ML20135B803
| ML20135B803 | |
| Person / Time | |
|---|---|
| Issue date: | 01/31/1997 |
| From: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| References | |
| NUREG-BR-0184, NUREG-BR-184, NUDOCS 9703030205 | |
| Download: ML20135B803 (289) | |
Text
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. United States i i l Nuclear Regulatory Commission '% ,,,, /
Regulatory Analysis l Technical Evaluation l Handbook l l
Final Report i
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Omce of Nielcar Regulatory Research January 1997 ,-
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s I NUREG/BR-0184 REGULATORY ANALYSIS TECHNICAL EVALUATION HANDBOOK
,-- $ FINAL . JANUARY 1997
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, United States i 1 Nuclear Regulatory Commission ,,,,, .M l
! Regulatory Analysis i Technical Evaluation
! Handbook Final Report i ,
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. Office of Nuclear Regulatory Research l
- January 1997 i
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n Abstract i,
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'j The purpose of this Handbook is to provide guidance to the regulatory analyst to promote preparation of quality regulatory analysis documents and to implement the policies of the Regulatory Analysis Guidelines of the U.S.
Nuclear Regulatory Commission (NUREG/BR-0058 Rev. 2). This Handbook expands upon policy concepts l included in the NRC Guidelines and translates the six steps in preparing regulatory analyses into implementable l methodologies for the analyst. It provides standardized methods of preparation and presentation of regulatory i analyses, with the inclusion of input that will satisfy all backfit requirements and requirements of NRC's Committee to Review Generic Requirements. Information on the objectives of the safety goal evaluation l processs and potential data sources for preparing a safety goal evaluation is also included. Consistent application l of the methods provided here will result in more directly comparable analyses, thus aiding decision-makers in evaluating and comparing various regulatory actions.
The handbook is being issued in loose-leaf format to facilitate revisions. NRC intends to periodically revise the handbook as new and improved guidance, data, and methods become available.
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O kv iii NUREG/BR-0184
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COBtents R
i Abstract.................................................................... iii {
l I i Foreword.................................................................... xv j
. r Acknowledgments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xvii ;
i u Abbreviations and Acronyms ........................................................ xix i t
l 1 Introduction ....................................... ...................... 1.1 l t
1.1 Purpose............................................................. 1.2 ;
4 L2 Regulatony Analysis Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2 l
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j 1.2.1 Key Terms and Cec @ . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . 1.2 i j .1.2.2 Steps in a Regulatory Analysis . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3 :
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! 1.3 Handbook overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.4 l f 1.4 Endnotes br Chapter 1 . . . . . . . . . . . . . ....................................... 1.5 1 i
- 2.1 2 2 Scope of a Regulatory Analysis . . . . . . . . . . . ....................................... i
, e i 2.1 When a Regulatory Analysis is Required . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1 l
{ 2.2 When a Back6t Regulatory Analysis is Required . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1 !
I 2.3 When a CRGR Regulatory Analysis is Required . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4 l a 2.4 Imel of Detail . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.6 :
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2.5 Units .............................................................. 2.7 1 2.6 Regulatory Relaxations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.7
] 2.7 Endnotes br Chapter 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ~ . . . . . . . . . . . . . . 2.9 s
j 3 Safety Goal Evaluation for Operation of Nuclear Pbner Plants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1 9
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j 3.1 Endnotes for Chapter 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2 r
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1 4 Regulatory Analysis Methods and Supporting Inbrmation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 4
4.1 Statement of the Problem and Objective . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 4.2 Identification and Preliminary Analysis of Alternative Approaches . . . . . . . . . . . . . . . . . . . . . . . . . 4.3 4 4.3 Estimation and Evaluation of Wlues and Impsets . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.5 s . 4.4 Presentation of Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.8
, 4.5 Decision Rationale . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.9 1 4.6 Implementation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.10
. 4.7 Endnotes for Chapter 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.11 l 5 Value-impact Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.1 4
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5.1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ............. 5.1 5.2 Methods . . . . . . . . . . . . . . ... ...................... .................... 5.1 5.3 Standard Analysis . . . . . . . . . .... ....................................... 5.2 5.4 Treatment of Uncertainty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... .......... 5.3 5.4.1 'lypes of Uncertainty . . . ........... .................. ............. 5.3 5.4.2 Uncertainty Versus Sensitivity Analysis . . . . . . . . . . . . . . . . . . . . . ............... 5.4 5.4.3 Uncertainty / Sensitivity Analyses . . . . . .............. .................... 5.5 5.4.3.1 N UREG-I ISO . . . . . . . . . . . . . . . . . . . . . . . . . . .................... 5.5 5.4.3.2 NUREG/CR-5381. . .......................................... 5.6 5.4.3.3 NUREG/CR-4832 . . . . . . . . . . . . . . . . . . . . . ....................... 5.7 5.4.4 Suggested Approach . .. ................... ....................... 5.7 5.5 Identification of Attributes . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... ................ 5.8 5.5.1 Public Health (Accident) . . . . . . . . ................................. ... 5.10 5.5.2 Public Health (Routine) . . . . . . . . . . . . . . . . . . . . . ......... .. . .......... 5.10 5.5.3 Occupationa! Health (Accident) .............. ............ ........... 5.10 5.5.4 Occupational Health (Routine) ....... .............. ..... ............. 5.10 5.5.5 Offsite Property . . ................................................ 5.11 5.5.6 Onsite Property . . . . .... . . ....... ... .......................... 5.11 5.5.7 Industry lmplementation . . . . . . . ...................................... 5.11 5.5.8 Industry Operation . . . . . . . .... .......... . ...................... 5.11 5.5.9 NRC Implementation ........... ......... .......................... 5.11 5.5.10 NRC Operation . . . . . . . . . ..... ....................... .......... 5.12 5.5.11 Other Government .......... ... .... .. .............. ... . .... 5.12 5.5.12 General Public . . . . . . . ... ............................. . ........ 5.12 5.5.13 Improvements in Knowledge . . .................... . ............. 5.12 5.5.14 Regulatory Efficiency . , . . . . . . . . . . . . . . .. ........ .............. 5.13 5.5.15 Antitrust Considerations . ... ..... ....... .. ..... .. . ........ 5.13 5.5.16 Safeguards and Security Considerations . . . . . . . . . . . . . ..... .............. 5.13 5.5.17 Environmental Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . ... ........... 5.13 5.5.18 Other Considerations . .... ...... .. ........... ... .... ..... .. 5.14 5.6 Quantification of Change in Accident Frequency . . . . . . . . . . . . ......................, 5.14 5.6.1 Identification of Affected Parameters . . . .. ...... . . .. ..... ......... 5.15 5.6.2 Estimation of Affected Parameter Values . . . . . . . ........ .. ............. . 5.18 5.6.3 CN.nge in Accident Frequency .. . ............ ... ...... ......... 5.19 5.7 Quantification of Attributes . ... .... ......... . .... ... ...... . ..... 5.20 5.7.1 Public Health (Accident) . . ... . ... . .. .. . . . . . ....... 5.22 5.7.1.1 Estimation of Accident-Related Health Effects . . . . . . . . . . . .. .... . 5.22 5.7.1.2 Monetary Valuation of Accident-Related Health Effects . .. .. ..... 5.26 5.7.1.3 Discounting Monetized Value of Accident Related Health Effects .... .. .. 5.26 NUREG/BR-0184 vi
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5.7.2 Public Health (Routine) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.27 l 1
1 5.7.2.1 Estimation of Change in Routine Exposure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.28 !
1 5.7.2.2 Monetary Valuation of Routine Exposure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.28 !
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i 5.7.3 Occupational Health (Accident) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.29 4
i 1 5.7.3.1 Estimation of Accident-Related Exposures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.30 5.7.3.2 Monetary Valuation of Accident.Related Exposures . . . . . . . . . . . . . . . . . . . . . . . 5.32 ,
i 5.7.3.3 Discounting Monctized Values of Accident.Related Exposures . . . . . . . . . . . . . . . . 5.32 l
- 5.7.4 Occupationa Health (Routine) ............................ . . . . . . . . ..... 5.34
- 5.7.4.1 Estimation of Change in Routine Exposure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.35 '
I l 5.7.4.2 Monetary Valuation of Routine Exposure . . . . . . . . . . . . . . . . . . . . . . . . ..... 5.37 i 5.7.4.3 Nonradiological Occupational Impacts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.37 l
I 5.7.5 Offsite Property . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.37 I
5.7.6 Onsite Property . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.40 1
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- 5.7.6.1 Cleanup and Decontammation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... 5.41 5.7.6.2 Long. Term Replacement Ibwer .................................... 5.43 ,
- 5.7.6.3 Repair and Refurbishment . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . ..... 5.45 ;
- 5.7.6.4 Total Onsite Property Damage Costs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.45 j 5,7.7 Industry Implementation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.49 l
5.7.7.1 Short. Term Replacement Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.51 j 5.7.7.2 Premature Facility Closing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.52 l 4
i 5.7.8 Industry Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.52 J
5.7.9 NRC implementation ................................................ 5.54 l
- 5.7.10 NRC Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.55
! 5.7.11 Other Governnvent .................................................. 5.56
' 5.58 5.7.12 General Public . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
5.7.13 Improvements in Knowledge . . . . . ......................,............... 5.58 4 5.7.14 Regulatory Efficiency . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.59 4
5.7.15 Antitrust Considerations . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.60 5.7.16 Safeguards and Security Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... 5.61
- 5.7.17 Environmental Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.61 5.7.18 Other Considerations ................. . . . . . . . . . . . . . . . . . . . . . . . ...... 5.61 l
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- 5.8 Summanzation of Value-impact Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.62 2
5.9 Endnotes for Chapter $ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.63 6 References ...................... ............... . . . . . . . . . . . . . . . . . ...... 6.1 vii NUREG/BR-0184
Contents Appendix A - Regulatory Analysis issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1 O
A.1 H uman Factors issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1 A.l.1 Results Documents ...... .......................................... A.2 A.I.2 Methods Documents .................... . ......................... A.4 A.2 Cumulative Accounting of Past and Ongoing Safety Improvements . . . . . . . . . . . . . . . . . . . . .... A.5 A.3 Use of Industry Risk and Cost Estimates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.6 Appendix B - Supplemental Information for Value-Impact Analyses . . . . . . . . . . . . . . . . . . ........... B.1 B.1 Numbers of Operating Power Reactors and Their Remaining Lifetimes . . . . . . . . . . . . . . . . . . . . . . B.1 B.2 Economic Discounting and Calculation of Present Value . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B.2 B.2.1 Discount Rate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B.2 B.2.2 Discrete Discounting ....... . ........................ ............ B.2 B.2.3 Continuous Discounting .. ........ ....... .......................... B.5 B.3 Occupational Exposure Experience . . . . . . .......................... ........... B.6 B.4 Calculational Method for Handbook Table 5.3, ' Expected Population Doses for Power Reactor Release Categories" ... ...................................... ........... B.17
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B.4.1 Introduction . . . . . . . . . .......... ................................. B.'17 B.4.2 MACCS input Parameter Assumptions . . . . . . . . . . . . . . . . . ................ .. B.24 Appendix C - Supplemental Information for Non-Reactor Regulatory Analyses . . . . . . . . . . . . .......... C.I C.! Facility Classes . . . . . . . . . . . . . . . . . . ...... ... .......................... C.2 C.l.1 Fuel Cycle Facilities ...................... ......................... C.2 C.I.2 Non-Fuel Cycle Facilities .... ...... ..... ... ....................... C.3 C.2 Quantification of Attributes ......... ...................................... C.3 C.2.1 Public Health (Accident) . . . . . . . .................. .................. C.4 C.2.1.1 Accident Frequencies ............................ ............. C.4 C.2.1.2 Population Doses from Accidents ............... ...... ........... C.10 C.2.1.3 Total Accident Risks . . ............ ........ .................. C.15 C.2.2 Public Health (Routine) .. ............. .. ......................... C.16 C.2.3 Occupationa1 Health (Accident) ..... ...... ............... .... ... . . C.17 C.2.4 Occupational Health (Routine) .. ........................ .. ...... C.17 C.2.5 Offsite and Onsite Pmperty . .. ..... ..... .......................... C.19 C.2.5.1 Fuel Cycle Facilities . . ... .. .. ................ ... ........ C.19 C.2.5.2 Non. Fuel Cycle Facilities ......... .... .... ................. . C.19 NUREG/BR-0184 viii
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C.3 A Preliminary Evaluation of the Economic Risk for Cleanup of Nuclear Material ucensee ;
Contamination Incidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.20 ;
,. C.4 Economic Risk of Contammation Cleanup Costs Resulting from Large Non-Reactor Nuclear l . Material Licensee Operations ................................................ C.21 j C.5 Prelimmary Characterization of Risks in the Nuclear Waste Management System Based on i information in the Literature ......... ............. ........................ C.22
- C.6 Preliminary Ranking of Nuclear Fuel Cycle Facilities on the Basis of Radiological Risks from Accidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.24 l . C.7 Cost-Benefit Analysis of United PuO21%11 cts as an Alternative Plutonium Shipping Form . . . . . . . . . C.26 l' C.8 A Regulatory Analysis on F . - y w_ for Fuel Cycle and Other Radioactive M aterial Licensees . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.26 C.9 Regulatory impact Analysis of Final Environmental Standards for Uranium Mill Tallings at Active Sites . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.29 C.10 Value-Impact Analysis of Accident Preventive and Mitigative Options for Spent Fuel Pbols ........ C.31 C.11 Nuclear Fuel Cycle Facility Accident Analysis Handbook (NUREG-1320) . . . . . . . . . . . . . . . . . . . C.33 C.12 Endnotes for Appendix C . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - C.34 Appendix D - Safety Goal Policy Statement and Back8t Rule . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D.1 !
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D.1 Safety Goals for the Operations of Nuclear Pbwer Plants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D.1 D.2 Backfit Rule . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D.11 a
l Appendix E - Index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1 l
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- 2.1 Decision tree to determine level of effort . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.8 ...
4.1 Standard format and content of regulatory analyses . . . . . . . . . 1
........................... 4.2 4.2 Steps in a value-impact analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.6 j 5.1 Summary of value-impact results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.62 C.1 Uranium process flow among fuel cycle facilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.36 C.2 Cleanup cost as a function oflicensed radionuclide quantity for non-reactor nuclear material licensees ................................................ ............. C.37 3 C.3 Normalized peak individual doses for reviewed studies of geologic waste disposal postclosure period . . . . C.37 C.4 Incremental cost of alternative control methods for uranium mill tailings . . . . . . . . . ......... ... C.38 4
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!n V) j Tables 2.1 Applications of backfit and CRGR regulatory analyses ....... . .. ...... . ...... .. .. 2.2 2.2 Checklist for specific backfit regulatory analysis requirements . . . . . .. .. .... . .... . 2.3 2.3 Checklist for specific CRGR regulatory analysis requirements . . . . . ... .. . . . . .. . 2.5 4.1 List of potential alternative actions . . ....... ....... . ..... . . . .. ....... 4.3 5.1 Checklist for identification of affected attributes .. . . ... .. .. . . .. . ... 5.9 5.2 Nuclear power plants risk assessments . . . . . . . . . ..... ... .... .. ... ... .... 5.16 5.3 Expected population doses for power reactor release categories . . . . . .... .. ... 5.23 5.4 Weighted population dose factors for the five NUREG-1150 power reactors ..... . . 5.25 5.5 Estimated occupational radiation dose from cleanup and decommissioning after a power reactor accident (person-rem or person cSv) . . .. ......... . . . . . .. . .. . .. 5.31 5.6 Weighted costs for offsite property damage for the five NUREG-1150 power reactors ... . . . . 5.38 5.7 Onsite property damage cost estimates (U) for future years (1993 dollars discounted to year ofimplementation) ............. ............... . . . ... ... 5.47 B.1 Numbers and lifetimes of operating nucien power plants . . ... . .. . . B.1 i B.2 Present value of a future dollar . . . . . . . . . ........ .. .... .... ...... . B.4 1 B.3 Present value of annuity of a dollar, received at end of each year . . . . . ..... . ......... B.4 l B.7 C Occupational dose rates by EEDB classification for PWR systems and components . .
Occupational dose rates by EEDB classification for BWR systems and components . .
B.12 B.5 (v) B.4 B.6 1991-1993 annual occupational exposure information for industrial radiographers . . ... .. ... B.18 B.7 1991-1993 annual occupational exposure information for byproduct manufacturers and distributors . ......... . . . ............ . . ..... . ... . . .. B.18 B.8 1991-1993 annual occupational exposure information for fuel fabricators . ... ... .... .. B.19 B.9 Annual occupational doses for low level waste disposal and spent fuel storage facilities, 1991-1993 .. . B.19 B.10 Summary of 1973-1993 annual occupational exposure information reported by commercial BWRs . . B.20 B.ll Summary of 1973-1993 annual occupational exposure information reported by commercial PWRs .. B.21 B.12 Summary of 1973-1993 annual occupational exposure information reported by commercial LWRs .. B.22 B.13 1993 numbers of employees and collective and average doses by occupation and personnel type at LWRs ........ ...... ... .. . . ....... .. . .. . .. B.23 C.S.1 Summary description of representative uranium fuel cycle facilities . . . .... . . . C.39 Frequency of contamination incidents for non-teactor nuclear material licensees .. C.41 C.1 .... .
Incident cleanup cost by material quantity class for non-reactor nuclear material licensees .. . C.41 C.2 .
C.3 Economic risk as a function of material application /use and licensed curie quantity for non-reactor nuclear materiallicensees .......... ..... .. .... .. . . .. . . C.42 Summary of economic risk at a reference uranium mill . . . ........ .. .. . C.43 C.4 .. .
Summary of economic risk at.a reference uranium hexafluoride conversion plant . ... C.44 C.5 . .
Summary of economic risk at a reference uranium fuel fabrication facility . .. .. . C.45 C.6 .. .
Summary of economic risk at a reference byproduct material manufacture / distribution facility ... . C.46 C.7 Summary of economic risk at a reference waste warehouse .. ... . C.47 C.8 . ... . .
Estimated 70-year population and worker exposures for repository construction . . C.47 C.9 . .
C.10 Radiation exposure from normal construction and operation for repository preclosure period ... .. . C.48
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Tables C.11 Total radiological worker fatalities from construction and emplacement periods of three alternative O
Repository Sites .. ............................ .... ........ ........... C.48 C12 Occupational dose during normal operation and from a shaft drop accident for repository preclosure period ........................................ ............. .. C48 Cl3 Public dose during normal operation and from a shaft drop accident for repository preclosure period . . . C.49 Cl4 Summary of repository accident releases, frequencies, consequences, and risk values for repository preclosure period, operations phase .......... .... ........ ........ ............ C.50 C15 Radiation exposure fmm accidents for repository preclosure period, operations phase .............. C.50 C16 Occupational dose during repository operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.51 C17 Summary of annual occupational exposures for spent fuel and HLW operation at a tuff repository ...... C.51 C18 Estimated 50-year whole-body dose commitment to the public, maximally exposed individual, and workers from accidents for repository preclosure period, operations phase ..................... C.52 C.19 Preliminary risk estimates for postulated ecidents at a repository in tuff for operations phase . . . . . . . . . C.53 C20 Frequencies and consequences of accident scenarios projected to result in offsite doses greater than 0.05 rem for repository preclosure period, operations phase .............. ............. . C.54 C21 Occupational dose during normal operation and from accidents during decommissioning and retrieval phases of a repository . . ... ............. ................................. C.55 C22 Comparison of normalized public accident risk values from various studies for repository preclosure period . . . . . . ..
............................................... C.55 C23 1985 Revised EPA estimates of 10,000-year health effects for 100,000-MTHM repositories in basalt, bedded salt, tuff and granite ....... ............ . ....... ....... ........... C.56 C24 70-year cumulative maximally exposed individual and regional population doses for the two peak dose periods for a tuff repository .. ..................... ........ ....... C.56 C25 Peak conditional cancer risks due to ingestion for the 100,000-year postclosure period for a 90,000-MTU spent fuel repository in bedded salt ...............,..................... C.57 C26 Radiation exposures from routine operations at the MRS facility ........................... C.57 C27 Radiological impacts of potential MRS facility accidents for sealed storage cask at the Clinch River Site for operations phase . . ..................................................... C.58 C28 Occupational dose from MRS facility operations . . . . . . . . . . . . ......................... C.58 C29 Summary of occupational doses fmm MRS facility operations . . . . . . . . . ...... ............ C.58 C30 Occupational dose estimates for selected MRS operations . . .............. ............. C.59 C31 Summary of MRS drywell risk analysis for operations phase . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.59 C32 Summary of results of MRS operations phase . . . . . ............................. .... C.60 C.33 Projected maximum individual exposures from normal spent fuel transpon by truck cask .................. ...................................... .... C.61 C.34 Projected maximum individual exposures from normal spent fuel transpan by rail cask ... ................... ..... ........................ ....... C.62 C35 Summary of results from the NRC for spent fuel shipments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.62 C36 Maximum individual radiation dose estimates for rail cask accidents during spent fuel transportation ... ......... .......... ................... ............... C.63 C.37 50-year population dose estimates for spent fuel rail cask accidents with no cleanup of deposited nuclides ........... ............................................. C.63 C38 Population radiation exposure from water ingestion for severe but credible spent fuel rail cask accidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......... C.64 C39 Summary of spent fuel truck and rail transponation risks ..... ........ ......... ....... C.64 C40 Summary of the routine transportation risks for the waste management system without an MRS facility . . . . . . . . . . . . . . . . . . . . . . ....... ........... ... ............ C.65 C41 Summary of the routine transportation risks for the waste management system with an MRS facility . . . . . . . . ... . .. .......... ..... ... ... ................. C.66 NUREG/BR 0184 xi:
Tables C42 Aggregated public risks for the preclosure phases of the waste management system without an M RS Facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.67 C.43 Aggregated occupational risks for the preclosure phases of the waste management system without an M RS facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.68 C44 Aggregated public risks for the preclosure phases of the waste management system with an MRS facility . . ........................................................ C 69 C45 Aggregated occupational risks for the preclosure phases of the waste management system with an M RS facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.70 C46 'Ibtal preclosure life-cycle risk estimates for the waste management system ..................... C.70 C47 Summary of annual and total life-cycle risk estimates for the waste managemenr system . . . . . . . . . . . . . C.71 C.48 Accident frequencies and population doses for milling in the nuclear fuel cycle . . . . . . . . . . . . . . . . . . C.72 C49 Accident frequencies and population doses for comersion in the nuclear fuel cycle . . . . . . . . . . . . . . . . C.72 C50 Accident frequencies and population doses for enrichment in the nuclear fuel cycle ............... C.72 C51 Accident frequencies and population doses for fuel fabrication in the nuclear fuel cycle . . . . . . . . . . . . . C.73 C52 MOX fuel refabrication radiological accident risk . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.73 C.53 Accident frequencies and population doses for MOX fuel retabrication in the nuclear fuel cycle ....... C.74 C54 ' Accident frequencies and population doses for MOX fuel refabrication in the nuclear fuel cycle ....... C.74
' C55 Accident frequencies and population doses for MOX fuel refabrication in the nuclear fuel cycle ....... C.75 C56 Fuel reprocessing radiological accident risk . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.75 C57 Accident frequencies and population doses for reprocessing in the nuclear fuel cycle . . . . . . . . . . . . . . . C.76 C58 Accident frequencies and population doses for reprocessing in the nuclear fuel cycle . . . . . . . . . . . . . . . C.77 C59 Accident frequencies and population doses for reprocessing in the nuclear fuel cycle . . . . . . . . . . . . . . . C.78 C60 Accident frequencies and popula: ion dases for reprocessing in the nuclear fuel cycle . . . . . . . . . . . . . . . C.78 C61 Accident frequencies and population doses for spent fuel storage in the nuclear fuel cycle ........... C.79
,_.x C62 Accident frequencies and population doses for solidified HLW storage in the nuclear fuel cycle . . . . . . . . C.79 C.63 Preclosure geologic waste disposal radiological accident risk . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.79 C.64 Transportation radiological accident risk . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.80 C65 Accident frequencies and population doses for transportation of spent fuel by rail and PuO2 by truck in the nuclear fuel cycle . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.80 C66 Accident frequencies and population doses for transportation in the nuclear fuel cycle . . . . . . . . . . . . . . C.81 C67 Accident frequencies and population doses for rail transportation in the nuclear fuel cycle ........... C.82 C.68 Accident frequencies and popu'ation doses for rail transportation in the nuclear fuel cycle . . . . . . . . . . . C.82 C69 Accident frequencies and population doses for rail transportation in the nuclear fuel cycle . . . . . . . . . . . C.82 C70 Normalized risk results for nuclear fuel cycle . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.83 C71 Capital equipment costs for fuel pellet fabrication . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C 84 C72 Capital equipment costs for powder reconstitution during fuel fabrication . . . . . . . . . . . . . . . . . . . . . . C.85 l C.73 Start-up operation costs for fuel fabrication . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.86 C74 Process operation costs for fuel fabrication ......................................... C.86 C.75 Summary of dose equivalent estimates for fabricating PuO2 powder to unfired pellets during fuel Fabrication ...................................................... C.87 C76 Summa y of dose equivalent estimates for reconstituting unfired PuO pellets back to powder during fuel fabrication ................................................. C.87 C77 Accident source terms and doses from uranium mill accidents ............................. C.88 C78 Ossite doses calculated for fuel fabrication plants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.88 C.79 Dose commitments from plutonium fuel fabrication facility accidents ........................ C.89 C80 Maximum ofsite individual dose commitments (Rem) from spent fuel reprocessing facility accidents .......................................................... C.89 C81 Calculated releases and doses from spent fuel storage accidents ............................ C.89 xili NUREG/BR.0184
Tables C.82 Maximum possession limits, release fractions, and doses due to a major facility fire O'
for radiopharmaceutical manufacturing . . . . . . . . . . . . . . . . . . . . . ...................... C.90 C.83 Maximum possession limits, release fractions, and doses due to a major facility fire for a radiopharmacy .............. ..... ... ................... . ........ . C.91 C.84 Maximum possession limits, release fractions, and doses due to a major facility fire for sealed source manufacturing ......... ...... ..... ...... ........ .. ............. C.92 C.85 Maximum possession limits, release fractions, and doses due to a major facility fire for university research laboratories . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.93 C 86 Waste warehousing airborne releases and doses due to a major facility the ..................... C.93 C.87 Alternative disposal standards for uranium mill tailings ................. ............... C.93 C.88 Alternative standards and control methods for existing uranium mill tailings piles . . . . . . . . . . . . . . . . . C.94 C.89 Alternative standards and control methods for new uranium mill tailings piles . . . . . . . . . ....... . C.94 C.90 Summary of values for alternative disposal standards for uranium mill tailings . . . . . ............. C.95 C.91 Cost-effectiveness of control methods for uranium mill tailings ...................... . . C 96 C.92 Summary of costs in millions of 1983 dollars for alternative disposal standards for uranium mili tailings , . ..... . . . ................ .. .... ... .. ........... C.97 C.93 Estimated risks from spent fuel pool fires . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.97 C.94 Offsite consequence calculations for spent fuel pool fires ......................... ..... C.98 C.95 Onsite property damage costs in dollars per spent fuel pool accident ......................... C.98 C.96 Incremental storage costs in 1983 dollars associated with limited loudensity racking in the primary spent fuel pool . . ...... ............................ .............. C.99 C.97 Summary of Parameters affecting attributes for the spent fuel pool inventory reduction option ......... . ........ ....................... ......... . C.100 C.98 Summary of industry-wide value-impact analysis of the spent fuel pool inventory reduction option ................. ......... .. ......................... C.101 C.99 Failure frequency for generic spent fuel pool cooling and makeup systems ......... ........... C.102 C.100 Value-impact for generic improvements to the spent fuel pool cooling system ......,...... .. .. C.103 C.101 Offsite property damage and heahh costs per spent fuel pool accident ......... ........... . C.103 C.102 Summary of industry-wide value-impact analysis of the spent fuel pool post-accident spray system ..... C.104 C.103 Facility descriptots for accident analysis .... . ... . .. ........................... C.105 C.104 Fuel manufacturing process descriptors . . . . . . . ... .... ......... .............. C.106 C.105 Fuel reprocessing process descriptors . . . . . . .. . . . . .. ....... ........... .......... C.107 C.106 Waste storage / solidification process descriptors . . .. ................................. C.108 C.107 Spent fuel storage process descriptors . . . . . ... .. ,... ... .... ...... ... ...... C.109 C.108 Behavior mechanisms for airbome particles . . .... .... .......................... C.110 C.109 Unscaled and scaled total accident risks to the public for non-reactor fuel cycle facilities .......... C.111 C.110 Preliminary occupational risk estimates for postulated accidents at a repository in tuff for preclosure operations phase of geologic waste disposal . . . . .. .......... .. .... .. .. C.!13 NUREG/BR-0184 xiv O
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$ i Foreword l I
t i 1his document is a Handbook to be used by the NRC and its contractors in the preparation of regulato y analyses to aid
! NRC decision-makers in deciding whether a proposed new regulatory requirement should be i=,M In addition, it is j i anticipmeed that the Handhook will be useful to the A.- States in their assessment of new regulatory requirements. t The Handbook is an updated and revised version of an earlier document, A Nandbookfor Hime-Inqpact Assessment l (NUREG/CR-3568), issued by the NRC in 1983. ;
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i I . The 1983 document is being updated in this Handbook to accomplish the bliowing objectives: l
- 7b renect the content d NRC's Regulatory Analysis Guidelines, NUREG/BR-0058'Rev. 2, issued in November i 1995. j i
- lb capand the scope of the Handbook to include the entire regulatory analysis process and to address facilities other :
l j than power reactors. . j
- Tb renect NRC experience and improvements in data and methodology since the 1983 Handbook was issued.
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- Tb reRect the guidance in the 1996 document, Economic Analysis ofFederal Regulations Under becutive Order 2
12866. This document was prepared by a liederalinteragency regulatory working group convened by the OfBce of ;
1 Management and Budget. i b NRC obtained review enmments on the draft Handbook from the bliowing orgamantions: Westinghouse Savannah River i Co., Brookhaven National Laboratory, Argonne National I.aboratory, and Science and Engineering Associates, Inc. The l enmmems of these organizations are rcSected in the Handbook. The draft version of the Handbook has also been used by
); NRC staff members since 1993 and staff enmments have been incorporatsd A draft version of the Handhook was made
'l 1 available to the public in September 1993 (58 FR 47160), but camments were not speci6cally requested.
! The Handbook is being issued in loose-leaf bimat to facilitate future revisions. NRC intends to periodicidly revise the i Handbook as new and improved guidance, data, r.nd methods become available. Cnmments on the Handimok from users i and the public are welcome at any time. Cnmments should be submitted to: Chief, Rules Review and Directives Branch, j Division of Freedom of Inbrmation and Publication Services, Mail Stop T-6 D59, U.S. Nuclear Regulatory Cn-minaion, Washington DC 20555-0001.
1 Thomas O. Martin, Chief l Regulation De&==' Branch i Division of Regulatory Applications 08 ice of Nuclear Regulatory Research t
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Acknowledgments Paci6c Northwest National laboratory (PNNL) provided technical and editorial support in preparation of this Handbook.
The principal PNNL technical contributors were R. H. Gallucci and P. L. Hendrickson. G. J. Konzek and P. J. Pelto of PNNL also contributed to the document. Helpful comments on an early draft of this Handbook were provided by W. S. Durant of Westinghouse Savannah River Co.; V. Mubayi of Brookhaven National laboratory; P. H. Kier, C. Mueller, S. Folga, J. Roglans-Ribas, E Monetde, and J. C. VanKuiken of Argonne National Iaboratory; F. Sciacca of Science and Engineering Associates, Ine; and a number ofinternal NRC reviewers.
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Abbreviations and Acronyms AC alternating current AE architect engineer AEC U.S. Atomic Energy Commission AEOD NRC Office for Analysis and Evaluation of Operational Data ANL Argonne National Laboratory ATWS anticipated transient without scram B&W Babcock & Wilcox BEIR biological effects of ionizing radiation BLS Bureau of Labor Statistics BLSV bulk liquids and scintillation vials BNL Brookhaven National Laboratory BWR Boiling Water Reactor CAP Clean Air Act Assessment Package CDF core damage frequency CE Combustion Engineering CFR Code of Federal Regulations CPCFB conditional probability of containment failure or bypass CRAC calculation of reactor accident consequences CRDM control rod drive mechanism
/7 CRGR Committee to Review Generic Requirements (d) cSv CVCS centisievert chemical and volume control system DRW dry radioactive waste DE dose equivalent DOE U.S. Department of Energy DUT U.S. Department of Transportation EA environmental assessment ECCS emergency core cooling system EDE effective dose equivalent EDO Executive Director for Operations EEDB energy economic data base EIS environmental impact statement EO Executive Order EPA U.S. Environmental Protection Agency EPRI Electric Power Research Institute FR Federal Register FSAR final safety analysis report FY fiscal year GDP gross domestic product GE General Electric ,
GEIS generic environmental impact statement Guidelines Regulatory Analysis Guidelines of the U.S. NRC GWe gigawatt electric HAF high aqueous feed HAW high activity waste i I kJ xix NUREG/BR-0184
Abbreviations and Acronyms HEP human ermr probability O
HEPA high efficiency particulate air HESAP human error sensitivity assessment of a PWR HFPP human factors program plan HLW high level w3ste HPCS high pressure core spray HVAC heating, ventilation, air conditioning -
ICRP International Commission on Radiological Protection IDCOR Industry Degraded Core Rulem*ing IEEE Institute of Electrical and Electronic Engineers i IPE individual plant exammation IPEEE individual plant examination of external events l IREP Interim Reliability Evaluation Program IRRAS Integrated Reliability and Risk Analysis System LAN low activity waste LCF latent cancer fatality LCS leakage control system LER licensee event report LHE latent health effect i
LOCA 1 loss of coolant accident '
LPCS low pressure core spray LQR licensed quantity released LWR light water reactor MACCS MELCOR Accident Consequence Code System MOV motor operated valve MOX mixed oxide fuel MRS monitored retrievable storage MT metric tons !'
MTHM metric tons of hazardous materials MTU metric tons of uranium MWe megawatt electric NCRP National Council on Radiation Protection and Measurements NEPA National Environmental Policy Act NHLW Non-HLW NMED !
Nuclear Material Ewnt Database NMSS Office of Nuclear Material Safety and Safeguards NPP nuclear power plant NPRDS Nuclear Plant Reliability Data System NRC U.S. Nuclear Regulatory Commission NRER non-reactor event report NRR Office of Nuclear Reactor Regulation OMB Office of Management and Budget PASNY Pbwer Authority of the State of New York PNNL Pacine Northwest National Laboratory PRA probabilistic risk assessment / analysis PSE Pmjekt Sicherkeitsstudien Entsorgung PV present value PWR pressurized w3ter reactor RCIC reactor core isolation cooling NUREG/BR-0184 xx
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Abbreviations and Acronyms i
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, N RECAP Rglacement Energy Cost Analysis Package
- REIRS Radiation Exposme Information and Reporting System
. RES Office of Nuclear Regulatory Research RHR residual heat removal RMIEP Risk Methods Integration and Evaluation Program e ROR Reduction-Oxidation Reactor RSS reactor safety study RSSMAP RSS Methodology Applications Program Rwo Regulatory Working aroup RWCU Reactor Water Cleanup -
- SARA system analysis and risk assessment
- - SGBD steam generator blowdown j SOTR steam generator tube rupture ,
SGIS standby gas treatment system l j SECY Staff Papers Before the Commission l St.CS standby liquid control system SRM Staff Requirements Memorandum i SRP Standani Review Plan
!- SST siting source term Staff NRC staff manbers TAP TMI Action Plan TASC The Analytic Sciences Corporation i TB 'Ibrbine Building
- THERP technique for human error raec prediction
! TMI Three Mile Island l TRU transuranic ,
l i URL uniform resource locator <
USI umesolved safety issue W Westinghouse
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l xxi NUREG/BR-0184
I V) 1 Introduction De past two decades have seen an increasing recognition that governmental actions need to account for their societal and economic impacts. As early as 1969, the National Environmental Policy Act required an assessment of environmental impacts of major federal actions including descriptions of alternatives and any unavoidable environmental insults. In December 1977, the U.S. Nuclear Regulatory Commission (NRC) established value-impact analysis guidelines (SECY-77-388A) to aid its decision-making. Executive Order 12291 was issued in February 1981 (46 FR 13193) requiring that executive agencies prepare regulatory impact analyses for all major rules and directing that regulatory actions be based on adequate information regarding the need for and consequences of proposed actions. Although the order was not binding on the NRC, the Commission decided to meet its spirit to enhance the effectiveness of NRC regulatory actions. Accordingly, in January 1983, the NRC issued Regulatory Analysis Guidelines (NUREG/BR-0058) for performing regulatory analyses for a broad range of NRC regulatory actions (NRC 1983c). These guidelines established a framework for 1) analyzing the need for and consequences of alternative regulatory actions,2) selecting a proposed alternative, and 3) documenting the analysis in an organized and understandable format. In December 1983, the NRC issued A Handbookfor Blue-Impact Assessment (NUREG/CR-3568 [Heaberlin et al.1983]) (hereafter called the "1983 Handbook"). Its basic purpose was to set out systematic pmcedures for performing value-impact assessments. Revision 1 to NUREG/BR4)058 (NRC 1984b) was issued in May 1984 to include appropriate references to the 1983 Handbook.
In 1995, NRC's guidance on preparing regulatory analyses was updated in Revision 2 to NUREG/BR-0058 (NRC 1995a),
hereafter referred to as the "NRC Guidelines" or simply the " Guidelines." Revision 2 was issued to reflect the NRC's experience implementing Revision 1 of the Guidelines; changes in NRC regulations since 1984, especially the backfit rule t 10 CFR 50.109) and the Commission's 1986 Policy Statement on Safety Goals for the Operation of Nuclear Power Plants l y} (58 FR 51735; October 4,1993); and procedural changes des (NRC 1986); advances and refinements in regulatory analysis techniques; regulatory guidance in Executive Order 12866 )
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This revision to NUREG/CR-3568 (hereafter called the " Handbook") has been prepared to accomplish several objectives.
First, the expanded guidance included in Revision 2 of the NRC Guidelines has.been incorporated. Second, the scope of l the Handbook has been increased to include the entire regulatory analysis process (not only value-impact analyses) and to l address not only power reactor, but also non-reactor applications.* Third, NRC experience and improvements in data and methodology since the 1983 Handbook have been incorporated. Fourth, an attempt has been made to make the Hand- '
i book more " user friendly." Fifth, the Handbook incorporates guidance included in the document Economic Analysis of hderal Regulations Under Executive Onfer 12866 (Regulatory Working Group 1996). This document, which superseded !
the Office of Management and Budget's (OMB's) " Regulatory Impact Analysis Guidance" (reference 6 in the NRC Guidelines), was prepared by a federal interagency regulatory working group.
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This Handbook has been designed to assist the analyst in preparing effective regulatory analyses and to provide for consis- )
tency among them. The guidance provided is consistent with NRC policy and, if followed, will result in an acceptable document. It must be recognized, however, that all conceivable possibilities cannot be anticipated. Therefore, the Hand-book guidance is intended to allow ficxibility in interpretation for special circumstances. It must also be recognized that regulatery analysis methods continue to evolve, along with the applicable data. The NRC and other federal agencies (e.g., l OMB, the U.S. Environmental Protection Agency [ EPA], and the U.S. Department of Transportation [ DOT]) continue to undertake research and development to improve the regulatory decision-making process.
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1.1 NUREGiBR-0184 i
1 Introduction
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1.1 Purpose O'
The purpose of this Handbook is to provide guidance to the regulatory analyst to promote preparation of high-quality regu-latory decision-making documents and to implement the policies of the NRC Guidelines. In fulfilling this purpose, there are several objectives of the Handbook.
First, the Handbook expands upon policy concepts included in the NRC Guidelines. The steps in preparing regulatory !
analyses are translated into implementable methodologies for the analyst. An attempt is made to provide the rationale behind current NRC policy to assist the analyst in understanding what the decision-maker will likely need in the regulatory analysis. Second, the Handbook has been expanded to address the entire regulatory analysis process, i.e., all six steps (see Handbook Section 1.2.2) identified in the NRC Guidelines. The 1983 Handbook only addressed value-impact analysis, just one element of a regulatory analysis. Also, unlike the 1983 Handbook, this Handbook addresses not only power reactor but also non-reactor applications.
Third, the Handbook has been updated to incorporate changes in policy and advances in methodology that have occurred l since the 1983 Handbook was issued. Considerable research has been conducted by the NRC and other agencies on various aspects of regulatory decision-making. Also, NRC staff experience has resulted in significant modifications to the l
regulatory analysis process. Advances resulting from the above have been appropriately incorporated in this Handbook. <
Fourth, the Handbook has consolidated relevant information regarding regulatory analyses. As mentioned above, many activities have improved the ability to make better decisions. The resulting information has been used in the preparation of this Handbook. Where the information is not presented explicitly, references lead the analyst to the appropriate documents.
Fifth, the Handbook provides standardized methods of preparation and presentation of regulatory analyses, including back- i fit and Committee to Review Generic Requirements (CRGR) regulatory analyses. Consistent application of the methods pmvided here will result in more directly comparable analyses, thus aiding decision-makers in evaluating and comparing various regulatory actions.
The Handbook cites numerous references throughout, often extracting information fmm them directly. Where practical, the bases for extracted information have been summarized from the references. However, this does not imply that the analyst should use the information exclusively without consulting the references themselves. Where supplied data seem to contradict the anrJyst's " common sense," examination of the references may be crucial.
1.2 Regulatory Analysis Overview 1
The following sections provide an overview of a regulatory analysis. Section 1.2.1 discusses key terms and concepts in a regulatory analysis. Section 1.2.2 discusses the appropriate steps.
1.2.1 Key 'Ibrms and Concepts Bacifitting. Backfitting is defined at 10 CFR 50.109(a)(1) as "the modification of or addition to systems, structures, com-ponents, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organi-zation required to design, construct or operate a facility; any of which may result from a new or amended provision in the O
NUREG/BR-0184 1.2
Introduction r~N Commission rules or the imposition of a regulatory staff position interpreting the Commission rules that is either new or different from a previously applicable staff position... ." Backfitting requirements apply only to production and utilization facilities as those terms are defined at 10 CFR 50.2.
Backfit Regulatory Analysis. A backfit argulatory analysis is a regulatory analysis prepared for a generic backfit. A back-fit regulatory analysis is prepared to meet the requirements of 10 CFR 50.109(c) and the NRC Guidelines.m CRGR Regulatory Analysis. A Committee to Review Generic Requirements (CRGR) regulatory analysis is a regulatory analysis that satisfies the requirements of the CRGR Charter and the NRC Guidelines. CRGR regulatory analyses are pre-pared for proposed actions within the CRGR scope as set out in Chapter !!I of the CRGR Charter. In general, the scope covers new or amended generic requirements and staff positions to be imposed on one or more classes of power reactors.
Plant-Specific Bac4 fit. A plant-specific backfit is a backfit applicable to a single facility. Backfits of this type are subject j to the requirements of NRC Management Directive 8.4 (NRC Manual Chapter 0514). !
Regulatory Analysis. A regulatory analysis is a stmetured evaluation of all relevarit factors associated with the makmg of a regulatory decision. As used by the NRC, a regulatory analysis consists of the six steps described in Handbook Section 1.2.2 and NRC Guidelines Chapter 4.
Safety Goal Evaluation. An evaluation prepared to determine whether a proposed generic safety enhancement backfit for o nuclear power plants meets the safety goal screening criteria in the Commission's safety goal policy statement (see
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%lue-Impact (Benefit-Cost) Analysis. A value-impact analysis is a balancing of the benefits (values) and costs (impacts) associated with a proposed action or decision. Values and impacts should be evaluated in monetary terms when feasible, resorting to qualitative terms where conversion to monetary equivalents cannot be done. A value-impact analysis is a substantial part of a regulatory analysis.
1.2.2 Steps in a Regulatory Analysis Chapter 4 of the NRC Guidelines pmvides n; )ix steps in a complete regulatory analysis, corresponding with the six elements to be included in a regulatory analysis. The first step is identifying the problem and establishing the analysis objective. The nature of the problem and its history, boundaries, and interfaces must be clearly established. He objective is the conceptual improvement sought by the proposed regulatory action. It is typically a qualitative statement establishing a basis for judging the results of the subsequent analysis elements.
The second step is identifying alternative approaches to the problem and doing a preliminary analysis of these approaches.
Development of a reasonably broad and comprehensive set of alternatives is required to ensure identi6 cation of all significant approaches. The initial set of alternatives is reduced by eliminating ones based on obvious feasibility, value, and impact considerations. Alternatives that cannot be clearly eliminated will be subjected to the next step (value-impact analysis).
The third step is estimating and evaluating values and impacts. Step 3 also includes preparation of a safety goal evaluation if the alternatives involve a proposed generic safety enhancement backfit to nuclear power reactors which is subject to the substantial additional protection standard at 10 CFR 50.109(a)(3). Safety goal evaluations are discussed in Chapter 3.
There are many factors that complicate this step (e.g., imperfect knowledge, many possible evaluation methods, and 1.3 NUREG/BR-0184
Introduction values and impacts that are difficult to quantify). Despite the difficulties, a best effon must be made to characterize the O
factors peninent to a decision. Even if values and impacts cannot be sufficiently characterized, use of consistent methods, data, and presentation can form an adequate basis on which to prioritize alternative regulatory actions. Much of this liandbook addresses this step.
He founh step is presenting results A tabular presentation is typically optimal, with the results displayed to facilitate comparison of the evaluated alternatives. Values and impacts not quantified in monetary terms also need to be presented.
The goal is to clearly convey the complex value-impact resn'ts to the decision-maker. It is also important to reveal the uncenainties associated with the results so that the decision-maker can assess the confidence associated with them. In this llandbook, step:: three and four are together referred to as value-impact analysis.
The fifth step is preparing the decision rationale for selecting the proposed action. In this step the analyst recommends and justifies an action based on the previous analyses. Any decision criteria used in the selection are identified.
The sir.!h and final step is developing a schedule for the activities that will be required to implement the proposed actions.
Implementadon activities could include such things as needed analyses, approvals, procurement, installation and testing, procedure development, training, and reporting. He schedule should be realistic and can include alternative schedules if apprtpnate.
1.3 Handbook Overview Chapter 1 provides introductory and conceptual information regarding the performance of a regulatory analysis and some historical perspective. The relationship of this Handbook with the NRC Guidelines and other NRC policy is established.
Chapter 2 explains the scope of regulatory analyses and the appropriate level of detail to be used.
Chapter 3 discusses the safety goal evaluation required by Chapter 3 of the NRC Guidelines for generic safety enhance-ment backfits to nuclear power reactors when the proposed backfit is subject to the substantial additional protection standard at 10 CFR 50.109(a)(3).
Chapter 4 presents the methodology to oe used in performance of a regulatory analysis.
Chapter 5 presents detailed guidance on the performance of the va1ue-impact analysis ponion of a regulatory analysis for both power reactor and non-reactor facilities.
Chapter 6 lists all Handbook references.
Appendix A discusses topics of panicular importance in agulatory ana;yses that are not covered specifically in other areas of the Handbook, especially human factors issues.
Appendix B contains supplementary information for the value-impact portion of a regulatory analysis.
Appendix C presents supplemental information on regulatory analyses for ncn-reactor facilities.
1.4 O
NUREG/BR4)184
. . . . - . - - . - . - . . . . . ~ . - . . _ . - . . . - - . . . _ . . _ _
4 Introduction t
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Appendix D reproduces the Safety Goals for the Operations of Nuclear Pbwer Plants Policy Statement and the Backfit Rule.
3-j Appendix E is an index to the Handbook.
{ 1.4 Endnotes for Chapter 1 i
) 1. De variety of non-reactor facility types and the relatively non-integrated sets of available information add difficulty
- to the preparation of regulatory analyses for non-reactor facilities. Appendix C repsesents an attempt to coordmate available information to provide guidance for conducting a non-reactor regulatory analysis, especially the value- ,
impact analysis segment, ne nature of regulatory analyses for non reactor facilities will continue to evolve as more i analyses are performed and more information becomes available.
2.
- As discussed in Section 2.2 of the Handbook, some backfit regulatory analyses fall within the scope of the CRGR Chaner, and therefore, are subject to the requirements for CRGR regulatory analyses as well. Commission approval j of Revision 6 to the CRGR Charter was announced in SECY 96 032 issued in March 1996, l
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v) 2 Scope of a Regulatory Analysis Most NRC regulatory actions require some form of analysis and supporting documentation, the exact nature of which is determined by the type of action. 'Ihis chapter discusses the scope of the particular type of analysis termed a " regulatory analysis," defined in Section 1.2.1.
2.1 When a Regulatory Analysis is Required Section 2.2 of the NRC Guidelines states that, in general, all mechanisms proposed to be used by the NRC to establish or communicate generic sequirements, guidance, requests, or staff positions that would affect a change in 'he use of resources by NRC licensees, include an accompanying regulatory analysis. Specific criteria for determining whether a regulatory analysis will need to be performed are also presented in Section 2.2 of the NRC Guidelines.
Section 2.1 of the NRC Guidelines makes it clear that a regulatory analysis is an integral part of NRC decision-making It is necessary, therefore, that the argulatory process begin as soon as it becomes apparent that some type of regulatory action by the NRC to address an identified problem may be needed.
Many regulatory analyses will fall into the classifications of backfit regulatory analyses and/or CRGR regulatory analyses.
Table 2.1 summarizes important characteristics of these two classifications of regulatory analyses. Additional information is provided in Sections 2.2 and 2.3 of this Handbook.
k An additional consideration impacts regulatory analyses involving generic safety enhancement backfits to nuclear power plants that are subject to the substantial additional protection standard at 10 CFR 50.109(a)(3). As discussed in Chapter 3 of the Guidelines, a safety goal evaluation is needed for these regulatory analyses. The result of this evaluation determines 1 the extent to which further development of the regulatory analysis is appmpriate.
2.2 When a Backfit Regulatory Analysis is Required The term "backfitting" is defined at 10 CFR 50.109(a)(1). Backfitting only applies to facilities licensed under 10 CFR Part 50. Such facilities are called production facilities or utilization facilities (these terms are defined at 10 CFR 50.2). A nuclear power plant is a utilization facility. For a detailed discussion of concepts related to backfitting, the reader is referred to the Backfirting Guidelines, NUREG-1409 (NRC 1990a). The guidrace provided in this Handbook applies to ;
generic backfits (defined in Section 1.2.1) and, in certain instances, plant-specific backfits as well (al.o defined in Section i 1.2.1). NRC Management Directive 8.4 should be consulted for requirements related to plant-specific backfits. I Ordinarily, any proposed action 6tting the definition of a backfit will require the preparation of a backfit regulatory analy-sis. The only instances where a backfit regulatory analysis will not be required for a proposed backfit are the three excep-tions identified at 10 CFR 50.109(a)(4). These exceptions are determinations by the Commission or NRC staff, as appropriate, that:
- a modification is necessary to bring a facility into compliance with a license or the rules or orders of the Commission, or into conformance with written commitments by the licensee; or
- tegulatory action is necessary to ensure that the facility provides adequate protection to the health and safety of the
[s\ public and is in accord with the common defense and security; or 2.1 NUREG/BR-0184 l
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Scope O
'Ihble 2.1 Applications of back8t and CRGR regulatory analyses Characteristic Back8t Regulatory Analyses CRGR Regulatory Analyses Facilities Production and utilization facill- Nuclear power plants; ties (e.g., nuclear power plants). Materials licensees (to the extent directed by the Executive Director of Operations [EDO] or the Director of the Office of Nuclear Macrial Safety and Safeguards [NMSS]).
