NCRP report no 120 dose control at nuclear power plants

Dose control a t nuclear power plants 1 recommendations of the National Council on Radiation Protection and Measurements... "BNL ALARA center experience with an information exchange s

NCRP REPORT No 120

NUCLEAR POWER PLANTS

Recommendations of the

NATIONAL COUNCIL O N RADIATION

PROTECTION AND MEASUREMENTS

Issued December 30, 1994

National Council on Radiation Protection and Measurements

791 0 Woodmont Avenue / Bethesda, Maryland 2081 4-3095

LEGAI NOTICE

This Report was prepared by the National Council on Radiation Protection and Measurements (NCRP) The Council strives to provide accurate, complete and useful information in its documents However, neither the NCRP, the members of NCRP, other persons contributing to or assisting in the preparation of this Report, nor any person acting on the behalf of any of these parties: (a) makes any warranty or representation, express or implied, with respect to the accuracy, completeness or usefulness of the information contained in this Report, or that the use of any informa- tion, method or process disclosed in this Report &ay not infringe on privately owned rights; or (b) assumes any liability with respect to the use of, or for damages resulting from the use of any information, method or process disclosed in this Report, under the Civil Rights Act of 1964 Section 701 et seq as amended 42 U.S.C Section 2000e

et seq (Title VII) or any other statutory or common law theory governing liability

Library of Congress Cataloging-in-Publication D a t a

National Council on Radiation Protection and Measurements

Dose control a t nuclear power plants 1 recommendations of the National Council on Radiation Protection and Measurements

Contents

Preface . 1 Introduction

1.1 Scope 1.2 Background

1.3 The ALARA Principle

1.4 Range of Applicability 1.5 Qualitative Aspects of ALARA

1.6 Quantitative Aspects of ALARA

1.7 Implementation

2 Nuclear Power Dose Experience

2.1 Status of Nuclear Power Generation

2.2 Exposures in the United States

2.3 Goals for 1990 and 1995

2.4 Comparisons with Other Countries

3 Quantitative Methods in Optimization

3.1 Introduction

3.2 Conceptual Approach

3.3 Valuation of Marginal Value of Dose Avoided

3.3.1 Valuations Based on Surveys and Past Practices

3.3.2 Impact of Required Crew Changes

3.3.3 Impact of Replacement Power Costs 3.3.4 Job-Specific Marginal Values of Dose Reduction :

3.4 Eqaluation of Options

3.4.1 Preliminary Screening of Options

3.4.2 Aggregating Present and Future Costs

and Benefits 3.4.3 Present-Value Calculation

3.5 Comparison of Options 3.6 Sensitivity Analyses

4 Management, Policy and Administration for an ALARA Program

4.1 Organization 4.2 Management Direction

vi 1 CONTENTS

4.2.1 Corporate Policy 4.2.2 Radiation Protection Manual

4.2.3 Procedures

4.2.4 Responsibilities

4.2.5 Goals and Objectives

4.3 Training 4.3.1 Radiation Protection Technicians and

Supervisors

4.3.2 Station Employees

4.3.3 Engineers

4.3.4 Managers and Supervisors 4.3.5 Examinations

4.3.6 Continuing Training

4.4 Monitoring Program Performance

4.4.1 Quantitative Indicators 4.4.2 Performance Monitoring

4.5 Assessment

5 Dose Control Principles in Reactor D e s i m and Modification

5.1 Introduction

5.2 ALARA Principle in the Design Process

5.3 Design Dose Objectives

5.4 Design Responsibilities and Organizational Structure for Implementing the Dose Control Principles

5.5 Radiation Dose-Reduction Technology

5.5.1 Source Reduction Design Factors

5.5.1.1 Cobalt Source Reduction

5.5.1.2 System Chemistry and Metallurgy

5.5.1.3 System Decontamination 5.5.1.4 Fuel Integrity

5.5.2 System Integrity and Contamination

5.5.3 Time, Distance and Shielding

5.5.3.1 Plant and Equipment Reliability 5.5.3.2 Ease of Maintenance,

Operation, Inspection and Access 5.5.4 System Layout

5.5.5 Remote Operation 5.5.6 Robotics

5.5.7 Shielding 6 Operational Considerations

6.1 Introduction

6.2 Radioactive Source Reduction

CONTENTS 1

6.2.1 Optimum pH

6.2.2 Exclusion of Extraneous Materials

6.2.3 Cobalt Reduction

6.2.4 Hydrogen Water Chemistry and Zinc Injection in Boiling Water Reactors

6.2.5 Hot Drain-Off of Feedwater Systems and Condensers Following Shutdown and Prior to Start-Up in Boiling Water

Reactors 6.2.6 Ultra-Fine Filters for Letdown System

6.2.7 Oxygen Injection into Boiling Water Reactor Feedwater

6.2.8 Iron Controls in Boiling Water Reactors

6.3 Contamination Control

6.3.1 Proper Maintenance and Operations

6.3.2 Optimization of Plant Contamination Control 6.3.3 Protective Clothing andfor Respirator

Optimization 6.3.4 Decontamination of Work Areas

6.3.5 Use of Tents Containment Bags and Glove Boxes

6.3.6 Use of Local Ventilation

6.3.7 Boiling Water Reactor Controlled

Shutdowns 6.3.8 Pressurized Water Reactor Shutdown Chemistry

6.3.9 Decontamination of Primary System and

Components 6.3.10 Fuel Cladding Integrity

6.4 External Exposure Control

6.4.1 External Exposure Control During Job Setup 6.4.1.1 Radiation Work Permit

6.4.1.2 Decontamination of Components 6.4.1.3 AleridAlarm Systems

6.4.1.4 Posting

6.4.1.5 Radiation and Airborne Radioactivity

Monitoring 6.4.1.6 Temporary Shielding

6.4.1.7 Pre-Job Briefing

6.42 External Exposure Control During Work

Activities 6.4.2.1 Job Supervisor

6.4.2.2 Radiation Protection Technician

Coverage

viil / CONTENTS 6.4.2.3 Auxiliary Operators Routine

Activities 92

6.4.2.4 Radiation Workers 92 6.4.2.5 Special Tooling and Robotics 93

6.4.2.6 Communications 93

6.4.2.7 Dose Tracking 93

6.4.3 Post-Job Activities 94

6.4.3.1 Dose Accounting 94

6.4.3.2 Post-Job Reviews 94

6.4.3.3 Documentation 94

6.4.3.4 ALARA Reports 94

6.5 Planning 95

6.5.1 General 95

6.5.2 Task Planning 95

6.5.2.1 Defining the Job Scope 96 6.5.2.2 Photos Video Tape and Video

Mapping 96

6.5.2.3 Radiological Surveys 96

6.5.2.4 Pre-Job Inspections Dry Runs 97 6.5.2.5 Dose Estimates 97

6.5.3 Outage Coordination 97

6.5.3.1 Radiological Support Personnel 98

6.5.3.2 Schedule Considerations 98

6.5.3.3 Coordination of Outage Tasks 98 Appendix A Pre-Job ALARA Briefing 100

Appendix B Pre-Job ALARA Checklist 101

Appendix C ALARA Reviews-Graded

Approach 103 References 104

The NCRP 115

NCRP Publications 123 Index 134

References

ACAD (1991) The Objectives and Criteria for Accreditation of Training i n the Nuclear Power Industry, ACAD 91-105 (National Academy for Nuclear

Training, Atlanta, Georgia)

AEC (1971) U.S Atomic Energy Commission "Light-water-cooled nuclear power reactors," Federal Register 35 FR 18385 (U.S Government Printing Office, Washington)

AECB (1991) Atomic Energy Control Board, Advisory Committee on Radia- tion Protection Application of the ALARA Process in the Regulation of Nuclear Activities, Report No INFO-0387:AC-2 (National Technical Infor-

mation Service, Springfield, Virginia)

AIF (1980) Atomic Industrial Forum A n Assessment of Engineering Tech- niques for Reducing Occupational Radiation Exposure a t Operating Nuclear Power Plants, prepared by AIF Subcommittee on Engineering

Techniques for Reducing Occupational Exposures (Atomic Industrial Forum, Bethesda, Maryland)

ANSIIANS (1985) American National Standards InstituteIAmerican Nuclear Society Guidelines on the Nuclear Analysis and Design of Con- crete Radiation Shielding for Nuclear Plants ANSIIANS 6.4-1985 (Ameri-

can Nuclear Society, La Grange Park, Illinois)

