8118

Nuclear Powerplant Standardization: LightWater Reactors

April 1981

NTIS order #PB81-213589

Library of Congress Catalog Card Number 81-600061

For sale by the Superintendent of Documents,U.S. Government Printing Office, Washington, D.C. 20402

Foreword

This assessment responds to a request by the House Committee on Interior andInsular Affairs and endorsed by the Senate Subcommittee on Nuclear Regulation toevaluate the extent to which nuclear powerplants can and should be standardized.The assessment provides the essential background material for a broad understand-ing of the nuclear industry, its institutions and their relationship to standardization.Items presented in the report include specific examples of the current state ofnuclear powerplant standardization, and four different concepts of standardizationand their potential impact on safety. These concepts represent a wide range of ap-proaches toward standardization and would entail greatly differing requirementsfor industry and regulators as well as differing implications for safety.

We are indebted to the participants in the workshop, reviewers of the final re-port, and numerous other individuals who gave extensively of their time and talentsin support of this assessment. Also, the contributions of several contractors, whoperformed background analyses, are gratefully acknowledged.

JOHN H. GIBBONSDirector

Reviewers of the Nuclear Power plantStandardization Report

Robert CiviakCongressional Research Service

L. C. DailDuke Power Co.

Peter R. DavisIntermountain Technologies, Inc.

Robert L. FergusonWashington Public Power Supply System

James F. MallayElectric Power Research Institute

David OkrentUniversity of California, Los Angeles

Oswald F. SchuetteUniversity of South Carolina

G. Thomas SeelyFluor Power Services

Dee H. WalkerOffshore Power Systems

Abraham WietzbergNUS Corp.

E. P. WilkinsonInstitute of Nuclear Power Operations

Workshop on Nuclear Powerplant Standardization

Harold W. Lewis, ChairmanUniversity of California

Myer BenderOak Ridge National Laboratory

Sidney A. BernsenBechtel Power Corp.

Dale G. BridenbaughMH B Technical Associates

Thomas CoxU.S. Nuclear Regulatory Commission

Jesse C. EbersoleConsultant

Darrell G. EisenhutU.S. Nuclear Regulatory Commission

Jerry GriffithU.S. Department of Energy

Joseph M. HendrieU.S. Nuclear Regulatory Commission

Stan JacobsStone & Webster Engineering Corp.

Edward O’DonnellEbasco Services, Inc.

John RaulstonTennessee Valley Authority

Don RoyBabcock & Wilcox Co.

A. Edward SchererCombustion Engineering, Inc.

Glenn SherwoodNuclear Energy Systems DivisionGeneral Electric Co.

William R. SpezialettiWestinghouse Electric Corp.

Wayne StiedeCommonwealth Edison

NOTE The reviewers and workshop participants provided advice and comment throughout the assessment,but they do not necessarily approve, dis-

approve, or endorse the report for which OTA assumes fuII responsibility

iv

Nuclear Powerplant Standardization Project Staff

Lionel S. Johns, Assistant Director, O T AEnergy, Materials, and International Security Division

Richard E. Rowberg, Energy Program Manager

Edward C. Abbott, * Project Director

Alan T. Crane, Program Coordinator

Administrative Staff

Marian Grochowski Lillian Quigg Edna Saunders

Supplements to the Staff

William Metz Dean Eckhoff

Consultant

NUS Corp., Rockville,

Barbara Levi

Md.

OTA Publishing Staff

John C. Holmes, Publishing Officer

John Bergling** Kathie S. Boss Debra M. Datcher

Patricia A. Dyson** Mary Harvey* * Joe Henson

Senior Fellow, Advisory Committee on Reactor Safeguards, Nuclear Regulatory Commission‘ ‘OTA cont rac t personal

v

ContentsChapter Page

Glossary . . . . . . . . . . . . . . . . . . . . . . . . . viiiAcronyms and Abbreviations . . . . . . . . . . ix

1. Summary . . . . . . . . . . . . . . . . . . . . . . . . 3Principal Findings. . . . . . . . . . . . . . . . . . 4Institutional Responses . . . . . . . . . . . . . 5

Current NucIearlndustry. . . . . . . . . . 5Nuclear Regulatory Commission . . . . 6Congressional Role. . . . . . . . . . . . . . . 6

2. Introduction ...,.... . . . . . . . . . . . . . . 11

3. The Nuclear Industry Today . . . . . . . . . 15Light Water Reactors. . . . . . . . . . . . . . . 15

Fission Rate Control . . . . . . . . . . . . . . 17Fission Product Containment. . . . . . . 19Auxiliary Feedwater Systems . . . . . . 20Decay Heat Removal . . . . . . . . . . . . 22Control Room Design. . . . . . . . . . 23Causes for Variations Among LWRS. 23

The Nuclear Power Industry. . . . . . . . 24Vendors . . . . . . . . . . . . . . . . . . . . . . 25Architect Engineering Firms. . . . . . . 25Construct ion Companies . . . . . . . . . . 26Industry Trends . . . . . . . . . . . . . . 28

4. The Nuclear Regulatory Commission’sRole . . . . . . . . . . . . . . . . . . . . . . . . . . 33

NRC’ S Current Standardization Program 33Reference Plant Concept. . . . . . . . . . 33Duplicate Plant Concept . . . . . . . . . . 34Manufacturing License Concept . . . . 34Replicate Plant Concept . . . . . . . . . . 35

Experience With the NRCStandardization Program. . . . . . . . . . 36

Current Status of Licensing . . 39NRC’s Future Role... . . . . . . . . . . . . . . 39

5. The Nuclear Industry’s ExperienceWith Standardization. . . . . . . . . . . . . 43

The Naval Reactor Program. . . . . 43The Standardized Nuclear Powerplant

System . . . . . . . . . . . . . . . . . 44The French Nuclear Program . . . . . . 46

Chapter Page

6.

The West German Operator TrainingProgram . . . . . . . . . . . . . . . . . . . . . . . 49

Policy Impacts of Four Approaches toStandardization . . . . . . . . . . . . . . . . . 53

Four Approaches . . . . . . . . . . . . . . . . . . 53Safety Benefits . . . . . . . . . . . . . . . . . . . . 54

Improved Training for PlantPersonnel . . . . . . . . . . . . . . . . . . . . 57

Relevance to a National Safety Goal. . . 58The Impact of Standardization on

Resolution of Generic Issues . . . . . . . 59Standardization and Antitrust . . . . . . . . 61Utilities and Standardization. . . . . . . . . 62Feasibility . . . . . . . . . . . . . . . . . . . . . . . 62

TablesTable No. Page1.2.3.

4

5

6

Auxiliary Feedwater Systems . . . . . . . . 21Nuclear Reactor Suppliers . . . . . . . . 25Content of an NSSS Standard DesignApplication. . . . . . . . . . . . . . . . . . . . 27Architect Engineering Responsibilityfor Nuclear Powerplants. . . . . . . . . . . . . 28NSSS/AE Combination of Light WaterReactors Under Construction orOn Order. . . . . . . . . . . . . . . . . . . . . . . . . 29Unresolved Safety Issues . . . . . . . . . 60

FiguresFigure No. Page1.

2.3.45

6

Boiling Water Reactor Core and VesselAssembly . . . . . . . . . . . . . . . . . . . . . . . . . 15Fuel Bundles . . . . . . . . . . . . . . . . . . . . . . 16Boiling Water Reactor . . . . . . . . . . . . . . 18Pressurized Water Reactor . . . . . . . . . . . 18Comparisons of Auxiiary FeedwaterSystem Reliability on the Loss of MainFeedwater System . . . . . . . . . . . . . . . . . . 22Training Patterns for a West GermanReactor Operator . . . . . . . . . . . . . . . . . . 50

vii

Glossry

Architect engineer.– A supplier of design and engi-neering services for construction projects (e.g.,powerplants, office buildings, bridges, etc.).

Auxiliary feedwater. –A standby system used to sup-ply the secondary (nonradioactive) side ofPWR’S steam generation with cooling water inthe event the main source of water fails.

Balance of plant.– The equipment, in addition to thenuclear steam supply system, which is neces-sary to produce electricity from a nuclear pow-erplant.

Boiling water reactor. -A power reactor in whichwater, used as a coolant and moduator, is al-lowed to boil in the core.

Control rod. –A rod or tube containing a materialthat readily absorbs neutrons used to controlthe power of a nuclear reactor.

Decay heat.– The heat produced by the decay of ra-dioactive nuclides or fission fragments.

Fission. —The splitting of a heavy nucleus into twoapproximately equal parts, accompanied by therelease of a relatively large amount of energyand one or more neutrons.

Heat sink.– Anything that absorbs heat; usually partof the environment such as a river, pond, or theatmosphere.

Light water reactor.– A reactor which uses ordinarywater as opposed to heavy water as a moder-ator and/or coolant.

Megawatt. -A unit of energy production or con-sumption commonly used to describe the gen-erating capacity of a powerplant.

Moderator.—A material such as water used in areactor to slow down high-velocity neutrons.

Nuclear steam supply system.–An arrangement ofequipment with a critical array of nuclear fuelwhich creates high-quality steam for runningturbine generators.

Pressurized water reactors. -A power reactor inwhich heat is transferred from the core to a heatexchanger by water kept under high-pressure toachieve high temperature without boiling,

Probabilistic risk assessment. –An approach to safe-ty analysis which assesses undesirable conse-quences and their likelihood.

Radioactivity. —The spontaneous decay or disinte-gration of a unstable atomic nucleus accompa-n ied by the emission of ionizing radiation.

Reactor. -A device in which a fission-chain reactorcan be initiated, maintained, and controlled.

Safety goal.– A quantitative or qualitative target foreither reliability or unreliability (risk).

System.– An arrangement of equipment utilized ina powerplant for a specific function (e. g., the re-actor protective system).

Vendor.– The supplier of the design and much ofthe equipment for the nuclear steam supplysystem.

Vlll

Acronyms and Abbreviations

ACRS

AEAECA F WASME

ANSIASLBBOPBWRCPDBAEPAEPRIFDAFSARGEI N P OLE RLWRM W e

— Advisory Committee on ReactorsSafeguards

— architect engineer— Atomic Energy Commission— auxiliary feedwater system– American Society of Mechanical

Engineers— American National Standards Institute– Atomic Safety Licensing Board– balance of plant— boiling water reactor— construction permit– design basis accident– Environmental Protection Agency– Electric Power Research Institute– final design approval– final safety analysis report– General Electric Co,— Institute of Nuclear Power Operations– licensee event report— I ight water reactor— megawatts electric

M W tNRCNSACNSSSOLOPSO T APDAPDDAPRAPSARPwcPWRRSSSARSDASIP

— megawatts thermal— Nuclear Regulatory Commission— Nuclear Safety Analysis Center— nuclear steam supply system— operating Iicenses– Offshore Power Systems– Office of Technology Assessment– preliminary design approval— preliminary duplicate design approval— probabilistic risk assessment– preliminary safety analysis report— power-worthiness certificate— pressurized water reactor— reactors safety study— safety analysis report— standard design approval— standard information package

SNUPPS – standardized nuclear unit powerplantsystem

TM I – Three Mile IslandT V A – Tennessee Valley Authority

ix

Chapter 1

SUMMARY

Chapter 1

SUMMARY

After 25 years of commercial development,nuclear power has entered a period of transi-tion. The results of the accident at Three MileIsland (TMI) have introduced sufficient uncer-tainties into the industry’s licensing and safetypractices so that it makes it difficult, if not im-possible, to get a new plant approved. At thesame time, the unexpectedly low-growth ratethat many utilities are encountering has de-terred them from ordering any new nuclearplants for the immediate future. However,even zero growth of demand would requiresome new replacement facilities by the early1990’s to maintain the present generating ca-pacity. If the uncertainties resulting from TM Iare resolved soon, the nuclear industry willhave a unique opportunity to reevaluate itsdirection and practices.

One of the peculiarities of the way that theindustry has developed is that commercial re-actors are built with an unusual degree of vari-ability and diversity. Essentially every reactor,with a few exceptions to date, has been cus-tom-designed and custom-built. The fact thatalmost every reactor is “one-of-a-kind” has ledto excessive difficulty in verifying the safety ofindividual plants and identifying particularproblems in transferring the safety lessonsfrom one reactor to another. It may also ac-count for the escalating costs and long lead-times associated with nuclear powerplants.

Many of these problems can be alleviated ifthe industry moves away from its “one-of-a-kind” practices toward a degree of standardi-zation in its design, construction, operation,and licensing practices. Several types of stand-ardization are possible, and this report exam-ines them. Some trends in this direction are al-ready occurring; the present lull could be usedto lay the groundwork for future standardiza-tion.

A minimal level of standardization is theadoption of criteria for performance, reliabil-

ity, and general design principles. This type ofstandardization is promoted by groups such asthe American National Standards Institute andthe American Society of Mechanical Engi-neers. At the other extreme, some fee I stand-ardization means the selection of one com-plete nuclear reactor as the “standard” ormodel, according to which al I other reactorsare to be built.

OTA evaluated four different approaches tostandardization of the present generation oflight water reactors (LWRS). These are:

●

●

●

The acceleration of present trends. Th iswou Id entail revitalizing and streamliningthe Nuclear Regulatory Commiss ion’s(NRC) current standardization programand emphasizing one-step I icensing.

The procedura l s tandard i zat ion . T h i smeans the use of universal “softwarepractices” such as common terminologyand format for plant procedures and simi-lar requirements for the training of plantpersonnel.

The standardization of the powerplant’snuclear systern and those systerns criti-cally necessary for the safe shutdown ofthe reactor— the safety-block concept.This might include the development ofsimilar designs for auxiliary feedwater andshutdown cooling systems.

The selection of a single standardized de-sign resulting from a fresh approach inte-grating the past 25 years of operating expe-rience from various reactors.

This report considers these four representa-tive approaches to standardization and ex-amines the major advantages and disadvan-tages of each concept.

4 ● Nljc/ear powerp/ant standardization

PRINCIPAL FINDINGS

Standardization can be an essential elementin maintaining a viable and safe program fornuclear energy. There are relatively few plantsbuilt as examples of the approaches to stand-ardization considered in this report, but thepresent trend in the nuclear industry is towardgreater standardization.

Standardization yields safety benefits thatare intuitive/y valid even if they cannot be dem-onstrated unambiguous/y. The common-sensenature of this benefit and its widespread ac-ceptance in the nuclear industry more thancounterbalance the paucity of data from thefew relevant examples. However, the extreme,“single-design” approach to standardizationcould pose so many institutional difficultiesand generic risks, that the problems would out-weigh the safety benefits.

Standardization has c/ear potential for timeand cost reductions and for gains in safety fornew nuclear p/ants. Several utilities and utilitygroups have attempted to build standardplants in the hopes of shorter licensing timeand reduced design and construction costs.Some improvements have been reported butthere have also been problems.

Standardization is not a panacea, and theother elements needed for a safe and efficientnuclear program should not be ignored. Otherelements include safe operating practices, pro-grams for effective preventive maintenance,and direction by responsible technical manag-ers.

Standardized plants constructed during dif-ferent time periods have diverged from theiroriginal design due to the changing regulatoryrequirements, industrial standards, and utilities’preferences. The characteristics of differentsites have dictated further divergence fromoriginal standardized designs.

The quality of the implementation of stand-ardization is just as important as the conceptitself in reaping potential benefits. A customplant can be safer than a standard plant if it isoperated and maintained in an exemplary fash-ion. Conversely, a standard plant will be safer

only if the designers and operators are highlymotivated, talented, and technically compe-tent.

The present trends of the industry towardgreater standardization wil l be great/y en-couraged by the implementation of sing/e-stagelicensing. Proposals have already been madefor the one-step issuance of a standard designapproval or “power-worthiness certificate” fornuclear plants, but they have not been imple-mented.

NRC is current/y devoting little time to theproblem of nuclear powerplant standardiza-tion. The implementation of the rules and re-quirements resulting from the accident at TM Iis occupying much of NRC’s time. If standardi-zation is to succeed at all, NRC must startplanning for it now during this period of slackgrowth in nuclear power. They must developplans for future standardization, includingpossible implementation of one-step licensing.In addition, the vendors should realize that do-mestic orders for nuclear steam supply systems(NSSS) may not occur over the next few years,and they should take this opportunity to re-view and improve their basic designs.

The adoption of a national safety goal isdesirable. This would be a stated goal, agreedon by society through some institution — Con-gress, NRC — as the level of safety acceptableto the Nation. As such, it goes beyond themore general statement in the present law. Theadequacy of NRC’s response to the accident atTMI, in the absence of such a definition (i.e.,how safe is safe enough), is impossible to as-sess and creates a large uncertainty in the li-censing process. NRC must begin to manage itsactivities in a manner so that prompt and con-sistent decisions on safety issues can be made.Participants in the nuclear industry agree inprinciple, on the desirability of a safety goal.