Type of Action New or amended rule or staff New or amended generic position covering modification of requirements and staff posi-or additions to systems, struc- tions to be imposed on one or tures, components, or design of a more classes of power reac-facility or the procedures or tors or materials licensees, organization required to design, including reductions in exist-constmet, or operate a facility ing requirements.
[with the three exceptions described at 10 CFR 50.109(a)(4)].
Type of Backfit Covered Backfits where there are substan- All backfits meeting other tialincreases in the overall pro- CRGR criteria, including tection of the public health and backfits considered necessary safety or the common defense to ensure adequate protection and security and the implementa- to public health and safety.
tion costs are justified in view of the increased protection.
- the regulatory action involves defming or redefining what level of protection to the public health and safety or common defense and security should be regarded as adequate.
When one of these exceptions is relied upon for not performing a backfit regulatory analysis, a written evaluation meeting the requirements of 10 CFR 50.109(a)(6) and Section IV.B(ix) of the CRGR Charter (for proposed actions within the scope of the CRGR) must be prepared. Also, costs are not to be considered m justifying the proposed action.
I A backfit regulatory analysis is similar to, and should generally follow the requirements for, a regulatory analysis.m There are certain ret uirements specific to a backfit regulatory analysis that are identified at 10 CFR 50.109(a)(3) and 10 ,
CFR 50.109(c). 'Itese requirements are identified in Table 2.2 and at appropriate parts of the Handbook. 'Ihble 2.2 also l cites where in the CFR the requirement is located and indicates where in the regulatory analysis the discussion of each NUREG/BR-0184 2.2 9 l
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'Ihble 2.2 Check!!st for speci8c back8t regulatory analysis requirements Section of the Regulatory
. CFR Citation Information Item to be included Analysis Where Item Should
('I1tle 10) la a Back8t Regulatory Analysis Normally be Discussed 50.109(a)(3) Basis and a determination that there is Basis - Presentation of Results a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived Determination - Decision Rationale from the backfit and that the direct and indirect costs of implementation for the affected facilities are justified in view of this increased protection.
50.109(c)(1) Statement of the specific objectives that Statement of the Problem the proposed backfit is designed to achieve. and Objectives 50.109(c)(2) General description of the activities that identificationof Alternatives would be required by the licensee or applicant to complete the backfit.
( ./ 50.109(c)(3) Potential enange in the risk to the public Estimation and Evaluation of from the accidental offsite release of Values and Impacts radioactive material.
50.109(c)(4) Potential impact on radiological exposure of Estimation and Evaluation of )
facility employees. Values and Impacts l
50.109(c)(5) Installation and continuing cost associated Estimation and Evaluation of with the proposed backfit, including the cost Values and Impacts of facility downtime or construction delay.
50.109(c)(6) Potential safety impact of changes in plant Estimation and Evaluation of or operational complexity, including the Values and Impacts relationship to proposed and existing regulatory requirements.
50.109(c)(7) Estimated resource burden on the NRC Burtlen - Estimation and Evaluation of associated with the proposed backfit and the Values and Impacts estimated availability of such resources.
Availability -Implementation 50.109(c)(8) Pbtential impact of differences in facility Presentation of Results type, design, or age on the relevancy and practicality of the proposed backfit, implementation b
2.3 NUREG/BR-0184
l Scope Thble 2.2 (Continued) ell l
l l-Section of the Regulatory ,
CFR Citation Information Item to be Included Analtsis Where Item Should (Title 10) in a Backfit Regulatory Analysis Norinally be Discussed 50.109(c)(9) Whether the proposed backfit is interim or Decision Rationale final and, if interim, the justification for ;
imposing the proposed backfit on an interim basis.
1 50.109(c) Consideration of how the backfit should be Implementation scheduled in light of other ongoing
)
regulatory activities at the facility.
item should nornally appear. The analyst must be sure to integrate the 10 CFR 50.109 requirements into the backfit regulatory analysis. Section 2.3 of the Guidelines requires that the findings required by 10 CFR 50.109 are to be ,
highlighted in a backfit regulatory analysis. The recommended method of highlighting backfit rule finding is a vertical line in the left margin adjacent to the text to be highlighted.
If the proposed backfit falls within the scope of the CRGR (as set out in Section III of the CRGR Charter), the information requirements identified in Section IV.B of the Charter and Section 2.3 of this Handbook should be incorporated into the '
backfit regulatory analysis. (inclusion of these items will, in effect, render the backfit regulatory analysis a CRGR regulatory analysis). A proposed backfit involving a new or amended generic requirement or staff position to be imposed on one or more classes of nuclear power reactor licensees or materials licensees. (to the extent directed by the EDO or the Director of NMSS) will ordinarily require CRGR review.
2.3 When a CRGR Regulatory Analysis is Required The CRGR has the responsibility to review and recommend to the EDO approval or disapproval of requirements or NRC staff positions to be imposed on one or more classes of power reactors and, in some cases, on nuclear materials licensees.
The review applies to requirements or positions which rc.3uce existing requirements or positions and proposals which increase or change requirements. The CRGR's purpose, membership, scope, operating pmcedures, and reporting require-ments are set out in the CRGR Charter. The most recent version of the Charter is Revision 6, issued in 1996 (NRC 1996c).
Section IV.B of the Charter lists the information that is required to be submitted to the CRGR for review of proposed actions within its scope. One item (identified in Section IV.B(v) of the Charter) is a segulatory analysis conforming to the direction in the NRC Guidelines and this Handbook.m There are other requirements included in Section IV.B as shown in Table 2.3. Table 2.3 includes the citation to the portion of the CRGR Charter where the requirement is found and also indicates where in the regulatory analysis the discussion of each item should normally appear. The analyst should
, generally ensure that each item in Table 2.3 is included in a regulatory analysis prepared for CRGR review. The items included in Table 2.3 are identified and discussed at appropriate parts of this Handbook. Section 2.3 of the Guidelines NUREG/BR-0184 2.4 O
Scope O
'Ihble 2.3 Checklist for specifle CRGR regulatory analysis requirements Section of the Regulatory CRGR Charter Infonnation Item to be Included in a Regulatory Analysis Where Iten Should Citation Analysis Prepared for CRGR Review Normally be Discussed IV.B(1) The proposed generic requirement or staff Implementation position as it is proposed to be sent out to licensees.
When the objective or intended result of a Identification of proposed generic requirement or staff position Alternatives can be achieved by setting a readily ~
quantifiable standard that has an unambiguous relationship to a readily measurable quantity and is enforceable, the proposed requirement ;
should specify the objective or result to be attained rather than prescribing how the objective or result is to be attained.
IV.B(iii) The sponsoring office's position on whether Presentation of Results (v" ) the proposed action would increase requirements or staff positions, implement existing )
requirements or start positions, or relax or ,
reduce existing requirements or staff positions. ;
1 IV.B(iv) The proposed method of implementation.0) Implementation '
l IV.B(vi) Identification of the category of power reactors Identification of or nuclear materials facilities / activities Alternatives ,
to which the generic requirement or staff ,
position will apply. j IV.B(vil) If the proposed action involves a power reactor See Table 2.2 backfit and the exceptions at 10 CFR 50.109(a)(4)
IV.B(viii) are not applicable, the items identified at 10 CFR 50.109(c) and the required rationale at 10 CFR 50.109(a)(3) are to be included (these items are included in Table 2.2)*
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Thble 2.3 (Continued)
Section of the Regulatory CRGR Charter Information Item to be Included in a Regulatory Analysis Where item Should Citation Analysis Prepared for CRGR Review Normally be Discussed IV.B(x) For proposed relaxations or decreases in Decision Rationale current requirements or staff positions, a rationale is to be included for the deter-mination that (a) the public health and safety and the common defense and security would be adequately protected if the proposed reduction in requirements or positions were implemented, and (b) the cost savings attributed to the action would be substantial enough to justify taking the action.0)
IV.B(xii) Preparation of an assessment of how the Estimation and Evaluation proposed action relates to the Commission's of Values and Impacts Safety Goal Policy Statement (see NRC Guidelines Chapter 3 and Handbook Chapter 3).
P requires that the findings required by the CRGR Chaner are to be highlighted in a CRGR regulatory analysis. De recommended method of highlighting CRGR Chaner fmdings is a vertical line in the right margin adjacent to the text to be highlighted.
2.4 Level of Detail An overview of NRC policy regarding the level of detail to be provided in regulatory analyses is provided in Chapter 4 of j the NRC Guidelines. The emphasis in implementation of the NRC Guidelines should be on simplicity, fletibility, and commonsense, both in terms of the type of information supplied and in the level of detail provided. De level of treatment given to a pardcular issue in a regulatory analysis should reflect how crucial that issue is to the bottom line recom-mendation of the regulatorj analysis. In all cases, regulatory analyses are to be sufficiently clear and detailed for use by NRC decision-makers and other interested parties.
With respect to the appropriate level of detail, the analyst must first determine the level of effort to be expended in analyz-ing the problem. A greater expenditure of effort will result in a greater expenditure of NRC resources, and vice versa.
The expenditure of resources to analyze a regulatory action is to be correlated with the safety and cost impacts of the action. Chapter 4 of the Guidelines lists factors that should be considered to determine the appropriate level of detail.
This Handbook presents direct guidance for performing what is termed a " standard' analysis. His is expected to encom-pass one to two person-months, a level of effon believed sufficient for many regulatory analyses. The Guidelines and this NUREG/BR-0184 2.6
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V) llandbook, including references suggested by this Handbook, should be sufficient for performing the analysis. Where larger levels of effon may be involved, this Handbook suggests additional methods and references which can be used.
These could entail major effons, possibly on the order of a person-year.
A decision tree has been developed to assist the analyst in determining the appmpriate level of effort to be applied in a par-ticular case (see Figure 2.1), if the NRC action will result in a regulatory burden on licensees, a regulatory analysis will typically be required. The level of effon will depend on the complexity of the issue. A complex issue would clearly jus-tify a major effort based on the significant impacts of the regulatory decision. If NRC management specifically direct that a major effort be undertaken, the decision is clear. If the issue is not complex, the standard analysis should suffice. The level of detail to be included in the reguictory analysis document can generally be expected to follow the level of effort expended in performing the analysis. The Guidelines establish the minimum requirements. In determining the appropriate level of detail, the best guidance is that the analyst view the presentation objectively from the point of view of the decision-maker.
In cases where there is uncertainty as to the correct level of detail, it is probably better to err on the side of providing too much information. A decision-maker can always filter out unnecessary information, but may have considerable difficulty filling in the blanks. Tables and figures should be used to the maximum extent possible to convey information, panicularly where the amount of information is substantial or where comparisons are involved.
2.5 Units O
() Regulatory 1996). Regulatoryanalyses should analyses affecting be one more than prepared consistently licensee should be prepared in dualwith NRC's (i.e., metric finalunits.
and English) metrication 1 Metric units should be shown first with the value in English units shown in parenthesis. Regulatory analyses affecting a i single licensee should use the system of units employed by the licensee.
2.6 Regulatory Relaxations NRC's position on regulatory analysis requirements for relaxation of regulatory requirements is in Section 2.2 of the Guidelines. Preparation of a regulatory analysis for a proposed relaxation is generally required. However, the backfit rule requirements in 10 CFR 50.109 and the safety goal evaluation process set out in Chapter 3 of the Guidelines are not applicable to proposed relaxations.
For all regulatory analyses of proposed relaxations, information should be presented in the decision rationale section (see Section 4.4) indicating whether:
- 1. The public health and safety and the common defense and security would continue to be adequately protected if the proposed reduction in requirements or positions were implemented.
- 2. The cost savings attributed to the action would be substantial enough to justify taking the action.
- 3. The proposed relt.2ation is optional or mandatory for affected licensees.
l l Inclusion of the three preceding items will satisfy the requirements in Section IV.B(x) of the CRGR Chaner.
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2.7 NUREG/BR-0184
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1 Yes (See Belov0 No 1r 1r i 2 Yes Major (See Below) 5 Enort No 1P Standard Effort
- 1. Has the Commission, EDO, or Office Director requested a major effort?
- 2. Are any of the following likely to occur:
- an annual effect on the economy of $100 million or more e
a major increase in costs or prices for consumers; individual industries; federal, state, or local government agencies or j geographic regions '
significant adverse effects on competition, empicyment, investment, productivity, innovation, or on the ability of I U.S.-based enterprises to compete with foreign-based enterprises in domestic or export markets e roughly comparable values and impacts e potential for considerable controversy, complexity, or policy significance?
Hgure 2.1 Decision tree to determine Im! cf d'ert O
NUREG/BR-0184 2.8
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2.7 Endnotes for Chapter 2
- 1. NRC's Final Policy Statement on the use of probabilistic risk assessment (PRA) in nuclear regulatory activities (NRC 1995b) includes the statement that where appropriate. PRA should be used to support a proposal for additional regulatory requirements in accordance with 10 CFR 50.109 (see Section 5.6).
- 2.Section IV.B(iv) of the CRGR Charter states that a regulatory analysis is not required for backfits within the scope of 10 CFR 50.109(a)(4).
- 3.Section IV.B(iv) of the CRGR Charter also requires the concurrence of the NRC Office of the General Counsel (and any comments) and the concurrence of affected program offices or an explanation of their non-concurre we in the proposed method of implementation. These concurrences and related information can be included in the transmittal memorandum to the CRGR and need not be included in the CRGR regulatory analysis.
- 4.Section IV.B(viii) of the CRGR Charter also requires, in the case of power reactor backfits, a determmation by the proposing office director that there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from the backfit and that the direct and indirect costs of imple-mentation for that facility are justified in view of this increased protection. A statement of this determination may be included in the transmittal memorandum to the CRGR rather than in the CRGR regulatory analysis. Guidance on application of the " substantial increase" standard is in Attachment 3 to the CRGR Charter.
- 5.Section IV.B(x) of the CRGR Charter requires the proposing office director to determine that conditions (a) and (b) are met for the proposed action. A statement of this determination may be included in the transmittal memorandum y/ to the CRGR rather than in the CRGR regulatory analysis.
2.9 NUREG/BR-0184
3 Safety Goal Evaluation for Operation of Nuclear Pbwer Plants The Commission has directed that NRC's regulatory actions affecting nuclear power plants be evaluated for conformity with NRC's Policy Statement on Safety Goals for the Operations of Nuclear Power Plants (NRC 1990b). The Safety Goal Policy Statement is reproduced in Appendix D. The Policy Statement sets out two qualitative safety goals and two quantitative objectives. Both the goals and objectives apply only to the risks to the public from the accidental or routine release of radioactive materials from nuclear power plants.
The qualitative safety gorls in the Policy Statement are e individual members of the public should be provided a level of protection from the consequences of nuclear power plant operation such that individuals bear no significant additional risk to life and health
- societal risks to life and health from nuclear power plant operation should be comparable to or less than the risks of generating electricity by viable competing technologies and should not be a significant addition to other societal risks.
The two quantitative objectives in the Policy Statement are to be used in determtmng achievement of the qualitative safety goals. The objectives are ;
i e the risk to an average individual in the vicinity of a nuclear power plant of prompt fatalities that might result from l reactor accidents should not exceed 0.1% of the sum of prompt fatality risks resulting from other accidents to which I memben of the U.S. population are generally exposed l e the risk to the population in the area near a nuclear power plant of cancer fatalities that might result from nuclear power plant operation should not exceed 0.1% of the sum of cancer fatality risks resulting from all other causes.
Chapter 3 of the NRC Guidelines contains specific information implementing the quantitative objectives which the analyst should carefully follow. ;
Section 3.1 of the Guidelines states that a safety goal evaluation is needed for a proposed generic safety enhancement backfit to nuclear power plants which is subject to the substantial additional protection standard at 10 CFR 50.109(a)(3).
Thus, proposals for a plant-specific backfit or for generic backfits within the exceptions at 10 CFR 50.109(a)(4)(i-lii) do not require a safety goal evaluation. Section 3.1 of the Guidelines also states that a safety goal evaluation is not needed for a proposed relaxation of a requirement affecting nuclear power plants.
Section 3.2 of the Guidelines states that a probabilistic risk assessment (PRA) should normally be used in performing a safety goal evaluation to quantify the risk reduction and corresponding values of a proposed new requirement.m NRC's Final Policy Statement on the use of PRA methods in nuclear regulatory activities (NRC 1995b) contains the following statement:
The Comtr>2sion's safety goals for nuclear power plants and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgments on the need for proposing and backfitting new generic requirements on nuclear power plant licensees.
Table 5.2 in this Handbook contains a list of PRAs and their characteristics which can potentially be used in performing safety goal evaluations. Additional sources of PRAs are Individual Plant Exammation (IPE) and Individual Plant Exam-ination of External Events (IPEEE) reports submitted to the NRC by nuclear power plant licensees (see Section 5.6.1).m x
3.1 NUREG/BR4184
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Section 3.3.1 of the Guidelines provides an illustration of when an IPE report can be used in a safety Boal evaluation. The example is that if a proposed backfit will only affect older boiling water reactors (BWRs), one or more IPEs conducted for older BWRs should be utilized in the evaluation. IPE and IPEEE reports are available through the NRC public document room (telephone: 202-634-3273 or 800-397-4209). A draft NUREG report was issued in late 1996 covering 1) insights gained from staff review of IPE reports, and 2) NRC's overall conclusions and observations including comparisons of IPE results with the Commission's safety goals (NRC 1996b). This report also contains a discussion of acceptable attributes of a quality PRA.
If conducted, a safety goal evaluation should be included in Section 3 of the regulatory analysis document which covers
' estimation and evaluation of values and impacts." The results of the safety goal evaluation should be included in Sec-tion 4 of the regulatory analysis document which covers ' presentation of results."
It is planned that additional supplementary material will be added to Chapter 3 of this Handbook in the future after more safety goal evaluation experience is gained.
As this version of the Handbook was being completed, a number of NRC staff activities were underway which relate to PRA use in safety goal evaluations and other NRC regulatory activities. These include o completion of the staff's review of licensee-submitted IPEs o
evaluation of these IPEs for potetial use in other regulatory activities, documented in a draft report to be published as NUREG-1560 (NRC 1996b) e development of guidance on the use of PRA in plant-specific requests for license changes, including regulatory guides for use by licensees in preparing applications for changes and standard review plans for use by the NRC staff in reviewing proposed changes.
These activities should result in a more consistent and technically justified application of PRA in NRC's regulatory process. This work, along with staff work planned for fiscal year (FY) 1997 to initiate improvements to the economic models now used in NRC's effsite consequence analyses (e.g., in NRC's MELCOR Accident Consequence Code System
[MACCS] code), should have a significant impact on the PRA-related ponions of this Handbook. Consequently, the discussion in this Handbook on the use of PRA and offsite consequence estimates should be viewed as interim guidance that may be relied upon until the Handbook is updated to acconunodate the NRC's new position on these regulatory issues.
The staff expect to initiate this update as the preceding PRA guidance nears completion.
3.1 Endnotes for Chapter 3
- 1. SECY-95-079 contains a status update of NRC's PRA implementation plan. SECY-95-280 contains a framework for applying PRA in reactor regulation.
- 2. SECY-96-051 (NRC 1996a) contains the following statement:
Licensees were not requested to calculate offsite health effects in Generic Letter 88-20 and, therefore, most of the IPE results cannot be urged directly to compare with the quantitative health objectives of the Commission's Safety Goals (i.e., early and latent cancer fatalities). However, all licensees did estimate two related risk mer.sures:
containment failure frequencies and radionuclide release frequencies. These results can be examined in light of other studies of similar scope where explicit comparisons of plant risks with safety goals were performed, specifically NUREG/BR-0184 3.2
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4 Regulatory Analysis Methods and Supporting Information A regulatory analysis consists of six elements:
, 1. Statement of the problem and objective.
- 2. IdentiScation and preliminary analysis of alternative approaches.
- 3. Estimation and evaluation of values and impacts (incorporating a safety goal evaluation in appropriate cases).
[ 4. Presentation of results.
- 5. Decision rationale.
d
- 6. Implementation.
Each of these elements is very briefly summanzed in Section 1.2.2 of this Handbook, and addressed in detail in the six major sections (4.1 through 4.6) in this chapter. He conceptual requirements associated with the segulatory analysis
- elements are also described. The safety goal evaluation process is discussed in Chapter 3.
l 'Ib promote consistency, standard format and content guidance for regulatory analysis documents have been developed as
~
shown in Figure 4.1. The six major sections of the regulatory analysis document are mandatory, as well as the basic information indicated for each. Subsections under each section may be included at the discretion of the analyst.
Additional information not indicated in Figure 4.1 may be included as appropriate. De guidance pmvided is intended to j allow the analyst the maximum amount of Aexibility within the constraint of ensunng reasonable consistency among i l
regt.latory analysis documents.
4.1 Statement of the Problem and Objective i
This element allows the analyst to carefully establish the character of the problem, its background, boundaries, l s
significance, and what is hoped to be achieved (the objective).
The character of the problem consists of several factors. A concise description of the problem or concern needs to be !
developed included in the description is 1) the basis for the decision that a problem exists (e.g., a series of equipment 4
failures during operation or a major incident that seveals an inherent design weakness), and 2) the fnnA= mental nature of
- the problem (e.g., inadequate design, inadequate inspection or maintenance, operator failure, failure to incorporate ade-quate human factors). Care should be taken to neither define the problem too broadly (making it difficult to target a regu-latory action) nor too narmwly (risking non-solution of the problem when the regulatory action is implemented). A l background discussion of the problem should be provided, including relevant items from Section 4.1 of the Guidelines.
If appropriate, a statement of why 1) market forces cannot alleviate the problem [see Section I.A of RWG (1996) for a dis-cussion of the role market forces play in regulatory decision-making], and 2) the NRC, as opposed to other organizations
]
(e.g., licensees, vendors, owners groups or state agencies), is considering action should be included. The scope of the problem should be discussed in terms of the classes oflicensees or facilities being affected, including their numbers, sizes,
< etc. Any distinction between NRC and Agreement State") licensees should be made. The implications of takmg no action (i.e., maintaining the status quo) should be identi6ed.
I l
4.1 NUREG/BR-0184
}
i
l Methods l l
1 Thble of Contents O
l Executive Summary '
1 Statement of the Problem Describe the nature of the problem, any relevant history, the boundaries and Objective of the problem, interfaces with other NRC activities, and a clear statement of the objective of the proposed action (see Section 4.1).
2 Identification and Preliminary Identify alternative approaches considered and those approaches Analysis of Alternative eliminated due to obvious reasons, provide the basis for eliminating Approaches to the Prvblem alternatives, clearly explain alternatives to be considered, and determine the level of effort to be applied (see Section 4.2).
3 Estimation and Evaluation If appropriate, evaluate compliance with the Safety Goals guidance (see Chapter 3 of Values and Impacts of the Guidelines and Handbook). Summarize methods used and results for all alternatives evaluated in the value-impact analysis (see Section 4.3).
4 Presentation of Results Present results for alternatives evaluated, including discussion of supplemental con-siderations, uncertainties in estimates, and results of sensitivity analyses (see Section 4.4). Present results of safety goal evaluation if conducted.
5 Decision Rationale Present the pref.rred altemative and the basis for selection, discuss any decision criteria used, identify and discuss the regulatory instrument to be used, and explain the statutory basis for the action (see Section 4.5).
6 Implementation Present implementation milestones and associated schedule; discuss the relation-ships of the proposed action to other ongoing or proposed activities (see Section 4.6).
References '
Appendixes (as needed)
Hgure 4.1 Standard format and content of regulatory analyses Establishment of problem boundaries entails the mahng of decisions as to how far the regulatory analysis will go in solv-ing the problem. Systems, equipment, and operational activities at licensed facilities are highly interrelated, and there are !
typically numerous ways of viewing any particular problem. For example, consider the failure of a particular type of valve that serves two different safety-related coolant injection systems and concurrently serves as a conainment isolation valve. The problem resulting from failure of the valve can be viewed as a system problem for either of the injection sys-tems or a problem related to isolation valves or systems, or it could be viewed as part of a larger problem, such as inade-quate maintenance or an inadequate quality assurance program.
Establishment of the appropriate boundaries can be a complicated matter. It is incumbent upon the regulatory analyst to identify other NRC programs (both ongoing and proposed) that could overlap or otherwise interface with the problem !
under consideration. The analyst should confer with those responsible for identified programs to determine appropriate boundaries. Interfacing programs should also be identified in the regulatory analysis document to facilitate communication between related programs.
NUREG/BR-0184 4.2 O
Methods
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i v) A statement of what is hoped to be achieved is also referred to as the objective. This is a concise statement of the concep-tual improvement sought by the proposed action. The objective should also be as specific as possible (assuring the public health and safety and minimizing occupational radiation exposures are two examples of objectives that are unacceptably broad). Precluding a fire from disabling redundant safety systems or reducing the probability of component failure to some particular value would be acceptably specific. Some elaboration may be required to show the reader how the objective would resolve the problem. The relationship of the objective to NRC's legislative mandatea, safety goalsm (NRC 1986), and most recent prioritization of generic safety issues (NUREG-0933 [NRC 1983b]) should be identified in appropriate cases.
4.2 Identification and Preliminary Analysis of Alternative Approaches 1 Identifying and evaluating alternative approaches to resolve problems is a key element in meeting the letter and spirit of NRC's regulatory analysis policy.
Developing a set of alternative approaches needs to be done early in the analysis process to help maintain objectivity and
! prevent premature drawing of conclusions.
- The initial set of alternatives should be broad and comprehensive, but should also be nfliciently different to provide meaningful comparison and to represent the spectrum of reasonable possibilities. Alternatives that are minor variations of I each other should be avoided. Table 4.1 contains a list of potential alternatiws that may be used to begin identification of alternatives; however, the analyst should recognize that this generic list cannot envision every possibility associated with I I
p) specific issues. Thking no action should be viewed as a viable alternative except in cases where action ha by legislation or a court decision. If a viable new alternative is identified after analysis has begun, it should be added to i
O the list of alternatives rod treated in the same manner as the original alternatives.
lhble 4.1 List of potential alternative actions
- Takmg no action (i.e., maintaming the status quo eliminate for all entries). l
- Installation of new equipment (various possibilities).
j
- Replacement of equipment (various possibilities).
- ModiScation of design.
- Modification of equipment.
- Removal of equipment.
.
- Change in invencry amount.
- Development of new procedures.
- Use of alternative processes.
- ModiScation of existing procedures. l
- Deletion of existing procedures. i
- Development of research programs to better understand the problem. l
- Facility staffing changes.
- Technical specification changes.
- Imposition oflicense conditions.
- Augmented or decreased NRC inspection.
- Varying requirements across licensee gr6ups.
1 l
4.3 NUREG/BR-0184 i
l
Methods Chapter 11 of the Regulatory Working Group's report Economic Analysis ofFederal Regulations Under Erecutive Onter O
12866 (RWG 1996) can be used in the identification and preliminary assessment of alternatives and to assist in deternmng which alternatives need to be subjected to a comprehensive value-impact analysis. The following six considerations adapted from the RWG report reflect principles included in Sections 4.2 and 4.6 of the NRC Guidelines:
- 1. Performance-oriented standards are generally preferred to engineering or design standards because performance standards generally allow licensees to achieve the regulatory objective in a more cost-effective manner.
(Section IV.B(i) of the CRGR Charter suppons performance-oriented standards.)
- 2. Different requirements for different segments or classes oflicensees should be avoided unless it can be shown that I there aie perceptible differences in the impacts of compliance or in the values to be expected from compliance.
- 3. Alternative levels of stringency should be considered to better understand the relationship between stringency and val-ues and impacts.
- 4. Alternative effective dates of regulatory compliance should be considered, with preference given to dates which favor cost-effective implementation of the regulatory action.
- 5. Alternative methods of ensuring cyliance should be considered, with emphasis on those methods which are most cost effective.
- 6. The use of economic incentives (e.g., fees, subsidies, penalties, marketable permits or offsets, changes in liabilities or pmperty rights, and required bonds, insurance, or warranties) instead of traditionally used command and control requirements should be considered in appropriate cases.
Once a broad and comprehensive list of alternatives has been developed, a preliminary analysis of the feasibility, values, and impacts of each alternative is performed. Some alternatives usually can be eliminated used on clearly exorbitant impacts in relation to values, technological infeasibility, severe enforcement or implementation problems, or other fairly obvious considerations. Reduction of the list of alternatives at this point in the analysis will reduce the resources needed to perform detailed evaluation of vabes and impacts. The regulatory analysis document should list all alternatives identified and considered, and provide a 'srief explanation of the reasons for eliminating certain alternatives during the preliminary analysis.
The level of analytical detail in the preliminary screening of alternatives need not be the same for all alternatives, particularly when one alternative can be shown to be clearly inferior or superior to the others. Rough estimates of values and impacts should be made using very simple analyses (in many cases, judgement may suffice). If several alternative actions are considered, comparison can be based on the " expected-value" of each.
Using the mugh estimates, and guidance provided by the Commission, the EDO, or the appropriate NRC office director, the significance of the problem should be estimated. This determination will usually result in a conclusion that a major or standart! effon will be expended to resolve the problem (see Figure 2.1). These two classifications are used to establish the level of detail to be provided in the regulatory analysis document and the amount of effort to be expended in perform-ing the value-impact analysis. The significance of the problem will also help determine the priority assigned to its resolution.
Alternative regulatory documents which could be used to sddress regulatory concerns should also be identified at this time.* The most common forms of documents include regulations, policy statements, orders, generic letters, and NUREG/BR-0184 4.4 O
Methods n
I C) '
regulatory guides. Alternatives could include issuance of new documents or revision or deletion of existing ones. Other implementation means should be considered when appropriate (e.g., submission of proposed legislation to Congress).
Regulatory document alternatives should only be subjected to detailed value-impact analysis if preliminary assessment indi-cates significant differences in the values or impacts among such alternatives. Otherwise, the means of implementing the proposed action should be discussed in the section of the regulatory analysis document covering implementation (see Section 4.6).
For altematives that survive preliminary screening and that require a backfit analysis according to 10 CFR 50.109(a)(3), a general description of the activities that would be required by the licensee or license applicant to complete the backfit should be prepared at this point in the regulatory analysis process. Preparation of this information will satisfy the require-ments at 10 CFR 50.109(c)(2) and Section IV.B(vii)(b) of the CRGR Charter.
The alternative approaches that irmain after the prelimmary analysis is completed will be subjected to a detailed value-impact evaluation according to the guidance presented in Section 4.3 below. Ahernative instruments will be subjected to detailed value-impact analysis only if the preliminary analysis indicates that significant differences among these alternatives exist.
4.3 Estimation and Evaluation of Wlues and Impacts This section provides gent.ral guidance on performance of a value-impact analysis. The value-impact portion of a f3 i regulatory analysis em npasses steps three and foue in the six-step regulatory analysi*: process discussed in Section 1.2.2.
(Q Detailed guidance on the value-impact analysis process is presented in Chapter 5 of this Handbook.
The following definitions of values and impacts (benefits and costs) are taken from NRC Guidelines Section 4.3 and used j in this Handbook:
Blues (Benefrrs). The beneficial aspects anticipated from a proposed regulatory action such as, but not limited to, the
- 1) enhancement of health and stfety,2) protection of the natural environment, 3) promotion of the efficient functioning of the economy and private markets, and 4) elimination or reduction of discrimination or bias.
Impacts (Costs). The costs anticipated from a proposed regulatory action such as, but not limited to, the 1) direct costs to NRC and Agreement States in administering the proposed action and to licensees and others in complying with the pro-posed action; 2) adverse effects on health, safety, and the natural environment; and 3) adverse effects on the efficient func-tioning of the economy or private markets.
The algebraic signs of values and impacts that can be quantified are provided in the description of attributes (see Section 5.5). l l
The process of selecting alternatives and performing a value-impact analysis is shown pictorially in Figure 4.2. Figure 4.2 shows each of the steps to be performed and the relationships among steps. The figure also indicates the section of this Handbook where each step is described in detail. The following discussion briefly explains each step. !
l For alterrfatives involving generic safety enhancement backfits to multiple operating nuclear power plants, the analyst begins with safety goal evaluation (i.e., whether core damage frequency (CDF) thresholds are satis 6ed or exceeded). j Based on the guidance provided in Chapter 3 of the Guidelines, the analyst determines whether or not to proceed with the Q ,l l 4.5 NUREG/BR-0184
Methods O
Anomsdhe (5eogon4.2)
V Identify Attrituos (Seedca5.5) 1 I QuentNy Change in Acoldent Frequency 3 r (Souden5.8)
Quentfy Anrtbuese 4 (Secean 5.7)
V Stanrnertas Flecults 4
($ecilon 5.8) a Yes More No , Present
, Allematives F Resulte 7 (SecNan (4)
Mgure 4.2 Steps in a value-impact analysis 1
value-impact analysis. If the safety goal evaluation of the proposed regulatory action results in a favorable determination, the analyst may presume that the substantial additional protection standard of 10 CFR 50.109(a)(3) is achievable (see Section 3.3.4 of the Guidelines).
Next, the analyst proceeds with the value-impact analysis by selecting one of the alternatives to be evaluated (see Section 4.2). For this alternative, those attributes that would be affected by implementation of the proposed action are identified. Attributes are standardized categories of values and impacts (e.g., public health [ accident] or industry implementation cost).
l NUREG/BR-0184 4.6 O
Methods
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'Ihe analyst should make every effon to use quantitative attributes relevant to the value-impact analysis. The quantifica-tion should employ monetary terms whenever possible. Dollar values should be established in real or constant dollar values (i.e., dollars of constant pmehasing power). If monetary terms are inappropriate, the analyst should strive to use other quantifiable values. However, despite the analyst's best effons at quantification, there may be some attributes which cannot be readily quantified. These attributes are termed " qualitative" and handled separately from the quantitative ones.
If appropriate, an estimate is made of the change in accident frequency which would result if the alternative were imple-mented. Parameters affected by the proposed action are identified, estimates are made for these affected parameters before and after implementation of the action, and the change in accident frequency is estimated by calculating the change in each affected accident sequence and summing them.*
l Estimates are made for those attributes which lend themselves to quantification using standard techniques. Obtaining the appropriate data may be more complicated when a major effon is being undertaken. In cases where a proposed action would result in significantly different attribute measures for different categories of licensees, separate estimates and evaluations should be made for each distinct category (e.g., older plants vs. newer plants). In backfit regulatory analyses, it is also required that the potential impact of differences in facility type, design, or age on the relevancy and practicality of the proposed backfit be evaluated [10 CFR 50.109(c)(8)].
Section 4.3 of the Guidelines identifies the need to consider attributes in terms of the different groups that may be affected by a proposed action. This 11andbook accommodates this need by the way that the suggested attributes are defined (e.g.,
impacts on the industry, the NRC, and other governmental units). If appropriate, qualitative considerations may also be evaluated. While these may be difficult to compare with the quantitative attributes, a consistent approach in their evalua-tion can result in a useful comparison among competing alternatives.
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v/ Section 4.3 of the Guidelines requires the use of best estimates. Oicn these are evaluated in terms of " expected value,"
the pmduct of the probability of some event occurring and the consequences which would occur assuming the event actually happens. Sometimes, measures other than the expected value may be appropriate, such as the mean, median, or some other point estinate. Ilowever, the expected value is generally preferred.
Section 4.3.2 of the Guidelines states that transfer payments such as insurance payments and taxes should not be included as impacts. Transfer payments are payments that reflect a redistribution of wealth rather than a social cost. Additional information on identifying transfer payments is in Section III.C.2 of the RWG repon (RWG 1996).
Depending upon the level of effort, either sensitivity or uncertainty analyses should be performed while quantifying the attributes to estimate the effect upon the results of variations in input parameters. liypothetical best- and worst-case conse-quences may be estimated for sensitivity analyses. The output from the sensitivity analyses is used to determine the impor-tance of various parameters and to approximate the uncertainties associated with the results. Actual uncenainty analyses should be more rigorous. A number of techniques are available, each with differences in usefulness of results and the amount of resources required. Uncertainty analyses should produce actual probability distributions for the overall results based on assumed distributions for selected input parameters. The differences between sensitivity and uncertainty analyses and their respective roles in regulatory analysis are discussed in Section 5.4.
At this point, the above steps are repeated if there is another alternative to be evaluated. If not, results for all evaluated alternatives are put into a form for presentation in the regulatory analysis document. Guidance for performing each of the above steps is provided in detail in Chapter 5.
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4.7 NUREG/BR-0184
l Methods 4.4 Presentation of Results O
I i
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l l The following items must be included in the presentation of results section of the regulatory analysis document for each l alternative:
results of the evaluation for compliance with the Safety Goal guidance, if appropriate (see Section 4.4 of the l Guidelines) j presentation of the net value (i.e., the algebraic sum of the attributes) using the discount rate procedures stated in Section 4.3.3 of the Guidelines and discussed in Sections 5.7 and B.2 of this Handbook f' estimates for each attribute for each alternative (the analyst can choose to present the estimates in tabular or graphica!
form if such presentation would aid the reader) I l
presentation of any attributes quanti 6ed in non-monetary terms in a manner to facilitate comparisons among alternatives .
1 the distribution of values and impacts on various groups if significant differences exist between recipients of values and those who incur impacts (see Section 4.4 of the Guidelines) e discussion of key assumptions and results of sensitivity analyses or tmcertainty analyses e
impacts on other NRC programs and federal, state, or local government agencies.
Key assumptions are to be specifically stated so that readers of the regulatory analysis have a clear understanding of the analysis and the decision-maker will be able to assess the confidence to place in the results. Sources and magnitudes of uncertainties in attribute estimates and the methods used to quantify sensitivity or uncertainty estimates should be discussed in all regulatory analyses.
For alternatives projected to result in significantly different attribute measures for different categories oflicensees, sepa-rate evaluations should be made for each distinct category. In cases where significant differences exist, their distributions with respect to the various groups invoked should be discussed.
He effects of the pmposed action on other NRC programs need to be assessed. These could include eliminating or creat-ing a need for other programs; use of limited NRC resources resulting in postponement or rescheduling of other programs; modifying accident probabilities resulting in changes to priority of, or need for, other programs; or developing information with a bearing on other programs. Effects on other government agencies, if any, should also be assessed and reported.
In cases where uncertainties are substantial or where important values cannot be quantified, alternatives that yield equiva-lent values may be evaluated band on their cost-effectiveness. This methodology should also be used when the levels of values are specified by statute.
Proposed actions subject to the backfit rule should be evaluated against the following two criteria from 10 CFR 50.109(a)(3):
Is there a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived fmm the backfit?
NUREG/BR-0184 4.8 O
Methods i s (vl
- Are the direct and indirect costs of implementation justified in view of this increased protection?
Guidance on application of the "substtntial increase" standard is in Attachment 3 to the CRGR Charter. Each alternative that meets both of the preceding criteria should be so indicated, and a discussion of why the criteria are met should be developed. Backfitting will be required by the NRC only if both criteria are met.
For CRGR regulatory analyses, the following information (from Table 2.3) should be included in the presentation of results:
- The sponsoring office's position on whether the proposed action would increase requirements or staff positions, implement existing requirements or staff positions, or relp or reduce existing requirements or staff positions.
4.5 Decision Rationale This element of the regulatory analysis provides the basis for selection of the recommended alternative over the other alter-natives considered. In selecting the preferred alternative, decision criteria are used and reported in the regulatory analysis document. Section 4.5 of the Guidelines gives the minimum set of decision criteria to be used, as well as other considerations. .
The net-value calculation is a compilation of all of the attributes that can be quantified in monetary terms. Certain attri- )
butes are generally quantified in other than monetary terms (e.g., public health [ accident], which is measured in person p)
(,
w rems of exposure) and converted to monetary terms with an established conversion factor (see Section 5.7.1.2). These attributes are included in the net-value calculation. To aid the decision maker, the net value is to be cornputed for each alternative.
In considering the net value, care must be taken in interpreting the significance of the estimate. An algebraically positive estimate would indicate that the action has an overall beneficial effect; a negative estimate would indicate the reverse.
However, if the net value is only weakly positive or negative, it would be inappropriate to lean strongly either way since minor errors or uncertainties could easily change the sign of the net value.
If the net value is calculated to be strongly positive or negative, the result can be given considerable significance since the variations in the assumptions or data would be much less likely to affect the sign of the net value. Even so, other consid-erations may overrule the decision supponed by the net value (e.g., qualitative factors such as those embodied in the
" qualitative" attributes).
Non-quantifiable attributes can only be factored into the decision in a judgmental way; the experience of the decision-maker will strongly influence the weight that they are given. These attributes may be significant factors in regulatory deci-1 sions and should be considered, if appropriate.
In addition to being the "best" alternative based on monetary and non-monetary considerations, the selected alternative must be within the NRC's statutory authority and, when applicable, consistent with NRC's safety goals and policy. A ,
showing of acceptable impact of the proposed action on other existing and planned NRC programs and requirements is also necessary. This will ensure that there are no negative safety impacts in other areas, that NRC resources are being used responsibly, and that all actions are adequately planned and coordinated. Any other relevant criteria may be used with adequate documentation in the regulatory analysis.
tph V
4.9 NUREG/BR 0184
Methods Recommended actions in backfit regulatory analyses must meet the two additional criteria from 10 CFR 50.109(a)(3),
O namely that 1) there is substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from the backfit, and 2) the direct and indirect costs of implementation are justified in view of this increased protection. The recommended action must be shown to meet these criteria, and, therefore, must be selected from those alternatives shown to meet the criteria.
Each proposed alternative should be reviewed to determine whether it is an interim or final action. In cases where the action is interim, it is n-cessary to develop an adequate justification for imposing the proposed backfit on an interim basis.
If such justification cannot be satisfactorily developed, the alternative should be dropped from further consideration.
For CRGR regulatory analyses, the following information (from Table 2.3) should be included in the decision rationale:
For proposed relaxations or decreases in current requirements or staff positions, a rationale for the determination that
- 1) the public health and safety and the common defense and security would continue to be adequately protected if the proposed reduction in requirements or positions wre implemented; and 2) the cost savings attributed to the action would be substantial enough to justify taking the action, and clearly outweigh any reduction in benefits.
Recommended actions in CRGR regulatory analyses involving proposed relaxations or decreases in current requirerrants or staff positions must meet the following two additional criteria found in Section IV.B(x) of the CRGR Charter: 1) the public health and safety and the conunon defense and security would continue to be adequately protected if the proposed reduction in requirements or positions were implemented, and 2) the cost savings attributed to the action would be substan-tial enough tojustify taking the action, and clearly outweigh any reduction in benefits. Also, the analysis must indicate whether the proposed relaxation or decrease in current requirements or staff positions is optional or mandatory.
4.6 Implementation O
An implementation schedule for the proposed action must be prepared. 'Ibe schedule Inust identify all major steps or actions to be taken by all affected parties (the NRC, Agreement States, licensees, and any others), and the dates or amounts of time allocated to accomplish each step. The schedule must be realistic and allow sufficient time for such fac-tors as needed analyses, approvals, procurement, installation and testing, and training. Anticipated downtime of licensee facilities to implement the proposed action must be specifically identified. Availability and lead time required for acquisi-tion and installation of new equipment and replacement parts must be addressed. For NRC planning purposes, short- and long-term actions are to be identified in such a way as to clearly differentiate the two.
I For backfit regulatory analyses, the implementation schedule should account for other ongoing regulatory activities at the i facility. The backfit regulatory analysis document should describe how this is accomplished in the recommended schedule.
For CRGR regulatory analyses, the proposed method of implementation and the proposed generic requirement or staff position as it is proposed to be sent out to licensees should be included in the implementation section (see Table 2.3).
The implementation section of the regulatory analysis document should also identify the proposed NRC instrument (e.g.,
1 rule, regulatory guide, policy statement) for implementing the proposed action and the reasons for selecting the proposed l instrument. The relationship of the proposed action to other NRC programs, actions, and requirements, both existing and l proposed, should be established. Tb the extent possible, the analyst should assess the effects of implementation of the pro-posed action on the priorities of other actions and requirements and the potential need to tevisit other regulatory analyses.
NUREG/BR-0184 4.10 O
Methods 4.7 Endnotes for Chapter 4
- 1. Agreement States are states which have entered into an agreement with the NRC under Section 274b of the Atomic Energy Act to assume regulatory authority over byproduct materials, source materials, and small quantities of spe-cial nuclear materials insufficient to form a critical mass.
- 2. The Commission has directed NRC staff to ensure that future regulatory actions involving generic safety -
enhar.ccments to nuclear power plants are evaluated for conformity with the NRC Safety Goals (NRC 1990b).
l l
- 3. NUREG/BR-0070 (NRC 1984a) discusses various types of formal NRC documents. Attachment 2 to the CRGR l Charter identifies mechanisms that can and cannot be used to establish, interpret, or communicate generic requirements or staff positions to licensees.
- 4. Although most actions are expected to affect risk through a change in accident frequency, some may change conse-quences instead. Evaluating the change in risk for these latter actions is discussed in Section 5.7.1.1.
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l 4.11 NUREG/BR-0184 F
O b
5 Value-Impact Analysis
'Ibe discussions presented in this chapter generally apply to both power reactor and non-reactor facilities. To simplify the presentation, the term " facility" has been selected to serw as the generic indicator for both types. Where the discussion is specific to power reactor versus non-reactor facilities, this will be indicated. Material supplemental to that presented in this chapter for power reactor and non-reactor value-impact analyses is included in Appendixes B and C, respectively.
5.1 Background
Value-impact analysis is one form of formal decision analysis, not necessarily binding. Formal decision methods can
- help the analyst and decision-maker clearly define and think through the problem ;
- segment complex pmblems into conceptually manageable portions
- provide a logical structure for the combination of issues contributing to a decision
- clearly display beneficial and detrimental aspects of a decision
- provide a record of the decision rationale, helping to provide documentation, defensibility, and reproducibility
- focus debate on the specific issues of contention, thereby assisting resolution
- provide a framework for the sensitivity testing of data and assumptions.
However, limitations must be noted. Formal decision methods cannot
- completely remove subjectivity
- guarantee that all factors affecting an issue are considered ,
e produce imambiguous results in the face of closely valued alternatives and/or large uncertainties I
- be used without critical appraisal of results; to use a decision analysis method as a black box decision-maker is both wrong and dangerous.
l 5.2 Methods The value-impact portion of a regulatory analysis encompasses the thini and fourth steps of the complete six-step regula- ;
tory analysis process discussed in Section 1.2.2. Value-impact analysis identifies and estimates the relevant values and .
impacts likely to result from a proposed NRC action. The methodology outlined in this chapter guides the systematic j definition and evaluation of values and impacts. It also provides guidance on the it: porting of results.
O 5.1 NUREG/BR-0184
Value-impact O
Values and impacts are classified as " attributes." Attributes are the principal components of value-impact assessment that are used to characterize the consequences of a proposed action. Any given NRC action can affect a large number of fac-tors within the public and private sectors. The attributes represent the factors that are most frequemly affected by a proposed NRC action. The attributes affected by any given proposed action will vary, however, and the analyst will have to determine the appropriateness of each attribute. Attributes, whether values or impacts, can have either positive or nega-tive algebraic signs, depending on whether the proposed action has a favorable or adverse effect. The sign conventions are as follows: favorable consequences are positive, advetse consequences are negative. Each attribute measures the change from the existing condition due to the proposed action. Attributes are discussed in detail in Sections 5.5 and 5.7.
Section 4.4 Of the Guidelines requires that the value-impact of an alternative be quantified as the " net value" (or " net bene-fit"). To h extent possible, all attributes, whether values or impacts, are quantified in monetary terms and added together (with the appropriate algebraic signs) to obtain the net value in dollars. The net value calculation is generally favored war other measures, such as a value-impact ratio or internal rate of return (RWG 1996,Section III. A.2).m The net-value method calculates a numerical value that is intended to summarize the balance between the favorable and unfavorable consequences of the proposed action. The basic perspective of the net-value measure is national economic efficiency. All values and impacts are added together and the total is intended to reflect the aggregate effect of the pro-posed action on the national economy. The net-value measure does not, and is not intended to, provide ary information about the distribution of values and impacts within the national economy. He values and impacts to all affected panies are simply added together.
Section 4.4 of the Guidelines states that if significant differences exist between recipients of values and those who incur impacts, the distribution of values and impacts on various groups should be presented and discussed.Section III.A.8 of the 1996 RWG report supports this position.
To calculate a net value, all attributes must be expressed in common units, typically dollars. Person-rems of averted expo-sure, a measure of safety value, is converted to dollars via a dollar / person-rem equivalence factor (see Section 5.7.1.2).
Net value is an absolute measure. It indicates the magnitude of the proposed action's contribution toward the specified goals. When faced with a choice between two mutually exclusive actions, the " optimal" decision is to select the action with the larger net value.
5.3 Standard Analysis Section 2.4 introduced the concept of a standard regulatory analysis, generally expected to encompass approximately one to two person-months of effort using specific guidance provided in this Handbook, ne standard analysis should be adequate for most regulatory analyses, requiring guidance only from the NRC Guidelines, Handbook, and appropriate references.
Sections 5.4-5.8 and Appendixes A, B, and C provide information for the level of detail deemed sufficient for a standard regulatory analysis. For those issues which require major levels of effort, this Handbook suggests additional methods and trferences which should prove useful. In general, the numerical values provided by this Handbook represent " generic" values which, in practice, apply better to multiple licensees than to individual licensees. For regulatory actions involving individual licensees, plant-specific values are reconunended. However, as these are often unavailable, the analyst may be limited in some cases to applying generic values to plant-specific cases.
5.2
Value-Impact rh t \
U 5.4 Treatment of Uncertainty Chapter 4 of the NRC Guidelines requires that uncertainties be addressed in regulatory analyses, both for exposure and cost measures. In addition, NRC's Final Policy Statement on the use of probabilistic risk assessment (PRA) in nuclear regulatory activities (NRC 1995b) states that sensitivity studies, uncertainty analysis, and importance measures should be used in regulatory matten, where practical within the bounds of the state-of-the-art. Uncenainties in exposure measures, especially those related to facility accid'ents, have traditionally been difficult to estimate. With respect to power reactor facilities, much has been written about uncertainty analysis in risk assessments. The more rigorous assessments typically provide an uncertainty analysis, usually performed via stochastic simulation on a computer. Briefly, the analyst determines probability distributions for as many of his i >put parameters as deemed necessary and practical. A computer code then samples values from each distribution randomly and propagates these values throt% h the risk equation to yield one result. When repeated a large number of times (at least several huadred), a probability distribution for the result is generated, from which the analyst can extract meaningful statistical values (e.g., mean, standard deviation, median, and upper and lower bounds for given confidence levels).
Risk assessments for non-reactor facilities often identify best estimates only. Some have provided uncenainty ranges (see Appendix C), but their development has generally been less rigorous than that for reactor facilities. On the positive side, :
accident scenarios for non-reactor facilities are much less complex than for power reactors, facilitating uncenainty estimation, at least from a calculational penpective. l i
This Handbook is not intended to provide basic information on probability and statistics, and therefore does not attempt to describe the details of uncertainty analysis techniques. The analyst needing information on these topics is referred to text-O books on probability and statistics, as well as the following references: Seiler (1987), Iman and Helton (1988), Morgan
's' j and Hention (1990), and DOE (1996). Instead, this Handbook presents a general discussion of the types of uncertainty that will be encountered in a regulatory analysis, primarily the value-impact ponion, and outlines some of the more recent approaches to deal with them.
