Nuclear Engineering and Design 192 (1999) 351–363

License renewal — an idea whose time has come. Hatchnuclear plant license renewal program: an actual example of

application of the license renewal rule to the IntakeStructure

Farook Chandiwala *, William P. Evans 1

Southern Company Ser6ices, 40 In6erness Center Parkway, Bin B007, Birmingham, AL 35242, USA

Abstract

After the NRC issued a revised license renewal rule in May 1995, the nuclear industry focussed on developinggeneric industry for implementing the rule and testing the guidance through various demonstration programs andwork products in conjunction with the NRC. In addition, plant-specific programs also proceeded forward. Theseactivities show that implementation issues continue to exist. Since the issuance of the rule, the NRC has issued a draftstandard review plan for license renewal (SRP-LR), Working Draft, September 1997. Southern Nuclear OperatingCompany (SNC) has begun development work on a license renewal application for Plant Hatch Units 1 and 2. PlantHatch Units 1 and 2 are BWR 4, Mark I plants whose operating licenses expire in 2014 and 2018, respectively. ThePlant Hatch initiative also involves teaming with other Boiling Water Reactors (BWRs) to develop the license renewaltechnology within the BWR fleet, and to support Plant Hatch by providing an oversight role for the applicationprocess. The teaming effort involved two other utilities, each being assigned to prepare a common report on amechanical system or a structure. The common report could be presented to the NRC with modifications to suit theindividual plants, thereby saving time and money, and hopefully resulting in quicker approval by the NRC. Thedesired license renewal process end result is a renewed license with up to a 20 year extension (10CFR 54.31(b)).© 1999 Elsevier Science S.A. All rights reserved.

www.elsevier.com/locate/nucengdes

1. Introduction

The Hatch License Renewal Program is a long-range program developed by Southern Nuclear

Operating Company (SNC). This program willhelp assure the development of aging managementenhancements in order to ensure the optimumservice life of the Hatch Nuclear Plant. The pro-gram strategies, and participation in teaming rela-tionships, are expected to provide opportunitiesfor early interaction with the NRC. The goals areto resolve, or at least better define, the areas ofuncertainty and to assess their impact on thePlant Hatch program.

* Corresponding author. Tel.: +1-205-992-6904; fax: +1-205-992-0348.

E-mail addresses: fchandiw@southernco.com (F. Chandi-wala), wpevans@southernco.com (W.P. Evans)

1 Tel.: +1-205-992-5663.

0029-5493/99/$ - see front matter © 1999 Elsevier Science S.A. All rights reserved.

PII: S0029 -5493 (99 )00129 -6

F. Chandiwala, W.P. E6ans / Nuclear Engineering and Design 192 (1999) 351–363352

This paper will concentrate on the preparationof the license renewal technical document for theIntake Structure prepared by SNC. The draftreport was circulated to the other two utilities andtheir comments incorporated. The purpose of thisreport is to document compliance of Plant Hatch,Monticello Nuclear Plant, and Peach Bottom Nu-clear Power station Intake Structures with therequirements of 10CFR Part 54.21 (the Rule) insupport of a License Renewal Application. Theprocess for developing this document is based onthe Nuclear Energy Institute’s document NEI 95-10 ‘Guideline to implement 10CFR Part 54, theLicense Renewal Rule’ (CFR, 1996) which isfounded on industry experience and expertise.This report provides an enveloping assessment ofthe three plants where practical. Plant-specific as-sessments of unique features are described sepa-rately. As a result of the technical document, aseparate licensing report will be developed forNRC approval on the Plant Hatch design.

The final rule requires a review of certain plantsystems, structures, and components (SSCs) toultimately determine if the effects of aging areadequately managed in the period of extendedoperation. The Hatch License Renewal programis outlined in the following sections.

1.1. Scoping process

SNC first identified the system/structure func-tions, using the Maintenance Rule system func-tions as a guide. SNC then applied the criteria of10CFR54.4 to each function to determine if thefunction was in the scope of the Rule.

