Nuclear Powerplant Standardization: Light

Water 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 and

Insular Affairs and endorsed by the Senate Subcommittee on Nuclear Regulation to

evaluate the extent to which nuclear powerplants can and should be standardized.

The assessment provides the essential background material for a broad understanding of the nuclear industry, its institutions and their relationship to standardization.

Items presented in the report include specific examples of the current state of

nuclear powerplant standardization, and four different concepts of standardization

and their potential impact on safety. These concepts represent a wide range of approaches toward standardization and would entail greatly differing requirements

for industry and regulators as well as differing implications for safety.

We are indebted to the participants in the workshop, reviewers of the final report, and numerous other individuals who gave extensively of their time and talents

in support of this assessment. Also, the contributions of several contractors, who

performed background analyses, are gratefully acknowledged.

JOHN H. GIBBONS

Director
Reviewers of the Nuclear Power plant

Standardization Report

Robert Civiak

Congressional Research Service

L. C. Dail

Duke Power Co.

Peter R. Davis

Intermountain Technologies, Inc.

Robert L. Ferguson

Washington Public Power Supply System

James F. Mallay

Electric Power Research Institute

David Okrent

University of California, Los Angeles

Oswald F. Schuette

University of South Carolina

G. Thomas Seely

Fluor Power Services

Dee H. Walker

Offshore Power Systems

Abraham Wietzberg

NUS Corp.

E. P. Wilkinson

Institute of Nuclear Power Operations

Workshop on Nuclear Powerplant Standardization

Harold W. Lewis, Chairman

University of California

Myer Bender

Oak Ridge National Laboratory

Sidney A. Bernsen

Bechtel Power Corp.

Dale G. Bridenbaugh

MH B Technical Associates

Thomas Cox

U.S. Nuclear Regulatory Commission

Jesse C. Ebersole

Consultant

Darrell G. Eisenhut

U.S. Nuclear Regulatory Commission

Jerry Griffith

U.S. Department of Energy

Joseph M. Hendrie

U.S. Nuclear Regulatory Commission

Stan Jacobs

Stone & Webster Engineering Corp.

Edward O’Donnell

Ebasco Services, Inc.

John Raulston

Tennessee Valley Authority

Don Roy

Babcock & Wilcox Co.

A. Edward Scherer

Combustion Engineering, Inc.

Glenn Sherwood

Nuclear Energy Systems Division

General Electric Co.

William R. Spezialetti

Westinghouse Electric Corp.

Wayne Stiede

Commonwealth Edison

NOTE The reviewers and workshop participants provided advice and comment throughout the assessment, but they do not necessarily approve, disapprove, or endorse the report for which OTA assumes fuII responsibility

iv
Nuclear Powerplant Standardization Project Staff

Lionel S. Johns, Assistant Director, O T A

Energy, 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 contract personal

v
Contents

Chapter Page

Glossary . . . . . . . . . . . . . . . . . . . . . . . . . viii

Acronyms and Abbreviations . . . . . . . . . . ix

1. Summary . . . . . . . . . . . . . . . . . . . . . . . . 3

Principal Findings. . . . . . . . . . . . . . . . . . 4

Institutional Responses . . . . . . . . . . . . . 5

Current NucIearlndustry. . . . . . . . . . 5

Nuclear Regulatory Commission . . . . 6

Congressional Role. . . . . . . . . . . . . . . 6

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

3. The Nuclear Industry Today . . . . . . . . . 15

Light Water Reactors. . . . . . . . . . . . . . . 15

Fission Rate Control . . . . . . . . . . . . . . 17

Fission Product Containment. . . . . . . 19

Auxiliary Feedwater Systems . . . . . . 20

Decay Heat Removal . . . . . . . . . . . . 22

Control Room Design. . . . . . . . . . 23

Causes for Variations Among LWRS. 23

The Nuclear Power Industry. . . . . . . . 24

Vendors . . . . . . . . . . . . . . . . . . . . . . 25

Architect Engineering Firms. . . . . . . 25

Construct ion Companies . . . . . . . . . . 26

Industry Trends . . . . . . . . . . . . . . 28

4. The Nuclear Regulatory Commission’s

Role . . . . . . . . . . . . . . . . . . . . . . . . . . 33

NRC’ S Current Standardization Program 33

Reference Plant Concept. . . . . . . . . . 33

Duplicate Plant Concept . . . . . . . . . . 34

Manufacturing License Concept . . . . 34

Replicate Plant Concept . . . . . . . . . . 35

Experience With the NRC

Standardization Program. . . . . . . . . . 36

Current Status of Licensing . . 39

NRC’s Future Role... . . . . . . . . . . . . . . 39

5. The Nuclear Industry’s Experience

With Standardization. . . . . . . . . . . . . 43

The Naval Reactor Program. . . . . 43

The Standardized Nuclear Powerplant

System . . . . . . . . . . . . . . . . . 44

The French Nuclear Program . . . . . . 46

Chapter Page

6.

The West German Operator Training

Program . . . . . . . . . . . . . . . . . . . . . . . 49

Policy Impacts of Four Approaches to

Standardization . . . . . . . . . . . . . . . . . 53

Four Approaches . . . . . . . . . . . . . . . . . . 53

Safety Benefits . . . . . . . . . . . . . . . . . . . . 54

Improved Training for Plant

Personnel . . . . . . . . . . . . . . . . . . . . 57

Relevance to a National Safety Goal. . . 58

The Impact of Standardization on

Resolution of Generic Issues . . . . . . . 59

Standardization and Antitrust . . . . . . . . 61

Utilities and Standardization. . . . . . . . . 62

Feasibility . . . . . . . . . . . . . . . . . . . . . . . 62

Tables

Table No. Page

1.

2.

3.

4

5

6

Auxiliary Feedwater Systems . . . . . . . . 21

Nuclear Reactor Suppliers . . . . . . . . 25

Content of an NSSS Standard Design

Application. . . . . . . . . . . . . . . . . . . . 27

Architect Engineering Responsibility

for Nuclear Powerplants. . . . . . . . . . . . . 28

NSSS/AE Combination of Light Water

Reactors Under Construction or

On Order. . . . . . . . . . . . . . . . . . . . . . . . . 29

Unresolved Safety Issues . . . . . . . . . 60

Figures

Figure No. Page

1.

2.

3.

4

5

6

Boiling Water Reactor Core and Vessel

Assembly . . . . . . . . . . . . . . . . . . . . . . . . . 15

Fuel Bundles . . . . . . . . . . . . . . . . . . . . . . 16

Boiling Water Reactor . . . . . . . . . . . . . . 18

Pressurized Water Reactor . . . . . . . . . . . 18

Comparisons of Auxiiary Feedwater

System Reliability on the Loss of Main

Feedwater System . . . . . . . . . . . . . . . . . . 22

Training Patterns for a West German

Reactor Operator . . . . . . . . . . . . . . . . . . 50

vii
Glossry

Architect engineer.– A supplier of design and engineering services for construction projects (e.g.,

powerplants, office buildings, bridges, etc.).

Auxiliary feedwater. –A standby system used to supply the secondary (nonradioactive) side of

PWR’S steam generation with cooling water in

the event the main source of water fails.

Balance of plant.– The equipment, in addition to the

nuclear steam supply system, which is necessary to produce electricity from a nuclear powerplant.

Boiling water reactor. -A power reactor in which

water, used as a coolant and moduator, is allowed to boil in the core.

Control rod. –A rod or tube containing a material

that readily absorbs neutrons used to control

the power of a nuclear reactor.

Decay heat.– The heat produced by the decay of radioactive nuclides or fission fragments.

Fission.—The splitting of a heavy nucleus into two

approximately equal parts, accompanied by the

release of a relatively large amount of energy

and one or more neutrons.

Heat sink.– Anything that absorbs heat; usually part

of the environment such as a river, pond, or the

atmosphere.

Light water reactor.– A reactor which uses ordinary

water as opposed to heavy water as a moderator and/or coolant.

Megawatt. -A unit of energy production or consumption commonly used to describe the generating capacity of a powerplant.

Moderator.—A material such as water used in a

reactor to slow down high-velocity neutrons.

Nuclear steam supply system.–An arrangement of

equipment with a critical array of nuclear fuel

which creates high-quality steam for running

turbine generators.

Pressurized water reactors. -A power reactor in

which heat is transferred from the core to a heat

exchanger by water kept under high-pressure to

achieve high temperature without boiling,

Probabilistic risk assessment. –An approach to safety analysis which assesses undesirable consequences and their likelihood.

Radioactivity. —The spontaneous decay or disintegration of a unstable atomic nucleus accompan ied by the emission of ionizing radiation.

Reactor. -A device in which a fission-chain reactor

can be initiated, maintained, and controlled.

Safety goal.– A quantitative or qualitative target for

either reliability or unreliability (risk).

System.– An arrangement of equipment utilized in

a powerplant for a specific function (e. g., the reactor protective system).

Vendor.– The supplier of the design and much of

the equipment for the nuclear steam supply

system.

Vlll
Acronyms and Abbreviations

ACRS

AE

AEC

A F W

ASME

ANSI

ASLB

BOP

BWR

CP

DBA

EPA

EPRI

FDA

FSAR

GE

INPO

LE R

LWR

M W e

— Advisory Committee on Reactors

Safeguards

— 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 t

NRC

NSAC

NSSS

OL

OPS

O T A

PDA

PDDA

PRA

PSAR

Pwc

PWR

RSS

SAR

SDA

SIP

— 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 powerplant

system

TM I – Three Mile Island

T 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 transition. The results of the accident at Three Mile

Island (TMI) have introduced sufficient uncertainties into the industry’s licensing and safety

practices so that it makes it difficult, if not impossible, to get a new plant approved. At the

same time, the unexpectedly low-growth rate

that many utilities are encountering has deterred them from ordering any new nuclear

plants for the immediate future. However,

even zero growth of demand would require

some new replacement facilities by the early

1990’s to maintain the present generating capacity. If the uncertainties resulting from TM I

are resolved soon, the nuclear industry will

have a unique opportunity to reevaluate its

direction and practices.

One of the peculiarities of the way that the

industry has developed is that commercial reactors are built with an unusual degree of variability and diversity. Essentially every reactor,

with a few exceptions to date, has been custom-designed and custom-built. The fact that

almost every reactor is “one-of-a-kind” has led

to excessive difficulty in verifying the safety of

individual plants and identifying particular

problems in transferring the safety lessons

from one reactor to another. It may also account for the escalating costs and long leadtimes associated with nuclear powerplants.

Many of these problems can be alleviated if

the industry moves away from its “one-of-akind” practices toward a degree of standardization in its design, construction, operation,

and licensing practices. Several types of standardization are possible, and this report examines them. Some trends in this direction are already occurring; the present lull could be used

to lay the groundwork for future standardization.

A minimal level of standardization is the

adoption of criteria for performance, reliability, and general design principles. This type of

standardization is promoted by groups such as

the American National Standards Institute and

the American Society of Mechanical Engineers. At the other extreme, some fee I standardization means the selection of one complete nuclear reactor as the “standard” or

model, according to which al I other reactors

are to be built.

OTA evaluated four different approaches to

standardization of the present generation of

light water reactors (LWRS). These are:

●

●

●

The acceleration of present trends. This

wou Id entail revitalizing and streamlining

the Nuclear Regulatory Commission’s

(NRC) current standardization program

and emphasizing one-step I icensing.

The procedural standardization. This

means the use of universal “software

practices” such as common terminology

and format for plant procedures and similar requirements for the training of plant

personnel.

The standardization of the powerplant’s

nuclear systern and those systerns critically necessary for the safe shutdown of

the reactor— the safety-block concept.

This might include the development of

similar designs for auxiliary feedwater and

shutdown cooling systems.

The selection of a single standardized design resulting from a fresh approach integrating the past 25 years of operating experience from various reactors.

This report considers these four representative approaches to standardization and examines the major advantages and disadvantages of each concept.
4 ● Nljc/ear powerp/ant standardization

PRINCIPAL FINDINGS

Standardization can be an essential element

in maintaining a viable and safe program for

nuclear energy. There are relatively few plants

built as examples of the approaches to standardization considered in this report, but the

present trend in the nuclear industry is toward

greater standardization.

Standardization yields safety benefits that

are intuitive/y valid even if they cannot be demonstrated unambiguous/y. The common-sense

nature of this benefit and its widespread acceptance in the nuclear industry more than

counterbalance the paucity of data from the

few relevant examples. However, the extreme,

“single-design” approach to standardization

could pose so many institutional difficulties

and generic risks, that the problems would outweigh the safety benefits.

Standardization has c/ear potential for time

and cost reductions and for gains in safety for

new nuclear p/ants. Several utilities and utility

groups have attempted to build standard

plants in the hopes of shorter licensing time

and reduced design and construction costs.

Some improvements have been reported but

there have also been problems.

Standardization is not a panacea, and the

other elements needed for a safe and efficient

nuclear program should not be ignored. Other

elements include safe operating practices, programs for effective preventive maintenance,

and direction by responsible technical managers.

Standardized plants constructed during different time periods have diverged from their

original design due to the changing regulatory

requirements, industrial standards, and utilities’

preferences. The characteristics of different

sites have dictated further divergence from

original standardized designs.

The quality of the implementation of standardization is just as important as the concept

itself in reaping potential benefits. A custom

plant can be safer than a standard plant if it is

operated and maintained in an exemplary fashion. Conversely, a standard plant will be safer

only if the designers and operators are highly

motivated, talented, and technically competent.

The present trends of the industry toward

greater standardization will be great/y encouraged by the implementation of sing/e-stage

licensing. Proposals have already been made

for the one-step issuance of a standard design

approval or “power-worthiness certificate” for

nuclear plants, but they have not been implemented.

NRC is current/y devoting little time to the

problem of nuclear powerplant standardization. The implementation of the rules and requirements resulting from the accident at TM I

is occupying much of NRC’s time. If standardization is to succeed at all, NRC must start

planning for it now during this period of slack

growth in nuclear power. They must develop

plans for future standardization, including

possible implementation of one-step licensing.

In addition, the vendors should realize that domestic orders for nuclear steam supply systems

(NSSS) may not occur over the next few years,

and they should take this opportunity to review and improve their basic designs.

The adoption of a national safety goal is

desirable. This would be a stated goal, agreed

on by society through some institution — Congress, NRC — as the level of safety acceptable

to the Nation. As such, it goes beyond the

more general statement in the present law. The

adequacy of NRC’s response to the accident at

TMI, in the absence of such a definition (i.e.,

how safe is safe enough), is impossible to assess and creates a large uncertainty in the licensing process. NRC must begin to manage its

activities in a manner so that prompt and consistent decisions on safety issues can be made.

Participants in the nuclear industry agree in

principle, on the desirability of a safety goal.

Enhanced standardization increases the likelihood of accurate risk assessment. The only

means to assure that a nuclear powerplant has

achieved a quantitative safety goal is through
Ch. l—Summary ● 5

the use of probabilistic risk assessment. lmproved risk assessment under standardization

is primarily due to the increased attention that

can be given to a few well-defined assessments

rather than many diversified ones.