5.4.1 Types of Uncertainty Vesely and Patsmuson (1984) identified seven categories of uncenainties in PRA, the majority of which, if treated at all, have only recently begun to receive attention. The seven categories are uncertainties in data, analyst assumptions, modeling, scenario completeness, accident frequencies, accident consequences, and interpretation. These seven categories, going from first to last, represent a progression from uncertainties in the PRA input to higher-level ,
uncenainties with the PRA results. Vesely and Rasmuson considered these categories to be generally applicable to any l modeling exercise, not just a PRA. Thus, they would also apply to the cost analysis ponion of the regulatory analysis.
The first category, data uncertainty, is the most familiar and most often treated. It can be divided into four groups: popu-lation variation, imprecision in values, vagueness in values, and indefiniteness in applicability. Population variation refen to parameter changes from scenario to scenario, usually due to physical causes. The variations occur among the random variables which, when treated as constants, give a false impression of the stability of the results. Parameter imprecision and vagueness refer to separate concepts. Imprecision occurs when only limited measurements are available from which to estimate parameter values. Vagueness occurs when definitive values or intervals cannot be assigned to parameters.
Indefinite applicability deals with the extrapolation of parameter values to situations different from those for which they were derived (e.g., extrapolating component failure data for normal environments to accident conditions).
The second category, analyst uncertainty, refers to variations in modeling and quantification which arise when different analysts perform different ponions of the analysis. Often included with data uncertainty, analyst uncenainty provides its N own separate contribution. Modeling uncertainty, the third category, arites from the indefmiteness in how comprehensive
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and how well characterized are the numerous models in the analysis. Do the models account for all significant variables?
How well do the models represent the phenomena? Is the dependence between two phenomena accurately modeled? Simi-lar to modeling uncenainty is completeness uncertainty, the fourth category. It dif= nnly in that it occurs at the initial, identification stage in the analysis. When the analytic " boundaries" are drawn at the start of the analysis, how can one be i sure that all "important* items have been included (e.g., the Three-Mile Island core-damage scenario was not specifically identified in PRAs until it had occurred)? Even if the imponant items have been included, are their interrelationships ade-quately defined (if ewn known)?
He last three uncenainty categories-those for accident frequencies and consequences, and interpretation-deal with the analytic output and results. Accident frequency uncertainties arise from two sources: variations between accidents of the j same type and limited knowledge of the data, models, and completeness. Accident consequence uncenainties parallel j those in accident frequency, except that they involve consequence modeling rather than frequency estimation. Interpreta- l tion uncertainty arises from the combination of all previous uncertainties plus the difficulty in conveying the information to the decision-maker. Even the most precise uncertainty analysis can be wasted if the meaning cannot be transferred to the l decision-maker. Often, this results fium difficulty in the way the results are presented. Ernst (1984) provides insight on reducing the uncenainty in interpretation of results.
5.4.2 Uncertainty Versus Sensitivity Analysis As defined by Vesely and Rasmuson, uncertainty and sensitivity analyses are similar in that both strive to evaluate the variation in results arising from the variations in the assumptions, models, and data. However, they differ in approach, scope, and the information they provide.
Uncertainty analysis attempts to describe the likelihood for different size variations and tends to be more fonnalized than sensitivity analysis. An uncertainty analysis explicitly quanti 5es the uncenainties and their relative magnitudes, but requires probability distributions for each of the random variables. The assignment of these distributions often involves as much uncenainty as that to be quantified.
Sensitivity analysis is generally more straightforward than uncertainty analysis, requiring only the separate (simpler) or simultaneous (more complex) changing of one or more of the inputs. Expert judgment is involved to the extent that the analyst decides which inputs to change, and how much to change them. This process can be streamlined if the analyst ,
knows which variables have the greatest effect upon the results. Variation of inputs one at a time is preferred, unless i multiple parameters are affected when one is changed. In this latter case, simultaneous variation is required. Hamby '
(1993) provides a detailed description of the most common techniques employed in sensitivity analysis.
Vesely and Rasmuson identify which of the seven types of uncertainths encountered in PRAs are best handled by uncer-tainty versus sensitivity analysis. ney are as follows:
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- 1. Data Uncertainty: Use uncertainty analysis for population variation and value imprecision, sensitivity analysis for i value vagueness and indefiniteness in applicability.
- 2. Analyst Uncertainty: Use sensitivity analysis.
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- 3. Modeling Uncertainty: Use sensitivity analysis. J I
- 4. Completeness Uncertainty: Use sensitivity analysis. l I
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- 5. Frequency Uncertainty: Use uncenainty analysis for variation from one accident to another, sensitivity analysis for the limited knowledge of the data, models, and completeness.
- 6. Consequence Uncertainty: Use uncertainty analysis for variation from one accident to another, sensitivity analysis for the limited knowledge of the data, models, and completeness.
- 7. Interpretation Uncenainty: Use sensitivity analysis.
5.4.3 Uncertainty / Sensitivity Analyses nree major NRC studies involving detailed uncertaimy/ sensitivity analyses were NUREG-ll50, Severe Accident Risks:
An Assessmentfor Rve U.S. Nuclear itnwr Plants (NRC 1991); NUREG/CR-5381, Economic Risk of Contamination Cleanup Costs Resultingfmm Isrge Non-Reactor Nuclear Material Licensee Opemtions (Philbin et al.1990); and NUREGICR-4832, Analysis of the LaSalle Unit 2 Nuclear lher Plant: Risk Methods Integration and Evaluation Pmgmm (RMIEP) (Payne 1992). The first and third studies address reactor facilities, the second non-reactor facilities.
The approach used in each study is summartzed below.
5.4.3.1 NUREG-1150 "An imponant characteristic of the PRAs conducted in support of this report [NUREG-1150] is that they have explicitly included an estimation of the uncertainties in the calculations of core damage frequency and risk that exist because of l incomplete understanding of reactor systems and severe accident phenomena." With this introduction, NUREG-1150 iden-J
[^} tified four steps in the performance of its uncertainty / sensitivity analysis: ;
- 1. Define the Scone. The total number of parameters that could be varied to produce uncertainty estimates was quite large and limited by computer capacity. Thus, only the most imponant sources were included, these sources being identified from previous PRAs, discussion with phenomenologists, and 'dmited sensitivity analyses. For those parame-ters important to risk and having large uncertainties and limited, if any, data, subjective probability distributions were generated by expert panels.
- 2. Define Specific Uncenainties. Each section of the risk assessment was conducted at a slightly different level of detail, none of w.nich to the degree involved in a mechanistic analysis. This resulted in the uncertain input parameters being "high level" or summary parameters, for which their relationships with their fundamental physical counterpart parameters were not always clear. This resulted in Vesely and Rasmuson's "modeling uncertainties." In addition, ,
" data uncenainties" arose from limited knowledge of some important physical or chemical parameters. NUREG-1150 l included both types of uncertainty, with no consistent effort to distinguish between them. j
- 3. Define Probability Distributions. Probability distributions were developed by several methods, paramount among these being " expert clicitation" (discussed below). " Standard" distributions employed in previous risk assessments were used when the expens' estimation was not needed.
- 4. Combination of Uncertainties. The Latin hypercube method, a specialized form of stochastic simulation, was employed to sample from the various probability distributions. The sampled values were propagated through the con-stituent analyses to produce probability distributions for core damage frequency and risk. Results were presented graphically as histograms and complementary cumulative distribution functions showing the mean, median, and two-sided 90% confidence intervals.
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A major innovation of the NUREG-IISO project was the development of a formal method for clicitation of expert judg-ment. Nine steps were involved:
- 1. Selection of Issues. He initial list of issues was identified from the important uncertain parameters speciSed by each plant analyst.
- 2. Selection of Experts. Seven expen panels were assembled to address issues in accident frequency (two panels), acci-dent progression and containment loading (three panels), containment structural response (one panel), and source terms (one panel). Selection was based on recognized expenise in the nuclear industry, the NRC and its contracton, l and academia. Each panel contained 3-10 expens.
- 3. Elicitation Training. Decision analysis specialists trained both the experts and analysis team members in elicitation methods, including the psychological aspects of probability estimation. He experts perfected their estimation tech- !
niques by conjuring probabilities for items for which 'true" values were known.
- 4. Presentation and Review of issues. The analysis staff formally presented the relevant issues to each panel mer the course of several days. Interactive discussions ensued.
- 5. Preparation of Exoert Analyses. Over a periods ranging from one to four months, each panel deliberated on its issues. However, each panel member arrived at his/her own quantitative results.
- 6. Expert Review and Discussion. At a final meeting, each expert presented his/her analysis which, in some cases, resulted in members modifying their preliminary results subsequent to the meeting.
- 7. Elicitation of Experts. Two analysis staff members, one trained in clicitation techniques, the other familiar with the technical subject, interviewed each expen privately. The expen's final quantitative results were documented.
- 8. Accrenation of Judcments. From each expen's results, the analysis staff composed probability distributions which were then aggregated to produce a single composite for each issue. Each expert was equally weighted in the composite.
- 9. Review by Experts. Each expert's pmbability distribution, as developed by the analysis staff from the expen's inter-view, was reviewed privately with that expert to correct any misconceptions that may have arisen. The probability distribution was then fin.dized, as was the composite.
5.4.3.2 NUREG/CR-5381 In NUREG/CR-5381, Philbin et al. took advantage of some of the convenient combinatorial properties of the lognormal distribution to facilitate a straightforward uncertainty analysis. NUREG/CR-5381 assessed the economic risk of cleanup costs resulting from non-reactor NRC licensee contamination incidents (see Section C.4). The calculational procedure involved three steps: estimating the frequency and cleanup cost of each accident scenario, taking their product to yield the
' cleanup risk' (probabilistically-weighted cleanup cost) per scenario, and summmg the scenario risks to ykld the total facility risk. The uncenainty analysis paralleled these three steps.
For both the accident frequency and cleanup cost, probability distributions were selected from the available data, if possi-ble, or by expert judgment. When using historical data to obtain frequency estimates, the assumption was made that the number of incidents for a specified scenano followed the Poisson distribution, his was deemed reasonable in light of the small number of incidents over a relatively large number of operating yean and the absence of any obvious trends. The O
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! Poisson point estimate incident rate was taken to be the historical rate, with two-sided 80% confidence bounds derived from the properties of the Poisson distribution.
When a calculational model was used to estimate the frequency, the uncertainty was based on expert judgment. Unless deemed inappmpriate, the frequency distribution was taken to be lognormal with an error factor of 10. If previous analyses provided only a frequency range, the distribution was again assumed to be lognormal, with the upper and lower bounds taken as the endpoints of this range. Thus, the point estimate (median, in this case) became their geometric mean.
For the cleanup costs, the point estimates were derived from historical data of calculational models. These costs were assumed to be lognormally distributed with error factors of 1.25.
Philbin et al. defended their choice of the lognormal as a " generically" representative probability distribution for several reasons. The lognormal has a minimum value of zem, a realistic limit on the minimum frequency and cost, and is skewed in a way which yields relatively wider error bounds on the upper than lower side. Thus, it produces an uncertainty band which is conservative. Also, the lognormal has two convenient combinatorial properties. De product of two lognormally distributed variables is lognormally distributed, while the sum can be appmximated by another lognormal provided one variable dominates the other.
The economic risk per accident scenario was estimated by propagating the frequency and cost uncertainties through their product. When both frequency and cost were lognormally distributed, this product was also lognormal. When the fre-quency distribution was Poisson, it was approximated by a lognormal to simplify the calculation. Each scenario thus resulted in an economic risk which was lognormally distributed. These were summed to yield the total economic risk per facility. The individual variances were summed and the resultant total economic risk was assumed to be appmximately lognormal (p) v one can see, a reasonable assumption if it was dominated by one scenario risk. Refe that this assumption was generally valid for three of the five facilities (i.e., one scenario risk contributed over 50% to the total facility risk). The Snal results were reported as two-sided 80% confidence bounds.
5.4.3.3 NUREGICR-4832 in NUREG/CR-4832, Payne generally followed an uncertainty / sensitivity calculational procedure similar to that employed in NUREG-ll50. The major contribution was the development of a new computer code, TEMAC (Iman and Shortencarier 1986) to perform the final quantification of the accident sequence uncertainties via the Latin hypercube sampling method. The TEMAC code also calculated various risk importance measures (Vesely et al.1983) and ranked the basic events by their contribution to mean core damage frequency.
Bree importance measures were estimated in NUREG/CR-4832. The first, risk reduction importance, calculates the decrease in the total core damage frequency which could result if a single basic event's probability were set to zero (i.e.,
the component could not fail or the event could not occur). The second, risk increase importance, calculates the increase l in the core damage frequency which could result if a single basic event's probability were set to one (i.e., the component !
would always fail or the event would always occur). The third, uncertainty importance, estimates the extent to which the uncertainty in the total core damage frequency depends upon the underlying uncertainty in a common contributor to a set of related basic events (e.g., a failure to actuate in all motor-operated valves). Rese importance measures represent a combination of sensitivity with uncertainty analyses which feature some of the better aspects of each.
5.4.4 Suggested Approach The value-impact portion of a regulatory analysis will often require use of an existing risk assessment for the estimation of some of the attributes. If the risk assessment has an uncertainty / sensitivity analysis accompanying it, the analyst should 1
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try to adapt it for use in the value-impact analysis. Unfortunately, this is often impractical for the standard analysis since the analyst does not have access to the computer code and numerous data and assumptions necessary to generate the resul-tant probability distributions.
When a detailed uncertainty / sensitivity analysis is not possible or practical, the following approach is suggested for the standard analysis. The standarti anrlysis should attempt to include an uncertainty / sensitivity analysis approaching the level of that conducted by Philbin et al. in NUREG/CR-5381 (see Section 5.4.3.2). This analysis can be done with varying degrees of formality and rigor. First, a systematic attempt should be made to identify all of the pertinent factors (assump-tions, data, models) that could affect the results. Since the number of such factors is usually very large, not all of them can be treated in detail. Nevertheless, it is useful to make a systematic effort at least to identify them. As a second step, the list of factors should be screened to select a subset for detailed examination. The screening process should concentrate on eliminating unimportant factors (for example, those that are known to contribute little to the overall uncertainty or those that have minimal effect on the bottom line results) and reducing the list to manageable size. Typically, the screening will be done on the basis of judgment and experience, but more formal methods and calculations may be appropriate in some circumstances (e.g., an abridged form of the " expert clicitation" procedure in NUREG-1150 [see Section 5.4.3.1]). The third step is to defme a set of cases to be evaluated. The most common approach is to define a best estimate, establish a range of interest for each factor, and then systematically vary the factors, one or more at a time. The results are then expressed as a range (low value, best estimate, high value) which indicates the effect on the output of variations in the factors, and thus provides some insight concerning uncertainties and their effects.
Uncertainty / sensitivity analysis for the cost measures is generally simpler than that for exposures. Complex accident sce-narios are not involved. Moreover, the analyst usually has a better " feel" for cost-related measures (e.g., labor rates, interest rates, and equipment costs) than for risk-related ones. Thus, such analyses require no more than the straight-forward variation of interest rates, labor hours, contingency factors, etc. However, the analyst is cautioned that, while the calculational techniques may be simple, wide ranges can still result.
To assist the analyst in performing uncenainty/ sensitivity analyses for the standard analysis, this Handbook provides high and low values for selected best estimates in the evaluation of certain attributes (see, for example, Section 5.7.3.1).
Should the analyst have access to better estimates, they should be used. In the cases where the analyst has access to a computerized assessment, the uncertainty / sensitivity analysis results obtainable via computer can be incorporated into the standard analysis. However, it is felt that more formal uncertainty / sensitivity analyses will only be practical for regulatory analyses requiring major efforts.
Finally, automated uncertainty calculations using default distributions are a feature of the FORECAST computer code for regulatory effects cost analysis (lepez and Sciacca 1996). Uniform, lognormal, and several user-specified probability distributions are options.
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5.5 Identification of Attributes For ever,< v6e-impact analysis to be performed, those attributes that could be affected by the proposed action must be )
identified. Once identified, the attributes may be quantified using the techniques presented in Sections 5.6 and 5.7. Note that the subsections of this section and Section 5.7 are numbered so as to correspond to one another in their discussions of i
the attributes. This section introduces the most commonly used attributes. Most of the attributes presented may be quantified in monetary terms, either directly or through use of a radiation exposure-to-money conversion factor (see Section 5.7.1.2). The remaining attributes are not readily quantifiable and are treated in a more qualitative manner.
However, the analyst shoul6 attempt quantitative estimation whenever possible, relying on qualitanve descriptions when no quantification is feasible. I O
NUREG/BR-0184 5.8
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Value-Impact l Table 5.1 is a checklist for identifying affected attributes. The analyst is encouraged to use this checklist when first deter-mining the attributes that will need to be evaluated. For each attribute listed, a check should be made ifit is affected.
Each affected attribute can then be evaluated according to the instructions included in Sections 5.6 and 5.7.
Thble 5.1 Checklist for identifleation of afected attributes Attribute Afected Public Health (Accident) O Public Health (Routine) O Occupational Health (Accident) O Occupational Health (Routine) O Offsite Property O 1
I Onsite Property O Industry implementation O Industry Operation O NRC Implementation O NRC Operation O Other Government O General Public 0 Improvements in Knowledge O Regulatory Efficiency O Antitmst Considerations O Safeguards and Security Considerations O Environmental Considerations O Other Considerations (Specify) O m
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5.5.1 Public Health (Accident)
This attribute is a value which measures expected changes in radiation exposures to the public due to changes in accident frequencies or accident consequences associated with the proposed action. For nuclear power plants, expected changes in radiation exposure should be measured ove: a 50-mile radius from the plant site. The appmpriate distance for other types oflicensed facilities should be determined on a case-by-case basis. In most cases, the effect of the proposed action would be to decrease public exposure. A decrease in public exposure (given in person-rems) assumes a positive sign. Therefore, this decrease multiplied by the monetary conversion factor ($/ person rem) will give a positive monetary value.
It is possible that a proposed action could increase public exposure due to potential accidents. In this case, the increase in public exposure (person-rems) assumes a negative sign. When this increase is multiplied by the monetary conversion factor ($/ person-rem), the resulting monetary term is interpreted as negative.
5.5.2 Public Health (Routine)
This attribute is a value which accounts for changes in radiation exposures to the public during normal facility operations (i.e., non-accident situations). It is expected that this attribute would not be affected as often in reactor regulatory analy-ses as in non-reactor ones. When used, this attribute would employ an actual estimate; accident probabilities are not involved.
Similar to the attribute for public health (accident), a decrease in public exposure would be positive. Therefore, the prod-uct of a decrease in exposure and the monetary conversion factor (assumed to be the same factor as that for public health
[ accident]) would be taken as positive. The product of an increase in public exposure and the monetary conversion factor would be taken as negative.
5.5.3 Occupational Health (Accident) l This attribute is a value which measures health effects, both immediate and long-term, associated wi;h site workers as a result of changes in accident frequency or accident mitigation. A decrease in worker radiological exposures is taken as positive; an increase in worker expcsures is considered negative.
1 As is the case for public exposure, t) e directly calculated effects of a particular action are given in person-tems. A mone- l tary conversion factor must be used to convert the effect into dollars. Under current NRC policy the value to be used is l
$2000 per person-rem (see Sectint 5.7.1.2). This value is subject to future revision. l l
5.5.4 Occupational Health (Routine) )
This attribute is a value which accounts for radiological exposures to workers during normal facility operations (i.e., non-accident situations). For many types of proposed actions, there will be an increase in worker exposures; sometimes this will be a one-time effect (e.g., installation or modification of equipment in a hot area), and sometimes it will be an ongoing effect (e.g., routine surveillance or maintenance of contaminated equipment or equipment in a radiation area).
Some actions may involve a one-time increase with an offsetting lowering of future exposures.
This antibute represents an actual estimate of health effects; accident probabilities are not relevant. As is true of other l types of exposures, a net decrease in worker exposures is taken as positive; a net increase in worker exposures is taken as negative. This exposure is also subject to the dollar per person-rem conversion factor (see Section 5.7.1.2).
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( l 5.5.5 Offsite Property This attribute is a value which measures the expected total monetary effects on offsite propeny resulting from the proposed action. Changes to offsite propeny can take various forms, both direct (e.g., land, food, and water) and indittet (e.g.,
tourism). This attribute is typically the product of the change in accident frequency and the propeny consequences resulting from the occurrence of an accident (e.g., costs of interdiction measures such as decontamination, cleanup, and evacuation). A reduction in offsite propeny damage is taken as positive; an increase in offsite property damage is considered negative.
5.5.6 Onsite Property This attribute is an impact which measures the expected monetary effects on onsite propeny, including replacement power (specifically for power reactors), decontamination, and refurbishment costs, fmm the poposed action. This attribute is typically the product of the change in accident frequency and the onsite propeny consequences given that an accident were to occur. A reduction in expected onsite property damage is taken at positive; an increase in onsite property damage is considered negative. Panicular cart should be taken in estimating dollar savings associated with this attribute because
- 1) values for this a' tribute are difficult to accurately estimate, and 2) estimated values can potentially significantly outweigh other values and impacts associated with an alternative. ,
i 5.5.7 Industry Implementation j i
This attribute is an impact which secounts for the projected net economic effect on the affected licensees to install or implement mandated changes. Costs will include pmcedural and administrative activities, equipment, labor, inaterials, and shudown costs, including the cost of replacement power in the case of power reactors (see Section 5.7.7.1), as Y appropriate. Additional costs above the status quo are considered negative; cost savings would be considered positive.
This attribute, and the following five, reflect actual estimated costs; accident probabilities are not involved. In this regard, these attributes are sneasured very differently from those associated with accident-related health effects and onsite and offsite propeny.
i 5.5.8 Industry Operation l This attribute is an impact which measures the projected net economic effect due to routine and recurring activities required by the proposed action on all affected licensees. If applicable, replacement pour costs (pour reactors only) directly attributable to the proposed action will be included. Additional costs above the status quo are taken to be negative; cost savings are taken to be positive.
Costs falling in this category, and those anociated with NRC operational considerations, generally occur over long periods of time (the facility lifetime). These costs are particularly sensitive to the discount factor used.
5.5.9 NRC Implementation This attribute is an impact which measures the projected net economic effect on the NRC to place the proposed action into operation. Costs already incuned, including all pre-decisional activities performed by the NRC, are viewed as " sunk" costs and are not to be included. Additional costs above the status quo are taken to be negative; cost savings are taken to be positive.
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l The NRC may seek compensation (e.g., license fees) from affected licensees to provide needed services; any O
compensation received should not be subtracted from the cost to the NRC because the NRC is the entitv consuming real resources (e.g., labor and capital) to meet its responsibilities. Any fees provided by licensees are viewed as transfer payments, and as such are not real costs from a societal perspective.
5.5.10 NRC Operation This attribute is an impact which measures the projected net economic effect on the NRC after the proposed action is implemented. Additional inspection, evaluation, or enforcement activities would be examples of such costs. Additional costs above the status quo are taken to be negative; cost savings are taken to be positive. As with industry operation costs, NRC operation costs generally occur over long periods of time and are sensitive to the assumed discount factor.
Here too, the NRC may seek compensation from the licensee to provide needed services; any compensation received should not be subtracted from the cost to the NRC.
5.5.11 Other Government This attribute is an impact which measures the net economic effect of the proposed action on the federal government (other than the NRC) and state and local governments resulting from the action's implementation or operation. Additional costs above the status quo are taken to be negative; cost savings are taken to be positive.
This attribute will be affected less often than some attributes, but can be teaterial in certain types of actions (e.g., changes to offsite emergency planning, provision of offsite services, and new requirements affecting Agreement States). The government entities may seek compensation from the licensee to provide the needed services; any compensation received should not be subtracted from the cost to the government units.
5.5.12 General Public This attribute is an impact which accounts for direct, out-of-pocket costs paid by members of the general putte as a result of implementation or operation of a proposed action. Examples of these costs could include items such as ircreased cleaning costs due to dust and construction-related pollutants, property value losses due to the action, or incanveniences (e.g., testing of evacuation sirens). Increases in costs from the status quo are taken to be negative; decreases in costs from the status quo are taken as positive.
This attribute is not related to the attribute associated with offsite property losses due to accidents. The general public attribute measures real costs that will be paid due to implemenation of the proposed action, subject to the uncenainties involved in estimation. These costs exclude taxes as they are simply transfer payments with no real resource commitment from a societal perspective. Any costs which are reimbursed by the applicant or licensee should be accounted for here and not duplicated under industry costs.
5.5.13 Improvements in Knowledge This attribute accounts for the potential value of new information, especially from assessments of the safety of licensee activities. Some NRC actions have as their goal the improvement in the state of knowledge for such factors as accident probabilities or consequences, with an ultimate objective of facilitating safety enhancement or reduction in uncertamty.
Quantitative measurement of improvements in knowledge depends largely on the type of action being investigated. The value of assessments directed at a fairly narrow problem (e.g., reducing the failure rate of a panicular component) may be NUREG/BR 0184 5.12
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- l l G l quantifiable in terms of safety or monetary equivalent. If this is the case, such values and impacts should be treated by other attributes and not included under this attribute. On the other hand, if potential values from the assessments are diffi-cult to identify or are otherwise not easily quantified, then they should be addressed under this attribute.
5.5.14 Regulatory Efficiency
- This attribute attempts to measure regulatory and compliance improvements resulting from the proposed action. These '
I may include changes in industry reporting requirements and the NRC's inspection and review efforts. Achieving consis-tency with international standards groups may also improve regulatory efficiency for both the NRC and the groups. This attribute is qualitative in nature.
l In some instances, changes in regulatory efficiency may be quantifiable, in which case the improvements should be l accounted for under other attributes, such as NRC implementation or industry operation. Regulatory efficiency actions that are not quantifiable should be addressed under this attribute.
5.5.15 Antitrust Considerations The NRC has a legislative mandate under the Atomic Energy Act to uphold U.S. antitrust laws. This qualitative attribute is included to account for antitrust considerations for those proposed actions that have the potential to allow violation of the antitrust laws.
If antitrust considerations are involved, and it is determined that antitrust laws could be violated, then the proposed action l must be reconsidered and, if necessary, redefined to preclude such violation. If antitrust laws would not be violated, then l5 i evaluation of the action may proceed based on other attributes. The decision as to whether antitrust laws could be violated
! must rely on a criterion of reasonable likelihood, since it is difficult to anticipate the consequences of a regulatory action with absolute certainty.
5.5.16 Safeguards and Security Considerations The NRC has a legislative mandate to maintain the common defense and security and to protect and safeguard national security information in its regulatory actions. This attribute includes such considerations.
In applying this attribute, it must be determined whether the existing level of safeguards and security is adequate and what effect the proposed action has on achieving an adequate level of safeguards atd security. If the effect of the proposed action on safeguards and security is quantifiable, then this effect should be included among the quantitative attributes.
Otherwise the contribution of the action will be evaluated in a qualitative way and treated under this attribute.
5.5.17 Environtnental Considerations Section 102(2) of the National Enviromnental Policy Act (NEPA) requires federal agencies to take various steps to l enhance environmental decision-making. NRC's procedures for implementing NEPA are set forth in 10 CFR Part 51.
Many of the NRC's regulatory actions are handled thmugh use of a generic or programmatic environmental impact state-ment (EIS), environmental assessment (EA), or categorical exclusion. If these vehicles are used, no further consideration is required in a regulatory analysis covering the same subject matter as the environmental document, although a summary of the most salient results of the environmental analysis should be included. Otherwise, an evaluation of the action with respect to its impact on the environment is required. Such an evaluation is usually handled separately from the value-impact analysis described in this Handbook. It could be the case that mitigation or other measures resulting from the Oh
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environmental review may result in cost increases that should be accounted for in the regulatory analysis. Alternatives examined in an EIS or EA should correspond as closely as possible to the alternatives exammed in the corresponding regulatory analysis.
5.5.18 Other Considerations The above set of attributes is believed to be reasonably comprehensive for most value-impact analyses. It is recognized that any panicular analysis may also identify attributes unique to itself. Any such attributes should be appropriately described and factored into the analysis.
5.6 Quantification of Change in Accident Frequency As expressed in this Handbook, the term " accident" should be viewed generally as an unplanned occurrence which potentially releases radioactive materials, applicable to both power teactor and non-reactor facilities. Discussions in this section assume familiarity with the concepts of risk as related to the nuclear industry, as well as knowledge of event- and fault-tree terminology. 'Ihe reader unfamiliar with these concepts or in need of review is directed to existing risk assessments or such standard references as the PRA Procedures Guide (NRC 1983a) and the Fault Tree Handbook (Vesely et al.1981). The NRC formally endorsed the use of PRA methods in nuclear regulatory activities with its issuance of a Final policy Statement in 1995 (NRC 1995b). The Policy Statement includes four elements, the first of which states that The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy.
SECY-95-079 contains a status update of NRC's PRA implementation plan. SECY-95-280 contains a framework for l
applying PRA in reactor regulation. As noted in Section 3, as this version of the Handbook was being completed a i number of NRC staff activities were underway which relate to PRA use in NRC regulatory activities. These include l
- completion of the staff's review oflicensee-submitted IPEs cvaluation of these IPEs for potential use in other regulatory activities, documented in a draft repon to be published as NUREG-1560 (NRC 1996b) e development of guidance on the use of PRA in plant-specific requests for license changes, including regulatory guides for use by licensees in preparing applications for changes and standard review plans for use by the NRC staffin reviewing proposed changes.
These activities should result in a more consistent and technically justified application of PRA in NRC's regulatory process. In particular, draft NUREG-1560 contains a detailed and explicit description of acceptable attributes of a quality PRA. The activities, along with staff work planned for FY 1997 to initiate improvemens to the economic models now used in NRC's offsite consequence analyses (e.g., the MACCS code), should have a significarn 1:npct on the PRA-related portions of this Handbook. Consequently, the discussion in this Handbook on the use of PRA and offsite consequence estimates should be viewed as interim guidance that may be relied upon until the Handbook is updated to accommad=* the NRC's new position on these regulatory issues. The staff expect to initiate this update as the preceding PRA guidance nears completion.
5.14
Value-Impact b
(
l Estimates of the change in accident frequency resulting from a proposed NRC action are based on the effects of the action
! on appropriate parameters in the accident " equation.*m Examples of these parameters might be system or component failure probabilities, including those for the facility's containment structure. The estimation process involves two steps:
- 1) identificat ir)n of the parameters affected by a proposed NRC action (see Section 5.6.1); and 2) estimation of the values of these affected parameters before and after the implementation of the action (see Section 5.6.2).
The parameter values are substituted in the accident equation to yield the base- and adjusted-case accident g=>
[ frequencies. The sum of their differences is the reduction in accident frequency due to the proposed NRC action.
t The process can 1,e viewed as follows. He frequency for accident sequence ij ism F, = { M,k n
where M, = the frequency of minimal cut set k for accident sequence i initiated by event J.
l A minimal cut set represents a unique combination of occurrences at lower levels in a structural hierarchy (e.g., compo-
! nent failures in power reactor systems) which produces an overall occurrence (e.g., reactor core <tamafe) at a higher level. ,
it takes the form of a product of these lower level occurrences. He affected parameters comprise one or more of the mul- l tiplicative terms in the minimal cut sets. Thus, the reduction in accident sequence ij's frequency is l
AF,= (F,), - (F,) .
'O =tn.u-n }
The reduction in accident frequency is the sum of the reductions for each affected accident sequence:
= ,{ ,{ { .
(M,( - (M,),,,,,)
Note that a negative reduction represents an increase in accident frequency from the base to the adjusted case (i.e., an increase resulting from the proposed action).
5.6.1 Identification of Affected Parameters The level of effort required to identify the parameters affected by implementation of an action depends primarily on the availability of one or more existing power reactor or non-reactor risk / reliability studies which include those parameters.
For nuclear power plants, Thble 5.2 provides a list of risk studies. The folkwing characteristics are included, as available:
a plant type (BWR/FWR and vendor) e external events inclusion (yes/no)
- year of commerelal operation
- program under which performed (if any)
- level of risk /seliability analysis *
- report reference v
5.15 NUREG/BR 0184 1
l l
l
Value-Impact
'Ihble 5.2 Nuclear power picnts risk assessments O
War Analysis External Plant Type Commercial IA,J' Events? Program References Brunswick-1/2 GE BWRs 1977/75 1 No Industry Apri!1988 Reviewed NUREG/CR-5465 l November 1989 Grand Gulf l GEBWR 1983 3 No NUREG-1150 NUREG/CR4550, V.6, September 1989
)
Brown et al.1990 1
Indian Ibint-2 W PWR 1974 3 %s Industry PASNY 1982 NRC Report NUREG/CR 1410 and 1411 August 1980 Reviewed NUREG/CR-2934, December 1982 i Reviewed NUREG/CR-0850, 1 November 1981 LaSalle GE BWR 1984 3 Ws Industry Call et al.1985 l County-1 RMIEP, NRC NUREG/CR4832, i 1992 and 1993 Peach GEBWR 1974 3 %s NUREG-1150 NUREG/CR4550, V.4, Bonom-2 August 1989 l (Also train level) Payne et al.1990 l I
Sequoyah-1 W PWR 1981 3 No NUREG-1150 NUREG/CR4550, V.5 l April 1990 l Gregory et al.1990 Surry-1 W PWR 1972 3 Ws NUREG-1150 NUREG/CR4550, V3, April 1990 Breeding et al.1990 Zion-1 W PWR 1973 3 No NUREG-Il50 NUREG/CR4550, V.7, May 1990 Park et al.1990 AP-600 W PWR *
- Reviewed by NRC 1992 System 80+
- Advanced reactor designs In addition to the studies shown in Table 5.2, IPE reports covering vulnerabilities to seven accidents and IPEEE reports can serve as additional references. Generic letter 88-20, issued in November 1988, required all holders of nuclear power plant operating licenses and construction permits to prepare IPE reports. Supplement 4 to General letter 88-20, issued in July 1991, requited these licensees to prepare IPEEE reports. IPE and IPEEE reports are available through the NRC NUREG/BR-0184 O
5.16
Value-Impact l f (v )/
Public Document Room. The status of the IPE and IPEEE programs is discussed in SECY-96-51 (NRC 1996a) and draft NUREG-1560 (NRC 1996b). NRC staff prepare an evaluation report documenting staff conclusions on each IPE and l IPEEE report submitted to NRC (NRC 1996a).
l l When evaluating generic power reactor issues, where many types of plants may be affected, the five risk assessments per-formed as part of the NUREG-1150 program (NRC 1991) are particularly useful. One of the primary objectives of that l
program was to " provide a set of (risk assessment) models and results that can suppon the ongoing prioritization of poten-tial safety issues and related research" (NRC 1991). As such, these provide a valuable resource for both quantitative and qualitative information on a set of five commercial nuclear power plants of different design.
Several computer codes containing reactor risk assessment information are also available which can be used to support regulatory analyses. Panicularly well suited to this type of analysis is the System Analysis and Risk Assessment (SARA) code (Stewart et al.1989), which contains the dominant accident sequences and cut sets for each of the NUREG-1150 plants. The Integrated Reliability and Risk Analysis System (IRRAS [ Russell and Sattison 1988]) is an integrated risk assessment software tool. Using this code, the analyst can create and analyze custom-made fault trees and event trees using a microcomputer.
In addition to these assessments of total plant risk / reliability, some studies focus on specific systems, accident initiators, or accident sequences. For certain actions, such specialized studies may be more appropriate for identifying affected parame-l ters than the various plant-wide assessments.
While risk / reliability assessments have been performed for selected non-reactor facilities, these are generally much less comprehensive than their power reactor counterpans. Available data for accident frequencies at non-reactor facilities have
! (fk been assembled into composite lists in Section C.2.1.1. They may be used as presented to identify affected parameters in a non-reactor accident equation, or as guides to the more detailed assessments from which they have been extracted.
l Additional information sources for non-reactor facility accidents may be found among the numerous Safety Analysis Repons conducted for U.S. Depanment of Energy (DOE) fuel-cycle facilities. For example, the DOE's Savannah River l
Site has roughly 30 such repons for fuel fabrication, chemical separation, research laboratories, analytical laboratories, waste handling, irradiated fuel storage, and radioactive material transponation.
At the simplest level, the standard analysis assumes that appropriate risk / reliability studies from which the affected parameters are easily identified are readily available. For example, all currently available reactor risk / reliability studies include accident sequences involving loss of emergency AC power. If the minimal cut sets used in the analytical modeling of these sequences contain parameters appropriate to an action related to loss of emergency AC power, then these risk / reliability studies (supplemented by any new studies published subsequent to this Handbook) would be appropriate for use in the standard analysis. The affected parameters can be readily identified, and the estimation of changes in accident frequency can proceed to the next step (parameter value estimation). Similarly, a major fire accident scenario has been investigated for most non-reactor facilities (see Section C.2.1.1). If a proposed action relates to reducing the fire potential at one or more types of non-reactor facilities, then these risk / reliability studies (supplemented by any new studies published subsequent to this Handbook) would be appropriate for use in the standard analysis. A useful source of data for l non-standard events at non-reactor facilities is that maintained at DOE's Savannah River Site (Durant et al.1988).
At a more detailed level, but still within the scope of a standard analysis, the identi5 cation of affected parameters may require more than direct use of existing risk / reliability studies. Existing studies may need to be modified without sacrific-ing their analytical consistency. The effon may involve performing an expanded or independent analysis of the accident sequences associated with an action, using previous studies only as a guideline, or several existing risk /relSbility studies may be combined to form some " composite" study more applicable to a generic action.
)
v 5.17 NUREG/BR-0184 i
Value-Impact O
Beyond the standard analysis lies the major effort, where identi6 cation of affected parameters requires the type of analysis associated with a much greater level of detail and, most likely, a significantly expanded scope. Typical of major efforts are NRC programs related to umesolved pour reactor safety issues. Such programs tend to be multi-year tasks con-ducted by one or more NRC contractors. Clearly, the expected degree of detail and quality of analysis made possible through a major effon to identify affected parameters should be " state-of-the-an,' significantly better ihan could be obtained from the standard effort.
5.6.2 Estimation of' Affected Parameter Values Presumably, the analyst has identified the parameters affected by action implementation. (If not, it is still possible to esti-mate changes in accident frequencies through expen opinion, discussed as pan of the standard analysis.) ne next step is to estimate the base- and adjusted-case frequencies / likelihoods of the affected parameters, which are then used to estimate the base- and adjusted-case total accident sequence frequencies. The sum of the differences between the base and adjusted cases is the reduction in accide . frequency resulting from the action (a negative reduction is an increase).
At the simplest level, the standard analysis assumes that frequencies / likelihoods for affected parameters are readily availa-ble or can be derived easily. The most convenient sources of data are the existing risk / reliability assessments; these pro-vide parameter frequencies / likelihoods in forms appropriate for accident frequency calculations (e.g., frequencies for initiators and unavailabilities or demand failure probabilities for subsequent system / component failures).
For power reactors, NUREG/CR-4639 (Gertman et al.1988) provides a Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR). Other data sources are available, including the Nuclear Plant Reliability Data System (NPRDS);* and the LERs. These may or may not repon data in the forms directly applicable as parameter frequencies /
likelihoods. For non-reactor facilities, failure rate data for non-reactor components are available from Dexter and Perkins (1982), Wilkinson et al. (1991), and Blanton and Eide (1993).
He derivation of frequencies / likelihoods from available data should require no more than standard statistical analysis tech-niques. In addition to statistics textbooks, other sources provide methods for deriving failure rates and probabilities more specifically for use in risk / reliability analyses. McCormick (1981)is a standard reference of this type. If derivation requires more detailed modeling, the analyst should consider the possibility of estimating frequencies / likelihoods through expert opinion. A formalized procedure like the Delphi tecimique may yield adequate estimates (Dalkey and Helmer 1963; Humphress and 12wis 1982). Also recommended are the " Formal Procedures for Elicitation of Expen Judgment,"
employed in the NUREG-1150 analyses (NRC 1991) and reviewed in Section 5.4.3.1.
Earlier, it was mentioned that an analyst unable to identify affected parameters for an action can still estimate changes in accident frequency. This removes the need for propagating the effect of change in individual risk parameters through the risk equation to obtain the accident frequency. It involves expen judgment of changes in accident frequency based on the total core-melt frequency of a representative nuclear power plant (although less applicable to the total radioactive release frequency for a non-reactor facility, see below). A formalized procedure like the Delphi method could be used to provide an overall consensus from expen estimates of percent changes in total accident frequency due to action implementation. l However, caution is advised, since direct estimation, as compared to more detailed calculations, can result in inaccurate i estimates.
Because of the nature of the radioactive material, its multiple locations, and near inconceivability of an accident capable of releasing the total inventory (except, possibly, an " external event"), estimating the frequency of total radioactive release from a non-reactor facility by expen judgment is difficult. It would be more realistic to use the experts to estimate fre-l quencies for individual release locations and initiators.
NUREG/BR-0184 5.18 O
Value-Impact
,a e i V)
Expen opinion may also play a prime role in estimating adjusted-case parameter values. Typically, existing data are applied to yield base-case values, leaving only engineering judgment for arriving at adjusted-case values. Consensus can reduce uncertainties, and the magnitudes of parameter values normally encountered in risk / reliability studies can serve as rough guidelines.
At a more detailed level, but still within the scope of a standard analysis, the analyst would be expected to conduct reason-ably detailed statistical modeling or extensive data compilation when frequencies / likelihoods for affected parameters are not readily available. While existing risk / reliability assessments may provide some data for use in statistical modeling, the level of detail required would normally be greater than they could provide. Statistical modeling may use stochastic simula- l tion methods and may also involve relatively basic statistical analysis techniques using extensive data. l l
Beyond the standard analysis lies the major effort, where estimation of affected parameter values requires much greater )
detail and a significantly expanded scope. When frequencies / likelihoods are unavailable for affected parameters, a major analytical effort is required. The analyst may need to develop specialized statistical models or possibly seek experimental data. On the other hand, data may be so abundant as to require extensive statistical analysis to produce a more workable base. Typically, both detailed statistical modeling and extensive data compilation will be required as part of a major effort. " State.of-the-an* data analysis techniques should be employed.
Estimation of adjusted-case affected parameter values should involve more than just expen opinion for a major effort.
Engineering judgment can be incorporated into an overall framework, but this framework should be analytical, not judgmental. If the need for expert opinion proves inevitable, only a rigorous application of the Delphi or other such l methods will suffice for a major ellort. l l
l
) 5.6.3 Change in Accident Frequency The change in accident frequency is a key factor for several of the value-impact analysis attributes. Having identified base-and adjusted-case values for the parameters in the plant risk equation affected by the proposed regulatory action, the ana-lyst proceeds to calculato the reduction in accident frequency as the sum of the differences between the base- and adjusted-case values far all affecte:1 accident sequences. Section 5.6 presented this calculation in the format of an equation.
Reduction in accident frequexy is algebraically positive; increase is negative (viewed as a negative reduction).
An error factor
- of at least five (typical for a 90% confidence level) on the best estimate of the reduction in total acci-dent frequency may be used to estimate high and low values for the sensitivity calculations in a standard analysis for power reactor facilities. If no better information is available, higher error factors (at least 10) can be assumed for non-reactor standard analyses. If better values are known (e.g., error factors from the specific risk assessment being used), they should be employed. Rigorously derived error factors via computer simulation would be appropriate for a major analysis beyond the standard scope.
NUREG/CR-2800 (Andirws et al.1983) provides a useful conceptual discussion on the calculation of change in core-melt accident frequency for power reactors, along with detailed examples. Such calculations would be typical of what is expected to be appropriate in the standard value-impact analysis portion of a regulatory analysis.
The FORECAST computer code for regulatory effects cost analysis (Impez and Sciacca 1996) allows input for the change in accident frequency.
A
/ \
(v/
5.19 NUREG/BR-0184
Value-Impact O
5.7 Quantification of Attributes The following sections provide specrk guidance in atimating the values of each attribute. However, before looking at specific attributes, them are several genenooncepts that need to be explored.
%!ue and impact estimates are performed relative M a baseline case, which is typically the no-action alternative. In estab-11shing the baseline case, an assumption should be made that all existing NRC and Agreement State requirements and written licensee commitments are already being implemerted and that values and impacts associated with these require-ments are not part of the incremental estimates prepared for the regulatory analysis. Similarly, the effects of formally proposed concurrent regulatory actions that are viewed as having a high likelihood of implementation need to be incorporated into the baseline before calculating the incremental consequences of the regulatory action under consideration.
The treatment of volunta:y incentives on the part of industry also has important implications on the baseline and therefore, the incremental consequences of the proposed action. Section 4.3 of the NRC Guidelines discusses the treatment of voluntary activities by affected licensees when establishing a baseline reference. Basically, analysts should give no credit for voluntary actions in makmg base case estimates. However, for completeness and sensitivity analysis purposes, the analyst should also display results with credit being given for voluntary actions by licensees.
Section 4.3 of the NRC Guidelines requires the use of best estimates. Often these are evaluated in terms of the mean or
" expected value," the product of the probability of some event occurring and the consequences which would occur assum-ing the event actually happens. Sometimes, measures other than the expected value may be appropriate, such as the median or even a point estimate. However, the expected value is generally preferred.
There are four attributes used in value-impact analysis for which expected value is normally calculated: public health (acci-dent), occupational health (accident), offsite property, and onsite property. All four of these attributes usually rely on esti-mation of the change in probability of occurrence of an accident as a result of implementation of the proposed action.
(Changes in the consequence of the accident [i.e., dose or cost) would also affect these attributes.)
Four attributes involve radiation exposure: 1) public health (accident), 2) public health (routine), 3) occupational health .
(accident), and 4) occupational health (routine). In quantifying each measure, the analyst should assess the reduction (or j risk averted) relative to the existing condition. For accident-related exposures, the measure will be probabilistically weighted (i.e., the potential consequence is multiplied by its pmbability of occurrence).m The non-accident terms (e.g.,
)
l routine occupatied exposure) are given in terms of annual expected effect. Both types of terms would be integrated over the lifetime of the affected facilities to show the total effect. Each of the attributes involving radiation exposure can be characterized in terms of person-rems, either avened by or resulting from implementation of the proposed action.
The four attributes associated with radiation exposure require a person-rem-to<!ollars conversion factor to be expressed monetarily (see Section 5.7.1.2). The remaining quantitative attributes are normally quantified monetarily in a direct manner. When quantified monetarily, attributes should be discounted to present value (see Section B.2 for a discussion of discounting techniques). This operation involves an assumption regarding the remaining lifetime of a facility. If appropriate, the effect of license renewal should be included in the facility lifetime estimate (see Section 4.3 of the Guidelines). The total dollar figures capture both the number of facilities invohed (in the case of generic rulemaking) and the economic lifetime of the affected facilities.
5.20
l Value-Impact d
l O \
L Based on OMB's guidance in Circular A-94, Section 4.3.3 of the Guidelines requires that a 7 % real (i.e., inflation- )
adjus:ed) discount rate be used for a best estimate. For sensitivity analysis, the Guidelines recommend a 3 % discount rate.
However, for certain regulatory actions involving a timeframe exceeding 100 years (e.g., decommissioning and waste dis-posal issues), Section 4.3.3 of the Guidelines stipulates the following:
I
. ..[T]he regulatory analysis should display results to the decision-maker in two ways. First . on a present worth basis using a 3 percent real rate, and second, by displaying the values and impacts at the time in which they are incurred with no present worth conversion. In this latter case, no calculation of the resulting net value... should be made.
l
' Qualitative" attributes do not lend themselves to quantification. To the degree to which the considerations associated with these attributes can be quantified, they should be; the quantification should be documented, preferably under one or more of the quantitative attributes. However, if the consideration does not lend itself to any level of quantification, then its treatment should take the form of a qualitative evaluation in which the analyst describes as clearly and concisely as possi-ble the precise effect of the proposed action.
To estimate values for the accident-related attributes in a regulatory analysis, the analyst ideally can draw from detailed risk / reliability assessments or statistically-based analyses. Numerous sources exist for power reactor applications (e.g., j see Section 5.6). To a lesser extent, Sections C.3-C.6 and C.10 provide similar data for non-reactor applications. Most l regulatory analyses for power reactor facilities are based on detailed risk / reliability assessments or equivalent statistically based analyses.
However, the analyst will sometimes find limited factual data or information sufficiently applicable only for providing a 1
( quantitative perspective, possibly requiring extrapolation. These may often involve non-reactor licensees since detailed
\s risk / reliability assessments and/or statistically-based analyses are less available than for power reactor licensees. Two .
examples illustrate this type of quantitative evaluation.
In 1992, the NRC performed a regulatory analysis for the adoption of a proposed rule (57 FR 56287; November 27,1992) concerning air gaps to avert radiation exposure resulting from NRC-licensed users of industrial gauges. The NRC found insufficient data to determine the averted radiation exposure. To estimate the reduction in radiation exposure should the rule be adopted, the NRC proceeded as follows. The NRC assumed a source strength of one curie for a device with a large air gap, which produces 1.3 rem /hr at a distance of 20 inches from a Cs-137 soutec. Assuming half this dose rate would be produced, on average, in the air gap, and that a worker is within the air gap for four hours annually, the NRC estimated the worker would receive 2.6 rem /yr. The NRC estimated that adopting the proposed air-gap rule would be i
cost-effective if 347 person-rem /yr were saved. At the estimated atrage savings of 2.6 person-rem /yr for each gauge licensee, incidents involving at least 133 gauges would have to be eliminated. Given the roughly 3,000 gauges currently
,; used by these licensees, the proposed rule would only have to reduce the incident rate by roughly 4 %, a value the NRC believed to be easily achievable. As a result, the NRC staff recommended adoption of the air-gap rule.
In 1992, the NRC responded to a petition from General Electric (GE) and Westinghouse for a rulemaking to allow self-guarantee as an additional means for compliance with decommissioning regulations. An NRC contractor estimated the 1 default risks of various types of financial assurance mechanisms, including the proposed self-guarantee. The contractor had to collect data on failure rates both of firms of diffetent sizes and of banks, savings and loans, and other suppliers of financial assurance mechanisms. The contractor estimated a default risk of 0.13% annually for the GE-Westinghouse l proposal, with a maximum default risk of only 0.055 % annually for third-party guarantors, specifically a small savings and loan issuing a letter of credit. Based on these findings, the NRC initiated a proposed rulemaking which would allow self-guarantee for certain licensees. The final rule was issued December 29,1993 (58 FR 68726).
f
\
Additional examples of this more limited type of quantitative approach to estimation can be found in Sections C.8 and C.9.
b 5.21 NUREG/BR-0184
Value-Impact O
5.7.1 Public Health (Accident)
Evaluating the effect on public health from a change in accident frequency due to proposed regulatory actions in a multi-step process. For each affected facility, the analyst first estimates the change in the public health (accident) risk associated with the action and reports this as person-rem avoided exposure. Reduction in public risk is algebraically positive; increase is negative (viewed as a negative reduction). Next the analyst convens person-rems to their monetary equivalent (dollars) and discounts to present value. Finally, the analyst totals the change in public health (accident) as expressed in discounted dollars over all affected facilities.
'The steps are as follows:
- 1. Estimate reduction in accident frequency per facility (see Section 5.6).
- 2. Estimate reduction in public health (accident) risk per facility (see Section 5.7.1.1).