The scoping process evaluated the system/struc-ture on a function-by-function basis against theeight specific License Renewal (LR) criteria(CFR, 1995, 1996). Any system/structure func-tions identified by the specific evaluations wasadded to the database and evaluated against theeight criteria. The eight criteria are:1. safety-related — reactor coolant pressure

boundary (10CFR 54.4(a)(1)(i))2. safety-related — safe shutdown (10 CFR 54.4

(a)(1)(ii))3. safety-related — prevent or mitigate the con-

sequences of accidents (10 CFR 54.4(a)(1)(iii))

4. nonsafety-related that affects safety-relatedfunctions (10 CFR 54.4(a)(2))

5. relied on to demonstrate compliance with 10CFR 50.48 (fire protection)

6. relied on to demonstrate compliance with 10CFR 50.49 (environmental qualification)

7. relied on to demonstrate compliance with 10CFR 50.62 (anticipated transients withoutscram)

8. relied on to demonstrate compliance with 10CFR 50.63 (station blackout)

NOTE: The Commission regulation for pressur-ized thermal shock (10 CFR 50.61) is not ad-dressed because it is not applicable to boilingwater reactors.

Fig. 1 is a scoping form for the Intake Struc-ture. This form includes the eight criteria listedabove, and determines whether the Intake Struc-ture is in scope for License Renewal.

1.2. Preparation of boundary diagrams

To identify clearly which structures, compo-nents and commodities (SSCs) are included withthe in-scope functions, plant drawings weremarked to establish boundaries that match thefunctions on the scoping sheet.

1.3. Screening process

SNC then determined which SSCs were withinthe function boundary and were necessary to ac-complish these in-scope functions. These SSCswere determined to be in-scope of License Re-newal if they were passive and long-lived. If so,the SSC would be evaluated for aging effects,which would have to be managed over the ex-tended license term.

1.4. Integrated plant assessment (IPA)

IPA was used to identify those structures, com-ponents and commodities that require an agingmanagement review (AMR) (the screening pro-cess), identify time limited aging analysis (TLAA)issues, perform AMRs to demonstrate that theeffects of aging will be properly managed, and toevaluate the TLAA issues.

F. Chandiwala, W.P. E6ans / Nuclear Engineering and Design 192 (1999) 351–363 353

Fig. 1. License Renewal Services Procedure Manual LRS 1-3.

F. Chandiwala, W.P. E6ans / Nuclear Engineering and Design 192 (1999) 351–363354

The purpose of the IPA is to provide reasonableassurance that the structures and components willperform their functions consistent with the currentlicensing basis throughout the extended operatingperiod. Very little is said in the rule about accept-able scoping and IPA methods. The rule requiresthat the IPA be documented in a license renewalapplication and submitted to the NRC for review.The Plant Hatch license renewal program is amulti-year program with a goal of having a renewalapplication ready for submittal by the end of 1999.Activities leading to the plant-specific process andprocedural documents are nearly complete. Scop-ing is complete and screening activities are proceed-ing. Completing the IPA and TLAA evaluationsfor the major plant structures and components isplanned for 1999.

2. Teaming initiatives

Three utilities agreed to cooperate in developinga standardized report for submittal to the NRC forthe Standby Liquid Control System (PECO Energyfor Peach Bottom), Recirculation Piping System(Northern States Power for Monticello) and theIntake Structure (Southern Nuclear for PlantHatch). For Monticello and Peach Bottom, infor-mation for the Intake Structure was obtained fromNorthern States Power Company (1992), PECO.

2.1. Reasons for teaming

The following areas of common ground, sharedpoints of view and/or issues of concern providedthe reasons for these utilities to participate in alicense renewal teaming program.1. Each of the utilities has BWR plants2. Teaming provides each utility a measure of

assurance that its programs, procedures, and/or positions are on the right track.

3. The utilities have a common philosophy on theimportant implementation issues.

4. Teaming could provide an opportunity to re-duce costs.

5. Teaming provides opportunities for meaning-ful mentoring peer review support.

6. A combined approach versus a generic ap-

proach could be submitted to facilitate obtain-ing NRC resources.

2.2. Benefits of the teaming effort

� Establishment of a success path (standard) forother BWRs to follow.

� Application of strategies that reduce the poten-tial for public intervention.

� Promotion of wider industry involvement.