Enhanced standardization increases the like-lihood of accurate risk assessment. The onlymeans to assure that a nuclear powerplant hasachieved a quantitative safety goal is through

Ch. l—Summary ● 5

the use of probabilistic risk assessment. lm-proved risk assessment under standardizationis primarily due to the increased attention thatcan be given to a few well-defined assessmentsrather than many diversified ones.

The safety benefits of improved procedures,through adoption of uniform reporting prac-tices and industrywide participation in reviewof operating experience, can easily be obtainednow. Substantial benefits can also be obtainedthrough standardization of training of plantpersonnel, even when considering the utility’sresponsibiIity for a diversity of plant types.

The four approaches to standardization arenot necessarily mutualy exclusive and mightbe explored in paral le l . The f i r s t two ap-proaches– acceleration of present trends andprocedural standard i z at ion — are alreadybeing pursued but could be further en-couraged. They can be accomplished with lit-tle, if any, disruption of the present structureof the industry.

The second two approaches – the unificationof “safety -block” systems and the adoption of asingle “standard” plant design — could bringabout significant and perhaps disruptivechanges in the institutions of the commercialnuclear industry, The safety-block approachwould transfer design responsibilities for cer-tain safety systems — e.g., the containment — to

a section of the industry not traditionallyresponsible for such systems. The single-stand-ard plant approach would reduce the two ma-jor participants in the industry—vendors andarchitect-engineers (A Es) — to suppliers of com-ponents and engineering services for the singlenational design.

The second two approaches could establishmore specific design criteria than currentlyexist and provide an “idea/” case for measuringfuture design criteria. The purpose would beserved whether or not the more standardizedplant design was actually implemented.

The U.S. Navy’s experience with standardiza-tion is not directly applicable to the commer-cial nuclear power industry. The naval reactorsprogram is the only U.S. example of a well-standardized program with considerable oper-ating experience, but the principles applied inthis program are not directly applicable to thecommercial industry, which has a diversity ofdesigners, AEs, and operators who functionmuch more independently than the partici-pants in the Navy program. The Navy’s safetyrecord is apparently due to strong central con-trol and the greater attention that can be fo-cused on a smaller number of reactor designs.

Standardization would aid the resolution ofsome of NRC’s generic safety issues, while theresolution of others would be unaffected.

INSTITUTIONAL RESPONSES

Current Nuclear Industry

The tasks of design, construction, and opera-tion are handled by diverse and independentorganizations, each with its own distinctivestyle and mode of business. The 75 commercialreactors now operating in the United Statesreflect this variety. However, in recent years,the industry has begun to reduce this diversityas designs have matured and to some extentconverged.

The two types of companies that togetherdesign the systems of a nuclear powerplant are

the manufacturers of the NSSS and AE firms.The four NSSS vendors design and manufac-ture the nuclear-related systems such as thereactor vessel and core, primary cooling sys-tem, and reactor protective system. The AEfirms (which number about 12) design the bal-ance of the plant, including the piping andelectrical layouts, auxiliary feedwater system,and the containment building. Both the NSSSvendor and the AE firm collaborate with theutility to produce a plant that meets the utili-ty’s specifications. In most cases, the AE firmalso serves as contractor for the pIant’s con-struct ion.

6 ● Nuclear Powerplant Standardization

In recent years, each NSSS vendor hasevolved basically one design for an NSSSwhich varies little from one order to another.The current variety of designs is due to thelarger number of AE firms than the NSSS ven-dors. Also, satisfying the different uti l it ies’design specifications creates additional varie-ty. Some AE firms have moved toward one de-sign with an interface package to match eachof the four NSSS designs. However, the designshave not moved toward greater similarity fromone AE to the next. Standardization wouldreduce the design effort of the AEs. This wou Idnot greatly reduce the total cost of the plantsince such efforts account for only a low frac-tion of the component and construction costs,but it would affect the AE’s business. Never-theless, AEs also serve as contractors and ac-cept some form of standardization as inevita-ble and in the best interests of the industry.

The NSSS vendors, AE firms, and util it iesshould continue to pursue a cooperative pro-gram of standardization, perhaps utilizing thecurrent trade associations. An alternativewould be the establishment of a joint utilityorganization that sets standards and design cri-teria which are more detailed than the currentNRC regulations. Neither of these conceptswill become a reality as long as the industry’sresources are stretched to meet NRC require-ments resulting from the accident at TMI.

Nuclear Regulatory Commission

Since 1973, NRC has had a program forlicensing standard nuclear reactors accordingto one of four definitions of standardization.The industry and utilities have participated ac-tively in hopes of a shorter and more predicta-ble licensing process. The gains in time andmanpower effort have only been marginal todate, although it may be premature to judgethe program’s success.

Industry observers believe that standardiza-tion will be hindered until NRC makes defini-tive rulings regarding which safety concernsare sufficient to warrant a design change in astandard reactor. Until that disciplined ap-

proach is achieved, no two “standard” reac-tors wilI remain alike.

The same basic criticism is leveled againstNRC in both its licensing and regulatory rolebecause it lacks clear direction for makingsafety rules. A long list of generic safety issuesare before the Commission, and several keysafety issues await the Commission’s ruling.The outcomes will remain unpredictable untilNRC establishes a safety goal to guide its deci-sions. Until regulatory and demand uncertain-ties are removed, no utility is likely to applyfor a new license — custom or standard.

Another step NRC might consider to encour-age standardization is the implementation ofstandard design approval, a concept for one-stage Iicensing (the current procedure is a two-stage process). NRC has considered the im-plementation of a standard design approvalwhich would involve submittal of informationthat is significantly more developed than thatnow provided for a preliminary design, butsomewhat less than that for a final design. TheGeneral Electric Co. has proposed a similarone-stage licensing program by which NRCwould grant a “power-worthiness certificate”to an acceptable design.

Congressional Role

Although no legislation has emerged fromCongress that directs a standardization effort,there remains considerable interest in whetherstandardization can improve nuclear safety.The findings of this study show that there is noquantifiable demonstration that standardiza-tion enhances safety but there is a strong “in-tuitive” feeling that it wil l . The issue thenbecomes the degree to which standardizationshould be pursued considering the tradeoffsbetween potential safety gains and possiblecosts as summarized above and discussed inthis report.

If Congress chooses to pursue the third orfourth approaches to standardization, legisla-t ion wi l l probably be necessary becauseneither the industry nor NRC will take thesesteps voluntarily. If Congress decides that the

Ch. l—Summary ● 7

forces of the marketplace restrained by thenumerous industrial standards are sufficient,then legislation mandating greater standardi-zation is probably not necessary. Action thatsupports this goal, either by legislation en-couraging it or setting-up incentives such asone-step Iicensing, could accelerate the trendand provide a clear policy statement aboutstandardization and nuclear safety. In this con-

nection, establishment of a nuclear safetygoal, by Congress, could be an important stepin encouraging standardization. Procedural oroperational standardization is also being pur-sued by the industry and util it ies. Congres-sional legislation is probably not necessary toachieve some degree of procedural standard-ization, but, again, could be encouraged by acongressional statement of national policy.

Chapter 2

INTRODUCTION

Chapter 2

INTRODUCTION

The nuclear industry that has developed inthe United States since 19s9 has grown up witha surprising degree of technical diversity. Allbut a handful of the 72 plants that are current-ly licensed for operation have been custom-de-signed and custom-built. A result of this prac-tice is that the plants must also be individuallylicensed, since the safety analysis of each is in-evitably different. When a utility decides tobuild a plant, it usually first hires an architect-engineering (AE) firm, then contracts with a re-actor manufacturer (one of the four existing“nuclear vendors”) to build the nuclear core,vessel, and control mechanisms, which repre-sent about 10 percent of the plant investment.Each vendor has a different design for its nu-clear system, so there are four different op-tions. Then the AE designs the balance of theplant (BOP):

● cooling systems;● feedwater systems;● steam systems;● control room; and● generator systems.

There are about 12 AEs presently designingnuclear plants in the United States, and eachhas its own preferred approach to these vari-ous systems. The AE’s approach will be tai-lored by past experience to be consistent withone vendor’s nuclear system, but not neces-sarily compatible with the systems of all four.In addition to the diversity due to the differentarchitect-vendor combinations, there is also adegree of variability due to the different mete-orological, seismic, and hydrological condi-tions at different plant sites.

Further var iabi l i ty i s introduced by thelength of the process (12 years) and the piece-meal approach that is taken to both design andl icensing. Because safety standards havegrown up with the nuclear industry rather thanbeing formulated in full and fixed fashionwhen the industry began, plant builders anddesigners have taken a “design-as-you-go” ap-proach to new plants in order to be able tomeet upgraded safety standards that might be

adopted during the period a pIant was underconstruction. For some years, the industry’spractice has been to start construction withthe design about 15-percent complete. On theregulatory side, the approach taken — to ac-commodate changing safety standards due toaccrued experience and improved analysis —has been to issue plant licenses in two steps, apreliminary step sufficient to start construc-tion and a final step necessary to start oper-ation. Both of these practices have inevitablyincreased the variation from one plant toanother. Even among plants intended to beidentical, but started at different times, signifi-cant design differences have occurred in thefinal plants.

Reducing the diversity that now exists in thenuclear industry would allow increased atten-tion to be given to improving each plant de-sign. It would also increase the amount ofoperating experience that would be availablefor a particular design and make it possible forimprovements at one plant to be immediatelyapplicable to an entire plant family.

Efforts to encourage standardization how-ever, have met with slow acceptance. Someargue that the many deviations from originaldesigns that now occur before plants operateindicate that neither the technology nor thelicensing process is sufficiently stabil ized tosupport standardization. Furthermore, the non-standardization that now exists in the industryis a direct resu It of the diversity that exists inthe marketplace, and a substantial move to-ward standardization could result in some re-structuring of the nuclear industry.

How substantial any move toward standardi-zation should be is one of the topics of thisreport. There is such a range of possible op-tions that l ie between the two logical ex-tremes—that either all plants be different orall be the same— that four different ap-proaches to standardization merit discussion.The different approaches represent greater de-grees of standardization, the last option beinga single design identical to all others in both its

11

12 ● Nuclear powerplant Standardization

nuclear and BOP systems. The approaches dif-fer in their technical, institutional, licensing,and safety implications. Some require stronglegislative action, while others rely predomi-nantly on trends already underway in theindustry.

This study was undertaken by requests ofthe House Committee on Interior and InsularAffairs and the Senate Subcommittee on Nu-clear Regulation. Some committee membersexpected that standardization would signifi-cantly improve the safety of the plants, andhelp create a stable licensing process in whichutil it ies would have confidence that theywould get their reactors approved. The acci-dent at Three Mile Island (TMI) contributed tothis expectation because both the local opera-tors and the Nuclear Regulatory Commission(NRC) personnel seemed to lack thoroughunderstanding of the reactor and had failed tolearn from similar experiences at related reac-tors. Ever-increasing licensing delays, especial-ly since TMI, reinforce the need to reexaminethe merits of standardization.

Congress is not the only institution inter-ested in standardization. NRC has also en-couraged standardization although recent ac-tions indicate that its priority at NRC has beenlowered. The NRC Advisory Committee on Re-actor Safeguards has maintained a strong in-terest in the subject. The nuclear industry hasalso been moving towards standardization asindividual companies have filed standardizedversions of their own designs with NRC. How-ever, such efforts have been directed more atunifying current practices than at maximizingsafety.

Any degree of standardization will requiredecisions as to the level of specification re-quired. The standard plants that have beenfiled with NRC specify flow diagrams, designdescriptions, and generic information, butdoes not include all the detailed informationrequired to actually build a plant. Completestandardization would require considerablygreater efforts before a design is approved andwould allow considerably less flexibility after-wards, but wou Id result in making plants vir-tually identical.

This is not an exhaustive study of standardi-zation. It is a broad scoping of four kinds ofstandardizations that could be considered andthe major advantages and disadvantages in-volved in each. In addition, the study examinesthe standardization of procedures and organi-zations to see if some advantages can begained without depending on new designs andplants. The retrofitting of existing plants to en-hance standardization or safety has not beenconsidered. OTA had staff and contractorsprepare background papers on NRC policy, theU.S. Navy’s experience with standardization,several plant systems that could be standard-ized, and the relation of standardization tosafety. These background papers were distrib-uted to the participants of a 2-day workshopheld to identify and discuss the issues of stand-ardization. The workshop included representa-tives of reactor manufacturers, AE companies,utilities, regulators, and concerned observersof nuclear power. This report is the result ofthe background papers, the conclusions of theworkshop, and further information received bythe staff. It has been reviewed by the workshopparticipants and by others.

Chapter 3

THE NUCLEAR INDUSTRY TODAY

Chapter 3

THE NUCLEAR INDUSTRY TODAY

LIGHT WATER REACTORS (LWRs)

To appreciate the degree to which standard i- erators, turbines, heat exchangers, etc.). Thezation could be improved (costs, savings, and different heat sources used are the combustionother benefits), present designs and the dif - of fossil fuel (e. g., coal, oil, natural gas) and fis-ferences among them must be understood. A sioning of nuclear fuel.steam electric station converts thermal energy(heat) to mechanical energy and finally to elec- The heat source in a commercial nuclear—.trical energy. This cycle of energy conversion plant is called a reactor cc

is common to all central thermal generating consists of an array of fuelstations and results in similar equipment being steel cylinder (the reactorused amongst facilities (e. g., feed pumps, gen- sustaining a controlled nuc

Figure 1 .—Boiling Water Reactor Core and Vessel Assembly

Vent andhead spray

Steamout let

Core sprayinlet

Low-pressurecoolant

injection inletCore spraysparger

Jet pumpassembly

Fuelssemblies

Jet pumprecirculationwater inlet

Vesselsupport skirt

Controlrod drives

In-coreflux monitor

SOURCE: General Electric Co

7 7 - 5 3 8 -

Steam dryerlifting lug

Steam dryeassembly

Steam separatorassemblyFeed water Holddowninlet ring

Feedwatersparger CEA shroud

Coreassembly

spray lineTop guide

Core shroudControlblade

Core plateRecirculationwater outlet

Shield wall

Upperguidestructure

Inletnozzle

$ 1Activecorelength I II

Control rod drivehydraulic lines

Instrumentation lengthassembly u

core (fig. 1). The corebundles (fig. 2) in avessel) capable ofear reaction.

u InstrumentationIu ~ _

nozzleSOURCE Combustion Engineering, Inc

. Alignmentkey

Coresupportbarrel

Outletnozzle

Coreshroud

Fuelassembly

Lowersupportstructure

Flowskirt

15


Upper I

plate

Figure 2.— Fuel Bundles

BWR Fuel

Fuel E x p a n s i o nbundle spring

Fuelchannel Plenum

Fuel rodspring

interimspacer Getter

Fuel pelletFingerspring Fuel rod

(typical Lower tie plate

of “4) N o o s e p i e ~ e

SOURCE General Electric Co

PWR FuelAlinement post

Upperend-fitting

Spacergrid

CEAguidetube

m assemblyTop view

1- Fuel rod

TActivefuellength

1Bottom view

L o w e r i n 9SOURCE Combustion Engineering, Inc

Upperend cap

Spring

Spacer

Dishedpellets

Fuelcladding

Spacer

Lowerend cap

The fuel bundles consist of square arrays of50 to 250 fuel rods about 1/2 -inch in diameterand 12 feet long. Each rod is filled with 1/2-inch-long fuel pellets containing slightly en-riched uranium dioxide, and 200 to 500 fuelbundles arranged in a circular array form thecore.

A nuclear reaction is initiated by the absorp-tion of a neutron in the nucleus of a fissionableatom (e. g., uranium-235, plutonium-239). Thefissionable atom splits, releases energy andmore than one neutron. These extra neutronsare then available to produce more fissionsand continue the reaction and the release ofenergy. This release of energy produces heatwithin the fuel which in turn is released to thecool ing water flowing through the core.

In a boiling water reactor (BWR), the typeshown in figure 3, this coolant is allowed toboil. The steam thus produced drives a turbine,which in turn yields electrical energy. In a pres-surized water reactor (PWR), shown in figure 4,the water that circulates through the core (theprimary coolant) is kept under pressure andnot allowed to boil. Instead, it transfers itsheat in a steam generator to a secondary cool-ing loop. Water in this steam generator thenboils, and its steam drives a turbine. In bothBWRS and PWRS, the steam emerging from theturbine is discharged to the main condenserwhere the steam condenses and the waste heatis rejected to a heat sink such as a coolingpond or tower, The condensed steam or waterthen returns to the reactor vessel (in a BWR) orto the steam generator (in a PWR) to begin thecycle over again. The conversion of steam toelectrical energy with turbines and generatorsis similar to nonnuclear steam electric stations.The systems used in this conversion are refer-red to as power generation or nonsafety-re-Iated. The major systems required for the nu-clear heat source — including some, but not all,of the safety-related systems— are defined. bythe industry as the nuclear steam supply sys-tem (N SSS).