- 3. Convert value of public health (accident) risk avoided (person-rems) per facility to monetary equivalent (dollars) via monetary valuation of health effects (see Section 5.7.1.2).
Z,u = RD, where Z,a4 = monetary value of public health (accident) risk avoided per facility-year before discounting ($/ facility-year)
Da = avoided public dose per facility-year (person-rem / facility-year)
R = monetary equivalent of unit dose ($/ person-rem).
- 4. Discount to present value per facility (dollars)(see Section 5.7.13).
- 5. Total over all affected facilities (dollars).
V,g, = W pg, where V,u, = discounted monetary value of public health (accident) risk avoided for all affected facilities ($)
W,w4 = monetary value of public health (accident) risk avoided per facility after discounting ($/ facility)
N = number of affected facilities.
If individual facility values rather than generic values are used, the formulations can be replaced with V,a4 =
N,W,,,
where i = facility (or group of facilities)index.
5.7.1.1 Estimation of Accident-Related IIealth Effects The results of the formulations given la Section 5.6 are reductions in accident frequency. These form the first portion of the public health (accident) risk estimate. For the standard analysis, the analyst would employ data developed in existing risk studies which include offsite effects, if possible. Such studies provide population dose factors that can be applied to accident release categories to yield dose estimates as follows:
NUREG/BR-0184 5.22 O
l
. . - _ , _ _ _ . . _ _ ._ _ . . _ _ . . _ _ . _ . . ~
4 Value-Impact J
Avoided Public Population Dose
' Reduction in Release 1 F o ease 4
Dose (person-rem / ihty-yr)
=
{
8*==
Category Frequency ' x (events / facility-yr)
%, (person-rem / event) l If the risk assessment being used by the analyst to estimate public health (accident) employs its own unique accident
- release categories with corresponding population dose factors, then these should be used. Otherwise, population dose fac-tors should be based on Thble 5.3 (see Appendix B.4 for development of this table). For non-reactor accidents, population dose factors for accident scenarios at selected facilities have been assembled into composite lists in Section C.2.1.2. An i error factor of at least five is considered appropriate for use in sensitivity studies.
- Thble 5.3 Expected population doses for power rear.or release categories ,
AccVist Progression Charo:seristics Population Dese Plant Release Categorf CF *thne PDS SP Bypen M GM TYPE gyp,,, (p,,,,,,,,,,) ,Ihrum RSURI CFatVB Rupture 6.15E+ 6 63
! RSUR2 Law CF IDSP leak 2.30E + 6 88 RSUR3 No CF No CF 2.50E +2 67 '
{
I RSUR4 Bypass Bypass Bypass 4.29E + 6 80
- RZ1 CFatVB Shallow Ruptune 5.77E+6 65 i RZ2 LamCF LOCA leak 1.31E+5 38 PWR RZ3 No CF No CF 3.31E + 2 67 A aW
- RZ4 Byp, ss Bypass Dry Bypass 4.80E+6 76 RSEQ1 CFdus "D 1.31E+7 50 i LOSP Dry CatRup j RSEQ2 CFat(B 5.77E+6 56 RSEQ3 Las CF Rupture 1.33E+5 42 Flooded -
RSEQ4 No CF LOCA No CP 4.06E + 2 71 RSEQS Bypass Bypass Dry Bypass 4.94E + 6 76 i BWR RPB1 LOSP 5.25E+6
! CF3tVB Early/Las Sm/None IMMth 80 1 RPB2 5.32E+6 4
ATWS Dry RPB3 CFdurCD None WWvent 3.26E+6 84 J
RPB4 Law CF Early/ Lam IMrup 1.13.G+ 6 92 RPB5 No CP LDSP None Sm/None Shallow NoCF 8.27E+3
- 62 l RPB6 CFatVB Early/Lau Large Dry DWMth 1.IIE+7 RLASI CFdurCD ' Dan Earty/ Law Sm/None Dry WMwrup 5.25E+6 80 A RLAS2 WMw-Ik 3.21E + 6 81 1 CFatVB Shallow RLAS3 DWrup 4.66E + 6 82 1
RLAS4 CPdurCD Dry 5.92E+6 73 RLAS5 Sm/None Shallow 1.75E+6 82 RLAS6 Large Dry CF-Ped 4.18E+6 73 l
j 5.23 NUREG/BR-0184
l Value-Impact Table 5.3 (Continued)
Accident Progression Characteristics Ibpulation Dose Dpe Category CF Time PDS SP Bypass CCI CF Mode
- 8 Bypass (hrson-Rem) Term BWR RLAS7 No CF l ' Dan None Sm/None Shallow No CF 3.33E + 2 65 RGG1 CFatVB Early/ Law 5,77E + 6 75 RGG2 CFdurCD None Flooded 2.74E + 6 90 RGG3 Law CF FISB Lale Only large 2.35E + 6 80 RGG4 CFdurCD Early/ Late 2.70E + 6 93 RGG5 No CF None No CF 1.lSE + 2 59 f*ne; The inhais R$UR, RZ, and R$EQ whr so Surry, Zum, and Sequoyah reasone conscres rapecimely bliomed by em release categwy number.
The latchais EPB, RLA3, and R00 nhr to hach Bottom. LaSalle, and Grand Gulf rolasse causarms respecdvely blioned by the reisene causary aumber.
Key-CF Time = Comanenent hilwe (CF ume)
CFstVB = CF at vesselineach (VBI CFdwCD = CP durms core damage (betro VB, ifit occurs)
LamCF = CF durms core comsmrmen 4meractions (CCD No CF = ao CF Bypass = bypau of comammem (usually droughout duraten of socadem)
LO$P = loss of offsin power LOCA = loss of coulam arcident Bypass a bypass of comalmnem (inwhcmg synens LOCA or suam generator tube rupture)
ATW5 = amicipsed Wansom withnut scram han = Trarmem STSB = short-orid sailon blackout CCI = The of malma core concrem imoracemns (CCl)
Dry = CCI oucws in a dry cavity Shallow = CCI axws in a met covtry (mmnemelly 5 P.. of mater)
Flanded = CCI occws in a dupiy Souded cavity (nommally 14 h. of water)
No CCI = There o no CCI(the debra bed is coolable with splemshable waar or no VB)
CF Mads = Conummem hilwe mods Ca:Rup
- Caastrophe ruptwo hilure Rupture = Ruptwo hilure of comamment Bypass = bypass of conminmes Laak = Laak kilwa of comammes No CF = no CF Whr@ = Rupture ebove the wetwell user level Wh Ik = Laek above de netwell waar level DWRup = Rgews in the drywell WWvant = %nting of the meteell CF-Ped = Rupture in the drywell wall, cawed by las hilwe of de reactor pedssmi DWMth = Met:4hrough of the drywell well by direct conmet of the moires ccre SP Bypass = Suppressen pool (SP) bypass Early% = SP is bypassed dram the time of VB throughout the accidem Nous = SP h never bypassed Lam Only = SP h ordy bypassed las in the accides (during CCI)
RB Bypass = Reactur building (RB) bypass Sm/None = Nominal or small lentage Wom t%s RB Large = Large leakags kom the RB w bypass of the kB (tr Grand Gulf, all comainmem hilwes were assumed to be abow the RB)
Should the nature of the issue require that the reduction in accident frequency be expressed as a single number, a single population dose factor, preferably one that has been pmbabilistically weighted to reflect those for all accident release cate-l gories, is generally needed. For this approach, the calculation of avoided public dose becomes:
Avoided Public Reduction in Population Dose
= Accident Fn:quency x Factor Dose DQ (person-rem / cihty-yr) , (events / facility-yr) , ,(person-rem / event)
5.24
i
+
Value-impact l I j
)
Mubayi et al. (1995) have calculated population doses weighted by the frequencies of the accident release categories for ;
the five power reactors analyzed in NUREG-1150 (NRC 1991). 'Ihese air listed in 'Ihble 5.4 based on Version 1.5.11.1 ;
of the MACCS computer code (Chanin et al.1993). The population doses have been calculated as the sum of those for .
emergency response and long-term pmtective action, defmed as follows- !
I
- For early consequences, an effective emergency response plan consisted of evacuation of all but 0.5 % of the population within a ten-mile radius at a specified speed and delay time following notification of the emergency.
- For long-term relocation and banning of agricultural products, the interdiction criterion was 4 rem to an individual over ;
five years (2 rem in year one, followed by 0.5 rem each successive year).
l i
For segulatory analyses involving nuclear power plants, doses should be estimated over a 50-mile radies from the plant !
site (see Guidelines Section 4.3.1). Doses for other distances can be considered in sensitivity analyses or special cases, {
and are available in Mubayi et al. (1995). :
It is possible that the proposed action will affect public health (accident) through a mitigation of consequences instead of l (or as well as) through a reduction in accident frequency.* Should this be the case, the previous general formulations are replaced with the following: ,
i l
Avoided Public , p Release Category , Category Population Dose m Frequency Dose Factor s, j cm, . ,
_p Release Category g Category Pbpulation' O %
cm,,.
Frequency Dose Factor ,g y-or Avoided Public , ' Accident g Population Dose' ,
' Accident g Population Dose' 1)cae Frequency Factor y Frequency Factor g Thble 5.4 Weighted population dose factors for the five NUREG-1150 power reactors j 1%rson-rem Within 50 miles Reactor Type from the Plant Zion PWR 1.95E+5 Surry PWR 1.60E+5 Sequoyah PWR 2.46E+5 1%ach Bottom BWR 2.00E+ 6 Grand Gulf BWR 1.931s+5 Average 1.99E+5 5.25 NUREG/BR-0184
Value-Impact O
Public risks from non-reactor accidents have been assembled into composite lists in Section C.2.1.3. These represent the products of accident frequencies and population dose factors, whether calculated as release category summations or single frequency and dose numbers.
Beyond the standant analysis lies the major effort. In parallel with the more involved effort to identify and quantify affected parameters in appropriate accident sequences (see Section 5.6) would be an equivalent effort to quantify popula-tion dose factors and possibly even specific health effects. Such effort at the " consequence end" of the risk calculation would increase the likelihood of obtaining representative results. Non-representative results can arise through the use of inappropriate or inapplicable dose calculations just as readily as through inappropriate logic models and failure data.
Several computer codes exist for estimation of population dose. Most for reactor applications have been combined under MACCS (Chanin et al. 1990,1993; Summers et al.1995a,b). Three codes for non-reactor applications are GENII (Napier et al.1988), CAP-88 (Beres 1990), and COMPLY (EPA 1989). There have a'so been recent upgrades to MELCOR itself for modeling severe accidents in light water reactors, including estimation of severe accident source terms and their sensitivities / uncertainties (Summers et al.1995a,b).
The GENIl code package determines individual and population radiation doses on an annual basis, as dose commitments, and as accumulated fium acute or chronic radionuclide releases to air or water. It has an additional capability to predict very-long-term doses from waste management operations for periods up to 10,000 years.
The CAP-88 code package is generally required for use at DOE facilities to demonstrate compliance with radionuclide air r: mission standards where the maximally exposed offsite individual is more than 3 km from the source [40 CFR 61.93(a)].
The code contains modules to estimate dose and risk to individuals and populations from radionuclides released to the air.
It comes with a library of radionuclide-specific data and provides the most flexibility of the EPA air compliance codes in terms of ability to input site-specific data. A personal computer version of the CAP-88 code package (Ptrks 1992) was released in March 1992 under the name CAP 88-PC and is also approved for demonstrating compliance at DOE facilities.
The COMPLY code is a screening model intended primarily for use by NRC licensees and federal agencies other than DOE facilities. It is approved for use by DOE facilities where the maximally exposed offsite individual is less than 3 km from the emissions source [40 CFR 61.93(a)]. The code consists of four screening levels, each of which requires increasingly detailed site-specific data to produce a more realistic (and less conservative) dose estimate. COMPLY runs on a personal computer and does not require extensive site-specific data.
5.7.1.2 Monetary Wluation of Accident-Related Ilealth Effects Section 4.3.3 of the Guidelines states that the conversion factor to be used to establish the monetary value of a unit of radiation exposure is $2000 per person-rem. This value will be subject to periodic NRC review. The basis for selection of the $2000 per person-rem value is set out in NUREG-1530 (NRC 1995d). The $2000 per person-rem value is to be used for routine and accidental emissions for both public and occupational exposure. Unlike past NRC practice, offsite property consequences are to be separately valued and are not part of the $2000 per person-rem value. Monetary conversion of radiation exposure using the $2000 per person-rem value is to be performed for the year in which the exposure occurs and then discounted to present value for purposes of evaluating values and impacts.
5.7.1.3 Discounting Monetized Value of Accident-Related IIcalth Effects The present value for accident-related health effects in their monetized form can be calculated as follows:
W,=CxZ pa pna NUREG/BR-0184 O
5.26
Value-Impact -
O where W = monetary value of public health (accident) risk avoided per facility after discounting ($/ facility)
C = [exp(-rt;) - exp(-rtf )]/r tr = years remaining until end of facility life ;
ti = years before facility begins operating l r = real discount rate (as fraction, not percent) 1 Z = monetary value of public health (accident) risk avoided per facility-year before discounting l
. ($/ facility-year). ]
If a facility is already operating, t will be zero and the equation for C simplifies to C= 1 - exp(-rt,) /r Should public health (accident) risk not be discounted in an analysis, r effectively becomes zero in the preceding equations.
. In the limit as r approaches zero, C = tr- t;(or, C = t when f ti = 0). This new value of C should be used to evaluate
, Wm in the undiscounted case. ]
The quantity Wm must be interpreted carefully to avoid misunderstandings. It does not represent the expected reduction in public health (accident) risk due to a single accident. Rather, it is the present value of a stream of potential losses ;
l extending over the remaining lifetime of the facility. Thus, it reflects the expected annual loss due to a single accident (this is given by the quantity Zm); the possibility that such an accident could occur, with some small probability, at any time over the remaining facility life; and the effects of discounting these potential future losses to present value. Since the
, pj
- t quantity Z only accounts for the risk of an accident in a representative year, the result is the expected loss over the facility life, discounted to present value.
V i The FORECAST computer code for regulatory effects cost analysis (Lopez and Sciacca 1996) allows input for the public health (accident) attribute.
- 5.7.2 Public Health (Routine) )
! 1 i
As with the public health (accident), the evaluation of the effect on public health from a change in routine exposure due to i proposed regulatory actions is a multi-step process. Reduction in exposure is algebraically positive; increase is negative (viewed as a negative reduction).
The steps are as follows:
- 1. Estimate reductions in public health (routine) risk per facility for implementation (D,u) and operation (Dno) (see Section 5.7.2.1).
- 2. Convert each reduction in public health (routine) risk per facility from person-rems to dollars via monetary evaluation of health effects (see Section 5.7.2.2):
G,u = RD,u G,,o - RD,o l
where G,u = monetary value of per-facility reduction in routine public dose required to implement the proposed action, before discounting ($/ facility) t 5.27 NUREG/BR-0184
Value-impact O
Gno = monetary value of annual per-facility reduction in routine public dose to operate following implementation of the proposed action, before discounting ($/ facility-year)
Dn " Per-facility reduction in routine public dose required to implement the proposed action (person-
. rem / facility)
D,,,o = annual per-facility reduction in routine public dose to operate following implementation of the proposed action (person-rem / facility-year)
R = monetary equivalent of unit dose ($/ person-rem).
- 3. Discount each reduction in public health (routine) risk per facility (dollars) [see Section B.2].
- 4. Sum the reductions and total over all facilities (dollars):
Vn, = N(Hm+Hyo) where V ma = discounted monetary value of reduction in public health (routine) risk for all affected facilities ($)
Hm = monetary value of per-facility reduction in mutine public dose required to implement the proposed action, after discounting ($/ facility)
Hno = monetary value of per-facility reduction in mutine public dose to operate following implementation of the proposed action, after discounting ($/ facility)
N = number of affected facilities.
Note the algebraic signs for Dm and Dno. A reduction in exposure is positive; an increase is negative. The dose for implementation (Dm) would normally be an increase and therefore negative.
If individual facility values rather than generic values are used, the formulations can be replaced with V,=
n N, (Hm, + Hyo) where i = facility (or group of facilities)index.
5.7.2.1 Estimation of Change in Routine Exposure A proposed NRC action can affect routine public exposures in two ways. It may cause a one-time increase in routine dose due to implementation of the action (e.g., installing a retrofit). It may also cause a change (either increase or decrease) in the recurring routine exposures after the action is ime'. -- i? For the standard analysis, the analyst may attempt to make exposure estimates, or obtain at least a sampit st Austry or community data for a validation of the estimates devel-oped. Baker (1995) provides estimates of populatica and individual dose commitments for reported radionuclide releases from commercial power reactors operated during 1991. Tichler et al. (1995) have compiled and reported teleases of radioactive rnaterials in airborne and liquid effluents from commercial Light Water Reactors (LWRs) during 1993. Data on solid waste shipments are also included. ' Itis report is updated annually. Routine public risks for non-reactor facilities have been assembled into composite lists in Section C.2.2.
5.7.2.2 Monetary %luation of Routine Exposure As with public health (accident) (Section 5.7.1.2), monetary valuation for public health (routine) employs the value of
$2,000! person-rem as the best estimate of the monetary conversion factor (R).
NUREG/BR-0184 5.28
Value-Impact p
(m)
The FORECAST computer code for regulatory effects cost analysis (Impez and Sciacca 1996) allows input for the public health (routine) attribute.
5.7.3 Occupational Health (Accident)
Evaluating the effect on occupational health from a change in accident frequency due to proposed regulatory actions is a multi-step process. Reduction in occupational risk is algebraically positive; increase is negative (viewed as a negative reduction).
The steps are as follows:
- 1. Estimate reduction in accident frequency per facility (see Section 5.6). j l
- 2. Estimate reduction in occupational health (accident) risk per facility due to the following (see Section 5.7.3.1)-
l
- "immediate" doses l
- long-term doses
- 3. Per facility, convert value of occupational hedh (accident) risk avoided (person-rems) to monetary equivalent (dollars) via monetary evaluation of health effew, ac to the following (see Section 5.7.3.2):
(N * "immediate" doses Zo i = RY io
(
k/
)
- long-term doses Zm = RYm I
where Z i o = monetary value of occupational health (accident) risk avoided per fL-ility-year due to "immediate" doses, before discounting ($/ facility-year) l Zm = monetary value of occupational health (accident) risk avoided per facility-year due to long-term doses, l before discounting ($/ facility-year) ;
Y io = avoided occupational "immediate" dose per facility-year (person-rem / facility-year)
Ym = avoided occupational long-term dose per facility-year (person-rem / facility-year)
R = tnonetary equivalent of unit dose ($/ person-rem).
- 4. Discount to present value per facility (dollars) (see Section 5.7.3.3).
- 5. Total over all affected facilities (dollars) using Von, = N (W,o + Wm) where Vowa = discounted monetary value of occupational health (accident) risk avoided for all affected facilities W io = monetary value of occupational health (accident) risk avoided per facility due to "immediate" doses, atter discounting ($/ facility)
Wm = monetary value of occupational health (accident) risk avoided per facility due to long-term doses, after discounting ($/ facility)
N = number of affected facilities, f3 I l v
5.29 NUREG/BR-0184
Value-Impact If individual facility values rather than generic values are used, the formulations can be replaced with l
V. = N(W , + WM l
where i = facility (or group of facilities) index.
5.7.3.1 Estimation of Accident-Related Exposures There are two types of occupational exposuit related to accidents: "immediate" and long-term. The first occurs at the j time of the accident and during the immediate management of the emergency. The second is a long-term exposure, l presumably at significantly lower individual rates, associated with the cleanup and refurbishment or decommissioning of the damaged facility. The value gained in the avoidance of both types of exposure must be conditioned on the change in frequency of the accident's occurrence (see Section 5.6).*
"Immediate" Doses
{
Licensing of nuclear facilities requires the licerise applicant to consider and attempt to minimize occupational doses.
Radiation protection in a reactor control ror,m is required to limit dose to 5-rem whole body under accident conditions (10 CFR 50, Appendix A, Criterion 19). The experience at the Three Mile Island (TMI) Unit 2 nuclear power plant 1 indicated that potential for significant oenipational exposures exists for activities outside the contml room during a power l reactor accident. (However, there was nc individual occupational exposure exceeding 5-rem whole body at TMI-2.) ,
i For the standard analysis specifically applied only to power reactor facilities, the analyst may employ the TMI or i Chernobyl experience. At TMI, the average occupational exposure related to the incident was approximately I rem. A collective dose of 1,000 person-rem could be attributed to the accident. This occurred over a 4-nonth span, after which !
time occupatiotal exposure was approaching pre-accident levels. An upper estimate for sensitivity analysis is obtained by assuming that the average individual receives a dose equal to that of the maximum individual dose at TMI. The ratio of maximum to average dose for TMI is 4.2 rem /l rem; therefore, the upper estimate for the collective dose can be taken as 4,200 person-rem. A lower estimate of zero indicates a case where no increase over the normal occupational dose occurs The DOE (1987) summarized results on the collective dose received by the populace surrounding the Chernobyl accident.
Average dose equivalents of 3.3 rem / person,45 rem / person, and 5.3 rem / person were estimated for residents within 3 km, between 3 km and 15 km, and between 15 km and 30 km of Chernobyl, respectively (Mubayi et al.1995, p. A-5).
Although none of these translates readily into an occupational dose as that for TMI, a reasonable, but conservative, assumption would be that the average worker received the average dose for persons closest to the plant (i.e.,
3.3 rem / person). For 1,000 workers, an average value of 3,300 person-rem is obtained, about three times that estimated for TMI-2. Given the greater severity of the Chernobyl accident, this seems reasonable. Using TMI's ratio of 4.2/1 for the maximum, an upper bound of 14,000 person-rem results. TMI's average value of 1,000 person-rem would appear to be a reasonable lower bound for Chernobyl.
Given the uncertainties in existing data and variability in severe accident parameters and worker response, the following is suggested as Do (immediate occupational dose) specifically for power reactor accidents:
Best estimate: 3,300 person-rem High estimate: 14,000 person-rem Low estimate: 1,000 person-rem >
NUREG/BR-0184 5.30 0
.__m._. _ - _ . _ _ _ _ . _ _ _ m _ __ _ _ _ _ _ _ _ _ _ . . _ _ _ . . . _ . . . _ _ _ . . . _ . _ . _ __
l 1-Value-impact j.
i
) l
!= . i l For a mailor effort beyond the standard analysis, specinc calculations to estimate onsite exposures br various accidents !
i could be performed. j Imag'ibrus Doses
- After the inimadiaw response to a major power reactor accident, a long process of cleanup and refurbishment or decom-
- missioni 4 will follow. Signi6 cant r-via==1 dose will result (individual exposures controlled by normal occupational 4 dose guidelines). The values for the standard analysis speci6cally applied only to power reactors are based on a study
. (Murphy and Holter 1982) ri fecommissioning a reference LWR followag postulated accidents. Thble 5.5 sn==arians the occupational doses esticuwd oy the study and is presented for perspectiw, j
! Since this Handbook focuses on avoidance of major large-scale accidents, use of the following long-term doses based on
- - Murphy and Holter (1982) is suggested speci6cally for power reactor accidents.
l Duo (long-term occupational):
i Best estimate: 20,000 person-rem High estimate: 30,000 person-rem j Iow estimate: 10,000 person-rem
- 'thble 5.5 Rati== sad1 -:- , ^" ' radiaties dose fhmm cisamep and staaa====la=la=ing after a power l reactor accident (person-ruun or persosH:ST)
Accident Accident Accident '
i l Activity Seemarlo 1" Seemario 2" Seemarlo 3" i ,
i
! Cleanup 670 4,580 12,100 Diamantlement and Decommissioning L23Q LQGQ LifG s-lbtal 1,900 7,640 19,760 i.
(a) Accident Scenario 1 - a small IAss of Coolant Accident (IhCA) in which F-- y Core
{~
Cooling System (ECCS) functions as intended. Some fuel :ladding ruptures, but no fuel melts. The containmant building is moderately mataminanad, but there is minimal physical l
- damage.
(b) Accident Scenario 2 - a small LOCA in which ECCS is delayed. Fifty percent of the fuel l
claddag ruptures, and some fuel melts. The matainmant building is extensively contaminated, but there is minimal physical damage. (This scenario is presumed to simulate the TMI-2 accident.)
l (c) Accident Scenario 3 - a major IDCA in which ECCS is delayed. ' All fuel claddmg ruptures, i and there is signi6 cant fuel melting and core damage.1he matainmant building is
', extensively mataminatad and physically damaged. The auxiliary building undergoes some contamination.
i l
1 l 5.31 NUREG/BR-0184 i
I _ ._ _, _ __ -
Value-Impact A mided D9ses e l l
Tb calcilate the avoided accident-related occupational exposures, both the "immediate" and long-term occupational dose are muPiplied by the reduction in accident frequency (see Section 5.6) which is postulated as a result of the proposed action.
Yo - AF D o Ym - AF Dm l where AF = reduction in accident frequency (events / facility-year)
Yo = avoided occupational "immediate" dose per facility-year (person-rem / facility-year)
Do = immediate occupational dose Ym = avoided occupational long-term dose per facility-year (person-rem / facility-year)
Duo = long-term occupational dose.
It is possible that the proposed action will mitigate accident-related occupational exposures instead of (or as well as) reducing the accident frequency. In any case, it is the change from current condition to that following implementation of the proposed action that is sought. The formulation above can be replaced with the more explicit formulation below:
Yo - (FDo)s - (FDo ),
Ym = (FDm)s - (FDm),
where F = accident frequency (events / facility-year)
S = status quo (current conditions)
A = after implementation of proposed action.
Occupational risks from non-reactor accidents have been assembled into composite lists for selected non-reactor facilities in Section C.2.3 As for the public risks from non-reactor accidents, these also represent the products of accident frequencies and dose factors.
5.7.3.2 Monetary Valuation of Accident Related Exposures The analyst should use the $2000 per person-rem conversion value discussed in Section 5.7.1.2 for the monetary valuation of accident-related exposures.
5.7.3.3 Discounting Monetired % lues of Accident-Related Exposures The present values for "immediate" and long-term accident-related exposures in their monetized forms are calculated in slightly different ways.
"Immediate" Doses For "immediate" doses, the present value is Wo- C x Z,o j NUREG/BR-0184 5.32 e i
Value-Impact n
where Wo = monetary value of occupational health (accident) risk avoided per facility due to "immediate" doses, after discounting ($/ facility) I C = [exp(-rtg)- exp(-rtf )]/r t, = yean remaining until end of facility life t,- = years before facility begins operating r = real discount rate (as fraction, not percent)
Z io = monetary value of occupational health (accident) risk avoided per facility-year due to "immediate" doses, before discounting ($/ facility-year).
If a facility is already operating, tj will be zero and the equation for C simplifies to C= 1 - exp(-rt,) /r
]
Should occupational health (accident) risk due to "immediate" doses not be discounted in an analysis, r efectively becomes zero in the preceding equations. In the limit as r approaches zero, C = tg- tg (or, C = tg when tg = 0). This new value of C should be used to evaluate W oi in the undiscounted case.
The quantity W,omust be interpreted carefully to avoid misunderstandings. It does not represent the expected reduction in occupational health (accident) risk due to "immediate" doses as the result of a single accident. Rather, it is the present value of a stream of potential losses extending over the remaming lifetime of the facility. Thus, it reflects the expected annual loss due to a single accident (this is given by the quantity Zm); the possibility that such an accident could occur, '
with some probability, at any time over the remaining facility life; and the effects of discounting these potential future g'~j losses to present value. Since the quantity Zoonly accounts for the risk of an accident in a representative year, the result is the expected loss over the facility life, discounted to present value.
14:eg 'Ibrun Doses For long-term doses, the present value is Wm= 2 Zm/mr exp(-rtj
=
1 - exp{-r(t, - t,) (1 - exp(-tm))
where Wm = monetary value of occupational health (accident) risk avoided per facility due to long term doses, after ,
discounting ($/ facility) !
m = years over which long-term doses accrue"0 r = real discount rate (as fraction, not percent) t, = years remaining until end of facility life t, = years before facility begins operating Zm = monetary value of occupational health (accident) nu avoided per facility-year due to long-term doses, before discounting ($/ facility-year).
If the facility is already operating, t, will be zero and the equation for Wms implifies to 2'
Wm= Zm/mr 1 - exp(rt,))[1 - exp(-rm)]
\
d 5.33 NUREG/BR-0184
Value-impact O
Should occupational health (accident) risk due to long-term doses not be discounted in an analysis, r effectively becomes zero in the preceding equations. In the limit as r approaches zero Wm=Zm(t, - t,)
or Wm=Zm t,, when t, = 0 The quantity W m must be interpreted carefully to avoid misunderstandings. It does not represent the expected reduction in occupational health (accident) risk due to long-term doses as a result of a single accident. Rather, it is the present value of a stream of potential losses extending over the remaining lifetime of the facility. Thus, it reflects the expected annual loss due to a single accident (this is given by the quantity Zm); the possibility that such an accident cocid occur, with some probability, at any time over the remaining facility life; and the effects of discounting the:: potential 'uture losses to present value. Since the quantity Zm only accounts for the risk of an accident in a representative year, the result is the expected loss over the facility life, discounted to present value.
The FORECAST computer code for regulatory effects cost analysis (Lopez and Sciacca 1996) allows input for the occupational health (accident) attribute.
5.7.4 Occupational Health (Routine)
As with occupational health (accident), the evaluation of the c5ct on occupational health from a change in mutine exposure due to proposed regulatory actions is a multi-step pmcess. Reduction in exposure is a!gebraically positive; increase is negative (viewed as a negative reduction).
The steps are as follows:
- 1. Estimate reductions in occupational heale l routine) risk per facility for implementation (Do,i) and operation (Dogo)
(see Section 5.7.4.1)
- 2. Convert each reduction in occupational health (routine) risk per facility from pe: son-rems to dollars via monetary evaluation of health effects (see Section 5.7.4.2):
Goa, = RDo,i Go ,o=RD o,o where Goni = monetary value of per-facility reduction in routine occupational dose to implement the proposed action, before discounting ($/ facility)
Gtao = monetary value of annual per-facility reduction in routine occupational dose to operate following implementation of the proposed action, before discounting ($/ facility-year)
Doni = per-facility reduction in routine occupational dose to implement the proposed action (person-rem / facility)
Dogo = annual per-facility reduction in routine occupational dose to operate following implementation of the proposed action (person-rem / facility-year)
R = manctary equivalent of unit dose ($/ person-rem).
- 3. Discount each reduction in occupational health (routine) risk per facility (dollars) (see Section B.2)*
NUREG/BR-0184 5.34
i l
I Value-Impact
'N 4
- 4. Sum the reductions and total over all facilities (dollars):
t O. = N (H o,, + H o,o)
- where V , = discounted monetary value of reduction in occupational health (routine) risk for all afected facilities ($)
Ho , = monetary value of per-facility reduction in routine occupational dose required to implement the proposed i action, after discounting ($/ facility) l . Homo = monetary value of per-facility reduction in routine occupational dose to operate following ,
{ implementation of the proposed action, after discounting ($/ facility) j
[ N = _ number of afected facilities.
I i )
Note the algebraic signs for Dani and Dono. A reduction in exposure is positive; an increase is negative. The dose for
' implementation (Domi) would normally be an increase and therefore negative. ]
If individual facility values rather than generic values are used, the formulations can be replaced with l V. - N, (Ho ,,, + Ho,o) i i 1 s where i = facility (or group of facilities) index. l i
5.7.4.1 Fahmetina of Change in Routine Exposure !
l
] A proposed NRC action can afect routine occupational exposures in two ways. It may cause a one-time increase in j
) routine dose due to implementation of the action (e.g., installing a retro 6t). It may also cause a change (either increase or i decrease) in the recurring routine exposures aAer the action is implemented. A new coolant system decontaminaten j technique, for example, may cause a small implemantation dose but may result in a decrease in annual exposures from -
} maintenance thereafter, i
For the standard analysis, the analyst may attempt to make exposure estimates, or obtain at least a sample of industry or other technical data for a validation of the estimmten developed. There are two components in the development of an exposure estimare: estimating the radiation Seid (rem / hour) and estimating the labor hours required. 1he product is the l exposure (person-rem). In developing operational estimates, the annual frequency of the activity is also required.
!- General estimates of radiation nelds can be obtained from a number of sources. For power reactors, Chapter 12 of the
! Final Safety Analysis Report (FSAR) for the plant will contain a partitioning of the poner plant into estimatad radiation j zones. Both summary tables and plant layout drawings are usually provided. Some FSARs provide exposure estimates for
- speci6c operational activities. The analyst must be cautioned that the FSAR values are calculated, not meastund. Actual
- data from operating facilities, as might be obtained from facility sure
- ;fs, weald have greater accuracy. Generic estimates ,
of dose rates for work on specific Pressunaed Water Reactor (PWE) and BW2 systems and components are provided by l 1
Beal et al. (1987) and included in Section B.3. These are used by Sciacca (1992) in NUREG/CR-4627 along with labor 4 l hours and occupational exposure estimmten for speci6c repair and modi 6 cation activities. Appropriate rderences are cited. j The FORECAST ::omputer code for regulatory efects cost analysis (Lopez ard Sciacca 1996) contains a database of ;
! ddault dose rates and ranges for both PWR and BWR systems.
[ Work in a radiation zone inevitably requires extra labor due to radiation exposure limits and lower worker efficiency.
3 Such inefficiencies arise from restrictive clothmg, rubber gloves, breathing through filtered respirators, standag on '
}
4 5.35 NUREG/BR 0184 4
1
, - ~ v
Value-Impact O
ladders or scaffolding, or crawling into inaccessible areas. In addition, paid breaks must be allowed for during a job.
Basically, there are five types of adjustment factors identified for work on activated or contaminated systems. LaGuardia et al. (1986) identify the following five time duration multipliers:
- 1. Height (i.e., work conducted at elevations, e.g., on ladders or scaffolds) = 10-20% of basic time duration
- 2. Respiratory Protection = 25-50% of basic time duration
- 3. Radiation Protection = 10-40% of basic time duration
- 4. Protective Clothing = 30% of adjusted time duration
- 5. Work Breaks = 8.33% of total adjusted time duration.
Sciacca (1992) provides information from which to estimate relevant labor productivity factors, whose values can vary with the status of the plant and work environment at the time of the action.
Keeping these factors in mind, the analyst can proceed with the estimation of implementation and operational doses. The implementation dose would be Do,i = - F , x W ,
where Doni = per-facility reduction in routine occupational dose required to implement the proposed action (person-rem / facility-year)
F, = radiation field in area of activity (rem / hour)
Wi = work force required for implementation (labor-hours / facility).
As mentioned earlier, implementation dose normally involves an increase, hence the negative sign in the equation.
The operational dose is the change from the current level; its formulation is D,o=(F,W o o A,), - (F, W o A,),
where Do,o = annual per-facility reduction in routine occupational dose to operate following implementation of the proposed action (person-rem / facility-year)
F, = radiation field in area of activity (rem / hour)
Wo = work forte required for activity (labor-hours / facility-activity)
Ar = number of activities (e.g., maintenance, tests, inspections) per year (activities / year)
S = status quo (current conditions)
A = after implementation of proposed action.
Again, note the algebraic sign for Do,o. As mentioned earlier, an operational dose reduction is positive; an increase is negative.
If the issue does not lend itself to the estimation procedure just presented, the analyst may use the following approximation specifically for reactor facilities. To estimate changes in routine operational dose, the analyst may directly estimate fractional changes for routine doses. The techniques for soliciting expert opinion discussed in Section 5.6.2 could be NUREG/BR-0184 5.36
. . _ _ , _ _ . _ . . . _ . _ _ _ _ . . _ _ _ _ - - - _ ~ _ _ _ _ _ _ _..__m. _ _ _ . _m.. _
1 Value-Impact 3(
employed. De average annual occupational dose for BWRs in 1993 was 330 person-rem / reactor and 0.31 person.
rem / worker (see 'Ihble B.9). For PWRs, the average was 194 person-rendreactor and 0.25 person-rem / worker (see
'thble B.10). De overall average annual occupational dose at LWRs in 1993 was 240 person-rem / reactor and 0.27
- person-rem /morker (see 'lhble B.11). Additional data on toutine occupational exposue for both power reactors and non .
reactor facilities are provided in Section B.3. Also, routine occupational risks for selected non-reactor facilities have been j assembled into composite lists in Section C.2.4.
l For a major esort beyond the standard analysis, the best source of data to estimate both the implementation and j operational exposures would be a thorough survey of health physicists at the afected facilities. This survey could be screened for bias and potential inflated value by a knowledgeable third party.
5.7.4.2 Monetary %luation of Routine Exposure The analyst should use the $2000 per person-rem conversion factor discussed in Section 5.7.1.2 for the monetary valuation of routine exposures.
, 5.7.4.3 Ne 'j=" Occupational Impacts ;
1
! In some cases, it will be possible to identify nonradiological occupational impacts associated with a proposed action.
l When possible, these should be identi6ed and included in the regulatory analysis. One source of data on the incidence of occupational injuries for various industries is the report Occupational injuries andIllnesses in the United States by Industry, published annually by the Department of Labor's Bureau of Labor Statistics (BLS). Data from thl report can be j % acce-I from the BLS Home Page on the Internet (URL: http:// stats.bts. gov:80/datahorne.htm). r 1
I Occupational injury data should be converted to a dollar valuation. The value of an injury should include medical costs !
and the value of lost production (RWG 1996, Section 5). The value of loss production is normally estimated using employee wage rates. Pain and suffering costs attributable to occupationalinjury can be identified qualitatively, but would not normally be quantified in dollar terms. Potential information sources for occupational injury valuation data are the
- National Center for Health Statistics (URL: http://www.cdc. gov /nchswww/nchshome.htm) and the publication Accident i Arcts published annually by the National Safety Council b? sed in Itaska, Illinois.
1
[
i 5.7.5 Ofsite Property j Estimating the efect of the proposed action upon ossite property involves three steps:
i j 1. Estimate induction in accident frequency (see Section 5.6), incorporating conditional probability of 3
containment /confmement failure, if applicable.
- 2. Estimate level of property damage.
i
- 3. Calculate reduction in risk to offsite property as V, = NAFD i I 2 where Vy = monetary value of avoided ossite property damage ($)
I N = number of affected facilities 4
i i \. ,
J
- 5.37 NUREG/BR-0184 4
Value-Impact O
AF = reduction in accident frequency (events / facility-year)
D = present value of property damage occurring with frequency F ($-year).
It is possible that the proposed action mitigates the consequences of an accident instead of, or as well as, reducing the acci-dent frequency. In that event, the value of the action is V,, - (NFD), - (NFD),
where F = accident frequency (events / facility-year)
S = status quo (current conditions)
A = after implementation of proposed action.
Reduction in offsite property damage costs (i.e., costs savings) is algebraically positive; increase (i.e., cost accruals) is negative (viewed as negative cost savings).
An important tool formerly used by the NRC to estimate power reactor accident consequences is the computer program CRAC2 (Ritchie et al.1985). More recently, the computer code MACCS (Chanin et al. 1990,1993; Summers et al.
1995a,b) has been developed to estimate power reactor accident consequences using currently available information.
MACCS was employed for the consequence analyses in NUREG-1150 (NRC 1991). The analyst interested in code descriptions for CRAC2 or MACCS is referred to the references cited.
For the standard analysis speci6cally applied only to power reactor facilities, estimates based on work by Mubayi et al.
(1995) can be employed. Mubayi et al. (1995) have developed costs for offsite consequences for the five power reactors analyzed in NUREG-1150 (NRC 1991). These costs have been weighted by the frequencies of the accident release categories for the five plants. The results (in 1990 dollars) are given in Ttble 5.6. ne analysis used Version 1.5.11.1 of the MACCS computer code (Chanin et al.1993) on a site-specific basis. Offsite costs have been calculated as the sum of those for emergency response and long-term protective action, defined as follows:
- For early consequences, an effective emergency response plan consisted of evacuation of all but 0.5 % of the popula-tion within a ten-mile radius at a specified speed and delay time following notification of the emergency.
Able 5.6 Weighted costs for offsite property damage for the five NUREG-1150 power reactors Cost (1990 $) Within 50 Miles Reactor DPe from the Plant Zion PMR 2.23E + 8 Surry P%R 2.30E + 8 Sequoyah P%R 3.19E + 8 Peach Bottom BMR 2.71E + 9 Grand Gulf BMR 1.87E + 8 Average 2.46E + 8 O
NUREG/BR-0184 5.38
Value-Impact O
- For long-term relocation and bannmg of agricultural products, the interdiction criterion was 4 rem to an individual over five years (2 rem in year one, folloud by 0.5 rem each successive year).
Cost valutz within 50 miles are to be used in the regulatory analysis. Alternative values reArcting shoner and long distances from the plant may be used for sensitivity analyses or special cases, and are available in Mubayi et al. (1 The present value for offsite pre >erty damage can be calculated as !
D=CxB where D = plead value of ossite propeny damage ($-year) !
C = [exp (-rt,) - exp (-n,)]/r t, = years reinaining until end of facility life t, = years before facility begins operating ,
r = real discount rate (as fraction not percent) l B = undiscounted cost of offsite propeny damage. '
If a facility is already operating, t, will be zero and the equation for C simplified to j l
C = 1 - exp(-n,), /r
' Should offsite property damage not be discounted in an analysis (e.g., when the time frame is sufficiently short to mitigate the need for discounting), r effectiwly becomes zero in the preceding equations. In the limit as r approaches acro, C = t, j
= t,(or, C = 1, when t, = 0). This new value for C should be used to evaluate D in the undiscounted case.
'Ihe quantity D must be interpreted carefully to avoid misunderstandings. It does not represent the expected offsite prop-eny damage due to a single accident. Rather, it is the present value of a stream of potential losses extending over the remaining lifetime of the facility. Thus, it reflects the expected loss due to a single accident (this is given by the quantity B); the possibility that such an accident could occur, with some probability, at any time over the emaining facility life; and the effects of discounting these potential future losses to present value. When the quantity D is multiplied by the annual fwquency of an accident, the result is the expected loss over the facility life, discounted to present value.
Costs for offsite prepeny damage from non-reactor accidents have been assembled in Section C.2.5. However, most are given as combined offsite and onsite damage costs and have not been as thoroughly estimawd as those by Mubsyi et al.
(1995) for offsite propeny damage frcm power reactor accidents.
At a more detailed level, but still within the scope of a standard analysis, the analyst can identify the affected facilities, then calculate the proper sum effect rather than relying on generic values. The following steps are required:
- 1. Identify affected facilities.
- 2. Identify reductions in accident frequency per facility.
- 3. Calculate value of property damage per facility.
5.39 NUREG/BJ.-0184
Value-Impact O
- 4. Calculate avoided propeny damage value per facility.
- 5. Sum avoided propeny damage over affected facilities.
In the 1983 Handbook, Heaberlin et al. made extensive use of NUREG/CR-2723 (Strip 1982) for offsite propeny cost estimation. Strip reponed the pmsent value of offsite health and property costs, onsite costs, and replacement power costs for accidents in release categories SSTI through SST3 for 91 U.S. power reactor sites. The offsite propeny costs were based on CRAC2 results, with 1970 population estimates and state-wide land use. The analyst may find the site-speci6c emphasis in Strip (1982) helpful in a more detailed value-impact analysis.
For a major effort beyond the standard analysis, it is recommended that the estimates be derived from information more site-specific than that used by Strip (1982). For power reactors, the MACCS code with the most recent data available should be used. This degree of effort would be relatively costly to conduct, both in terms of computer costs and data col-lection and interpretation costs. Howewr, it would provide the highest degree of reliability.
Burke et al. (1984) examined the offsite economic consequences of severe LWR accidents, developing costs models for the following:
- population evacuation and temporary sheltering, including food, lodging, and transportation
- emergency phase relocation, including food, housing, transportation, and income losses e intermediate phase relocation, beginning immediately after the emergency phase e long-term protective actions, including decontamination of land and propeny and land area interdiction
- health effects, including the two basic approaches (human capital and willingness-to-pay).
Tawil et al. (1991) compared three computer models for estimating offsite propeny damage from power reactor accidents.
Two of the models are the CRAC2 and MACCS codes; the third is the computer code DECON (Tawil et al.1985). Three accident severity categories-SST1-SST3-are considered for the six Pasquill atmospheric stability categories (A-F).
Offsite propeny damage is calculated for each pairing at cleanup levels from 10 through 200 rems. A study is also performed comparing the effect of modeling offsite damage to radii of 50 and 500 miles. It indicates that the choice of radius is significant only for the SST1 accident category, the diffemnces being quite pronounced.
The FORECAST computer code for regulatory effects cost analysis (Lopez and Sciacca 1996) allows input for the offsite property attribute.
5.7.6 Onsite Property Section 4.3.1 of the NRC Guidelines states that onsite propeny damage cost savings (i.e., avened onsite costs) need to be included in the value-impact analysis. In the net-value formulation it is a positive attribute.
Estimating the effect of the proposed action on onsite propeny involves three steps:
- 1. Estimate reduction in accident frequency (see Section 5.6).
- 2. Estimate onsite propeny damage.
O NUREG/BR-0184 5.40
i l
Value-linpact t f
i
- 3. Calculate reduction in risk to onsite property as V, = N AFU t
where Voi, = monetary value of avoided onsite property damage ($) :
N = numberof affected facilities AF = reduction in accident frequracy (events / facility-year) jl U = present value of property damage humus with figy F ($-year).
- Reduction in onsite property damage costs (i.e., costs savings) is algebraically positive; merease (i.e., cost ecruals) is negative (viewed as negatiw cost savings). 'la
.i For the standard analysis, it is comenient to treat onsite property costs under three categories: 1) cleanup and decontami-nation, 2) long-term replacement power, and 3) repair and refurbishment. Each of these categories is considered below ,
for power reactors with the focus on large-scale core-melt accidents. Additional categories of costs have been considered .i by Mubayi et al. (1995) and Burke et al. (1984) as outlined in Section 5.7.6.4, but they were either found to be speculative !
or contributed small fractions to the costs identified below. I i
5.7.6.1 Cleanup and h==*==a eann i
Cleanup and decontamination of a nuclear facility, especially a power reactor, followmg a medium or severe accident can l
be extremely expensive. For example, Mubgi et al. (1995) report that the total cleanup and decontamination of TMI-2 l
) cost roughly $750 million (in 1981 dollars). Murphy and Holter (1982) estimatM cleanup costs for a reference PWR and j d BWR for the following three accident scenarsos: I Scenario 1 - a small LOCA in which ECCS functions as intended. Some fuel cladding ruptures, but no fuel melts.
The containment building is moderately contammatM but there is minimal physical damage.
Oma 2 - a small LOCA in which ECCS is delayed. Half of the fuel cladding ruptures, and some fuel melts. The coo atment building is extensively conimmmarM, but there is minimal physical damage.
Scenario 3 - a major lhCA in which ECCS is delayed. All fuel claddag ruptures, and there is signi6 cant fuel melt-ing and core damaged. 'Ihe containment building is extensively containmatM and physically d-a-ad. The auxiliary building undergoes some contammation.
I In 1981 dollars, Murphy and Holter estimated the following cleanup costs:
Scenario PWR BWR 1 $1.05E+8 $1.28E+8 2 $2.24E+8 $2.28E+8
.3 $4.04E+ 8 $4.21E+8 l 1
Mubayi et al. (1995) consider the TMI-2 accident to lie between Sanarios 2 and 3, lying closer to Scenario 3 in terms of l
the contamination and damage to the core. Murphy and Holter's costs were somewhat less than those actually realized at :
TMI. Mubayi et al. (1995) attribute the difference to three factors:
5.41 NUREG/BR4)184
Value-Impact O
- 1. The start of the TMI cleanup was delayed by 2.5 years due to regulatory and financial requirements. Murphy and tfolter assumed no additional delays between the accident and stan of the cleanup. Mubayi et al. (1995) consider this somewhat unrealistic.
- 2. Decontamination at TM1 required facilities not included in Murphy and Holter's reference plants (e.g., a hot chemis-try laboratory, containment recovery service building, and comment center / temporary personnel access facility).
- 3. TMI required additional decontamination of the containment building after the reactor was defueled. Murphy and Holter excluded this in their analysis.
When these three factors are considered, the results from Murphy and Holter become reasonably consistent with the actual TMI cleanup costs ($7.50E+8 in 1981 dollars).
Burke et al. (1984) produced a very rough estimate of $1.7 billion (in 1982 dollars) for the cleanup and decontammation costs following a severe power reactor accident. An uncertainty range of appmximately 50% was assigned, bringing the lower bound reasonably in line with the actual TM1 cleanup cost. A study by Konzek and Smith (1990) updated the cleanup costs associated with Murphy and Holter's Scenario 3. Costs ranging from $1.22E+9 to $1.44E+9 (in undis-counted 1989 dollars) were estimated, based on real escalation rates of 4% to 8% during the cleanup period. A base cost of $1.03E+9 was estimated assuming no real escalation during the cleanup period.
After converting the costs to undiscounted 1993 dollars, the cost reported by Mubayi et al. (1995) for TMI is $1.2E+9, the base estimate from Konzek and Smith (1990) is $1.2E+9, and the estimate from Burke et al. (1984), which doubled the cost of TMI, is $2.5E+9. Based on these references, the total onsite cost estimates given in Section 5.7.6.4 are based on $1.5E+9 (undiscounted) for cleanup and decontamination (Cco in the equations that follow). For sensitivity analysis, lower and upper bounds of $1.0E+9 and $2.0E+9 are recommended for evaluating severe accident effects.
Assuming the $1.5E+9 estimate is spread evenly over a 10-year period for cleanup (i.e., constant annual cost of Cco/ni =
$1.5E+8 in the equation below, with Cco = $1.5E+9 and m = 10 years), and applying a 7% real discount rate, the cost translates into a net present value of $1.1E+9 for a single event. This quantity is derived from the following equation (see Section B.2.3):
PVco = [Cco / mr] (1 - exp(-rm))
where PVco = net present value of cleanup and decontamination costs for single event ($)
Cco = total undiscounted cost for single accident in constant year dollars ($)
m = years required to return site to pre-accident state r = real discount rate (as fraction, not percent).
Before proceeding, this present value must be decreased by the cleanup and decontamination costs associated with normal reactor end-of-life. The Yankee Atomic Electric Co. (NRC 1995c), Sacramento Municipal Utility District (NRC 1994),
and Portland General Electric Co. (1995) provided the following estimates to the NRC for decommissioning their Yankee Rowe, Rancho Seco, and Tmjan nuclear power plants, respectively: $3.41E+8 (1991 dollars), $2.80E+8 (1991 dollars),
and $4.15E+8 (1993 dollars). These suggest a value of approximately $0.4E+9 (1993 dollars) for " normal" cleanup and decommissioning. The analyst can also consult Bierschbach (1995) for estimating PWR decommissioning costs and Bierschbach (1996) for estimating BWR decommissioning costs.
When spread evenly over the same 10-year period at a 7 % real discount rate, this translates into a net present value of
$0.3E+9. However, since this value would "normally" be applied at reactor end-of-life (i.e.,24 years later, using the NUREG/BR-0184 5.42
Value-impact (v/
estimate from Table B.1), the net present value (at the same 7% real discount rate) is reduced to $0.06E+9. Since this amom ',o only 5% of the net present value for cleanup and decontamination following a severe accident ($1.lE+9), it can be generally ignored.