3. Description of intake structure

The Intake Structure for each of these plants isan open cycle circulating water system that facili-tates the intake of cooling water to the maincondenser. The Intake Structure is part of theultimate heat sink and is classified as a Class 1structure. The service conditions at this structure,including exposure to water, weather, freezing andthawing, and flowing water, are representative ofthe most challenging conditions to structuraldurability occurring at the plant. The Intake Struc-ture, constructed primarily of reinforced concrete,is separated into two general compartments con-sisting of a screen well and a pump well. Pumpslocated in the pump well include emergency orplant service water (PSW) pumps, residual heatremoval service water (RHRSW) pumps, and fireprotection pumps. Much of the Intake Structure isbelow the water surface to facilitate pump suctionfrom the cooling water source, the Altamaha River.

The intended function of the Intake Structure isto protect equipment for safe plant shutdown fromthe influence of environmental conditions such asflooding, earthquakes, and tornadoes. It supportssafety and non-safety related equipment and sys-tems during all operating modes under all applica-ble loading conditions.

There are some variances between the IntakeStructure designs for plants Hatch, Monticello, andPeach Bottom. These variances are indicated in acompliance report (10 CFR54.21.).

3.1. E6aluation boundaries

Initially, a boundary was defined for the struc-

F. Chandiwala, W.P. E6ans / Nuclear Engineering and Design 192 (1999) 351–363 355

Fig. 2. LRS 1-4 Boundary Description Package for Civil/Mechanical Disciplines.

ture. In the case of civil structures, the typicalboundary of the structure included in the evalua-tion was the entire building including concrete floorslabs, sub-foundations, foundations, external andinternal walls, roofs, curbs, and equipment pads.This includes all rebar, anchor bolts, miscellaneousembedded steel shapes and plates, and grout pads.Structural steel shapes (other than for mechanicaland electrical supports), structural steel spray bar-riers, connections, stairs, ladders, handrails, andgrating are included. Metal doors, frames, andhatches are included in this evaluation boundaryalso. Items such as pipe supports will be evaluatedwith the system they support. Cable trays andelectrical supports were evaluated on a ‘spacesapproach’.

Boundaries for structures, civil, and mechanicalitems were assigned a boundary package number,and used for identification on their respectiveboundary drawings. A Boundary Package Formwas then prepared and reviewed to define theboundary as described in the previous paragraph

(Fig. 2). The boundary for the Intake Structure mayinterface with other boundaries, and this interfaceis included in the Boundary Package Form. Theseforms were developed in-house by SNC.

3.2. Screening

After the boundaries were defined, screening ofthe components was performed. The structuralcomponents, such as concrete, masonry, steel, andother metals, if any, were input in the screeningforms, along with the internal and/or externalenvironments to which these components are sub-jected. Since most of the civil items are passive andlong-lived, all of these were included. The formsshown in Figs. 3–5 show the results of the screeningprocess.

3.3. Aging management

The screening process in Section 3.2 identifies theplausible and non-plausible aging mechanisms,

F. Chandiwala, W.P. E6ans / Nuclear Engineering and Design 192 (1999) 351–363356

which are to be addressed in the next step. Forthe teaming effort, a table was prepared for allthree plants. This table (a sample page shown inTable 1) was divided into three sections; concrete,steel, and miscellaneous. The aging mechanismdescription, along with the governing parameters,taken from (EPRI, 1994) were listed in this table.Information from NUREG (1996) was also uti-lized. These were used as a yardstick for determin-ing the plausibility of an actual effect of theIntake Structure.

The table identifies the aging mechanisms forthe components and commodity groupings of theIntake Structure. Methods to prevent detrimentaleffects by plausible aging mechanisms are de-

scribed for each plant. The table makes a com-parison of the aging effects between the threeplants to demonstrate the consistency oruniqueness.

3.4. Aging effects

The only aging effect found to be common toall three plants was corrosion of steel. However,Plant Hatch also demonstrated corrosion of em-bedded steel in concrete as a plausible effect.Cracking of masonry block walls was evaluatedonly for Peach Bottom and Monticello plants.Also, scouring only was a credible effect forMonticello.

Fig. 3. LRS 1-5 Screening Evaluation Record.

F.

Chandiw

ala,W

.P.