The safety benefits of improved procedures,

through adoption of uniform reporting practices and industrywide participation in review

of operating experience, can easily be obtained

now. Substantial benefits can also be obtained

through standardization of training of plant

personnel, even when considering the utility’s

responsibiIity for a diversity of plant types.

The four approaches to standardization are

not necessarily mutualy exclusive and might

be explored in parallel. The first two approaches– acceleration of present trends and

procedural standard i z at ion — are already

being pursued but could be further encouraged. They can be accomplished with little, if any, disruption of the present structure

of the industry.

The second two approaches – the unification

of “safety -block” systems and the adoption of a

single “standard” plant design — could bring

about significant and perhaps disruptive

changes in the institutions of the commercial

nuclear industry, The safety-block approach

would transfer design responsibilities for certain safety systems — e.g., the containment — to

a section of the industry not traditionally

responsible for such systems. The single-standard plant approach would reduce the two major participants in the industry—vendors and

architect-engineers (A Es) — to suppliers of components and engineering services for the single

national design.

The second two approaches could establish

more specific design criteria than currently

exist and provide an “idea/” case for measuring

future design criteria. The purpose would be

served whether or not the more standardized

plant design was actually implemented.

The U.S. Navy’s experience with standardization is not directly applicable to the commercial nuclear power industry. The naval reactors

program is the only U.S. example of a wellstandardized program with considerable operating experience, but the principles applied in

this program are not directly applicable to the

commercial industry, which has a diversity of

designers, AEs, and operators who function

much more independently than the participants in the Navy program. The Navy’s safety

record is apparently due to strong central control and the greater attention that can be focused on a smaller number of reactor designs.

Standardization would aid the resolution of

some of NRC’s generic safety issues, while the

resolution of others would be unaffected.

INSTITUTIONAL RESPONSES

Current Nuclear Industry

The tasks of design, construction, and operation are handled by diverse and independent

organizations, each with its own distinctive

style and mode of business. The 75 commercial

reactors now operating in the United States

reflect this variety. However, in recent years,

the industry has begun to reduce this diversity

as designs have matured and to some extent

converged.

The two types of companies that together

design the systems of a nuclear powerplant are

the manufacturers of the NSSS and AE firms.

The four NSSS vendors design and manufacture the nuclear-related systems such as the

reactor vessel and core, primary cooling system, and reactor protective system. The AE

firms (which number about 12) design the balance of the plant, including the piping and

electrical layouts, auxiliary feedwater system,

and the containment building. Both the NSSS

vendor and the AE firm collaborate with the

utility to produce a plant that meets the utility’s specifications. In most cases, the AE firm

also serves as contractor for the pIant’s construct ion.
6 ● Nuclear Powerplant Standardization

In recent years, each NSSS vendor has

evolved basically one design for an NSSS

which varies little from one order to another.

The current variety of designs is due to the

larger number of AE firms than the NSSS vendors. Also, satisfying the different utilities’

design specifications creates additional variety. Some AE firms have moved toward one design with an interface package to match each

of the four NSSS designs. However, the designs

have not moved toward greater similarity from

one AE to the next. Standardization would

reduce the design effort of the AEs. This wou Id

not greatly reduce the total cost of the plant

since such efforts account for only a low fraction of the component and construction costs,

but it would affect the AE’s business. Nevertheless, AEs also serve as contractors and accept some form of standardization as inevitable and in the best interests of the industry.

The NSSS vendors, AE firms, and utilities

should continue to pursue a cooperative program of standardization, perhaps utilizing the

current trade associations. An alternative

would be the establishment of a joint utility

organization that sets standards and design criteria which are more detailed than the current

NRC regulations. Neither of these concepts

will become a reality as long as the industry’s

resources are stretched to meet NRC requirements resulting from the accident at TMI.

Nuclear Regulatory Commission

Since 1973, NRC has had a program for

licensing standard nuclear reactors according

to one of four definitions of standardization.

The industry and utilities have participated actively in hopes of a shorter and more predictable licensing process. The gains in time and

manpower effort have only been marginal to

date, although it may be premature to judge

the program’s success.

Industry observers believe that standardization will be hindered until NRC makes definitive rulings regarding which safety concerns

are sufficient to warrant a design change in a

standard reactor. Until that disciplined approach is achieved, no two “standard” reactors wilI remain alike.

The same basic criticism is leveled against

NRC in both its licensing and regulatory role

because it lacks clear direction for making

safety rules. A long list of generic safety issues

are before the Commission, and several key

safety issues await the Commission’s ruling.

The outcomes will remain unpredictable until

NRC establishes a safety goal to guide its decisions. Until regulatory and demand uncertainties are removed, no utility is likely to apply

for a new license — custom or standard.

Another step NRC might consider to encourage standardization is the implementation of

standard design approval, a concept for onestage Iicensing (the current procedure is a twostage process). NRC has considered the implementation of a standard design approval

which would involve submittal of information

that is significantly more developed than that

now provided for a preliminary design, but

somewhat less than that for a final design. The

General Electric Co. has proposed a similar

one-stage licensing program by which NRC

would grant a “power-worthiness certificate”

to an acceptable design.

Congressional Role

Although no legislation has emerged from

Congress that directs a standardization effort,

there remains considerable interest in whether

standardization can improve nuclear safety.

The findings of this study show that there is no

quantifiable demonstration that standardization enhances safety but there is a strong “intuitive” feeling that it will. The issue then

becomes the degree to which standardization

should be pursued considering the tradeoffs

between potential safety gains and possible

costs as summarized above and discussed in

this report.

If Congress chooses to pursue the third or

fourth approaches to standardization, legislation will probably be necessary because

neither the industry nor NRC will take these

steps voluntarily. If Congress decides that the
Ch. l—Summary ● 7

forces of the marketplace restrained by the

numerous industrial standards are sufficient,

then legislation mandating greater standardization is probably not necessary. Action that

supports this goal, either by legislation encouraging it or setting-up incentives such as

one-step Iicensing, could accelerate the trend

and provide a clear policy statement about

standardization and nuclear safety. In this connection, establishment of a nuclear safety

goal, by Congress, could be an important step

in encouraging standardization. Procedural or

operational standardization is also being pursued by the industry and utilities. Congressional legislation is probably not necessary to

achieve some degree of procedural standardization, but, again, could be encouraged by a

congressional statement of national policy.
Chapter 2

INTRODUCTION
Chapter 2

INTRODUCTION

The nuclear industry that has developed in

the United States since 19s9 has grown up with

a surprising degree of technical diversity. All

but a handful of the 72 plants that are currently licensed for operation have been custom-designed and custom-built. A result of this practice is that the plants must also be individually

licensed, since the safety analysis of each is inevitably different. When a utility decides to

build a plant, it usually first hires an architectengineering (AE) firm, then contracts with a reactor manufacturer (one of the four existing

“nuclear vendors”) to build the nuclear core,

vessel, and control mechanisms, which represent about 10 percent of the plant investment.

Each vendor has a different design for its nuclear system, so there are four different options. Then the AE designs the balance of the

plant (BOP):

● cooling systems;

● feedwater systems;

● steam systems;

● control room; and

● generator systems.

There are about 12 AEs presently designing

nuclear plants in the United States, and each

has its own preferred approach to these various systems. The AE’s approach will be tailored by past experience to be consistent with

one vendor’s nuclear system, but not necessarily compatible with the systems of all four.

In addition to the diversity due to the different

architect-vendor combinations, there is also a

degree of variability due to the different meteorological, seismic, and hydrological conditions at different plant sites.

Further variability is introduced by the

length of the process (12 years) and the piecemeal approach that is taken to both design and

licensing. Because safety standards have

grown up with the nuclear industry rather than

being formulated in full and fixed fashion

when the industry began, plant builders and

designers have taken a “design-as-you-go” approach to new plants in order to be able to

meet upgraded safety standards that might be

adopted during the period a pIant was under

construction. For some years, the industry’s

practice has been to start construction with

the design about 15-percent complete. On the

regulatory side, the approach taken — to accommodate changing safety standards due to

accrued experience and improved analysis —

has been to issue plant licenses in two steps, a

preliminary step sufficient to start construction and a final step necessary to start operation. Both of these practices have inevitably

increased the variation from one plant to

another. Even among plants intended to be

identical, but started at different times, significant design differences have occurred in the

final plants.

Reducing the diversity that now exists in the

nuclear industry would allow increased attention to be given to improving each plant design. It would also increase the amount of

operating experience that would be available

for a particular design and make it possible for

improvements at one plant to be immediately

applicable to an entire plant family.

Efforts to encourage standardization however, have met with slow acceptance. Some

argue that the many deviations from original

designs that now occur before plants operate

indicate that neither the technology nor the

licensing process is sufficiently stabilized to

support standardization. Furthermore, the nonstandardization that now exists in the industry

is a direct resu It of the diversity that exists in

the marketplace, and a substantial move toward standardization could result in some restructuring of the nuclear industry.

How substantial any move toward standardization should be is one of the topics of this

report. There is such a range of possible options that lie between the two logical extremes—that either all plants be different or

all be the same— that four different approaches to standardization merit discussion.

The different approaches represent greater degrees of standardization, the last option being

a single design identical to all others in both its

11
12 ● Nuclear powerplant Standardization

nuclear and BOP systems. The approaches differ in their technical, institutional, licensing,

and safety implications. Some require strong

legislative action, while others rely predominantly on trends already underway in the

industry.

This study was undertaken by requests of

the House Committee on Interior and Insular

Affairs and the Senate Subcommittee on Nuclear Regulation. Some committee members

expected that standardization would significantly improve the safety of the plants, and

help create a stable licensing process in which

utilities would have confidence that they

would get their reactors approved. The accident at Three Mile Island (TMI) contributed to

this expectation because both the local operators and the Nuclear Regulatory Commission

(NRC) personnel seemed to lack thorough

understanding of the reactor and had failed to

learn from similar experiences at related reactors. Ever-increasing licensing delays, especially since TMI, reinforce the need to reexamine

the merits of standardization.

Congress is not the only institution interested in standardization. NRC has also encouraged standardization although recent actions indicate that its priority at NRC has been

lowered. The NRC Advisory Committee on Reactor Safeguards has maintained a strong interest in the subject. The nuclear industry has

also been moving towards standardization as

individual companies have filed standardized

versions of their own designs with NRC. However, such efforts have been directed more at

unifying current practices than at maximizing

safety.

Any degree of standardization will require

decisions as to the level of specification required. The standard plants that have been

filed with NRC specify flow diagrams, design

descriptions, and generic information, but

does not include all the detailed information

required to actually build a plant. Complete

standardization would require considerably

greater efforts before a design is approved and

would allow considerably less flexibility afterwards, but wou Id result in making plants virtually identical.

This is not an exhaustive study of standardization. It is a broad scoping of four kinds of

standardizations that could be considered and

the major advantages and disadvantages involved in each. In addition, the study examines

the standardization of procedures and organizations to see if some advantages can be

gained without depending on new designs and

plants. The retrofitting of existing plants to enhance standardization or safety has not been

considered. OTA had staff and contractors

prepare background papers on NRC policy, the

U.S. Navy’s experience with standardization,

several plant systems that could be standardized, and the relation of standardization to

safety. These background papers were distributed to the participants of a 2-day workshop

held to identify and discuss the issues of standardization. The workshop included representatives of reactor manufacturers, AE companies,

utilities, regulators, and concerned observers

of nuclear power. This report is the result of

the background papers, the conclusions of the

workshop, and further information received by

the staff. It has been reviewed by the workshop

participants 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.). The

zation could be improved (costs, savings, and different heat sources used are the combustion

other benefits), present designs and the dif - of fossil fuel (e. g., coal, oil, natural gas) and fisferences 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 fuel

stations and results in similar equipment being steel cylinder (the reactor

used amongst facilities (e. g., feed pumps, gen- sustaining a controlled nuc

Figure 1 .—Boiling Water Reactor Core and Vessel Assembly

Vent and

head spray

Steam

out let

Core spray

inlet

Low-pressure

coolant

injection inlet

Core spray

sparger

Jet pump

assembly

Fuel

ssemblies

Jet pump

recirculation

water inlet

Vessel

support skirt

Control

rod drives

In-core

flux monitor

SOURCE: General Electric Co

77-538-

Steam dryer

lifting lug

Steam drye

assembly

Steam separator

assembly

Feed water Holddown

inlet ring

Feedwater

sparger CEA shroud

Core assembly

spray line

Top guide

Core shroud

Control

blade

Core plate

Recirculation

water outlet

Shield wall

Upper

guide

structure

Inlet

nozzle

$ 1

Active

core

length I II

Control rod drive

hydraulic lines

Instrumentation length assembly u 

core (fig. 1). The core

bundles (fig. 2) in a

vessel) capable of

ear reaction.

u Instrumentation I

u ~ _

nozzle

SOURCE Combustion Engineering, Inc

. Alignment

key

Core

support

barrel

Outlet

nozzle

Core

shroud

Fuel

assembly

Lower

support

structure

Flow

skirt

15
16 ● Nuclear powerplant Standardization

Upper I

plate

Figure 2.— Fuel Bundles

BWR Fuel

Fuel Expansion

bundle spring

Fuel

channel Plenum

Fuel rod spring

interim

spacer Getter

Fuel pellet

Finger

spring Fuel rod

(typical Lower tie plate

of “4) Noosepie~e

SOURCE General Electric Co

PWR Fuel

Alinement post

Upper

end-fitting

Spacer

grid

CEA

guide

tube

m assembly Top view

1- Fuel rod

T

Active

fuel

length

1 Bottom view

Lowerin9

SOURCE Combustion Engineering, Inc

Upper

end cap

Spring

Spacer

Dished

pellets

Fuel

cladding

Spacer

Lower

end cap

The fuel bundles consist of square arrays of

50 to 250 fuel rods about 1/2 -inch in diameter

and 12 feet long. Each rod is filled with 1/2-

inch-long fuel pellets containing slightly enriched uranium dioxide, and 200 to 500 fuel

bundles arranged in a circular array form the

core.

A nuclear reaction is initiated by the absorption of a neutron in the nucleus of a fissionable

atom (e. g., uranium-235, plutonium-239). The

fissionable atom splits, releases energy and

more than one neutron. These extra neutrons

are then available to produce more fissions

and continue the reaction and the release of

energy. This release of energy produces heat

within the fuel which in turn is released to the

cool ing water flowing through the core.

In a boiling water reactor (BWR), the type

shown in figure 3, this coolant is allowed to

boil. The steam thus produced drives a turbine,

which in turn yields electrical energy. In a pressurized water reactor (PWR), shown in figure 4,

the water that circulates through the core (the

primary coolant) is kept under pressure and

not allowed to boil. Instead, it transfers its

heat in a steam generator to a secondary cooling loop. Water in this steam generator then

boils, and its steam drives a turbine. In both

BWRS and PWRS, the steam emerging from the

turbine is discharged to the main condenser

where the steam condenses and the waste heat

is rejected to a heat sink such as a cooling

pond or tower, The condensed steam or water

then returns to the reactor vessel (in a BWR) or

to the steam generator (in a PWR) to begin the

cycle over again. The conversion of steam to

electrical energy with turbines and generators

is similar to nonnuclear steam electric stations.