The total onsite cost estimates shown in Section 5.7.6.4 integrate this net present value over the average number of remaining service years (24 years) using the following equation:
U co =(PVeo / r] (1 - exp(-rt,)]
i where Uco = net present value of cleanup arJ decontamination over life of facility ($-year) t, = years remaining until end of facility life.
The integrated cost is $1.3E+ 10 over the life of a power reactor. This cost must be multiplied by the accident frequency (F, expressed in events per facility-year), and the number of reactors, to determine the expected value of cleanup and l decontamination costs. To determine averted costs, the reduction in accident frequency AF is applied as outlined in '
Section 5.7.6.
For comparison, these costs can also be estimated for less severe accidents as defmed by Murphy and Hoher's Scenarios 1 and 2. The estimates shown in the following table were obtained by using $1.1E+9 (1993 dollars) as a base value for Scenario-3 PVco costs, and applying the same relative fractions as shown in Murphy and Holter's (1982) results for Scenario-1 and 2 costs. The results from Murphy and Holter were not used directly because of the factors cited by in\
( ./
Mubayi et al. (1995) in compansons of those estimates with actual cleanup and decontamination costs at TMI.
Scenario PVr,m Urn I
l 1 $3. l E + 8 $3.7E+9 2 $6.0E + 8 $7. l E + 9 l 3 $1.1 E + 9 $1.3E + 10 The issue of license renewal has only moderate implications for the integrated cost estimates (Uco). With longer operating lifetimes, the reactors are at risk for more years, and the costs would be expected to increase accordingly. However, because the additional costs are discounted to present worth terms, the effect is not substantial. For example, an additional life extension of 20 years would only increase the value of Uco for a Scenario-3 accident 15% from $1.3E+10 to
$1.5 E + 10, 5.7.6.2 Iong-Term Replacement Ft>wer Replaced power for short-term reactor outages is discussed in Section 5.7.7.1. Following a severe power reactor accident (replacement power need be considered only for electrical generating facilities), replacement power costs must be considered for the remaining reactor lifetime.o23 Argonne National Laboratory (ANL) has developed estimates for long-term replacement power costs based on simulations of production costs and capacity expansion for representative pools of utility systems (VanKuiken et al.1992). VanKuiken et al. examined replacement energy and capacity costs, including purchased energy and capacity charges required to pro-vide the same level of system reliability as available prior to the loss of a power reactor (VanKuiken et al.1993). In the event of a permanent shutdown, it was assumed that a reactor would be replaced by one or more alternative generating units, after an appropriate delay for planning and construction.
/
5.43 NUREG/BR.0184
Value-Impact O
Capacity expansion and production cost simulations were performed for six representative power reactors over 40-year study periods. The results were used to estimate replacement power costs for each of 112 reactors which, at the time of the study, were expected to be in operation by 1996. Cost estimates for each reactor reflect the remammg lifetimes, reactor sizes, and ranges in shon-term replacement energy costs (as encountered in each utility). Averages were deter-mined by summing the individual reactor costs and dividing by the number of reactors evaluated. Characteristics for the
' generic" reactor cited in Section 5.7.6.4 reflect an average unit size of 910-MWe and average life remaming of 24 years for reactors currently operating and planned.
Simulation results were first used to estimate the present value costs of single accidents occurring in each year of remaining facility lifetimes (quantity PV,, used in the discussions that follow). Each of these net present values represents a summation of annual replacement power costs incurred from the year of the assumed accident to the final year of service. For example, the average net present value for an event occurring in 1993 is $1.lE+9. For 1994, the cost is
$1.0E +9, and for 1995, the cost is $0.9E+9. The decline in costs with each successive yea- reflects present value considerations and the fact that there att fewer remaining service years requiring replacement power.
The following equation can be used to approximate the average value of PVa, for alternative discount rates.
PV,, = [$1.2E + 8 / r) 1 - exp(-n,)
where PV,, = net present value of replacement power for a single event ($).
The $1.2E+8 value used in the above equation has no intrinsic meaning. It is treated in the equation similar to an equivalent annual cost, but it is actually a substitute for a string of non-constant replacement power costs that occur over the lifetime of the generic reactor after an event that takes place in 1993. The equation is only presented here for examining the effects of alternate discount rates and remaining reactor lifetimes.
The above equation for PVg, was developed for discount factors in the range of 5%-10%. Unlike the equations for PV co and Uco, the equation for PV a, diverges from modeled results at lower discount rates. At a discount rate of 3 % the recommended value for PVa ,is $1.4E+9, as compared with the equation estimate of $1.lE+9. For discount rates between 1% and 5% the analyst is urged to make linear interpolations using $1.6E &9 at 1% and $1.2E+9 at 5%. At higher discount rates the equation for PVa, provides recommended estimates of f 4.2E+9 at 5 % and $1.0E+9 at 10%.
The results that are applied in Section 5.7.6.4 sum the single-event costs over a'1 1 years of reactor service. While these summations were calculated directly fmm simulation results, ANL found that the outcomes could be closely approximated with the equation that follows. The squared term in this equation servt:s as a proxy for the fact that costs for events in future years decline due to the reduced number of remaining service years for which replacement power is required:
U,, - f,,/ r) (1 - exp(-n,)f where Um, = net present value of replacement power over life of facility ($-year).
Replacement power costs for the generic unit are estimated to be approximately $10 billion over the life of the facility. An uncertainty range for this average is estimated at appmximately 20%. However, the range of estimates for specific power reactors varies directly with unit size, remaining life, and replacement energy costs. For example, costs were estimated to be $7.5 billion for the 1040-MWe Zion-2 reactor, assuming 16 years of remaining opemting life. Zion-2 is in a power pool with approximately average replacement energy costs. In contrast, costs for Big Rock Point were $120 million due to its smaller size (67-MWe), shoner remaining life (8 years assumed), and average replacement energy costs. At the upper NUREG/BR-0184 5.44
I I
Value-Impact l i
(
p)
U limit were costs of $24 billion for the 1090-MWe Nine Mile Point 2 unit, assuming 34 years of service remaining. Nine Mile Point 2 is in a power pool with above average replacement energy costs. )
As noted for PV,,, the equation for U,, was developed for discount rates ranging from 5%-10%. For lower discount rates, linear interpolations for U,, are recommended between $1.9E+10 at 1 % and $1.2E+10 at 5 %. The equation for U,, yields the recommended values of $1.2E+10 at 5% and $0.8E+10 at 10%, based on PV,, values described previously.
As discussed in Section 5.7.6.4, these summed costs must be multiplied by the accident frequency (expressed in events per facility-year) to determine the expected value of replacement power costs for a typical reactor. To determine the value of reductions in the accident fr:quency due to regulatory actions, the total integrated costs must be multiplied by the reduction in accident frequency AF and the number of reactors affected (N).
The issue of license renewal has a much more significant impact on replacement power costs than on cleanup and decontamination costs. Extending the operating life by an additional 20 years would increase the net present value of a single event (PV,,) by about 38%, and would increase the present value of costs integrated over the reactor life (U,,) by about 90% (VanKuiken et al.1992). Thus, a license renewal period of 20 years would mean the generic reactor would have a remaining life of 44 years, PV,, would be estimated to be $1.5E+9, and U,, would be approximately $1.9E+10 (1993 dollars).
For less severe accidents such as characterized by Scenario-1 events, the analyst is referred to Section 5.7.7.1 which addresses short-term replacement energy costs. Replacement capacity costs, which contribute to severe accident costs, are A not incurred for more temporary reactor shutdowns.
I V} 5.7.6.3 Repair and Refurbishment in the event of recoverable accidents (i.e., for Scenario 1, but not Scenarios 2 or 3), the licensee will incur costs to repair /
replace damaged components before a facility can be returned to operation (these costs are not included in the total onsite l cost estimates for severe ac.idents as addressed in Section 5.7.6.4). Burke et al. (1984) have estimated typical costs for equipment repair on the order of $1,000/hr of outage duration, based on data from outages of varying durations at reactors. They suggest an upper bound of roughly 20% of the long-term replacement power costs for a single event.
Mubayi et al. (1995) observe that the $1,000/hr figure corresponds closely to the repair costs following the Browns Ferry fire and also to the TM1-1 steam generator retubing outage costs.
5.7.6.4 Tbtal Onsite Property Damage Costs l Based on the information included in Sections 5.7.6.1 and 5.7.6.2, ANL has estimated the total cost due to onsite property j damage following a severe reactor accident for the Zion-2 reactor and a " generic" 910-MWe reactor assumed to have a remabmg hfe of 24 years. Total costs are assumed to consist of cleanup and decontamination costs and replacement power costs (repair and refurbishment costs are not included for severe accidents). The total costs described below correspond to the " risk-based" costs as defined by Mubayi et al. (1995):
"... risk-based cost, the discounted net present value of the risk over the remaining life of the plant, which is proportional to the accident frequency (F]..."
The risk-based costs (quantities U, Uco, and U,, in the equations that follow) must be interpreted carefully to avoid misunderstandings. They do not represent the expected onsite property damage due to a single accident. Rather, they are the present value o? a stream of potential losses extending over the remaining lifetime of the facility. Thus, they reflect f,- ,
T the expected loss due a a single accident (given by quantities PVco and PV,,); the possibility that such an accident could
[
\
5.45 NUREG/BR-0184
Value-Impact i
occur, with some small probability, at any time over the remaining facility life; and the effects of discounting those e
potential future losses to the present value. When the quantity U is multiplied by the annual accident frequency, the result is the expected loss over the facility life, discounted to the present value.
The estimates for total risk-based costs attributed to regulatory actions that occur in 1993, expressed in 1993 dollars assuming a 7% real annual discount rate, are as follows:
Variable Cost Component Ziq1-1 " Generic" Reactor U,, Replacement Power $0.7E+10 x F $1.0E+10 x F Uco Cleanup & Decontamination $1.0E+10 x F $1.3E+10 x F U Total $1.7E+10 x F $2.3E+10 x F Alternate values of U may be approximated for different discount rates, years of operation remaming, and estimates for Cc3 and PV,,. However, for changes in discount rate or final year of operation, the analyst is cautioned to revise the esti-mates for PVa, using the equation described in Section 5.7.6.2 prior to re-estimating U from the equation that follows.
Also, for discount rates lower than 5 %, PV,, and U,, should be estimated from interpolation guidelines presented in Section 5.7.6.2 rather than from the equations. The relationship that defines total lifetime costs is U=Uco + U,,
= Cco /mr ) fl - exp(-rt,)) (1 - exp(-rm)) + (PV,,/r] 1 - exp(-rt,))
2 where U = total net present value of onsite property damage ($-year).
The procedure outlined in Section 5.7.6 may be used to evaluate averted onsite property damage using these estimates.
For illustration, assume that the reduction in severe accident frequency (AF) is 1.0E-6 and the number of reacton affected (N) is 111. The total averted onsite damage costc would be Vo, = NAFU = (Ill)(1.0E-6)($2.3E + 10) = $2.6E + 6 The value of this redaction in accident frequency is $2.6 million (net present value in 1993 dollars).
The $2.3E+10 value used above is an appropriate generic estimate for regulatory requirements that become effective in 1993 and that affect severe accident probabilities in that year. For regulatory actions that affect accident frequencies in future years, the cost estimates must be adjusted to recognize that the number of reactor-years at risk and the number of service years requiring replacement power are reduced. 7hble 5.7 shows how these factors affect cost estimates for the 10-year period of 1993-2002. The results are expressed as net present values discounted to the year that the rulemaking is assumed to take effect.
To illustrate the use of these estimates, assume a reduction in accident frequency of 1.0E-6 begins in 1998 and affects all 111 of the remaining reactors. The revised astimate for U would be $1.9E+10 and the total averted onsite damage costs for this reduction in frequency would be Vo, = (Ill)(1.0E-6)($1.9E + 10) = $2.1E + 6 (1993 dollan)
O NUREG/BR-0184 5.46 i
l
. _ . _ . _ . . . . . _ . _ _ _ _ _ . _ _ _ _ . _ . . _ _ _ _ _ _ _ . . _ , _ _ _ _ _ ~ . _ _ _ . . _ .
l Value-Impact O
hble 5.1 Onsite property damage cost wi==*= (U) for future years (1993 dollars discounted to year of implesmentation) ;
i
. I
- Cleanup and Decontammation l (Uen) Replacement Ptmer (U ,) Tbtal(U) l 1993 $1.3E+10 $1.0E+10 $2.3E+10 ;
1994 $1.2E+10 $9.6E+9 $2.2E+10 ;
) 1995 $1.2E+10 $9.1E+9 $2.1E+10 1996 $1.2E+10 $8.6E+9 $2.1E+10 j 4
1997 $1.1E+10 $8.1E+9 $1.9E+10 7
1998 $1.1E+ 10 $7.6E+9 $1.9E+ 10 g.
i 1999- $1.1E+10 $7.1E+9 $1.8E+10 l
! '2000 $1.1E+10 $6.6E+9 $1.8E+10 2001. $1.0E+10 $6.2E+9 $1.6E+10 2002 $1.0E+10 $5.7E+9 $1.6E+10
- This would indicate that the reduction in accident frequency valued at $2.6 million beginning in 1993 would be valued at F $2.1 million if introduced in 1998 (1993 dollars adjusted to 1998).
The following linear equation provides approximate cost estimates for implementation later than 10 years in the future.
ne result represents net present value (1993 dollars) discounted to the year of implemantation, ne analyst must adjust l the 1993 dollars for general in8ation if costs are to be expressed in alternate reference-year dollars. (See Section 5.8 for i
information on adjusting dollar years.)
U = $2.3E + 10 - ($6.7E + 8) (t, - 1993) where t, = year of reduction in accident frequency.
Thus, for regulatory actions that would affect accident probabilities for 86 reactors remaining in service in 2010, the revised estimate for U would be U = $2.3E + 10 - ($6.7E + 8)(2010 - 1993) l
= $1.2E + 10 (1993 dollars adjusted to 2010)
The total averted onsite damages costs for a reduction in accident frequency of 1.0E-6 would be V, = (86)(1.0E -6)($1.2E + 10)
= $1.0E + 6 (1993 dollars adjusted to 2010) 5.47 NUREG/BR-0184
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This example also illustrates that the number of reactors at risk and the average remammg years of reactor service change in the evaluation of future regulatory initiatives. Because of the distribution oflicense expiration dates, the average remaining reactor life does not decrease on a one-to-one basis with each successive year in the future.
For 20-year license renewal considerations, the estimates for U discussed above should be increased by approximately 50%. In 1993, Uco would be estimated at $1.5E +10 (versus $1.3E+10 for 40-year license), r.nd Un would be estimated to be $1.9E+ 10 (wrsus $1.0E+10 for 40-year license). His yields a total of $3.4E+10 (1993 dollars) as compared with
$2.3E+10 for the 40-year license assumption.
Costs for onsite propeny damage from non-reactor accidents have been assembled in Section C.2.5. However, most are given as combined offsite and onsite damage costs.
For a major effon beyond the standard analysis, there are two general ways to achieve a greater level of detail: 1) the analysis can be conducted for individual facilities or groups of similar facilities, using site-specific information; 2) the analysis can provide cost information in much greater detail. With regard to the first approach, the most relevant site-speciSc information includes the cost of long-term replacement power and the value of the facility and equipment at risk, taking into account the remaining useful life of the facility. De analyst is referred to VanKuiken et al. (1992) for funher detail on average shutdown costs for different categories of reactors (e.g., by region, reactor supplier, architect engineer, etc.), and guidance for scaling costs for different unit sizes and remammg lifetimes.
With regard to providing greater detail on the cost information, the major cost elements (in addition to replacement power) are likely to include decontamination and other cleanup costs and repair or replacement of plant and equipment that is physically damaged. Other costs relate to transporting and disposing of contaminated materials and equipment, and startup costs. Costs for me Storing the site for radiation and fixing contamination at the site will likely be insignificant relative to the other costs. The analyst is referred to Murphy and Holter (1982), and the follow-up study by Konzek and Smith (1990), for detailed cost estimates to decontammate a nuclear power reactor following a postulated accident.
Burke et al. (1984) examined the onsite economic consequences of severe LWR accidents, developing cost models for the following:
replacement power, drawing information mainly from Buchring and Peerenboom (1982) (which has been updated by VanKuiken et al. [19921) e plant decontamination, including both medium and large conseluence events
- plant repair, spanning small to large consequence events e early decommissioning for medium and large consequence events worker health effects and medical care, primarily for medium and large consequence events electric utility " business" (i.e., costs resulting from changed risk perceptions in financial markets and the need to replace the income once produced by the operating plant after a power plant is permanenity shutdown)
- nuclear power " industry" (i.e., costs resulting from elimination or slowed growth in the U.S. nuclear power industry due to altered policy decisions and risk perceptions following a severe accident) onsite litigation (i.e., " legal fees for the time and effon of those individuals involved in the litigation process").
O NUREG/BR-0184 5.48
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The first three categories of costs have been covered in Sections 5.7.6.1-5.7.6.3. The other categories are covered elsewhere in this Handbook or are considered to be either speculative or small in magnitude relative to replacement power, cleanup and decontamination, and repair costs.
The FORECAST computer code for regulatory effects cost analysis (lepez and Sciacca 1996) allows input for the onsite property attribute.
I 5.7.7 Industry Implementation This section provides procedures for computing estimates of the industry's incremental costs to implement the proposed I action. Estimating incremental costs during the operational phase that follows the implementation phase is discussed in Section 5.7.8 Incremental implementation costs measure the additional costs to industry imposed by the regulation; they ,
are costs that would not have been incurred in the absence of that regulation. Reduction in the net cost (i.e., cost savings) I is algebraically positive; increase (i.e., cost accrual) is negative (viewed as negative cost savings). Both NRC and Agreement State licensees should be addressed, as appropriate. :
1 1
In general, there are three steps that the analyst should follow in order to estimate industry implementation costs-1 Step 1 - Estimate the amount and types of plant equipment, materials, and/or labor that will be affected by the proposed )
action.
Step 2 - Estimate the costs associated with implementation.
A ;
(J) Step 3 - If appropriate, discount the implementation costs, then sum (see Section B.2).
In preparing an estimate of industry implementation costs, the analyst should also carefully consider all cost categories that i may be affected as a result of implementing the action. Example categories include I
- land and land-use rights
- structures
- hydraulic, pneumatic, and electrical equipment
- radioactive waste disposal
- health physics
- monitoring equipment
- personnel construction facilities, equipment, and services
- engineering services
- recordkeeping
- procedural changes O
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a license modi 6 cations i e staff training / retraining l
- administration l
- facility shutdown and restart l
- replacement power (power reactors only) l
- reactor fuel and fuel services (power reactors only)
- items for averting illness or injury (e.g., bottled water or job safety equipment).
Note that transfer payments (see Section 4.3) should not be included.
For the standard analysis, the analyst should use consolidated informa: ion to estimate the cost to industry for implementing the action. Sciacca (1992) is a prime source of such information, providing not only cost estimates, but also labor hours, cost rates, and adjustment factors, mainly for reactor facilities. Appropriate references are cited by Sciacca. The FORECAST computer code for regulatory effects cost analysis (Lopez and Sciacca 1996) incorporates much of the j information assembled by Sciacca (1992) into a computer database for the analyst's use in estimating industry implementa-tion as well as other costs.
Step 1 - Estimate the amounts and types of plant equipment, materials, and/or labor that will be affected by the proposed action, including not only physical equipment and craft labor, but professional staff labor for design, engineering, j quality assurance, and licensing associated with the action. If the action requires work in a radiation zone, the analyst should account for the extra labor required try radiatica exposure limits and low worker efficiency due to l
awkward radiation protection gear and tight quarters (see discussion of labor productivity in Section 5.7.4.1).
- When performing a sensitivity analysis, but not for the best estimate, the analyst should include contingencies, such as the most recent greenfield construction project contingency allowances supplied by Robert Snow Means Co., Inc. (1995). They suggest adding contingency allowances of 15 % at the conceptual stage,10% at the schematic stage, and 2% at the preliminary working drawing stage. The FORECAST computer code (1epez and Sciacca 1996) contains an option to include an allowance for uncertainty and cost variations at the summary cost level. The Electric Power Research Institute (EPRI 1986) offers guidelines for use in estimating the costs for "new and existing power generating technologies," EPRI suggests applying two separate contingency factors, one 1 for " projects" to cover costs resulting from more detailed design, and one for " process" to cover costs associated l with uncertainties of implementing a commercial-scale new technology.
1 Step 2 - Estimate the costs associated with implementation, both direct and indirect. Direct costs include materials, i equipment, and labor used for the construction and initial operation of the facility during the implementation phase. Indirect costs include required services. The analyst should identify any significant secondary costs that may arise. One-time component replacement costs and associated labor costs should be accounted for here. For additional information on cost categories, especially for reactor ficilities, see Schulte et al. (1978) and United ,
Engineers and Constructors, Inc. (1979; 1988a, b).
l l
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i Step 3 - If appropriate, discount the costs, then sum. If costs occur at some future time, they should be discounted to yield present values (see Section B.2). If all costs occur in the first year or if present value costs can be directly r ' estimated, discounting is not required. Generally, implementation costs would occur shortly after adoption of the l proposed action.
! When performing value-impact analyses for non-reactor facilities, the analyst will encounter difficulty in fmding consolidated information on indust,y implementation costs comparable to that for power reactors. Comprehensive data
! sources such as Sciacca (1992) and the references ftom which he drew his information are generally unavailable for non-reactor facilities. Some speci6c information for selected non-reactor facilities is in Sections C.7-C.10. De types of non--
- , reactor facilities (see Section C.1) are quite diverse. Furthermore, within each type, the facility layouts typically lack the limited standardization of the reactor facilities. Dese combine to leave the analyst pretty much "on his own" in developing industry implementation costs for non-reactor facilities. The analyst should follow the general guidelines given 4 in this Handbook section. Specific data may be best obtainM through direct contact with knowledgeable sources for the facility concerned, possibly even the facility personnel themselves.
[
E i For a major effort beyond the standard analysis, the analyst should obtam very detailed information, in terms of the cost categories and the costs themselves. The analyst should seek guidance from NJ.C contractors or industry sources experi-i enced in this area (AE firms, etc.). The incremental costs of the action should be de6ned at a fmer level of detail. De
) analyst, should refer to the code of accounts in the Energy Economic Data Base (EEDB [ United Engineers and Con-structors, Inc.1988b]) or Schulte et al. (1978) to prepare a detailed account of implementation costs.
' 5.7.7.1 Short 'Ilenn Regdacement Pbwer 1 For power reactors, the possibility that implementation of the proposed action may result in the need for short-term replacement power must be addressed. Section 4.3.2 of the Guidelines indicates that replacement power costs are to be incorporated into a regulatory analysis when appropriate. Unlike the long-term costs associated with severe power reactor
- accidents discussed in Section 5.7.6.2, the replacement power costs associated with industry implementation of a j regulatory action would be short-term.
For a " typical" 910-MWe reactor operating at an average capacity factor of 60%-65%, VanKuiken et al. (1992) suggests l
- $310,000/ day (1993 dollan) as su average cost for short-term replacement power The 60%-65 % range in capacity factor
- is representative of annual averages, accounting for unplanned outage periods and planned outage periods for maintenance and refueling. However, if the timing of a short-term shutdown coincides with a time wher
- a power reactor is expected to
- be fully operational, then a higher average cost per day is more appropriate. At a capacity factor of 100%, the average
, cost for the typical reactor is estimarM to be $480,000/ day (1993 dol'ars). l
- At a more detailed level, VanKuiken et al. (1992) project the seasonal replacement power costs for potential short-term
- shutdowns of 112 nuclear power plants over the five-year period from 1992 through 1996. These costs are estimated from l probabilistic production-cost simulations of pooled utility-system operations. Average daily replacement power costs are j presented by season for each of the 112 plants. The 20 U.S. power pools containmg these plants are identi6ed along with their following characteristics
- total system capacity, annual peak load, annual energy demand, annual load factor, prices 4 for fuels, and mix of generation by fuel type.
l De sensitivity of replacement power costs to changes in oil and gas prices is quantified for each power pool. De effects
!- of multiple plant shutdowns are addressed, with the replacement power costs quantified for each pool assuming all plants -
I within the pool are shutdown.
i i
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The replacement power cost information compiled in an analogous but earlier study by WaKuiken et al. (1987) has subse-quently been incorporated into two cost analysis computer codes. 'Ibe Replacement Energy Cost Analysis Package (RECAP [VanKuiken et al.1994]) determines the replacement energy costs associated with short-term shutdowns of nuclear power plants, and can be applied to determine average costs for general categories based on location, unit type (e.g., BWR), constructor, utility, and other differentiating criteria. Plant-specific costs are also available, and can be evaluated for user-specified outage durations and alternative capacity factor assumptions. FORECAST (Lopez and Sciacca l 1996), a computer code for regulatory effects cost analysis, provides the user with the capability to estimate replacement power costs in current year dollars. Sciacca (1992) also provides a discussion and data for use in estimating replacement power costs based on this earlier study by VanKuiken et al. (1987).
Imposition of a new regulation often requires that a nuclear power plant be shutdown while the modification takes place.
If the requirement is needed to meet adequate protection, the analyst can assume that the required downtime is independent of any scheduled downtime, thereby realizing full replacement power costs. However, the modification often is not needed to meet adequate protection, enabling it to be completed during already scheduled downtime. Only if the time needed to perform the modification exceeds that allotted for the scheduled downtime should any replacement power costs accrue, these being solely due to the excess time.
The most likely scenario permits the modiScation to be accommodated completely within already scheduled downtime, and this has frequently been the policy adopted by the NRC. As a result, no replacement power costs accrue. While this i
assumption holds for a modification performed in the absence of other; required by new regulations, it tends to underestimate the cost of multiple modifications resulting from the cumulative effect of new NRC requirements. When multiple modifications are perfortr* ,.1, as they often are, the originally scheduled downtime may be insufficient to I accommodate all of them. Usually, this results from the limited number of available maintenance personnel and space restrictions for nearby component repair or service. !
Historic data indicate roughly 15 days per year, or 17% and 25 % of the annually scheduled downtime for PWRs and BWRs, respectively, can be atsbuted to the cumulative impact of new regulatory requirements. Assuming the contribu-tion of each regulatory requirement to the incremental downtirne equals the overall percentage increase, one can assign a prorated share to that requhement (i.e.,17% for PWRs,25% for BWRs, or roughly 20% for LWRs in general). For example, if a regulatory iequirement requires one-week of reactor shutdown time,1.2 days (PWRs),1.8 days (BWRs), or 1.4 days (LWRs) of additional downtime and, thus, replacement power costs would accrue.
5.7.7.2 Premature &cIlity Closing Several nuclear power plants have been voluntarily shut down prior to the expiration of their operating licenses.
Normally, a decommissioning cost of approximately $0.3E+9 (1993 dollars) would be associated with an end-of-life shutdown (see Section 5.7.6.1). However, if a proposed regulatory requirement is expected to result in a premature shutdown, this cost is shifted to an earlier time with an associated net increase in its present value. For example, if a plant with an estimated t years of remauung life is prematurely closed, the net increase in present value, for a real discount rate of r, becomes ($0.3E +9) [1 - 1/(1 +r)'].
Thus, a plant closed 20 years early will incur an ad.iitional cost of $0.2E+8 for a 7 % real discount rate.
5.7.8 Industry Operatiore This section provides procedures for estimating industry's incrementa! costs during the operating phase (i.e., after implementation) of the proposed action. The incremental costs measuie the additional costs to industry imposed by the proposed action; they are costs that would not have been incurred in the sbsence of the action. Reduction in the net cost NUREG/BR-0184 O
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(i.e., cost savings) is algebraically positive; increase (i.e., cost accrual) is negative (viewed as negative cost savings).
Both NRC and Agreement State licensees should be addressed, as appropriate.
l In general, there are three steps that the analyst should follow in order to estimate industry operation costs:
S. 1 - Estimate the amount and types of plant equipment, materials, and/or labor that will be affected by the proposed action.
I Step 2 - Estimate the associated costs.
Step 3 - Discount the costs over the remaining lifetimes of the affected facilities, then sum (see Section B.2).
Cos': incurred for operating and maintammg facilities may include, but are not limited to, the following:
- maintenance of land and land-use rights
- maintenance of structures e operation and maintenance of hydraulic, pneumatic, and electrical equipment j
- scheduled radioactive waste disposal and health physics surveys e
,e mN o scheduled updates of records and procedures 4
- scheduled inspection and test of equipment
- scheduled recertification/ retraining of facility personnel 1
- associated recurring administrative costs e scheduled analytical updates. j The FORECAST computer code for regulatory effects cost analysis (Lopez and Sciacca 1996) allows user input for industry (licensee) operation costs.
For the standard analysis, the analyst should proceed as follows:
Step 1 - Estimate the amount and types of plant equipment, materials, and/or labor that will be affected by the proposed regulation, including professional staff time associated with reporting requirements and compliance activities.
Possible impacts on a facility's capacity factor should be considered. The analyst may consult with engineering and costing experts, as needed. The analyst could seek guidance from NRC contractors, architect engineering firms, or utilities.
Step 2 - Estimate the associated operation and maintenance costs. The analyst should consider direct and indirect effects of the action; for example, the action could have an impact on plant labor, which, in turn, could affect administrative costs.
Step 3 - Discount the total costs over the remaining lifetime of the affected facilities (see Section B.2).
/
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5.53 NUREG/BR-0184
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Much of the discussion on industry implementation costs in Section 5.7.7 for non-reactor facilities applies here for operation costs. Again, the analyst will generally not find consolidated cost information comparable to that for power reactors facilities. As before, Sections C.7-C.10 provide some limited data. However, the analyst may again need to rely !
on " engineering judgement," although specific data may be available through direct contact with cognizant industry / ,
contractor personnel.
l l
For a major effort beyond the standard analysis, the analyst should seek specific guidance from contractor or industry i sources experienced in this area. The user may wish to use contractors who have developed explicit methodologies for i
estimating operat ng and maintenance costs. The following references can provide useful information for industry opera-tion costs: Budwani (1969); Carlson et al. (1977); Clark and Chockie (1979); Eisenhauer et al. (1982); EPRI (1986);
NUS Corporation (1969); Phung (1978); Roberts et al. (1980); Stevenson (1981); and United Engineers and Constructors, Inc. (1979; 1988a, b).
5.7.9 NRC Implementation Once a proposed action is defined and the Commission endorses its application, the NRC will incur costs to implement the action. Implementation costs refer to those " front-end" costs necessary to realize the proposed action. All costs associated with pre-decisional activities by the NRC are viewed as " sunk" costs and are excluded from the NRC implementation costs. Reduction in the net cost (i.e., cost savings) is algebraically positive; increase (i.e., cost accrual) is negative (viewed as negative cost savings).
Implementation costs to the NRC may arise from developing procedures, preparing aids, and taking other actions to assist in or assure compliance with the proposed action." The analyst should determme whether the proposed action will be implemented entirely by the NRC or in cooperation with one or more Agreement States. Implementation costs shared by Agreement States may reduce those of the NRC and are discussed in Section 5.7.11.
NRC implementation costs include only the incremental costs resulting from adoption of the proposed action. Examples of these costs are as follows:
developing guidelines for interpreting the proposed action and developing enforcement procedures e
preparing handbooks for use by the NRC staff responsible for enforcement and handbooks for use by others responsible for compliance o supporting and reviewing a licensee's change in technical speciScations e conducting initial plant inspections to validate implementation.
Sciacca (1992) and the FORECAST computer code for regulatory effects cost analysis (Lopez and Sciacca 1996) assist the analyst in calculating these and "other" implementation costs implementation costs may include labor costs and cc.tead, purchases of equipment, acquisition of materials, and the cost of tasks to be carried out by outside contractors. Equipment and materials that would be eventually replaced during operation should be included under operating costs (see Section 5.7.10) rather than implementation costs.
Three steps are necessary for estimating NRC implementation costs:
NUREG/BR-0184 5.54 O
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b 4 Step 1 - Determine what steps the NRC must take to put the proposed action into efect.
Step 2 - Determine the requirements for NRC stas, outside contractors, materials, and equipment.
Step .I- Estimate the costs of the required resources, discount if appropriate, then sum (see Section B.2).
Implementation is likely to afect a number of NRC branches and oEces. For example, the OEce of Nuclear Regulatory Research (RES) may dewlop a regulatory guide, the OEce of Nuclear Reactor Regulation (NRR) may review any licensee submissions, and the NRC Regional OEces may inspect against some portion of the guide in operating facilities, la developing estimates for the implementation costs, the analyst is encouraged to contact all of the NRC components likely to be afected by the proposed action.
For the standard analysis, the analyst should identify the major tasks that must be petformed to get the proposed rule implemented, major pieces of equipinent (if any) that must be acquired, and major costs of snaterials. Major tasks are then =====d to estimate the approximate lew! of esort (in professional stas person-hours) a~~wy to complete them.
The number of person-hours for each task is multiplied by the appropriate NRC labor rate and then summed over all of the tasks. In 1996 dollars, the average NRC labor rate (salary and benefits plus allocated agency management and support) is
$67.50/ person-hr."
Similarly, the costs to complete tasks that would be contracted out also need to be estimatal. In order to obtam a waaanahly good approximation of contractor costs, the analyst should contact the NRC component that would be responsi-ble for contractag for the tasks. Finally, the costs of major pieces of equipment and quantities of materials are added to the labor and contract costs.
When other data are ucavailabte, the analyst may assume as an approximation that for a non-controversial amendment to an existing rule or regulatics implementation will require the fodowmg: a total of one professional NRC stas person-year at a cost of $122,000/ person-year (in 1996 dollars), no additional equipment, and no additional materials. For a new rule or segulation, it is much more discult to supply a tough but reasonable estimate of the implementation cost, because the level of efort and types and quantities of ra-hiaary and materials can vary dramatically. One recourse would be to use as a proxy the implementation costs for a recently adopted regulatory requhement that is similar to the proposed measure.
'the relatiw similarity of the two requirements should be judged with respect to the efort required to implement the proposed measure.
For a major esort beyond the standard analysis, a more detailed and complete accounting would be ==~% The analyst can request the responsible NRC oEce to provide available information, such as paper submittals or secords of initial l i
inspections.
i 5.7.10 NRC Operation AAer a proposed action is implemented, the NRC is likely to incur operating costs. These are the recurnng costs that are necessary to ensure continued compliance. For czample, addag a new regulation may require that NRC perform periodic inspections to ensure complhece. ~1he analyst should determine whet 5r operations resulting from the proposed action will be conducted entirely by the NRC or in cooperation with one or more Ag sement States. Reduction in the net cost (i.e., cost savings) is algebraically positiw; increase (i.e., cost accrual) is negatiw (viewed as negatiw cost savings).
"Ibere are three steps for estimating NRC operating costs:
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Step 1 - Determine the activities that the NRC must perform after the proposed action is implemented.
O l
1 Step 2 - Estimate NRC staff labor, contractor support and any special equipment and material required.
Step 3 - Estimate the costs of the required resources, discount (usually over the remaining lifetimes of the affected ,
facilities, as for industry operation costs) to yield present value, then sum (see Section B.2). )
In deternumng the required post-implementation activities, the analyst should carefully examine the proposed action, asking such questions as the following: {;
How is compliance with the proposed action to be assured?
Is periodic review of industry performance required?
- What is an appmpriate schedule for such review?
Does this action affect ongoing NRC programs, and, if so, will it affect the costs of those programs?
Since recurring costs attributable to the proposed action may be incurred by several NRC branches and offices, the analyst is encouraged to contact all of the NRC components likely to be affected. The FORECAST computer code for regulatory effects cost analysis (Impez and Sciacca 1996) allows user input for NRC operation costs.
For the standard analysis, the analyst should obtain estimates of the number of full-time equivalent pmfessional NRC staff person-hours that would be required to ensure compliance with the proposed rule. Each person-hour should be costed at
$67.50/ person-hr (in 1996 dollars) (refer to endnote 14). Major recurring expenditures for special equipment and materials, and for contractors, should be added. Since operating costs are recurring, they must be discounted as described in Section B.2, usually over the remaining lifetimes of the affected facilities.
A major effort beyond the standard analysis would pmceed along the lines described above, except that greater detail would be provided to account for acquisitions of special equipment and materials.
5.711 Other Government This ratribute measures costs to the federal government (other than the NRC) and state (including Agreement State) and local governments. The discussion parallels that for NRC implementation and operation in Sections 5.7.9-5.7.10.
Reduction in the net cost (i.e., cost savings) is algebraically positive; increase (i.e., cost accrual) is negative (viewed as negative cost savings).
Implementation costs to the federal (non-NRC) government and to state and local governments may arise fmm developing procedures, preparing aids, supporting license amendments, and taking action to assure compliance with the proposed action. For example, placing roadside evacuation route signs for the possibility of a radioactive release from a nearby power reactor would require expenditures from selected government agencies. As another example, requiring criminal investigation checks for nuclear reactor personnel may require resources of the Federal Bureau of Investigation. When estimating the implementation costs, the analyst should be aware that they may differ between Agreement and non-Agreement States. Such differences should be taken into account in preparing cost estimates.
Three steps are needed to estimate the other government implementation costs:
NUREG/BR-0184 5.56 O
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v Step 1 - Determine what steps the other governments must take to put the proposed action into effect.
Step 2 - Determine the requirements for government staff, outside contractors, materials, and equipment.
Step 3 - Estimate the costs of the required resources, discount if appropriate, then sum (see Section B.2).
Implementation is likely to affect a number of government branches and offices. In developing estimates for the
. implementation costs, the analyst is encouraged to contact all of the government components likely to be affected by the proposed action. The FORECAST computer code for regulatory effects cost analysis (14pez and Sciacca 1996) allows input for other government costs.
For the standard analysis, the analyst should identify the major tasks that must be performed to get the proposed rule implemented, major pieces of equipment (if any) that must be acquired, and major costs of materials. Major tasks are then assessed to estimate the approximate level of effon (in professional staff person-hours) necessary to complete them.
The number of person-hours for each task is multiplied by the appropriate labor rate and then summed over all of the tasks.
Similarly, the costs to complete tasks that would be contracted out also need to be estimated. In order to obtain a reasonably good approximation of in-house and contractor costs, the analyst should contact the government agencies that l would be responsible for carrying out or contracting for the tasks. Finally, the costs of major pieces of equipment and quantities of materials are added to the labor and contract costs.
fN After a proposed action is implemented, the federal (non-NRC) government and state and local governments may incur
, (V) operating costs. These are the recurnng costs that are necessary to ensure continued compliance new regulation may require that other government agencies in addition to the NRC perform periodic inspec mns to ensure compliance. The analyst should determine whether operations resulting from the proposed action will be s onducted entirely by the NRC or in cooperation with one or more other government agencies.
The three steps for estimating the other government operating costs are Step 1 - Determine the activities that the other governments must perform after the proposed action is implemented.
Step 2 - Estimate government staff labor, contractor support, and any special equipment and material required. l l
Step 3 - Estimate the costs of the required resources, discount (usually over the remaining lifetimes of the affected facili-ties, as for NRC operation costs) to yield present value, then sum (see Section B.2),
in determining the required post-implementation activities, the analyst should carefully examme the proposed action, ask-ing such questions as the following:
- Does compliance with the proposed action require non-NRC cooperation?
1
- Is periodic review of industry performance required beyond that of the NRC7
- What is an appropriate schedule for such review?
- Does this action affect ongoing government programs, and, if so, will it affect the costs of those programs?
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5.57 NUREG/BR-0184
i Value-Impact Since recurring costs attributable to the pmposed action may be incurred by several government branches and offices, the O.l analyst is encouraged to contact all components likely to be affected.
For the standard analysis, the analyst should obtain estimates of the number of full-time equilent professional staff person-hours that would be required to ensure compliance with the proposed rule. Each person-hour should be costed at the appropriate labor rate (an average NRC labor rate of $67.50/ person-hr [in 1996 dollars] maybe used as a substitute if no more specific value is available [ refer to endnote 14]). Major recurring expenditures for special equipment and materials, and for contractors, should be added. Since operating costs are recurring, they must be discounted as described in Section B.2, usually over the remaining lifetimes of the affected facilities.
I A major effort beyond the standard analysis would proceed along the lines described above, except that a more detailed '
and complete accounting would be expected. The analyst could request the responsible government agencies to provide available information.
5.7.12 General Public This attribute measures costs incurred by members of the general public, other than additional taxes, as a tesult ofimple-mentation of a proposed action. Taxes are viewed simply as transfer payments with no real resource commitment from a societal perspective. Reduction in the net cost (i.e., cost savings) is algebraically positive; increase (i.e., cost accrual) is negative (viewed as negative cost savings).
Typically, costs to the general public cover such items as increased cleaning due to dust and construction-related pollutants, property value losses, or inconveniences, such as testing of evacuation sirens. Care must be taken not to double count for general public and other govemment costs. If a cost could be assigned to either group, it should be assigned where more appropriate, the analyst remembering not to account for it again in the other attribute.
The two steps to estimate costs to the general public are as follows:
Step 1 - Identify the adverse impacts incurred by the general public to implement the proposed action.
Step 2 - Estimate the costs associated with these adverse impacts, discount if appropriate, then sum (see Section B.2).
This attribute is not expected to be one commonly affected by regulatory actions. However, if relevant, the standard ;
analysis would require the analyst to identify the major activities to implement the proposed actior. that will result in '
adverse impacts to the general public. Public records or analogous experience from other commu litics could be used as information sources to estimate the costs to the general public. '
i 5.7.13 Improvements in Knowledge l l
This attribute relates primarily to proposals for conducting assessments of the safety oflicensee activities. At least four major potential benefits are derived from the knowledge produced by such assessments:
e I improvements in the materials used in nuclear facilities I
l e improvement or development of safety procedures and devices o
production of more robust risk assessments and safety evaluations, reducing uncertainty about the relevant processes j i
NUREG/BR-0184 5.58 O\I l
Value-Impact
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/3 i
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N>
- improvement in regulatory policy and regulatory requirements.
To the extent that the effects of regulatory actions can be quan:ified, they should be treated under the appropriate quantita-tive attributes. On the other hand, if the effects from the assessments are not easily quantified, the analyst still has the burden of justifying the effort and providing some indication of its effect. If necessary, this justification would be expressed qualitatively under this attribute. An effort should be made to identify the types of values and impacts that are likely to accrue and to whom.
Consider the following statement:
This assessment effort has a reasonable prospect of reducing our uncertainty regarding the likelihood of contain-ment failure resulting from hydmgen burning. Such an accident may be a significant source of risk. 'Ibe know-ledge from the proposed assessments would enable us to assess more accurately the overall accident risk posed by nuclear reactors, and this in turn should benefit the public through better policy decisions.
While this statement describes why the proposed assessment is needed, no information is provided for evaluating the merits of the proposed assessment.
Providing answers to the following questions would help to fill this information gap:
- What are the likely conscquences of a hydrogen-burning accident?
- To what extent would the proposed assessment reduce the uncertainty in the likelihood of a hydrogen-burning
( , accident?
%J
- Given our current information, what is the contribution of hydrogen burning to overall accident risk?
The above questions are specific to a particular topic. For the broader problem of providing a value-impact analysis of an assessment proposal, it is recommended that the analyst be responsive to the following list of more general questions:
- Whzt are the objectives?
- If the assessment is successful in meeting its objectives, what will be the social benefits?
- Is there a time constraint on the usefulness of the results?
- Who will benefit from the results, by how much, and when?
- What is the likelihood that the assessment will fail to meet its objectives within the time and budget constraints?
- What will be the social costs (and benefits) if the assessment is not successful, or if the assessment is not undertaken?
5.7.14 Regulatory Efficiency Regulatory efficiency is an attribute that is frequently difficult to quantify. If it can be quantified, it should be included under one or more of the other quantifiable attributes. If quantification is not practical, regulatory efficiency can be treated in a qualitative manner under this attribute. For example achieving consistency with international standards groups o may increase regulatory efficiency for both the NRC and the groups. Ilowever, this increase may be difficult to quantify.
I I v
5.59 NUREG/BR-0184
Value-Impact j
O If necessary, this justificaion would be expressed qualitatively under this attribute. An effon should be made to identify the types and recipients of values and impacts likely to accrue. If the proposed NRC action is expected to have major cffects on regulatory efficiency, then a proper evaluation of these effects may require a level of effon commensurate with their magnitude. This may mean expending resources to obtain the judgments of expens outside of the NRC if the neces-sary expenise is not available in-house.
i To obta n unful information, the analyst can solicit expert opinion in a number of ways. A general discussion of those methods ind others is found in Quade (1975), especially Chapter 12, "When Quantitative Models are inadequate." One ;
way i io convene the expens in a round-table discussion with the objective of reachmg a consensus. This technique has j
- scane of the drawbacks of a committee meeting--often the assumptions are not made explicit, and strong-willed (or strong- 1 voiced) individuals often carry undue v,eight.
l Another way of pooling expert opinion in a systematic manner is to use one of the numerous procedures for iterative group :
decision-makmg. For example, the Delphi technique (Dalkey and Helmer 1963; Humphress and lewis 1982) is a proce-dure that features an anonymous exchange of information or expen opinion. This approach is designed to encourage the modification of earlier answers by each expen so that a group consensus can be achieved. Even if consensus is not achieved, information is produced that allows the analyst to compile statistical estimates of the responses.
Whether the assessment is performed by a panel of experts or by the analyst, the following are questions that might be considered in order to focus on that assessment:
- Does this action conflict with any other NRC/ federal / state directives?
Are there any nuclear facilities for which (or conditions under which) this action might have unexpected or undesirable consequences?
- Do you foresee any major enforcement problems with this action or regulation?
- What son of adjustments might indastry undertake to avoid the regulation's intended effects?
- How will the regulation impact productivity in the nuclear / electric utility industries?
How will this action affect facility licensing times?
How will this action affect the regulatory process within the NRC (and/or other regulatory agencies)?
5.7.15 Antitrust Considerations This qualitative attribute is not expected to be one commonly affected by re alatory actions. However, the NRC does have a legislative mandate in Section 105 of the Atomic Energy Act to uphi d the antitrust laws. Thus, this attribute can be relevant for those proposed actions which may potentially violate the m cust laws. If applicable, antitrust considera-tions should be explored with the NRC Office of the General Counsei carly in the analysis to preclude analyzing an issue clearly in conflict with these laws. If antitrust considerations are involved, and it is determined that antitrust laws would be violated, then the proposed action must be reconsidered and, if necessary, redefined to preclude such violation.
NUREG/BR-0184 5.60 O
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Value-Impact (v )
5.7.16 Safeguards and Security Considerations Safeguards and security considerations include protection of the common defense and security and safeguarding restricted data and national security information. In more practical terms, this mee.ns providing adequate physical security and safeguards systems to prevent the diversion of certain types of fissionable and radioactive materials, the perpetration of acts of radiological sabotage, and the theft by unauthodzed individuals of restricted data or national security information.
The NRC has a legislative mandate in the Atomic Energy Act to assure the objectives mentioned above. Through its regulations and regulatory guidance, the NRC has established a level of protection deemed to satisfy the legislative mandate. As is the case for adequate protection of the health and safety of the public, this level of protection must be maintained without consideration of cost.
While quantification of safeguanis and security changes may be difficult, the analyst should attempt quantification when feasible. If this pmcess is impossible, the analyst may proceed with a qualitative analysis under this attribute.
Section 5.7.14, where methods of evaluating expert opinion are discussed, may be helpful.
5.7.17 Environmental Considerations l l
Section 102 of the National Environmental Policy Act (NEPA) requires federal agencies to consider environmental impacts in the performance of their regulatory missions. NRC's regulations implementing NEPA are in 10 CFR Part 51. Any documentation prepared to satisfy NEPA and Part 51 should be coordinated with any regulatory analysis documentation covering the same or similar subject matter as much as possible.
O J Environmental impacts can have monetary effects (e.g., environmental degradation, mitigation measures, environmental !
enhancements), which could render potential alternative actions unacceptable or less desirable than others. Therefore, at a l minimum, such effects should be factored into the value-impact analysis, at least to the extent of including a summary of the results of the envimnmental analysis.
Many of the NRC's regulatory actions are subject to categorical exclusions as set forth in 10 CFR 51.22. In these cases, detailed environmental analyses are not performed, and there will be no environmental consideration to factor into the reg-ulatory analysis, in some cases, a generic or programmatic environmental impact statement (EIS) is prepared. If such is the case, Section 5.3 of the Guidelines allows portions of the EIS to be referenced in lieu of performing certain elements of the regulatory analysis. In the remaining cases, it may be that the regulatory analysis alternative being considered will initiate the requirement for review of environmental effects. For purposes of the regulatory analysis document, the preferred approach to be used in this situation is to perform a preliminary environmental analysis, identifying in general terms anticipated environmental consequences and potential mitigation measures. The results of this preliminary analysis should be quantified under the appropriate quantitative attributes, if possible, or addressed qualitatively under this attribute, if not quantified.
5.7.18 Other Considerations There may be other considerations associated with a particular proposed action that are not captured in the preceding descriptions. Possible examples might include the way in which the proposed action meets specific requirements of the Commission, EDO, or NRC office director that requested the regulatory analysis; the way in which the proposed action would help achieve NRC policy; or advantages or detriments that the proposed action would have for other NRC programs ,
and actions. If quantifiable, the effect should be included in essentially the same way as in the quantitative attributes. i Because such considerations would be expected to be unusual, some additional discussion in the regulatory analysis document should be provided.
(a i
)
i a 1 5.61 NUREG/BR-0184 i
I
Value-Impact Re analyst needs to give thoughtful consideration to the possible effects of the proposed action. Some of the effects may O
not be immediately obvious. He analyst may wish to consult with other knowledgeable individuals to aid in the identifica-tion of all significant effects. These considerations need to be presented clearly to facilitate the reader's understanding of the issues.
When quanti 6 cation of effects is not feasible, the analyst may still be able to provide some indication of the magnitude to facilitate comparison among alternatives, and comparison with quantifiable attributes. Comparative language (greater than, less than, about equal to) can be very helpful in achieving this objective, as long as the analyst can make the neces-sary judgements. Consultation with experts or other knowledgeable individuals may be required.
5.8 Summarization of Value-Impact Results Having completed the value-impact analysis for one or more alternatives of the proposed action, the analyst should sum-manze the results for each alternative using a summary table such as that shown as Figure 5.1. Such a tabular
'ntle of Proposed Action / Date Summary of Pmblem and Pmoosed Solution:
- "I * **" * ***'
Quantitative attribute Low
- Best* High*
Public Health
. Accident Occupational Routine l Offsite Pmperty Onsite Implementation Industry Operation Implementation NRC l Operation l Other Government General Public NET VALUE (Sum) ;
(a) Low estimates correspond to the worst case, i.e., highest costs and lowest benefits, relative to the baseline case.
(b) Best estimates are normally the expected value, but could be other point estimates st'th as the mean or median (see Section 4.3 of the Guidelines).
(c) High estimates correspond to lowest cost estimates and highest benefit estimates.
Comments: Discuss any other attributes considered, compliance with Safety Goal guidance, special considerations, etc.
Hgure 5.1 Summary of value-impact results NUREG/BR-0184 5.62 O
l l
, Value-Impact O
I presentation provides a uniform format for recording the results of the evaluation of ali quantitative attributes plus a comments section to discuss other attributes considered, compliance with the Safety Goal guidance, special considerations, etc. It displays the results for the net-value measure, discussed in Section 5.2.