E6ans

/N

uclearE

ngineeringand

Design

192(1999)

351–

363357

Table 1Sample sheet from teaming effort for aging management: component and component groupings subject to aging management review (teaming effort-aging mechanisms andmanagement-intake structure)

CONCRETE

Actual MonticelloActual hatch parameters Actual Peach BottomAging effects Governing parametersAging mechanism descrip-parameters (Ref. 8a)tion per Ref. 1a parameters (Ref. 9a)(Ref. 2a, U.N.O.)

Freeze–thaw damage (scaling, To avoid effects provideSec. 4.1.1 N N Nair content between 3 to Hatch Plant is located inNb Conc. Contains appropriatecracking and spelling) in: Air content 4.1 to

‘negligible’ region, hence6%; Weathering indexB 4.9%; Water/cement amount of entrained air (3–Foundations100 day-inches/year for ratio between 0.50 andfreeze–thaw is not a con- 6%) necessary for freeze–Exterior concrete below

0.56‘negligible’ region cern thaw resistance,gradesufficient cement and re-Sec. 4.2.1.1 (1) Sec. 4.1.1Exterior concrete abovequired water–cement ratiograde

Interior concrete slabs to reduce permeabilityItem 1, page 4

Sec. 4.1.3 Effects realized when: Ext. conc. above gradeAggressive chemical action N(increased porosity, crak- — N

P Chlorides\500 ppm, Ground water aggressive Chlorides=17.9 ppm Ning, spalling, and losschemical action — N Chloride=13.7–21.5 ppmstrength) in:Chlorides=10 ppm Sulfates=58.5 ppmSulfates\1500 ppm,Foundations

and/or pH B5.5 Sulfates=20 ppm pH 8.0Exterior concrete below Sulfates=10.3–41 ppmSec. 4.1.1pH 6.6 pH 7.2–7.6grade Ref. 5, Table 2.1–10 Item 3, page 6Exterior concrete above

gradeInterior concrete slabsMasonry block walls

Effects realized when:Sec. 4.1.10 N P N (See narrative)Scouring or abrasion (re-Continuously exposed Max velocity=10.7 fps inMax. velocity=5 55 fpsP Conc/struct. exposed tomoved cement and aggre-to flowing water HPSW piping. Velocities inconstinously flowing watergate) in:

the larger open channeltransporting materialsExterior concreteSec. 4.1.1, pages 4–9Ref. 5, Table 3.8–12 pump bays will be signifi-which abrade the con-

cantly less than the pipecrete.flow velocity noted above.Cavitation damage occursVelocities are well below 40with velocity \40 fps orfps and 25 fps thresholds\25 fps in closed con-for cavitation damage Itemduits6, page 8

Aging managementc: Periodic visual inspec- None required, however reg-None required; however,Scouring or abrasion tion of intake structurevisual Inspection done by ular inspections performed

bays and intake canal every refueling outage (24Hatch Structural Monitor-months) Item 6, page 9retaing walls per-ing Program not to exceed

formed per RG 1.1275 yearsas indicated in Sec. 6.2

a References: 1 Class I Structures License Renewal Industry Report, Rev. 1, EPRI TR-103842. 2 Integrated Plant Assessment of Hatch Units 1 and 2, Intake Structure LRDemonstration, May 29, 1996. 3 Life Cycle Management License Renewal Program, Implementation Guide, Aging Assessment of Plant Structures, Westinghouse Owners Group(WOO). 4 Life Cycle Management License Renewal Program, Aging Assessment Process for Structures and Components, Westinghouse Owners Group (WOO). 5 Hatch FSAR,Unit 2. 6 Hatch FSAR, Unit 1. 7 General Concrete Specifications for Hatch Units 1 and 2, PDCR-PSOC-54-001-SPEC 1. 8 Aging Management Review, Peach Bottom SafetyRelated Pump Structure, Units 2 and 3. 9 Monticello Nuclear Generating Plant, An Evaluation of the Civil Structures important to License Renewal, Rev 0, April 1992. 10 ConcreteStructures Aging Reference Manual, Project no. 5472-01, COSTAR Project Expert Panel Review (Later).

b P, plausible; N, not plausible; NC, not considered; NA, not applicable.c Aging management for plausible mechanisms is defined immediately after the mechanism, and occurs in bold type.