The byproducts from the fission process in-clude unstable nuclei (fission products) whichdecay to more stable nuclei by emitting an en-ergetic particle or gamma ray. This decay proc-

Ch. 3—The Nuclear Industry Today ● 17

Photo credit Atomic Includes Forum, Inc

A refueling crane operator lowers a fresh fuel bundle into the core of a boiling water reactor. To the right of the fuel bundleare two of the four vessel penetrations that route steam from the reactor to the turbine

ess produces heat at a much lower rate (severalpercent of the fission process), but it continueseven after the reactor is shut down.

The fission process carries the unique prob-lems of fission rate control, f ission productcontainment, and decay heat removal. Sys-tems normally associated with these processesare known as “safety-related” systems sincethey are the ones depended on to prevent orcontrol accidents that could endanger the pub-lic. Several safety-related systems are dis-

cussed here with the purpose of understandingthe relationship of safety to standardization.

Fission Rate Control

The rate of the fission reaction is controlledby materials that absorb neutrons without fis-sioning and, therefore, absorb the neutronsavailable for fission. These absorbers are com-monIy referred to as “poison s.”

The term “control rod” refers to a mechani-cal device containing an absorber with a fixed


Figure 3.— Boiling Water Reactor (BWR)

Turbilgener

neato

Condenserc o o l i n gwate r

SOURCE: Atomic Industrial Forum, Inc.

S t e a m Steam line

Turbinegenerator

●

Condensercoolingwater


Photo credit Atomiic Indusrial Forum, Inc

The major portions of the power conversion train are located within the turbine building. 1) Main turbine, convertssteam’s thermal energy to rotational mechanical energy. The thermal energy is generated in the core by fissioning

nuclear fuel. 2) Main generator, converts rotational mechanical energy to electrical energy. 3) Generator alterix,maintains the generator’s rotating electric field

geometric shape. Another form of poison issoluble in water and added to the primarycoolant. I n pressurized water reactors thesesoluble poisons are used in both safety andpower generation systems. In boiling water re-actors they are only used in safety systems.

There are differences in designs between re-actor vendors in both the control rod and itsmechanical drive. PWRS use tubular controlrods that are inserted into the fuel bundle. I nBWRS the control rod is in the shape of a cruci-form which is inserted between fuel bundles.In either case, the rod and its mechanical driveare a “standard” design peculiar to each ven-dor.

If the fission rate increases above a predeter-mined level (greater than the rate at whichheat can be removed by the coolant), the fis-sion process is stopof the control rodscalled a “scram”).power excursionssystems and scram“reactor protection

ped by the rapid insertionthis function is commonly

The systems that senseo actuate the protective the reactor are called

system s.” These systemshave undergone a careful evolutionary designchange with changes in state-of-the-art elec-tronics — e.g., one vendor has changed the sys-tem’s analog signal processor to one usingdigital computers. Although these designs arestandard to each vendor, they have not been

“locked-in” to one design insulated from ad-vances in the applicable technology.

Fission Product Containment

The radioactive fission products must not bereleased to the environment in excess ofFederal regulations because they could harmthe general public and the plant’s personnel.Several barr iers ex ist between the f i ss ionfragments and the environment. They are:

● fuel pellet;. fuel rod (i. e., cladding);● reactor vessel and primary coolant piping;. primary containment; and

Photo credit Atomic Industrial Forum, Inc

One-half-inch long fuel pellets ( < 1/2 in. diam.) containingslightly enriched uranium dioxide


● secondary containment (on BWRS andsome PWRS).

Each barrier is a backup to the one before inthe event of failure —e. g., failure of the fuelrod as a boundary is mitigated by the reactorvessel and associated piping. In addition, pene-trations in the primary containment (e. g., forventilation ducts, piping, etc. ) have isolationvalves (normally two) which close automati-cally on signals indicating potential fuel fail-ures. The barriers listed can generally be de-scribed as passive (e. g., the fuel rod has no ac-tive components), or active (e. g., the isolationvalves require motive power to shut and re-quire process signals for automatic actuation).

During the Three Mile Island (TMI) accidenta hydrogen explosion caused a pressure pulsethat actuated the containment isolation sys-tem. The system’s sensors and relays changedelectrical states and signaled the containmentisolation valves to shut. The signal was of shortduration (4 minutes) and eventually cleared,allowing the operator to “reset” the contain-ment isolation system, thereby returning theelectric portion of the system to its previous“standby” state. 1 On resetting, the contain-ment isolation valves for the containmentsump opened, allowing contaminated waterfrom inside the containment to flow to the aux-iliary building. This may have caused an inad-vertent release of gaseous activity into the en-vironment through the exhaust ventilation inthe auxiliary building. The simple resetting ofthe isolation signals should not have causedthe containment valves to open.

A post-TMl requirement was to review thisproblem and ensure that each containmentisolation system would not automaticalIy openisolation valves when the initiating signal wasreset. ~ A review of a selected number of re-sponses to this requirement shows that this wasa problem at some reactors but not at others.

‘ E Iectrlc Power Research Institute, NLIC Iear Safety AnalysIsCenter, “Analy$ls of Three Mile Island, Unit 2 Accident, ”NSAC-I , j Ulv 1979

‘Nuc lea r Regu la to r y Commiss ion , “NRC ActIon PlanDeveloped as a Result of the TM I-2 Accident, ” VOI 1,NUREG-0660, January 1980

This lack of standardization in containmentisolation systems required a detailed review ofeach plant’s containment isolation system andresulted in a unique fix for each similar prob-lem that was discovered. The Nuclear Regula-tory Commission (NRC), in turn, had to stretchits limited resources to review each design todetermine whether or not a modification wasrequired. This lengthened the time and re-duced the depth of the review.

Auxiliary Feedwater Systems(PWRS only)

Auxiliary feedwater (AFW) systems are de-signed to remove decay heat when the reactoris shut down but at high pressure (normallygreater than 400 lb/in 2). The design criteria forthem are usually established by the NSSS ven-dor while the detailed design responsibil ityusually rests with the architect engineer (A E).AFW systems are required to be available onloss of main feedwater. The inadvertent isola-tion of this system was a possible contributorto the accident at TMI. Valves on the outlet ofthe pumps were found shut and they isolatedthe pumps from the steam generators. The op-erator eventualIy opened these valves (approx-imately 7 minutes into the accident). 3

In addition, the unavailability of a plant’sAFW system is an important and significantcontributor to the overaIl risk of any particularPWR. As mentioned earlier, the generation ofheat from fission products must be removed ordissipated to ensure that the integrity of thepassive containment boundaries is maintained.I n a PWR, the methods available at high reac-tor pressures for decay heat removal are theAFW system; some PWRS are also able to usean alternative method incorporating the high--pressure injection pumps. ’ The former methodis preferable because the AFW system is on thenonradioactive side of the plant. The lattermethod is often called “feed and bleed” andmay requ ire discharging radioactive primarycoolant onto the containment floor. The latter

‘E Iectrlc Power Research Institute, op cit‘Nuclear Regulatory Commission, “Generic Evaluation of

Small Break Loss of Coolant Accident Behavior In Babcock &WIICOX Design 177-F Operating Plants, j anuary 1980


was the primary heat removal mechanism dur-ing the initial phases of the accident at TMI. 5

In response to TMI, NRC conducted a de-tailed review of AFW systems in PWRS to iden-tify deficiencies in existing systems by assess-ing their relative reliability under loss of mainfeedwater. The results of a portion of the studyare presented in table 1 and figure 5.6 Table 1shows the diversity in an AFW system for onePWR vendor. Note that only one plant hadautomatic system initiation and most plantsdiffer in the number of pumps of each type. Adirect result of this diversity is shown in figure5, Quantitative reliability assessments on 33existing AFW systems show there is a widespread in the likelihood that the AFW systemwilI fail on the interruption of main feedwater.As with the primary containment isolationproblem, the design solutions to this problemare many and have very few elements in com-mon. In addition, the acceptability of the sys-tem is impossible to judge in the absence of aspecific reliability goal. Therefore, the designsolutions are unique to each plant and subject

‘t lec trl{ Power Rewarch I nst Itute, of) c it

“N uc lea r Regulatory Corn m Iss.ion, “ G enertc E val uatlon ofFeeclwater Transient\ and Small Break L055 of Coolant AccidentsIn Combustion E ngineerlng Design Operating Plants, ” NUREG-06 ]5, J anuarv 1980

to arbitrary judgment. If these systems weremore standard than they are today, therewould not be such a wide divergence in relia-bility; therefore, mandated engineered fixes tothe design would be easier to implement andreview.

A reduction in the diversity of AFW systemdesigns alleviates the above-mentioned prob-lems. Two items are encouraging in this areaand illustrate the industry’s progress towardstandard system designs. First, a review of ex-isting standard designs supplied and docketedby the AE’s show a marked increase in stand-ardization of auxiliary feedwater systems com-pared to those in existing plants—design isdocketed when it is formally submitted to NRCand the administrative process for review andapproval begins. Ten AEs have designed anauxiliary feedwater system that is applicableto all PWRS. Therefore, this results in a singleAE’s design that is applicable to all threePWRS. A further step toward standardizationof auxiliary feedwater systems is the approvalby the American National Standards Instituteof a design standard for these systems. 7 Th is

7 A m e r i c a n N a t i o n a l S t a n d a r d s I n s t i t u t e , “ A u x i l a r y F e e d w a t e r

Systems for Pressurized Reactors, ” AN SIIANS 51 10, November

1979

Table 1 .—Auxiliary Feedwater Systems

Number ofpumps/type

Plant AIE of drive Capacity

Arkansas Nuclear One, Unit 2 Bechtel 1 steam-driven Steam: 575 gal/rein @ 2,800 ft1 motor-driven Motor: 575 gal/rein @ 2,800 ft

Calvert Cliffs 1 & 2 Bechtel 2 steam-driven 700 gal/rein @ 1,100 lb/in2a eachper unit

Ft. Calhoun 1 Gibbs & 1 steam-driven Steam: 260 gal/rein@ 2,400 ftHill

1 motor-driven Motor: 260 gal/rein @ 2,400 ft

Maine Yankee Stone & 1 steam-drivenWebster

2 motor-drivenMillstone 2 Bechtel 1 steam-driven

2 motor-drivenPalisades Bechtel 1 steam-driven

1 motor-drivenSt. Lucie 1 Ebasco 1 steam-driven

2 motor-driven

Steam: 500 gal/rein@ 1,100 lb/in2g

Motor: 1,500 gal/rein @ (each) 1,100 lb/in2gSteam: 600 gal/rein @ 2,437 ft

Motor: 300 gal/rein @ (each) 2,437 ftSteam: 415 gal/rein@ 2,730 ftMotor: 415 gaI/min @ 2,730 ft

Steam: 500 gal/rein @ 1,200 lb/in2

Motor: 250 gal/rein @ (each) 1,200 lb/in2

AFW system mode ofi nit iat ion

Automat ic

Manual

Semiautomaticmotor-driven

Pump manuallyconnected to diesel

generatorManual

Manual

Manual

Manual

SOURCE Nuclear Regulatory Commlsslon


Figure 5.—Comparisons of Auxiliary FeedwaterSystem Reliability on the Loss of Main Feedwater

System (LMFW)

Plant

3 <

SOURCE: Nuclear Regulatory Commision.

standard was approved late in 1979 and tookabout 3 years to develop through the “consen-sus” process. As encouraging as these items ap-pear, they lack the quantitative reliability cri-teria needed to remove the arbitrariness in reg-ulatory judgments regarding their adequacy.

Decay Heat Removal

At low-reactor pressures ( less than 400Ib/in2), redundant methods of decay heat re-moval prevent the uncontrolled heatup of thecore. The systems remove decay heat by con-tinuously circulating water through the coreand rejecting the heat through heat exchangersto the ultimate heat sink (e. g., cooling tower,pond, lake, etc.),

The heat removal function operates in twomodes: 1 ) “emergency core cooling” during ac-cident conditions, and 2) normal “shutdowncooling” when the pIant is not producing elec-tricity. In the emergency core-cooling mode,the systems operate automatically to providecooling. In the shutdown cooling mode, theoperator sets up the system manually in ac-cordance with the procedures for shuttin g

down the plant. The design responsibility ofthese systems rests with the vendor. There isvery little difference between plants of thesame vendor. For light water reactors (LWRS)there are four basic residual heat removaldesigns which are standardized. These designsall comply with the “general design criteria, ”which are part of the Federal code (10 CFR)governing the design, construction, and opera-tion of commercial reactors.

However, cr i t ics of these designs havepointed out that, due to the lack of specificityin the requirements, the fundamental problemof decay heat removal during the normal shut-down cooling mode has been overlooked. n in-stead, the operator is required to use his witand ingenuity to overcome built-in design com-plexities for the simple purpose of removingdecay heat during plant malfunctions when aloss-of-coolant accident does not occur. Awell-publicized example of this is the Brown’sFerry fire where decay heat removal dependedon nonsafety-related equipment arranged in amanner not previously considered necessaryfor shutdown conditions. Even though thesestandardized residual heat-removal systemsexist for both PWRS and BWRS and conform tothe existing design criteria, their adequacyunder nonaccident conditions is questionable.In fact, the West Germans have added to theirAmerican-designed PWRS an extra “bunkered”decay, heat removal system independent ofthe safety-related systems used during loss-of-coolant accidents.

As the various criteria for decay heat remov-al illustrate, NRC’s general design criteria (sup-plemented by the existing standards and regu-

“Common Mode Failure of Light Water ReactorSystems. What Has Been Learned,” for Energy Analysis,May 1980


Iations) may not be adequate for routine oper-ations during adverse plant conditions (e. g., aplant fire). Some suggest this deficiency resultsfrom the lack of specificity in the criteria.Therefore, standardizing designs, without in-creasing the level of detail in the criteria andaccounting for past operating experiences,may not make future standard plants any saferthan the existing operating ones. New NRCrulemaking actions in the wake of the accidentat TM I point this out.

Control Room Design

Because the accident at TM I highlightedconcern over operator error, greater attentionis being placed on the control room design. I nthe past, control room designs have varied agreat deal from plant to plant. One reason wasthe considerable input from the utilities, whichhave preferred to maintain a degree of similari-ty between their nuclear plants and other typesof power-generating plants. Even before theTMl accident, control room designs for futureplants incorporated some of the followingfeatures: 9

●

●

●

●

consideration of functional grouping ofthe reactor control panels;location and layout of individual controlson each panel in a logical common sensemanner;compliance with regulatory criteria forseparation and instalIation of safety-gradecontrol equipment; andutiI ization of state-of-the-art computerand display technology to aid the opera-tor in the evaluation and control of theplant’s condition.

Since TMI, NRC has required all operatingreactor Iicenses and applicants for operatinglicenses to perform a detailed control roomdesign review to identify and correct deficien-cies.10 These reviews, which are expected totake 1 year, are to include, among other things,

4“Surmortlng Intormatlon t o r t he Backg round Papers o n.,Nuclear Powe~plant Stancfardlzatlon, ” submitted to the Of f I ceot Techno logy Assessment, September 1980

‘ ( ] Nuc]ear R e g u l a t o r y C o m m i s s i o n , “ N R C A c t I o n Plan

Developed a~ a Result of the TN! I 2 Ac cldent, ” OP clt

an assessment of control room layout and con-sideration of human factors that influence op-erator effectiveness.

These requirements may indirectly lead togreater standardization of control room de-signs. Three of the four vendors offer speciallydesigned control rooms that incorporate“human engineering” features. Most recentcontrol room designs by AE firms incorporatesome human engineering. Utilities are likely tofind it too expensive to custom-design theirown control rooms for new facilities. Thus, thenumber of d i f ferent control room des ignsshould be reduced in the future.

Causes for Variations Among LWRS

Aside from the two major types of l ightwater reactors (BWRS and PWRS), there aremany possible variations in design rangingfrom minor deviations in piping layouts todifferent numbers of steam generators to thedifferent types of heat sink. Some of the majorv a r i a t i o n s s t e m f r o m t h e v a r i e d d e s i g n sevolved by the three vendors supplying PWRSand from the range of reactor sizes desired tobe built –e.g., Westinghouse reactors all haveone standard loop design for the primary cool-ant system, and plants of different sizes resultby including two, three, or four such loops inparallel. By contrast, Combustion Engineeringand Babcock & Wilcox have two loops in everyplant but meet different power requirementsby varying the size of pumps and steam gener-ators.