The systems used in this conversion are referred to as power generation or nonsafety-reIated. The major systems required for the nuclear heat source — including some, but not all,

of the safety-related systems— are defined. by

the industry as the nuclear steam supply system (N SSS).

The byproducts from the fission process include unstable nuclei (fission products) which

decay to more stable nuclei by emitting an energetic 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 bundle

are two of the four vessel penetrations that route steam from the reactor to the turbine

ess produces heat at a much lower rate (several

percent of the fission process), but it continues

even after the reactor is shut down.

The fission process carries the unique problems of fission rate control, fission product

containment, and decay heat removal. Systems normally associated with these processes

are known as “safety-related” systems since

they are the ones depended on to prevent or

control accidents that could endanger the public. Several safety-related systems are discussed here with the purpose of understanding

the relationship of safety to standardization.

Fission Rate Control

The rate of the fission reaction is controlled

by materials that absorb neutrons without fissioning and, therefore, absorb the neutrons

available for fission. These absorbers are commonIy referred to as “poison s.”

The term “control rod” refers to a mechanical device containing an absorber with a fixed
18 ● Nuclear Powerplant Standardization

Figure 3.— Boiling Water Reactor (BWR)

Turbil

gener

ne

ato

Condenser

cooling

water

SOURCE: Atomic Industrial Forum, Inc.

Steam Steam line

Turbine

generator

●

Condenser

cooling

water
Ch. 3—The Nuclear Industry Today ● 19

Photo credit Atomiic Indusrial Forum, Inc

The major portions of the power conversion train are located within the turbine building. 1) Main turbine, converts

steam’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 is

soluble in water and added to the primary

coolant. I n pressurized water reactors these

soluble poisons are used in both safety and

power generation systems. In boiling water reactors they are only used in safety systems.

There are differences in designs between reactor vendors in both the control rod and its

mechanical drive. PWRS use tubular control

rods that are inserted into the fuel bundle. I n

BWRS the control rod is in the shape of a cruciform which is inserted between fuel bundles.

In either case, the rod and its mechanical drive

are a “standard” design peculiar to each vendor.

If the fission rate increases above a predetermined level (greater than the rate at which

heat can be removed by the coolant), the fission process is stop

of the control rods

called a “scram”).

power excursions

systems and scram

“reactor protection

ped by the rapid insertion

this function is commonly

The systems that sense

o actuate the protective

 the reactor are called

system s.” These systems

have undergone a careful evolutionary design

change with changes in state-of-the-art electronics — e.g., one vendor has changed the system’s analog signal processor to one using

digital computers. Although these designs are

standard to each vendor, they have not been

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

Fission Product Containment

The radioactive fission products must not be

released to the environment in excess of

Federal regulations because they could harm

the general public and the plant’s personnel.

Several barriers exist between the fission

fragments 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.) containing

slightly enriched uranium dioxide
20 ● Nuclear Powerplant Standardization

● secondary containment (on BWRS and

some PWRS).

Each barrier is a backup to the one before in

the event of failure —e. g., failure of the fuel

rod as a boundary is mitigated by the reactor

vessel and associated piping. In addition, penetrations in the primary containment (e. g., for

ventilation ducts, piping, etc. ) have isolation

valves (normally two) which close automatically on signals indicating potential fuel failures. The barriers listed can generally be described as passive (e. g., the fuel rod has no active components), or active (e. g., the isolation

valves require motive power to shut and require process signals for automatic actuation).

During the Three Mile Island (TMI) accident

a hydrogen explosion caused a pressure pulse

that actuated the containment isolation system. The system’s sensors and relays changed

electrical states and signaled the containment

isolation valves to shut. The signal was of short

duration (4 minutes) and eventually cleared,

allowing the operator to “reset” the containment isolation system, thereby returning the

electric portion of the system to its previous

“standby” state. 1 On resetting, the containment isolation valves for the containment

sump opened, allowing contaminated water

from inside the containment to flow to the auxiliary building. This may have caused an inadvertent release of gaseous activity into the environment through the exhaust ventilation in

the auxiliary building. The simple resetting of

the isolation signals should not have caused

the containment valves to open.

A post-TMl requirement was to review this

problem and ensure that each containment

isolation system would not automaticalIy open

isolation valves when the initiating signal was

reset. ~ A review of a selected number of responses to this requirement shows that this was

a problem at some reactors but not at others.

‘ E Iectrlc Power Research Institute, NLIC Iear Safety AnalysIs

Center, “Analy$ls of Three Mile Island, Unit 2 Accident, ”

NSAC-I , j Ulv 1979

‘Nuclear Regulatory Commission, “NRC ActIon Plan

Developed as a Result of the TM I-2 Accident, ” VOI 1,

NUREG-0660, January 1980

This lack of standardization in containment

isolation systems required a detailed review of

each plant’s containment isolation system and

resulted in a unique fix for each similar problem that was discovered. The Nuclear Regulatory Commission (NRC), in turn, had to stretch

its limited resources to review each design to

determine whether or not a modification was

required. This lengthened the time and reduced the depth of the review.

Auxiliary Feedwater Systems

(PWRS only)

Auxiliary feedwater (AFW) systems are designed to remove decay heat when the reactor

is shut down but at high pressure (normally

greater than 400 lb/in 2

). The design criteria for

them are usually established by the NSSS vendor while the detailed design responsibility

usually rests with the architect engineer (A E).

AFW systems are required to be available on

loss of main feedwater. The inadvertent isolation of this system was a possible contributor

to the accident at TMI. Valves on the outlet of

the pumps were found shut and they isolated

the pumps from the steam generators. The operator eventualIy opened these valves (approximately 7 minutes into the accident). 3

In addition, the unavailability of a plant’s

AFW system is an important and significant

contributor to the overaIl risk of any particular

PWR. As mentioned earlier, the generation of

heat from fission products must be removed or

dissipated to ensure that the integrity of the

passive containment boundaries is maintained.

I n a PWR, the methods available at high reactor pressures for decay heat removal are the

AFW system; some PWRS are also able to use

an alternative method incorporating the high-

-pressure injection pumps. ’ The former method

is preferable because the AFW system is on the

nonradioactive side of the plant. The latter

method is often called “feed and bleed” and

may requ ire discharging radioactive primary

coolant 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
Ch. 3—The Nuclear Industry Today ● 21

was the primary heat removal mechanism during the initial phases of the accident at TMI. 5

In response to TMI, NRC conducted a detailed review of AFW systems in PWRS to identify deficiencies in existing systems by assessing their relative reliability under loss of main

feedwater. The results of a portion of the study

are presented in table 1 and figure 5.6 Table 1

shows the diversity in an AFW system for one

PWR vendor. Note that only one plant had

automatic system initiation and most plants

differ in the number of pumps of each type. A

direct result of this diversity is shown in figure

5, Quantitative reliability assessments on 33

existing AFW systems show there is a wide

spread in the likelihood that the AFW system

wilI fail on the interruption of main feedwater.

As with the primary containment isolation

problem, the design solutions to this problem

are many and have very few elements in common. In addition, the acceptability of the system is impossible to judge in the absence of a

specific reliability goal. Therefore, the design

solutions 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 of

Feeclwater Transient\ and Small Break L055 of Coolant Accidents

In Combustion E ngineerlng Design Operating Plants, ” NUREG06 ]5, J anuarv 1980

to arbitrary judgment. If these systems were

more standard than they are today, there

would not be such a wide divergence in reliability; therefore, mandated engineered fixes to

the design would be easier to implement and

review.

A reduction in the diversity of AFW system

designs alleviates the above-mentioned problems. Two items are encouraging in this area

and illustrate the industry’s progress toward

standard system designs. First, a review of existing standard designs supplied and docketed

by the AE’s show a marked increase in standardization of auxiliary feedwater systems compared to those in existing plants—design is

docketed when it is formally submitted to NRC

and the administrative process for review and

approval begins. Ten AEs have designed an

auxiliary feedwater system that is applicable

to all PWRS. Therefore, this results in a single

AE’s design that is applicable to all three

PWRS. A further step toward standardization

of auxiliary feedwater systems is the approval

by the American National Standards Institute

of a design standard for these systems. 7 This

7American National Standards Institute, “Auxilary Feedwater

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

1979

Table 1 .—Auxiliary Feedwater Systems

Number of

pumps/type

Plant AIE of drive Capacity

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

1 motor-driven Motor: 575 gal/rein @ 2,800 ft

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

per unit

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

Hill

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

Maine Yankee Stone & 1 steam-driven

Webster

2 motor-driven

Millstone 2 Bechtel 1 steam-driven

2 motor-driven

Palisades Bechtel 1 steam-driven

1 motor-driven

St. 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/in2g

Steam: 600 gal/rein @ 2,437 ft

Motor: 300 gal/rein @ (each) 2,437 ft

Steam: 415 gal/rein@ 2,730 ft

Motor: 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 of

i nit iat ion

Automat ic

Manual

Semiautomatic

motor-driven

Pump manually

connected to diesel

generator

Manual

Manual

Manual

Manual

SOURCE Nuclear Regulatory Commlsslon
22 ● Nuclear Powerplant Standardization

Figure 5.—Comparisons of Auxiliary Feedwater

System Reliability on the Loss of Main Feedwater

System (LMFW)

Plant

3 <

SOURCE: Nuclear Regulatory Commision.

standard was approved late in 1979 and took

about 3 years to develop through the “consensus” process. As encouraging as these items appear, they lack the quantitative reliability criteria needed to remove the arbitrariness in regulatory judgments regarding their adequacy.

Decay Heat Removal

At low-reactor pressures (less than 400

Ib/in2

), redundant methods of decay heat removal prevent the uncontrolled heatup of the

core. The systems remove decay heat by continuously circulating water through the core

and rejecting the heat through heat exchangers

to the ultimate heat sink (e. g., cooling tower,

pond, lake, etc.),

The heat removal function operates in two

modes: 1 ) “emergency core cooling” during accident conditions, and 2) normal “shutdown

cooling” when the pIant is not producing electricity. In the emergency core-cooling mode,

the systems operate automatically to provide

cooling. In the shutdown cooling mode, the

operator sets up the system manually in accordance with the procedures for shuttin g

down the plant. The design responsibility of

these systems rests with the vendor. There is

very little difference between plants of the

same vendor. For light water reactors (LWRS)

there are four basic residual heat removal

designs which are standardized. These designs

all comply with the “general design criteria, ”

which are part of the Federal code (10 CFR)

governing the design, construction, and operation of commercial reactors.

However, critics of these designs have

pointed out that, due to the lack of specificity

in the requirements, the fundamental problem

of decay heat removal during the normal shutdown cooling mode has been overlooked. n instead, the operator is required to use his wit

and ingenuity to overcome built-in design complexities for the simple purpose of removing

decay heat during plant malfunctions when a

loss-of-coolant accident does not occur. A

well-publicized example of this is the Brown’s

Ferry fire where decay heat removal depended

on nonsafety-related equipment arranged in a

manner not previously considered necessary

for shutdown conditions. Even though these

standardized residual heat-removal systems

exist for both PWRS and BWRS and conform to

the existing design criteria, their adequacy

under nonaccident conditions is questionable.

In fact, the West Germans have added to their

American-designed PWRS an extra “bunkered”

decay, heat removal system independent of

the safety-related systems used during loss-ofcoolant accidents.

As the various criteria for decay heat removal illustrate, NRC’s general design criteria (supplemented by the existing standards and regu-

 “Common Mode Failure of Light Water Reactor

Systems. What Has Been Learned,” for Energy Analysis,

May 1980
Ch. 3—The Nuclear Industry Today ● 23

Iations) may not be adequate for routine operations during adverse plant conditions (e. g., a

plant fire). Some suggest this deficiency results

from the lack of specificity in the criteria.

Therefore, standardizing designs, without increasing the level of detail in the criteria and

accounting for past operating experiences,

may not make future standard plants any safer

than the existing operating ones. New NRC

rulemaking actions in the wake of the accident

at TM I point this out.

Control Room Design

Because the accident at TM I highlighted

concern over operator error, greater attention

is being placed on the control room design. I n

the past, control room designs have varied a

great deal from plant to plant. One reason was

the considerable input from the utilities, which

have preferred to maintain a degree of similarity between their nuclear plants and other types

of power-generating plants. Even before the

TMl accident, control room designs for future

plants incorporated some of the following

features: 9

●

●

●

●

consideration of functional grouping of

the reactor control panels;

location and layout of individual controls

on each panel in a logical common sense

manner;

compliance with regulatory criteria for

separation and instalIation of safety-grade

control equipment; and

utiIization of state-of-the-art computer

and display technology to aid the operator in the evaluation and control of the

plant’s condition.

Since TMI, NRC has required all operating

reactor Iicenses and applicants for operating

licenses to perform a detailed control room

design review to identify and correct deficiencies. 10 These reviews, which are expected to

take 1 year, are to include, among other things,

4“Surmortlng ., Intormatlon tor the Background Papers o n

Nuclear Powe~plant Stancfardlzatlon, ” submitted to the Off Ice

ot Technology Assessment, September 1980

‘

(] Nuc]ear Regulatory Commission, “NRC Act Ion Plan

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

an assessment of control room layout and consideration of human factors that influence operator effectiveness.

These requirements may indirectly lead to

greater standardization of control room designs. Three of the four vendors offer specially

designed control rooms that incorporate

“human engineering” features. Most recent

control room designs by AE firms incorporate

some human engineering. Utilities are likely to

find it too expensive to custom-design their

own control rooms for new facilities. Thus, the

number of different control room designs

should be reduced in the future.

Causes for Variations Among LWRS

Aside from the two major types of light

water reactors (BWRS and PWRS), there are

many possible variations in design ranging

from minor deviations in piping layouts to

different numbers of steam generators to the

different types of heat sink. Some of the major

variations stem from the varied design s

evolved by the three vendors supplying PWRS

and from the range of reactor sizes desired to

be built –e.g., Westinghouse reactors all have

one standard loop design for the primary coolant system, and plants of different sizes result

by including two, three, or four such loops in

parallel. By contrast, Combustion Engineering

and Babcock & Wilcox have two loops in every

plant but meet different power requirements

by varying the size of pumps and steam generators.

Other variations in design result from sitespecific factors. Reactors built in regions subject to earthquakes must be designed for

higher reaction loadings for such features as

the containment structure and mechanical and

electrical equipment. The meteorology associated with a particular site affects plant design

(possibly mandating a secondary containment)

because of concern over the patterns of dispersion of any radioactive gases released from the

plant. Flood and tornado hazards may also

have some effect on plant design. Duke Power

has cited an example of site-specific require-
24 ● Nuclear powerplant Standardization

ments that caused major divergences between

two standard units that were built at its

McGuire site and two units intended to be

identical, but built later at its Catawba site.

Differences between plant characteristics at

the two sites were forced by: 1 ) rulings of the

Environmental Protection Agency (EPA) and

NRC, and 2) changes in industry standards during the period of design.

EPA required Catawba to use cooling towers

rather than once-through-cooling. Cooling towers are less efficient. The additional power

consumed by the fans and the higher temperature of the cooling water in the condensers affected the design of other plant systems. The

overall power rating will be reduced from

1,180 to 1,145 MWe.