F. Chandiwala, W.P. E6ans / Nuclear Engineering and Design 192 (1999) 351–363358

Fig. 4. LRS 1-5 Screening Evaluation Record.

3.4.1. Loss of material due to scouring, abrasionand ca6itation ( for Monticello only)1. This degradation mechanism requires flowing

water.2. Abrasion and cavitation are potentially signifi-

cant degradation mechanisms for the pumpstructure concrete below the water line.

3.4.2. Loss of material due to corrosion1. Corrosion of reinforcing steel could result

from the intrusion of aggressive ions or acids,leaching of alkaline products through cracks,or carbonation.

2. Corrosion of embedded steel could result fromcontinued exposure to water or high humidity

in the presence of oxygen.3. Corrosion of reinforcing and embedded steel

can affect the foundation, exterior concretebelow grade, exterior concrete above grade,interior concrete slabs, and the support ofsafety related equipment.

4. Design features, such as concrete mix design,control of curing depth of cover over thereinforcement, minimum reinforcement to con-trol cracking, and protective coatings affectthe potential for reinforcing and embeddedsteel corrosion.

Hence, for the Intake Structure, corrosion is acredible aging effect. However, no corrosion hasbeen observed in reinforcing steel.

F. Chandiwala, W.P. E6ans / Nuclear Engineering and Design 192 (1999) 351–363 359

3.4.3. Deformation and cracking due to settlement1. Settlement occurs mainly during construction

and in the first several months afterconstruction.

2. Except for those structures founded on exist-ing bedrock, the possibility of changes in siteconditions might affect settlement.

3. Settlement curves for Plant Hatch have flat-tened out since 1988, and settlement is not acredible aging effect. The building settlementprogram (Georgia Power Company, 1992) willbe kept in place due to a commitment to theNRC.

3.4.4. Cracking of masonry block walls ( forMonteicello and Peach Bottom plants only)1. Cracking of masonry block walls can be

caused by shrinkage, loading effects, or deflec-tion of supporting structures (e.g. supportingfloors or foundation).

2. Shrinkage and cracking of masonry blockwalls is generally small and occurs at blockjoints in the earlier stages of plant operation.

3. Excessive cracking can reduce the structuralstrength of the wall and corrosion of embed-ded or reinforcing steel may occur as a resultof ingress of moisture and oxygen via cracks.

Fig. 5. LRS 1-5 Screening Evaluation Record.

F. Chandiwala, W.P. E6ans / Nuclear Engineering and Design 192 (1999) 351–363360

Thus, cracking of masonry block walls is acredible aging effect for Monticello and PeachBottom plants.

4. Aging management programs for renewal

Several in-house programs are conducted byeach utility to address the aging managementfor items that require it in order to satisfy theLicense Renewal rule.

4.1. Structural monitoring program (SMP) (SNC,1997, 1998)

1. SMP per the Maintenance Rule (10CFR50.65) is a specific program using qualifiedcivil engineers to perform actual visual walk-downs of plant structures, including the Intakestructure.

2. The goal of the program is to demonstratethat acceptable performance is being main-tained and that the identified structure or com-ponents remain capable of performing theirintended function.

3. If degraded and/or age-related conditions areidentified for a particular structure or compo-nent, the condition is further evaluated andadditional inspections and testing will be im-plemented as appropriate to rectify the de-graded or age-related condition.

4. These walkdowns are performed at regularintervals. These intervals may vary accordingto the structure or component, dependingupon the extent of observed degradation.

5. Per the SMP, buried and/or underwater com-modities are visuallly inspected during oppor-tune excavations (such as during maintenanceor new equipment installation), dive team in-spections, or during low water levels (such ashistorical low river occurrences in thesummer).

6. Historically collected data is used as a com-parison with current conditions to determinewhether aging effects are effectively managedand if corrective measures need to be taken.

4.2. Cracking of masonry block walls (Monticelloand Peach Bottom plants only)

1. It is unlikely that cracking would initiate orpropagate subsequent to the first five yearsafter installation. Therefore, an appropriateaction is to inspect a sample of the masonrywalls prior to the license renewal term. Ifcracks are noted that have structural signifi-cance, further inspections will follow. Anyidentified cracks will be repaired or analyzedfor acceptance.