Other variations in design result from site-specific factors. Reactors built in regions sub-ject to earthquakes must be designed forhigher reaction loadings for such features asthe containment structure and mechanical andelectrical equipment. The meteorology associ-ated with a particular site affects plant design(possibly mandating a secondary containment)because of concern over the patterns of disper-sion of any radioactive gases released from theplant. Flood and tornado hazards may alsohave some effect on plant design. Duke Powerhas cited an example of site-specific require-


ments that caused major divergences betweentwo standard units that were bui l t at i t sMcGuire site and two units intended to beidentical, but built later at its Catawba site.Differences between plant characteristics atthe two sites were forced by: 1 ) rulings of theEnvironmental Protection Agency (EPA) andNRC, and 2) changes in industry standards dur-ing the period of design.

EPA required Catawba to use cooling towersrather than once-through-cooling. Cooling tow-ers are less efficient. The additional powerconsumed by the fans and the higher tempera-ture of the cooling water in the condensers af-fected the design of other plant systems. Theoverall power rating will be reduced from1,180 to 1,145 MWe.

Duke Power also intended the decay heat re-moval systems for these standard plants to be

the same, however, the EPA ruling cited aboveforced the Catawba decay heat removal heatexchangers to be larger than those at MCGuire ire.In addition, NRC took a new regulatory posi-tion requiring the Catawba units to have an in-dependent suction from the reactor coolantsystem for each of the two trains of decay heatremoval. The MCGUire ire units have a single suc-tion supplying both trains. Finally, the industrystandards changed in the time period of designof the four units, causing variation in the char-acteristics of such items as pumps and reliefvalves. ’ These are a few of the many examplesof similar modifications. However, such designchanges may not be great enough to inhibitsome of the benefits of standardized plants.

1“’Supporting Information for the Background Papers onNuclear Powerplant Standardliation, ” op cit

THE NUCLEAR POWER INDUSTRY

The major participants in the process ofdesigning and constructing a nuclear powerplant are the:

● electric util ity;● NSSS vendor;● AE; and● construction company.

The total number of companies involved maywel l be in the hundreds, but these foureffectively control the major decisions.

If and when a utility determines that it needsnew control-station generating capacity, itusually hires an architect-engineer firm to helpestimate costs and other considerations of thevarious options. Eventually, alternative powersystems —e.g., solar, wind, etc. — may be con-sidered, but at present few utilities have anyoptions other than coal or nuclear for large,new power supplies. The cost comparison in-cludes fixed-price bids from some or all of thefour NSSS vendors. The utility then contractswith one of the NSSS vendors to supply thenuclear components, and an AE firm (usually,

but not always, the same one) to design thebalance of the plant (BOP). The utility alsohires the construction company (often, but notalways, the AE firm). The AE and constructioncompanies work on a cost-plus basis since it isimpossible to predict in advance exactly whatlevel of effort will be required. In some cases(usually large utilities), the utility may act as itsown AE or constructor or both.

The process outlined above and the partici-pants described below represent the industryas it operated several years ago. No plantshave been ordered for several years and feware expected for the next few years. Somechanges may be expected if a resurgence of or-ders occurs, particularly if a policy of stand-ardization is enforced. For instance, OffshorePower Systems, a subsidiary of Westinghouse,offers a complete nuclear powerplant. In thiscase, the Westinghouse reactor is mounted ona barge and sold to the utility complete withall systems required to operate the reactor andgenerate electricity. The only AE involvementwould be in site preparation. A somewhat simi-


Iar scope of supply will be available at GeneralEIectric Co, (C E), which expects to offer a com-plete “nuclear island. ” The nuclear islandconsolidates the GE BWR and auxiliary equip-ment into one standard design and includes allof the buiIdings and structures that have radio-logical significance. Some AEs offer standardBOP designs which interface with the standardNSSS.

Vendors

Four companies manufacture NSSS for nu-clear LWRS. These are listed in table 2 togetherwith the number of plants built and on order,and the total generating capacity of theseplants. 12 GE makes a BWR while the otherthree companies make PWRS. BWRS and PWRSare clearly quite different facilities that willcall for quite different systems, components,and layouts. However, the three PWRS are alsoquite different. The number of loops for agiven power level may vary as can the size ofthe reactor, the means of controlling it, andthe design philosophy of the systems servicingit. All three PWRS are the end product of twodecades of somewhat divergent evolutionarydevelopment. Even though conceptually simi-lar, the engineering approaches to the variousdesign problems have been so sufficiently dif-ferent that each NSSS is quite distinctive.

Since the NSSS is only one part of a largecomplex of systems comprising a powerplant,design of other systems may be assigned toeither the vendor or the AE at the discretion ofthe utility owner. I n recent years however, a

1 2U S D e p a r t m e n t o f E n e r g y , “ N u c l e a r P o w e r P r o g r a m i n -

format ion and Data, ” May 1980

uniform scope of responsibi l i ty has comeabout through actions by NRC. When NRC[formerly the Atomic Energy Commission) wasbeginning to encourage standardization of nu-clear powerplants in the early 1970’s, it formu-lated a detailed program for docketing stand-ard plants for review and approval. The ven-dors at that time decided to limit their scopeof design responsibility to those componentswhich they planned to market as a standard-ized responsibility (i. e., those components thatwere proving competitive). As a result, NRCdeveloped the list of systems shown in table 3as the NSSS standard plant scope to be dock-eted by each vendor. Note that the list ofsystems is largely the same for each vendor. 13

Architect Engineering Firms

The remaining systems necessary for a func-tioning plant are referred to as the BOP. SomeAE firms in accordance with NRC’s programsubmitted standard plant designs for the BOP.Each firm’s BOP design is matched to the NSSSthrough “interface criteria. ” The BOP designsvary from one firm to another, but each firm’sBOP design is generally applicable to any PWRby adjusting parameters (e.g., pressures andflow rates) to meet the interface criteria.BWRS require a separate class of BOP designs.

The NSSS represents about 10 percent of thetotal plant, and the AEs design the remaining90 percent. The cost of the plant design isabout 10 percent of the total plant cost. Thereare also considerably more AEs than vendors,

1‘Nuclear Regulatory Commlsslon, “Programmatic informa-tion for the Licensing of Standardized Nuclear Power Plants, ”WASH-1 341 and amendment 1, August 1974

Table 2.—Nuclear Reactor Suppliers

Commercial plants Under construction On order

Manufacturer Number MWe Number MWe Number MWe

Westinghouse . . . . . . . . . . . . . . . . . . . 27 20,063 38 41,454 3 2,590General Electric . . . . . . . . . . . . . . . . . . 24 17,758 28 30,101 7 8,304Combustion Engineering . . . . . . . . . . 8 6,361 15 17,893 6 7,490Babcock & Wilcox . . . . . . . . . . . . . . . . 9 7,885 8 7,947 3 3,790Other . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1,230 — — — —

Total. . . . . . . . . . . . . . . . . . . . . . . . . . 71a 53,297 89 97,396 19 22,174

aDoes not Include Indian Point 1 or Humboldt Bay

SOURCE U S Department of Energy

26 ● Nuclear Powerplants Standardizationl

Basic primary loop configuration of standard

Photo credit: Combustion Engineering Inc.

NSSS (3,800 MWth class)illustrating standard features of design

so the present diversity of designs is due moreto the AEs. The AEs and their share of thebusiness is indicated in table 4.

Experienced AEs have preexisting designsthat can be tailored to specific site characteris-tics and utility needs. As will be discussed fur-ther, some of these designs have achieved suf-ficient maturity; the AEs have developedstandard plants for some or all of the NSSSS.The use of such standard plants reduces the re-quired design effort (which as stated before isa moderate fraction of the total cost) and alsoreduces the schedule and uncertainty of I icens-ing. These gains become questionable if theutility insists on too many modifications to suitits particuIar desires. Greater standardizationwould affect the relationships of involved

“U S Department of Energy, op cit

firms by reducing the role of thenearly to that of the purchaser of

utilities morea stock item.

Construction Companies

With only a few exceptions, the companiesthat build nuclear powerplants are the sameAE firms that design them. Thus, they will notbe identified separately here. The role of theconstruction company is to build the plant ac-cording to the design and specifications of theAE and the NSSS vendor. Theoretically, twoplants built to the same design would be iden-tical, but in actual fact, minor differencesdevelop at the work site. A subcontractor maydeviate slightly from his blueprint due to un-foreseen interferences, buildup of tolerances,problems with field fits, or the unavailabilityof a component. These changes are performedunder the supervision of a responsible engineer


Table 3.—Content of an NSSS Standard Design Application” fI. .

Babcock & WilcoxA. Reactor

1. Fuel assemblies2. Reactor vessel internals3. Control assemblies4. CRDMS

B. Reactor Coolant System(including layout andanalysis)1. Reactor vessel2. Reactor coolant pump3. Steam generator (not

beyond nozzles)4. Main piping5. Pressurizer (including

safety valves)6. Pressurizer relief system7. Inservice inspection8. Equipment supports (not

including embeddedanchorage)

C. Emergency Core CoolingSystems

D. Instrumentation andControls for the NSSSC

1. Main control room panelboard (including allintegral equipment)

2. l&C equipment racks andpanels

3. Reactor control andprotection systems(including actuationsystems)

4. Nuclear Instrumentationsystem

5. Process l&C (includingcontrol valves)

E. Electric Powerc

1. CRDM power supply2. Pressurizer heater

controls

F. Auxiliary Systems1. Special handling

equipment for fuel andreactor vessel internals

2. Makeup and purificationsystem

3. Chemical addition andboron recovery system

4. Steam generatorcirculating system

5. Decay heat removalsystem

G. Startup Test Program forNSSS Items

Combustion EngineeringA. Reactor

1. Fuel assemblies2. Reactor vessel internals3. Control element

assemblies4. Control element drive

mechanisms

B. Reactor Coolant System(including layout andanalysis)1. Reactor vessel2. Reactor coolant pump3. Steam generator (not

beyond nozzles)4. Main piping5. Pressurizer (including

safety valves)6. Inservice inspection7. Equipment supports (not

including embedded

c.

D.

anchorage)

Emergency Core CoolingSystems

Instrumentation andControls for the NSSSC1. Main control room panel

board (including allintegral equipment)

2. l&C equipment racks andpanels


4. Neutron monitoringsystem


E. Electric Powerc

1. Control element drivemechanism power supply

2. Pressurizer heatercontrols



2. Chemical and volumecontrol system

3. Shutdown cooling system


General ElectricA. Reactor

1. Fuel assemblies2. Reactor vessel internals3. Control assemblies4. CRDMS5. Control rod drive

hydraulic system

B. Reactor Coolant System(including layout andanalysis)1. Reactor vessel2. Recirculation pumps3. Recirculation piping and

MSL piping (including butnot beyond secondisolation valve)

4. Safety/relief valves5. Inservice inspection6. Equipment supports (not

including embeddedanchorage)


D. Instrumentation andControls for the NSSSC1. Main control room panel

board (including allintegral equipment)

2. i&C equipment racks andpanels


4. Nuclear instrumentationsystem


E. Auxiliary Systems1.

2.

3.

4.

5.

6.

7.

Special handlingequipment for fuel andreactor vessel internalsStandby liquid controlsystemReactor core isolationcooling systemMSLIV leakage controlsystemReactor water cleanupsystemResidual heat removalsystemPressure regulationsystem

F. Startup Test Program forNSSS Items

WestinghouseA. Reactor

1. Fuel assemblies2. Reactor vessel internals3. Control assemblies4. CRDMS (including missile

shield and ventilation)

B. Reactor Coolant System(including layout andanalysis)1.2.3.

4.5.

6.7.8.

Reactor vesselReactor coolant pumpSteam generator (notbeyond nozzles)Main pipingPressurizer (includingrelief and safety valves)Pressurizer relief tankInservice inspectionEquipment supports (notincluding embeddedanchorage)


D. Instrumentation andControls for the NSSSC1.

2.

3.

4,

5.

Main control room panelboard (including allintegral equipment)l&C equipment racks andpanelsReactor control andprotection systems(including actuationsystems)Nuclear instrumentationsystemProcess l&C (includingcontrol valves) -

E. Electric Powerc

1. CRDM power supply2. Pressurizer heater

controls



2. Chemical and volumecontrol system

3. Boron recycle system4. Emergency boration

system5. Residual heat removal

system


aThe Items to be addressed In an NSSS SSAR are listed by major systems, components, and structures Items more detailed In nature will be handled on a case”by-case basis

bFor ~a~h ,tem Ilsted, the NSjSS SSAR s h o u l d preSent the fUflCtlOnal cfewlptm. design requirements, drawings and diagrams. safety evaluation, and Interface

requirements With the exception of the layout. analysts, and supports for the reactor coolant system, other design aspects such as layout, structural

conslderatlons, supports. plplng analysls, protection against floodlng, pipe whip, missile protection, cabllng layout,ventilation requirements, Instrument cabllng

and plplng, etc should be addressed In the BOP SSAR.Clncludes the equipment ,tems only for the NSSS, not the Interconnecting plplng and cablingdDeslgn Provlslons to accommodate Inser’vice lnsPectlon

SOURCE. Nuclear Regulatory Commlsslon


Table 4.—Architect Engineering Responsibility for Nuclear Powerplants

Commercial plants Under construction On order

Architect Engineer Number MWe Number MWe Number MWe

Bechtel . . . . . . . . . . . . . . . . . . . . . . . . . 27 20,099 21 22,564 6 7,494Burns & Roe . . . . . . . . . . . . . . . . . . . . . 4 3,184 2 2,163 1 350Black &Veatch . . . . . . . . . . . . . . . . . . . — — — 2 2,300Brown & Root . . . . . . . . . . . . . . . . . . . . —

—— 2 2,500

Ebasco . . . . . . . . . . . . . . . . . . . . . . . . .— —

4 2,676 8 8,003 1 1,150Gilbert/Commonwealth. . . . . . . . . . . . — — 3 3,310Gibbs & Hill. . . . . . . . . . . . . . . . . . . . . .

— .1 457 2 2,222

Gilbert Associates . . . . . . . . . . . . . . . .— —

3 2,114 —Fluor Power Services. . . . . . . . . . . . . .

— — —3 1,595 —

Sargent & Lundy. . . . . . . . . . . . . . . . . .— — —

8 5,626 13 13,310 2 2,240Stone&Webster . . . . . . . . . . . . . . . . . 9 5,859 11 10,797 4 4,800United Engineers.. . . . . . . . . . . . . . . . 4 3,480 4 4,836Tennessee Valley Authority . . . . . . . .

— —4 4,343 13 15,896

Utility owner a . . . . . . . . . . . . . . . . . . . . 4 3,864 10 11,795 ‘ 3 3,840

Total. . . . . . . . . . . . . . . . . . . . . . . . . . 7 1b

53,297 89 97,396 19 22,174alncludes Niagara Mohawk Power Corp. Public Service Electric&GasCo., American Electric Power Service Corp Pacific Gas & Electric Co..and Duke power Corp.bDoes not Includ lndian Polnt l or Humboldt Bay

SOURCE” Off Ice of Technology Assessment

Photo credit Atomic Industrial Forum, Inc

A milestone in the construction on a nuclear powerplant isthe setting of the reactor vessel within the containment.

In this photo, a PWR vessel is being lowered into position.The steam generators have already been set in place

in the background

which prevents the subcontractor from arbi-trarily changing the design. However, stringentlevels of stapractices arconstruction

The roles

Idardization might frustrate suchd lengthen the time required forof the plant.

Industry Trends

]f the utility and these three parti-cipants are not fixed. Some utilities do some orall of the AE design work themselves; the Ten-nessee Valley Authority (TVA), Duke Power,and American Electric Power are examples.The utility is responsible for licensing, but itcan delegate the bulk of this task to the AE andvendor if it chooses. Standardization wouldtend to diminish utility involvement in licens-ing. AEs would also have a less pivotal role.

Current trends in standardization will be dis-cussed in the following chapters, but it shouldbe noted from table 4 that the dominance ofseveral AEs may help ensure a certain degreeof standardization even in the absence of anyofficial action. Only four AEs (not counting theutilities) have more than four projects under-way: Bechtel, Ebasco, Sargeant & Lundy, andStone & Webster. The current experience andexpertise of these four (plus one or two others)will likely attract utilities to them when and ifthey begin to order new plants. Any resump-


tion of orders is likely to be at a relatively slowrate compared to the peak years of the late1960’s and early 1970’s. These four to six AEscould probably handle all the renewed busi-ness, and the utilities would most likely con-centrate their orders on them. In that event,the number of different possible combinationsof BOP plus NSSS would be sharply reduced.

Table 5 shows the present combinations ofNSSS vendors and AEs for LWRS under con-

struction or on order. Instead of being 56 possi-ble combinations, there are 22 NSSSIAE, plus 4NSSS/TVA, and 2 other utility designs. ” If GEsucceeds in completing Iicensing its nuclearisland, if AEs having a smaller share of themarket are excluded, and if most of the re-maining ones have approved standard designs,the total number of combinations could beless than 10.

