGR, the Staff will take another look atthe rulemaking package relative to the industry's comments.

An 8-page vugraph by the NRC Staff is available upon request from the SERCH Staff.


SERVICE FOR EVALUATING REGULATORY CHANGESi

MEETING: NRC Staff Meeting With NUMARC on Proposed 10 CFR Part 100, Appendix A

DATE: June 17, 1992 APPLICABILITY: ALWR

NUMBER: MS.92-111 KEY SUBJECT: Siting

_C Staff:

Andrew Murphy, Chief, Structural & Seismic Engineering Branch (SSEB), Office of NuclearRegulatory Research (RES)

Nilesh Chokshi, SSEBRoger Kenneally, SSEBRichard McMullen, SSEBB.D. Liaw, Deputy Director, Division of Engineering Technology (DET), Office of Nuclear

Reactor Regulation (NRR)Goutam Bagchi, Chief, Structural & Geosciences Branch (ESGB), DET, NRRPhyllis Sobel, ESGBJohn Craig, Chief, License Renewal Project Directorate (PDLR), NRRRani Watkins, PDLRFrancis Akstulewicz, PDLR

Indust_ Representatives Included:

Ray Ng, NLrMARCJohn Butler, NUMARCCarl Stepp, EPRIMarty McCann, JBAJohn Jacobson, Yankee AtomicDonald Moore, Southern Nuclear ServicesRobert Whorton, South Carolina Electric & GasSam Stone, Tennessee Valley Authority

Industry representatives met with the NRC Staff on June 17 to continue discussion of the proposedPan 100, Appendix A revision. This meeting was a follow-up to the meet/rig held onApril 23, 1992 (see SERCH MS-92-78) where the industry expressed concern that the wording ofthe proposed rule (requiring that siting be based on a dual deterministic and probabilistic analysisapproach) is too inflexible, would not allow for evolving technology to be used in the future, andwould not result in a more stable licensing process. In this meeting, the industry offered an outlineof the siting approach it feels should be used. The industry's approach, which would be reflected

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in a regulatory guide (RG DG-1015), consists of an integrated analysis using probabilistic anddeterministic techniques.

Before the industry presentation began, Murphy commented that the rulemaking package wasforwarded to the Commission on June 12 along with a SECY' requesting approval for issuing thepackage for comment. Ng pointed out that NUMARC has forwarded a letter to Eric Beckjord,Director of RES, requesting that the wording of Appendix B of the rule be adjusted prior toissuance to allow for more flexibility in using probabilistic and deterministic techniques, and askingfor an early release of the package.

NUMARC wants the wording of Appendix B to simply identify the types of investigations anddeterminations that would need to be done. The acceptable methods would be outlined in theregulatory guides. In NUMARC's view, this would provide a robust and stable decision-makingframework for siting.

According to Stepp, the siting decision guidelines in RG DG-1015 should address the uncertaintyin interpreting earthquake phenomena; should provide information that facilitates the evaluationand review of site-specific seismic hazard acceptability by considering all new site-specific data; mustbe generally accepted by the regulated, regulator, and the technical/professional community; mustbe reproducible (the use of a different approach on every site as was done in the past is notacceptable); and should be robust with respect to evolving understanding of earthquake processesand phenomena.

In the industry's view, regulatory decision-making stability can be better achieved with an integrateddecision-making approach for selecting the SSE ground motion. Such an approach would be basedon new site-specific geological, seismological, and geophysical information, and existing seismicsource interpretations. The steps for carrying out this approach would be as follows:

1. Compile site-specific geological, seismological, and geophysical information.

2. Determine whether the existing seismic source interpretations (i.e., the LLNL and EPRIgeneric seismic sources) are robust with respect to new information.

3. Assess the site-specific SSE ground motion using probabilistic seismic hazard analysis(PSHA) procedures, existing sources, and new ground motion attenuation models.

4. Determine the magnitudes and distances at selected ground motion frequencies for a) thecomposite seismic hazard and b) the seismic hazard source by source.

5. Evaluate the derived magnitude and distance information against the new geological,seismological, and geophysical data for the site.

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6. If needed, perform a separate seismic hazard analysis based on the new information toverify that the derived hazard is consistent with the site's PSHA results.

According to Stepp, the industry is currently performing trial applications of this approach byevaluating the SSE ground motion for several existing sites as well as sites located in proximity tohigh seismic regions. The trial evaluations will assess the magnitudes and distances for the totalhazard (i.e., from all the sources in the area) and for each seismic source.

Bagchi initially seemed uncomfortable with the industry's suggestion to eliminate the deterministicside of the dual siting approach. He did not see why the deterministic analysis should be eliminatedsince it has y/elded adequate results in the siting of existing plants. Stepp responded that thedeterministic approach has resulted in the use of a broad range of SSEs from one site to anotherwhich is a reflection of the uncertainty/inconsistent interpretation or limitation of the existingPart 100, Appendix A siting approach. Stepp added that deterministic analysis factors into theindustry's proposed approach, but it is only one small aspect of a more probabilistic-based approach.

Chokshi thought that the industry's approach is consistent with the proposed rule. Kenneallyagreed. However, Murphy pointed out that the industry basically wants NRC to remove the "and"from the rule (i.e., the rule's reference to the use of deterministic _ probabilistic analyses). Hisconcern in doing this is that it would allow an applicant to use an entirely probabilistic approach(i.e., with no deterministic analysis factored in to the results) and this would be unacceptable to theNRC. Stepp suggested precluding this by simply having the rule say that probabilistic anddeterministic methods will be used.

At this point, Bagchi commented that he now appreciates the industry's concern. He pointed outthat the proposed rule is not intended to require the use of two completely parallel paths with acomparison of their results at the end. In his view, the industry's approach (i.e., a comb/nation ofprobabilistic and deterministic analysis) would be acceptable under the proposed rule. However,Stepp pointed out that leaving the wording in the proposed rule as is would allow legal interventionbecause it implies the use of parallel deterministic and probabilistic approaches.

Murphy closed the meeting by noting that, preliminarily, he does not object to the use of industry'ssiting approach; upon further evaluation of the approach, the Staff may endorse it in the regulatoryguide. In response to Murphy's question on when the industry expects to submit the approach inwriting, Ng indicated that the industry will be ready to present the results from its trial applicationson July 10. A formal submittal of the approach would probably be made sometime toward the endof the expected comment period of the proposed rule (i.e., in mid-fall).

A 14-page handout is available upon request from the SERCH Staff.

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