5. Demonstration that the effects of aging areadequately managed

Scouring, abrasions, cavitation, corrosion ofembedded steel, and cracking of masonry blockwalls have been identified as potentially signifi-cant aging effects for the Intake Structure. Thissection demonstrates that the Structural Moni-toring Program will appropriately manage theseaging effects, such that there is reasonable assur-ance that the Intake Structure’s intended func-tion will be maintained consistent with thecurrent licensing basis (CLB) during the periodof extended operation.

5.1. Corrosion

The effects of corrosion of embedded steel aremanaged using on-site inspections. The condi-tions of concrete surfaces are examined to evalu-ate for deterioration and continued serviceabilityof the concrete. The concrete structures are ex-amined for structural cracking resulting fromoverstress due to applied loads, shrinkage, andtemperature effects. Water passages and under-water concrete surfaces are visually inspected forsilt, obstructions, debris, and surface condition(including erosion, rust stains or significantstructural cracks). The examination and on-siteinspections are performed periodically throughthe Structural Monitoring Program. For PlantsHatch and Monticello, these inspections con-form to Regulatory Guide 1.127. These activitiesprovide reasonable assurance that the Intake

F. Chandiwala, W.P. E6ans / Nuclear Engineering and Design 192 (1999) 351–363 361

Structure is able to perform its intended functionconsistent with the CLB during the period ofextended operation.

5.2. Implementation

The frequency and attributes of the activities ofStructural Monitoring Program ensure early detec-tion of degraded performance and mitigation bytimely corrective action to ensure that the IntakeStructure can perform its intended function. Theprogram activities will appropriately manage theeffects of aggressive chemical attack, and corro-sion, such that there is reasonable assurance thatthe Intake Structure’s intended function will bemaintained consistent with the CLB during theperiod of extended operation.

Potentially significant aging effects identifiedabove and programs used to manage their effectsare shown in the table below:

Aging man-Aging effects Frequencyofagementinspectionsprogram

Not credibleStructuralScouring, abra-Monitoring at Hatchsions, andProgramcaviation

Corrosion of re- Every out-StructuralMonitoring age cycle (@inforcing and

Hatch)Program, andembeddedsteel also in accor-

dance withRegulatoryGuide 1.127

Cracking of ma- Before startStructuralMonitoring of license re-sonry blockProgram (forwalls newal pro-

gram (NotMonticelloand Peach applicable to

Hatch)Bottom only)

6. License renewal walkdowns

To obtain extra assurance of the acceptability of

the Structural Monitoring Program as an AgingManagement Program, a walkdown of the IntakeStructure was performed by two civil engineers.This walkdown was performed per the checklistshown on Table 2. A few cases of rusted supportbase plates and bolts, and missing bolts wereobserved. These were listed and transmitted to thePlant Hatch site. It is anticapted that Monticelloand Peach Bottom will take similar action. TheIntake Structure was walked down by three South-ern Company Services (SCS) engineers and anengineer from LCM Engineering, PLLC, who isassisting in the preparation of the License RenewalApplication to the NRC. Besides the Intake Struc-ture, several other structures and components werealso walked down.

7. Time limited aging analysis (TLAA)

In order to be included in the license renewalprogram, a structure, system, or component hasto be passive and long-lived. A passivecomponent is defined as one with no change instate or moving parts. For example, all valves(except valve bodies) are active. To assure thesecond part (i.e. long-lived), design calculationswere reviewed to ascertain that they did notindicate a replacement life of less than fortyyears. Approximately 3000 civil calculationswere reviewed for the whole plant, and onlyseven civil calculations needed further review(none for the Intake Structure). The majority ofthe calculations that needed further review werein the mechanical areas, such as piping thermalfatigue stress analyses.

8. Conclusion

The information from several walkdownreports, industry reports, and use of the SNCinhouse screening criteria mentioned above,indicates that the structures and commoditiesare generally in good condition. Plant Hatchwill be notified of items discovered to haveexhibited substantial deterioration.