‘‘U S Department of Energy op clt

Table 5.—NSSS/AE Combination of Light Water Reactors Under Construction or On Order

Bechtel . . . . . . . . . . . . . . . . .Burns & Roe . . . . . . . . . . . . .Black & Veatch. . . . . . . . . . .Brown & Root . . . . . . . . . . . .Ebasco . . . . . . . . . . . . . . . . .Gilbert/Commonwealth. . . .Gibbs & Hill. . . . . . . . . . . . . .Gilbert Associates. . . . . . . .Utility Owner. . . . . . . . . . . . .Fluor Power Services. . . . . .Sargent & Lundy. . . . . . . . . .Stone & Webster . . . . . . . . .United Engineers . . . . . . . . .Tennessee Valley

Authority. . . . . . . . . . . . . . .

CombustionWestinghouse General Electric Engineering Babcock & Wilcox

6 10 6 5—

—2 — — —4 4 —

2 — — —

7 — 6 —— — — —

8 7 — —

5 6 2 22 — — 2

3 6 2 2

SOURCE ffice of Technology Assessment

Chapter 4

THE NUCLEAR REGULATORYCOMMISSION’S ROLE

—

Chapter 4

THE NUCLEAR REGULATORY COMMISSION’S ROLE

The Nuclear Regulatory Commission (NRC)licenses all commercial nuclear reactors andmonitors them for safe operation. Thus, NRC isa natural agency both to promote standardiza-tion and to benefit from it. NRC recognizes theadvantages of standardization: from its view-point, it would expedite the licensing processand save staff time and attention; it would en-hance public health and safety; and it mightbenefit construction through the earlier availa-bility of final design documents and construc-tion experience.

A standardization program was first insti-tuted in 1973 by the former Atomic Energy

Commission. The program, with some changes,still operates under NRC today, as will be de-scribed further. Vendors, architect engineers(AEs), and utilities have participated in thestandardization program, however, it has hadonly marginal success in reducing Ieadtimes ormanpower efforts. At the present time, stand-ardization is accorded low priority at NRC. Infact, all licensing at NRC is at a virtual stand-still, not only because of decline of new plantorders but also because of the many uncertain-ties over the outcome of unresolved safetyissues. These topics will form the content ofthis chapter.

NRC’S

All plants currently

CURRENT STANDARDIZATION PROGRAM

must be reviewed atboth the preliminary safety analysis report(PSAR) stage and at the final safety analysisreport (FSAR) stage. Thus, both custom andstandard plant applications follow a two-stagereview process.

For the custom plant, the utility applicantmust submit a PSAR, including a general planfor the plant and many details about the par-ticular site. If the PSAR is approved, the utilityis granted a construction permit (C P). In thesecond stage, the utility applicant must file anFSAR that describes in greater design detail thereactor as it is actually being built. The FSARhas considerably more detail about the typesand characteristics of the actual componentsof the balance of plant (BOP) than does thePSAR. Acceptance of the FSAR and inspectionof the completed plant result in the issuance ofan operating Iicense (O L).

The licensing for a standard plant is also atwo-stage process but may take a shortcut byone of the following four methods. 1

Reference Plant Concept

Under this concept, a vendor or an AE firmmay apply for approval of an entire facility, ora major portion of it, outside the context of aparticular ut i l i ty appl icat ion. Once NRCreviews and accepts the reference systemdesign, it issues a preliminary design approval(PDA). The PDA can then be referenced by autility to build a specific plant at the CP stage.

A similiar procedure exists for the vendor orAE firm to obtain a final design approval(FDA), which can then be referenced in theFSAR submitted by the utility applicant at theOL stage. Once the FDA is issued for either anuclear steam supply system (NSSS) andlorBOP, the PDA is no longer needed. The way li-censing would then work is that the uti l itywould reference the FDA for a CP. The utility’sOL for the plant would, therefore, only requirean audit of the constructed plant to assurecompliance with the FDA. (The Iicensing of thePalo Verde plants in Arizona should give Com-bustion Engineering an FDA for its standardNSSS.)

‘Nuc lear Regula tory Commiss ion, “Rev iew of the Commiss ion

Program for S tandard iza t ion o f Nuc lear P o w e r P l a n t s a n dRecommendations to Improve Standardardizations Concepts, ”NUREG 0427, February 1978

33

34 Nuclear Powerplant Standardization

Duplicate Plant Concept

Under this concept, NRC receives a numberof applications for construction and operationof nuclear powerplants of essentially the samedesign to be built at different sites by one ormore utilities. Initially, the concept applied toapplications received within a few months ofone another. As modified, the concept allowsNRC staff to issue a preliminary duplicate de-sign approval (PDDA) for the first duplicateplant approved at the CP stage and a final du-plicate design approval (FDDA) at the OLstage.

Manufacturing License Concept

The manufacturing license concept involvesthe submittal of an application for a numberof identical nuclear powerpl ants which wouldbe manufactured at one location and movedto a different location for operation. An ap-plication for a manufacturing license is sub-mitted by a vendor and includes a report thatis similar to a safety analysis report (SAR) ex-cept that it is designated a design report. Theutility-applicant and site-specific informationare reviewed on each application that refer-ences the manufacturing Iicense application.

Ch. 4—The Nuclear Regulatory Commission’s Role ● 35

This concept is specifically applicable to theOffshore Power Systems (OPS) approach. Theapproach includes the NSSS and BOP, manu-factured in Florida, towed to a permanent sitefor mooring and connection to the electricalgrid.

Replicate Plant Concept

The replicate plant concept involves thesubmittal of an application by a utility appli-cant for a nuclear powerplant of essentiallythe same design as one in which the staff’sreview has resulted in the issuance of a safetyevaluation report. The nuclear powerplant pre-

viously reviewed by the staff is referred to asthe base plant, and the new plant is referred toas the replicate plant.

NRC has considered, but not yet imple-mented, a program by which a standard plantcould be reviewed only once before it is li-censed. This involves the concept of a stand-ard design approval (SDA). NRC staff believesthat single-stage licensing review is desirablefrom the standpoint of the public, industry,and NRC. The advantage of a single-stage li-censing review from the NRC staff’s viewpointis that it is based on more complete informa-tion and a single set of regulatory require-ments.

36 . Nuclear PowerPlant Standardization

From the utility applicant’s standpoint, con-struction can proceed on the basis of an NRCstaff-approved design that will not be sub-jected to a second review. From the public’sviewpoint, a more complete understanding ofthe facility is available at the beginning. ln-tervenors should be able to frame more specif-ic contentions based on the more detaileddesign.

One problem in the formulation of any sin-gle-stage licensing procedure is that the AEfirms would have difficulty in supplying, at anearly stage, the level of detail typical of anfinal design. As one AE firm put it, they believeone-stage Iicensing:

. . . can be an effective tool in increasing li-censing predictability if executed at the prop-er level of detail so as not to tie up the AEs, theutilities, and the regulators in paperwork thatwould result from the inevitable changes nec-essary to complete the design and construc-tion of a plant.

To make a single-stage review applicable tothe entire plant design, NRC would issue anSDA in lieu of a combined PDA and FDA. TheSDA concept involves the submittal of in-

EXPERIENCE WITH THE NRC

Under the reference plant concept there hasbeen quite a lot of activity for NSSSS, but thisis of marginal value for standardizing reactorsbecause the vendors’ NSSS systems are alreadyfixed in design. Five AE firms have submittedBOP designs under the reference system con-cept, but none of these have yet been used.

Under the duplicate plant concept, two ma-jor projects have been undertaken. With threeplants being planned at each of three sites,Duke Power has more experience with this con-cept than other utilities. Duke is also unusualin that it serves as its own AE. A consortium ofut i l i t ies (Standardized Nuclear Unit PowerPlant System (SNUPPS)) is also making consid-erable use of the duplicate plant concept.Originally planned at six plants, the group hasnow cut its number of plants planned under

formation that is significantly more developedthan that now provided for a preliminarydesign but somewhat less than that for a finaldesign. The SDA would of necessity be limitedin some areas to complete functional speci-fications rather than to actual design drawingsand specifications to avoid possible antitrustproblems with equipment suppliers. A supple-mentary NRC staff-audit function would be re-quired during plant construction to verify thatthe actual components—features installed orconstructed —adequately meet the approvedfunctional specifications. To date the SDAsingle-stage review concept has not been im-plemented.

The General Electric Co. (GE) has proposedthe similar concept of a “power-worthinesscertificate” (PWC), in analogy to the air-worthi-ness certif icate granted to aircraft by theFederal Aviation Administration. The majordifference between PWC and SDA concepts isthe scope of hardware licensed. The minimumscope for the PWC is the NSSS plus the otherradiologically significant systems and struc-tures. This is contrasted to the NSSS or B O Pminimum for the SDA.

STANDARDIZATION PROGRAM

the concept to three. Four other applicationsfor pairs of plants under the concept havebeen made.

OPS is the only applicant that has requesteda manufacturing license, which is an applica-tion for a license to manufacture eight iden-tical plants. The licensing process has beencompleted except for the new requirementsimposed by NRC as a result of the TM I acci-dent. This post-TMl review has been held up byNRC as it has for other pending CP. OPS pres-ently has no plants on order and probably mustobtain a manufacturing Iicense before it canaccept orders.

Initially, five util it ies applied for l icensesunder the replicate plant concept, but onlyone of these applications remains active. Sin-gle-stage licensing has not been implemented.


illustration credit General Electric Co

Cutaway view of the reactor building for the BWR nuclear island. The nuclear island containsthose structures of radiological significance

In summary to date, there has been partici-pation in NRC’s standardization program by allfour vendors, by five AE firms, and initially by10 utilities (though some have since canceled).It should be noted that all BOP constructiondone under NRC standardization programs hasbeen under the duplicate plant concept. Nodoubt most participants did so in the hope ofreducing Iicensing times, increasing predict-

ability that designs would be accepted forlicensing, and lessening construction costs.

By and large, it is too early to judge whethermany of these hopes will be realized. DukePower has encountered some difficulties withits duplicate plant efforts (see ch. 3, pp. 23-24).However, SNUPPS does report it has cut con-struction costs about 10 percent by stand-ardizing.


An NRC study of the standardization pro-gram revealed that savings in the effort neededto review applications, the primary objectiveof NRC’s program, have been minimal so far.2

The number of questions asked during a reviewis considered a key indicator of the difficultyof processing a nuclear plant application. Thestandard reviews over the period studied tookas many as 12.6 man-years and as many as1,060 questions, compared to 6.3 man-yearsand 700 questions for a custom design. Notehowever, that the standard design review in-cluded a review of the basic design; subse-quent reviews of referencing applicationsshould be expected to be much shorter. Datafor duplicate plant reviews indicate a substan-tial reduction in staff and industry effort.

In an interv iew of the NRC staff , OTAlearned that the staff strongly supports the cur-rent approach to standardization but hopesthat fewer standard plant designs than thenumber submitted to date will eventually re-sult from the efforts of NRC. The staff ex-pressed little or no support for a single stand-ard nuclear plant designed and supported byFederal Government agencies. They did ex-press great interest in single-stage licensing asrepresented by SDA but recognized that itposes greater problems to implement for BOPdesigns than for the NSSSS , mainly because ofthe traditional engineering procedures fol-lowed by the AEs and their utility customers.

The NRC staff feel that the nuclear power in-dustry would be improved by having both Gov-ernment and industry maintain a firm commit-ment to I imit changes to an approved standarddesign to those clearly needed for publichealth and safety reasons.

The staff fe l t that nonstandard des ignsresulted in confusion in understanding acci-dent conditions such as those experienced atTMI. If that plant had been a standard design,the accident could have been analyzed withfar less confusion and with more certainty.

NucIear ReguIatory Commission, op c it

Finally, the NRC staff believes that the pres-ent hearing process is a large impediment tothe full realization of the benefits of stand-ardization. Under present procedures, a stand-ardized design with safety features reviewed inpublic licensing hearings and accepted by theNRC staff can be reevaluated and perhapschanged in future hearings. This process ofadversary cross-examination may cause the in-dustry and regulators to perform better andmore thorough safety evaluations of proposednuclear powerplants than would be performedin the absence of public scrutiny. To reducethe opportunity for public rehearing wouldenhance efforts toward standardization at theexpense of public input.

The alternative in the NRC current standard-izations program is the submittal of a finaldesign by any qualified applicant to the Com-mission for rulemaking. 3 This allows NRC toreview and approve a plant design withouthaving received an individual CP application.Once approved, utilities could reference thisdesign and start construction after demonstrat-ing the acceptability of the proposed site. Aprocedure similar to this is being used for themanufacturing license although a final NRCreview and possible hearing will be requiredpr ior to the towing of the in i t ia l f loat ingnuclear plant from the manufacturing facility.Each site for a floating nuclear pIant must belicensed. Since a public hearing is mandatoryduring rulemaking on the design, later hearingsat the CP and OL stages would be limited toissues unrelated to the approved design (e. g.,envi ronmental impact and ut i l i ty compe-tence). Presently, no applicant has requestedrulemaking under the current standardizationprogram, probably because of the expense in-volved and the possible public perception thatrulemaking on a design was a means of bypass-ing the statutory requirements of the AtomicEnergy Act for public hearings.

‘Code of Federal Regulations, title 10, pt 50, app O (40FR-2977), Jan 17,1975


CURRENT STATUS OF LICENSING

S ince TMI, new plant orders have dis-appeared. Meanwhile, NRC has been under in-tense pressure to investigate or rule on manysafety issues. These two factors have com-bined to force standardization programs intothe background. Currently, such programs areunder the Standardization and Special Proj-ects Branch at NRC and are manned by a verysmall staff with virtually no budget. However,the work that must be done includes review ofCombustion Engineering’s application for anFDA, which is the first one to be requestedunder the reference system concept [the 3-unitPalo Verde application references the Com-bustion Engineering FDA). Another important

action awaiting the standardization branch isthe complete acceptance review of the appli-cation by GE in March 1980 for an FDA for itsnuclear-island designs. Six plants now underconstruction reference this design.

Although other matters do require heavydemands on its staff, NRC should be awarethat current steps must be taken, both to con-solidate the gains begun under the standard-ization program and to plan for a possiblefuture of renewed interest in nuclear power. Inparticular, NRC should be giving more atten-tion to the implementation of some form ofsingle-stage Iicensing.

NRC’S FUTURE ROLE

One criticism sometimes leveled at NRC isthat it is not disciplined or consistent in itsdecisions regarding which safety concerns aresufficient to warrant design changes or evenreactor retrofits. In d u s t r y observers, in par-ticular, worry that unless NRC is more disci-plined, reactors initially designed as similarplants may grow apart because of changingregulations. Adoption of a safety goal wouldcertainly help NRC arrive at consistent andmore predictable decisions regarding designchanges. As generally used now, the conceptof a safety goal — which might be either quanti-tative or qualitative— is the definition of anoptimum level of safety as a focus for the li-censing process. It would consider both indi-vidual and societal risk, and include somemethod of measuring the effectiveness of thesafety standards prevailing at any particulartime.

In the licensing for either custom or stand-ard plants, NRC has currently introduced an at-mosphere of uncertainty. Many safety issuesawait rulemaking by NRC. Until NRC rules onthe issues or unless NRC adopts interim crite-ria, nuclear plant designers will be uncertainhow to design a plant that can be licensed.

One example of these current safety issues per-tains to degraded cores. The objective of thedegraded core rulemaking, to commencesome-time during the second half of 1981, is todetermine whether fundamental changes arerequired in reactor design to prevent or miti-gate a melted core from penetrating the con-tainment and entering the outside environ-ment. NRC has not provided any interim guide-lines for what designs are acceptable until therulemaking is completed.

As another example, NRC has recently ruledthat applicants analyze all accidents of a cer-tain class (called “class 9“ accidents). Un-fortunately, NRC has not defined these acci-dents well enough or sufficiently narrowedthat class of accidents for them to be reason-ably analyzed.

These two examples are a few of many thatindicate that NRC is not managing its activitieseffectively at this time. Uncertainties or ambi-guities such as those mentioned will impede alllicensing— standard or otherwise. The adop-tion of a safety goal might facilitate the manydecisions NRC has to make.

Chapter 5

THE NUCLEAR INDUSTRY’SEXPERIENCE WITH STANDARDIZATION

Chapter 5

THE NUCLEAR INDUSTRY’S EXPERIENCEWITH STANDARDIZATION

THE NAVAL REACTOR PROGRAM

The Naval Reactors Program under Adm.Hyman Rickover has responsibility for 125 op-erating nuclear-powered ships, with 36 auth-orized or under construction. The U.S. Navyhas attempted to maintain as much standardi-zation as practicable, with particular emphasison the similarity of control rooms, instrumen-tation, operating procedures, and training pro-grams, All operators attend the same nuclearpower school, and manuals used for trainingare of the same type as those on shipboard.However, the specific propulsion-plant designsmay vary because of the different sizes andmilitary functions of the vessels they mustpower. The Navy’s nuclear-powered ships in-clude attack submarines, ballistic missile sub-marines, aircraft carriers, and cruisers. At least11 classes of ships are built or authorizedunder the Naval Nuclear Propulsion Program.Even within a single class, some variation hasresulted as new technologies develop andbecome installed on later models of a givenclass.