All dollar measures should be present valued and expressed in terms of the same year. This may require conversion of ,
2 some dollar values from whatever years in which they have been expressed to one common year. Sciacca (1992) describes l techniques for these conversions. He Gross Domestic Product (GDP) price de8ator can be used to convert historical l nominal dollars to dollars of one common year. Financial publications, such as National Eronomic Dends by the Federal Reserw Bank of St. Louis, supply implicit price de8ators for the GDP, through the current year. GDP price de8ator information from the Federal Reserve Bank of Chicago is also available at the following Internet address:
http:// gopher. great-lakes. net:2200/0/ partners /ChicagoFed/econind/.
4 When reconting the low and high estimates for an attribute, the analyst should generally record tha 1s, est and highest estimates if multiple estimates are made. For example, suppose the analyst calculated a best estimate of -$5.0E+5 for NRC implementation cost (the negative value indicates the cost will be an expense rather than a savings). The analyst then
, performed two separate sensitivity analyses, obtaining the following sets oflow (nwre negative) and high (less negative) estimates:
Low Estimate Hiah Estunate Sensitivity A -$7.5E +5 -$2.5E+5 Sensitivity B -$1.0E + 6 -$3.0E+5 The analyst should record the lowest (most negative) and highest (least negative) estimates in Figure 5.1 (i.e., -$1.0E+6 ll (p) and -$2.5E+5, respectively), even though each comes from a different sensitivity analysis.
t/
The net value is the required value-impact measure (see Section 5.2). Its calculation is the sum of the present value of all the quantitative attributes. Information on computing present value is in Section B.2. A positive net value result indicates an overall cost savings for the proposed action. A negative net value result indicates the opposite. As mentioned in Section 5.2, the net value is an absolute measure, resecting the magnitude of the proposed action's contribution toward the
- specified goals. The results of the value-impact assessment can be displayed as a ratio and in tables and/or graphs, in addition to a summary table for additional perspectives.
$ 5.9 Endnotes for Chapter 5
- 1. Section 4.4 of the Guidelines allows the analyst to display the results of a value-impact analysis as a ratio of values to impacts, all expressed in dollars. The numerator would sum the estimates for all quantifiable attributes classified as values, while the denominator would do likewise for impacts. Section 4.4 of the Guidelines views a value-impact ratio as supplemental to the net value, not as a replacement, i
- 2. He term ' equation" is loosely used to indicate anything from a single mathamatical expression (e.g., one for a major fire at a non-reactor facility) to a complete computer analysis (e.g., a core damage assessment for a power reactor).
- 3. The double index notation indicates that an initiating event j can lead to several accident sequences 1.
O i V 5.63 NUREG/BR-0184 4
a w
l i
Value-Impact I 4.
O l l
Level 1 analyses generally produce a list of core-damage accident sequences, together with the overall core-Amage accident frequency as their final product. Level 2 analyses take the Level I analyses one step further by evaluating the containment response to the accident sequences and the associated contamment release magnitudes. Ievel 3 analyses take the Level 2 analyses one step further by evaluating the public risk associated with the centamment ,
release frequencies and magnitudes. As a result, Level 3 analyses are the preferred tools for evaluating the effect of l a proposed action on public risk. l
- 5. Developed by the Southwest Research Institute, San Antonio, Texas.
- 6. An error factor f is used as follows to estimate upper and lower bounds, presuming a positive value for the best estimate: i I
Upper Bound = Best Estimate x f !
Lower Bound = Best Estimate / f l
- 7. As discussed in Section 5.7.1.1, public health (accident) may be affected through a mitigation of consequences instead of (or as well as) a reduction in accident probability.
- 8. Andrews et al. (1983) provide a conceptual discussion of assessing the risk for this type of proposed action
- 9. The equations included in this Handbook (e.g., Section 5.7.1.3) apply a discounting term to doses associated with both implementation and operational impacts. In practice, the implementation dose may be of such short duration that discounting is not necessary. Its inclusion here is in recognition that, in some cases, implementation may extend over a longer period than one year.
- 10. NRC has required its contractors to estimate onsite dose rates in the Surry and Grand Gulf risk assessments during low power and shutdown operations (Brown et al.1992; Jo et al.1992).
- 11. Based on ANL estimates, a cleanup period as long as 10 years may be needed following a major power reactor accident (see Section 5.7.6.1). Ieng-term doses will occur over some portion of this time.
- 12. Accidents at non-reactor nuclear facilities could also lead to the need for replacement services of the same type provided by the facility where the accident occurred. 1
- 13. NRC implementation costs associated with facility closure may be increased if the facility closes prematurely (see I Section 5.7.7.2).
- 14. The $67.50 hourly rate is derived from June 1996 data and the technique described in Abstract 5.2 of Sciacca (1992). 1 1
l l
NUREG/BR-0184 5.64 O 1
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O NUREG/BR-0184 6.4
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, Appendix A 4
i Regulatory AnalysisIssues a
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i Appendix A
[
Regulatory AnalysisIssues
[
his appendix addresses three topics of particular interest in' connection with the performance of regidatory analyses.
Owing to the special nature or extensiwness of these topics, it was judged best to discuss them here rather than in the main
. body of this Handbook, as has been done with other issues. The topics are human facion issues, cuandative accounting of past and ongoing sahty improvements, and use of industry risk and cost estimates A.1 Human Enctors Issues Regulatory analyses involving proposed actions related to human factors issues chen prove to be diEcult to quantify, especially with regard to risk-related attributes. This degree of diEculty varies to the extent that the human factors issue in " concrete" or " abstract." For example, an issue propssing to clarify standard procedures for hardware inapartian can be perceived as fairly concrete, inspection personnel can be expected to perbrm more ediciently with less likelihood of -
error during the inspection procedure. This would decrease the likelihood of overlooking a hardware desset. Such an issue can be translated into a reduced unavailability for selected hardware companants, several of which most likely appear in a facility risk equation. - For such a human factors issue, the expected improvement can be tressed as an improvement in the reliability of the hardware itself. Thus, this " concrete" human factors issue can be analysed in a mannar aimilar to any other hardware issue.
As an " abstract" example, consider a human factors issue pmposing to revise management guidelines for a power plant.
Disculty is foreseen in directly linkmg this action to parameters in a plant risk equation. One approach might he to assume some small improvement in the portion of the unavailability due to human error in each risk parameter a appro-priate. De analysis then could proceed as in a hardware issue, except that many parameters might be affected, thereby complicating the calculations. Studies completed by Ramanta et al. (1981,1989) and Andrews et al. (1985), diacnamad in .
Section A.1.1, provide results which can facilitate these types of calculations.
As an alternative, an approach similar to that discussed in Section 5.6.2 may be appropriale. For fairly " nebulous" issues (i.e., ones where the reductions in accident frequency [and/or risk) are diEcult to quantify duectly via a facility risk equation) expert judgment of the changes in the accident frequency (and/or risk) can be based on the total accident frequency (and/or risk). Employing informal procedures or a formalized one such as the Delphi method (Dalkey and Helmer 1%3; Humphress and Lewis 1982), the analyst can obtain a consensus estimate of the percent change in total acci-dent frequency (and/or risk) due to implementation of the proposed action. This may be the best that can be done for the more _" abstract" human factors issues.
Several studies have been conducted to address quantification of human error probabilities (HEPs) for nuclear power plant sisk analyses. The initial standard for human error analysis, subsequently named the 'Ibchnique for Human Error Race Prediction (THERP), was established by the complementary documenta NUREG/CR-1278 (Swain and Ontemann 1983) and NUREG/CR-2254 (Bell and Swain 1983). Swain and Outtmann (1983) developed a handbook of human performance models and procedures for estimating HEPs, including numerical values, for application in nuclear power plant risk 0
A.1 NUREG/BR-0184
Appendix A analyses. In its sister document (NUREG/CR-2254), Bell and Swain (1983) detailed a standard procedure to conduct a e
human reliability analysis for nuclear power plants, emphasizing an event tree approach which utilizes results from NUREG/CR-1278. Swain (1987) supplemented the THERP with a simplified version in NUREG/CR-4772, intended "to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other performance characteristics which are sufficiently accurate for many probabilistic risk assessments."
l Additional studies which can assist the analyst in performing a regulatory analysis, particularly the value-impact portion, for a human factors issue can be grouped into two categories:
- 1. Documents addressing methods to estimate HEPs, sometimes including numerical results for applying these methods (see Section A.I.2). The previous studies plus a trio by Stillwell et al. (1982), Seaver and Stillwell (1983), and Comer et al. (1984) am examples of these " methods" documents.
- 2. Documents presenting the results of quantifying the impact of HEPs on a nuclear power plant's overall core-melt fre-quency and/or public risk (see Section A.I.1). A pair of studies by Samanta et al. (1981,1989) and one by Andreu et al. (1985) are examples of these "results" documents.
Documents from each group have been reviewed, and summaries are pmvided in the remainder of this appendix section.
We begin with studies from the second group.
A.1.1 Restilts Documents In a pair of studies, Samanta et al. (1981,1989) evaluated the sensitivity of selected risk parameters to changes in HEPs for a pair of representative PWRs. The first study (NUREG/CR-1879 [Samanta et al.1981]) quantified the effect of changing HEPs for the Surry PWR on the following parameters: system unavailability, accident sequence frequency, core-melt frequency, and release category frequency. The Human Error Sensitivity Assessment of a PWR (HESAP) computer code was developed to model the human errors in fault trees based on the Sutry plant as modeled in WASH-1400 (NRC 1975a). HEPs were both increased and decreased by factors of 3,10,20, and 30 relative to selected base-case values. Numerous tables and figures give the results of simultaneously varying all HEPs by these factors in terms of the changes in the four risk parameters listed above.
in addition, Samanta et al. (1981) estimated the sensitivity of core-melt and release category frequencies to changes in probabilities for generic classes of human error (e.g., operator error, maintenance error, and errors of omission /
commission). Also, individual human errors were ranked relative to one another in terms of their structural importance to core-melt frequency and their reliability importance to core-melt and release category frequencies (Vesely et al.1983).
The results are conveniently presented as tables and figures.
The second study (NUREG/CR-5319 [Samanta et al.1989]) updated the first using the more recent, and more detailed, Oconee PWR risk assessment performed by EPRI and Duke Power Co. (1984). Only the portion of the Oconee risk assessment pertaining to internal events was employed by Samanta et al. External events were not included. The effect of changing IIEPs on the following risk parameters was evaluated: accident sequence frequency, core-melt frequency, and core-melt bin frequency (somewhat analogous to release category frequency). Statistical methods were employed to esti-mate factors by which HEPs could be both increased and decreased realistically. Factors ranging as high as 26 were calculated, depending upon the type of human error (an additional degree of resolution relative to the first study).
O NUREG/BR-0184 A.2
Appendix A A
( )
%.J Human errors were divided into the following oserlapping categories for the sensitivity analysis:
o Eming - when the human error occurs relative to the accident initiating event or transient e Accident Initiator - which accident initiating event is related to the human error o System - the system in which the human error occurs o IVrsonnel - which individuals are responsible for the human error o Omission / Commission - whether the hutaan error is one where a needed action is not performed (omission) or one where an improper action is performed (commission) o Event 7)pe - relating the human error to the category assigned in the 0:once risk assessment (EPRI and Duke Power Co.1984) 1 i
o Location - where the personnel most responsible for the human error are located o Activity - which type of nuclear power plant activity relates to the huraan error o Dependence - whether or not the huraan error results from another human error l
A)o NRC Pmgram - which NRC inspection area may detect the occurren:e of the human error.
I 1
\V The sensitivity of the thste risk parameters mentioned above to changes in HEPs for these various categories are conven-l iently presented as figures in NUREG/CR-5319. All HEPs within each category were simultaneously varied relative to l i
the base-case value from the Oconee risk assessment. In addition, the effect of simultaneous variation of all HEPs on the three risk parameters was evaluated. He results we e compared with those from the first study.
Both these studies provide information which would be useful in human factors issues where categories of HEPs would be affected. For ex. ample, plant-wide impmvements in maintenance procedures or more stringent testing of reactor operators would be expected to reduce all HEPs falling within the appropriate categories. These two studies provide relative values for the change in selected risk parameters for such simultaneous variation of HEPs. Most human factors issues appear to j be of this " global" nature, hence the usefulness of the studies' results.
The NRC (NRC 1983b), with assistance from Pacific Northwest National Laboratory (PNNL) (Andrews et al.1983), has been systematically prioritizing generic safety issues since 1982, many of which involve human factors for nuclear power plants. Simple methods were initially established to handle human factors issues which fell into the "concirte" and
" abstract" categories discussed earlier in this appendix section. The earlier discussion summarizes the approach that was taken in the prioritization assessments. NUREG/CR-2800 and its supplements (Andrews et al.1983) provide numemus examples of human factors issues analyzed using these simple methods. In 1985, Andrews et al. conducted a study (NUREG/CR-2800, Supplement 3) in which they 1) developed an alternative approach to prioritizing human factors issues and 2) prioritized the elements of the 1983 Human Factors Program Plan (HFPP) developed by the NRC.
The development of the alternative human factors methodology by Andrews et al. (1985) involved investigation of four attributes of human factors analyses: 1) the general guidelines used by the decision-making panel in the initial prioritiza-tions,2) the impact of using alternate representative plants, 3) human factors modeling related to maintenance and plant availability, and 4) human factors data bases. For the first attribute, decision-making basis was documented in terms of
)
(d A.3 NUREG/BR-0184
i Appendix A plant-related guidelines, human error assumptions, independence of human factors issues, and cost guidance. For the O.
second attribute, the differences in core-melt frequency resulting from reducing HEPs for three different representative l plants, the Oconee and Calvert Cliffs PWRs and the Grand Gulf BWR, as modeled by their Reactor Safety Study l
Methodology Application Program (RSSMAP) studies (Kolb et al.1981; Hatch et al. 1981,1982) was quantified. For the !
third attribute, new maintenance and plant availability models were developed and tested. For the fourth attribute, !
available human factors data bases were examined and found to be only of limited use in prioritization analyses.
)
Andrews et al. (1985) also prioritized the following six elements of the 1983 HFPP: 1) staffing and qualifications,
- 2) training, 3) licensing examinations, 4) procedures, 5) man-machine interfaces, and 6) management and organization. ,
Eigl' teen generic safety issues were divided among the six elements. For each, expen opinion on the effects on HEPs and !
costs resulting fmm resolution was solicited through a structured series of questionnaires. The consensus changes in HEPs !
were transformed into public risk changes via the Oconee and Grand Gulf RSSMAP models. Public risk, industry, and NRC cost estimates for implementing the HFPP as a whole and for implementing each specific element were calculated and used to assign priorities to the six elements.
As in the studies by Samanta et al. (1981,1989), this study by Andrews et al. (1985) provides information which would be useful to human factors issues where categories of HEPs would be affected. It provides relative values for the change in core-melt frequency and public risk for simultaneous variation of HEPs. In addition, since a comprehensive program for human factors improvements has been examined, estimates of maximum possible reductions in public risk and increases in industry and NRC costs attainable by implementing such a program are available. Individual issues within each element of the HFPP were also examined, with their public risk reductions and industry and NRC cost increases evaluated.
Therefore, this information is available for several types of human factors issues.
A.1.2 Methods Documents In NUREG/CR-2255, Stillwell et al. (1982) :eviewed probability assessment and psychological scaling techniques that could be used to estimate human error probabilities in nuclear power plant operations. The techniques rely on expert opinion and can be used where data do not exist or are inadequate. An extensive literature search was performed, and the results are discussed under two categories: 1) subjective pmbability assessment, and 2) psychological scaling. While this report is primarily a qualitative overview of the various techniques, it provides useful background as to which ones would be appropriate and when, as well as serving as a reference document for additional information.
The first category examined by Stillwell et al. considered seven aspects of subjective probability assessment: 1) use of expert judgment for assessing probabilities,2) probabilistic assessment techniques,3) use of multiple experts in assessing probabilities,4) problems and biases in the assessment of subjective probability, 5) training probability assessors, 6) new methods for resolving inconsisat judgmeats, and 7) defining and structuring judgments. The second category compared
)
the following five techniques of psychological scaling, with emphasis on their validity and reliability: 1) paired compari- ;
sons, 2) ranking, 3) sorting, 4) rating, and 5) fractionation. l in a follow-on report (NUREG/CR-2743), Seaver and Stillwell (1983) described and evaluated the following five pro-cedures for employing expert opinion to estimate HEPs for nuclear power plant operations: 1) paired comparisons,
- 2) rankmg and rating, 3) direct numerical estimation,4) indirect numerical estimation, and 5) multiattribute utility measurement. The following criteria were used to evaluate these techniques: quality of judgments, difficulty of data collection, empirical support, acceptability, theoretical justification, and data processing. Quantitative guidance on the implementation of these procedures is provided, along with situational constraints (e.g., the number of HEPs to be estimated) which impact the choice of a procedure.
O I NUREG/BR-0184 A.4 i
_ _ _ _ _ _ _ - - _ _ __ _ _ _ _ _ - _ _ _ _. .~
i Appendix A ,
i .
o Third in this series of studies was NUREG/CR-3688, in which Comer et al. (1984) examined selected techniques for psychological scaling, first introduced by Stillwell et al. (1982) in NURECWR-2255. Two techniques-direct numerical estimation and paired comparison scaling-were evaluated in detail. Comer et al. answered the following 11 questions as i a result of their study:
- 1. Do psychological scaling techniques produce consistent judgments from which to estimate HEPs?
- 2. Do psychological scaling techniques produce valid HEP estimates?
- 3. Can the data collected using psychological scaling techniques be generalized?
- 4. Are the HEP estimates that are generated from psychological scaling techniques suitable for use in probabilistic risk assessments and the human reliability data bank?
- 5. Can psychological scaling procedures be used by persons who are not experts to generate HEP estimates?
- 6. Do the experts used in the psychological scaling process have confidence in their ability to make judgments?
- 7. Is there any difference in the quality of estimates obtained from the two scaling techniques?
- 8. Is there any difference in the results based on the type of task that is being judged?
- 9. Do education and experience have any effect of the expens' judgments?
- 10. How should the paired comparison scale be calibrated into a probability scale?
- 11. Can reasonable uncertainty bounds be estimated judgmentally?
l The HEPs for 35 BWR tasks that were estimated as part of the study are also presented.
These three studies provide guidance on the estimation of HEPs by expert judgment. Although intended for estimating HEPs directly, the techniques presented in these three studies are readily adapted to estimating changes in llEPs by expert j judgment, typically what is needed to quantify the value-impact of a human factors issue. 'Ibchniques such as these can be l used to estimate the changes in individual or families of HEPs. Subsequently, they can be combined with knowledge on l the overall effect of more global changes in HEPs on core-melt frequency and public risk as provided by studies such as those of Samanta et al. (1981,1989) and Andrews et al. (1985).
A.2 Cumulative Accounting of PaSt and Ongoing Safety Improvements When performing a regulatory analysis, an analyst should be aware of previous or ongoing safety improvements which already have impacted or bear the potential to impact the status quo for the issue being addressed. Incorporation of such improvements could be accommodated if there existed a " master" risk assessment (or a few " masters") deemed representa-tive of all facilities for which all previous safety improvements have been included and the baseline risk recalculated.
Since this currently is not practical, the analyst must resort to a "best effort" approach in accounting for preexisting or concurrent impacts, consistent with NRC policy regarding the treatment of voluntary activities by affected licensees (see NRC Guidelines Section 4.3).
O V
A.5 NUREG/BR-0184
1 l
l l
Appendix A I
Ol During Step 1 of the regulatory analysis (see Section 4.1), the analyst should make a thorough effort to identify any previous or ongoing safety improvements which may impact the issue under consideration. For example, an analyst I
addressing proposed improvements in diesel generator performance at pcrwer reactors should be aware of any diesel gen- ;
erator improvements already addressed in station blackout (SBO) considerations. To the extent possible, the analyst l should modify the risk equation of the plant chosen as representative to reflect the upgraded status quo from these other safety improvements. The analyst can then proceed to assess the difference between this new status quo and the proposed improvements from the issue under consideration. The analyst should also seek out and use (when appropriate) the most recent risk assessments (including IPE and IPEEE reports) affecting the facilities impacted by the issues under consideration (see Table 5.2).
An attempt to accommodate "dependences" between issues was informally tried during the Prioritization of Safety Issues Program (Andrews et al.1983). Issues of "high" rank were divided into " families" with similar issue resolutions (e.g.,
diesel generator reliability and SBO were assigned to an electrical family). The issues within each family were exammed for all pairwise combinations where Issue A was implemented before Issue B and vice versa. Within these families, few dependent pairs were found and, for those found, the dependent effects were generally small (< 10%). A similar approach could be taken, although the analyst may wish to consider greater than pairwise combinations if necessary.
A.3 Use ofIndustry Risk and Cost Estimates As a general rule, analysts can use risk and cost data prepared by industry sources provided the analyst can independently attest to the reasonableness of the data.
Table 5.2 in Section 5.6.1 lists nuclear power plant risk / reliability studies (other than IPE and IPEEE reports) for use in regulatory analyses for power reactors. Several studies have been performed by the nuclear industry (i.e., the utilities themselves and/or their contractors). Theoretically, some bias may exist depending upon the source of the study (NRC contractor or industry). Some indication of such bias may be obtained by comparing studies performed for the same plant by different sources. However, one would have to take care not to attribute differences to bias if plant changes, more recent data, or different analytical methods are the reasons for differing results. The issue of bias may often be rendered useless to debate since the analyst may not have a wide choice of representative plants with existing risk / reliability studies.
The analyst should always opt for the most representative plant, whether its risk / reliability study was performed by an NRC contractor or industry. The same considerations apply to regulatory analyses for non-reactor facilities, to the extent that representative risk / reliability studies are available (see Sections 5.6.1 and C.2.1.1).
Wider choice may be available to the analyst for cost estimates, and the analyst may be faced with different costs from equally valid sources. A sensitivity analysis may be best in which the analyst uses each set of costs for those attributes most strongly affected. However, should the analyst have reason to believe one set to be more representative than the other, the more representative set should be selected. The analyst may still use the other set in a sensitivity study should it be deemed appropriate.
i l
O NUREG/BR-0184 A.6 i
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- l O
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1 1 1 1
I I
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- Appendix B 1
l SupplementalInformation For Blue-Impact Analyses 5
i i
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s 4
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Q Appendix B 1
4
- SupplementalInformation For Value-Impact Analyses
- This appendix presents data on the number of operating power reactors and their remaining lifetimes, methods of eco-nomic discounting and present value calculation, data on occupational exposure experience at nuclear power plants and i some non-reactor facilities, additional cost information, and a description of the calculational snethod used to generate ;
Thble 5.3, " Expected Population Doses," br power reactor plant damage states. These can be used by the analyst to l support his evaluation of attributes durin the value-impact analysis portion of a regulatory analysis.
B.1 Numbers of Operating 1%er Reactors and Their Remaining Lifetimes Thble B.1 liars the numbers of operating power reactors and their semaining lifetimes relative to 1993. The lifetimes are based on tiie years in which the Operating Licenses cuntntly expire, as reported in NUREG-1350, Vol.4 (NRC 1992).
Thble B.1 lists the plants by ver. dor and reactor type.
'Ihble B.1 Numben and lifetimes of operating nuclear power plants !
Number of Average Remaining
- Reactor Supplier Type Operating Units (N) Lifetime (T) (years)*
Westinghouse PWR 52 25.4 General Electric BWR 37 23.3 Combustion PWR 15 23.7 Engineering Babcock and PWR 7 21.4 Wilcox N T(years)
All PWRs 74 24,7 AllIPNPa 37 23.3 All Plants 111 24.2 (a) Relative to 1993.
B.2 Economic Discounting and Calculation of Present Value To evaluate the economic consequences of proposed regulatory actions, the costs incuned or saved over a period of years must be summed.
a B.1 NUREG/BR-0184
Appendix B O I This summation cannot be donc directly because an amount of money available today has greater value than the same amount at a future date. There are several reasons for this difference in value:
the present amount of money can be invested and t'n mtal amount increased through accumulated interest certain consumption today is superior to contingent consumption in the future the option of present or future consumption is superior to future consumption alone.
A method known as
- discounting" is used to compare amounts of money expended at differen: times. The result of dis-counting is called the "present value," the amount of money that must be invested today to achieve a specified sum in the future. Tb perform the discounting procedure, the analyst must know three parameters:
- the discount rate e
the time period over which discounting is to be performed
- the amount of money or value that is to be discounted.
B.2.1 Discount Rate The appropriate discount rate to use is often a controversial issue in the application of value-impact analysis. NRC Guide-lines Section 4.3.3 states that the discount rates specified in the most recent version of OMB Circular A-94 are to be used in preparing regulatory analyses. Circular A-94 currently specifies use of a real discount rate (r) of 7% per year (OMB 1992). NRC Guidelines Section 4.3.3 further states that a discount rate of 3 % should be used for sensitivity analysis to indicate the sensitivity of the results to the choice of discount rate.
When the time horizon associated with a regulatory action exceeds 100 years, Section 4.3.3 of the Guidelines specifies that the 7% real discount rate should not be used. Instead the net value should be calculated using the 3 % real discount rate.
In addition, the results should be displayed showing the values and impacts at the time they are incuned with no discounting (see Section 5.7).
OMB Circular A-94 defines the term " discount rate" as the interest rate used in calculating the present value of expected yearly benefits and costs. When a real discount rate is used as specified in Section 4.3 of the Guidelines, yearly benefits and costs should be in real or constant dollars. Circular A-94 defines "real or constant dollar values" as economic units measured in terms of constant purchasing power. A real value is not affected tr/ general price inflation. Real values can be estimated by deflating nominal values with a general price index, generally the GDP deflator as discussed in Section 5.8.
B.2.2 Discrete Discounting The following formula is used to determine the present value (PV) of an amount (F,) at the end of a future time period:
PV = F/(1 + r)',
where r = the real annual discount rate (as fraction, not percent) t = the number of years in the futum in which the costs occur.
B.2
Appendix B G
For czample,*to determine how much $750 to be received 25 year: ('t) hence is worth today, using a 7% real discount rate
. (r), the formula yields PV = $750/(1 & .07)" = ($750)(0.184)
= $138 ;
Hble' B.2 contains values of the discount factor 1/(1 + r)' for discount rates (r) of 3% and 7% and for various values of t, l
. the number of years. Tb fmd the present value of a stream of costs and revenues, the analyst should record the costs and )
i revenues occurring in each year. Then, for each year, the net cost is determined by simply adding algebraically the costs '
and revenues for that year. After this has been done for each year, the net cost in each year is discounted to the present using hble B.2. The sum of these present values is the present value of the entire stream of costs and revenues. A sam-pie use of this formula in value-impact analysis would be in C mising the PV ofimplementation costs for industry and the NRC which occur in the future.
7he above formula is used for discounting single amounts backward in time. However, some of the costs encountered in value-impact analysis recur on an annual basis. These include not only industry and NRC operating costs, but also the monetized values of the annual per-facility reductions in routine public and occupational dose due to operation (see Sections 5.7.2 and 5.7,4). Such costs can be discounted by the use of the following annuity formula (only if they are the same amount for each time period):
PV = Ca[(1 + ry- 1]/r(1 + ry where C3 = ' identical annual costs r = .the real discount rate (as fraction, not percent) ,
- t = the number of years over which the costs recur.
l l
For example, if the increase in annual industry costs is $1,000, due to increased maintenance expenses, with a 7 % real discount rate for 20 years, starting at the present time, the present value of these costs is L PV = ($1,000)[(1 + .07)"- 1]/(.07)(1 + .07)"
= ($1,000)(10.6) = $10,600 Table B.3 contains values of the annuity discount factor: [(1 + r)'- 1]/r(1 + ry, for real discount rates (r) of 3 % and 7%
and for various values of t, the number of years over which the costs are incurred, in most cases, operating costs will start to be incurred at some date in the future, after which the real costs will be constant on an annual basis for the remaining life of the facility. To discount the costs in this situation, a combination of the above two methods or formulas is needed. For example, given the same $1,000 annual cost for a 20-year period at a 7 % real i discount rate, but starting five years in the future, the formula to calculate the PV is PV = ($1,000)[(1 + ry:- 1]/r(1 + ry (1 + ry:
where r = 7% discount mie (i.e., .07/yr) i ti = 5 years ,
12 = 20 years for annuity period. l l
Therefore, PV = ($1,000)(10.6)(0.713) = $7,560.
I B.3 NUREG/BR-0184 e
- n- -- -m .
Appendix B O
Table B.2 Present value of a future dollar (yearly Table B.3 Present value of annuity of a dollar, compounding) received at end of each year (yearly compounding)
Year 3% 7% Year 3% 7%
1 0.971 0.935 1 0.971 0.935 i 2 0.943 0.873 2 1.91 1.81 3 0.915 0.816 3 2.83 2.62 4 0.889 0.763 4 3.72 3.39 5 0.863 0.713 5 4.58 4.10 6 0.838 0.666 6 5.42 4.77 7 0.813 0.623 7 6.23 5.39 8 0.789 0.582 8 7.02 5.97 9 0.766 0.544 9 7.79 6.52 10 0.744 0.508 10 8.53 7.02 11 0.722 0.475 11 9.25 7.50 12 0.701 0.444 12 9.95 7.94 13 0.681 0.415 13 10.6 8.36 14 0.661 0.388 14 11.3 8.75 15 0.642 0.362 15 11.9 9.11 16 0.623 0.339 16 12.6 9.45 17 0.605 0.317 17 13.2 9.76 18 0.587 0.296 18 13.8 10.1 19 0.570 0.277 19 14.3 10.3 20 0.554 0.258 20 14.9 10.6 25 0.478 0.184 25 17.4 11.7 30 0.412 0.131 30 19.6 12.4 40 0.307 0.0668 40 23.1 13.3 50 0.228 0.0339 50 25.7 13.8 O
NUREG/BR-0184 B.4
I Appendix B )
l 7hbles B.2 and B.3 contain the appropriate discount factors to be multiplied together. Additional background on discrete ;
discounting can be found in EPRI (1986), DOE (1982), and Wright (1973).
B,2,3 Continuous Discounting l Discrete discounting, as discussed above, deals with costs and revenues that occur at discrete instances over a period of time. For most regulatory analyses, discrete discounting and the present value factors shown in 'lables B.2 and B.3 can be used. Technically, discrete discounting does not correctly account for consequences that occur constantly, but the difer-ence is viewed as minimal, and the additional eNort is generally not warranted in a standard regulatory analysis. !
i Continuous discounting should be used in regulatory analyses beyond the standard analysis when costs and revenues occur j continuously over a period of time, such as those which must be weighed by an accident frequency over the remaining life j of a facility. The accident frequency is a continuous variable, although the real cost of the accident consequences is ;
constant. j The formula for continuous discounting is derived from the discrete discounting formula as follows. Assume that in one l period (t), the time will be subdivided into n intervals. The formula for discrete discounting, with a real discount rate of r, is 1/(1 + r/n)". As we subdivide the time period into an infmite number of intervals in the limit, we would abandon dis- ;
crete intervals altogether and so set tic limit as i l
lim 1/(1 + r/n)" = exp(-r) {
n-+oo For t periods, instead of one period as above, the formula becomes exp(-n), where r and t are defmed over the same time period.
The monetired values for the reductions in public and occupational dose due to accidents, as well as the avoided onsite and ossite property damage costs, require continuous discounting. To calculate the present value for the public health (acci-dent) and ofsite property attributes, when the monetary value or cost Co can occur with a frequency f, Strip (1982) pro-vides the following formula:
tg f C,f exp(-rt)dt = C f[exp(-rt,)- exp(-st,)]/r t, l wherc _ ti = time of onset of accident risk tr = time of end of accident risk. j For public (accident) risk, the product C o f is replaced by Z,,a representing the monetary value of avoided risk before dis-counting ($/ facility-yr [see Section 5.7.1.3}). As an example, assume the monetary value of avoided public risk due to an accident is $1.0E+4/ facility-yr (C,f = $1.0E+4). The facility is operational (t, = 0) with a remaining lifetime of 25 years (t, = 25). For an annual discount rate of 7% (t = .07/yr) the present value of avoided risk (monetized) becomes B.5 NUREG/BR-0184
Appendix B O
PV = ($1.0E+4/yr)[exp {-(.07)(0)}
- exp {-( 07)(25)}]/(.07/yr)
= ($1.0E+4)(ll.8)
= $1.18E+5/ facility To determine the present value of a reduction in offsite property risk, the frequency (f in the general equation above) is replaced with the frequency reduction (Af). As an example, let the frequency reduction (Af) be 1.0E-5/facili:y-yr and the cost (Co) be $1,0E+9. The annual discount rate is 7 % (r = .07/yr), and the reduction in accident frequency takes place 5 years in the future (t, = 5) and will remain in place for 20 years (t, = 5 + 20 = 25). The present value of the avoided offsite property damage becomes PV = ($1.0E+9)(1.0E-5/yr)[exp(-(.07)(5)} - exp(-(.07)(25)}]/(.07/yr)
= ($1.0E+9)(1.0E-5)(7.58) = $7.58E+4/ facility Tb calculate present values for the occupational health (accident) and onsite property attributes, the continuous discounting formula must be modified. The modifications account for the fact that 1) some components of severe accident costs are not represented by constant annual charges as noted in Section B.2.2, and 2) the single-event present values must be reintegrated because the accident costs and risks would be spread over a period of time (e.g., over the rerammg plant life-time for replacement power costs and over the estimated 10 years for cleanup and decontamination following a severe accident, for onsite property damage). Sections 5.7.3.3 and 5.7.6.4 address these modifications and provide estimation guidelines for regulatory initiatives that affect accident frequencies in current and future yean.
B.3 Occupational Exposure Experience Two documents contain considerable information related to occupational exposure experience at nuclear power plants and some non reactor facilities. In the first (NUREG/CR-5035), Beal et al. (1987) state the following concerning generic dose rate data for use in regulatory analyses:
"...The NRC is generally concerned with the average exposures potentially experienced at all piants within a specific class (i.e., BWRs, PWRs, or PWRs manufactured by a particular vendor), rather than with the exposures at a specific plant. Therefore, it is desirable to have a generic dose-rate data base available to NRC analysts for making radiation exposure estimates."
The dose rates have been classified by Beal et al. (1987) according to the EEDB (United Engineers and Constructors, Inc. l 1988b) code-of-accounts for nuclear power plant systems and components. The analyst can estimate the radiation expo- i sure as the product of the estimated labor hours for work on a specific EEDB system / component and the dose rate for that system / component. Tables B.4 and B.5 list occupational dose rates for PWR and BWR systems and components, respectively, by EEDB classification.
Chapter 4 of NUREG/CR-5035 provides illustrative examples of the estimation of occupational radiation exposure for specific tasks at a power plant. Labor-hour estimates are obtained from the EEDB (United Engineers and Constructors, Inc.1986). Adjustments to account for differences in labor productivity are taken from Riordan (1986). If hardware is to ,
be removed, and/or a learning curve is to be involved, these effects are accounted for using information from Sciacca
{
et al. (1986). '
l 1
NUREG/BR4)184 B.6 0\,
l l
l
.. . . . - - - . --- , . . - . , . . - . - = . . . . ~ . - . . - _ . . . - - . . . ~ . . . .
Appendix B
'Ddde B.4 C-
^'
' dose rates by EEDB dhh for PWR systems and components (Beal et al.1987)
Average l EEDB Dose Rate
- Code 4fh Description (ar/br) l REACTOR EQU1PMENT 221.122 Reactor Vessel Closure & Attachmanta 650 f Reactor Vessel Studs, Fasteners, Seals, l 221.123 l & Gaskets 140 1 800 i 221.131-2 Reactor Vessel Upper and 1mer Internals '
1 221.211 Contml Rods -
221.212 Contiel Rod Driws 1400 221.213 Control Rod Driw Missile Shield -
221.214 CRDM Seismic Supports -
MAIN HEAT TRANSFER TRANSPORT SYSTEM l !
222.1111 Main Coolant Pumps & Driw 65 2 i 222.118 Main Coolant Pumps Instr. & Control .
222.119 Main Coolant Pumps Foundations / Skids 40 222.12 Reactor Coolant Piping System 270 222.1321 Steam Generators l
- at manway and inside steam generator 5100
-manway vicinity and general area 110-222.1431 Pressuriser 95 222.1432 Pressurimr Relief Thnk 32 222.148 Pressuriser instrumentation & Control 15 222.149 Pressuriser Foundation / Skids RESIDUAL HEAT REMOVAL SYSTEM 223.111 RHR Pumps & Drives 45 223.121 RHR Heat F barrs 35 223.15,16,17 RHR Piping System 65 223.18 RHR instrumentation & Control 45
! SAFETY INJECr10N SYSTEM 223.311 Safety injection System Pumps and, Drives 8 223.312 Boron injection Pumps and Drive 223.331 Accumulator Thnk 6 ,
223.332-3 Boron injectionihnks 70 1 223.334 Refueling Water Storage Thnk <1 223,35,36,37 Safety injection System Piping System 55 l
I i
B.7 NUREG/BR-0184 i
Appendix B O
'Ihble B.4 (Continued)
Average EEDB Dose Rate
- Code-of Account Description (mr/hr) 223.38 Safety Injection System Instr. & Control 5 CONTAINMENT SPRAY SYSTEM 223.411 Containment Spray Pumps & Motors 15 223.421 Contamment Spray Heat Exchanger -
223.431 Containment Spray Additive Thnk <1 223.45,46,47 Containment Spray Piping System 25 223.48 Containment Spray Instrument. & Control 120 COMBUSTIBLE GAS CONTROL SYSTEM 223.55,56,57 Combustible Gas Control System Piping 10 223.58 Combustible Gas Contml System Instr. & Control 10 223.591 Hydrogen Recombiner 10 LIQUID WASTE SYSTEM Primary Equipment Drain System 224.1111-33 Tanks, Pumps, & Motors 250 224.1141 Equipment Drain Filter 50 224.115.116,117 Equipment Drain Piping 35 Miscellaneous Drain Waste System 224.1211-32 Thnks, Pumps, & Moton 170 224.1241-3 Waste Filters, Demineralizers, & R/O Units 150 224.125,126,127 Misc. Waste Piping System 75 Detergent Waste System 224.1311 32 Thnks, Pumps, & Motors 2 224.1241 4 Waste Filten, Demineralizers, & R/O Units 3 224.135,136,137 Detergent Waste Piping System 2 Chemical Waste System 224.1411-31 Thnks, Pumps, & Motors 60 224.144 Purification & Filter Equipment -
225.145,146,147 Chemical Waste Piping System 13 NUREG/BR-0184 O
B.8
. . - - .- .- ~. . - _
- Appendix B
'Ihble B.4 (Continued)
Average EEDB Dose Rate
- l Code-of-Account Descripilon (mr/hr)
Steam Generator Blowdown System I 224.1511-3 Tanks, Pumps, & Heat Feh==ers 3 224.15141-4 Demineralizers and Filters 4 224.151,1516,1517 S.G.B.D. Piping System 8 224.1518 S.G.B.D. Instrument. & Control 2 Regen. Chemical Waste System 224.1611-32 'Ihnks, Pumps, & Motors -
224.1641-3 Demineralizers, Filters, & Evaporator 100
} 224.165,166,167 Regen. Waste Piping System --
224.171 Chemical Feed Package (tks., pumps, piping, etc) 2 224.18 Liquid Waste System Instr. & Control 2 RADIOACTIVE GAS WASTE PROCESSING SYSTEM p(- 224.2111-32 224.2141 Radioactive Gas Compressors, Drives, & Decay 'Ihnks Recombiner Packages 7
2 224.2142 Gas Waste Vent Filter 3 224.215,216,217 Radioactive Gas Waste Piping System 2 224.218 Radioactive Gas Waste Instr. & Control -
SOLID WASTE SYSTEM Dry Active Waste Volume Reduction 224.3111-32 'Ihnks, Pumps, & Motots 120 224.3141 Filters 2000 Volume Reduction and Solidification System i 224.325,326,327 Solid Waste System Piping 7 224.328 Solid Waste System Instrument. & Control 2 FUEL HANDLING AND S'IORAGE 225.111-4 New and Spent Fuel Cranes and Hoists 25 225.131-2 Transfer Systems ,
210 225.31-2 Reactor Service & Fuel Storage Pool Service Platform 13 225.41 New Fuel Storage Racks <1 225.42 Spent Fuel Storage Racks -
225.4311-45 Spent Fuel Pool Cleaning & Purification Equipment 85 V
B.9 NUREG/BR-0184
Appendix B O
'Ihble B.4 (Continued)
Average EEDB Dose Rate
- Code-of-Account Description (mr/hr) 225.435,436,436 Spent Fuel Pool Clean. & Purif. Piping System 15 225.438 Spent Fuel Pool Clean. & Purif. System Instrument & Control --
INERT GAS SYSTEM 226.11 H 2/N2Gas Supply System 20 REACIOR MAKEUP WATER SYSTEM 226.311 Reactor Makeup Water Pumps & Drives 4 226.331 Reactor Makeup Water Tank 120 226.35,36,37 Reactor Makeup Water Piping System 20 226.38 Reactor Makeup Water System Instr. '& Control 3 COOLANT TREATMENT & RECYCLE 226.4111-5 Chemical & Volume Control System Pumps, Motors, & Equipment 13 226.4121-8 CVCS Heat Transfer Equipment 80 226.4121-7 CVCS Thnks and Pmssure Vessels 140 l
226.4141-5 CVCS Purification and Filtration Equipment 1800 226.415,416,417 CVCS Piping System 95 226.418 CVCS Instr. & Contml 21 226.4191-2 Foundations & Skids for Boron System Equipment 22 226.4211-33 Boron Recycle System Pumps, Motors, Tanks,
& Equip. 100 226.4241-7 Boron Recycle System Purif, & Filter Equipment 38 226.425,426,427 Boron Recycle Piping System -
226.428 Boron Recycle Instrument. & Control 3 FLUID LEAK DETECTION SYSTEM 226.6 Fluid Leak Detection System -
B.10
J a+-
i
- Appendix B l Thble B.4 (Continued) j Average
'EEDB Dose Rate
- Code-of-Account Description (mr/br) l AUXILIARY COOL.ING SYSTEMS l Nuclear Service Water System j 226.7111-2 Safeguanis Cooling 'Ibwer Pumps, Equip, I
& Cooling Tower -
226.715,716,717 Cooling Tbwer Piping System 80 j 226.718 Cooling 'Ibwer instr. & Contrcl -
Primary Component Cooling Water 226.7211-31 Prim. Comp. Cooling Water Pumps, Motors
& Equip. Thnks 2 j 226.725,726,727 Prim. Comp. Cool. Water Piping System 25 l
226.728 Prim. Comp. Cool. Water Instr. & Contml -
- r CRDM = Connel Rod Drive Mechanian .
CVCS = Chemical and %lume Control Systm !
EEDB = Energy Economic Dam Base j mr = millisem i SGBD = Stam Generneor Blondown
- Average of across-plant " typical' values i
t B.11 NUREG/BR-0184
Appendix B O
'Ihble B.5 Occupational dose rates by EEDB classification for BWR systems and components (Beal et al.1987)
Average EEDB Dose Rate
- Code-of-Account Description (mr/hr)
REACIOR EQUIPMENT 221.122-133 Reactor Vessel Closure & Attachments, Studs, Fasteners, Seals, Gaskets, Core Support, and Shroud Assembly -
221.134 Jet Pump Assemblies 4400 221.135 Fluid Distribution Assemblies 210 221.136 Steam Dryer Assembly 800 221.211 Control Rods 170 221.212 Control Rod Drives 110 MAIN HEAT TRANSFER TRANSPORT SYSTEM 222.I111 Reactor Recirculation Pumps & Motors 90 222.15,16,17 Recirculation Piping System 240 222.18 Reactor Recirculation Instrument. & Control 200 RESIDUAL '* EAT REMOVAL SYSTEM 223.11 RHR Pumps & Drives 60 223.12 RHR Heat Exchangers 320 223.14 RHR Puri6 cation & Filtration Equipment -
223.15,16,17 RHR Piping System 100 223.18 RHR Instrumen:ation & Control 80 REACIDR CORE ISOLATION COOLING SYSTEM l 223.21-24 RCIC Pumps, Motors, & Equipment 90 l
223.25,26,27 RCIC Piping System 100 223.28 RCIC Instrumentation & Control -
HIGH PRESSURE CORE SPRAY SYSTEM 223.31-34 HPCS Pumps, Motors, & Strainers 30 223.35,36,37 HPCS Piping System 100 223.38 HPCS Instrumentation & Control 20 .
l O
NUREG/BR-0184 B.12
Appendix B
,x
'Ihble B.5 (Continued)
- Average
- EEDB Dose Rate
- l Code-of-Account Description (mr/hr) i LOW PRESSURE CORE SPRAY SYSTEM a ,
l 223.41-44 LPCS Pumps, Motors, & Strainers 15 4
223.45,46,47 LPCS Piping System 190 223.48 LPCS Instrumentation & Control --
I A
I
)
COMBUSTIBLE GAS CONTROL SYSTEM
)
223.55,56,57 Combustible Gas Control System Piping System 1 l
223.58 Combust. Gas Control System Instr. & Control ---
]
223.591 Hydrogen Recombiner 20 ;
3 STANDBY LIQUID CONTROL SYSTEM
- j. 223.61 Standby Liquid Control System Pump & Motor 1 j
s 223.631 SLCS Main Storage Tank 5 223.632 SLCS Test Tank --
I 223.65,66,67 SI.CS Piping System 55 223.68 SLCS Instrumentation & Control -
t STANDBY GAS TREATMENT SYSTEM 223.711 722 SGIS Fans, Motors, Heat Transfer & Equipment -
223.74 SG13 Purification & Filtration Equipment 1 223.75,76,77 SUIS Piping System -
223.78 SGTS Instrumentation & Control -
LIQUID WASTE SYSTEM High Purity System 224.111-113 High Purity Collection Tanks, Pumps, Motors, & Equipments 280 224.114 High Purity Waste Filter, Dem'.neralizers -
224,115,116,117 High Purity Waste Piping Sysem 10 B.13 NUREG/BR4184
Appendix B O
Thble B.5 (Continued)
Average EEDB Dose Rate
- Code-of-Account Description (mr/hr)
Low Purity System 224.121-123 Iew Purity Collection Tanks, Pumps, Motors, & Equipment 190 224.124 low Purity Waste Evaporaton Demineralizers and Filters --
224.125,126,127 low Purity Waste Piping System 60 Detergent Waste System 224.131-133 Detergent Waste Tanks, Pumps, Moton,
& Equipment 40 224.134 Detergent Waste Filter, Demineralizers, R/O Unit Package 65 224.135,136,137 Detergent Waste Piping System 2 Chemical Waste System 224.141 143 Chemical Waste Tanks, Pumps, Moton,
& Equipment 40 224.144 Chemical Waste Purification & Filter Equipment ---
224.145,146,147 Chemical Waste Piping System --
Cleanup Floor Drain Waste System 224.15 Cleanup Floor Drain Waste Pumps, Motors, & Eq. -
Chemical Waste Train 224.16 Regen. Waste Pumps, Motors, Equipment,
& Piping -
224.17 Misc. Radwaste Equipment --
224.18 Liquid Waste System Instrument & Control --
RADIOACTIVE GAS WASTE PROCESSING 224.211-214 Gas Waste Processing System Equipment ---
224.215,216,217 Radioactive Gas Waste Piping System 10 224.218 Radioactive Gas Waste Instmment & Control -- I i
l 1
I i
G; NUREG/BR-0184 B.14 ;
1 I
i Appendix B i
'Ihble B.5 (Continued) l Average ;
EEDB Dose Rate
- Codeef Account Description (ar/br) ,
SOLID WASTE SYSTEM 224.321 Dry Active Waste Wlume Reduction Centnfuge, Pumps, Motors, & Equipment ,
200 224.322-324 Solid Waste System Equipment, 'Ihnks, Puri8 cation
& Filtration -
224.325,326,327 Solid Waste System Piping System 250 224.328 Solid Waste System Instruments, & Control -
l FUEL HANDLING AND STORAGE 225.11 Fuel Handling Equipment, Cranes, & Hoists 20 225.12-14 Fuel Handling 'Ibols, Transfer Systems, & Machmes -
225.2-3 Remote Viewing Equipment, Refueling Platform, Fuel Handling Platform 4 i 225.41-42 Fuel Storage Equipment & Racks -
225.431-434 Spent Fuel Pool Cleamng & Puri6 cation Pumps Motors Equipment, Filters, & Demineralisers 400-225.435,436,437 Spent Fuel Pool Clean. & Purif. Piping Systems 40 225.438 Spent Fuel Pool Clean. & Purif. Piping System Instrument & Cont -
REACIOR WATER CLEANUP SYSTEM 226.41-42 RWCU System Pumps, Motors, & Heat Fehaarrs 120 226.43 RWCU 'lhnks & Pressure Vessels 2 226.44 RWCU Puri6 cation & Filter Equipment 80 226.45,46,47 RWCU Piping System 120 226.48 RWCU System Instrument & Control --
1 FLUID LEAK DETECTION SYSTEM j 226.6 Fluid leak Detection System --
AUXILIARY COOLING SYSTEMS 226.71 Essential Service Water System -
226.72 Closed Cooling Water System -
226.731-732 Plant Chilled Water System Pumps, Motors,
& Heat Transfer Equipment 80 B.15 NUREG/BR-0184
Appendix B O
'Ihble B.5 (Continued)
Average EEDB Dose Rate
- Code +f-Account Description (mr/hr) 226.734 Purification & Filtration Equipment -
226.735,736,737 Plant Chilled Water Piping System -
226.738 Plant Chilled Water Instrument & Control -
FEED HEATING SYSTEM 234.1 Feed Water Heaters 1 234.211 Feed Water Pumps 2 234.25 Feed Water Piping 70 234.26 Feed Water Valves 850 OTHER TURBINE PLANT EQUIPMENT 235.115 Main Vapor System Piping 50 235.116 Main Vapor System Valves 260 235.117 Main Vapor System Misc. Piping 2 235.118 Main Vapor System Instrument & Control 100 235.21 Main Steam / Reheat Vents & Drains 16 235.35 T.B. Closed Cooling Water System Piping 20 235.4 Demin. Water Makeup System 1 235.631 Neutralization System Tank 1 HPCS = High Pressure Core Spray 1.PCS = 1m Pressure Core Spray mr = rnillirem RCIC = Reactor Core Isolation Cooling RWCU = ReacerWaerCleanup SGI3 = Standby Gas'Deatment Sysem SLCS = Standby Liquid Control System TB = "ntrbine Building
- Average of across-plant ' typical
- values O
NUREG/BR-0184 B.16
i l; Appendix B LC l
l 'Ibe NRC mamtains occupsional exposum data in the Radiation Exposure Information and Reporting System (REIRS).
i 'Itc followinig six categories of ficensees have reported occupational exposure data:
i 1. power reactors (LWRs)
- 2. industrial radiographen
- 3. fuel processors, fabricators, and reprocesson
- 4. manufacturers and dhtributors of byproduct material
- 5. indet spent fuel storage installations
- 6. facilities for land disposal of low level waste.
' Annual reports br 1993 were received from 360 NRC limaaae of which 114 were operalors of power reactors. Raddatz and Hagemeyer (1995) have compiled and prec t. the 1993 and previous years' data in the second document related to occupational exposure experience of NRC-licensed facilities. No data from Ag-_- - r State licensees are included in the report.