ANSIIANS (1987) American National Standards InstituteIAmerican Nuclear Society Qualification and Training of Personnel for Nuclear Power Plants, ANSIIANS 3.1-1987 (American Nuclear Society, La Grange

Park, Illinois)

ANSIIANS (1988) American National Standards InstituteIAmerican Nuclear Society Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants, ANSVANS 3.2-1988 (Ameri-

can Nuclear Society, La Grange Park, Illinois)

BAKER, D.A (1993) Dose Commitments Due to Radioactive Releases from

Nuclear Power Plant Sites i n 1989, U.S Nuclear Regulatory Commission

NUREGICR-2850, PNL-4221,ll (National Technical Information Service, Springfield, Virginia)

BAUM, J.W (1991a) "ALARA at nuclear power plants," pages 165 to 181

in Proceedings of the 24th Midyear Topical Meeting of the Health Physics Society on Implementation of Current NCRP and ICRP Guidance and Revised 10 CFR Part 20, Jorgensen, D.P., Seagondollar, L.W and Watson,

J.E., Jr., Eds (Health Physics Society, McLean, Virginia)

BAUM, J.W (199lb) "Valuation of dose avoided at U.S nuclear power plants," Nucl Plant J 9, 40-47

REFERENCES / 105

BAUM, J.W (1994) Value ofpublic Health and Safety Actions, U.S Nuclear Regulatory Commission NUREG/CR6213, BNL-NUREG-52413 (National Technical Information Service, Springfield, Virginia)

BAUM, J.W and HORAN, J.R (1985) Summary of ComparativeAssessment

of U.S and Foreign Nuclear Power Plant Dose Experience, U.S Nuclear Regulatory C o m m i s s i o n NUREGICR-4381, BNL-NUREG-51918 (National Technical Information Service, Springfield, Virginia)

BAUM, J.W and K H A N , T.A (1986) Occupational Dose Reduction at Nuclear P1ants:Annotated Bibliography of Selected Readings i n Radiation Protection and ALARA, U.S Nuclear Regulatory Commission NUREGI CR-3469, BNL-NUREG-51708, 3 (National Technical Information Ser- vice, Springfield, Virginia)

BAUM, J.W and KHAN, T.A (1992) "BNL ALARA center experience with

an information exchange system on dose control at nuclear power plants," pages 270 to 282 in NEA Workshop on Work Management i n Occupational Dose Control (Organization for Economic Cooperation and Development, Nuclear Energy Agency, Paris)

BAUM, J.W and MATTHEWS, G.R (1985) Compendium of Cost-Effective- ness Evaluations of Modifications for Dose Reduction at Nuclear Power Plants, U.S Nuclear Regulatory Commission NUREG/CR-4373 (National Technical Information Service, Springfield, Virginia)

BAUM, J.W and SCHULT, D.A (1984) Occupational Dose Reduction at Nuclear P1ants:Annotated Bibliography of Selected Readings in Radiation Protection and RLARA, U.S Nuclear Regulatory Commission NUREGI CR-3469, BNL-NUREG-51708, 1 (National Technical Information Ser- vice, Springfield, Virginia)

BAUM, J.W and WEILANDICS, C (1985) Occupational Dose Reduction at Nuclear Plants: Annotated Bibliography of Selected Readings i n Radiation Protection and ALARA, U.S Nuclear Regulatory Commission NUREGI CR-3469, BNL-NUREG-51708, 2 (National Technical Information Ser-

vice, Springfield, Virginia)

BAUM, J.W., DIONNE, B.J and KHAN, T.A., Eds (1989) Proceedings of the International Workshop on New Developments i n Occupational Dose Control a n d m Implementation a t Nuclear Power Plants and Similar Facilities, U S Nuclear Regulatory Commission NUREGICP-0110, BNL- NUREG-52226 (National Technical Information Senrice, Springfield, Vir- ginia)

BENGEL, P.R and FOLTMAN, A.J (1986) 'T h e TMI-2 remote technology program," pages 49 t o 60 in Proceedings of the Workshop on Requirements

of Mobile Teleoperators for Radiological Emergency Response and Recou- ery, Report No ANLIEES-TM-261 (Argonne National Laboratory, Argonne, Illinois)

BENGTSSON, G and HOGBERG, L (1988) "Status o f achievements reached in applying optimisation of protection i n prevention and mitiga- tion of accidents in nuclear facilities," i n Proceedings of a n Ad Hoc Meeting

on the Application of Optimisation of Protection i n Regulation and Opera- tional Practices (Organization for Economic Cooperation and Develop- ment, Nuclear Energy Agency, Paris)

106 / REFERENCES

BENGTSSON, G and MOBERG, L (1993) "What is a reasonable cost for protection against radiation and other risks?" Health Phys 64, 661-666 BENINSON, D and GONZALEZ, A.J (1981) "Optimization of nuclear safety systems," pages 449 to 455 in Proceedings of a n International Con- ference on Current Nuclear Power Plant Safety Issues, Volume 11, IAEA- STIIPUB/566 (International Atomic Energy Agency, Vienna)

BERGMANN, C.A and LAUDERMAN, E.I (1984) Cobalt Release from PWR Valves, EPRI NP-3445 (Electric Power Research Institute, Palo Alto, California)

BERTHET, A., BOUSSARD, P., LOCHARD, J., BRISSAUD, A., ROLLIN,

P and LEFAURE, C (1992) "Valems de r6f6rence de l'unit6 de dose collective professionnelle pour la mise en oeuvre de la politique "ALARA"

dans les centrales nucleaires d'Electricit6 de France," Radioprotection 27,411-421

BESLU, P., ANTHONI, S., BRISSAUD, A., RIDOUX, P., CHEVALIER, C and SAURIN, P (1989) V e n d of plant radiation fields of French reactors: Analysis and perspectives," pages 1 to 6 in Water Chemistry of Nuclear Reactor Systems 5, 2 (British Nuclear Energy Society, London)

BOC (1989) U.S Bureau of Census Statistical Abstract of the United States,

1989, 109th ed (U.S Government Printing Office, Washington)

BROOKS, B.G (1985) Occupational Radiation Exposure a t Commercial Nuclear Power Reactors 1983 Annual Report, U.S Nuclear Regulatory Commission NUREG-0713, 5 (National Technical Information Service, Springfield, Virginia)

CHARBONNEAU, S (1987) "Nuclear products and services," Nucl Eng

28, 121-126

CLARK, M.J., FLEISHMAN, A.B and WEBB, G.A.M (1981) Optimisation

of the Radiological Protection of the Public, NRPB-R12O (National Radio- logical Protection Board, Chilton, Didcot, Oxon, United Kingdom) COHEN, J.J (1970) "Plowshare: New challenge for the health physicist," Health Phys 19, 633-639

COHEN, B.L (1980) "Society's valuation of life saving in radiation protec- tion and other contexts," Health Phys 38, 33-51

COHEN, S.C., GOLDIN, D.J., GOLDIN, A.S and EDWARDS, D.W (1986)

"Occupational radiation exposure implications of NRC-initiated multi- plant actions," pages 269 to 279 in Proceedings ofASME IANS Bi-Annual Nuclear Power Conference: Safety and Reliability (American Society of Mechanical Engineers, New York)

COMLEY, G.C.W and ROOFTHOOFT, R (1988) "Recent chemistry studies

a t the Belgian PWRs, DOEL 3 and DOEL 4," page 9 in Proceedings of EPRI Seminar on PWR Water Chemistry and Radiation Field Control (Electric Power Research Institute, Palo Alto, California)

CRAWFORD, G.S (1993) "Remote handling equipment aids Bruce," Nucl Eng Int 38,33-34

DIONNE, B.J and BAUM J.W (1985) Occupational Dose Reduction and

ALARA a t Nuclear Power Plants: Study on High-Dose Jobs, Radwaste Handling, and AURA Incentives, U.S Nuclear Regulatory Commission

REFERENCES 1 107

NUREGICR-4254, BNL-NUREG-51888 (Brookhaven National Labora- tory, Upton, New York)

DIONNE, B.J., MEINHOLD, C.B., KHAN TA and BAUM J.W (1990)

Occupational Dose Reduction at Department of Energy Contractor Facili- ties: Study ofALARA Programs-Status 1989 (Brookhaven National Labo-

ratory, Upton, New York)

DUBOURG, M (1985) "Designing for minimum man-rem," Nucl Eng Inter

Kingdom," pages 121 to 132 in Proceedings of the International Workshop

on New Developments in Occupational Dose Control and ALARA Imple- mentation at Nuclear Power Plants and Similar Facilities, U.S Nuclear