Duke Power also intended the decay heat removal systems for these standard plants to be

the same, however, the EPA ruling cited above

forced the Catawba decay heat removal heat

exchangers to be larger than those at MCGuire ire.

In addition, NRC took a new regulatory position requiring the Catawba units to have an independent suction from the reactor coolant

system for each of the two trains of decay heat

removal. The MCGUire ire units have a single suction supplying both trains. Finally, the industry

standards changed in the time period of design

of the four units, causing variation in the characteristics of such items as pumps and relief

valves. ’ These are a few of the many examples

of similar modifications. However, such design

changes may not be great enough to inhibit

some of the benefits of standardized plants.

1

“’Supporting Information for the Background Papers on

Nuclear Powerplant Standardliation, ” op cit

THE NUCLEAR POWER INDUSTRY

The major participants in the process of

designing and constructing a nuclear power

plant are the:

● electric utility;

● NSSS vendor;

● AE; and

● construction company.

The total number of companies involved may

well be in the hundreds, but these four

effectively control the major decisions.

If and when a utility determines that it needs

new control-station generating capacity, it

usually hires an architect-engineer firm to help

estimate costs and other considerations of the

various options. Eventually, alternative power

systems —e.g., solar, wind, etc. — may be considered, but at present few utilities have any

options other than coal or nuclear for large,

new power supplies. The cost comparison includes fixed-price bids from some or all of the

four NSSS vendors. The utility then contracts

with one of the NSSS vendors to supply the

nuclear components, and an AE firm (usually,

but not always, the same one) to design the

balance of the plant (BOP). The utility also

hires the construction company (often, but not

always, the AE firm). The AE and construction

companies work on a cost-plus basis since it is

impossible to predict in advance exactly what

level of effort will be required. In some cases

(usually large utilities), the utility may act as its

own AE or constructor or both.

The process outlined above and the participants described below represent the industry

as it operated several years ago. No plants

have been ordered for several years and few

are expected for the next few years. Some

changes may be expected if a resurgence of orders occurs, particularly if a policy of standardization is enforced. For instance, Offshore

Power Systems, a subsidiary of Westinghouse,

offers a complete nuclear powerplant. In this

case, the Westinghouse reactor is mounted on

a barge and sold to the utility complete with

all systems required to operate the reactor and

generate electricity. The only AE involvement

would be in site preparation. A somewhat simi-
Ch. 3—The Nuclear Industry Today ● 25

Iar scope of supply will be available at General

EIectric Co, (C E), which expects to offer a complete “nuclear island. ” The nuclear island

consolidates the GE BWR and auxiliary equipment into one standard design and includes all

of the buiIdings and structures that have radiological significance. Some AEs offer standard

BOP designs which interface with the standard

NSSS.

Vendors

Four companies manufacture NSSS for nuclear LWRS. These are listed in table 2 together

with the number of plants built and on order,

and the total generating capacity of these

plants. 12 GE makes a BWR while the other

three companies make PWRS. BWRS and PWRS

are clearly quite different facilities that will

call for quite different systems, components,

and layouts. However, the three PWRS are also

quite different. The number of loops for a

given power level may vary as can the size of

the reactor, the means of controlling it, and

the design philosophy of the systems servicing

it. All three PWRS are the end product of two

decades of somewhat divergent evolutionary

development. Even though conceptually similar, the engineering approaches to the various

design problems have been so sufficiently different that each NSSS is quite distinctive.

Since the NSSS is only one part of a large

complex of systems comprising a powerplant,

design of other systems may be assigned to

either the vendor or the AE at the discretion of

the utility owner. I n recent years however, a

1 2U S Department of Energy, “Nuclear Power Program information and Data, ” May 1980

uniform scope of responsibility has come

about through actions by NRC. When NRC

[formerly the Atomic Energy Commission) was

beginning to encourage standardization of nuclear powerplants in the early 1970’s, it formulated a detailed program for docketing standard plants for review and approval. The vendors at that time decided to limit their scope

of design responsibility to those components

which they planned to market as a standardized responsibility (i. e., those components that

were proving competitive). As a result, NRC

developed the list of systems shown in table 3

as the NSSS standard plant scope to be docketed by each vendor. Note that the list of

systems is largely the same for each vendor. 13

Architect Engineering Firms

The remaining systems necessary for a functioning plant are referred to as the BOP. Some

AE firms in accordance with NRC’s program

submitted standard plant designs for the BOP.

Each firm’s BOP design is matched to the NSSS

through “interface criteria. ” The BOP designs

vary from one firm to another, but each firm’s

BOP design is generally applicable to any PWR

by adjusting parameters (e.g., pressures and

flow rates) to meet the interface criteria.

BWRS require a separate class of BOP designs.

The NSSS represents about 10 percent of the

total plant, and the AEs design the remaining

90 percent. The cost of the plant design is

about 10 percent of the total plant cost. There

are also considerably more AEs than vendors,

1

‘Nuclear Regulatory Commlsslon, “Programmatic information 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 2,590

General Electric . . . . . . . . . . . . . . . . . . 24 17,758 28 30,101 7 8,304

Combustion Engineering . . . . . . . . . . 8 6,361 15 17,893 6 7,490

Babcock & Wilcox . . . . . . . . . . . . . . . . 9 7,885 8 7,947 3 3,790

Other . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 more

to the AEs. The AEs and their share of the

business is indicated in table 4.

Experienced AEs have preexisting designs

that can be tailored to specific site characteristics and utility needs. As will be discussed further, some of these designs have achieved sufficient maturity; the AEs have developed

standard plants for some or all of the NSSSS.

The use of such standard plants reduces the required design effort (which as stated before is

a moderate fraction of the total cost) and also

reduces the schedule and uncertainty of I icensing. These gains become questionable if the

utility insists on too many modifications to suit

its particuIar desires. Greater standardization

would affect the relationships of involved

“U S Department of Energy, op cit

firms by reducing the role of the

nearly to that of the purchaser of

utilities more

a stock item.

Construction Companies

With only a few exceptions, the companies

that build nuclear powerplants are the same

AE firms that design them. Thus, they will not

be identified separately here. The role of the

construction company is to build the plant according to the design and specifications of the

AE and the NSSS vendor. Theoretically, two

plants built to the same design would be identical, but in actual fact, minor differences

develop at the work site. A subcontractor may

deviate slightly from his blueprint due to unforeseen interferences, buildup of tolerances,

problems with field fits, or the unavailability

of a component. These changes are performed

under the supervision of a responsible engineer
Ch. 3—The Nuclear Industry Today ● 27

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

Babcock & Wilcox

A. Reactor

1. Fuel assemblies

2. Reactor vessel internals

3. Control assemblies

4. CRDMS

B. Reactor Coolant System

(including layout and

analysis)

1. Reactor vessel

2. Reactor coolant pump

3. Steam generator (not

beyond nozzles)

4. Main piping

5. Pressurizer (including

safety valves)

6. Pressurizer relief system

7. Inservice inspection

8. Equipment supports (not

including embedded

anchorage)

C. Emergency Core Cooling

Systems

D. Instrumentation and

Controls for the NSSSC

1. Main control room panel

board (including all

integral equipment)

2. l&C equipment racks and

panels

3. Reactor control and

protection systems

(including actuation

systems)

4. Nuclear Instrumentation

system

5. Process l&C (including

control valves)

E. Electric Powerc

1. CRDM power supply

2. Pressurizer heater

controls

F. Auxiliary Systems

1. Special handling

equipment for fuel and

reactor vessel internals

2. Makeup and purification

system

3. Chemical addition and

boron recovery system

4. Steam generator

circulating system

5. Decay heat removal

system

G. Startup Test Program for

NSSS Items

Combustion Engineering

A. Reactor

1. Fuel assemblies

2. Reactor vessel internals

3. Control element

assemblies

4. Control element drive

mechanisms

B. Reactor Coolant System

(including layout and

analysis)

1. Reactor vessel

2. Reactor coolant pump

3. Steam generator (not

beyond nozzles)

4. Main piping

5. Pressurizer (including

safety valves)

6. Inservice inspection

7. Equipment supports (not

including embedded

c.

D.

anchorage)

Emergency Core Cooling

Systems

Instrumentation and

Controls for the NSSSC

1. Main control room panel

board (including all

integral equipment)

2. l&C equipment racks and

panels

3. Reactor control and

protection systems

(including actuation

systems)

4. Neutron monitoring

system

5. Process l&C (including

control valves)

E. Electric Powerc

1. Control element drive

mechanism power supply

2. Pressurizer heater

controls

F. Auxiliary Systems

1. Special handling

equipment for fuel and

reactor vessel internals

2. Chemical and volume

control system

3. Shutdown cooling system

G. Startup Test Program for

NSSS Items

General Electric

A. Reactor

1. Fuel assemblies

2. Reactor vessel internals

3. Control assemblies

4. CRDMS

5. Control rod drive

hydraulic system

B. Reactor Coolant System

(including layout and

analysis)

1. Reactor vessel

2. Recirculation pumps

3. Recirculation piping and

MSL piping (including but

not beyond second

isolation valve)

4. Safety/relief valves

5. Inservice inspection

6. Equipment supports (not

including embedded

anchorage)

C. Emergency Core Cooling

Systems

D. Instrumentation and

Controls for the NSSSC

1. Main control room panel

board (including all

integral equipment)

2. i&C equipment racks and

panels

3. Reactor control and

protection systems

(including actuation

systems)

4. Nuclear instrumentation

system

5. Process l&C (including

control valves)

E. Auxiliary Systems

1.

2.

3.

4.

5.

6.

7.

Special handling

equipment for fuel and

reactor vessel internals

Standby liquid control

system

Reactor core isolation

cooling system

MSLIV leakage control

system

Reactor water cleanup

system

Residual heat removal

system

Pressure regulation

system

F. Startup Test Program for

NSSS Items

Westinghouse

A. Reactor

1. Fuel assemblies

2. Reactor vessel internals

3. Control assemblies

4. CRDMS (including missile

shield and ventilation)

B. Reactor Coolant System

(including layout and

analysis)

1.

2.

3.

4.

5.

6.

7.

8.

Reactor vessel

Reactor coolant pump

Steam generator (not

beyond nozzles)

Main piping

Pressurizer (including

relief and safety valves)

Pressurizer relief tank

Inservice inspection

Equipment supports (not

including embedded

anchorage)

C. Emergency Core Cooling

Systems

D. Instrumentation and

Controls for the NSSSC

1.

2.

3.

4,

5.

Main control room panel

board (including all

integral equipment)

l&C equipment racks and

panels

Reactor control and

protection systems

(including actuation

systems)

Nuclear instrumentation

system

Process l&C (including

control valves) -

E. Electric Powerc

1. CRDM power supply

2. Pressurizer heater

controls

F. Auxiliary Systems

1. Special handling

equipment for fuel and

reactor vessel internals

2. Chemical and volume

control system

3. Boron recycle system

4. Emergency boration

system

5. Residual heat removal

system

G. Startup Test Program for

NSSS Items

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 should 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 ,temsonly for the NSSS, not the Interconnecting plplng and cabling

dDeslgn Provlslons to accommodate Inser’vice lnsPectlon

SOURCE. Nuclear Regulatory Commlsslon
28 ● Nuclear Powerplant Standardization

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 7,494

Burns & Roe . . . . . . . . . . . . . . . . . . . . . 4 3,184 2 2,163 1 350

Black &Veatch . . . . . . . . . . . . . . . . . . . — — — 2 2,300

Brown & Root . . . . . . . . . . . . . . . . . . . . —

—

— 2 2,500

Ebasco . . . . . . . . . . . . . . . . . . . . . . . . .

— —

4 2,676 8 8,003 1 1,150

Gilbert/Commonwealth. . . . . . . . . . . . — — 3 3,310

Gibbs & 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 2,240

Stone&Webster . . . . . . . . . . . . . . . . . 9 5,859 11 10,797 4 4,800

United Engineers.. . . . . . . . . . . . . . . . 4 3,480 4 4,836

Tennessee 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 22,174

a

lncludes 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 is

the 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 arbitrarily changing the design. However, stringent

levels of sta

practices ar

construction

The roles

Idardization might frustrate such

d lengthen the time required for

of the plant.

Industry Trends

]f the utility and these three participants are not fixed. Some utilities do some or

all of the AE design work themselves; the Tennessee Valley Authority (TVA), Duke Power,

and American Electric Power are examples.

The utility is responsible for licensing, but it

can delegate the bulk of this task to the AE and

vendor if it chooses. Standardization would

tend to diminish utility involvement in licensing. AEs would also have a less pivotal role.

Current trends in standardization will be discussed in the following chapters, but it should

be noted from table 4 that the dominance of

several AEs may help ensure a certain degree

of standardization even in the absence of any

official action. Only four AEs (not counting the

utilities) have more than four projects underway: Bechtel, Ebasco, Sargeant & Lundy, and

Stone & Webster. The current experience and

expertise of these four (plus one or two others)

will likely attract utilities to them when and if

they begin to order new plants. Any resump-
Ch. 3—The Nuclear Industry Today ● 29

tion of orders is likely to be at a relatively slow

rate compared to the peak years of the late

1960’s and early 1970’s. These four to six AEs

could probably handle all the renewed business, and the utilities would most likely concentrate their orders on them. In that event,

the number of different possible combinations

of BOP plus NSSS would be sharply reduced.

Table 5 shows the present combinations of

NSSS vendors and AEs for LWRS under construction or on order. Instead of being 56 possible combinations, there are 22 NSSSIAE, plus 4

NSSS/TVA, and 2 other utility designs. ” If GE

succeeds in completing Iicensing its nuclear

island, if AEs having a smaller share of the

market are excluded, and if most of the remaining ones have approved standard designs,

the total number of combinations could be

less 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. . . . . . . . . . . . . . .

Combustion

Westinghouse General Electric Engineering Babcock & Wilcox

6 10 6 5

—

—

2 — — —

4 4 —

2 — — —

7 — 6 —

— — — —

8 7 — —

5 6 2 2

2 — — 2

3 6 2 2

SOURCE ffice of Technology Assessment
Chapter 4

THE NUCLEAR REGULATORY

COMMISSION’S ROLE
—

Chapter 4

THE NUCLEAR REGULATORY COMMISSION’S ROLE

The Nuclear Regulatory Commission (NRC)

licenses all commercial nuclear reactors and

monitors them for safe operation. Thus, NRC is

a natural agency both to promote standardization and to benefit from it. NRC recognizes the

advantages of standardization: from its viewpoint, it would expedite the licensing process

and save staff time and attention; it would enhance public health and safety; and it might

benefit construction through the earlier availability of final design documents and construction experience.

A standardization program was first instituted in 1973 by the former Atomic Energy

Commission. The program, with some changes,

still operates under NRC today, as will be described further. Vendors, architect engineers

(AEs), and utilities have participated in the

standardization program, however, it has had

only marginal success in reducing Ieadtimes or

manpower efforts. At the present time, standardization is accorded low priority at NRC. In

fact, all licensing at NRC is at a virtual standstill, not only because of decline of new plant

orders but also because of the many uncertainties over the outcome of unresolved safety

issues. These topics will form the content of

this chapter.