Data limitations att discussed in Chapter 2 of Raddutz and Hagemeyer (1995), prior to the presentation of the processed results. Annual exposure data are given for the six facility classes listed above. Annual occupational exposme data for p 1991-1993 are tabulated in Mies B.6 to B.8 for industrial radiographers, manufacturers and distributors of typroduct f material, and fuel fabricators. For low level waste disposers and i%* spent fuel storers, the annual number of
\
workers with measurable doses and the collective and awrsge doses for 1991-1993 are shown in Mle B.9. For power reactors, the annual occupational exposure data from 1973 through 1993 are presented br BWRs, PWRs, and LWRs in Mies B.10 to B.12, respectively.
+
Chapter 4 of PaMan and Hagemeyer (1995) exammen occupational exposure data at LWRs in more detail. Included are annual whole body dose distributions; plant rankings by the collective dose per reactor; and the average, median, and extreme values of the collective dose per reactor. Mle B.13 lists the numbers of employees and collective and average doses for 1993 as a function of occupation and personnel type for LWRs B.4 Calculational Method for Table 5.3, " Expected Pbpulation Doses for Pbwer Reactor Release Categories" The information in this section is from the letter report, "MACCS Economic Consequence Tables for Regulatory !
Applications" (%ung 1995) prepared for the NRC. It provides an overview of the calculations and assumptions used in the preparation of Mle 5.3. Wung's results represent mean results conditional on the occurrence of each release category.
B.4.1 Introduction The MACCS Wrsion 1.5.11.1 was used to complete the calculations performed for the analysis reported in Wung (1995).
MACCS was designed to assess the potential off-site dose, health, and economic consequences of postulated nuclear power plant (NPP) accidents. Interdiction criteria specified by the user determine the dose levels at which long-term mitigative actions are implemented.
B.17 NUREG/BR-0184
Appendix B O
'Ihble B,6 1991-1993 annual occupational exposure information for industrial radiographers (Raddatz and Hagemeyer 1995)
Wrken Collective herage Number Number of with Dose Measurable of Monitored Measurable (person cSv or Dose (cSv or War Type of License Licenses Mrkers Doses person-rem) rems) 1993 Single location 39 673 183 23 0.13 Multiple locations 137 4,046 2,824 1,603 0.57 !
Total 176 4,721 3,007 1,627 0.54 l 1992 Single location 48 771 182 37 0.20 ,
Multiple locations 198 5,392 4,082 1,827 0.45 !
Total 246 6,703 4.265 1,864 0.44 j 1
1991 Single location 56 822 338 44 0.13 Multiple 'ntation 192 5,998 4,311 2,116 0.49 lbtal 248 6,820 4,649 2,160 0.46 l
'hth 41.7 1991-1993 annual occupational exposure information for byproduct manufacturers and distributors (Raddatz and Hagemeyer 1995)
M rkers Collective herage Number Number of with Dose (penon- Measurable of Monitored Measurable cSv or Dose (cSv or War Type of License Licenses Mrkers Doses peswn-rem) rem) 1993 M & D-Broad 8 2,455 925 512 0.55 M & D-Limited 50 2,458 1,329 168 0.13 Tbtal $8 4.913 2,254 680 0.30 1992 M & D-Broad il 3,632 1,674 718 0.43 M & D-Limited 56 1,578 576 72 0.13 Total 67 5.210 2,250 784 0.35 1991 M & D-Broad 12 3,732 1,443 674 0.47 M & D-Limited 46 1,198 513 47 0.09 Total 58 4,930 1,956 721 0.37 I 1
I O'
NUREG/BR-0184 B.18
l i Appendix B i >
i Able B.8 1991-1993 annual occupational exposure infonnation for !
fuel fabdcators Gladdetz and Haganeyer 1995)
I W'rkers e % l Number Nimmber of with Dese (person- 14asurable ;
of Meeltered Measurable rems or Dose (rusas l hr Type of License Licenses Workers Dessa person-cSv) or cSv) ,
1993 Uranium Fuel Fab 8 9,649 2,611 339 0.13 '
1992 Uranium Fuel Fab 11 8,439 5,061 545 0.11 )
1991 Uranium Fuel Fab 11 11,702 3,929 378 0.10 hble B.9 Annual occupational doses for low level waste disposal and spent fuel storage facilities,1991-1993
[Raddatz and Hagesneyer 1995] I J
Workers with . Average ,
measurable Collective dose measurable '
Licensee War doses (person-cSv) dose (cSv)
I.ow Level Waste Disposers 1991 147 39 0.27 l 1992 82 37 0.45 l
1993 76 21 0.27 Independent Spent Fuel Storers 1991 24 4 0.17 1992 85 11 0.13 1993 52 14 0.26 l
]
i B.19 NUREG/BR 0184
Z >
l E Table B.10 Summary of 1973-1993 annual occupational exposure information reported }
$ by commercial BWRs (Raddatz and IIagemeyer 1995) ii.
5r 9 CD W
b Average g Collective Average No. Average Annual Collective No. of Average Dose Per Personnel Collec- Average Average Iksuber Doses Workers Gross Dose Per Reactor Vtth ttve Dose Electrtetty Maxtsue of (person- With Electricity Worker (person- Measurable per W-yr Generated Dependable Reactors c$v er Measurable Generated (c5v or c5v or Doses (person-c$v Per Reactor Capactty Year included person-ree) Doses (W-yr) ree) person-ree) Per Reactor /W-yr) ( W yr) liet (We) 1973 12 4,564 5,340 3,393.9 0.85 380 445 1.34 283 438 1974 14 7,095 8,769 4,060.2 0.81 507 626 1.75 290 485 1975 18 12,611 14,607 5,786.4 0.86 701 812 2.18 321 595 1976 22 12,300 16,604 8,137.9 0.74 559 755 1.51 370 630 1977 23 19,041 21,388 9,102.5 0.89 828 930 2.09 396 637 1978 25 15,273 20,278 11,856.0 0.75 611 811 1.29 474 660 1979 25 18,325 25,245 11,671.0 0.73 733 1,010 1.57 467 660 1980 26 29,530 34,094 10,868.2 0.87 1,136 1,311 2.72 418 663 1981 26 25,471 34,755 10,899.2 0.73 980 1,337 2.34 419 663 tc 1982 26 24.437 32,235 10,614.6 0.76 940 1,240 2.30 408 663 S 1983 26 27,455 33,473 9.730.1 0.82 1,056 1,287 2.82 374 663 1984 27 27,097 41,105 10,019.2 0.66 1,004 1,522 2.70 371 754 1985 29 20,573 38,237 12,284.0 0.54 709 1,319 1.67 424 775 1986 30 19,349 37,928 12,102.1 0.51 645 1,264 1.60 403 786 1987 32 16,717 41.737 15,109.0 0.40 522 1,304 1.11 472 832 1988 34 17,983 40,305 16,665.4 0.45 529 1,185 1.08 490 845 1989 36 15,549 44,360 17,543.5 0.35 432 1,232 0.89 487 857 1990 37 15,780 41,577 21,336.1 0.38 426 1,124 0.74 577 862 1991 37 12,005 38,492 21,505.8 0.31 324 1,040 0.56 581 860 1992 37 13,309 42,095 20,592.2 0.32 360 1,138 0.65 557 859 t 1993 37 12,221 38,309 21,995.6 0.31 330 1.062 0.56 594 798
- Includes only those reactore that had been in comuneretal operetton for at lease one full year as of Decaster 31 of each of the indicated years. and all figures are uncorrected for multiple reporting of translent Individuals.
l O O O
Table B.11 Summary of 1973-1993 annual occupational exposure infonnation reported j by commercial PWRs (Raddatz and Hagemeyer 1995) i j Awrego e Annual Collective Average leo. Average !
Collective No. of Average Deee Per Pereennel Collac- Average Average thesher Baees Werkere Gross Dese Per teacter with ttve Deee Electricity Itseless, of (persen- With Electricity Worker (person- Itseeurable per Inf-yr Generated - Bayeseeable anectors cSv er floseurable Generated (c5v er c5v er Ossee (persen-c5v Per Reacter Cepecity Year incleded pereen-ree) Deses (Inf-yr) rem) person-tem) Per Anector /Inf-yr) (Inf-yr) liet (Inse) 1973 12 9,398 9,440 3,770.2 1.00 783 787 2.49 314 544 l
- 1974 19 6,555 9,370 6,530.7 0.70 345 493 1.00 344 591 -
1975 26 8,268' 10,884 11,982.5 0.76 318 419 0.69 461 647 ;
1976 30 13,807 17,588 13,325.0 0.79- 460 586 1.04 444 701 j 1977 34 13.467 20,878 17,345.8 0.65 396 614 0.78 510 688 i 1978 39 16,528 25,700 19,840.5 0.64 424 659 0.83 509 706 ;
1979 42 21,657 38,828 18,255.0 0.56 516 924 1.19 435 746 l 1980 42 24,265 46,237 18,289.3 0.52 578 1,101 1.33 435 746 1981 44 28,673 47,351 20,553.7 0.61 652 1,076 1.40 467 752 m 1982 48 27,753 52,146 22,140.6 0.53 578 1.086 1.25 461 777 b 1983 49 29,017 52,173 23,195.5 0.56 592 1,065 1.25. 473 785 '
~
1984 51 28,138 56,994 26,478.4 0.49 552 1,118 1.06 519 809 j 1985 53 22,469 54,633 29,470.7 0.41 424 1,031 0.76 556 820 1986 60 23,032 62,995 33,593.0 0.37 384 1,050 0.69 560 878 1987 64 23,684 62,597 37,007.3 0.38 370 978 0.64 578 900 ;
i 1988 68 22,786 62,921 42,929.7 0.36 335 925 0.53 631 885 '
1989 71 20,381 63,894 44,679.5 0.32 287 900 0.46 629 897 .
1990 73 20,812 67,081 46,955.6 0.31 285 919 0.44 643 907 i 1991 74 16,510 60,269 51,942.6 0.27 223 814 0.32 702 913 l 1992 73 15,985 61,048 53,419.8 0.26 219 836 0.30 732 923 .
1993 73 14,142 56,588 50,480.6 0.25 194 775 0.28 692 919 ;
i
- Includes only these reacters that had been in camerclei operetten for et least one full year es of Deceedier 31 of each of the indtested years and all figures are uncorrected for multiple reporting of tronelent Individuals.
{
z ,
p.
W r
- O
, = a 1 M e ,
T = i
- _ _ . - - _ _ _ - - . _ _ _ _ _ _ . _ _ . _ _ _ _ - _ _ _ . _ _ _ _ _ _ - - -_- _ _- _ . . _ _ _-___ _I
2: >
$ Table B.12 Summary of 1973-1993 annual occupational exposure information reported by commercial LWRs (Raddatz and IIagemeyer 1995) g E "
6
- Average
$ ennual Collective Average No. Average Copective llo. of Average Dose Per Persormel Collec- Average Average Percent of Number Don's Workers Gross Dose Per Reactor With tive Dese Electricity Mentase Maulass of (p<rson With Electricity Worker (person- Moseurable per Oti-yr Generated Dependable Dependeble Reactors c5, or Measurable Generated (c$v or c5v or Deses (persen-c5v Per Reactor Capacity Capacity Year Included person-rom) Doses (Iti-yr) rem) persor:-rom) Per Reactor /Inf-yr) (Int-yr) Net (Info) Achieved 1973 24 13,962 14,780 7,164.1 0.94 582 616 1.95 299 491 61%
1974 33 13,650 18,139 10,590.9 0.75 414 550 1.29 321 546 59%
1975 44 20,879 25,491 17,768.9 0.82 475 579 1.18 404 626 651 1976 52 26,107 34,192 21,462.9 0.76 502 658 1.22 413 671 62%
1977 57 32,508 42,266 26,448.3 0.77 570 742 1.23 464 667 70%
1978 64 31,801 45,978 31,696.5 0.69 497 718 1.00 495 688 72%
1979 67 39,982 64,073 29,926.0 0.62 597 956 1.34 447 714 63%
1980 68 53,795 80,331 29,157.5 0.67 791 1,181 1.84 429 714 60%
1981 70 54,144 82,106 31,452.9 0.66 773 1,173 1.72 449 719 63%
m 1982 74 52,190 84,381 32,755.2 0.62 705 1,140 1.59 443 737 60%
w 1983 75 56,472 85,646 32,925.6 0.66 753 1,142 1.72 439 743 59%
1984 78 55,235 98,099 36,497.6 0.56 708 1,258 1.51 468 790 59%
1985 82 43,042 92,870 41,754.7 0.46 525 1,133 1.03 509 804 63%
1986 90 42,381 100,923 45,695.1 0.42 471 1,121 0.93 508 847 605 1987 96 40,401 104,334 52,116.3 0.39 421 1,087 0.78 543 877 62%
1988 102 40,769 103,226 59,595.1 0.39 400 1,012 0.68 584 871 67%
1989 107 35,930 108,254 62,223.0 0.33 336 1,012 0.58 582 883 66%
1990 110 36,592 108,658 68,291.7 0.34 333 988 0.54 621 892 70% i 1991 111 28,515 98,761 73,448.4 0.29 257 890 0.39 662 895 74% '
1992 110 29,294 103,143 74,012.0 0.28 266 938 0.40 673 901 75%
1993 110 26,363 95,896 72,476.2 0.27 240 872 0.36 659 878 '75%
' Includes only tivee reactors Get had been in commercial operetten for et least one full year es of December 31 of each of the indicated years, and all ftpures are w=w-crected for muittple reporting of transtant Individuela.
P O O _
O -
f O m
' l Table B.13 1993 numbers of esuployees and Collective and average doses by occupation and personnel type at LWRs (Raduatz and Hagesneyer 1995)
TOTE PER WORK PUNCTION l UTILITT E N LOVEES CONTRACT tmAEERS tunt JNe STAfl0N EMPLOVEES Pets 0N-cSV E OF TO'AL-Jos ruscTacN PEnseN-es , E Or YOTR PEaseN ese E Or t0TE PEasou-cSV E OF TOTR 00lLING M T H KACTORS __ ,
1,209 9.9E 81 0.7E 459 3.35 1,749 14.3E '
REACTOR WP5 & sMRV 4,12T 50.21 neUTINE NEINTENANCE 2,140 17.5E 199 1.4E 3,788 31.1E 8.9E 35 0.35 T23 5.9E 865 T.11 IN-SERVIE INEPECTieN 107 659 5.45 175 1.45 1,453 11.9E 2,287 18.8K ,
SPECI R m INTENautE 291 2.4X MSTE PROCESStas 154 1.35 9 0.11 128 1.0E ,
2.0E TT 0.E 539 4.45 STF T.21 REfUELINE M1 596 4.R T,090 55.11 12,196 100.0E TOTE 4,510 37.0E PM5mRI25 MTB MACTWE 0.2K 470 3.3E 1.M9 8.65 REACTet EPS & sNRW MT 5.2K 31 5,872 35.11 1,590 11.05 ees 4.2X 2,SF3 19.9E meWTIM mRNTm 1,652 11.41 2,006 13.9E IN-5ERVIM INEPECTIeu 16T 1.2X 1st 1.3E 592 4.11 192 1.35 2,805 19.4X 3,589 24.8E SPECIE mlNTomaars 375 2.4X tmSTE PROCESSINS 141 1.1E 9 0.11 29T 1.4X
{ (
w REFW Llus det 3,962 26.FE 4.2E 254 1,202 1.8E 4.9E 1,305 9,312 9.0K 64.41 2,143 14,45F 15.0E 100.0K a TO:n e
ALL LI N T % M ER REACT 1RS T.3K 0.45 929 3.5E 2,997 11.21 l REACTcR ess & suRV 1,957 112 42.0E 3,730 14.0K SOT 3.05 6,661 25.0E 11,199 neu!!NE hFf4TENANCE 2,871 10.8E IN-sERVIC INSPECTIe5 27% 1.0K 222 0.85 2,375 s.9E 1,251 4.TE 36T 1.4X 4,254 16.0E 5,87T 22.0E SPECI R mINTENANCE det 2.5X i unsTE PeoRissINS 316 1.2E 18 0.15 335 1.3X 845 3.2K 351 1.3X 1,864 6.9E 3,060 11.4E REFUELING 14,402 61.55 26,653 100.01 I TOTE 8,373 31.41 1,875 T.0E i
'J: i C
70 tri O
'O ,
iB 1
- =
I h !
. CO
i I
l Appendix B l
O The following scenario assumptions and input data were selected by Young and NRC staff as a basis for the analysis:
- 1. 80th pementile weather data, as defined in 3 raft NUREG/CR-6295 (Davis et al.1995) were used as the meteorological input data.
- 2. The site data for the analysis were chosen to represent an 80th percentile NPP site in terms of the population density surToundm' g the site.
- 3. Calculations were performed for each of BWR and PWR source terms defined by Nourbakhsh (1992) as representative of severe LWR accident source terms.
- 4. NUREG-1150 (NRC 1991) emergency response assumptions were implemented as reported in NUREG/CR-4551 (Sprung et al.1990).
- 5. The values assigned to the MACCS food ingestion model input parameters PSCMILK, PSCMH, and GCMAXR are those values recommended by Mubayi as corrections to the values used in the NUREG-IISO analyses (Mubayi 1994).
The PSCMILK and PSCMH parameters define the levels of ground contammation above which crops are interdicted for accidents occurring during the growing season. GCMAXR defines the levels of ground contammation above which land is restricted from agricultural production.
- 6. Consequence values represent mean results and consequences within a 50-mile radius of the release.
B.4.2 MACCS Input Parameter Assumptions NUREG-1150 MACCS input parameter values as provided and discussed in Sprung et al. (1990) were applied in the calculations except for those parameters discussed below. In addition, the values recommended by Mubayi (1994) as cor-rections to the NUREG-1150 values for MACCS input parameters PSCMILK, PSCOITI, and GCMAXR were used.
I Meteorological Data Ontryear of meteorological data from Charleston, South Carolina was selected from Davis et al. (1995) to represent the conservative case (80th percentile) weather data. Wind roses were defined in the EARLY input file. The peak sector was assigned a 15 % frequency, the adjacent sectors a frequency of 11 %, and the remammg sectors were assigned a frequency l
of 4.85%. The wind rose sector containing the maximum population for the site was defined as the peak sector. The definition of the wind roses for the site is consistent with the method used to define the 80th percentile wind rose in Davis (1995).
Site Data Population and land use, data for the Peach Bottom NPP, as defined by the SECPOP90 software package, was imple-mented in this analysis (liumphreys 1995). The population data provided by SECPOP90 is based on 1990 data. Peach Bottom is at the 84th percentile in terms of U.S. NPP site population density within 30 miles and the 79th percentile in terms of population density within 20 miles (Young 1994). Peach bottom is located within the state of Ibnnsylvania.
NUREG/BR-OlB4 B.24 O'
Appendix B CN (V).
Source 'Ierm Calculations were performed for all of the source-term release categories defmed by Nourbakhsh (1992). The accident progression characteristics of these release estegories were extracted from Gregory (1995). The analyst is referred to these two references for a detailed discussion of the derivation and application of these source term release categories.
Protection Actions The duration of the emergency phase was defined as four rather than seven days as in the NUREG-1150 analysis.
. The dose criterion for bot spot and normal relocation during the emergency phase was defined as 0.01 Sv. The values 4
assigned to these variables in NUREG-1150 were 0.5 Sv and 0.25 Sv, respectively.
The remaming emergency response input parameter values implemented in Young's analysis are the same as those applied
- in the NUREG-1150 Nch Botpm analysis. Ninety nine and one-half percent of the population is assumed to evacuate within 10 miles of the NPP. Tse evacuating population is assumed to disappear at 20 miles from the NPP. The delay time between the notification of off-site emergency response officials to initiate protective actions (input parameter OALARM) and the beginning of evacuation is assumed to be 1.5 hrs. The population is assumed to evacuate at a speed of 4.8 meters per second. It is assumed that the 0.5 % of the populatica not evacuating was relocated based on 0.01 Sv dose criterion for relocation.
Discounting f~hi The MACCS code economic model is not designed to discount doses incurred in the years following the accident release.
t O Consequently, it was not possible to include discounting in the calculations performed for Young's analysis without completing major modifications to the MACCS code.
"Long-term" doses incurred over the period of time following the first year after the accident were tabulated to assess the portion of the total population dose which could be significantly impacted by the discounting of accident costs. The inte-gration period for the calculanon of the population 40se resulting from groundshine and resuspension during the long-term phase is 1E+6 years. The luel of contammation modeled in the long-term environment is dependent upon the half-life of the released radionuclides as d the weathering terms input by the user. The prpulation dose received from food ingestion is dependent upon the long- erm transfer factor for each nuclide and crop of concern. The consequences calculated in Young's analysis are based on 1990 census and statistical data applied for the calculation of population dose and per per-son. The data indicate that the population dose incurred over the long term comprises between 50% and 93% of the total population dose for 94% of the source-term categories.
l n
B.25 NUREG/BR-0184
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' Appendix C a
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l Supplemental Information for Non-Reactor
- Regulatory Analyses i
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- Appendix C Appendix C Supplemental Information for Non-Reactor Regulatory Analyses This appendix provides supplemental information for performing a regulatory analysis for non-reactor facilities, both fuel and non-fuel cycle. The procedure is essentially the same as that described in Chapters 2 through 5. However, the variety of facility types and the relatively non-integrated sets of available information lend dif5culty to performing a value-impact analysis in the more straightforward manner as that for power reactors.' This appendix represents a compilation of information to aid the preparation of a regulatory analysis applicable to non-reactor facilities. 'Ihe nature of regulatory analyses for non-reactor facilities will continue to evolve as more analyses are performed and more information becomes available.
As discussed in Section 4.3, the analyst should strive to use quantitative attributes when performing a regulatory analysis -
for non-reactor licensees. The Commission has determined, for example, that PRA should be used for analyses involving materials licensees when the potential safety consequences warrant its use, sufficient data are available, and the licensees can reasonably be expected to be capable of performing such analyses (NRC 1996c). However, it should be recognized l that there are many benefits of improved regulation of non-reactor facilities that do not lend themselves to quantification. 4
/ For example, increased confidence in the margin of safety may be a nonquantifiable benefit of a particular proposed I
( regulatory requirement. As noted in Section 4.5, nonquantifiable benefits and costs can be significant elements of a regulatory analysis and need to be considered by the analyst and decision maker as appropriate.
l The approach taken in this appendix has been to first review the relevant literature in sufficient detail to permit the regulatory analyst to judge the value of each report (see Sections C.3-C.11). Tables and figures containing potentially useful data have been extracted from the reports and included in this appendix. Reviews of non-reactor regulatory anat ses that have been performed comprise Sections C.8-C.!!.
I Based on the review of the literature, guidance on the performance of the value-impact analysis portion of a regulatory analysis has been developed. It is presented at the front of this appendix in the form of composite listings developed from the tables and figures to focus the relevant data for the analyst (see Sections C.1 and C.2). These should be used to direct the analyst's search for information that may be needed in the value-impact analysis. In some cases, the analyst may fmd values differing by several orders of magnitude, presumably the result of varying assumptions between the source documents. The analyst may wish to consult the references before selecting which value to use, especially since these tables are intended to direct analysts to appropriate sources, rather than to be used prima facie.
To assist the analyst, the tables and figures from which the data have been extracted to form these composites are referenced with the data. These composites are not intended to replace the original tables and figures, or the reports from which these tables and figures have been extracted. The analyst needing more detail should refer to the tables and figures, j 4
or the actual reports, directly. The analyst should also be aware that the composite listings combine data from multiple tables and Sgures, most of which were developed with differing sets of assumptions. Thus, the analyst may wish to use a specific table or figure, rather than a composite listing, when performing the analysis.
Two relatively recent data sources not cited in the Appendix C tables are also potentially available to the analyst. The first n data source is the Nuclear Material Events Database (NMED) administered by the NRC Office for Analysis and
(% )\
C.1 NUREG/BR-0184
Appendix C O
Evaluation of Operational Data (AEOD). He NMED contains information from materials, fuel cycle, and nonpower reactor licensees on events such as personnel radiation overexposures, medical misadmmistrations, losses of radioactive material, and potential criticality events. These data sources can be used to supplement and, when appropriate, supersede the information in the Appendix C tables. He second is the Bulletin 91-01 Event 'IYacking System administered by NMSS. NRC's Bulletin 91-01 requested reports from fuel cycle licensees relating to 1) loss or substantia' degradation of a j criticality safety control, and 2) conditions with a possible criticality hazard which have not been analyzed, j
Re analyst should also be aware of Attachment 3 to the CRGR Charter which provides guidance on the application of the
" substantial increase" standard at 10 CFR 50.109(a)(3). Footnote 13 in Revision 6 of the CRGR Charter states that i although 10 CFR 50.109 does not directly apply to facilities not licensed under Part 50, "much of the guidance in Attach-ment 3 is applicable and should be considered by the staffin evaluating qualitative factors that may contribute to the justifi-cation of proposed backfitting actions directed to nuclear materials facilities / activities." j l
C.1 Facility Classes l Review of the literature discussed in Sections C3-C11 suggested that non-reactor facilities would most appropriately be divided into two groups: fuel-cycle facilities and non-fuel cycle facilities. This grouping is defined in this section and i employed throughout the presentation on attribute quantification in Section C2. I C.I.1 Fuel Cycle Facilities A division of fuel cycle facilities was made by Pelto et al. In the unpublished PNNL study from 1983 reviewed in Section C.6. He facilities were classified into the following 13 groups: O,l.
l
- 1. mining 8. spent fuel storage
- 2. milling 9. HLW (high level waste) storage
- 3. conversion 10. TRU (transuranic) waste storage
- 4. enrichment 11. geologic waste disposal
- 5. fuel fabrication 12. shallow land waste disposal
- 6. MOX (mixed oxide) fuel refabrication 13. transportation.
- 7. fuel reprocessing Table CS.1, extracted from Schneider et al. (1982), pmvides a summary description of each of these 13 groups. It is accompanied by Figure Cl, also extracted from Schneider et al. (1982), which shows the uranium process flow and rela-tionship among the 13 groups.
Potential accidents during uranium mining do not yield much higher releases than incurred during normal operation.
Philbin et al. (1990) (see Section C4), Pelto et al. (see Section C6), McGuire (1988) (see Section C.8), and the EPA (1983) (see Section C9) addressed uranium mills. The followmg tables present data related to uranium milling: C4, O
NUREG/BR-0184 C.2
Appendix C g
w C48, C70, C77, and C87-C92. Figure C.4 also provides informatior, on uranium milling. UF conversion was exanuned by Philbin et al. (1990), Pelto et al., and McGuire (1988). Tables C.3, C49, and C70 present data related to
)
UF. conversion. i l
Enrichment facilities have been addressed by Pelto et al. and McGuire (1988). Ables C.50 and C70 provide data. Fuel fabrication has been exanuned by Philbin et al. (1990), Pelta et al., Mishima et al. (1983) (see Section C.7), McGuire (1988), and Ayer et al. (1988)(see Section C11). Relevant data are presented in the following tables: C6, C.51, C70-C76, C78-C.79, and C103-C104. Pelto et al. and Ayer et al. (1988) have addressed MOX fuel refabrication. Seven tables, C52-C55, C70, and C103-C.104, contain MOX information. Fuel reprocessing was exammed by Pelto et al., l McGuire (1988), and Ayer et al. (1988). Tables C56-C60, C70, C80, C.103, and C.105 provide relevant data. Spent )
fuel sto. age was examined by Daling et al. (1990) (see Section CS, Pelto et al., McGuire (1988), Jo et al. (1989) (see )
Section C10), and Ayer et al. (1988). Data are provided in the following tables: C26-C32, C44-C45, C61, C70, C81, C93-C.103, and C107.
)
j l
Philbin et al. (1990), Pelto et al., and Ayer et al. (1988) addressed HLW storage. The following tables contain relevant information: C62, C70, C103, and C106. No literature on TRU storage was reviewed. Daling et al. (1990) and l Pelto et al. exammed geologic waste disposal. Data are presented in the following tables: C9-C25, C.42-C.45, C.63, and C70. Figure C3 also provides data for geologic waste disposal. No literature on shallow land waste disposal was reviewed. Daling et al. (1990) and Pelto et al. addressed transportation. Ables C33-C45 and C64-C70 contain rele-vant information.
C.1.2 Non-Fuel Cycle Facilities
("j A division of non-fuel cycle facilities is in NUREG/CR-4825 (Ostmeyer and Skinner 1987) (see Section C.3). The facilities were classified into the following four groups based on the application /use of the licensed nuclear material:
- research, teaching, experimental, diagnostic, and therapeutic facilities, including hospitals, universities, medical groups, and physicians i
- measurement, calibration, and irradiation facilities, including users of sealed sources l
- manufacturing and distribution facilities employing byproduct and source materials, such as radiopharmaceuticals a service organizations, including waste repackagers, processors, and disposers.
Ostmeyer and Skinner (1987) (see Section C3) exammed all four groups. Relevant data are provided in Tables C.1-C3 and Figure C.2. Philbin et al. (1990) addressed large manufacturers / distributors of nuclear byproducts (Group 3) and waste warehouses (Group 4). Tables C.7 and C8 present information. McGuire (1988) examined Groups 1,3, and 4.
Relevant data are provided in Tables C82-C84 (Group 1), C85 (Group 3), and C86 (Group 4).
C.2 Quantification of Attributes The procedure to quantify the attributes appropriate to the value-impact analysis portion of a regulatory analysis for non-reactor facilities is discussed in Section 5.7. Based on the information from the literature survey (see Sections C3-C.11),
specific quantitative data are presented in this section for use with the following six attributes when performing the value-impact analysis portion of a non-reactor regulatory analysis:
/ i
.)
C.3 NUREG/BR-0184
Appendix C O
- 1. public health (accident)
- 2. public health (routine)
- 3. occupational health (accident)
- 4. occupational health (routine)
- 5. offsite property
- 6. onsite property.
Note that the last two attributes are discussed together rather than separately due to the nature of the anilable information.
C.2.1 Public Health (Accident)
The quantification of public health (accident) involws both frequencies and population doses associated with accident scenarios. Because non-reactor facilities tend to be much simpler in system configuration than power reactors, the number of potential accidents is much smaller, simplifying the scope of the accident analysis. However, accident frequency and population dose data are typically less available than for power reactors. This section extracts relevant frequency and dose data from Sections C.3-C.10. Also included are estimates of the total risk from accidents, as available.
C.2.1.1 Accident Frequencies The literature review yielded accident frequencies for both fuel and non-fuel cycle, non-reactor facilities. Composite listings have been assembled in this section.
Fuel Cycle Facilities Accident frequencies have been estimated for ten of the 13 non-reactor fuel cycle facilities listed in Section C.I. Only mining, TRU waste storage, and shallow land waste disposal haw been excluded (see Section C1.1).
For URANIUM MILLING, estimated frequencies for eight accident scenario; are in Table C.4, both as best estimates and 80% confidence bounds. Three of these scenario frequencies are also estimated in Table C.48, as follows:
- 1. solvent extraction fire = 4E-4 to 0.003/ facility-yr
- 2. retention pond failure with slurry release = 0.04/ facility-yr
- 3. slurry release from distribution pipe = 0.01/ facility-yr.
Except for the second, these estimates lie at least partially within the uncertainty ranges listed in Table C.4. For the retention pond failure with slurry release, the estimate of 0.04/ facility-yr slightly exceeds the upper bound in Table C.4.
1 For UF, CONVERSION, estimated frequencies for nine accident scenarios are in Table C.S. both as best estimates and 80% confidence bounds. Six of these scenario frequencies are also estimated in Thble C.49, as follows:
NUREG/BR-0184 C.4 l
Appendix C m.
- 1. uranyl nitrate evaporator explosion = IE-4 to 0.001/ facility-yr
- 2. hydrogen explosion during reduction = 0.001 to 0.05/ facility-yr
- 3. solvent extraction fire = 4E-4/ facility-yr
- 4. release from UF cylinder = 0.03/ facility-yr
- 5. distillation valve rupture = 0.05/ facility-yr
- 6. waste pond release = 0.02/ facility-yr.
Except for the last, these estimates lie within the uncertairry ranges listed in Table C5. For the waste pond release, the estimate of 0.02/ facility-yr is slightly below the lower bound in Table C5.
For ENRICHMENT, estimated frequencies for four accident scenarios are in Table C50. For FUEL FABRICATION, estimated frequencies for ten accident scenarios are in Tables C6 and C.51. Table C.6 lists them as both best estimates and 80% confidence bounds. The estimates for the ten scenarios are as follows [ parentheses () denote confidence bounds from Able C6):
- 1. minor facility release = 0.21/ facility-yr (0.15 to 0.32) from Table C6
- 2. large spills due to accidents or natural phenomena = 0.024/ facility-yr (0.015 to 0.044) from Table C6
'}
- 3. transportation accident = 0.0028/ facility-yr (0.0026 to 0.0030) from Table C.6
- 4. hydrogen explosion in reduction farnace = 0.01/ facility-yr (0.002 to 0.05) from hble C6 and 0.002 to 0.05/ facility-yr from Table C51
- 5. major fire = 2.lE-4/ facility-yr (1.2E-4 to 5.lE-4) from Table C6 and 2E-4/ facility-yr from Able C51
- 6. criticality = 0.0033/ facility-yr (5.0E-4 to 0.011) from Table C6 and SE-4/ facility-yr from Table C.51
- 7. release from hot UF cylinder = 0.021/ facility-yr (0.011 to 0.081) from Table C6 and 0.03/ facility-yr from Table C.51
- 8. fire in a roughing filter = 0.01/ facility-yr from Able C51
- 9. failure of valves and piping = 0.004/ facility-yr from Table C51
- 10. was: Ttention pond failure = 0.002 to 0.02/ facility-yr from Table C.51.
For MOX FUEL REFABRICATION, estimated frequencies for 14 accident scenarios are in Tables C53-C55. The estimates for these scenarios are listed below. Note that the values listed from Table C53 are those associated with normal high ef5ciency particulate air (HEPA) filtration. The corresponding estimates with HEPA filter failure are 1,000 times lower:
A (v)
C.5 NUREG/BR-0184 l
l
Appendix C O
- 1. > design basis earthquake = SE-6/ facility-yr(Table C54)
- 2. aircraft crash = 3E-7/ facility-yr (C54) and 1.5E-9/ facility-yr (C55)
- 3. hydrogen explosion in ROR (reduction-oxidation reactor) = 0.002 to 0.05/ facility-yr (C53),0.001/ facility-yr (C54),
and 0.005/ facility-yr (C55)
- 4. hydrogen explosion in sintering furnace = 0.001/ facility-yr (C54) and 0.005/ facility-yr (C55)
- 5. hydrogt.n explosion in wet scrap = 0.01/ facility-yr (C.53),0.005/ facility-yr (C54), and 3E-4/ facility-yr (C55)
- 6. ion-exchange resin fire = 1E-4 to 0.1/ facility-yr (C.53) and SE-4/ facility-yr (C54)
- 7. loaded fmal filter failure = 2E-4/ facility-yr (C54) !
- 8. criticality = 3E-5 to 0.008/ facility-yr (C53),6E-5/faciUty-yr (C54), and 6E-5/ facility-yr (C55)
- 9. powder shipping container spill = 3E-5/ facility-yr (C55)
- 10. exothermic reactions in powder storage = 1.5E-6/ facility-yr (C.55)
- 11. major facility fire = 2E-4/ facility-yr (C.53)
- 12. fire in waste compaction glove box = 0.01/ facility-yr (C53)
- 13. glove failure = 1/ facility-yr (C53)
- 14. severe gicve box damage = 0.01/ facility-yr (C53).
For FUEL REPROCESSING, estimated frequencies for 20 accident scenarios are in Ables C.57-C60. The estimates for these scenarios are listed below. Note that values from Table C57 are those associated with normal HEPA filtration.
The corresponding estimates with HEPA filter failure are generally 1,000 times lower, except where noted. Also note that values from Able C59 assume HEPA filter failure, except where noted.
- 1. loss of fuel storage pool water = 3E-6/ facility-yr (hble C58) l
- 2. lon-exchange resin fire and explosion = IE-4 to 0.1/ facility-yr (C.57, with frequencies 1E+5 times lower with '
HEPA filter failure) and SE-4/ facility-yr (C.58)
- 3. criticality = 3E-5 to 0.008/ facility-yr (C57), 6E-5/ facility-yr (C.58), and 2E-5/ facility yr (C59, without HEPA filter consideradon)
- 4. hydrogen explosion in high aqueous feed (HAF) tank = IE-5/ facility-yr (C57, with frequency 100 times lower with HEPA failure), 7E-5/ facility-yr (C58), 3E-6/ facility-yr (C59), and 1E-5/ facility-yr (C60)
- 9. fire in low level waste = 0.01/ facility-yr (C58)
C.6
i l
Appendix C
! n
!(V) l
- 6. fuel assembly drop = 0.01 to 0.1/ facility yr (C57),0.002/ facility-yr (C.58),0.0012/ facility-yr (C59, without HEPA consideration), and 0.01/ facility-yr (C.60)
- 7. explosion in HLW calciner = IE-6/ facility-yr (C57), SE-10/ facility-yr (C58, assuming HEPA filter failure), l 2E-7/ facility yr (C59), and IE-6/ facility-yr (C60) ,
- 8. krypton cylinder rupture = IE-4/ facility-yr (C58) and 1.3E-4/ facility-yr (C59, without HEPA consideration)
- 9. explosion in high activity waste (HAW) concentrator = IE-5/ facility-yr (C.57),4E-8/ facility-yr (C59), and IE-5/ ;
facility-yr (C.60)
- 10. solvent fire in codecontamination cycle = 1E-6 to 1E-4/ facility-yr (C.57) and 1E-6/ facility-yr (C60)
- 11. explosion in low activity waste (LAW) concentrator = IE-4/ facility-yr (C57) and IE-4/ facility-yr (C.60) l
- 13. solvent fire in plutonium extraction cycle = IE-6 to IE-4/ facility-yr (C57, with frequencies IE+5 times lower with HEPA failure)
- 14. dissolver seal failure = IE-5/ facility-yr (C57) r 1
( 3) 15. release from hot UF. cylinder = 0.05/ facility-yr (C.57, without HEPA consideratil V 16. solvent fire in hydrogen concentrator = 2E-6/ facility-yr (C.59) ;
- 17. red oil explosion in fuel product concentrator = 4E-8/ facility-yr (C.59)
- 18. explosion in fuel product denitrator = 4E-9/ facility-yr (C59)
For SPENT FUEL S'IORAGE, estimated frequencies for 17 accident scenarios are in hbles C.31, C32, C.61, C93, C.97, and C99. Data from Tables C31, C32, and C61 have been combined into 14 accident scenarios whose frequen- l cies are listed below. Note that the values taken from hble C31 correspond to the drywell storage concept only.
Ables C.93 and C.97 present frequencies for two additional scenarios-spent fuel pool fires due to seismic and cask drop initiators. Table C99 addresses one more scenario, deriving failure frequencies for four different configurations of a spent fuel pool cooling and makeup system:
- 1. collision during highway transport = 2E-4/ facility-yr (Table C.32, without fire, cask storage concept),
2E-5/ facility-yr (C32, without fire, drywell storage concept), 2E-6/ facility-yr (C.32, with fire, cask storage), and 2E-7/ facility-yr (C.32, with fire, drywell storage)
A \
lJ C.7 NUREG/BR-0184
Appendix C O
- 3. fuel assembly drop = 0.1/ facility-yr (C32),9E-4/ facility-yr (C61, for PWRs), and 0.006/ facility-yr (C61, for BWRs)
- 4. transpon cask drop = 0.004/fxility-yr (C32, cask storage),0.07/ facility-yr (C32, drywell storage), IE-4/ facility-yr (C61, PWRs), and 2.5E-4/ facility-yr (C61, BWRs)
- 5. cask venting during transport = 0.002/ facility-yr (C32, cask storage) and 0.03/ facility-yr (C.32, drywell storage)
- 6. canister drop during emplacement = 1.7E-8/ facility-yr (C31) and 1E-6/ facility-yr (C.32, drywell storage)
- 7. canister shear during emplacement = 2E-6/ facility-yr (C.32, drywell storage)
- 8. cask drop during emplacement = IE-5/ facility-yr (C32, cask storage)
- 9. airplane crash = 4.0E-10/ facility-yr (C31, without fire),7.4E-9/ facility-yr (C31, with fire), 6E-9/ facility-yr (C32, with fire, cask toppled, cask storage), 9E-9/ facility-yr (C32, with fire, cask storage), 2E-7/ facility-yr (C32, one fuel assembly, with fire, drywell storage), and 2E-8/ facility-yr (C32,10 assemblies, with fire, drywell storage)
- 10. canhquake = 4.8E-9/ facility-yr (C31, without fuel pin failure),4.3E-8/ facility-yr (C31, with pin failure),
4E-6/ facility-yr (C32, 24 assemblies, cask storage), 4E-8/ facility-yr (C32, 2,400 assemblies, cask storage),
8E-6/ facility-yr (C32, one assembly, drywell storage), 8E-7/ facility-yr (C32,10 assemblies, drywell storage), and 2E-8/ facility-yr (C32, 2,400 assemblies, drywell storage)
- 11. transporter collision during emplacement = 1.7E-8/ facility-yr (C31, without fire) and 6.lE-7/ facility-yr (C31, with fire)
- 12. transporter collision during retrieval = 0.0089/ facility-yr (C31, without pin failure or fire),0.028/ facility-yr (C31, with pin failure, without fire),1.4E-4/ facility-yr (C31, without pin failure, with fire), and 1.4E-4/ facility-yr (C31, with pin failure and fire)
- 13. transponer motion with canister panially i place = 0.086/ facility-yr (C31, during emplacement),0.0089/ facility-yr (C31, during retrieval, without pin failure), and 0.14/ facility-yr (C31, during retrieval, with pin failure)
- 14. canister drop during retrieval = 0.11/ facility-yr (C31).
For III3V STORAGE, estimated frequencies for three accident scenarios are in Table C62 (after gmuping by pairs).
For GEOIDGIC WASTE DISPOSAL, estimated frequencies for 18 accident scenarios are in Tables C14, C19, and C20. Note that Table C20 divides earthquake-induced accidents into nine categories, which are listed below as 18a-18i.
The estimates for the 18 scenarios are as follows:
- 1. fuel truck crash into HLW area = 2.0E-6/ facility-yr (Table C14)
- 2. fuel tnick crash into cladding waste area = 2.0E-6/ facility-yr (Cl4)
- 3. fuel truck crash into non-IILW (NHLW) area = 2.0E-6/ facility-yr (C14)
- 4. airplane crash = 1.0E-7/ facility-yr (Cl4) and <2.0E-10/ facility-yr (C19)
O NUREG/BR-0184 C.8
Appendix C g
(' J' ,
l
- 5. elevator drop = 4.0E-8/ facility-yr (C14)
- 6. fuel assembly drop = 0.1/ facility-yr (C19) and 1.E-8/ facility-yr (C20, dmp into hot cell with HVAC failure)
J 7 NHLW pallet drop = 0.050/ facility-yr (C14)
- 8. fmal filter failure = 0.003/ facility-yr (C14)
- 9. shipping cask drop = SE-6/ facility-yr (C20, with cask breach)
- 10. open consolidated fuel container drop = IE-9/ facility-yr (C20, with HVAC failure)
- 11. container drop in storage vault = 3E-8/ facility-yr (C.20, with failure to activate filtration system)
- 12. nuclear test = <0.001/ facility-yr (C.19)
- 13. loadinFdock fire = < l.0E-7/ facility-yr (C19, spent fuel) and < l.0E-7/ facility-yr (C.19, HLW)
- 14. waste handling ramp fire = < 1.0E-7/ facility-yr (C19)
- 15. emplacement drift fire = < 1.0E-7/ facility-yr (C19)
- 16. flood = 0.01/ facility-yr (C19)
- 17. tornado = <9.1E-11/ facility-yr (C19) 18 earthquake = <0.0013/ facility-yr (C.19) 18a. crane fails, falling on or dropping cask in receiving area = SE-8/ facility-yr (C20) 18b. train falls on cask = SE-8/ facility-yr (C20) 18c. structural object falls on fuel in cask unloading cell = SE-7/ facility-yr (C20) 184. crane fails, falling on or dropping fuel in cask unloading cell = IE-6/ facility-yr (C20) 18e. structural object falls on fuel in consolidation cell = SE-7/ facility-yr (C20) 18f. crane fails, falling on or dropping fuel in consolidation cell = IE-6/ facility-yr (C20) 18g. structural object falls on fuel in packaging cell = SE-7/ facility-yr (C20) 18h. crane fails, falling on or dropping fuel in packaging cell = IE-6/ facility-yr (C20, with HVAC failure) 18i. structural object falls on fuel in transfer tunnel = SE-7/ facility-yr (C20).
For TRANSPORTATION, it is convenient to identify three categories based on the material being shipped: spent fuel, plutonium oxide, and HLW. For spent fuel transportation, estimated frequencies for eight accident scenarios are in Tables C65-C69. The estimates for the scenarios are as follows:
- 1. leakage of coolant from spent fuel cask during rail shipment = 3E-4/ shipment (Table C65),6.4E-6/ shipment (C69, impact fails cask seal, fuel failure),1.2E-6/ shipment (C69, side impact fails pressure relief valve, fuel failure),
6.4E-6/ shipment (C69, end impact fails pressure relief valve, fuel failure), and 1.2E-6/ shipment (C69, side impact fails cask seal, fuel failure)
- 2. release from a collision during rail shipment = 2E-8 to 9E-6/ shipment (C65),9E-6/ shipment (C.67), and IE-4/yr 3 (C68, with closure errors)
) l v i C.9 NUREG/BR-0184 l l
Appendix C O
- 3. release from a collision followed by release of fuel from the cask during rail shipment = 2E-10 to 9E-8/ shipment (C65), 2E-5/yr C68, for 50-80 km/hr collision), 3E-4/yr (C68,80-100 km/hr), 8E-5/yr (C.68, with 1000*C fire for
> 1 hr), and 2E-5/yr (C.68, 800*C for >2 hr)
- 4. loss of gases from inner cavity = 9E-6/ shipment (C66, rail shipment) and 2E-5/ shipment (C66, truck)
- 5. loss of confinement and 50% fuel damage = 4E-7/ shipment (C66, without fire, rail),2E-9/ shipment (C66, with fire, rail), 2E-7/ shipment (C66, without fire, truck), 2E-9/ shipment (C66, with fire, truck), 4E-7/ shipment (C67, without fire, rail), and 3E-9/ shipment (C.67, with fire, rail)
- 6. loss of neutron shielding during rail shipment = 2E-5/ shipment (C67)
- 3. fire during rail shipment = IE-4/yr (C68,1000*C for > 1 hr) and 2E-5/yr (C68,800'C for >2 hr).
For plutonium oxide transportation, estimated frequencies for six accident scenarios are in Tables C.65 (three scenarios for rail shipment) and C66 (three scenarios for truck shipment). For HLW transportation by rail, estimated frequencies for five accident scenarios are in Tables C66 and C.67.
Non-Fuel Cycle Facilities For RESEARCH, TEACHING, EXPERIMENTAL, DIAGNOSTIC, AND THERAPEUTIC FACILITIES, Table C.1 contains an estimated overall accident frequency of 2.3E-4/ facility-yr. For MEASUREMENT, CALIBRA-TION, AND IRRADIATION FACILITIES, Thble C.1 contains an estimated overall accident frequency of 1.8E-4/ facility-yr. For MANUFACTURING AND DISTRIBUTION FACILITIES EMPLOYING BYPRODUCr AND SOURCE MATERIALS, estimated frequencies for eight accident scenarios are in 7hble C7, both as best estimates and 80% confidence bounds. Table C1 also contains an overall estimate of 0.0026/ facility-yr, which is noticeably less than the sum of the eight accident frequencies from Table C7. For SERVICE ORGANIZATIONS (waste warehouses),
estimated frequencies for six accident scenarios are in Table C8, both as best estimates and 80% confidence bounds.
McGuire (1988) estimated the frequency of a major radioactive release for a non-reactor facility to be IE-4/yr, assumed applicable to either fuel- or non-fuel cycle facilities (see Section C8).
C2.1.2 Population Doses from Accidents Unlike accident frequencies, literature review yielded population doses from accidents only for non-reactor fuel cycle facilities. However, safety analysis reports conducted for various DOE non-fuel cycle facilities (e.g., those at the Savannah River Site) contain population doses from accidents. If available, the analyst could use these for particular facilities.
Fuel Cycle Facilities Estimated population doses from accidents for 10 of the 13 non-reactor fuel cycle facilities listed in Section C.1 are included in this section. Estimates for mining, TRU waste storage, and shallow land waste disposal are not included. For NUREG/BR-0184 C.10 O
Appendix C URANIUM MILLING, estimated population doses from three accident scenarios are in Table C48. For UF. CONVER- ;
SION, estimated population doses from six accident scenarios are in Thble C49. For ENRICHMENT, estimated popu- l lation doses from four accident scenarios are in Table C.50. For FUEL FABRICATION, estimated population doses from seven accident scenarios are in Table C.51.
For MOX FUEL REFABRICATION, estimated population doses from 14 accident scenarios are in 'Ihbles C53-C.55. )
The estimates for these scenarios are listed below. Note that the values listed from Table C53 are those associated with i normal HEPA filtration. The corresponding estimates with HEPA filter failure are generally 1E+5 times higher, except where noted.
- 1. > design basis earthquake = IE+5 person-rem (Table C54)
- 2. aircraft crash = 3E+4 person-rem (C54) and 500 person-rem (C.55)
- 3. hydrogen explosion in Reduction-Oxidation Reactor (ROR) = 0.031 person-rem (C53), SE-9 person-rem (C54), and 1.lE-11 person-rem (C55)
- 5. hydrogen explosion in wet scrap = 0.16 person-rera (C53),2E-6 person-rem (C.54), and 1.1E-11 person-rem (C.55)
- 6. lon-exchange resin fire = 0.0092 person-rem (C53) and 2E-9 person-rem (C.54)
[ \
() 7. loaded final filter failure = 0.3 person-rem (C54)
- 8. criticality = 0.38 person-rem (C53, with dose 1100 times higher with HEPA filter failure),5 person-rem (C54), and 2 person-rem (C55)
- 9. powder shipping container spill = 1.1E-11 person-rem (C55)
- 10. exothermic reactions in powder storage = IE-10 person-rem (C55)
- 11. major facility fire = 1.6 person-rem (C53, with dose 9E+4 times higher with HEPA failure)
- 12. fire in waste compaction glove box = 0.0031 person-rem (C53)
- 13. glove failure = 1.3E-5 person-rem (C53)
- 14. severe glove box damage = 0.061 person-rem (C53).
For FUEL REPROCESSING, estimated population doses from 20 accident scenarios are in Tables C57-C60. The esti-mates for these scenarios are listed below. Note that values from Table C59 assume HEPA filter failure, except where noted.