Regulatory Commission NUREGICP-0110, BNL-NUREG-52226 (National Technical Information Service, Springfield, Virginia)

EICKELPASCH, N and LASCH, M (1986) "Investigations on transport and activation of corrosion products in the 2 x 1300 MWel twin boiling

water reactors of Gundremmingen," pages 55 to 58 in Proceedings of

Water Chemistry of Nuclear Reactor Systems 4 , 2 (British Nuclear Energy

Society, London)

EPRI (1987) Electric Power Research Institute Guidelines for Permanent

BWR Hydrogen Water Chemistry Installations-1987Revision, EPRI NP-

5283-SR-A (Electric Power Research Institute, Palo Alto, California)

EPRI (1989) Electric Power Research Institute NOREM Wear Resistant,

Iron-Based Hard Facing Alloys, EPRI NP-6466 (Electric Power Research

Institute, Palo Alto, California)

EPRI (1990a) Electric Power Research Institute Cobalt Reduction Guide-

lines, EPRI NP-6737 (Electric Power Research Institute, Palo Alto, Cali-

fornia)

EPRI (1990b) Electric Power Research Institute PWR Primary Water

Chemistry Guidelines: Revision 2, EPRI NP-7077 (Electric Power Resea~ch

Institute, Palo Alto, California)

EPRI (1993a) Electric Power Research Institute Cobalt Reduction Guide-

lines Revision 1, EPRI-TR-103296 (Electric Power Research Institute, Palo

Alto, California)

EPRI (1993b) Electric Power Research Institute PWR Primary Shutdown

and Startup Chemistry, EPRI TR-101884 (Electric Power Research Insti-

tute, Palo Alto, California)

EPRI (1993~) Electric Power Research Institute PWR Secondary Water

Chemistry Guidelines-Revision 3, EPRI TR-102134 (Electric Power Research Institute, Palo Alto, California)

EPRI (1993d) Electric Power Research Institute BWR Primary System

Activity Transients During Plant Shutdowns, EPRI TR-103536 (Electric

Power Research Institute, Palo Alto, California)

108 / REFERENCES

EPRI (1994) Electric Power Research Institute BWR Normal Water Chem-

istry Guidelines-1993 Revision Normal and Hydrogen Water Chemistry,

Report Summary, EPRI TR-103515 (Electric Power Research Institute, Palo Alto, California)

ERDA (1976) Energy Research and Development Administration MORT

User's Manual, ERDA-76145-4, SSDC-4 (National Technical Information

Service, Springfield, Virginia)

FIGLHUBER, D., BRETTSCHUH, W and KISON, H (1984) "Plant inspec-

tion experience and the equipment needed," pages 285 to 298 in Nuclear

Power Plant Outage Experience, IAEA STYPUB 669 (International Atomic

Energy Agency, Vienna)

GAWRON, J.P (1989) "INPO programs in dose control," pages 451 to 465

in Proceedings of the International Workshop on New Developments in

Occupational Dose Control and ALARA Implementation at Nuclear Power Plants and Similar Facilities, Baum, J.W., Dionne, B.J and Khan, T.A.,

Eds U.S Nuclear Regulatory Commission NUREGICR-0110, BNL- NUREG-52226, (National Technical Information Service, Springfield, Vir- ginia)

GILCHRIST, R.L., SELBY, J.M and WEDLICK, H.L (1978) Technical

Guidelines for Maintaining Occupational Exposures as Low as Practicable: Phase I-Summary of Current Practices, PNL-2663 (Pacific Northwest

Laboratories, Richland, Washington)

GOLD, R.E (1987) Proceedings: 1985 Workshop on Primary-Side Stress

Corrosion Cracking of PWR Steam Generator Tubing, pages 4-1 to 4-11,

EPRI NP-5158 (Electric Power Research Institute, Palo Alto, California)

HEARD, D.B and FREEMAN, R.J (1983) Cobalt Contamination Resulting

from Value Maintenance, EPRI NP-3220 (Electric Power Research Insti-

tute, Palo Alto, California)

HEDGRAN, A and LINDELL, B (1970) "PQR-A special way of thinking," Health Phys 19, 121 (abs)

HORAN, J.R., BAUM, J.W and DIONNE, B.J (1985) Proceedings of a n

International Workshop on Historic Dose Experience and Dose Reduction

(ALARA) at Nuclear Power Plants, U.S Nuclear Regulatory Commission

NUREGICP-0066, BNL-NUREG-51901 (National Technical Information Services, Springfield, Virginia)

HUMMELGREN, L (1990) "Cobalt reduction in BWR," pages 187 to 188

in Proceedings of the International Workshop on New Developments i n

Occupational Dose Control and ALARA Implementation at Nuclear Power Plants and Similar Facilities, Baum, J.W., Dionne, B.J and Khan, T.A.,

Eds., U.S Nuclear Regulatory Commission NUREGICP-0110, BNL- NUREG-52226 (National Technical Information Service, Springfield, Vir- ginia)

IAEA (1985a) International Atomic Energy Agency Assigning a Value to

Transboundary Radiation Exposure, Safety Series No 67 (International

Atomic Energy Agency, Vienna)

IAEA (1985b) International Atomic Energy Agency Design Aspects ofRadi-

ation Protection for Nuclear Power Plants-A Safety Guide, Safety Series

No 50-SG-D9 (International Atomic Energy Agency, Vienna)

55, Annals of the ICRP 20 (Pergamon Press, Elmsford, New York) ICRP (1991) International Commission on Radiological Protection 1990

Recommendations of the Znternational Commission on Radiological Pro- tection, ICRP Publication 60, Annals of the ICRP 21 (Pergamon Press, Elmsford, New York)

JONES, R.L., WOOD, C.B., COWAN, R.L., HEAD, R.A., LINN, C.C and WONG, T.L (1987) "Hydrogen injection mini-tests completed a t eight U.S BWR's," Nucl Eng Int 32, 25-26

JUDGE, J.J (1992) "Remote operated vehicles-a driving force for improved outages," Nucl Eng Int 37, 34-36

KATHREN, R.L., SELBY, J.M and VALLARIO, E.J (1980) A Guide to

Reducing Radiation Exposure to as Low as Reasonably Achievable, DOE1

EVl1830-T5 (U.S Department of Energy, Washington)

KATHREN, R.L., MASSE, F., MOSSMAN, K., ROESSLER, G a n d SCHIAGER, K (1993) "Scientific and public issues committee position statement: radiation dose limits for the general public," Health Phys Newsl XM, No 5, 13-17

KAURIN, D.G., KHAN, T.A., SULLIVAN, S.G and BAUM, J.W (1993)

Occupational Dose Reduction at Nuclear Plants: Annotated Bibliography

of Selected Readings i n Radiation Protection and ALARA, U.S Nuclear

Regulatory Commission NUREGICR-3469, BNL-NUREG-51708, 7 (National Technical Information Service, Springfield, Virginia)

KHAN, T.A and BAUM, J.W (1988) Worldwide Activities on the Reduction

of Occupational Exposure at Nuclear Power Plants, U.S Nuclear Regula-

tory Commission NUREGICR-5158, BNL-NUREG-52086, 1 (National Technical Information Service, Springfield, Virginia)

KHAN, T.A and BAUM, J.W (1989) Occupational Dose Reduction at Nuclear Plants: Annotated Bibliography of Selected Readings i n Radiation

Protection and ALARA, U.S Nuclear Regulatory Commission NUREGI

CR-3469, BNL-NUREG-51708, 4 (National Technical Information Ser- vice, Springfield, Virginia)

KHAN, T.A and W, C (1994) Data Base on Reduction Research Projects for Nuclear Power Plants, U.S Nuclear Regulatory Commission NUREGI

CR-4409, BNL-NUREG-51934, 5 (National Technical Information Ser- vice, Sprhglleld, Virginia)

110 / REFERENCES

KHAN, T.A., TAN, H., BAUM, J.W andDIONNE, B.J (1990) Occupational

Dose Reduction at Nuclear Plants: Annotated Bibliography of Selected Readings in Radiation Protection and ALrlRA, U.S Nuclear Regulatory Commission NLTREGICR-3469, BNL-NUREG-51708,5 (National Techni- cal Information Service, Springfield, Virginia)