NRC’S

All plants currently

CURRENT STANDARDIZATION PROGRAM

must be reviewed at

both the preliminary safety analysis report

(PSAR) stage and at the final safety analysis

report (FSAR) stage. Thus, both custom and

standard plant applications follow a two-stage

review process.

For the custom plant, the utility applicant

must submit a PSAR, including a general plan

for the plant and many details about the particular site. If the PSAR is approved, the utility

is granted a construction permit (C P). In the

second stage, the utility applicant must file an

FSAR that describes in greater design detail the

reactor as it is actually being built. The FSAR

has considerably more detail about the types

and characteristics of the actual components

of the balance of plant (BOP) than does the

PSAR. Acceptance of the FSAR and inspection

of the completed plant result in the issuance of

an operating Iicense (O L).

The licensing for a standard plant is also a

two-stage process but may take a shortcut by

one of the following four methods. 1

Reference Plant Concept

Under this concept, a vendor or an AE firm

may apply for approval of an entire facility, or

a major portion of it, outside the context of a

particular utility application. Once NRC

reviews and accepts the reference system

design, it issues a preliminary design approval

(PDA). The PDA can then be referenced by a

utility to build a specific plant at the CP stage.

A similiar procedure exists for the vendor or

AE firm to obtain a final design approval

(FDA), which can then be referenced in the

FSAR submitted by the utility applicant at the

OL stage. Once the FDA is issued for either a

nuclear steam supply system (NSSS) andlor

BOP, the PDA is no longer needed. The way licensing would then work is that the utility

would reference the FDA for a CP. The utility’s

OL for the plant would, therefore, only require

an audit of the constructed plant to assure

compliance with the FDA. (The Iicensing of the

Palo Verde plants in Arizona should give Combustion Engineering an FDA for its standard

NSSS.)

‘Nuclear Regulatory Commission, “Review of the Commission

Program for Standardization of Nuclear Power Plants an d

Recommendations to Improve Standardardizations Concepts, ”

NUREG 0427, February 1978

33
34 Nuclear Powerplant Standardization

Duplicate Plant Concept

Under this concept, NRC receives a number

of applications for construction and operation

of nuclear powerplants of essentially the same

design to be built at different sites by one or

more utilities. Initially, the concept applied to

applications received within a few months of

one another. As modified, the concept allows

NRC staff to issue a preliminary duplicate design approval (PDDA) for the first duplicate

plant approved at the CP stage and a final duplicate design approval (FDDA) at the OL

stage.

Manufacturing License Concept

The manufacturing license concept involves

the submittal of an application for a number

of identical nuclear powerpl ants which would

be manufactured at one location and moved

to a different location for operation. An application for a manufacturing license is submitted by a vendor and includes a report that

is similar to a safety analysis report (SAR) except that it is designated a design report. The

utility-applicant and site-specific information

are reviewed on each application that references the manufacturing Iicense application.
Ch. 4—The Nuclear Regulatory Commission’s Role ● 35

This concept is specifically applicable to the

Offshore Power Systems (OPS) approach. The

approach includes the NSSS and BOP, manufactured in Florida, towed to a permanent site

for mooring and connection to the electrical

grid.

Replicate Plant Concept

The replicate plant concept involves the

submittal of an application by a utility applicant for a nuclear powerplant of essentially

the same design as one in which the staff’s

review has resulted in the issuance of a safety

evaluation report. The nuclear powerplant previously reviewed by the staff is referred to as

the base plant, and the new plant is referred to

as the replicate plant.

NRC has considered, but not yet implemented, a program by which a standard plant

could be reviewed only once before it is licensed. This involves the concept of a standard design approval (SDA). NRC staff believes

that single-stage licensing review is desirable

from the standpoint of the public, industry,

and NRC. The advantage of a single-stage licensing review from the NRC staff’s viewpoint

is that it is based on more complete information and a single set of regulatory requirements.
36 . Nuclear PowerPlant Standardization

From the utility applicant’s standpoint, construction can proceed on the basis of an NRC

staff-approved design that will not be subjected to a second review. From the public’s

viewpoint, a more complete understanding of

the facility is available at the beginning. lntervenors should be able to frame more specific contentions based on the more detailed

design.

One problem in the formulation of any single-stage licensing procedure is that the AE

firms would have difficulty in supplying, at an

early stage, the level of detail typical of an

final design. As one AE firm put it, they believe

one-stage Iicensing:

. . . can be an effective tool in increasing licensing predictability if executed at the proper level of detail so as not to tie up the AEs, the

utilities, and the regulators in paperwork that

would result from the inevitable changes necessary to complete the design and construction of a plant.

To make a single-stage review applicable to

the entire plant design, NRC would issue an

SDA in lieu of a combined PDA and FDA. The

SDA concept involves the submittal of inEXPERIENCE WITH THE NRC

Under the reference plant concept there has

been quite a lot of activity for NSSSS, but this

is of marginal value for standardizing reactors

because the vendors’ NSSS systems are already

fixed in design. Five AE firms have submitted

BOP designs under the reference system concept, but none of these have yet been used.

Under the duplicate plant concept, two major projects have been undertaken. With three

plants being planned at each of three sites,

Duke Power has more experience with this concept than other utilities. Duke is also unusual

in that it serves as its own AE. A consortium of

utilities (Standardized Nuclear Unit Power

Plant System (SNUPPS)) is also making considerable use of the duplicate plant concept.

Originally planned at six plants, the group has

now cut its number of plants planned under

formation that is significantly more developed

than that now provided for a preliminary

design but somewhat less than that for a final

design. The SDA would of necessity be limited

in some areas to complete functional specifications rather than to actual design drawings

and specifications to avoid possible antitrust

problems with equipment suppliers. A supplementary NRC staff-audit function would be required during plant construction to verify that

the actual components—features installed or

constructed —adequately meet the approved

functional specifications. To date the SDA

single-stage review concept has not been implemented.

The General Electric Co. (GE) has proposed

the similar concept of a “power-worthiness

certificate” (PWC), in analogy to the air-worthiness certificate granted to aircraft by the

Federal Aviation Administration. The major

difference between PWC and SDA concepts is

the scope of hardware licensed. The minimum

scope for the PWC is the NSSS plus the other

radiologically significant systems and structures. This is contrasted to the NSSS or B O P

minimum for the SDA.

STANDARDIZATION PROGRAM

the concept to three. Four other applications

for pairs of plants under the concept have

been made.

OPS is the only applicant that has requested

a manufacturing license, which is an application for a license to manufacture eight identical plants. The licensing process has been

completed except for the new requirements

imposed by NRC as a result of the TM I accident. This post-TMl review has been held up by

NRC as it has for other pending CP. OPS presently has no plants on order and probably must

obtain a manufacturing Iicense before it can

accept orders.

Initially, five utilities applied for licenses

under the replicate plant concept, but only

one of these applications remains active. Single-stage licensing has not been implemented.
Ch. 4—The Nuclear Regulatory Commission’s Role ● 37

illustration credit General Electric Co

Cutaway view of the reactor building for the BWR nuclear island. The nuclear island contains

those structures of radiological significance

In summary to date, there has been participation in NRC’s standardization program by all

four vendors, by five AE firms, and initially by

10 utilities (though some have since canceled).

It should be noted that all BOP construction

done under NRC standardization programs has

been under the duplicate plant concept. No

doubt most participants did so in the hope of

reducing Iicensing times, increasing predictability that designs would be accepted for

licensing, and lessening construction costs.

By and large, it is too early to judge whether

many of these hopes will be realized. Duke

Power has encountered some difficulties with

its duplicate plant efforts (see ch. 3, pp. 23-24).

However, SNUPPS does report it has cut construction costs about 10 percent by standardizing.
38 ● Nuclear Powerplant Standardization

An NRC study of the standardization program revealed that savings in the effort needed

to review applications, the primary objective

of NRC’s program, have been minimal so far.2

The number of questions asked during a review

is considered a key indicator of the difficulty

of processing a nuclear plant application. The

standard reviews over the period studied took

as many as 12.6 man-years and as many as

1,060 questions, compared to 6.3 man-years

and 700 questions for a custom design. Note

however, that the standard design review included a review of the basic design; subsequent reviews of referencing applications

should be expected to be much shorter. Data

for duplicate plant reviews indicate a substantial reduction in staff and industry effort.

In an interview of the NRC staff, OTA

learned that the staff strongly supports the current approach to standardization but hopes

that fewer standard plant designs than the

number submitted to date will eventually result from the efforts of NRC. The staff expressed little or no support for a single standard nuclear plant designed and supported by

Federal Government agencies. They did express great interest in single-stage licensing as

represented by SDA but recognized that it

poses greater problems to implement for BOP

designs than for the NSSSS , mainly because of

the traditional engineering procedures followed by the AEs and their utility customers.

The NRC staff feel that the nuclear power industry would be improved by having both Government and industry maintain a firm commitment to I imit changes to an approved standard

design to those clearly needed for public

health and safety reasons.

The staff felt that nonstandard designs

resulted in confusion in understanding accident conditions such as those experienced at

TMI. If that plant had been a standard design,

the accident could have been analyzed with

far less confusion and with more certainty.

NucIear ReguIatory Commission, op c it

Finally, the NRC staff believes that the present hearing process is a large impediment to

the full realization of the benefits of standardization. Under present procedures, a standardized design with safety features reviewed in

public licensing hearings and accepted by the

NRC staff can be reevaluated and perhaps

changed in future hearings. This process of

adversary cross-examination may cause the industry and regulators to perform better and

more thorough safety evaluations of proposed

nuclear powerplants than would be performed

in the absence of public scrutiny. To reduce

the opportunity for public rehearing would

enhance efforts toward standardization at the

expense of public input.

The alternative in the NRC current standardizations program is the submittal of a final

design by any qualified applicant to the Commission for rulemaking. 3 This allows NRC to

review and approve a plant design without

having received an individual CP application.

Once approved, utilities could reference this

design and start construction after demonstrating the acceptability of the proposed site. A

procedure similar to this is being used for the

manufacturing license although a final NRC

review and possible hearing will be required

prior to the towing of the initial floating

nuclear plant from the manufacturing facility.

Each site for a floating nuclear pIant must be

licensed. Since a public hearing is mandatory

during rulemaking on the design, later hearings

at the CP and OL stages would be limited to

issues unrelated to the approved design (e. g.,

environmental impact and utility competence). Presently, no applicant has requested

rulemaking under the current standardization

program, probably because of the expense involved and the possible public perception that

rulemaking on a design was a means of bypassing the statutory requirements of the Atomic

Energy Act for public hearings.

‘Code of Federal Regulations, title 10, pt 50, app O (40

FR-2977), Jan 17,1975
Ch. 4—The Nuclear Regulatory Commission’s Role ● 39

CURRENT STATUS OF LICENSING

Since TMI, new plant orders have disappeared. Meanwhile, NRC has been under intense pressure to investigate or rule on many

safety issues. These two factors have combined to force standardization programs into

the background. Currently, such programs are

under the Standardization and Special Projects Branch at NRC and are manned by a very

small staff with virtually no budget. However,

the work that must be done includes review of

Combustion Engineering’s application for an

FDA, which is the first one to be requested

under the reference system concept [the 3-unit

Palo Verde application references the Combustion Engineering FDA). Another important

action awaiting the standardization branch is

the complete acceptance review of the application by GE in March 1980 for an FDA for its

nuclear-island designs. Six plants now under

construction reference this design.

Although other matters do require heavy

demands on its staff, NRC should be aware

that current steps must be taken, both to consolidate the gains begun under the standardization program and to plan for a possible

future of renewed interest in nuclear power. In

particular, NRC should be giving more attention to the implementation of some form of

single-stage Iicensing.

NRC’S FUTURE ROLE

One criticism sometimes leveled at NRC is

that it is not disciplined or consistent in its

decisions regarding which safety concerns are

sufficient to warrant design changes or even

reactor retrofits. Industr

y observers, in particular, worry that unless NRC is more disciplined, reactors initially designed as similar

plants may grow apart because of changing

regulations. Adoption of a safety goal would

certainly help NRC arrive at consistent and

more predictable decisions regarding design

changes. As generally used now, the concept

of a safety goal — which might be either quantitative or qualitative— is the definition of an

optimum level of safety as a focus for the licensing process. It would consider both individual and societal risk, and include some

method of measuring the effectiveness of the

safety standards prevailing at any particular

time.

In the licensing for either custom or standard plants, NRC has currently introduced an atmosphere of uncertainty. Many safety issues

await rulemaking by NRC. Until NRC rules on

the issues or unless NRC adopts interim criteria, nuclear plant designers will be uncertain

how to design a plant that can be licensed.

One example of these current safety issues pertains to degraded cores. The objective of the

degraded core rulemaking, to commence

some-time during the second half of 1981, is to

determine whether fundamental changes are

required in reactor design to prevent or mitigate a melted core from penetrating the containment and entering the outside environment. NRC has not provided any interim guidelines for what designs are acceptable until the

rulemaking is completed.

As another example, NRC has recently ruled

that applicants analyze all accidents of a certain class (called “class 9“ accidents). Unfortunately, NRC has not defined these accidents well enough or sufficiently narrowed

that class of accidents for them to be reasonably analyzed.

These two examples are a few of many that

indicate that NRC is not managing its activities

effectively at this time. Uncertainties or ambiguities such as those mentioned will impede all

licensing— standard or otherwise. The adoption of a safety goal might facilitate the many

decisions NRC has to make.
Chapter 5

THE NUCLEAR INDUSTRY’S

EXPERIENCE WITH STANDARDIZATION
Chapter 5

THE NUCLEAR INDUSTRY’S EXPERIENCE

WITH STANDARDIZATION

THE NAVAL REACTOR PROGRAM

The Naval Reactors Program under Adm.

Hyman Rickover has responsibility for 125 operating nuclear-powered ships, with 36 authorized or under construction. The U.S. Navy

has attempted to maintain as much standardization as practicable, with particular emphasis

on the similarity of control rooms, instrumentation, operating procedures, and training programs, All operators attend the same nuclear

power school, and manuals used for training

are of the same type as those on shipboard.

However, the specific propulsion-plant designs

may vary because of the different sizes and

military functions of the vessels they must

power. The Navy’s nuclear-powered ships include attack submarines, ballistic missile submarines, aircraft carriers, and cruisers. At least

11 classes of ships are built or authorized

under the Naval Nuclear Propulsion Program.

Even within a single class, some variation has

resulted as new technologies develop and

become installed on later models of a given

class.

The designs are formulated at one of two,

Government-owned, contractor-run laboratories—the Bettis Atomic Power Laboratory

(operated by Westinghouse) and the Knolls

Atomic Power Laboratory (operated by General Electric (GE)). Designers at these laboratories must obtain the approval for their

designs from Naval Reactors headquarters,

and the designers are held responsible for

study and improvements in the designs even

after the reactor has been built. Presently, submarines are built at two commercial shipyards.

With regard to commercial nuclear powerplant standardization, Rickover made two observations before the President’s Commission

on the Accident at Three Mile Island on July

23, 1979. The first was on the desirability that

the utilities “unite to establish a separate technical organization which could provide a more

coordinated and expert technical input and

control for the commercial nuclear power program than is presently possible for each utility

with its limited staff. ”

Among the things such an organization

could do are:

●

●

☛

●

●

●

develop the standards and specifications

utilities should require for design and construction of their plants;

provide direct, indepth technical assistance to utilities in design, construction,

and operational questions;

establish recommended staffing requirements for operation of nuclear plants—

e.g., at times there may be only a single

operator with no supervision present in

the control room of an operating plant.