The designs are formulated at one of two,Government-owned, contractor-run labora-tories—the Bettis Atomic Power Laboratory(operated by Westinghouse) and the KnollsAtomic Power Laboratory (operated by Gen-eral Electric (GE)). Designers at these labora-tor ies must obtain the approval for thei rdesigns from Naval Reactors headquarters,and the designers are held responsible forstudy and improvements in the designs evenafter the reactor has been built. Presently, sub-marines are built at two commercial shipyards.

With regard to commercial nuclear power-plant standardization, Rickover made two ob-servations before the President’s Commissionon the Accident at Three Mile Island on July23, 1979. The first was on the desirability thatthe utilities “unite to establish a separate tech-nical organization which could provide a more

coordinated and expert technical input andcontrol for the commercial nuclear power pro-gram than is presently possible for each utilitywith its limited staff. ”

Among the things such an organizationcould do are:

●

●

☛

●

●

●

develop the standards and specificationsutilities should require for design and con-struction of their plants;provide direct, indepth technical assist-ance to utilities in design, construction,and operational questions;establish recommended staffing require-ments for operation of nuclear plants—e.g., at times there may be only a singleoperator with no supervision present inthe control room of an operating plant.Also, that operators may be assigned andactually carry out unrelated duties whileon watch. These are contrary to Navypractice;develop a comprehensive training and re-training program, including lesson plans,qualification requirements, etc., for utili-ties to use in training their people. Thismust be based on what is needed and notgeared solely to passing licensing exami-nations. It should cover al I types of per-sonnel, not just operators;provide trained technical teams to per-form periodic audits of nuclear stationsand critically evaluate the plants and thequalifications and performance of person-nel; andadvise utilities on technical safety ques-tions.

In the same testimony before the Three MileIs land (TMI) Commiss ion, R ickover recom-mended that “plant designs, equipment, con-trol rooms, training, etc., should be standard-ized insofar as practicable. ” Rickover notedthat much more standardization seems practi-

43


cal on new plants than old ones (where it mightnevertheless be possible to achieve some de-gree of standardization of control rooms, in-strumentation, etc.), and that standardizationshould have two distinct benefits. First, henoted, that bet te r des igns shou ld resu l tbecause a larger number of engineering man-hours could be applied to standardized de-signs, and, with a larger number of identicaloperating systems, operational experiencewould “provide a valuable source of informa-tion that can be used to improve the designand procedures and establish a more effectivepreventive maintenance program for a l lplants. ” Second, he noted, the use of standarddesigns would make it possible to train operat-ing and inspection personnel more effectively.

However, Rickover did not advocate themost extreme form of standardization. “In ad-vocating more standardization I am not sayingthat there should be one single design. I havestandardized in my program as far as practica-ble. Even then we have a number of designs tosuit the different power ratings and ship typesand to take advantage of new developmentsand technology which have become avail-able. ”

With regard to a new technical organization,the utilities have jointly funded the Institutefor Nuclear Power Operations (I NPO), which isundertaking to prepare models for operatortraining programs, and will establish trainingprogram criteria, accrediting industry trainingprograms, and performing in-plant evalua-

tions. INPO hopes these programs will be morespecific than those of the Nuclear RegulatoryCommission (NRC). The models will be recom-mendations, not requirements, for the utilities.Another collective organization funded by theutilities is the Nuclear Safety Analysis Center(N SAC), recently created by the Electric PowerResearch Institute to provide more technicalassistance to the uti l it ies. The commercialnuclear industry has, therefore, strengthenedits organizations along the lines suggested byRickover although none has the total authoritythat the Navy exercises over its reactors’ pro-gram. The benefit of these organizations is dif-ficult to judge because of preoccupation withthe implementation of the requirements result-ing from the accident at TM I and the shortlength of time (about 1 year) of their existence.Their success will depend on the quality of thepersonnel in the organization and the willing-ness of the utilities to accept their assistanceresponsibly.

With regard to standardized plant designs,the current ly available standard designsdocketed with NRC represent an improvementin decreasing the number of designs that arecommercially offered. A greater reliance on ajoint utility organization that sets design stand-ards and criteria that are more detailed thanthose in NRC regulations is desirable. The im-plementations of such a concept in the near fu-ture may be extremely clifficult due to the cur-rent high level of regulatory activity in areasother than standardization.

STANDARDIZED NUCLEAR UNIT POWERPLANT SYSTEM

The Standardized Nuclear Unit Power PlantSystem (S NUPPS) is a consortium of utilitiesorganized to build identical nuclear plants atdifferent sites across the country. It is theclosest project to full standardization withplants under construction. SNUPPS projectmanagement is handled by a contractual ar-rangement with Nuclear Projects, Inc., and ahierarchy of utility companies. The five utilitycompanies have entered into separate butnearly identical contracts with Bechtel Power

Corp. (lead architect engineers (AE)), Westing-house (supplier of the nuclear steam supplysystem (N SSS)), General EIectric Co. (supplierof turbine generators), and Nuclear Projects,Inc.

SNUPPS or iginal ly was a consort ium ofpower utilities that made an application for sixunits at four sites. One unit was withdrawnshortly after application. Another (the Sterlingunit) was canceled because of a lessening o f

Ch. 5—The Nuclear Industry’s Experience with Standardization ● 45

Photo credit Bechtel Power Corp

A mock-up of the equipment inside the containment is used to minimize problems with equipment layout and pipeor cable tray interference. It also serves as a planning aid during construction of the plant. The model seen here

is of SNUPPS. The long cylindrical vessels with the “J” shaped tube at the top are the steam generators.The reactor vessel mock-up would be surrounded by the portion of the containment seen here

demand, restrictions of local governments, anduncertainties in the Federal regulatory pro-cedures. The design and construction of bothCallaway 1 and Wolf Creek units are over 60-percent complete, however, both have suf-fered time delays and substantial cost in-creases. The time delays have resulted fromfinancial considerations and Federal regula-tory concerns, while cost increases have oc-curred primarily due to recent unusually highinflation rates.

The SNUPPS project is based on identicalunits with no shared systems. If two units wereto be constructed at the same site they wouldbe identical but separate units. For each plant,Westinghouse produced a standard informa-tion package in order that Bechtel could de-

sign and engineer the balance of plant withminimum changes to the NSSS. This approachfacilitates the orderly progression of designdrawings and the ordering of equipment.

All plants will have identical portions calledthe “power block”, this consists of the reactorb u i l d i n g ( c o n t a i n m e n t ) , f u e l b u i l d i n g , t u r b i n e

building, hot machine shop, auxi l iary bui lding,diesel generator building, control room build-ing, and radwaste building. Structures andcomponents outside the power block differ forthe various plants and are not under control ofSNUPPS .

In the l icens ing process, the project i smanaged by a single project manager and re-view team within NRC. In addition, the Ad-


visory Committee on Reactor Safeguards as-signed a subcommittee to review the standardportions of SNUPPS and when the AtomicSafety and Licensing Board hearing for two ofthe units were held, previously resolved issueswere not re-reviewed. This sharing of licensingresources allows more licensing personnel toprovide a greater indepth review than wouldhave been possible with a customized applica-tion for several plants. In addition, there was areduction in the questions asked by NRC froman average of 700 for a customized plant to anaverage of 150 per single SNUPPS unit.

During the procurement for the units, onlyproven materials, equipment, and systems areto be used; American National Standards in-stitute and other appropriate standards are tobe strictly followed. Power block purchasesare centralized — i.e., with few exceptions thesame supplier and the identical item for a par-ticular function are used for all plants. Thisallows interchangeabil ity of parts betweenplants. These are common industrial practices.

During construction, a considerable amountof standardization is maintained. Detailedmodels and photographs of the models of the

standard plants are used in the construction ef-fort. This method has eliminated much inter-ference and many delays while providing aconsiderable surety of proper constructiontechniques. Construction equipment commonto SNUPPS plants is shared by the constructioncrews.

Standard preoperational procedures, start-up, and functional operating procedures arebeing prepared for the SNUPPS plants. Also,simulators will be available for the SNUPPSpIants and operat ing exper iences wi l l beshared among SNUPPS utilities’ personnel.

The participants in the SNUPPS programclaim the SNUPPS plants wi l l be bui l t forabout 1(1-percent less than if they were custom-ized plants. Further, they feel the plants will besafer because of the standardization effort.However, there are no hard data to substanti-ate this claim, only an intuitive feeling that themore man-hours spent on a particular systemdesign the safer it will be. ’

‘Nlcolax A Petrick, “Progress Report on the SNUPPS NuclearStat Ions,” Nuclear Projects, Inc , Nuclear Englneerlng lnterna-tlonal, November 1977

THE FRENCH NUCLEAR PROGRAM

The French have developed a consensus ofgovernment energy policy makers that is sup-ported, almost totally, by all four major politi-cal parties. The French nuclear program hassome of the same problems as other nations —e.g., opposition by organized citizen groups,some difficult public relation situations, andsome technological shortfalls; however, theyhave maintained a firm commitment to theirpolicy of “tout nucleaire” (i. e., decommissioningoi l - f i red electr ic i ty generat ion plants andbuilding coal-fired, hydrostorages and mostlynuclear powerplants). The French policy wasformulated by their perception for the need toreduce dependence on foreign supply of oil(which in 1973 supplied France 67 percent ofits energy needs). Further, the French have onlyvery limited supplies of oil, natural gas, and

coal within their boundaries. The French con-dit ion i s qui te di f ferent f rom the UnitedStates— i.e., there is no clear political consen-sus on the need for nuclear power in theUnited States, partly because there is an in-digenous supply of oil, natural gas, and coalwithin U.S. borders.

The choice between the two commercialtypes of light water reactors —e. g. (boilingwater reactor (BWR) and pressurized waterreactor (PWR)) — using enriched uranium wasmade on the basis of price. The French enteredinto a competitive program between Europeanholders of l icenses for the manufacture ofAmerican designed plants. Framatome held aWest inghouse l icense and Alsthom, a GElicense. The latter group had a significant dis-

ch. 5—The Nuclear Industry’s Experience with Standardization ● 47

Photo credit Electricite de France

Four identical nuclear units are shown under construction in Blayais, France. The first unit, as seen in the background,is scheduled to produce electricity in 1981, just 6 years after construction started. By 1985 there will be

80 such units supplying 52,000 MWe for an area no larger than the State of Texas

advantage in the competition by the fact thatit does not own heavy forging facilities for re-actor vessel construction. For this reason, Als-thom either had to call upon their competitor,Framatome, or contract abroad. The BWR linewas therefore dropped not because of thePWRS technical superiority but to ensure a suf-f ic ient work load for the French indust r ia lgroup in charge of construction. The French in-dustry was, therefore, restructured into oneconstructor of nuclear steam supply compo-nents (Framatome) and one constructor of tur-bine generators (Alsthom). In addition, the na-tional electric utility is the only French AE,thereby making the standardization of nuclearpowerplants easier in France than it would bein the United States.

The French recognize four major safety-related advantages for standardization:

1.

2.

3.

4.

a more thorough investigation of safety-related matters is possible when multipleunits are involved;

experience in design, manufacture andconstruction, and operation can be trans-ferred from unit to unit;

more designer time becomes available tospend time working with a new generationstandardization series; and

regulators can spend more time inquiringabout site-specific considerations, theneed for new units, and the ability of theutility owner to operate the unit.

48 . Nuclear Powerplant Standardization

. .

Photo credit Electricite France

The Paluel site, Normandy, France consists of four 1,300-MWe units. The concrete walls ofthe containment and auxiliary buildings were erected during the early stages of construction.

The first unit should produce electricity y sometime in 1983

Also, the French recognize at least three ma-jor difficulties with standardization of nuclearpowerplants:

1. problems involved with one unit of a se-ries propagates to other units in the seriesand may require expensive and time-con-suming redesign and back-fitting;

2. site considerations may require substan-tial design differences between units of astandardized series; and

3. the optimal balance between design sta-bility and technological upgrading is dif-ficult to determine (i. e., a definition is

needed of safety enhancement or cost re-duction required before a new technolog-ical achievement can be incorporated .2

Overall, the French are satisfied with theirchoice and consider that the advantage ofstandardization (especially those related tosafety and economics) far outweighs those dif -ficulties.

‘MIChel Pecgner, “How,One Organization Runs the Whole in-dustry, ” Commissarat a L Energle Atomique (CEA), Nuclear En-gIneerlng International, December 1976

Ch. 5—The Nuclear Industry’s Experience with Standardization ● 49

THE WEST GERMAN OPERATOR TRAINING PROGRAM

A possible model for standardization oft rain ing and cert i f icat ion of personnel incommercial nuclear powerplants is the WestGerman program. The West Germans train andcertify their operators for both conventionaland nuclear powerplants in a powerplantschool called the Kraftwerkschu le. This is ajoint organization of owners of large power-plants with 116 members from six differentcountries. The primary purpose of the school isto provide professional and advanced trainingin six different technical areas for powerpl antpersonnel in maintenance and operation. Theschool was founded by a parent organizationcalled the Technical Association of PowerPlant Operators, formed as a result of a severeboiler explosion in 1920.

The formal training for a plantworker takes3 years and is integrated into the operation ofthe powerpl ant. Training consists of theoryand practice with a final exam for certificationin the operation of powerplant systems. Figure6 shows the progression for a nuclear plant-worker from initial certification by the Kraft-werkshule to shift supervisor.

The professional competence of the opera-tors and shift supervisor is regulated by officialgovernment guidelines. The minimum person-nel complement for a nuclear powerplant con-trol room is a shift supervisor, a deputy shiftsupervisor, and a powerplant reactor operator.The shift supervisor must be an engineer and

his deputy must be at least qualified as a Kraft-werkmiester (see figure 6). All three require ad-ditional special nuclear training including sim-ulator training, and practical nuclear power-plant experience.

As in the United States, the plant’s superin-tendent is responsible for the selection andtraining of the powerplant team. The superin-tendent assesses the workers’ capabilities anddetermines who will eventually be qualified asa plant attendant, plant operator, or shift su-pervisor. Unlike his counterpart in the UnitedStates, the West German superintendent pickshis candidates from a pool of workers whohave completed a standardized training pro-gram established by the owners of the power-plants under the guidelines of the government.In this country, the closest organization to theWest German program that has uniform train-ing and certification for its reactor operators isthe U.S. Navy. Many utilities rely heavily onthe Navy for qualified plant operators. This de-pendence can create manpower shortages inan area vital to the national defense andallows the utility to abrogate some of its re-sponsibility for a complete and total trainingprogram for new operators with no nuclear ex-perience.

‘0 Schwarz and G Schiegel, “Combining Theory and Practicein West German y,” NucIear Engineering International, March

1980

-


Figure 6.—Training Patterns for a West German Reactor Operator

Krattwerksmeister training

theoretical trainingat KWS 1580 h

II

.

Reactor Operator.,I

Krattwerker training

practical training inpowerplant 36 monthstheoretical training inpowerplant or at KWS590 h

M Mechanical engineeringElectrical engineeringMeasurement andcontrol instrumentationNuclear engineeringKraftwerksschule

SOURCE Nuclear Engineering International

Chapter 6

POLICY IMPACTS OF FOURAPPROACHES TO STANDARDIZATION

Chapter 6

POLICY IMPACTS OF FOUR APPROACHES TOSTANDARDIZATION

The s i tuat ion presented in the previouschapters is one of an industry which has beenslowly evolving toward a greater degree of sim-ilarity in its products. Stringent standardiza-tion is very difficult in the commercial nuclearindustry where the tasks of design, supply andconstruction, operation, and regulation are un-dertaken by multiple and often independentorganizations. Nevertheless, the designs for-mulated by the nuclear steam supply system(NSSS) vendors and architect-engineer (AE)firms are slowly converging toward a singledesign for each company. Several utilities andutility consortia have attempted to constructmultiple reactors based on a single design. TheNuclear Regulatory Commission (NRC) has for

some years defined special licensing for fourcategories of “standard” plants defined inchapter 4. These steps have been taken volun-tarily over a 10-year period, because the indus-try perceives they will produce lower costs,shorter l icensing times, and more reliableplants. Increasingly, both Government andindustry personnel have concluded that amore rapid move to standardization may in-crease the safety of nuclear plants. They alsorecognize that the industry will not move morerapidly toward standardization unless externalforces push it in that direction. Four represen-tative approaches to standardization are usedhere to provide a framework for this analysis.