KHAN, T.A., W L I N , D.S., LANE, S.G and BAUM, J.W (1991) Occupa-

tional Dose Reduction at Nuclear Plants: Annotated Bibliography of Selected Readings in Radiation Protection and-, U.S Nuclear Reg- ulatory Commission NUREGICR-3469, BNL-NUREG-51708,6 (National Technical Information Service, Springfield, Virginia)

KUMAZAWA, S., NELSON, D.R and RICHARDSON, A.C.B (1984) Occu-

pational Exposure to Ionizing Radiation in the United States-A Compre- hensive Review for the Years 1960-1985, EPA 52011-84-005 (U.S Environmental Protection Agency, Washington)

LEDERBERG, J (1971) "Squaring an infinite circle: Radiobiology and the value of life," Bull A Sci 27, 43-46

LEINE, L (1984) "Design for maintainability," pages 141 to 155 in Nuclear

Power Plant Outage Experience, LAEA STVPUB/669 (International Atomic

Energy Agency, Vienna)

LIN, C.C and SMITH, F.R (1988) BWR Cobalt Deposition Studies: Final

Report, EPRI NP-5808 (Electric Power Research Institute, Palo Alto, Cali-

fornia)

LISTER, D.H and DAVIDSON, R.D (1989) Corrosion-Product Release in

LWRs: 1984-1985 Progress Report, EPRI NP-4741 (Electric Power

Research Institute, Palo Alto, California)

LISTER, D.H and GODIN, M.S (1990) The Effect of Dissolved Zinc on the

Transport of Corrosion Products in PWRs," EPRI NP-6949-D (Electric

Power Research Institute, Palo Alto, California)

LOCHARD, J and BENEDITTINI, M (1987) Expositions Professionnelles

dans les Reacteurs Globaux entre 1975 et 1985, Report 10 (Centre d'etude sur l'evaluation de la Protection dans le Domaine Nucleaire, Fontenay- aux-Roses, France)

LUNDGREN, K and ELKERT, J (1990) "ALARA recommendations based

on Swedish BWR experience," pages 483 to 492 in Proceedings of the

International Workshop on New Developments in Occupational Dose Con- trol and ALARA Implementation at Nuclear Power Plants and Similar Facilities, Baum, J.W., Dionne, B.J and Khan, T.A., Eds., U.S Nuclear

Regulatory Commission NUREGICP-0110, BNL-NUREG-52226 (National Technical Information Service, Springfield, Virginia)

MANGENO, J.J and TYRON, A.E (1988) Occupational Radiation Expo-

sure from U.S Naval Nuclear Propulsion Plants and Their Support Facili- ties, NT-88-2 (US Department of the Navy, Washington)

MANN, B.J and COHEN, S.C (1986) Estimating Collective Dose in Nuclear

Facilities with Emphasis on the Design Process, AIFMESP-039 (Atomic

Industrial Forum, Bethesda, Maryland)

MARBLE, W.J and COWAN, R.L (1991) "Mitigation of radiation buildup

in the BWR by feedwater," in Proceedings of the 1991 JMF International

REFERENCES / 111

Conference on Water Chemistry in Nuclear Plants, CONF-910447 (Japan Atomic Industrial Forum, Inc., Tokyo)

MARBLE, W.J., WOOD, C.B., LEIGHTY, C.E and GREEN, T.A (1986)

"BWR radiation buildup control with ionic zinc," pages 144 to 151 in Proceedings of ASME IANS Bi-Annual Nuclear Power Conference: Safety

a n d Reliability (American Society of Mechanical Engineers, New York) MARCHL, T.F and REITZNER, U (1992) "Chemistry parameters influ- encing the dose rate buildup in BWR plants," pages 33 to 38 in Water Chemistry of Nuclear Reactor Systems 6 (British Nuclear Energy Soci- ety, London)

MARJN, A (1988) "The cost of avoiding death-nuclear power, regulation and spending on safety," Royal Bank of Scotland Review 157, 20-36 MILLER, T.R (1990) "The plausible range for the value of life-red herrings among the mackerel," J Forensic Econ 3, 17-39

MOCHIZUKI, H (1985) "Status of development and application of robots for nuclear plants," pages 99 to 113 in Proceedings of a n International Workshop on Historic Dose Experience a n d Dose Reduction (AWLRA) a t Nuclear Power Plants, U.S Nuclear Regulatory Commission NUREGI CP-0066, BNL-NUREG51901 (National Technical Information Service, Springfield, Virginia)

NASINRC (1990) National Academy of Sciences/National Research Council, Committee on the Biological Effects of Ionizing Radiation Health Effects of Exposure to Low Levels of Ionizing Radiation, BEIR V (National Academy Press, Washington)

NCRP (1954) National Committee on Radiation Protection Permissible Dose from External Sources of Ionizing Radiation, NCRP Report No 17, out of print

NCRP (1978) National Council on Radiation Protection and Measurements Operational Radiation Safety Program, NCRP Report No 59 (National Council on Radiation Protection and Measurements, Bethesda, Mary- land)

NCRP (1983) National Council on Radiation Protection and Measurements Operational Radiation Safety-Training, NCRP Report No 71 (National Council on Radiation Protection and Measurements, Bethesda, Mary- land)

NCRP (1989) National Council on Radiation Protection and Measurements Exposure of the U.S Population from Occupational Radiation, NCRP Report No 101 (National Council on Radiation Protection and Measure- ments, Bethesda, Maryland)

NCRP (1993) National Council on Radiation Protection and Measurements Limitation of Exposure to Ionizing Radiation, NCRP Report NO 116 ( N a t i o n a l Council o n R a d i a t i o n Protection a n d M e a s u r e m e n t s , Bethesda, Maryland)

NEELEY, V.I and WALKER, S.M (1990) BWR Nondestructive Evaluation Source Book: Final Report, EPRI NP-6879-D (Electric Power Research Institute, Palo Alto, California)

NEI (1986) Nuclear Engineering International "Optimization of radiation exposure in BWRs," Nucl Eng Int 31, 27-28

112 / REFERENCES

NEI (1993a) Nuclear Engineering International World Nuclear Industry Handbook 1993 (Nuclear Engineering International, Sutton, Surrey,

United Kingdom)

NEI (1993b) Nuclear Engineering International "Are we reaching the limits

of automated inspection in Japan?" Nucl Eng Int 38, 34

NISHINO, Y., NAGASE, M., SAWA, T., UGHIDA, S and OSHUMI, K

(1992) "Reactions of iron crud with metallic ions under BWR water condi- tions," pages 63 to 68 in Water Chemistry of Nuclear Reactor Systems 6,

1 (British Nuclear Energy Society, London)

NRC (1975) U.S Nuclear Regulatory Commission "Numerical guides for design objectives and limiting conditions for operation to meet the criterion

"as low a s is reasonably achievable, for radioactive material in light-water- cooled nuclear power reactor effluents," 10 CFR 50, Appendix I, 40 FR

19439, May 5, 1975; amended 40 FR 40516, September 4, 1975; 40 FR

33029, December 19, 1975 (US Government Printing Office, Washington) NRC (1977a) U.S Nuclear Regulatory Commission "Calculation of annual doses to man from routine releases of reactor effluent for the purpose of evaluating compliance with 10 CFR part 50, Appendix I," NRC Regulatory Guide 1.109, Revision 1 (U.S Government Printing Office, Washington) NRC (1977b) U.S Nuclear Regulatory Commission Operating Philosophy for Maintaining Occupational Radiation Exposures as Low as is Reason- ably Achievable NRC Regulatory Guide 8.10, Revision 1-R (U.S Govern-

ment Printing Office, Washington)

NRC (1978) U.S Nuclear Regulatory Commission Information Relevant

to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will be as Low as is Reasonably Achievable, NRC Regulatory

Guide 8.8, Revision 3 (U.S Government Printing Office, Washington) NRC (1993) U.S Nuclear Regulatory Commission.AL4RA Levels for Efflu- ents from Materials Facilities, NRC Regulatory Guide 8.37 (U.S Govern-

ment Printing Office, Washington)

NRC (1994) U.S Nuclear Regulatory Commission "Radiological criteria for decommissioning; proposed rule," Federal Register, 10 CRF Part 20

(U.S Government Printing Office, Washington)

NRC (1995) U.S Nuclear Regulatory Commission Proceedings of the Third International Workshop on the Implementation of ALARA a t Nuclear Power Plants NUREG/CP-0143, BNL-NUREG-52440, in press

NRPB (1993) National Radiological Protection Board Values of Unit Collec- tive Dose for Use in the 1990s, Doc NRPB 4, No 2, pages 75 to 80 (National