Also, that operators may be assigned and

actually carry out unrelated duties while

on watch. These are contrary to Navy

practice;

develop a comprehensive training and retraining program, including lesson plans,

qualification requirements, etc., for utilities to use in training their people. This

must be based on what is needed and not

geared solely to passing licensing examinations. It should cover al I types of personnel, not just operators;

provide trained technical teams to perform periodic audits of nuclear stations

and critically evaluate the plants and the

qualifications and performance of personnel; and

advise utilities on technical safety questions.

In the same testimony before the Three Mile

Island (TMI) Commission, Rickover recommended that “plant designs, equipment, control rooms, training, etc., should be standardized insofar as practicable. ” Rickover noted

that much more standardization seems practi43
44 ● Nuclear Powerplant Standardization

cal on new plants than old ones (where it might

nevertheless be possible to achieve some degree of standardization of control rooms, instrumentation, etc.), and that standardization

should have two distinct benefits. First, he

noted, that better designs should result

because a larger number of engineering manhours could be applied to standardized designs, and, with a larger number of identical

operating systems, operational experience

would “provide a valuable source of information that can be used to improve the design

and procedures and establish a more effective

preventive maintenance program for al l

plants. ” Second, he noted, the use of standard

designs would make it possible to train operating and inspection personnel more effectively.

However, Rickover did not advocate the

most extreme form of standardization. “In advocating more standardization I am not saying

that there should be one single design. I have

standardized in my program as far as practicable. Even then we have a number of designs to

suit the different power ratings and ship types

and to take advantage of new developments

and technology which have become available. ”

With regard to a new technical organization,

the utilities have jointly funded the Institute

for Nuclear Power Operations (I NPO), which is

undertaking to prepare models for operator

training programs, and will establish training

program criteria, accrediting industry training

programs, and performing in-plant evaluations. INPO hopes these programs will be more

specific than those of the Nuclear Regulatory

Commission (NRC). The models will be recommendations, not requirements, for the utilities.

Another collective organization funded by the

utilities is the Nuclear Safety Analysis Center

(N SAC), recently created by the Electric Power

Research Institute to provide more technical

assistance to the utilities. The commercial

nuclear industry has, therefore, strengthened

its organizations along the lines suggested by

Rickover although none has the total authority

that the Navy exercises over its reactors’ program. The benefit of these organizations is difficult to judge because of preoccupation with

the implementation of the requirements resulting from the accident at TM I and the short

length of time (about 1 year) of their existence.

Their success will depend on the quality of the

personnel in the organization and the willingness of the utilities to accept their assistance

responsibly.

With regard to standardized plant designs,

the currently available standard designs

docketed with NRC represent an improvement

in decreasing the number of designs that are

commercially offered. A greater reliance on a

joint utility organization that sets design standards and criteria that are more detailed than

those in NRC regulations is desirable. The implementations of such a concept in the near future may be extremely clifficult due to the current high level of regulatory activity in areas

other than standardization.

STANDARDIZED NUCLEAR UNIT POWERPLANT SYSTEM

The Standardized Nuclear Unit Power Plant

System (S NUPPS) is a consortium of utilities

organized to build identical nuclear plants at

different sites across the country. It is the

closest project to full standardization with

plants under construction. SNUPPS project

management is handled by a contractual arrangement with Nuclear Projects, Inc., and a

hierarchy of utility companies. The five utility

companies have entered into separate but

nearly identical contracts with Bechtel Power

Corp. (lead architect engineers (AE)), Westinghouse (supplier of the nuclear steam supply

system (N SSS)), General EIectric Co. (supplier

of turbine generators), and Nuclear Projects,

Inc.

SNUPPS originally was a consortium of

power utilities that made an application for six

units at four sites. One unit was withdrawn

shortly after application. Another (the Sterling

unit) was canceled because of a lessening of
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 pipe

or 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, and

uncertainties in the Federal regulatory procedures. The design and construction of both

Callaway 1 and Wolf Creek units are over 60-

percent complete, however, both have suffered time delays and substantial cost increases. The time delays have resulted from

financial considerations and Federal regulatory concerns, while cost increases have occurred primarily due to recent unusually high

inflation rates.

The SNUPPS project is based on identical

units with no shared systems. If two units were

to be constructed at the same site they would

be identical but separate units. For each plant,

Westinghouse produced a standard information package in order that Bechtel could design and engineer the balance of plant with

minimum changes to the NSSS. This approach

facilitates the orderly progression of design

drawings and the ordering of equipment.

All plants will have identical portions called

the “power block”, this consists of the reactor

building (containment), fuel building, turbine

building, hot machine shop, auxiliary building,

diesel generator building, control room building, and radwaste building. Structures and

components outside the power block differ for

the various plants and are not under control of

SNUPPS.

In the licensing process, the project is

managed by a single project manager and review team within NRC. In addition, the Ad-
46 ● Nuclear Powerplant Standardization

visory Committee on Reactor Safeguards assigned a subcommittee to review the standard

portions of SNUPPS and when the Atomic

Safety and Licensing Board hearing for two of

the units were held, previously resolved issues

were not re-reviewed. This sharing of licensing

resources allows more licensing personnel to

provide a greater indepth review than would

have been possible with a customized application for several plants. In addition, there was a

reduction in the questions asked by NRC from

an average of 700 for a customized plant to an

average of 150 per single SNUPPS unit.

During the procurement for the units, only

proven materials, equipment, and systems are

to be used; American National Standards institute and other appropriate standards are to

be strictly followed. Power block purchases

are centralized — i.e., with few exceptions the

same supplier and the identical item for a particular function are used for all plants. This

allows interchangeability of parts between

plants. These are common industrial practices.

During construction, a considerable amount

of standardization is maintained. Detailed

models and photographs of the models of the

standard plants are used in the construction effort. This method has eliminated much interference and many delays while providing a

considerable surety of proper construction

techniques. Construction equipment common

to SNUPPS plants is shared by the construction

crews.

Standard preoperational procedures, startup, and functional operating procedures are

being prepared for the SNUPPS plants. Also,

simulators will be available for the SNUPPS

pIants and operating experiences will be

shared among SNUPPS utilities’ personnel.

The participants in the SNUPPS program

claim the SNUPPS plants will be built for

about 1(1-percent less than if they were customized plants. Further, they feel the plants will be

safer because of the standardization effort.

However, there are no hard data to substantiate this claim, only an intuitive feeling that the

more man-hours spent on a particular system

design the safer it will be. ’

‘Nlcolax A Petrick, “Progress Report on the SNUPPS Nuclear

Stat Ions,” Nuclear Projects, Inc , Nuclear Englneerlng lnternatlonal, November 1977

THE FRENCH NUCLEAR PROGRAM

The French have developed a consensus of

government energy policy makers that is supported, almost totally, by all four major political parties. The French nuclear program has

some of the same problems as other nations —

e.g., opposition by organized citizen groups,

some difficult public relation situations, and

some technological shortfalls; however, they

have maintained a firm commitment to their

policy of “tout nucleaire” (i. e., decommissioning

oil-fired electricity generation plants and

building coal-fired, hydrostorages and mostly

nuclear powerplants). The French policy was

formulated by their perception for the need to

reduce dependence on foreign supply of oil

(which in 1973 supplied France 67 percent of

its energy needs). Further, the French have only

very limited supplies of oil, natural gas, and

coal within their boundaries. The French condition is quite different from the United

States— i.e., there is no clear political consensus on the need for nuclear power in the

United States, partly because there is an indigenous supply of oil, natural gas, and coal

within U.S. borders.

The choice between the two commercial

types of light water reactors —e. g. (boiling

water reactor (BWR) and pressurized water

reactor (PWR)) — using enriched uranium was

made on the basis of price. The French entered

into a competitive program between European

holders of licenses for the manufacture of

American designed plants. Framatome held a

Westinghouse license and Alsthom, a GE

license. 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 that

it does not own heavy forging facilities for reactor vessel construction. For this reason, Alsthom either had to call upon their competitor,

Framatome, or contract abroad. The BWR line

was therefore dropped not because of the

PWRS technical superiority but to ensure a sufficient workload for the French industrial

group in charge of construction. The French industry was, therefore, restructured into one

constructor of nuclear steam supply components (Framatome) and one constructor of turbine generators (Alsthom). In addition, the national electric utility is the only French AE,

thereby making the standardization of nuclear

powerplants easier in France than it would be

in the United States.

The French recognize four major safetyrelated advantages for standardization:

1.

2.

3.

4.

a more thorough investigation of safetyrelated matters is possible when multiple

units are involved;

experience in design, manufacture and

construction, and operation can be transferred from unit to unit;

more designer time becomes available to

spend time working with a new generation

standardization series; and

regulators can spend more time inquiring

about site-specific considerations, the

need for new units, and the ability of the

utility 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 of

the 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 major difficulties with standardization of nuclear

powerplants:

1. problems involved with one unit of a series propagates to other units in the series

and may require expensive and time-consuming redesign and back-fitting;

2. site considerations may require substantial design differences between units of a

standardized series; and

3. the optimal balance between design stability and technological upgrading is difficult to determine (i. e., a definition is

needed of safety enhancement or cost reduction required before a new technological achievement can be incorporated .2

Overall, the French are satisfied with their

choice and consider that the advantage of

standardization (especially those related to

safety and economics) far outweighs those dif -

ficulties.

‘MIChel Pecgner, “How,One Organization Runs the Whole industry, ” Commissarat a L Energle Atomique (CEA), Nuclear EngIneerlng International, December 1976
Ch. 5—The Nuclear Industry’s Experience with Standardization ● 49

THE WEST GERMAN OPERATOR TRAINING PROGRAM

A possible model for standardization of

training and certification of personnel in

commercial nuclear powerplants is the West

German program. The West Germans train and

certify their operators for both conventional

and nuclear powerplants in a powerplant

school called the Kraftwerkschu le. This is a

joint organization of owners of large powerplants with 116 members from six different

countries. The primary purpose of the school is

to provide professional and advanced training

in six different technical areas for powerpl ant

personnel in maintenance and operation. The

school was founded by a parent organization

called the Technical Association of Power

Plant Operators, formed as a result of a severe

boiler explosion in 1920.

The formal training for a plantworker takes

3 years and is integrated into the operation of

the powerpl ant. Training consists of theory

and practice with a final exam for certification

in the operation of powerplant systems. Figure

6 shows the progression for a nuclear plantworker from initial certification by the Kraftwerkshule to shift supervisor.

The professional competence of the operators and shift supervisor is regulated by official

government guidelines. The minimum personnel complement for a nuclear powerplant control room is a shift supervisor, a deputy shift

supervisor, and a powerplant reactor operator.

The shift supervisor must be an engineer and

his deputy must be at least qualified as a Kraftwerkmiester (see figure 6). All three require additional special nuclear training including simulator training, and practical nuclear powerplant experience.

As in the United States, the plant’s superintendent is responsible for the selection and

training of the powerplant team. The superintendent assesses the workers’ capabilities and

determines who will eventually be qualified as

a plant attendant, plant operator, or shift supervisor. Unlike his counterpart in the United

States, the West German superintendent picks

his candidates from a pool of workers who

have completed a standardized training program established by the owners of the powerplants under the guidelines of the government.

In this country, the closest organization to the

West German program that has uniform training and certification for its reactor operators is

the U.S. Navy. Many utilities rely heavily on

the Navy for qualified plant operators. This dependence can create manpower shortages in

an area vital to the national defense and

allows the utility to abrogate some of its responsibility for a complete and total training

program for new operators with no nuclear experience.

‘0 Schwarz and G Schiegel, “Combining Theory and Practice

in West German y,” NucIear Engineering International, March

1980
-

50 ● Nuclear Powerplant Standardization

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

Krattwerksmeister training

theoretical training

at KWS 1580 h

I

I

.

Reactor Operator .,

I

Krattwerker training

practical training in

powerplant 36 months

theoretical training in

powerplant or at KWS

590 h

M Mechanical engineering

Electrical engineering

Measurement and

control instrumentation

Nuclear engineering

Kraftwerksschule

SOURCE Nuclear Engineering International
Chapter 6

POLICY IMPACTS OF FOUR

APPROACHES TO STANDARDIZATION
Chapter 6

POLICY IMPACTS OF FOUR APPROACHES TO

STANDARDIZATION

The situation presented in the previous

chapters is one of an industry which has been

slowly evolving toward a greater degree of similarity in its products. Stringent standardization is very difficult in the commercial nuclear

industry where the tasks of design, supply and

construction, operation, and regulation are undertaken by multiple and often independent

organizations. Nevertheless, the designs formulated by the nuclear steam supply system

(NSSS) vendors and architect-engineer (AE)

firms are slowly converging toward a single

design for each company. Several utilities and

utility consortia have attempted to construct

multiple reactors based on a single design. The

Nuclear Regulatory Commission (NRC) has for

some years defined special licensing for four

categories of “standard” plants defined in

chapter 4. These steps have been taken voluntarily over a 10-year period, because the industry perceives they will produce lower costs,

shorter licensing times, and more reliable

plants. Increasingly, both Government and

industry personnel have concluded that a

more rapid move to standardization may increase the safety of nuclear plants. They also

recognize that the industry will not move more

rapidly toward standardization unless external

forces push it in that direction. Four representative approaches to standardization are used

here to provide a framework for this analysis.

FOUR APPROACHES

Acceleration of Present Policies. –An incentive program to accelerate the present trends

in the industry could reduce the number of designs substantially. In the first place, such a

program could reduce the number of designs

to one for each designer— i.e., 4 NSSS designs

and 4 to 12 balance-of-plant (BOP) designs, depending on the number of AEs that remain active in the nuclear field. Only a few AEs have

developed BOP designs for the boiling water

reactor (BWR) and General Electric Co. ’s (GE)

completed design for a nuclear island approach based on the BWR make it likely that

future BWRS will be of one design. For the

pressure water reactors (PWRS) produced by

the other three vendors, each AE would have

basically the same BOP design tailored to

meet the various interface criteria. Thus, the

possible number of different reactor plants

would be in the range of 5 to 13. The lower

number could result if the utilities agreed

on design features and specific criteria for a

standard BOP. Any AE could design a BOP

conforming to these agreed-on criteria and the

existing regulatory requirements. NRC could

then offer one-step licensing for any utility

referencing this “standard” in a license application. The time to implement this level of

standardization would equal the time to

formulate the criteria and implement one-step

licensing– about 1 to 3 years.

Procedural and Organizational Standardization.– One advantage of standardization

would be that it would allow personnel training, operating procedures, terminology, etc., to

be specified in greater detail for a larger body

of plants. Adoption of more universal practices would allow operators of different plants

to learn more from the experiences of one

another and would facilitate audits. Even without identical hardware, for the existing generation of powerplants, the “software” practices

could be made more alike. NRC has some

standards for such practices and private

groups, such as the Institute for Nuclear Power

Operations (IN PO) and the Nuclear Safety

Analysis Center (NSAC), are currently evaluating operating practices with a view toward

upgrading them. If the Government wished to

53
54 . Nuclear Powerplant Standardization

do more, a starting point would be to examine

NRC’s current standards to see if they could be

more precisely specified and more universally

applied. An examination of the West German

standardized training and certifications program for nuclear powerplant personnel discussed earlier, might be appropriate.