FOUR APPROACHES

Acceleration of Present Policies. –An incen-tive program to accelerate the present trendsin the industry could reduce the number of de-signs substantially. In the first place, such aprogram could reduce the number of designsto one for each designer— i.e., 4 NSSS designsand 4 to 12 balance-of-plant (BOP) designs, de-pending on the number of AEs that remain ac-tive in the nuclear field. Only a few AEs havedeveloped BOP designs for the boiling waterreactor (BWR) and General Electric Co. ’s (GE)completed design for a nuclear island ap-proach based on the BWR make it likely thatfuture BWRS wil l be of one design. For thepressure water reactors (PWRS) produced bythe other three vendors, each AE would havebasically the same BOP design tailored tomeet the various interface criteria. Thus, thepossible number of different reactor plantswould be in the range of 5 to 13. The lowernumber could result if the util it ies agreedon design features and specific criteria for astandard BOP. Any AE could design a BOPconforming to these agreed-on criteria and theexisting regulatory requirements. NRC could

then offer one-step licensing for any utilityreferencing this “standard” in a license ap-plication. The time to implement this level ofstandardizat ion would equal the t ime toformulate the criteria and implement one-stepl icensing– about 1 to 3 years.

Procedural and Organizational Standardiza-t ion.– One advantage of standardizat ionwould be that it would allow personnel train-ing, operating procedures, terminology, etc., tobe specified in greater detail for a larger bodyof plants. Adoption of more universal prac-tices would allow operators of different plantsto learn more from the experiences of oneanother and would facilitate audits. Even with-out identical hardware, for the existing genera-tion of powerplants, the “software” practicescould be made more alike. NRC has somestandards for such pract ices and pr ivategroups, such as the Institute for Nuclear PowerOperations (IN PO) and the Nuclear SafetyAnalysis Center (NSAC), are currently evalu-ating operating practices with a view towardupgrading them. If the Government wished to

53

54 . Nuclear Powerplant Standardization

do more, a starting point would be to examineNRC’s current standards to see if they could bemore precisely specified and more universallyapplied. An examination of the West Germanstandardized training and certifications pro-gram for nuclear powerplant personnel dis-cussed earlier, might be appropriate.

Standardization of the NSSS Design Plus a Safe-ty Block. –One of the major reasons or stand-ardization is to allow more attention to a smal-ler number of designs and especially to safety-related systems (e. g., auxiliary feedwater andcontainment solation systems). One possibleapproach to standardization is to define thoseportions of the BOP that are necessary to bringthe reactor safely to a cold shutdown condi-tion and to allow only four variants (one foreach NSSS) of this so-called “safety block. ”Under this approach, the safety block wouldinclude 25 to 50 percent more equipment andhardware than the present NSSS. This versionof standardization represents a significant de-viation from the current mode of doing busi-ness and would require either a redefinition ofresponsibilities as now specified by NRC andperhaps some legislative action. To achievethis level of standardization, either the Gov-ernment or industry would have to define whatcomponents belong in the safety block, sub-ject the particular designs to some criteria ofsafety and reliability, and transfer responsibil-ity for them from the AE firms to the NSSS ven-dors or to a design team composed of both.

SAFETY

Almost all of the potential safety benefits ofstandardization are proffered on the basis ofintuition rather than experience. Few relevantexamples of standardization exist, and nonedemonstrate unambiguously that the safetyachieved results f rom the standardizat ionrather than from other factors — e.g., the safetyrecord of the naval nuclear reactors programprobably results as much or more from the U.S.Navy’s central control and other factors asfrom any similarity among its various reactorplants. Some of the arguments for the safetybenefits may break down in the extreme case

Critics of the approach suggest this transfer ofresponsibi l i ty for safety systems, normal lyunder the control of the AE firms, may place aburden on some of the NSSS vendors for whichthey are neither qualified nor prepared andthereby significantly alter the present structureof the nuclear industry. Some of the essentialsafety systems (e. g., the containment) requiredesign and construction skills for which theAEs are uniquely qualified. The safety-blockapproach is similar to that proposed by GE,with its “nuclear-island” concept; this wouldtake about 3 to 5 years to implement.

One Single-Standard Plant. –The ultimate instandardization would be to select only oneplant design according to which all future re-actors would be built. Such standardizationwould have to be accomplished by legal stat-ute and would completely alter the presentstructure of the commercial nuclear industry.To implement this concept of standardizationone must decide who would have overall re-sponsibility for the design, what the design cri-teria should be, and what would be the criteriaand time scale for incorporating modificationsinto the standard design. It would require from6 to 10 years to design and an equivalent timeto construct this single, national reactor.

Even for a single-design approach, site-spe-cific factors such as seismology, meteorology,and hydrology would require modifications insome of the reactor plants.

BENEFITS

of standardization — e.g., the one single-stand-ard pIant concept is seen by some as an oppor-tunity for a fresh objective look at commercialreactor design while it is viewed by others as adangerous commitment to a possibly flawed,single design. The following discussion is an ex-amination of the arguments in favor of stand-ardization and the extent to which these argu-ments apply to the four previously defined ap-proaches to standardization.

Enhanced Design Review. –Most people inthe nuclear industry or within NRC concur that

Ch. 6—Policy Impacts of Four Approaches to Standardization ● 55

the attention given to a particular designshould increase as the number of designs de-creases. The incentive program towards stand-ardization should allow more concentration ofattention within the designer firms. Movingtowards a safety-block concept or single-stand-ard design would primarily benefit regulatorssuch as NRC by greatly reducing the number ofdifferent reactors it would have to understandand regulate. Those advocating a single na-tional design feel that its major advantagewould be the design attention devoted to it.Designers could start afresh, yet benefit fromthe experience gained during the many yearsof operation with light water reactors (LWRS).Similarly, design attention to a safety-blockdesign may lead to a safer product. Oneshould keep in mind that the quality of atten-tion paid to a design is as important as thequantity of designers or safety analysts study-ing it. It is also possible that the reduction inthe number of reactor designs might merely re-sult in a proportional reduction in the numberof designers.

A design-review mechanism known as prob-abilistic risk assessment (PRA) has receivedconsiderable attention since the Three MileIsland (TMI) accident. The use of this tech-nique in assessing auxiliary feedwater systemreliability was discussed earlier in chapter 3.On a larger scale, PRA involves the steps ofidentifying hazards, hazardous activities andaccident sequences, and quantifying the prob-ability of accident sequences and the magni-tude of their consequences. The determinationof risk for a nuclear plant involves all parts ofthe plant and its operation. The NSSS, the BOP(e. g., the control room, containment, powerconversion system, and electrical systems),and utiIity-operator aspects (i. e., the operatingand maintenance procedures and the elec-trical grid), all are important in determiningoverall plant accident risks.

What sequences dominate r i sk can bestrongly dependent on the details of plantdesign and operation. Subsequent to the reac-tor safety study (RSS) (WASH-I 400) which con-sidered two reactors in detail, NRC sponsoredan RSS methodology applications program

which looked at four additional reactors.While the results of this work have not beenpublished, preliminary results indicate thatconsiderable differences in accident se-quences exist compared to the one consideredin WAS H-1 400. These differences are due to:

●

●

●

Not

safety systems un ique to the p lantstudied;safety systems performing functions dif-ferent than in WASH-1400; andmuItiple success options for a given func-tion requiring different levels of systemsuccess.

only were unique plant sequences found,preliminary results indicate that the dominantsequences vary from plant to plant.

Therefore, the major impact of standardiza-tion on probabilistic risk assessment would beto avoid industry manpower limitations in theevaluation of all plants to the degree neededto maximize plant reliability and safety. Thefewer number of plants needing evaluation thegreater the quality and detail of the risk assess-ment for a given amount of resources. In addi-tion, a greater understanding of the insightsparticular to risk assessment would be ob-tained. In retrospect, the RSS (WASH-1400)yielded considerable insight to the TM1-typeaccident (e. g., a small break, loss of coolantaccident), to the recent Browns Ferry partialscram and to the contributions of human er-rors to reactor accidents in general. If it wereapplicable to all reactors, these design prob-lems might have been anticipated and there-fore prevented by early corrective action.

Increased Awareness and Applicability of Op-erational Experience. –This possible safety ben-efit should be realized to various degrees forany of the four approaches to standardization.Naturally, the fewer the differences among re-actors, the more the overlap of experience.The accident at TM I provides a positive exam-ple, by which reactors of similar design havelearned to watch for a similar sequence ofevents. On the other hand, many incidents—

‘Nuclear Regulatory Commission, “Reactor Safety Study AnAssessment of Accident Risks in Commercial Nuclear Power-plants, ” NUREG-75/014, WAS H-1 400, October 1975


such as the Brown’s Ferry partial scram — arestill caused by specific piping or instrument er-rors which may be peculiar to that plant alone.

One central mechanism by which variousnuclear plant operators learn from the ex-perience of others is by the Licensee EventReports submitted to NRC. 2 The greater thes imi lar i ty among plants—even i f i t i s onlymore similar terminology or procedures—theeasier it should be to understand these eventsand to decide to which other plants they po-tentially relate.

There is no inherent reason why operators ofcustom plants should learn as much from oper-ating standard plants as other plants, but moreinterpretation is required to decide where eachexperience is relevant. It has been reportedthat an incident at the Davis Besse plant, was aprecursor to the TM I accident, but no warningwas issued. Standardization would not elim-inate such omissions automatically but couldease the burden of deciding which reportableevents were especially important to whichplants.

The feedback provided by the naval nuclearreactors program is a key element in the safetyof their program, and it is achieved despiteconsiderable variation among naval reactors.Currently, NRC and the industry are striving toimprove the feedback of plant experience.NRC has established the Office for Analysisand Evaluation of Operational Data. The Of-fice reviews all reportable events from reactorsand users of byproduct material. NSAC hascreated a communication and evaluation net-work used by operators of commercial reac-tors to inform one another of significant opera-tional occurrences.

Regardless of the organization, one of thedifficulties experienced with reviewing oper-ating data is that of interpreting the relevanceof a specific component failure at one plant tothe safety of another plant using a similar butnot identical component. The interpretationmay be easier if the component used is iden-

2 nuclear ReguIatory Commission, “Reporting of OperatingInformation – Appendix, A Technical Specification,” RegulatoryGuide 116 (revision 4), August 1975

tical in all plants, but the plants themselvesdiffer significantly. Experience to date hasshown that emphasis on feedback of operatingdata by the reactor vendors (GE, in particular)has markedly improved plant availability. Onecharacteristic of responsible plant manage-ment is its willingness and ability to identifyand to correct the generic or recurrent prob-lems underlying all unusual occurrences in itsnuclear powerplants. In a more standardizednuclear industry there would be no questionabout the importance of taking the broad viewof al l ident i f ied problems. A more stand-ardized industry would potentially permit arelatively small group of experienced engi-neers to review the data generated by oper-ating experience, looking for the generic im-plications of apparently “random” failures. Atpresent, the heterogeneous nuclear industryprovides generic assessment of operating expe-rience by means of various user groups. Ex-amples include the BWR Mark I containmentowners’ group; and the GE, Westinghouse,Babcock & Wilcox, and Combustion Engineer-ing owner’s groups; The formation of thesegroups results in part from an interest in thefree flow of information on solutions to theircommon problems.

While increased standardization would fur-ther help in the identification and resolution ofsafety issues, it would also increase the risk ofsystematic oversight of potential problems. Asa matter of policy, electric utilities plan diver-sity into their generating mix, both fossil andnuclear, and among the several reactor de-signs. This course has been amply vindicatedby the many generic shutdowns that have oc-curred without loss of a major part of thenuclear generating capacity. A nonnuclearanalogy would be the obvious consequence ofhaving a standardized U.S. jumbo jet, such asthe DC-10, grounded when a generic engine-mounting defect is discovered. The degree ofnuclear standardization needed to produceopt imum benef i ts i s a subject for furtherevaluation.

The greatest increase in health and safetycomes from the review and evaluation of oper-ating and construction experience on one sin-


gle-plant design. However, the institutionalbarriers and the possibility of systematic over-sight of safety problems may outweigh anysafety benefits accrued through the feedbackof data on one “accepted” design. With regardto procedural and organizational standardiza-tion, the benefit acheived through uniformreporting and review practices can be easilyobtained with little if any disruption in the in-stitutions regulating and operating commer-cial reactors.

Improved Training for Plant Personnel

The impact of the approaches to standard-ization of improving plant training is easilyanalyzed by considering three of the conceptsunder one heading “hardware standardiza-tion. ” The order of increasing hardware stand-ardization would be:

1. acceleration of present trends;2. NSSS plus safety block; and3. single-plant design.

The other approach, procedural standardiza-tion, is considered by itself as the standardiza-tion of the management processes as distinctfrom hardware. In addition, other institutionalfactors not normally considered part of anidealized, formal training program must betaken into account.

The basis for the procedures for design, con-struction, and operation of a nuclear power-plant is the Code of Federal Regulations, in-dustry standards, and NRC’s rules and regula-tions. Each applicant for a license establishes aset of administrative procedures that imple-ment the letter and intent of these rules andregulations. For an operating reactor, one partof there administrative procedures deals withthe selection, training, and qualification of theplant’s employees —e. g., these procedures de-scribe the general employee training require-ments as well as those for technicians andoperators. Each member of the plant staff issubjected to some train ing with di f ferentdegrees of intensity and depth according tothe position fi l led. Currently, there is widediversity in the training programs resultingfrom the way the utilities interpret the basic re-

quirements when establishing their administra-tive procedures —e. g., the requirements for alicensed operator to requalify on a yearly basisinclude the performance of 10 major changesin the plant’s status from the operator’s con-sole. Some utilities meet the requirement bysimply counting the startups or shutdowns theoperator has performed over the past year.Others send the operator to a plant simulatorfor as long as 2 weeks for intensive retraining.New requirements resulting from the accidentat TM I have specified in detail the types ofmanipulations necessary for this requalifica-t ion.3 In addition, these manipulations will re-quire the use of a plant simulator.

Greater standardization in operator trainingprograms than what currently exists wouldease the administrative burden on implemen-tation and auditing of this new requirement.Also, the effectiveness of the requirement overthe next few years would be easier to judge ifthe change were made from training programswhich had more in common. Greater hardwarestandardization would make the detailed pro-cedural level of these training programs morealike but would be unlikely to increase their ef-fectiveness or ease the administrative burden.

Standardization of hardware would makeselected improvements possible in trainingplant personnel. One area in which this couldoccur is the use of plant simulators. A simu-lator consists of a mockup of the control roomwith indicators, gages, and other instrumentsand devices driven by a computer. The oper-ator’s manipulations of the switches in themockup are monitored by the computer ,which simulates the reactions of the plant onthe mockup instrumentation. If greater hard-ware standardization were used in the nuclearindustry, more plant operators could use thesame simulator and fewer plant-specific simu-lators would be needed. Standardization of thehardware would also increase the analyticalcapability of simulators to deal with off-nor-mal transients when a transient occurs at one

‘ N u c l e a r R e g u l a t o r y C o m m i s s i o n , N R C A c t i o n P l a n D e v e l -

oped as a Result of the T MI-2 Accident, N U R E G-0660, ” May1980


plant and operators at other plants need to betrained for possible reoccurrence of the sametype of event. Another benefit is that the in-corporation of an actual event into the simu-lator’s computer would be easier –e.g., all ac-tual transients could be incorporated into thesimulator without the necessity of incorporat-ing specific differences in plant operatingcharacteristics resulting from different de-signs. The difficulty encountered by the vari-ous vendors in simulating the TM I accident ontheir own simulators (as an aid to operatortraining) was an example of of this.

However, all of these advantages must beviewed in the context of the existing mix o fgeneration common to most utilities and re-gional differences in the utilities’ service areas.The additions of several nuclear powerplantsof s tandard des ign may not s impl i fy theutilities training program if the current pro-

gram is determined by the diversity in existingoperating units. Among most electrical utili-ties, any “standard” plant would be unique asa source of power generation because it wouldbe different from existing plants. It would com-plicate rather than simplify the existing train-ing program. Unless a utility makes a substan-tial use of a single design in its operatingsystem, the value of hardware standardizationin improving the utility’s training program willbe minimal.

Procedural standardization in personnel se-lection, training, and requalification may bedifficult if there are significant differences inState labor laws, union contracts, or State reg-ulatory requirements. However, consideringthe current generation mix of each utility, thisstandardization approach appears to be theeasiest to implement with substantial benefitsin personnel training.