Radiological Protection Board, Chilton, Didwt, Oxon, United Kingdom) NSC (1993) National Safety Council Accident Facts, 1992 ed (Chicago, Illi-

REFERENCES 1 113

Facilities, Baum, J.W., Dionne, B.J and Khan, T.A., Eds., U.S Nuclear Regulatory Commission NUREGICP-0110, BNL-NUREG-52226 (National Technical Information Service, Springfield, Virginia)

OCKEN, H and WOOD, C.B (1992) Radiation-Field Control Manual-

1991 Revision, EPRI TR-100265 (Electric Power Research Institute, Palo Alto, California)

OLDFIELD, M.A (1988) "Health physics i n developing the design of Sizewell B," Nuclear Europe 8, 16-17

OTWAY, H.J., BURNHAM, J.B and SOHRDING, R.K (1970) "Economic

vs biological risk as reactor design criteria," IEEE Trans Nucl Sci NS-

PETTIT, P.J., Ed (1981) Compendium Design Features to Reduce Occupa-

tional Radiation Exposure a t Nuclear Power Plants (Atomic Industrial Forum, Bethesda, Maryland)

POTHIER, N.E and METCALFE, R (1989) "Reliable reactor coolant pump seal performance-the station's role," Trans Am Nucl Soc 59, 86-87

RADDATZ, C.T and HAGEMEYER, D (1993) Occupational Radiation

Exposure at Commercial Nuclear Power Reactors and Other Facilities

1991, U.S Nuclear Regulatory Commission NUREG-0713 13 (National Technical Information Service, Springfield, Virginia)

RAMSAY, I, and KHAN, A.H (1992) "Assigning a monetary value to the person-sievert," HSD-HP-92-35, unpublished (Ontario Hydro, Pickering, Ontario, Canada)

REMARK, J.F (1989) A Review of Plant Decontamination Methods: 1988 Update, EPRI NP-6169 (Electric Power Research Institute, Palo Alto, Cali- fornia)

RUFFORD, N (1986) "Robotics take heat out of reactors," N Civ Eng 718,24-26

SAGAN, L.A (1972) "Human costs of nuclear power," Science 177,487-493 SMEE, J.L and BEAMAN, T.A (1984) "Chemical decontamination of the steam generator channel heads at the R.E Ginna station," Radiat Prot Manage 1, 75-84

SRPI (1991) Swedish Radiation Protection Institute "The monetary value

of collective dose reduction (a-value), statement from meeting of the Nordic

Radiation Protection Authorities," page 2 in Stralskyddsnytt (Swedish

Radiation Protection Institute, Stockholm)

UKAEA (1987) United Kingdom Atomic Energy Authority Radiological

Guidelines for the Design and Operation of URAEA Plant, Authority Code

of Practice, SRD R456 (United Kingdom Atomic Energy Authority, London)

UESUGI, N and OKANO, H (1987) "Toshiba looks to save skilled labour," Nucl Eng Int 32, 38-39

114 / REFERENCES

VANKUIKEN, J.C., GUIZIEL, K.A., WILLING, D.L andBUEHRING, W.A

(1992) Replacement Energy Costs for Nuclear Electricity-Generating Units

in the United States: 1992-1996, U.S Nuclear Regulatory Commission

NUREGICR-4012, ANL-AA-30 3 (National Technical Information Service, Springfield, Virginia)

WADA, T and WHELAN, E.P (1986) Development of Cobalt-Free Hard-

Facing Alloys for Nuclear Applications: 1985 Progress, EPRI NP-4775

(Electric Power Research Institute, Palo Alto, California)

WAGNER, D.S and THOMASSON, F.L (1989) ALARA Assessment of Use

of Extended Life Versus Standard Light Bulbs at Nuclear Power Plants, Khan, T.A and Baum J.W., Eds., U.S Nuclear Regulatory Commission NUREGICR-4409, BNL-NUREG-51934,3, BNL Data Base ID: T-121 (National Technical Information Senrice, Springfield, Virginia)

WAHLSTROEM, B.G (1987) "Minimizing occupational exposure a t Finnish nuclear power plants," Trans Am Nucl Soc 54, 187

WILSON, R., VIVIAN, G.A., CHASE, W.J., ARMITAGE, G and SENNEMA, L.J (1986) "Occupational dose reduction experience in Ontario hydro nuclear power stations," Nuclear Tech 72, 590

WOOD, C.B (1985) "Experience with the LOMI chemical decontamination system," Radiat Prot Manage 2, 51-58

WOOD, C.B (1986) Manual of Recent Techniquesfor LWR Radiation-Field

Control, EPRI NP-4505-SR (Electric Power Research Institute, Palo

Alto, California)

ZIMMER, J.J., ANSTINE, L.D and BURLEY, E.L (1985) Radiological

Effects of Hydrogen Water Chemistry, EPRI NP-4011 (Electric Power

Research Institute, Palo Alto, California)

Preface

This Report represents the first scientific committee effort by the National Council on Radiation Protection and Measurements (NCRP) devoted solely to radiation protection at nuclear power plants The members of the Committee represent a mixture of corpo- rate management, operational and regulatory personnel

This Report is another in the series of reports produced under the umbrella of NCRP Scientific Committee 46 on Operational Radiation Safety There is a wealth of material presented in this Report drawn from the more than 35 years of practical experience in nuclear power operation in this country and Canada The world's literature on the subject has been reviewed and pertinent material has been incorpo- rated into this Report The application of the principle of ALARA ( a s low a s reasonable achievable), a s promulgated by national (NCRP) and international [International Commission on Radiologi- cal Protection (ICRP)] committees, is the central theme and opera- tional examples of its use are provided

Serving on NCRP Scientific Committee 46-9 for the preparation

of this Report were:

John W Baum, Chairman

Brookhaven National Laboratory

Upton, New York

Members

William R Kindley Dennis M Quinn

Institute of Nuclear Power New York Power Authority Operations Buchanan, New York

Atlanta, Georgia

Thomas D Murphy Alan K Roecklein

U.S Nuclear Regulatory U.S Nuclear Regulatory

iv / PREFACE

Consultant

Bruce J Dionne

Brookhaven National Laboratory

Upton, New York

NCRP Secretariat

James A Spahn, Jr., Senior Staff Scientist

Cindy L O'Brien, Editorial Assistant

The Council wishes to express its appreciation to the Committee members for the time and effort devoted to the preparation of this Report

Charles B Meinhold

President, NCRP

1 Introduction

1.1 Scope

This Report addresses the implementation of dose control for occu- pational exposure associated with operation and maintenance of United States nuclear power plants The Report is based on practical experience from 35 y of operation of nuclear power plants throughout the world as well as the extensive literature available I t is based

on techniques and programs which have been generally successful

in controlling and reducing doses, and on the application of the principle ofALARA (as low as reasonably achievable) in its programs The emphasis is on exposure of workers, the main contributor to collective dose1 at these plants, rather than on exposures to the public or from emergency situations Exposures to the public have resulted in collective doses less than one percent of occupational collective doses during t h e 1980s (Baker, 1993) The Report is intended for a wide audience of students, utility management, plant management, engineers, health physicists and other radiation pro- tection specialists

This Report provides a multidisciplinary approach to the principles

of dose control a t nuclear power plants In Section 2, it provides a background and history of the dose control efforts in nuclear power plants Section 3 presents quantitative methods that can be used a s part of the decision-making process in satisfying dose limits and optimizing radiation exposure, and a s guidelines for t h e use of these methods

One key objective of the use of the principle of ALARA a t nuclear

power plants is to ensure that management of the utility and plant is committed to implementing a program for optimizing doses Section 4

provides an approach management can use to organize, direct and administer an ALARA program Also in Section 4, the Report pro- vides ways to assess the effectiveness of these programs, including evaluation techniques, program reviews, procedure reviews and accountability

'Collective dose and dose as used throughout this Report refer to the "effective" doses, and "effective dose equivalents" specified in recent ICRP and NCRP reports

2 1 1 INTRODUCTION

An integral part of any effective dose control program is assurance

Specific design considerations that have proven successful in reduc- ing doses at nuclear power plants are discussed in Section 5 Section 6 discusses operational considerations necessary for implementing dose control during the operational lifetime of the plant