Standardization of the NSSS Design Plus a Safety Block. –One of the major reasons or standardization is to allow more attention to a smaller number of designs and especially to safetyrelated systems (e. g., auxiliary feedwater and

containment solation systems). One possible

approach to standardization is to define those

portions of the BOP that are necessary to bring

the reactor safely to a cold shutdown condition and to allow only four variants (one for

each NSSS) of this so-called “safety block. ”

Under this approach, the safety block would

include 25 to 50 percent more equipment and

hardware than the present NSSS. This version

of standardization represents a significant deviation from the current mode of doing business and would require either a redefinition of

responsibilities as now specified by NRC and

perhaps some legislative action. To achieve

this level of standardization, either the Government or industry would have to define what

components belong in the safety block, subject the particular designs to some criteria of

safety and reliability, and transfer responsibility for them from the AE firms to the NSSS vendors or to a design team composed of both.

SAFETY

Almost all of the potential safety benefits of

standardization are proffered on the basis of

intuition rather than experience. Few relevant

examples of standardization exist, and none

demonstrate unambiguously that the safety

achieved results from the standardization

rather than from other factors — e.g., the safety

record of the naval nuclear reactors program

probably results as much or more from the U.S.

Navy’s central control and other factors as

from any similarity among its various reactor

plants. Some of the arguments for the safety

benefits may break down in the extreme case

Critics of the approach suggest this transfer of

responsibility for safety systems, normally

under the control of the AE firms, may place a

burden on some of the NSSS vendors for which

they are neither qualified nor prepared and

thereby significantly alter the present structure

of the nuclear industry. Some of the essential

safety systems (e. g., the containment) require

design and construction skills for which the

AEs are uniquely qualified. The safety-block

approach is similar to that proposed by GE,

with its “nuclear-island” concept; this would

take about 3 to 5 years to implement.

One Single-Standard Plant. –The ultimate in

standardization would be to select only one

plant design according to which all future reactors would be built. Such standardization

would have to be accomplished by legal statute and would completely alter the present

structure of the commercial nuclear industry.

To implement this concept of standardization

one must decide who would have overall responsibility for the design, what the design criteria should be, and what would be the criteria

and time scale for incorporating modifications

into the standard design. It would require from

6 to 10 years to design and an equivalent time

to construct this single, national reactor.

Even for a single-design approach, site-specific factors such as seismology, meteorology,

and hydrology would require modifications in

some of the reactor plants.

BENEFITS

of standardization — e.g., the one single-standard pIant concept is seen by some as an opportunity for a fresh objective look at commercial

reactor design while it is viewed by others as a

dangerous commitment to a possibly flawed,

single design. The following discussion is an examination of the arguments in favor of standardization and the extent to which these arguments apply to the four previously defined approaches to standardization.

Enhanced Design Review. –Most people in

the nuclear industry or within NRC concur that
Ch. 6—Policy Impacts of Four Approaches to Standardization ● 55

the attention given to a particular design

should increase as the number of designs decreases. The incentive program towards standardization should allow more concentration of

attention within the designer firms. Moving

towards a safety-block concept or single-standard design would primarily benefit regulators

such as NRC by greatly reducing the number of

different reactors it would have to understand

and regulate. Those advocating a single national design feel that its major advantage

would be the design attention devoted to it.

Designers could start afresh, yet benefit from

the experience gained during the many years

of operation with light water reactors (LWRS).

Similarly, design attention to a safety-block

design may lead to a safer product. One

should keep in mind that the quality of attention paid to a design is as important as the

quantity of designers or safety analysts studying it. It is also possible that the reduction in

the number of reactor designs might merely result in a proportional reduction in the number

of designers.

A design-review mechanism known as probabilistic risk assessment (PRA) has received

considerable attention since the Three Mile

Island (TMI) accident. The use of this technique in assessing auxiliary feedwater system

reliability was discussed earlier in chapter 3.

On a larger scale, PRA involves the steps of

identifying hazards, hazardous activities and

accident sequences, and quantifying the probability of accident sequences and the magnitude of their consequences. The determination

of risk for a nuclear plant involves all parts of

the plant and its operation. The NSSS, the BOP

(e. g., the control room, containment, power

conversion system, and electrical systems),

and utiIity-operator aspects (i. e., the operating

and maintenance procedures and the electrical grid), all are important in determining

overall plant accident risks.

What sequences dominate risk can be

strongly dependent on the details of plant

design and operation. Subsequent to the reactor safety study (RSS) (WASH-I 400) which considered two reactors in detail, NRC sponsored

an RSS methodology applications program

which looked at four additional reactors.

While the results of this work have not been

published, preliminary results indicate that

considerable differences in accident sequences exist compared to the one considered

in WAS H-1 400. These differences are due to:

●

●

●

Not

safety systems unique to the plant

studied;

safety systems performing functions different than in WASH-1400; and

muItiple success options for a given function requiring different levels of system

success.

only were unique plant sequences found,

preliminary results indicate that the dominant

sequences vary from plant to plant.

Therefore, the major impact of standardization on probabilistic risk assessment would be

to avoid industry manpower limitations in the

evaluation of all plants to the degree needed

to maximize plant reliability and safety. The

fewer number of plants needing evaluation the

greater the quality and detail of the risk assessment for a given amount of resources. In addition, a greater understanding of the insights

particular to risk assessment would be obtained. In retrospect, the RSS (WASH-1400)

yielded considerable insight to the TM1-type

accident (e. g., a small break, loss of coolant

accident), to the recent Browns Ferry partial

scram and to the contributions of human errors to reactor accidents in general. If it were

applicable to all reactors, these design problems might have been anticipated and therefore prevented by early corrective action.

Increased Awareness and Applicability of Operational Experience. –This possible safety benefit should be realized to various degrees for

any of the four approaches to standardization.

Naturally, the fewer the differences among reactors, the more the overlap of experience.

The accident at TM I provides a positive example, by which reactors of similar design have

learned to watch for a similar sequence of

events. On the other hand, many incidents—

‘Nuclear Regulatory Commission, “Reactor Safety Study An

Assessment of Accident Risks in Commercial Nuclear Powerplants, ” NUREG-75/014, WAS H-1 400, October 1975
56 ● Nuclear Powerplant Standardization

such as the Brown’s Ferry partial scram — are

still caused by specific piping or instrument errors which may be peculiar to that plant alone.

One central mechanism by which various

nuclear plant operators learn from the experience of others is by the Licensee Event

Reports submitted to NRC. 2 The greater the

similarity among plants—even if it is only

more similar terminology or procedures—the

easier it should be to understand these events

and to decide to which other plants they potentially relate.

There is no inherent reason why operators of

custom plants should learn as much from operating standard plants as other plants, but more

interpretation is required to decide where each

experience is relevant. It has been reported

that an incident at the Davis Besse plant, was a

precursor to the TM I accident, but no warning

was issued. Standardization would not eliminate such omissions automatically but could

ease the burden of deciding which reportable

events were especially important to which

plants.

The feedback provided by the naval nuclear

reactors program is a key element in the safety

of their program, and it is achieved despite

considerable variation among naval reactors.

Currently, NRC and the industry are striving to

improve the feedback of plant experience.

NRC has established the Office for Analysis

and Evaluation of Operational Data. The Office reviews all reportable events from reactors

and users of byproduct material. NSAC has

created a communication and evaluation network used by operators of commercial reactors to inform one another of significant operational occurrences.

Regardless of the organization, one of the

difficulties experienced with reviewing operating data is that of interpreting the relevance

of a specific component failure at one plant to

the safety of another plant using a similar but

not identical component. The interpretation

may be easier if the component used is iden2 nuclear ReguIatory Commission, “Reporting of Operating

Information – Appendix, A Technical Specification,” Regulatory

Guide 116 (revision 4), August 1975

tical in all plants, but the plants themselves

differ significantly. Experience to date has

shown that emphasis on feedback of operating

data by the reactor vendors (GE, in particular)

has markedly improved plant availability. One

characteristic of responsible plant management is its willingness and ability to identify

and to correct the generic or recurrent problems underlying all unusual occurrences in its

nuclear powerplants. In a more standardized

nuclear industry there would be no question

about the importance of taking the broad view

of all identified problems. A more standardized industry would potentially permit a

relatively small group of experienced engineers to review the data generated by operating experience, looking for the generic implications of apparently “random” failures. At

present, the heterogeneous nuclear industry

provides generic assessment of operating experience by means of various user groups. Examples include the BWR Mark I containment

owners’ group; and the GE, Westinghouse,

Babcock & Wilcox, and Combustion Engineering owner’s groups; The formation of these

groups results in part from an interest in the

free flow of information on solutions to their

common problems.

While increased standardization would further help in the identification and resolution of

safety issues, it would also increase the risk of

systematic oversight of potential problems. As

a matter of policy, electric utilities plan diversity into their generating mix, both fossil and

nuclear, and among the several reactor designs. This course has been amply vindicated

by the many generic shutdowns that have occurred without loss of a major part of the

nuclear generating capacity. A nonnuclear

analogy would be the obvious consequence of

having a standardized U.S. jumbo jet, such as

the DC-10, grounded when a generic enginemounting defect is discovered. The degree of

nuclear standardization needed to produce

optimum benefits is a subject for further

evaluation.

The greatest increase in health and safety

comes from the review and evaluation of operating and construction experience on one sin-
Ch. 6—Policy Impacts of Four Approaches to Standardization ● 57

gle-plant design. However, the institutional

barriers and the possibility of systematic oversight of safety problems may outweigh any

safety benefits accrued through the feedback

of data on one “accepted” design. With regard

to procedural and organizational standardization, the benefit acheived through uniform

reporting and review practices can be easily

obtained with little if any disruption in the institutions regulating and operating commercial reactors.

Improved Training for Plant Personnel

The impact of the approaches to standardization of improving plant training is easily

analyzed by considering three of the concepts

under one heading “hardware standardization. ” The order of increasing hardware standardization would be:

1. acceleration of present trends;

2. NSSS plus safety block; and

3. single-plant design.

The other approach, procedural standardization, is considered by itself as the standardization of the management processes as distinct

from hardware. In addition, other institutional

factors not normally considered part of an

idealized, formal training program must be

taken into account.

The basis for the procedures for design, construction, and operation of a nuclear powerplant is the Code of Federal Regulations, industry standards, and NRC’s rules and regulations. Each applicant for a license establishes a

set of administrative procedures that implement the letter and intent of these rules and

regulations. For an operating reactor, one part

of there administrative procedures deals with

the selection, training, and qualification of the

plant’s employees —e. g., these procedures describe the general employee training requirements as well as those for technicians and

operators. Each member of the plant staff is

subjected to some training with different

degrees of intensity and depth according to

the position filled. Currently, there is wide

diversity in the training programs resulting

from the way the utilities interpret the basic requirements when establishing their administrative procedures —e. g., the requirements for a

licensed operator to requalify on a yearly basis

include the performance of 10 major changes

in the plant’s status from the operator’s console. Some utilities meet the requirement by

simply counting the startups or shutdowns the

operator has performed over the past year.

Others send the operator to a plant simulator

for as long as 2 weeks for intensive retraining.

New requirements resulting from the accident

at TM I have specified in detail the types of

manipulations necessary for this requalification.

3 In addition, these manipulations will require the use of a plant simulator.

Greater standardization in operator training

programs than what currently exists would

ease the administrative burden on implementation and auditing of this new requirement.

Also, the effectiveness of the requirement over

the next few years would be easier to judge if

the change were made from training programs

which had more in common. Greater hardware

standardization would make the detailed procedural level of these training programs more

alike but would be unlikely to increase their effectiveness or ease the administrative burden.

Standardization of hardware would make

selected improvements possible in training

plant personnel. One area in which this could

occur is the use of plant simulators. A simulator consists of a mockup of the control room

with indicators, gages, and other instruments

and devices driven by a computer. The operator’s manipulations of the switches in the

mockup are monitored by the computer,

which simulates the reactions of the plant on

the mockup instrumentation. If greater hardware standardization were used in the nuclear

industry, more plant operators could use the

same simulator and fewer plant-specific simulators would be needed. Standardization of the

hardware would also increase the analytical

capability of simulators to deal with off-normal transients when a transient occurs at one

‘Nuclear Regulatory Commission, NRC Action Plan Developed as a Result of the T MI-2 Accident, N U R E G-0660, ” May

1980
58 ● Nuclear Powerplant Standardization

plant and operators at other plants need to be

trained for possible reoccurrence of the same

type of event. Another benefit is that the incorporation of an actual event into the simulator’s computer would be easier –e.g., all actual transients could be incorporated into the

simulator without the necessity of incorporating specific differences in plant operating

characteristics resulting from different designs. The difficulty encountered by the various vendors in simulating the TM I accident on

their own simulators (as an aid to operator

training) was an example of of this.

However, all of these advantages must be

viewed in the context of the existing mix o f

generation common to most utilities and regional differences in the utilities’ service areas.

The additions of several nuclear powerplants

of standard design may not simplify the

utilities training program if the current program is determined by the diversity in existing

operating units. Among most electrical utilities, any “standard” plant would be unique as

a source of power generation because it would

be different from existing plants. It would complicate rather than simplify the existing training program. Unless a utility makes a substantial use of a single design in its operating

system, the value of hardware standardization

in improving the utility’s training program will

be minimal.

Procedural standardization in personnel selection, training, and requalification may be

difficult if there are significant differences in

State labor laws, union contracts, or State regulatory requirements. However, considering

the current generation mix of each utility, this

standardization approach appears to be the

easiest to implement with substantial benefits

in personnel training.

RELEVANCE TO A NATIONAL SAFETY GOAL

The question of the need for quantitative

safety goals to ensure that adequate levels of

nuclear powerplant safety are achieved is a

longstanding one. The Atomic Energy Act of

1954 and the Energy Reorganization Act of

1974 established the legislative basis for NRC

regulation to ensure the safe use of commercial nuclear power. In response to the legislative mandate, NRC regulations require, as a

part of issuing a nuclear powerplant construction permit, that a finding be made that “the

proposed facility can be constructed and

operated at the proposed location without undue risk to the health and safety of the public’” and as a part of issuing an operating

license that a finding be made “that there is

reasonable assurance that the activities authorized by the operating license can be conducted without endangering the health and

safety of the public. ”5

The principles used by NRC are based on a

“defense-in-depth” approach to the plant

4CFR 1050, sec 5035

5CFR 1050, sec 5057

design. Reactor safety as practiced in accord

with these principles is defined in NRC’s regulations, safety guides, branch technical positions, and related industry standards. These

provide an extensively documented licensing

process that has helped the nuclear industry to

achieve an impressive record with regard to

public health and safety. In this process, many

safety requirements and calculational methods have been identified. Following NRC rules

establishes that plants adequately meet

specific safety requirements and satisfy the requirements of the legislative mandate. This deterministic process is based on implied but

unstated probabilities. For instance, a qualitative probabilistic judgment was made many

years ago that the large rupture of a reactor

pressure vessel in LWRS was unlikely enough

that it did not have to be considered in the

design. In the intervening years a quantitative

basis has been provided to support that qualitative judgment, The NRC licensing process is

now considering other factors that arise from

accidents of greater severity than the designbasis accidents (DBA). Consideration of such
Ch. 6—Policy Impacts of Four Approaches to Standardization ● 59

accidents will require a different type of analysis than the traditionally conservative approaches taken in the assessment of DBAs. The

use of PRA techniques is rapidly coming into

use for this purpose. Quantitative criteria for

acceptable levels of risk, or safety goals, are

needed if all the benefits of PRA are to be

realized. PRA is an acceptable quantitative

method of showing compliance with a well–

defined safety goal.