RELEVANCE TO A NATIONAL SAFETY GOAL

The question of the need for quantitativesafety goals to ensure that adequate levels ofnuclear powerplant safety are achieved is alongstanding one. The Atomic Energy Act of1954 and the Energy Reorganization Act of1974 established the legislative basis for NRCregulation to ensure the safe use of commer-cial nuclear power. In response to the leg-islative mandate, NRC regulations require, as apart of issuing a nuclear powerplant construc-tion permit, that a finding be made that “theproposed facil ity can be constructed andoperated at the proposed location without un-due risk to the health and safety of the pub-lic’” and as a part of issuing an operatinglicense that a finding be made “that there isreasonable assurance that the activities au-thorized by the operating license can be con-ducted without endangering the health andsafety of the public. ”5

The principles used by NRC are based on a“defense-in-depth” approach to the plant

4CFR 1050, sec 50355CFR 1050, sec 5057

design. Reactor safety as practiced in accordwith these principles is defined in NRC’s regu-lations, safety guides, branch technical posi-tions, and related industry standards. Theseprovide an extensively documented licensingprocess that has helped the nuclear industry toachieve an impressive record with regard topublic health and safety. In this process, manysafety requirements and calculational meth-ods have been identified. Following NRC rulesestabl i shes that plants adequately meetspecific safety requirements and satisfy the re-quirements of the legislative mandate. This de-terministic process is based on implied butunstated probabilities. For instance, a quali-tative probabilistic judgment was made manyyears ago that the large rupture of a reactorpressure vessel in LWRS was unlikely enoughthat it did not have to be considered in thedesign. In the intervening years a quantitativebasis has been provided to support that quali-tative judgment, The NRC licensing process isnow considering other factors that arise fromaccidents of greater severity than the design-basis accidents (DBA). Consideration of such


accidents will require a different type of anal-ysis than the traditionally conservative ap-proaches taken in the assessment of DBAs. Theuse of PRA techniques is rapidly coming intouse for this purpose. Quantitative criteria foracceptable levels of risk, or safety goals, areneeded if all the benefits of PRA are to berealized. PRA is an acceptable quantitativemethod of showing compliance with a well–defined safety goal.

U.S. activities relating to the establishmentof a national safety goal are going on withinthe NRC, the Advisory Committee on ReactorSafeguards (ACRS), the nuclear industry ingeneral, and the national technical and scien-tific community. There are also internationalactivities in this area. Possible variations ingoal forms that have been considered include:single v. multiple goals, quantitative v. quali-tative goals, and individual v. societal goals. G

The goal-setting process can be divided into

6S Levine, “TM I and the Future of Reactor Safety, ” Atomic ln-

Cfu$trlal Forum International Publ{c Affairs Workshop, Stock-holm, Sweden, ] une 1980

two broad phases, the initial phase in which awide range of goal elements and alternativestrategies are identified, and the second phasein which the effort is directed toward winnow-ing down the elements and strategies for moreindepth analysis and decisionmak ing.

in demonstrating compliance with any safe-ty goal, a high level of confidence in therelated risk assessments will be necessary. Ahigh level of confidence will also be necessaryto achieve public acceptance. PRA techniquesare relatively new and there are too few ski I ledpractitioners for it to be applied routinely forreactor safety assessment. If design standardi-zation were to result in a large reduction in thenumber of designs to be reviewed, PRA couldbe applied more comprehensively to showcompliance with a safety goal. By the sametoken, as the development of PRA techniquescontinues, confidence in their application willincrease and the number of skil led practi-tioners will become very much larger. It maythen be possible to address a wider range ofdesigns and this aspect of standardizationwould be less important.

THE IMPACT OF STANDARDIZATION ON RESOLUTION OFGENERIC ISSUES

A December 1977 amendment to the EnergyReorganization Act of 1977 required NRC tosubmit to Congress a list of unresolved safetyissues and plans for their resolution. Progresson resolution is to be included in NRC’s annualreport to Congress. Prior to that, NRC haddeveloped task-action plans for a multitude ofoutstanding topics, many of which were notconsidered unresolved safety issues. 1 nJanuary 1979, NRC submitted a report to Con-gress identifying 17 unresolved safety issuesand their related task-action plans. 7 A more re-cent plan updates the status of these issuesand plans.8 The 17 issues are listed in table 6.

‘Nuclear Regulatory Commlsslon, “ Identification of Unre-solved Safety Issues Relating to Nuclear Powerplants, ”NUREG-O51O, January 1979

“Nuclear Regulatory Commission, “Task Action Plan for Unre-SOlVed Safety lssues Related to Nuclear powerplants, ”NUREG-0649, February 1980

As a result of the many investigations of theTM I accident, NRC published an action plan inMay 1980.9 This report contains actions to becarried out by each nuclear plant owner andthe NRC. One might consider these as genericsafety issues; however, they are resolved issuesin that specific action is called for. Also, theseactions are applied to all operating plants, aswell as those under construction. Thus, stand-ardization would not have changed these ac-tion plans.

As an example of the effect of standardiza-tion on a safety issue, consider item 1 of table6, “water hammer. ” The phenomenon is simi-

lar to the banging of steam-heated radiatorscommonly found in old homes or office build-

‘Nuclear Regulatory Commission, “NRC Action Plan Devel-oped as a ResuIt of the TM I-2 Accident, ” NUREG-0660, May1980

.


Table 6.—Unresolved Safety Issues

1. Water hammer2. Asymmetric blowdown loads on the reactor coolant

system3. Pressurized water reactor steam generator tube

integrity4. BWR Mark I and Mark II pressure suppression

containment5. Anticipated transients without scram6. BWR nozzle cracking7. Reactor vessel materials toughness8. Fracture toughness of steam generator and reactor

coolant pump supports9. System Interactions in nuclear powerplants

10. Environmental qualification of safety-related electricalequipment

11. Reactor vessel pressure transient protection12. Residual heat removal requirements13. Control of heavy loads near spent fuel14. Seismic design criteria15. Pipe cracks in boiling water reactors16. Containment emergency sump reliability17. Station blackout

SOURCE: Nuclear Regulatory Commission.

ings. Occurrences have been attributed torapid condensation of steam pockets, steam--driven slugs of water, pump startup with par-tially empty l ines, and rapid-valve motion.Much of the problem might therefore be re-solved by piping arrangement to assure filledlines and prevent steam pockets. This would,of course, be easier to resolve in standardizedlayouts as opposed to those of differing plantdesigns. Although there has been no release ofradioactivity outside the plant’s boundarybecause of a water-hammer incident, the fre-quency of such events and the potential safetysignificance of the systems involved causedNRC to consider the water-hammer problemsignificant. Were most plants of standardizeddesign, modifications to prevent recurrence ofmany safety-related problems could be carriedout more rapidly as fewer designs need be ex-amined.

Resolution of another issue, related to con-tainment emergency sump reliability, wouldalso be quicker if designs were standardized.Although NRC has issued guidance for con-tainment sump design and testing, there arestill concerns about blockage of sump filtersand loss of abil ity to draw water from thesump. With fewer designs to investigate, the

emergency sump reliability issue could be set-tled much quicker.

The previous discussion indicates that stand-ardization would have facilitated resolution ofsome of the unresolved safety issues and there-fore improved nuclear powerplant safety. Onthe other hand, there are issues that would beunaffected by standardization. For instance,the disclosure by Virginia Electric Power Co.that asymmetric loads in the reactor vesselsupports and vessel internals caused by a PWRpipe break could cause a safety problem, wasthe result of studies with computer codes usingmore detailed analytical models. In otherwords, advances in the state of the art un-covered a problem. In that case the discoverywould have occurred at about the same time inthe advancement of the technology, whetheror not standardization had been implemented.

Finally, several situations have occurredwhere s imi lar i t ies in plant standardizat ionresulted in many nuclear plants experiencing

the same problem — a lesser degree of similari-ty (i. e., less standardization) could have limitedthe number of plants involved. One exampleof this was the realization that hydrodynamicloads on the suppression pool associated withloss-of-coolant accidents and safety-reliefvalve discharge were not considered in thedesign of Mark I and Mark II BWR contain-ment. These loads affected 24 Mark I and 11Mark II plants. Another example is the BWRnozzle-cracking problem associated with feed-water systems of many BWRS of similar de-sign—18 of 21 units inspected had cracks infeedwater nozzles.

For the most part, these generic issues arosewhen operating experience or advances in thestate of the art uncovered a problem, a dis-covery which would have occurred at aboutthe same time in the advancement of the tech-nology, with or without standardization. Resol-ution of some of the issues would be expeditedif affected nuclear plants were more stand-ardized, while resolution of other issues wouldnot be affected had standardization beenmore prevalent.


STANDARDIZATION AND ANTITRUST

As noted in chapter 3, the AEs normallyenter into a contract with the utility to provideengineering services for the proposed nuclearpIant including procurement of material forthe BOP. However, the utility selects the NSSSfrom the four available vendors based on com-petitive bidding. The reactor, much like theturbine generator, is considered for the pur-pose of procurement as a large single piece ofequipment. The utility normally does not in-volve itself with the selection of the vendor’ssupplier o the r than to as su re they arequalified. In many cases, the vendor may havealready completed procurement through exist-ing contracts with its suppliers. On the otherhand, the BOP equipment and materials areprocured by competitive bidding for eachplant to satisfy the State agencies regulatingthe utilities.

In order to perform safety reviews of pro-posed nuclear plants, the NRC staff prefers tohave as much detailed design as possible. Thelevel of detail provided by the vendors is suffi-cient for this purpose, even before actual con-struction of the plant begins. However, the AEcannot supply as detailed a design as can theNSSS vendor because the procurement of ma-terial and detailed design work has generallynot been completed at the time the CP isissued.

The exclusion of any qualified supplier ofplant equipment due to licensing requirementsfor a standard design is a breach of antitrustlaw. Increasing the level of detail in design forthe BOP to the same level found in the NSSSwould exclude qualified suppliers from themarket place, due to the differences in busi-ness methods.

By taking into account the antitrust dueprocess in the setting of standards for plantsystems and equipment, the antitrust problemcan be eliminated. Due process in standards-making according to the Department of J us-tice includes:’”

10john l-l Sherrefleldt Department of JustIce, “Standards forStandards-Makers (Washington, D C Department of Justice,American National Standards Institute, March 1978)

●

●

●

●

adequate notice of the proposed adoptionof a standard;standards development meetings shouldbe open to the public;the standards-setting body should have anaffirmative obligation to seek consumerand small business opinion; andmembership on standards developmentcommittees should represent a balancedcross-section of all affected parties.

The development of standards which specifysufficient detail to perform a safety review byknowledgeable engineers under the aboveguidelines should be sufficient to satisfy theconcern over anticompetitive practices andprotect the health and safety of the public. Asubcommittee of the Atomic Industrial Forumis currently working on a proposed revision tothe current NRC guidance on information re-quired for a safety analysis report for single-stage licensing. In addition, at least two AEsand one vendor are considering similar pro-posals.

Of the four standardization approaches con-sidered, the continuation of present policieswith refinement already being considered bythe industry is the least likely to create prob-lems with antitrust. The safety-block conceptwould not create any more difficulties thanthe acceleration of present policies, althoughit would place more of the total plant underthe design control of the vendors to the exclu-sion of the AE. However, the AE’s role as anengineering services contractor would be af-fected since design work encompasses onlyabout 10 percent of the total cost of the facili-ty. The “national single design” could forceone or more NSSS vendors from the market-place. The specifications for the design couldbe written to allow the vendors to remain com-petitive suppliers under contract to the util-ity for equipment and systems. Each vendorwould have to evaluate its interest in the sup-ply business, based in part on the similarity ofthe national design components to its own.However, the single-design standardization ap-proach has the greatest antitrust problems due

62 ● Nuc/ear Powerp}ant standardization

to the reduction of the NSSS vendors to sup- tions of their product line from the nationalpliers and the possible exclusion of large por- single-plant design.

UTILITIES AND STANDARDIZATION

A utility which operates and maintains anuclear powerplant is uniquely responsible tothe Federal and State Governments for the pro-tection of public health and safety. In addi-tion, the uti l ity is responsible to the stock-holders for the efficient operation of the plantand the protection of plant investment inequipment and fuel supply (i. e., the reactor’score). These are not mutually exclusive goalsand measures which protect the core, increaseplant availabil ity, and protect the public.Because of this unique relationship betweenthe utility, its stockholders, and government,nuclear utilities should actively participate inthe formulation of any standard design or ap-proach to standardization.

Over the past 25 years, some utilities thathave purchased nuclear powerplants have hadminimum influence on their design due in partto the lack of expertise in nuclear design en-gineering. Therefore, these uti l it ies placedheavy reliance on the judgment of the AEs andvendors to protect their financial and regula-tory interests. Other uti l it ies, such as DukePower and Tennesee Valley Authority haveacted as their own designers and have main-tained a strong influence in the design andconstruction of their plants. It is also this lattergroup of uti l it ies which have maintained a

strong commitment to standardization as evi-denced by their recent construction record forduplicate plants. However, having only a fewutilities committed to standardization may notbe enough to reap its benefits if a resurgencein new plant orders occurs.

A utility organization could, over the next 2or 3 years, develop standards and criteria fornew plants which incorporate the cumulativeoperating experience of the industry. Thesecriteria should concentrate on safe, conser-vative designs and reemphasize the past prac-tice of simply meeting licensing requirements.This effort would result in a set of criteria foreveryone (e. g., designers, operators, and regu-lators) and lend consistency to their actions,Common, understandable objectives could beestablished which concentrate on the reali ssues of safety and rel iabi l i ty. The effortshould include input from AEs, vendors, andperhaps NRC. Inclusion of NRC should belimited to their role as regulators not designersor operators.

Once the criteria are set, standard designscould be developed. Future construction dock-ets could then be limited to these designs andthereby allow the marketplace to l imit thenumber. Single-stage Iicensing would be a con-siderable inducement to the whole process.

FEASIBILITY

Of the approaches to standardization con-sidered, the acceleration of present trends andprocedural standardization are the most feasi-ble to achieve. These approaches work withinthe existing structures and motivations of thecommercial nuclear industry. Organizationssuch as NSAC and INPO have already beenestablished as a result of the TM I accident andare in excellent positions to develop and pro-

mote these forms of standardization. In addi-tion, these institutions were established by theutilities and the utilities are solely responsiblefor their success or failure. Such utility organi-zations could fill the role described previouslyfor the development of design standards andcriteria. The burden for standardization shouldrest with the utilities as they are ultimatelyresponsible for commercial nuclear power and


also have the most to lose in the event of anaccident.

As discussed earlier, trends in the industryover the past 25 years have led to some stand-ardization. This trend can be greatly accel-erated by implementing single-step licensing(or NRC’s standard-design approval) and regu-lating the industry in a consistent well-definedfashion. The development and implementationof a safety goal would certainly assist theregulation of the industry. However, its ab-sence should not deter the development of thestandards and criteria necessary for the nextgeneration of nuclear powerpl ants.

Under the safety-block concept, the vendor,either alone or in conjunction with an AE,would develop and obtain regulatory approvalof a standard design which consolidates in asingle design certain parts of the plant whichtraditionally have been split between the ven-dor and the AE. This would enable one de-signer or design group to have total systemresponsibility for the entire nuclear part of theplant and to better anticipate the impact ofvarious events on the entire plant. This ap-proach would eliminate a number of interfacesthat create difficulties in design and licensing,s ince al l the systems crucial for l icens ingwould be inside the safety-block portion of theplant. Approval of the power-generating sys-tems should be wholly routine. The safetyblock approach should therefore facilitate thelicensing process and allow a more thoroughdesign approval to take place. In either case,the AE firms would retain the bulk of theirfunction. This concept would require the ven-dor and perhaps the AE to expand their scopeof design responsibilities and accept the result-ing additional liability, The utility, therefore,would have to accept a lower degree of in-volvement than under the acceleration of pres-ent policies.

The single-standard design would requirecreating an entirely new design organization.This has the very real possibility of disruptingthe existing institutions which design, con-struct, operate, and regulate nuclear plants.Given the possibil ity of replicating an un-detected safety flaw in all the plants of asingle-standard design and the necessity ofrelating operating experience to the mixed setof plants already in place, the safety benefitsof such an approach are doubtful. The single-design approach has the greatest problemswith antitrust as well. The existing AtomicEnergy Act would have to be drast ical lymodified to enforce this approach and wouldtransfer the incentive and responsibil ity fordesign improvements f rom the indus t r ia lparticipants, who now have the responsibility,to an umbrella design organization. There is noprivate industry in the United States that hasundergone such a radical change. The net ef-fect of imposing a single design on the utilitiesis impossible to judge.

An alternative approach is to have a sepa-rate body go ahead with the design of a “na-tional reactor” or “yardstick” design, evenwithout a commitment to actually build them.This exercise would allow a comparison withexisting designs and possibly would bring im-provements to them. Such a design would haveto recognize the problems associated withcombining components or systems in ways notpreviously done and without any operationalexperience base for its performance. Such ayardstick could more easily be achieved bytightening the existing criteria to meet theutilities requirements for availability, reliabili-ty, and safety. This yardstick could then beused outside the licensing and regulatoryframework to measure the relative weaknessesor strengths of existing designs.

o