The NCRP has provided recommendations on many aspects of radiation protection in several reports that form the basis for the

limits for exposure to ionizing radiation, the NCRP, in its Report

No 116 (NCRP, 1993), states that the application of dose limits as specified is not sufficient in itself As explained, radiation protection

nomic and social factors being taken into account Keeping exposures ALARAwill generally result in levels substantially below dose limits Due to the large amounts of radioactivity contained in nuclear power plant systems, maintenance operations include both high- and low- dose jobs Consequently, dose control efforts are needed to both sat-

Some of the past reports and documents that address certain key aspects of the dose control process are the following: NCRP Report

No 59 (NCRP, 1978) which provides an overview of the elements of

(NCRP, 1983) which gives additional guidance on operational radia- tion safety training In Report No 101 (NCRP, 1989), the NCRP summarized exposure of the United States population from occupa- tional radiation The annual collective dose equivalent for nuclear power plant personnel represents approximately 25 percent of the collective dose equivalent received by all monitored workers The U.S Nuclear Regulatory Commission (NRC), in Regulatory Guide 8.8 (NRC, 1978), proposed a qualitative approach to implementation

of the ALARA principle a t nuclear power plants without specific cost-benefit analysis The United States commercial nuclear electric generating industry has also, through the Institute of Nuclear Power Operations CINPO), addressed radiological protection programs

Several ICRP publications provide important details for the philo- sophical and practical implementation of dose control principles ICRP Publication 22 (ICRP, 1973) discussed the implication of ICRP

1.3 THE ALARA PRINCIPLE / 3

recommendations that doses be kept ALARA ICRP Publication 37

(ICRP, 1983) provides details on the cost-benefit approach to the optimization of radiation protection ICRP Publication 55 (ICRP,

1989) provides a thorough discussion of additional quantitative approaches for the optimization process, including multiattribute utility analysis and multicriteria outranking analysis ICRP Publica- tion 60 (ICRP, 1991) and NCRP Report No 116 (NCRP, 1993) provide comprehensive updates on biological and conceptual bases for radia- tion protection practices, and update the basic recommendations of the two organizations

The ALARA Center at Brookhaven National Laboratory has pub- lished a series of bibliographies (Baum and Khan, 1986; Baum and Schult, 1984; Baum and Weilandics, 1985; Kaurin et al., 1993; JChan and Baum, 1989; Khan et al., 1990; 1991) of selected readings,in radiation protection and ALARA in a project aimed a t collection and dissemination of information on dose control a t nuclear power plants The Center also maintains a number of related data bases (Baum and Khan, 1992) Information from the data bases on dose reduction research and health physics technology projects for nuclear power plants are published periodically (Khan and Yu, 1994)

1.3 The ALARA Principle

The goal of radiation protection is to reduce the probability of radiation-induced diseases in persons exposed to radiation (somatic effects) and in their progeny (genetic effects) to a degree that is reasonably achievable and acceptable in relation to the benefits from the activities that involve such exposure For doses below the individ- ual dose limits, the probability of detrimental effects is assumed to

be proportional to dose based on known mechanisms of biological damage and limited human epidemiological data Hence, radiation protection practice requires that exposures to ionizing radiation be kept not only below established limits, but also to levels which are ALARA, economic and social factors being taken into account The ALARA principle was used a s early as 1954 in NCRP Report No 17

(NCRP, 1954) Over the years it has changed mainly in the terminol- ogy used to express the concept (Baum et al., 1989), and with greater emphasis on the quantitative aspects of the process (ICRP, 1983; 1989)

A basic difficulty encountered in implementing the ALARA princi- ple involves making judgments about what is meant by "low" and

"reasonably achievable." Generally, these judgments are qualitative

4 1 1 INTRODUCTION

During normal operations, and during the planning of many, if not most, jobs, qualitative judgment based on experience with similar jobs or activities is sufficient and appropriate to achieve ALARA conditions However, when new jobs, especially those involving potentially large (greater than 0.01 Sv) exposures, are being planned,

it will need to be recognized that quantitative input to the decision process is important Substantial progress has been made in develop- ing appropriate quantitative methods

At the time that this Report is being written, there is continuing social pressure to reduce risks ofnuclear accidents, radioactive waste and, to a lesser degree, to continue to reduce personnel radiation exposures a t nuclear power plants Thus, the management a t many nuclear plants has been willing to apply large resources to reduce occupational exposures, not only to protect workers health and safety, but also because of the benefit from reduced societal criticism

1.4 Range of Applicability

NCRP, t h e process of keeping doses ALARA is usually regarded a s extending from the dose limit to background dose levels, or a t least

to levels considered negligible As worker doses are reduced, the marginal costs of additional dose reduction should equal the benefits

of avoided detriment The process, then, should be self-limiting, because a t sufficiently low doses, even t h e costs of evaluation approach or exceed the benefits expected At this point, the optimiza- tion process dictates that no further action is justified Therefore, in principle, no negligible dose level needs to be defined or applied, though one may be desirable i n administrative or regulatory contexts

At the other extreme, where doses approach the administrative

or regulatory limit, other factors become important A s individual

inequitable distribution of potential risk among the workers and to avoid legal, licensing, personnel and public relations costs that may

be incurred if doses are judged not to be ALAFtA As a result, larger monetary expenditures can be justified to avoid individual doses near the limits These costs are often associated with the P term a s specified in ICRP publications that relate to nonobjective health detriments (ICRP, 1983; 1989) They could also be considered as the cost of ensuring that dose limits are not exceeded In this case, they

in practice, the distinction is not important The process of finding

1.4 RANGE OF APPLICABILITY / 5

the most cost-effective method of meeting dose limits and the process

of optimization below the limits are so similar and, in some ways overlapping, t h a t separation is not possible For example, many workers receive doses on a variety of jobs throughout the year If their doses are likely to approach dose limits, the value of dose avoided is affected on each job (i.e., the employer may be willing to pay more to avoid dose to this worker than for the worker not near the limits)

However, the value of dose avoided related to a specific piece of equipment or specific job may involve both workers likely to receive doses near the limit and those likely to accumulate much lower doses throughout the year In principle, then, the value of dose avoided for these jobs should be a weighted average of values determined by all the workers who receive doses from the equipment or job In practice, worker doses and equipment related doses are generally not broken down and analyzed in the detail needed for precise deter- mination of value of dose avoided for each piece of equipment or each job Rather, values are developed that are applicable to specific high-dose jobs and are used in decisions related to these jobs The same or lower values may be applied plant-wide, depending on man- agement's qualitative judgment concerning overall plant perfor- mance, costs and budgetary constraints

Throughout this Report, considerable emphasis is placed on the ALARA principle, since it needs, and is receiving, increasing atten- tion as collective and individual doses continue to decline in the nuclear power industry As individual doses are lowered, the costs

of protection shift from being determined primarily by the need to avoid the dose limits, to a need to achieve an optimum balance between costs and benefits This is the dose region in which the ALARA principle is most important However, in many cases, the term "dose control" is employed in this Report to better reflect the fact that considerable emphasis must still be placed on reducing exposures with the aim of avoiding doses near the limits

An important aspect of the ALARA principle is its breadth of applicability The ALARA principle is not limited to reducing expo- sures of individual radiation workers, but appropriately includes radiation workers collectively, nonradiation workers, members of the general public and the environment as well Exposures to individ- uals, as well as the total population dose, are therefore considered Thus, even though a particular task can be accomplished with rela- tively low individual doses, the task may not in fact conform to the ALARA philosophy if the collective effective dose to the group or the general population is excessive when costs and net benefits are considered

6 1 1 INTRODUCTION

1.5 Qualitative Aspects of ALARA

Radiation protection practices have long emphasized avoiding unnecessary exposures (NCRP, 1954) This principle, now referred

procedures and practices that involve qualitative judgments which are based on many years of experience Thus, the most important

protection philosophy that relies heavily on a "culture" that includes education, training and experience This leads to a way of operation

on a day-to-day basis that includes a constant awareness and atten- tion to the avoidance of unnecessary exposures This Report includes material that illustrates how many of these qualitative aspects are commonly implemented in current nuclear power plants The mate- rial is not intended to be comprehensive in covering these basic practices, but rather is intended to focus on those needing special emphasis in nuclear power plants

1.6 Quantitative Aspects of AZSLRA

of cost-benefit analysis is employed often in the nuclear power indus- try due to the relatively larger individual and collective doses usually encountered, and to the large monetary costs that can result if doses cause either an extension ofthe time the plant is shut down or require hiring and training additional crews to avoid individual worker doses that might exceed the dose limits

of dose avoided (see Section 3 for details) They may derive one value for an individual high-dose job and another for plant-wide application Resulting values are currently expressed in terms of