U.S. activities relating to the establishment

of a national safety goal are going on within

the NRC, the Advisory Committee on Reactor

Safeguards (ACRS), the nuclear industry in

general, and the national technical and scientific community. There are also international

activities in this area. Possible variations in

goal forms that have been considered include:

single v. multiple goals, quantitative v. qualitative 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 lnCfu$trlal Forum International Publ{c Affairs Workshop, Stockholm, Sweden, ] une 1980

two broad phases, the initial phase in which a

wide range of goal elements and alternative

strategies are identified, and the second phase

in which the effort is directed toward winnowing down the elements and strategies for more

indepth analysis and decisionmak ing.

in demonstrating compliance with any safety goal, a high level of confidence in the

related risk assessments will be necessary. A

high level of confidence will also be necessary

to achieve public acceptance. PRA techniques

are relatively new and there are too few ski I led

practitioners for it to be applied routinely for

reactor safety assessment. If design standardization were to result in a large reduction in the

number of designs to be reviewed, PRA could

be applied more comprehensively to show

compliance with a safety goal. By the same

token, as the development of PRA techniques

continues, confidence in their application will

increase and the number of skilled practitioners will become very much larger. It may

then be possible to address a wider range of

designs and this aspect of standardization

would be less important.

THE IMPACT OF STANDARDIZATION ON RESOLUTION OF

GENERIC ISSUES

A December 1977 amendment to the Energy

Reorganization Act of 1977 required NRC to

submit to Congress a list of unresolved safety

issues and plans for their resolution. Progress

on resolution is to be included in NRC’s annual

report to Congress. Prior to that, NRC had

developed task-action plans for a multitude of

outstanding topics, many of which were not

considered unresolved safety issues. 1 n

January 1979, NRC submitted a report to Congress identifying 17 unresolved safety issues

and their related task-action plans. 7 A more recent plan updates the status of these issues

and plans.8 The 17 issues are listed in table 6.

‘Nuclear Regulatory Commlsslon, “ Identification of Unresolved Safety Issues Relating to Nuclear Powerplants, ”

NUREG-O51O, January 1979

“Nuclear Regulatory Commission, “Task Action Plan for UnreSOlVed Safety lssues Related to Nuclear powerplants, ”

NUREG-0649, February 1980

As a result of the many investigations of the

TM I accident, NRC published an action plan in

May 1980. 9 This report contains actions to be

carried out by each nuclear plant owner and

the NRC. One might consider these as generic

safety issues; however, they are resolved issues

in that specific action is called for. Also, these

actions are applied to all operating plants, as

well as those under construction. Thus, standardization would not have changed these action plans.

As an example of the effect of standardization on a safety issue, consider item 1 of table

6, “water hammer. ” The phenomenon is similar to the banging of steam-heated radiators

commonly found in old homes or office build-

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

1980
.

60 ● Nuclear Powerplant Standardization

Table 6.—Unresolved Safety Issues

1. Water hammer

2. Asymmetric blowdown loads on the reactor coolant

system

3. Pressurized water reactor steam generator tube

integrity

4. BWR Mark I and Mark II pressure suppression

containment

5. Anticipated transients without scram

6. BWR nozzle cracking

7. Reactor vessel materials toughness

8. Fracture toughness of steam generator and reactor

coolant pump supports

9. System Interactions in nuclear powerplants

10. Environmental qualification of safety-related electrical

equipment

11. Reactor vessel pressure transient protection

12. Residual heat removal requirements

13. Control of heavy loads near spent fuel

14. Seismic design criteria

15. Pipe cracks in boiling water reactors

16. Containment emergency sump reliability

17. Station blackout

SOURCE: Nuclear Regulatory Commission.

ings. Occurrences have been attributed to

rapid condensation of steam pockets, steam--

driven slugs of water, pump startup with partially empty lines, and rapid-valve motion.

Much of the problem might therefore be resolved by piping arrangement to assure filled

lines and prevent steam pockets. This would,

of course, be easier to resolve in standardized

layouts as opposed to those of differing plant

designs. Although there has been no release of

radioactivity outside the plant’s boundary

because of a water-hammer incident, the frequency of such events and the potential safety

significance of the systems involved caused

NRC to consider the water-hammer problem

significant. Were most plants of standardized

design, modifications to prevent recurrence of

many safety-related problems could be carried

out more rapidly as fewer designs need be examined.

Resolution of another issue, related to containment emergency sump reliability, would

also be quicker if designs were standardized.

Although NRC has issued guidance for containment sump design and testing, there are

still concerns about blockage of sump filters

and loss of ability to draw water from the

sump. With fewer designs to investigate, the

emergency sump reliability issue could be settled much quicker.

The previous discussion indicates that standardization would have facilitated resolution of

some of the unresolved safety issues and therefore improved nuclear powerplant safety. On

the other hand, there are issues that would be

unaffected by standardization. For instance,

the disclosure by Virginia Electric Power Co.

that asymmetric loads in the reactor vessel

supports and vessel internals caused by a PWR

pipe break could cause a safety problem, was

the result of studies with computer codes using

more detailed analytical models. In other

words, advances in the state of the art uncovered a problem. In that case the discovery

would have occurred at about the same time in

the advancement of the technology, whether

or not standardization had been implemented.

Finally, several situations have occurred

where similarities in plant standardization

resulted in many nuclear plants experiencing

the same problem — a lesser degree of similarity (i. e., less standardization) could have limited

the number of plants involved. One example

of this was the realization that hydrodynamic

loads on the suppression pool associated with

loss-of-coolant accidents and safety-relief

valve discharge were not considered in the

design of Mark I and Mark II BWR containment. These loads affected 24 Mark I and 11

Mark II plants. Another example is the BWR

nozzle-cracking problem associated with feedwater systems of many BWRS of similar design—18 of 21 units inspected had cracks in

feedwater nozzles.

For the most part, these generic issues arose

when operating experience or advances in the

state of the art uncovered a problem, a discovery which would have occurred at about

the same time in the advancement of the technology, with or without standardization. Resolution of some of the issues would be expedited

if affected nuclear plants were more standardized, while resolution of other issues would

not be affected had standardization been

more prevalent.
Ch. 6—Policy Impacts of Four Approaches to Standardization ● 61

STANDARDIZATION AND ANTITRUST

As noted in chapter 3, the AEs normally

enter into a contract with the utility to provide

engineering services for the proposed nuclear

pIant including procurement of material for

the BOP. However, the utility selects the NSSS

from the four available vendors based on competitive bidding. The reactor, much like the

turbine generator, is considered for the purpose of procurement as a large single piece of

equipment. The utility normally does not involve itself with the selection of the vendor’s

supplier other than to assure they are

qualified. In many cases, the vendor may have

already completed procurement through existing contracts with its suppliers. On the other

hand, the BOP equipment and materials are

procured by competitive bidding for each

plant to satisfy the State agencies regulating

the utilities.

In order to perform safety reviews of proposed nuclear plants, the NRC staff prefers to

have as much detailed design as possible. The

level of detail provided by the vendors is sufficient for this purpose, even before actual construction of the plant begins. However, the AE

cannot supply as detailed a design as can the

NSSS vendor because the procurement of material and detailed design work has generally

not been completed at the time the CP is

issued.

The exclusion of any qualified supplier of

plant equipment due to licensing requirements

for a standard design is a breach of antitrust

law. Increasing the level of detail in design for

the BOP to the same level found in the NSSS

would exclude qualified suppliers from the

market place, due to the differences in business methods.

By taking into account the antitrust due

process in the setting of standards for plant

systems and equipment, the antitrust problem

can be eliminated. Due process in standardsmaking according to the Department of J ustice includes:’”

10john l-l Sherrefleldt Department of JustIce, “Standards for

Standards-Makers (Washington, D C Department of Justice,

American National Standards Institute, March 1978)

●

●

●

●

adequate notice of the proposed adoption

of a standard;

standards development meetings should

be open to the public;

the standards-setting body should have an

affirmative obligation to seek consumer

and small business opinion; and

membership on standards development

committees should represent a balanced

cross-section of all affected parties.

The development of standards which specify

sufficient detail to perform a safety review by

knowledgeable engineers under the above

guidelines should be sufficient to satisfy the

concern over anticompetitive practices and

protect the health and safety of the public. A

subcommittee of the Atomic Industrial Forum

is currently working on a proposed revision to

the current NRC guidance on information required for a safety analysis report for singlestage licensing. In addition, at least two AEs

and one vendor are considering similar proposals.

Of the four standardization approaches considered, the continuation of present policies

with refinement already being considered by

the industry is the least likely to create problems with antitrust. The safety-block concept

would not create any more difficulties than

the acceleration of present policies, although

it would place more of the total plant under

the design control of the vendors to the exclusion of the AE. However, the AE’s role as an

engineering services contractor would be affected since design work encompasses only

about 10 percent of the total cost of the facility. The “national single design” could force

one or more NSSS vendors from the marketplace. The specifications for the design could

be written to allow the vendors to remain competitive suppliers under contract to the utility for equipment and systems. Each vendor

would have to evaluate its interest in the supply business, based in part on the similarity of

the national design components to its own.

However, the single-design standardization approach 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 national

pliers and the possible exclusion of large por- single-plant design.

UTILITIES AND STANDARDIZATION

A utility which operates and maintains a

nuclear powerplant is uniquely responsible to

the Federal and State Governments for the protection of public health and safety. In addition, the utility is responsible to the stockholders for the efficient operation of the plant

and the protection of plant investment in

equipment and fuel supply (i. e., the reactor’s

core). These are not mutually exclusive goals

and measures which protect the core, increase

plant availability, and protect the public.

Because of this unique relationship between

the utility, its stockholders, and government,

nuclear utilities should actively participate in

the formulation of any standard design or approach to standardization.

Over the past 25 years, some utilities that

have purchased nuclear powerplants have had

minimum influence on their design due in part

to the lack of expertise in nuclear design engineering. Therefore, these utilities placed

heavy reliance on the judgment of the AEs and

vendors to protect their financial and regulatory interests. Other utilities, such as Duke

Power and Tennesee Valley Authority have

acted as their own designers and have maintained a strong influence in the design and

construction of their plants. It is also this latter

group of utilities which have maintained a

strong commitment to standardization as evidenced by their recent construction record for

duplicate plants. However, having only a few

utilities committed to standardization may not

be enough to reap its benefits if a resurgence

in new plant orders occurs.

A utility organization could, over the next 2

or 3 years, develop standards and criteria for

new plants which incorporate the cumulative

operating experience of the industry. These

criteria should concentrate on safe, conservative designs and reemphasize the past practice of simply meeting licensing requirements.

This effort would result in a set of criteria for

everyone (e. g., designers, operators, and regulators) and lend consistency to their actions,

Common, understandable objectives could be

established which concentrate on the real

issues of safety and reliability. The effort

should include input from AEs, vendors, and

perhaps NRC. Inclusion of NRC should be

limited to their role as regulators not designers

or operators.

Once the criteria are set, standard designs

could be developed. Future construction dockets could then be limited to these designs and

thereby allow the marketplace to limit the

number. Single-stage Iicensing would be a considerable inducement to the whole process.

FEASIBILITY

Of the approaches to standardization considered, the acceleration of present trends and

procedural standardization are the most feasible to achieve. These approaches work within

the existing structures and motivations of the

commercial nuclear industry. Organizations

such as NSAC and INPO have already been

established as a result of the TM I accident and

are in excellent positions to develop and promote these forms of standardization. In addition, these institutions were established by the

utilities and the utilities are solely responsible

for their success or failure. Such utility organizations could fill the role described previously

for the development of design standards and

criteria. The burden for standardization should

rest with the utilities as they are ultimately

responsible for commercial nuclear power and
Ch. 6—Policy Impacts of Four Approaches to Standardization ● 63

also have the most to lose in the event of an

accident.

As discussed earlier, trends in the industry

over the past 25 years have led to some standardization. This trend can be greatly accelerated by implementing single-step licensing

(or NRC’s standard-design approval) and regulating the industry in a consistent well-defined

fashion. The development and implementation

of a safety goal would certainly assist the

regulation of the industry. However, its absence should not deter the development of the

standards and criteria necessary for the next

generation of nuclear powerpl ants.

Under the safety-block concept, the vendor,

either alone or in conjunction with an AE,

would develop and obtain regulatory approval

of a standard design which consolidates in a

single design certain parts of the plant which

traditionally have been split between the vendor and the AE. This would enable one designer or design group to have total system

responsibility for the entire nuclear part of the

plant and to better anticipate the impact of

various events on the entire plant. This approach would eliminate a number of interfaces

that create difficulties in design and licensing,

since all the systems crucial for licensing

would be inside the safety-block portion of the

plant. Approval of the power-generating systems should be wholly routine. The safety

block approach should therefore facilitate the

licensing process and allow a more thorough

design approval to take place. In either case,

the AE firms would retain the bulk of their

function. This concept would require the vendor and perhaps the AE to expand their scope

of design responsibilities and accept the resulting additional liability, The utility, therefore,

would have to accept a lower degree of involvement than under the acceleration of present policies.

The single-standard design would require

creating an entirely new design organization.

This has the very real possibility of disrupting

the existing institutions which design, construct, operate, and regulate nuclear plants.

Given the possibility of replicating an undetected safety flaw in all the plants of a

single-standard design and the necessity of

relating operating experience to the mixed set

of plants already in place, the safety benefits

of such an approach are doubtful. The singledesign approach has the greatest problems

with antitrust as well. The existing Atomic

Energy Act would have to be drastically

modified to enforce this approach and would

transfer the incentive and responsibility for

design improvements from the industrial

participants, who now have the responsibility,

to an umbrella design organization. There is no

private industry in the United States that has

undergone such a radical change. The net effect of imposing a single design on the utilities

is impossible to judge.

An alternative approach is to have a separate body go ahead with the design of a “national reactor” or “yardstick” design, even

without a commitment to actually build them.

This exercise would allow a comparison with

existing designs and possibly would bring improvements to them. Such a design would have

to recognize the problems associated with

combining components or systems in ways not

previously done and without any operational

experience base for its performance. Such a

yardstick could more easily be achieved by

tightening the existing criteria to meet the

utilities requirements for availability, reliability, and safety. This yardstick could then be

used outside the licensing and regulatory

framework to measure the relative weaknesses

or strengths of existing designs.

o