F. Chandiwala, W.P. E6ans / Nuclear Engineering and Design 192 (1999) 351–363362

Table 2Walkdown checklist

Checklist unit cArea

PerformanceEnvironment Credit programs/procedureshistory

Humidity Temperature Traffic Radiation Dynamic or cycli-

cal loading Ultra–violet expo-

sure Wind

Walls Columns CeilingCONCRETE BeamsSlab/floors Equip. Pads InsertsSpallingCrackingRebar corrosionSurface pittingExposure of aggre-

gateMisalignmentScalingStaining

Beams Embeds PlatesColumns Nuts Bolts. etc.STEELCorrosionRough SurfaceLoose materialFoul OdorStained SurfaceSludge accumul.Metal BulgingGaps underbaseplatesDeformed members

DEGRADATION OF PROTECTIVE COATINGSCracking Softened Coat-

ing Disbondment or Exposed Sub-

stratepeelingFoul OdorBlistering

Corrosion penetrat- Discoloration ing Coating

POLYMER APPLICATIONS IN PLANT STRUCTUREScaulking, concrete crack repair, seal. adhesiveButyl rubber

Silicone Rubber caulking, tank seals, gaskets water sealsO-rings, vibration dampersNeopreneO-rings, formed seals, gasketsHypalon

Expanded polypropylene Expansion joint sealssealant, caulking, adhesivesVinyl polymergaskets, seals. vibration isolation mountsEthylene propylene rubber

F. Chandiwala, W.P. E6ans / Nuclear Engineering and Design 192 (1999) 351–363 363

Acknowledgements

The authors wish to thank the authors of thepaper in Pierce et al. (1998) for permission to usesome of the material in this presentation.

References

EPRI, 1994. Class I Structures License Renewal IndustryReport, EPRI TR-103842, Revision 1, Electric Power Re-search Institute, Palo Alto, CA, July 1994.

CFR, 1995. Code of Federal Regulations, 10CFR Part 54,Requirements of Renewal of Operating Licenses for Nu-clear Power Plants, Office of the Federal Register, Na-tional Archives and Records Administration, USGovernment Printing Office, Washington DC, May 1995.

CFR, 1996. Industry Guideline for Implementing the Require-ments of 10CFR Part 54, The License Renewal Rule, NEI95-10, Revision 0, Nuclear Energy Institute, WashingtonDC, March 1996.

NUREG, 1996. Report on Aging of Nuclear Power PlantReinforced Concrete Structures, NUREG/CR-6424(Draft), Oak Ridge National Laboratory, Oak Ridge, TN,February 1996.

SNC, 1998. Structural Monitoring Program for the Mainte-

nance Rule, Rev. 4, for Plants Hatch, Farley and Vogtle,prepared for Southern Nuclear Company by SouthernCompany Services, Birmingham AL, June 1998.

Georgia Power Company, 1992. Hatch Procedure 01-RC-SUV-001-S, Monitoring Settlement of Structures, Rev. 0,prepared by Georgia Power Company, Atlanta, GA, forPlant E.I. Hatch, March 1992.

SNC, 1997. Hatch Procedure 52PM-MME-006-0S, Rev. 6,Intake Structure Pit Inspection/Diving Activities, preparedby Southern Nuclear Company, Birmingham, AL for PlantE.I. Hatch, June 1997.

Demonstration of Compliance with the Technical Require-ments of the License Renewal Rule (10CFR54.21) (Team-ing effort report for Hatch, Peach Bottom and MonticelloIntake Structures, under preparation by Southern Com-pany Services, Birmingham, AL-Hatch Project).

Pierce, C.R., Hornbuckle, J., Lenhert, D.F., Plant HatchLicense Renewal Program, a technical paper presented atthe Sixth International Conference on Nuclear Engineer-ing, ICONE-6, (excerpts reproduced with permission of theauthors), May 1998.

Monticello Nuclear Generating Plant-An evaluation of thecivil structures important to License Renewal, Rev. 0,Northern States Power Company, Minneapolis, MN, April1992.

Structural Topical Design Baseline Document for Peach Bot-tom Atomic Power Station, Units 2 and 3, P-T-01, Rev. 4,PECO Energy Company, Wayne, PA (in preparation).

.