"dollars per person-rem" (dollars per person-Sv or dollars per person- mSv will be used throughout this Report) These values are generally much larger than the value that might be deduced for objective health detriment [a in the ICRP (1983) cost-benefit equations] They are also generally higher than values considered sufficient to cover the other detriments such as consideration of dose distribution among workers usually covered by the p term in ICRP equations The reason for this is that the values of dose avoided in nuclear power plants also reflect the industry's commitment to providing

a safe working environment and the increased monetary costs of

1.7 IMPLEMENTATION / 7

production that are affected by the high doses being encountered Although these latter costs may not normally be considered "detri- ment" costs, they are oRen included in the costs of radiation protec- tion Such costs oRen result from a need to avoid exceeding radiation protection effective dose limits and related operational requirements These costs, caused by the need to avoid limits, affect the value assigned to dose avoided and are therefore included in cost-benefit evaluations related to implementation of dose control at United States nuclear power plants

It is important to bear in mind that these "values of dose avoided"

in this Report are specific to the nuclear power plants in the United States at the present time, are plant specific, and are not likely to apply to other industries, or even in the nuclear power industry in future years if plant collective doses and maximum individual doses continue to decline

However, even larger monetary values of dose avoided could apply

if annual and lifetime dose limits continue to be lowered since more workers may then be near the individual dose limits This could cause the costs attributed to dose reduction or the hiring of replacement workers to increase further

It is also important to realize that monetary values for dose avoided cited in this Report are not intended as recommendations of the NCRP, but rather are merely illustrative of current practice in the

nuclear power industry In other industries, where worker doses are further below the limits, a much lower value of dose avoided may

be deduced This subject is discussed in more detail in Section 3.3

and will be considered in a related NCRP operational radiation safety programs report currently in preparation

1.7 Implementation

There are differences in the conduct of dose control programs depending on utility organization and worker exposure experience at individual plants Pressure for reducing doses is particularly strong when plants experience collective doses above the average Decisions

to reduce dose are based on both prudence and quantitative cost- benefit analyses During the 1980s, most nuclear power plant opera- tors made substantial efforts to reduce exposures As a result, both individual and collective exposures were reduced by a factor of about two in United States, Japanese, German and Canadian plants (see Section 2)

Both qualitative and quantitative methods and techniques for implementing dose control and the ALARA principle are included

2 Nuclear Power Dose

Experience

2.1 Status of Nuclear Power Generation

The first United States nuclear plant to generate significant amounts of electricity was the Shippingport Plant in Pennsylvania, which was a government-owned pressurized water reactor (PWR) operated by Duquesne Light Company I t began operations i n December 1957, was converted to a light-water breeder reactor in

1977 and was permanently shut down after the light-water breeder experiments were completed i n October 1982 Subsequently, it was disassembled and removed from the site Other demonstration plants with a fairly long history of operations include the Dresden 1 Plant,

a 207 MWe (megawatt electric) boiling water reactor (BWR) operated

by the Commonwealth Edison Company in Illinois from August 1960 until it was shut down in 1978; the Big Rock Point Plant in Michigan,

a 69 MWe BWR which was placed in operation by Consumers Power Company in December 1965 and continues to operate; the Yankee Rowe Plant, a 175 MWe PWR that was operated by Yankee Atomic Electric in Massachusetts from 1961 to 1993 A few other small demonstration plants were operated in the early 1960s for short periods

power plants that could generate about 106 GW (gigawatt) ofelectric- ity, or about 21 percent of total United States generation, and 31 percent of world-wide nuclear capacity These values were projected

to change by less than four percent by the year 2000 (NEI, 1993a)

Of the 110 plants, 37 were licensed BWR and 73 were licensed PWR plants There were seven PWR plants and one BWR plant that had construction permits Several units have been prematurely shut down in addition to those mentioned above These are the Hanford

N Reactor (a dual purpose facility), a light water graphite moderated unit; Shoreham; Indian Point 1; Three Mile Island 2; Rancho Seco and Fort St Vrain Information on the decontamination and decom- missioning of these facilities should be useful in judging the effective- ness of previous designs for decontamination and decommissioning

10 / 2 NUCLEAR POWER DOSE EXPERIENCE

2.2 Exposures in the United States

Consistent with the growth in the number of nuclear plants used

to generate electricity, the contribution to United States occupational collective effective dose from the nuclear fuel cycle increased from

an almost insignificant level in 1960 to a level that was a major contributor to occupational doses by 1980 Increased maintenance needed on older plants and a number of plant modifications man- dated by the NRC also contributed to this increase (Cohen et al.,

Matthews, 1985)

Table 2.1 shows radiation exposure data reported by utilities to the NRC for the years 1973 to 1991 (Raddatz and Hagemeyer, 1993) These d a t a show t h a t total collective dose leveled off a t about

was established starting in 1985, even though the number of nuclear power plants continued to increase Figure 2.1 shows this same trend of decreasing exposures while the gross electricity produced increases in the same period Consequently, the exposure per nuclear power plant showed a decreasing trend in the 1980s as shown in Figures 2.2 and 2.3, with a factor of four reduction by 1992 Although workers were permitted (under then existing regula-

cumulative limit of 5 (age-18) rem, most utilities used a 5 rem y-'

or lower administrative limit through 1984 After 1984 all utilities used a 50 mSv y-I or lower administrative limit and instituted provis- ions to restrict all workers including contractors from exceeding

50 mSv y-' while on the utility's station ( a station may have more than one nuclear power plant) However, the number of workers exceeding 50 mSv y-' is not easily determined from the present reporting system since transient workers may accumulate doses a t two or more stations Thus, the number of workers receiving more than 50 mSv y-I may be greater than the number reported by sta- tions, since a transient worker could accumulate a large fraction of

50 mSv y-' a t each of two or more stations, and the sum would exceed 50 mSv y-l Table 2.2 shows both the number of workers

TABLE 2.1-Summary of annual information reported by United States commercial light-water cooled reactors 1973 to 1991"

(Raddatz and Hagemeyer, 1993)

Average Number of Average Average Personnel Collective Average Annual Number of Gross Collective with Dose per Electricity Average Number of Collective Workers with Electricity Average Dose Dose per Measurable MWe y Generated per Rated Reactors Doses Measurable Generated per Worker Reactor Doses per (person- Reactor Capacity Net Year Included (person-Sv) Doses (MWe y) (mSv) (person-Sv) Reactor mSv/MWe y) (MWe y) (MWe)

2.3 GOALS FOR 1990 AND 1995 1 15 reported to exceed 50 mSv y-l a n d t h e number t h a t exceeded

50 mSv y-l when the doses a t more than one station are considered [Note that such workers were not permitted to exceed the cumulative limit of 5 (age-18) rem Five rem is equal to 50 mSv1

To accomplish reduction in collective dose and the number of work- ers exceeding 50 mSv y-l, United States utilities implemented sev- eral dose reduction techniques Without such techniques, it is likely that the collective dose would have risen significantly because of the design modifications for improved safety introduced following the accident a t Three Mile Island 2 in 1979, stricter environmental quali- fications for instrumentation, installation of improved fire and seis- mic safety systems, and special "in-service" inspections (ISI) required

a t increasing frequencies by the NRC Many of the specific techniques used by utilities to achieve this dose reduction were discussed a t the international workshops held a t Brookhaven National Laboratory (Baum et al., 1989; Horan et al., 1985; NRC, 1995) and form much

of the basis for the following sections of this Report

2.3 Goals for 1990 and 1995

In 1985, the commercial nuclear power industry in the United States established a set of 5 y goals as one method to encourage safety improvements ~ncluded among these goals were collective dose goals (person-Sv per unit) for BWR and PWR plants To help

TABLE 2.2-Annual whole-body doses exceeding 50 mSu at nuclear power facilities

(Raddatz and Hagemeyer, 1993)

Reported Number Corrected NumbeP

"orrection is based on statistical analyses of dose distributions and numbers of

transient workers (see Brooks, 1985)

16 / 2 NUCLEAR POWER DOSE EXPERIENCE

establish these goals, three groups were constituted to make recom- mendations These groups were composed of selected (1) reactor plant executives, (2) nuclear steam system vendors, suppliers and architect-engineers, and (3) international participants of INPO The combined recommendations of these three groups were provided to the industry Based on these recommendations, each utility estab- lished goals for 1990 The individual utility goals were averaged to establish industry wide goals The 1990 average goals for BWR and PWR plants were 4.69 person-Sv per unit and 2.88 person-Sv per unit, respectively Average collective doses for United States plants from 1980 through 1993 are shown on Figures 2.2 and 2.3

Comparing the 1990 goals with the 1984 exposure data, the latest data available to the industry when the goals were set in 1985, shows the goals were aggressive and challenging in that an approximately

50 percent reduction in exposure was needed to meet them

Establishment of goals for key performance indicators has contin- ued, and a set of 5 y goals was established for 1995, again by industry consensus For BWRs the goal is 2.55 person-Sv per reactor The goal for the PWRs is 1.85 person-Sv per reactor These goals provide incentive to reduce collective doses However, no study has been published to demonstrate that the goals are consistent with the

ALARA or optimization process, which is the preferred "goal."

2.4 Comparisons with Other Countries

Several comparisons have been made between collective doses in the United States stations and those in several other countries

1984; Lochard and Benedittini, 1987) A comparison for 1990 is illus- trated in Figure 2.4.' The reactors compared in Figure 2.4 are PWRs, pressurized heavy-water reactors (Canada), or BWRs Gas-cooled reactors, such a s those operated by Nuclear Electric in the United Kingdom, are not included because they have significantly lower radiation levels due to their different designs, and because t h e United States has none in operation for comparison

The Canadian plants, owned and operated by Ontario Hydro, pro- vide an outstanding example of the success achieved by a large utility (16 operating nuclear power plants) that made large dose reductions

to reach the status shown in Figure 2.4 The Canadian heavy-water moderated reactors are of quite different design than reactors with

'Personal communication, T.A Khan, Brookhaven National Laboratory, Upton, New York (1992)

0

Fig 2.4 Comparison between collective doses in the United States and several other countries

18 / 2 NUCLEAR POWER DOSE EXPERIENCE

light-water moderation (PWRs) more commonly utilized in t h e United States and other countries The heavy-water reactors produce much more tritium, which is a major additional problem in control- ling exposure with this type of system Also, the reactor is refueled

during operation (without shutdown), using a highly sophisticated refueling machine Early versions of this reactor design, located at the Douglas Point Station, resulted in collective doses about four times higher than United States doses per unit of electricity gener- ated during 1967 to 1970 However, through a broad major manage- ment commitment to improved design for dose control, the Canadian utility successfully reduced doses from about 40 person-mSv MWe y-'

Much of this improvement was achieved by eliminating alloys with

a high cobalt content from the primary system, the addition of shield- ing, improvements in water purification systems, improvements in air-drying systems (for tritium control), improved reliability and ease of maintenance, and the use of fewer workers The number of workers per reactor was reduced from about 600 in 1970 to about

300 in 1982 The Canadian reduction in dose is an outstanding example of what can be achieved by proper design and operation if

of the importance management placed on the control of dose a t that time was the monetary value used to judge acceptability or "break- even" in the cost-benefit evaluations A value of $1,000 per person-mSv avoided was employed, which was high for that time period This value was based on the cost of manpower replacement that was being experienced in the plants with large collective dose

Doses a t Japanese plants during 1978 to 1982 were high, due in part, to a policy which required more extensive dismantling and testing of components than was required in other countries during annual shutdowns for preventive maintenance More recently, changes in this policy and other dose reduction improvements have lowered Japanese collective dose by more than a factor of two Some

of this reduction may also be attributed to the beneficial effects of earlier efforts in preventive maintenance

Doses a t French plants have generally been low for several rea- sons First, all but a few French plants are PWRs In the United States, PWRs generally have about half the collective dose equivalent

of BWRs of similar size Second, the French plants are, on average, a few years newer than United States PWRs Since corrosion products buildup with time, collective dose from maintenance, testing, inspec- tions and replacement of components tends to increase with the age

of the plant A significant portion of the difference is due to these factors Third, the French plant designs are standardized with more

2.4 COMPARISONS WITH OTHER COUNTRIES 1 19 units per station The standardized plants and multiple units make design and use more cost effective for automatic and remote tooling Therefore, work crews can be trained more thoroughly and used more effectively Recent increases in doses a t French plants are attributable to plant aging and the need for 10 y ISIs Doses are now higher than a t United States PWRs

Similar increases in doses for the Taiwan plants in 1990 are attrib-

operations during that year and the fact that values are based on only six units

Considering PWR evolution since 1975, for the United States, Japan, and to a smaller degree, Sweden, Belgium, and plants in western Germany, the average annual collective dose per reactor

or per unit electricity produced first increased and then decreased (Wilson et al., 1986) By contrast, values are stable or rising slowly

in Switzerland and France

Exposure controls in Sweden and Finland have been consistently good (Horan et al., 1985) The BWR plants in these countries are designed by a Swedish steam supplier, ASEA-Atom (now called ABB Atom) The most recent of these plants have internal recirculation pumps which eliminate a major source of exposure due to leaks from the pump seals, and eliminate the piping needed with external pumps The plant design and layout provide better shielding and segregation of radioactive components They provide for adequate work space for both routine and special maintenance Both countries also have excellent programs for controlling the chemistry of the cooling water, which result in water purity and pH control among the best in the world Careful chemistry is important since water impurities have a major impact on the generation of corrosion prod- ucts, intergranular stress corrosion, corrosion related failure and component degradation Sweden also has used advanced electronic dosimetry to read dosimeters a t various work stations routinely; to track effective dose by worker, location and specific task; and to provide updated information on accumulated dose of workers with respect to dose limits The success in Sweden may also be due to the goal of 2 person-mSv MWe-' installed capacity, suggested in the 1970s by the Swedish National Institute of Radiation Protection This is equivalent to about 3 person-mSv MWe y-I generated, a n ambitious but apparently achievable goal (Baum and Horan, 1985) These are some of the reasons for the country-to-country differ- ences, but there are others that relate to policies, management and operational practices, regulatory requirements, special equipment and plant design These differences demonstrate, moreover, that countries w i t h lower exposures may b e u s i n g cost-effective

20 / 2 NUCLEAR POWER DOSE EXPERIENCE

approaches of potential benefit to countries with higher exposures

I t can be concluded, for example, that some plants in the United States could further reduce occupational exposure by implementing several of the techniques used by countries with lower exposures Furthermore, all countries could benefit from incorporating these concepts into new designs The following sections discuss some of the more important concepts and techniques developed and used

in the United States and other countries to reduce occupational exposures over the last 10 y Some of the factors contributing to low

doses are listed here and discussed in detail in Sections 5 and 6:

control of oxygen and pH in the primary system

minimization of the cobalt content of primary system components exposed to coolant water

minimization of the formation of corrosion products by maintain- ing high primary and secondary system water purity

plant design, layout, and component segregation and shielding management interest and commitment

number of workers and in-depth worker training

use of remote and robotic tools

standardization of plant

decontamination of the primary system

pretreatment of surfaces (passivation) to minimize corrosion and deposition

addition of Zn to BWR (and possibly PWR) primary water systems

addition of hydrogen toBWR primary systems to avoid intergran- ular stress corrosion cracking of primary system components

3 Quantitative Methods in

Optimization

3.1 Introduction

Optimization of radiation protection which involves consideration

of both qualitative and quantitative factors, is only one element i n the decision process However, it is important that a quantification process be employed for radiation protection decisions, when possi- ble, in order to achieve consistency This is most important when doses are large or costs are significant relative to doses that may be avoided Even when a decision is based primarily on nonquantifiable social, political or regulatory factors, quantitative analyses of options may still be valuable in judging the relative merits of the various options

Quantitative approaches to the optimization of radiation protec- tion have been described in detail in a number of publications (Clark

et al., 1981; ICRP, 1983; 1989) A recent ICRP publication (ICRP, 1989) provides a structured approach to solving problems, including descriptions of quantitative decision-aiding techniques and exam- ples of their application

The basic ICRP system of dose limitation requires justification

of a practice, optimization of radiation protection (keeping doses

ALARA) and keeping worker doses below limits These concepts are strongly endorsed by the NCRP (1993) Justification is achieved

by requiring that the net benefit of the practice be positive after considering costs of production, protection and radiation detriment

I n addition, a n optimal level of radiation protection is to be

achieved, i.e., collective doses are to be controlled to "a level ALARA."

The ALARA condition is predicated on the assumption that the prac- tice has met the justification criterion and is constrained only by dose limits and the objective of maximizing net benefit

A major difficulty arises i n the application of these principles to the practice of generating electricity by nuclear power plants This difficulty is due to the number of workers who receive or may receive doses near administrative or legal dose limits The value of dose