- 1. loss of fuel storage pool water = 50 person-rem (Table C.58)
- 2. lon-exchange resin fire and explosion = 0.36 person-rem (C57, with normal HEPA filtration),1800 person rem
( (C.57, with failed HEPA filtration), and 0.2 person rem (C.58) k C.11 NUREG/BR-0184
Appendix C O
- 3. criticality = 0.030 person-tem (C57, normal HEPA),0.035 person-rem (C.57, failed HEPA),5 person-rem (C58), l and 2 person-rem (C59, without HEPA filter consideration)
- 4. hydrogen explosion in HAF tank = 1600 person-rem (C57, normal HEPA),1700 person-rem (C57, failed HEPA),
0.07 person-rem (C58), 9E-4 person-rem (C59), and 490 person-rem (C60)
- 5. fire in low level waste = 0.1 person-rem (C58)
- 6. fuel assembly drop = 0.013 person-rem (C57, normal HEPA),1300 person-rem (C.57, failed HEPA),0.1 person-rem (C.58),0.05 person-rem (C59, without HEPA consideration), and 0.0020 person-rera (C.60) l
- 7. explosion in HLW calciner = 4300 person-rem (C.57, normal HEPA),1.3E+4 person-rem (C57, failed HEPA),
6E+6 person-rem (C58, assuming HEPA filter failure),0.2 person-rem (C59), and 510 person-rem (C60)
- 8. krypton cylinder rupture = 50 person-rem (C58) and 40 person-rem (C59, without HEPA consideration)
- 9. explosion in HAW concentrator = 430 person-rem (C57, normal HEPA),9500 perron-rem (C57, failed HEPA),
0.008 person-rem (C59), and 57 person-rem (C60)
- 10. solvent fire in codecontammation cycle = 23 person-rem (C57, normal HEPA),56 person-rem (C57, failed HEPA), i and 2.6 person-rem (C60) '
- 11. explosion in LAW concentrator = 28 person-rem (C57, normal HEPA),48 person-rem (C.57, failed HEPA), and !
3.2 person-rem (C60) I
- 13. solvent fire in plutonium extraction cycle = 3.lE-4 person-rem (C57, normal HEPA) and $20 person-rem (C57, failed HEPA)
- 14. dissolver seal failure = 0.023 person-rem (C57, normal HEPA) and 2300 person-arm (C57, failed HEPA)
- 15. release from hot UF. cylinder = 1.5 person-rem (C57, without HEPA consideration)
- 16. solvent fire in hydrogen concentrator = 7E-4 person-rem (C59)
- 17. red oil explosion in fuel product concentrator = 6E-4 person-rem (C.59)
- 18. explosion in fuel product denitrator = 0.012 person-rem (C59)
- 20. hydrogen explosion in fuel product denitrator fuel tank = 0.012 person-rem (C59).
For SPENT FUEL SIORAGE, estimated population doses from 18 accident scenarios are in Tables C.27, C31, C32, C61, C94, and C.101 Those from Tables C27, C31, C32, and C61 have been combined into 14 accident scenarios whose population doses are listed below. Note that the values taken from Table C27 are those for total body population dose. The values taken from Table C.31 correspond to the drywell storage concept only. Also note that Tables C31 and NUREG/BR-0184 C.12
1 Appendix C C32 are quantified in terms of latent cancer fatalities (LCFs) rather than person-rems. These can be transformed into person-rems via a typical conversion factor such as 200 health effects (or LCFs) per IE+6 person-rems, or inversely 5,000 person-rem / health effect.m Tabic C94 presents population doses for two additional scenarios-spent fuel pool fires
~
due to seismic and cask drop initiators, whose estimated frequencies are in Table C.93-in terms of an " average" and
" worst" case. Thble C.101 addresses two more scenarios, another " average" and ' worst" case, deriving population doses for four pairings of the accident scenarios and selected mitigative options. ;
- 1. collision during highway transport = 0.1 LCF (Table C32, without fire, cask storage concept),0.004 LCF (C32, i without fire, drywell storage concept),0.5 LCF (C32, with fire, cask storage), and 0.02 LCF (C32, with fire,
, drywell storage)
!' 2. tornado = 0.04 LCF (C.32, cask storage) and 0.04 LCF (C32, drywell storage) i
- 0.3 person-rem (C.61, for BWRs) i i
. 4. transport cask drop = 0.006 person-rem (C.27),4E-4 IIF (C32, cask storage),4E-4 LCF (C.32, drywell storage), l l 2 person-rem (C61, PWRs), and 1.8 person-rem (C61, BWRs) l
- 5. cask venting during transport = 0.1 LCF (C32, cask storage) and 0.004 LCF (C.32, drywell storage)
N i
- 7. canister shear during emplacement = 0.004 LCF (C32, drywell storage)
}
8 cask drop during emplacement = 0.006 person-re.n (C27) r.nd 0.004 LCF (C32, cask storage) i
- 9. airplane crash = 0.7.6 LCF (C31, without fire),1.3 LCF (C.31, with fire),0.5 LCF (C.32, with fire, cask toppled, ,
, cask storage). 0.5 LCF (C.32, with fire, cask storage),0.02 IIF (C32, one fuel assembly, with fire, drywell 4
storage), and 0.2 LCF (C32,10 assemblics, with fire, drywell storage)
, 10. earthquake = 0.061 LCF (C31, without fuel pin failure),3.3 LCF (C31, with pin failure),0.1 LCF (C32,24 assemblies, cask storage),10 LCF (C32,2400 assemblies, cask storage),0.004 LCF (C32, one assembly, drywell ]
storage),0.04 LCF (C32,10 assemblies, drywell storage), and 2.4 LCF (C32, 2400 assemblies, drywell storage) i I
- 11. transporter collision during emplacement = 3.4E-5 LCF (C.31, without fire) and 0.0019 LCF (C31, with fire) l 12. transporter collision during retrieval = 5.9E-7 LCF (C31, without pin failure or fire), 3.8E-5 LCF (C31, with pin failure, without fire), 2.6E-6 LCF (C31, without pin failure, with fire), and 2.6E-4 LCF (C31, with pin failure and fire)
- 13. transporter motion with canister partially in place = 0.018 LCF (C31, during emplacement), 5.9E-7 LCF (C31, during retrieval, without pin failure), and 0.0016 LCF (C.31, during retrieval, with pin failure)
- 14. canister drop during attrieval = 9.9E-7 LCF (C.31).
For HLW SIORAGE, estimated population doses from three accident scenarios (after grouping by pairs) are in Table C.62. For GEOIDGIC WASTE DISPOSAL, estimated population doses from 19 accident scenarios are in l
l C.13 NURFO!9R-0184
Ar gndix C O
Tables Cl*, C15, C18, and C.19. Note that Table CIS reports population doses as person-mrems. These are listed as person-rems below. Also note that Ables C18 and C19 generally provide the same values (and are referenced as com-ing from Table C19), except where noted.
- 1. fuel truck crash into HLW area = 2000 person-rem (Table C14)
- 2. fuel truck crash into cladding waste area = 2.0 person-rem (C14)
- 3. fuel truck crash into NHLW area = 40 person rem (C14)
- 4. airplane crash = 4000 person-rem (C14) and 110 person-rem (C19)
- 5. elevator drop = 0.050 person-rem (C14)
- 6. fuel assembly drop = 2.99 person-rem (C15) and 8.0E-5 person-rem (C19)
- 7. NHLW pallet drop = 0.80 person-rem (C14)
- 8. final filter failure = 2.0 person-rem (C14)
- 9. HLW drop = 0.175 person-rem (C15)
- 10. spent fuel handling = 1.29 person-rem (C.15)
- 11. remote TRU drop = 1.98E-4 person-rem (C15)
- 12. contract TRU puncture = 6.70E-8 person-rem (C15)
- 13. nuclear test = 0.0031 person-rem (C19) 14 loading dock fire = 0.0068 person-tem (C19, spent fuel) and 9.2E-4 person-rem (C19, HLW)
- 15. waste handling ramp fire = 3.6E-7 person-rem (C18) and 4.8E 7 person-rem (C19)
- 16. emplacement drift fire = 3.6E-7 person-rem (C18) and 4.8E-7 person-rem (C19) j
- 17. flood = 1.2E-9 person-rem (C.19) l 1
- 18. tornado = 0.0031 person-rem (C19)
- 19. earthquake = 0.0031 person-rem (C.19).
For TRANSPORTATION, it is convenient to identify three categories based on the material being shipped: spent fuel, plutonium ox de, and HLW. For spent fuel transportation, estimated population doses from eight accident scenarios are in i
Tables C.37, C38, and C.65-C69. The estimates for these scenarios are listed below. Note that the values reported from Table C37 are the totals from inhalation, plume gamma, and ground gamma pathways. The values listed beknv correspond to those for the urban area given in Thble C37. The corresponding values for the rural area in hble C37 are 640 times lower. Also note that Table C38 reports population doses from the water ingestion pathway.
NUREG/BR-0184 C.14 l
l
l Appendix C b
- 1. leakage of coolant from spent fuel cask during rail shipment = 5.8E-4 person-rem (Table C65),680 person-rem (C.69, impact fails cask seal, fuel failure),1900 person-tem (C69, side impact fails pressure relief valve, fuel failure),1900 person-rem (C.69, end impact fails pressure relief valve, fuel failure), and 680 person-rem (C69, side impact fails cask seal, fuel failure)
- 2. release fmm a collision during rail shipment = 939 person-rem (C37),182 person-rem (C38),1.9E+4 person-rem (C65),1.7E-6 person-rem (C.67), and 1.1 person-rem (C68, with closure errors)
- 3. release from a collision followed by release of fuel from the cask during rail shipment = 1.35E+4 person-rem (C37, with fire),1.12E+5 person-rem (C.37, with fire and fuel oxidation), 6870 person-rem (C.38, fire), 6.3E+4 person-rem (C38, fire and oxidation), 2.7E+4 person-rem (C65),0.28 person-rem (C.68, for 50-80 km/hr collision),0.28 person-rem (C68,80-100 km/hr),0.20 person-rem (C68, with 1000'C fire for > 1 hr), and 0.20 person-rem (C.68, 800'C for > 2 hr) 1 l
- 4. loss of gases from inner cavity = IE-6 person-rem (C.66, rail shipment) and SE-9 person-rem (C66, truck)
- 5. loss of confinement and 50% fuel damage = 0.1 person-rem (C.66, without fire, rail),2000 person-rem (C66, with
, fire, rail),100 person-rem (C66, without fire, truck), 600 person-rem (C66, with fire, truck), 0.5 person-rem (C67, without fire, rail), and 1700 person-rem (C67, with fire, rail)
, 6. loss of neutron shielding during rail shipment = 8E-7 person-rem (C67) j kO) 7. fall during rail shipment = 0.28 person-rem (C68, for 25 to 40 m fall) and 0.28 persoI
- 8. fire during rail shipment = 0.20 person-rem (C.68,1000'C for > 1 hr) and 0.20 person-rem (C.68,800'C for !
> 2 hr).
For plutonium oxide transportation, estimated population doses from six accident scenarios are in Tables C65 (three scenarios for rail shipment) and C.66 (tlure scenarios for truck shipment). For HLW transportation by rail, estimated ;
population doses from five accident scenarios are in Thbles C66 and C.67. i l
McGuire (1988) estimated the population doses from a major radioactive release for a non-reactor facility to be 40 and 800 person-rem for an effective dose equivalent (EDE) of 5 rems at distances of 100 and 1,000 m, respectively. These can be assumed applicable to either fuel- or non-fuel cycle facilities (see Section C.8).
C2.1.3 Tbtal Accident Risks
~
Total public risks from all accident scenarios have been estimated for 10 of the 13 non-reactor fuel cycle facilities listed in Section C1. Many of these estimated risks are in Thble C.70 after scaling on a consistent basis for comparison (see Section C6). Tables C14, C19, C.31, C32, C35, C.42, and C44 contain additional estimates. The estimates in these eight tables have been assembled into the following table, modeled after Table C.70. The risks from Tables C14, C19, C.31, C32, C35, C.42, and C44 are listed as " unscaled" values, after convening units of health effects or fatalities into person-tems via a conversion factor of 5,000 person-rem / health effect." The " normalized" risks from Table C70 are listed as " scaled" values in Thble C109.
Estimated public risks from three accident scenarios during the postclosure period of GEOIDGIC WASTE DISPOSAL in terms of 10,000-yr health effects for four geologic media are in Table C23. These can be summed to yield the fol-p lowing total public risks:
C.15 NUREG/BR-0184
1 Appendix C O
- basalt = 28.43 heahh effects a tuff = 3.44 health effects 1
- bedded salt = 6.57 health effects
- granite = 9.85 health effects.
These can be converted into person-rems as mentioned above.
C.2.2 Public Health (Routine) l There is considerably less literature on routine public health risks than on accidental risks for non-reactor applications.
For SPENT FUEL STORAGE, estimated routine public risks during the operations and decommissioning phases at a monitored retrievable storage (MRS) facility are in Table C44 in terms of latent health effects (LHEs) per year. These can be transformed into person-rem /yr via a typical conversion factor such as 5,000 person-rem / health effect.m Table C26 also provides the routine public risk during operations at an MRS facility,20 person-rem /yr (total body).
For GEOLOGIC WASTE DISPOSAL, estimated routine public risks during the construction, operations, and decommis-sioning phases of the preclosure period at a repository are in Tables C.9, CIO, Cl3, C42, and C44. Note that the velues in Tables C.9 and C10 are given in terms of the 70- and 50-year dose commitments, respectively. The value from Table Cl3 is taken for the ' reference" case. Also note that the values in Tables C42 and C44 are given in terms of LHE/yr, which can be converted into person-rem /yr as discussed above. Tables C42 and C.44 address the waste man-agement system without and with an MRS facility, respectively. The mutine public risks have been estimated as follows:
- 1. construction = 0.0068 erson-rem T (Table C9, salt medium),100 person-rem (C9, granite),15 person-rem (C9, basalt),38 person-rem (C9, shale), 2.0E+4 person-rem (C10),1E-5 LHE/yr (C.42), and 1E-5 LHE/yr (C44)
- 2. operations = 3.9E+5 person-rem (C10),1.5E-5 person-rem /yr (Cl3),9E-4 LHE/yr (C42), and 8E-7 LHE/yr (C44)
For the postclosure period of geologic waste disposal, estimated routine public risks are in Tables C23 and C24.
Table C23 provides the 10,000-yr health effects for an undisturbed repository in four geologic media. Able C.24 pm-vides 27,000- and 250,000-yr population doses to four body organs resulting from ingestion of drinking water.
For TRANSPORTATION, estimated routine public risks are in Tables C35, C40-C.42, and C44. The values in Table C35 apply exclusively to spent fuel shipment. Tables C40 and C41 present values for both spent fuel and HLW shipment by : ruck and rail to three repository locations for the waste management system without and with an MRS facility, respectively. The risks are given in health effects, which can be converted into person-rems as previously discussed. The values in Tables C.42 and C44 apply to both spent fuel and HLW shipment, assuming that 30% of the spent fuel is shipped by truck and 70% by rail, while all HLW is shipped by rail. Note that the values in T bles C42 and C44 are given in terms of LHE/yr. These can be transformed into person-rem /yr via a typical conversion factor such as 5,000 person-rem / health effect.* Tables C42 and C.44 address the waste management system without and with an MRS facility, respectively. The routine public risks have been estimated as follows:
- spent fuel by truck = 93.80 person-rem /yr (C35, in 1975) and 565.0 person-rem /yr (C35,1985)
- spent fuel by rail = 7.78 person-rem /yr (C35,1975) and 298.0 person-rem /yr (C35,1985)
O NUREG/BR-0184 C16
Appendix C kp) w/
o spent fuel and HLW combined = 0.09 LHE/yr (C42) and 0.03 LHE/yr (C44).
C2.3 Occupational Health (Accident)
There is less literature available on occupational compared to public health risks due to accidents. Information is panicu-larly scarce for non-fuel cycle facilities. Information for fuel cycle facilities is discussed below.
Estimated risks to the worker from accidents are shown below for four of the 13 non-reactor fuel cycle facilities listed in Section C1: MOX fuel refabrication, fuel reprocessing, spent fuel storage, and geologic waste disposal (Fullwood and Jackson 1980).
MOX FUEL REFABRICATION = 7.0E-4 person-rem /GWe-yr FUEL REPROCESSING = 1.0E-4 person-rem /GWe-yr.
For SPENT FUEL SIORAGE, estimated occupational risks due to accidents during the operations and decommissioning phases at an MRS facility are in Table C45. The values are in terms of LHE/yr, which can be transformed into person-rem /yr via a typical conversion factor such as 5,000 person-rem / health effect.m For GEOIDGIC WASTE DISPOSAL, occupational risks due to accidents have been estimated for aggregates of sce-narios during the operations, decommissioning, and retrieval phases in the preclosure period. The estimates are in Tables C21 (decommissioning and retrieval), C43 (operations, without an MRS facility), and C45 (operations, with an
[ MRS facility). The latter two tables provide values in terms of LHE/yr, which can be transformed into person-rem /yr as
(") mentioned above. Table C12 presents an occupational risk estimate for a shaft drop accident during the operations phase.
The information in Tables C18 and C.19 provide both frequencies and worker doses for individual accident scenarios during the operations phase of the preclosure period. These can be converted into occupational risk estimates in a manner similar to that employed in 'Pable C19 for public risk, as shown in Table C110.
C2.4 Occupational Health (Routine)
There is limited literature available on routine occupational health risks. Information for non-fuel cycle facilities is particularly scarce. Information for non-reactor fuel cycle facilities is discussed below.
Estimated risks to the worker from routine operations are included below for four of the 13 non-reactor fuel cycle facilities listed in Section C.1: fuel fabrication, spent fuel storage, geologic waste disposal, and transportation. For FUEL FABRICATION, estimated occupational doses for fabricating PuO2 Powder into unfired pellets and reconstituting the pellets back to powder are in Tables C75 and C76, respectively. Average values and ranges are provided.
For SPENT FUEL SIDRAGE, Table C.45 provides the total routine estimated occupational risks (in LHE/yr) for the operations and decommissioning phases at an MRS facility. These can be transformed into person-rem /yr via a typical conversion factor such as 5,000 person-rem / health effect.m Daling et al. (1990) provide estimates for the decommissioning phase at an MRS facility of 120 person-rem for drywell storage and 128 person-rem for cask storage (see Section CS). Totals for the operations phase at an MRS facility are also provided in Tables C28 and C29, and can be calculated from Table C30. Tables C28-C30 also list the routine occupational risks for separate activities during the operations phase. Note that Table C28 gives these in terms of person-rem /l 000 metric tons of uranium (MTU); C29 lists them in person-rem /yr; and C30 lists them in person-mrem /l 000 MTU (converted to person-rem /l,000 MTU below). The composites for the seven activities from Tables C28-C30 are as follows:
/~'N,
\ \
\d C.17 NUREG/BR-0184
Appendix C O
- 1. receipt, inspection, and unloading = 58 person-rem /1,000 MTU (hble C.28),148.0 person-rem /yr (C29),
0.135 person-rem /l.000 MTU (C30, fmm tmck), and 0.025 person-rem /1,000 MTU (C.30, rail)
- 2. consolidation and packaging = 15 person-rem /1,000 MTU (C28),6.2 person-rem /yr (C.29),0.0036 person-rem /1,000 MTU (C.30, for fuel), and 0.0011 person-rem /1,000 MTU (C30, non-fuel)
- 3. emplacement in storage area = 20 person-rem /l.000 MTU (C.28, including retrieval from storage area) and 7.2 person-tem /yr (C29)
- 4. maintenance / monitoring in storage area = 2 person-rem /1,000 MTU (C28) and 5.3 person-rem /yr (C29)
- 5. retrieval from storage area = 20 person-rem /l.000 MTU (C28, including emplacement) and 7.1 person-rem /yr (C29)
- 6. transfer to process cells = 4.0 person-rem /yr (C.29)
- 7. shipment to repository = 140.9 person-rem /yr (C29).
For GEOIDGIC WASTE DISPOSAL, total estimated routine occupational risks for the construction, operations, decommissioning, and retrieval phases of the preclosure period are in Tables C9, C11, C12, C16, C17, C.21, C43, and C.45. The estimates from Table C9 are in terms of the 70-yr dose commitment; Able C11 reports fatalities over 5-yr construction and 26-yr operations phases; T bles C43 and C.45 give values in terms of WE/yr for the waste man-agement system without and with an MRS facility, respectively. Both fatalities and WE/yr can be trarafurmed into person-rem /yr via a typical conversion factor such as 5,000 person-rem / health effect.m The values from Tables C12 and C21 are taken for the " reference" case. The mutine occupational risks have been estimated as follows:
- 1. construction = 0.18 person-rem (Table C9, salt medium),5,000 person-rem (C9, granite),6,200 person-rem (C9, basalt),1,900 person-rem (C9, shale),0.014 fatality (Cll, salt),0.77 fatality (Cll, tuff),1.6 fhtalities (Cll, basalt),0.1 WE/yr (C43), and 0.1 WE/yr (C45)
- 2. operations = 1.5 fatalities (C11, salt),5.0 fatalities (C11, tuff),7.3 fatalities (C.11, basalt),902 person-rem /yr (C12),0.02 WE/yr (C43), and 0.02 WElyr (C45)
- 3. decommissioning = 6 person-rem /yr (C.21),0.03 WE/yr (C43), and 0.03 WE/yr (C45)
- 4. retrieval = 163 person-rem /yr (C21).
Table C17 lists the routine occupational risks for separate activities during the operations phase at a tuff repository.
Table C16 does likewise for four of the activities listed in Table C17. The estimates fmm hble C16 are as follows:
- 1. teceiving = 44.8 person-rem /yr
- 2. handling and packaging = 6.9 person-rem /yr
- 3. transfer to underground facilities = 6.0 person-rem /yr
- 4. emplacement in boreholes = 12.4 person-rem /yr for vertical emplacement and 8.7 person-rem /yr for horizontal.
O NUREG/BR-0184 C18
. - . . - - - ..--- . . -. . - - . .-. .- .~ . - . - . ~ . . . -
Appendix C
! These values agree well with the corresponding ones in Table C17, 1
For TRANSPOKI'ATION, hbles C.43 and C45 contain estimated routine occupational risks for the waste management i'
system without and with an MRS facility, respectively. These values are given in LHE/yr which can be converted into person-rem /yr as mentioned above.
I C.2.5 Offsite and Onsite Property -
l
! The offsite and onsite property attributes are eranunarl together in this section for non-reactor facilities bemme most of j l the estimates reported in the literature have grouped the associated costs together as cleanup costs. When such costs are ;
j . multiplied by the accident frequencies, measures of economic risk from accidents are obtained. Several of the reviewed - '
- . reports contain economic risk estimates from accidents.
C2.5.1 Fuel Cycle Facilities ;
i information is included below on estimated cleanup costs and/or economic risks have been estimated for five of the 13 non-reactor fuel cycle facilities listed in Section C.1: uranium milling, UF. conversion, fuel fabrication, spent fuel l storage, and transportation. Estimates for URANIUM Mrt IJNG, UF CONVERSION, and FUEL FABRICATION '
are provided in hbles C4-C.6, respectively. Each table provides a best estimate and 80% confidence bounds for the cleanup cost (in 1989 dollars) associated with each accident scenario at the reference facility. Each cost is multiplied by the corresponding estimate for the scenario frequency (also given as a best estimate and 80% confidence bounds) to yield the best estimate and 80% con 6dence bounds for the economic risk associated with each scenario. These scenario risks !
are then summed to give the best estimate and 80% con 6dence bounds for the total economic risk fmm accidents at the i reference facility.
For SPENT FUEL S'IDRAGE, Dble C94 contains estimates of the offsite property damage in 1983 dollars for two -
accident scenarios: spent fuel pool 6res due to seismic and cask drop initiators. Frequency estimated are in ;
hble C93-in terms of an " average" and " worst" case. hble C95 contains estimates of the onsite property damage in !
1983 dollars corresponding to these same two scenarios. Able C101 contains estimates of offsite pmperty damage in l
1983 dollars for four pairings of accident scenarios and selected mitigative options. For TRANSPORTATION of spent j
~
fuel by rail, ranges of estimated cleanup costs for three accident scenarios in 1984 dollars are in Daling et al. (1990) (see Section C5).
C2.5.2 Non-I%ael Cycle Phellities Estimated cleanup costs (presumably in 1986 dollars) which can be associated with the FDUR NON-REACIDR NON-FUEL CELE FACILITIES listed in Section C1 are in Figure C1 and Table C2. They are expressed as functions of the licensed material quantity for both an " average" and " worst-case" release (see Section C3). For all but the service organizations, the average costs are multiplied by the accident frequencies for the corresponding facilities estimated in hble C1 to yield economic risk as a function of licensed material quantity for each of the remaining three facilities in hble C3.
hbles C.7 and C8 contain best estimates and 80% confidence bounds for the cleanup cost (in 1989 dollars) associated with each accident scenario at a REFERENCE MANUFACIURING AND DISTRIBUTION FACILITY EMPIAN-
' ING BYPRODUCT AND SOURCE MATERIAIE and SERVICE ORGANIZATIONS (waste warehouses). Each cost is multiplied by the corresponding estimate for the scenario frequency (also given as a best estimate and 80% confidence I d
C.19 NUREG/BR-0184 i
Appendix C O
bounds) to yield the best estimate and 80% confidence bounds for the economic risk associated with each scenario. These scenario risks are then summed to give the best estimate and 80% confidence bounds for the total economic risk from accidents.
C.3 A Preliminary Evaluation of the Economic Risk for Cleanup of Nuclear Material Licensee Contamination Incidents (NUREG/CR-4825)
In NUREG/CR-4825 (Ostmeyer and Skinner 1987) and a subsequent document (NUREG/CR-5381 [Philbin et al.1990],
see Section C.4), the economic risk of cleanup costs resulting from non-reactor NRC licensee contamination incidents was evaluated. This first study focused only on incidents where the cleanup cost was < $2E+6. Owing to the preliminary nature of this study, little information was assembled on tl.e frequencies, severities, and costs associated with the contamination incidents. The analysis objective was to provide a technical basis upon which to develop a financial cov-erage schedule for a rulemaking which would require certain nuclear material licensees to demonstrate adequate financial coverage for contamination cleanup. The analysis sought to provide three products:
- 1. a rational method to classify licensees according to the potential magnitude and frequency of contamination incidents
- 2. a model to rank the classes of licensees according to potential incident costs
- 3. estimates of the economic risk for licensees in each class.
Three indices were proposed to classify the licensees:
- 1. application /use of the licensed material
- 2. the licensed curie (Ci) activity
- 3. the nuclear material form.
Each class was further divided as follows:
- Class 1 I. research, teaching, experimental, diagnostic, and therapeutic facilities, including hospitals, universities, medical gmups, and physicians II. measurement, calibration, and irradiation facilities, including users of scaled sources III. manufacturing and distribution facilities employing bypmduct and source materials, such as radiopharmaceuticals IV. service organizations, including waste repackagers, pmcessors, and disposers V. non-reactor fuel cycle facilities, handling source and special nuclear material facilities, such as uranium or j thorium ore processors. !
- Class 2 l
This class was subdivided into seven categories ranging from facilities licensed to handle quantities _< 0.01 Ci to ones licensed to handle > 1,000 Ci, with each subclass spanning a factor of 10 in licensed Ci quantity. _ i l
1 NUREG/BR-0184 C.20 l
l
Appendix C p
i
- Class 3
- 1. licensees handling sealed sources II. licensees handling non-encapsulated Group A sources (i.e., sources whose potential release fraction is < 0.1) l III. licensees handling non-encapsulated Group B sources (i.e., sourecs whose potential release fraction is 10.1.).
Frequencies of contamination incidents were determined for the Class-1 licensees using historic data from the NRC's Non-Reactor Event Report (NRER) database (spanning 1980-1986 at the time of the study). These frequencies are tabulated in ;
Table C.I. Costs were developed from 19 historic events and orderaf-magnitude estimates for selected groups oflicensee l incidents. They have been plotted as a function of licensed Cl quantity in Figure C.1 for two cases:
- 1. a " worst" case, where 100% of the licensed quantity was assumed to be released
- 2. an " average" case, where only 15% of the licensed quantity was assumed to be released.
Cleanup costs were assigned to five of the seven divisions of Class-2 licensees at the geometric midpoints of each division's ra' ige from Figure C.I. These are listed in Table C.2 for both the worst (licensed quantity released [LQR]) and average cases.
The economic risk was defined as the product of the incident frequency (according to index Class 1) and the cleanup cost (according to Index Class 2). Using the incident frequencies from Table C.1 and the average cleanup costs from Table C.2, the economic risk per Class-1/ Class-2 licensee is tabulated in Table C.3. Division IV from Class I was excluded due to the lack of available data for frequency estimation. Division V from Class I was excluded because the incidents required cleanup costs .2. $2E+6, which fell outside the study scope.
Also provided in NUREG/CR4825 were the following:
a tabulation of the contammation incidents from the NRER database (1980-1986) and the NRC's OMIT and Fuel Cycle databases (pre-1980), in NUREG/CR4825 Appendix B a tabulation of the historic cost data for cleanup, in NUREG/CR4825 Appendix C the development of a simple cost model which estimates cleanup cost from contaminated floor space, in NUREG/CR-4825 Appendix D.
C.4 Economic Risk of Contamination Cleanup Costs Resulting from Large Non-Reactor Nuclear Material Licensee Operations (NUREG/CR-5381)
In NUREG/CR-5381 (Philbin et r'.1990) and (NUREG/CR4825 [Ostmeyer and Skinner 1987], see Section C.3), the economic risk of cleanup costs resdting from non-reactor NRC licensee contammation incidents was evaluated. This latter study focused only on incidents at large non-reactor licensees where the cleanup cost was .2. $2E+6. Five categories of non-reactor licensees were identified, with a reference facility chosen for each:
- 1. uranium mines and mills, represented by the White Mesa Mill in Blanding, Utah, as described in NUREG/CR-5381 Appendix A O
b C.21 NUREG/BR 0184
Appendix C l
- 2. uranium hexafluoride (UF.) conversion plants, mpresented by the Sequoyah Plant in Gore, Oklahoma, as described in 9l NUREG/CR-5381 Appendix B j
- 3. uranium fuel fabrication facilities, represented by the Westinghouse Facility in Columbia, South Carolina, as described in NUREG/CR-5381 Appendix C
- 4. large manufacturers and/or distributors of nuclear byproducts, represented by the DuPont Facility in Nonh Billerica, ,
Massachusetts, as described in NUREG/CR-5381 Appendix D l
- f. nuclear waste warehouses, represented by ADCO Services in Tmley Park, Illinois, as described in NUREG/CR-5381 Appendix E.
The approach taken in NUREG/CR-5381 consisted of the following steps:
1 describe each reference facility, postulating accident scenarios for each process in terms of the radioactive material releases, incident frequencies, decontamination efforts required, and decontammation costs for property cleanup and waste disposal defme incidents from historic data and systems analysis, covering the risk-dommant ones (i.e., the range from high frequency-low consequence events to those with low frequencies but high consequences; decontammation models were employed for the latter pair when historic data were unavailable)
- calculate the economic risk in 1989 dollars as the sum of the products of frequency and cost for each incident, including uncertainty analysis. In essence, the economic risk is the expected cost to decontammate the property in the event of a radioactive release at the facility.
Where available, historic data for actual or similar facilities were used to estimate the incident frequencies and cleanup costs. In lieu of these, historic data fmm related industries were employed. Mathematical models were developed to esti-mate frequencies and costs where no historic data were available. For each point estimate, upper and lower bounds were specified for an 80% confidence interval. These were propagated to yield 80% confidence bounds on both the individual scenario economic risk and the total economic risk for the sum of all the scenarios for a facility.
Ables C.4-C.8 list the incident scenarios, consequence descriptions, cleanup costs, annual frequencies, and annual eco-nomic risks for each of the reference facilities. The uncertainty bounds are included for the latter three parameters. As part of the reference facility descriptions, the radioactive inventories and curies released per accident are tabulated in Appendices A-E of NUREG/CR-5381. The contamination incidents for all five licensee classes based on NRC's NRER, OMIT, and Fuel Cycle databases are listed in Appendix F to NUREG/CR-5381. The NRER database included incidents from 1980 onward, while the others included only pre-1980 incidents. The OMIT database focused on non-fuel cycle activities, while the Fuel Cycle database addressed non-reactor fuel cycle operations. Note that neither Table C.5 nor Table C.6 includes a major UF. release that occurred at the Sequoyah nuclear power plant. Only accidents at uranium hexafluoride conversion plants and ful fabrication facilities were considered in the development of hbles C.5 and C.6.
C.5 Preliminary Characterization of Risks in the Nuclear Waste Management System Based on Information in the Literature (PNL-6099)
In PN1 6099, Daling et al. (1990) surveyed literature on the following three components of the nuclear waste management system to develop a preliminary characterization of the associated risks:
NUREG/BR-0184 C.22
Appendix C
\
p) v
- the waste repository (in tuff, salt, and basalt media) i
- the MRS facility
- the transportation system supporting both of these.
1 Five risk categories were dermed, of which only those associated with radiological exposure are of interest in this appendix:
)
1 1
- 1. public and occupational risks from radiological release accidents l l
- 2. public and occupational risks from radiological exposure during routine operations I
- 3. economic risks resulting from radiological release accidents. l l
l For the repository, both the preclosure (construction, operations, decommissioning, and retrieval phases) and postclosure !
periods were addressed. For the MRS facility, the construction, operations, and decommissioning phases were examined.
For the transportation system, only operations were considered. Construction and decommissioning of transport equip-ment were not addressed.
For each component of the waste management system, descriptions for reference facilities and processes were developed, primarily based on conceptual designs (see Chapter 3 of PNI-6099). These were used to form composite risk estimates
,m from all the reviews on a consistent basis by scaling to the reference facilities. Daling et al. (1990) first presented r, levant (V) datacomposites, taken from the reviewed documents prior to their combination into composite risk estimates as scaled for the reference facilities, were provided.
The repository preclosure period has been fairly well eramined with respect to risk estimation. Tables C.9-C.11 list exposures for the construction phase. The operations phase has been addressed extensively, as indicated by the data presented in 'Ihbles C.12-C.20. Limited information was available on the latter two phases of the preclosure period (decommissioning and retrieval). Able C.21 summarizes this information. Data for the repository preclosure period on a normalized basis is compared in 'Ihble C.22.
The repository postclosure period also has been examined quite well, although the estimates are usually very uncertain due l to the extremely long time scale considered. Table C.23 lists the health effects associated with four accident scenarios for a waste repository in four different geologic media. Table C.24 lists accumulated doses by body organ for a repository in a tuff medium. Conditional cancer risks from ingestion for six different accident scenarios are given in hble C.25.
For the MRS facility, no radiological risks exist during the construction phase. Radiological risks arise during the l operations phase "Ihbles C.26 and C.27 provide 50-year dose commitments during the operations phase under routine and accident conditions, respectively. For the three accident scenarios listed in Table C.27, the following frequencies were assumed: 1) fuel assembly drop - reasonable chance of occurring annually; 2) shipping cask drop - reasonable chance of occurring once during the facility lifetime; and 3) storage cask drop - unlikely to occur, but requiring consideration.
Occupational doses for standard activities during the operations phase are tabulated in hbles C.28-C.30. For drywell storage in the MRS facility, operations phase risks from selected accident scenarios are shown in hble C.31. Operations phase risks due to accidents for both drywell and cask storage concepts are listed in Table C.32. The following radiologi-cal risks to the worker from routine operations during the decommissioning phase were estimated: 120 person-rem for
,. drywell storage and 128 person-rem for cask storage.
( &,
V C.23 NUREG/BR-0184
I Appendix C )
O !
The radiological risks fmm transportation have been examined extensively. Dose ra'es and total doses under normal (non-accident) shipping conditions for spent fuel transport by truck and rail cask are listed in Tables C.33 and C.34. Note that both tables were based on a shipping cask modeled as an infinite line source. Thus, the doses reported are reasonable from 3 m to 15 m but probable overestimates beyond 40 m away. Radiological risks are given in Table C.35. Dose i estimates from selected accidents during rail shipment of spent fuel are provided in Tables C.36-C.38. Transportation I risks under both normal and accident conditions have been combined for truck and rail shipments of spent fuel in l Table C.39. The risks encour.tered during routine transportation (i.e., non-accident) for a waste management system without and with an MRS facility are listed in Tables C.40 and C.41, respectively, for both spent fuel and HLW shipment. I A range of cleanup costs (1984 dollars) were estimated for three accident classes for spent fuel transportation by rail:
- 1) impact = $2.0E+5 - $9.5E+6; 2) impact with burst = $1.4E+6 - $7.0E+7; and 3) impact with burst and oxidation
= $1.3E+7 - $6.2E+8.
The radiological risks from all three components of the waste management system were converted into composite estimates for the reference facilities assuming a throughput of 3,000 MTU/yr, a maximum repository capacity of 70,000 MTU, and a conversion factor of 2.0E-4 LHE per person-tem.m Public and occupational risks from the preclosure period of the waste management system without an MRS facility are tabulated in Tables C.42 and C.43, respectively. The corres-ponding risks for the system with an MRS facility are provided in Tables C.44 and C.45, respectively. Total risks for the preclosure period are given in Table C.46. Table C.47 summarizes the annual and total life-cycle risks for the entire waste management system.
C.6 Preliminary Ranking of Nuclear Fuel Cycle Facilities on the basis of Radio-logical Risks from Accidents In an unpublished PNNL study, Pelto et al. examined the risk to the public and plant worker from radiological accidents at non-reactor nuclear fuel cycle facilities. The study was essentially a literature survey, similar to that of PNL-6099 (Daling et al.1990 [see Section C.5]), but focusing on all non-reactor fuel cycle facilities, rather than just those associated with nuclear waste management. The 13 categories of non-reactor fuel cycle facilities listed in Section C.I.1 were identified.
Representative non-reactor fuel cycle facilities were selected for each of the 13 categories based on actual facilities or con-ceptual designs provided by Schneider et al. (1982). These representative descriptions, including site characteristics, were combined with the ALLDOS computer code (Strenge et al.1980) to scale the consequences of radioactive release on a consistent basis. Radiological risk was measured in whole body person-rem /GWe-year (i.e., in terms of the annual requirements of a 1,000-MWe [1-GWe] LWR) as the 50-year population dose commitments for selected organs, based only on the airborne pathway. Although the source documents reviewed by Pelto et al. were dated prior to 1983, they are felt to provide at least conservative results. Any subsequent refinements to the facilities would have tended to reduce risks ;
based on " lessons learned." 4 i
Fullwood and Jackson (1980) estimated the radiological risk to the plant worker, citing the fr' lowing pair of values:
- 1) 7.0E-4 person-rem /GWe-year for MOX fuel refabrication, and 2) 1.0E-4 person-rem /GW . car for fuel reprocessing.
The remaining literature addressed public risk as discussed below.
Cohen and Dance (1975) performed a risk analysis for uranium milling, yielding an expected pcpulation dose (public risk) of about 0.001 person-rem /GWe-year mainly due to the release of mill tailings slurry. Three accident scenarios were identified, and their frequencies and population doses were estimated as tabulated in Table C.48. Cohen and Dance also l performed a risk analysis for the conversion phase of the fuel cycle, obtaining an expected population dose ranging from 7.6E-4 to 0.0056 person-rem /GWe-year mainly due to a hydrogen explosion during the reduction step. Six accident scenarios were identified, and their frequencies and population doses were estimated as provided in Table C.49. !
l NUREG/BR-0184 C.24 )
l l
l
Appendix C
/%
i )
V Cohen and Dance (1975) also give risk estimates for enrichment and fuel fabrication. For enrichment, the expected popu-lation dose ranged fmm 0.0025 to 0.0037 person-rem /GWe-year, dominated by release from a hot UF. (uranium hexaflu-oride) cylinder. The frequencies and population doses from the four accWe m 'os considered for this phase of the fuel cycle are tabulated in Table C.50. For fuel fabrication, the expected popotation dose ranged widely fmm 4.8E-5 to 0.010 person-rem /GWe-year, again dominated by release from a hot UF cylinder. Seven accident scenarios were identi-fled and quantified as shown in Table C.51.
Cohen and Dance (1975), Erdmann et al. (1979), and Fullwood and Jackson (1980) addressed the public risk associated with MOX fuel refabrication. The ranges of expected population dose are listed along with the dominant risk contributors in Table C.52. Tables C.53-C.55 present the seven or eight accident scenarios considered for this phase of the fuel cycle, along with the associated frequencies and population doses. The relatively low risk and population doses estimated by Fullwood and Jackson (1980) indicated that results were sensitive to modeling assumptions. The same set of studies also examined the public risk associated with the fuel reprocessing phase of the fuel cycle, yielding the ranges of expected population dose and dominant risk contributors given in Table C.56. Eight to 12 accident scenarios were identified and i quantified for this phase; these are listed and quantified in Tables C.57-C.59. Six accident scenarios from a study by l Cooperstein et al. are presented in Table C.60, although a public risk estimate was not generated in the report.
Karn-Bransle-Sakerhat (1977), the DOE (1979), and Erdmann et al. (1979) addressed the spent fuel storage phase of the nuclear fuel cycle, estimating expected population doses ranging from 1.7E-6 to 8.9E-5 person-rem /GWe-year, dominated by either a fuel basket or fuel assembly drop accident. The frequency and population dose for the fuel assembly drop accident in Erdmann et al. (1979) were taken from their analysis for the fuel reprocessing phase (see Table C.58). Karn- 1 Bransle-Sakerhat (1977) identified and quantified fuel transfer basket and fuel assembly drop accidents, as indicated in j m Table C.61. The public risk fmm HLW storage accidents was exanuned by Smith and Kastenberg (1976), who reported fq, }
an expected population dose of 2.3E-4 person-rem /GWe-year mainly due to a major rupture of a waste canister combined with the independent failure of one HEPA filter. Six accident scenarios were identified, and their frequencies and .
population doses were estimated as tabulated in Table C.62. l Geologic waste disposal has been the subject of several risk studies. Two of the studies, DOE (1979) and Erdmann et al.
)
(1979), were reviewed by Pelto et al. The expected population doses varied widely between these two studies for the pre- j closure period of geologic disposal, as indicated in Table C.63. The Analytic Sciences Corporation (TASC 1979) reviewed the peak individual dose (rem / year) to the critical organ during the postclosure period as determined frorr. other studiei. Figure C.2 summarizes these results. Erdmann et al. (1979) estimated an expected population dose of 5.0E-11 person rem /GWe-year for the postclosure period.
Risks ssociated with the transportation phase of the nuclear fuel cycle have been investigated by Cohen and Dance (1975),
Erdmann et al. (1979), Fullwood and Jackson (1980), the DOE (1979), the NRC (1975a,1975b,1976,1977), Berman et al. (1978), the U.S. Atomic Energy Commission (AEC 1972), and Hodge and Jarrett (1974). Table C.64 summarizes the expected population doses from accidents during plutonium oxide, spent fuel, and HLW shipment. Table C.65 lists the frequencies and population doses for accident scenarios associated with spent fuel and plutonium oxide transportation, by rail and truck, respectively, as determined by Cohen and Dance (1975). Erdmann et al. (1979) identified accident scenarios for four transportation systems: spent fuel by rail and truck, plutonium oxide by truck, and HLW by rail. The associated frequencies and population doses are tabulated in Table C.66. Fullwood and Jackson (1980) examined rail shipment of spent fuel and HLW, identifying and quantifying the accident scenarios presented in Table C.67. Projekt Sitherkeitsstudien Entsorgung (PSE 1981) and Elder (1981) identified and quantifici transportation accident scenarios for rail shipment of spent fuel (Tables C.68 and C.69), although they did not convert these estimates into expected population doses.
Having surveyed available literature and extracted the quantitative information deemed representative of non-reactor fuel O cycle risks, Pelto et al. then scaled the risk estimates on a consistent basis for the purpose of comparison. Site-specific s
C.25 NUREG/BR-0184
Appendix C conditions for the representative facilities were input to the ALLDOS computer code to yield the public risks fmm each O
nuclear fuel cycle element as summarized in Table C.70. Those elements with comparable risks were grouped together ;
into two categories as follows: 1) conversion, enrichment, MOX fuel refabrication, fuel reprocessing, spent fuel storage, j and transportation, with expected population doses from 0.012 to 0.27 person-rem /GWe-year; and 2) milling, fuel ,
fabrication, HLW (solidified) storage, and geologic waste disposal (preclosure period), with expected population doses from 4.0E-5 to 0.00$0 person-rem /GWe-year.
C.7 Cost-Benefit Analysis of Unfired PuO2 Pellets as an Alternative Plutonium Shipping Form (NUREG/CR-3445)
N UREG/CR-3445 (Mishima et al.1983) is of interest not so much for the value-impact analysis performed (which was fairly preliminary), but for the data presented on industry costs and occupational exposure incurred during the pelletizing and reconstitution processes for PuO2 . Mishima et al. (1983) considered the potential costs of altering the current practice of shipping PuO 2as a powder to one where it is shipped as unfired pellets. The pellets would then be reconstituted into powder following receipt at the fuel fabrication facility. Direct costs (measured in 1983 dollars) consisted of equipment, labor, redesign of process and transport procedures, supplies, services, and additional transport costs. A facility throughput of 20 kg/ day was assumed.
Capital equipment costs for pellet fabrication and powder reconstitution are listed in Tables C 71 and C.72s respectively.
Tables C.73 and C.74 present operating costs associated with the startup and pmcess, respectively, for both pellet fabrica-tion and powder reconstitution. Indirect costs (occupational doses) are summarized in Tables C.75 and C.76 for pellet fabrication and powder reconstitution, respectively.
C.8 A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees (NUREG-1140)
In NUREG-1140, McGuire (1988) performed a regulatory analysis covering emergenev preparedness for non-reactor nuclear facilities, both fuel and non-fuel cycle. It contained five of the six steps required in a regulatory analysis, omitting only the last (implementation). The regulatory analysis began with the following statement of the problem:
"Should the NRC impose additional emergency preparedness requirements on certain fuel cycle and other radio-active material licen.<ees for dealing with accidents that might have offsite releases of radioactive material?"
The objective was to answer this question and, if answering yes, determine how to impose the requirements.
The identification and prelimmary analysis of alternative approaches to the problem came next. A description of the proposed actions and justification for their need were spelled out. Three altematives were cited:
- 1. adopting a regulation containing the proposed requirements
- 2. imposing the requirements by license condition
- 3. imposing no new requirements (the status quo, or baseline, case).
NUREG/BR-0184 C.26 O
Appendix C l U
As part of the preliminary analysis, McGuire (1988) established the following criterion for deeming an accident signifi- ,
cant. A release causing a person outside the plant along the plume centerline to receive an EDE > 1 rem, a thyroid dose l
> 5 rems, or an intake of soluble uranium > 2 mg would constitute a significant accident. These values were chosen ;
from the lower ends of the dose ranges for which the EPA states that protective actions should be considered. Fifteen l classes of licensees were identified, from which those which could have significant accidents were identified for further analysis. Those identified consisted of the following:
- Fuel Cycle Facilities !
- uranium mills
- UF. conversion plants
- enrichment plants uranium fuel fabrication plants
- plutonium fuel fabrication plants spent fuel storage facilities l spent fuel reprocessing plants ;
nuclear fuels research facilities (special nuclear materials). I Byproduct Material Facilities (only those handling large enough quantities of unsealed radioactive material so that the need for offsite emergency preparedness should be considered) l
- radiopharmaceutical manufacturers sealed soutre manufacturers.
(__/ For the estimation and evaluation of values and impacts, McGuire (1988) performed the following three steps for each facility class:
I
- 1. survey the accident history, including similar facilities in the database l
- 2. quantify the accident source terms, using NRC analyses of several severe accidents possible at non-reactor facilities
- 3. calculate the offsite dose via a " standard" dose calculation (i.e., assume a release fraction, atmospheric dispersal j model, and three exposure pathways [ inhalation and cloud- and ground-shine]). l l
The number of licensees potentially affected consisted of 14 fuel cycle and about 17 byproduct material licensees. Of the three alternatives approaches to the problem identified earlier, the first two would have the same values and impacts, and the third represented the fueline case for comparison. Thus, only one value-impact analysis was performed, with the value measured in tern:; cf public risk reduction.
Two cases were considered for estimating the risk reduction. The first assumed a release occurred with an EDE of 5 rems at a distance of 100 m under the Pasquill Class F atmospheric stability condition and a wind speed of 1 m/s. Under these conditions, the area over which the EDE would exceed 1 rem was estimated to be 0.006 mi2 For a typical population 2
density of 3000/m at the facilities, about 20 people would be in the estimated area, with 80% (16) indoors and the remainder (4) outdoors. An outdoor person would receive an average dose of about 3 rems, while one indoors would receive 1/2 of that due to protection from the building. For the base case, this amounted to a total collective dose of about 40 person-rems. The dose savings was assumed to be 1/2 of that, or about 20 person-rems. If 0.0001 cancer death occurred per rem, the number of lives saved would be about 0.002 for the worst meteorology, or about 2E-4 for an overall
' average meteorology, p
1 i 8
\
C.27 NUREGIBR-0184
Appendix C O
To estimate the fr pncy of a major release, McGuire (1988) used statistics from the insurance industry. A fire loss occurred in unsprinklered commercial and industrial facilities at a rate of about 0.006/yr. Where available, sprinklers failed at a rate of 0.038/ demand. Thus, a reasonable estimate of the fire loss rate for a sprinklered facility (typical of radioactive licensees) would be about 0.006/yr x 0.038, or 2E-4/yr. Assuming additional site-specific factors would halve this rate, an estimate of IE-4/yr was generated for the frequency of a major radioactive release. When multiplied by the consequence estimate of 2E-4 life saved on average, an estimate of 2E-8 life saved per facility per year was obtained as the public risk reduction. In monetary terms, this translated to $0.2/ facility-yr, assuming a value of $1E+7/ life.
The second case analyzed was essentially equivalent to the first, except that the 5-rem EDE was now assumed at a distance of 1,000 m. This translated into an increase in the area over which the EDE would exceed I rem to 0.15 mi2, encompas-sing 450 people. Retaining the other assumptions from Case 1, the public risk reduction for Case 2 was estimated at 4E-7 life saved per facility per year, or $4/frility-yr.
Costs to implement the proposed action were based on data from two radiopharmaceutical manufacturers, coupled with the assumption that the licensee would be required to have a 50-page plan containing instructions for what to do in the event of an emergency such as a fire. The initial setup would cost $84,000 (58,400/yr spread over 10 years) for a small program and $550,000 ($55,000/yr) for a large program. Labor costs were assumed to be included as 1/2 to 2/3 of these costs at a rate of $30/hr. For either p:ogram, the annual operating cost would be $18,000. Thus, the industry costs were estimated to be about $26,000/ facility-yr for a small program and $73,000/yr for a large one. The NRC cost to review and inspect the plan was estimated to be $4,000/ facility-yr, yielding total cost estimates of about $30,000/ facility-yr (small program) and $77,000/ facility-yr (large program).
For the presentation of results, McGuire utilized a simple table, as follows:
Licensee Size Cost Benefit Small $30,000/ facility-yr 50.2/ facility-yr Large $77,000/ facility-yr $4/ facility-yr The expected life savings amounted to 2E-8/ facility-yr for small licensees and 4E-7/ facility-yr for large ones. Roughly 20-30 small and 2-3 large licensees could be expected to achieve these savings. These results clearly indicated that the potential risk reduction to the public was very small.
The decision rationale for this regulatory analysis was summarized as follows:
"The cost of this [ emergency] preparedness may not be justified in terms of protecting public health and safety.
Rather, we would justify it in terms of the intangible benefit of being able to reassure the public that, if an accident happens, local authorities will be notified so they may take appropriate actions."
"Although emergency preparedness for fuel cycle and other radioactive material licensees cannot be shown to be cost effective, the NRC feels that such preparedness represents a prudent step which should be taken in line with the NRC's philosophy of defense-in-depth, to minimize the adverse effects which could result from a severe accident at one of its facilities."
McGuire (1988) also presented dose tables for various accident releases at selected fuel and non-fuel cycle facilities.
Tables C.77-C.81 address selected fuel cycle facilities. Tables C.82-C.86 present doses for non-fuel cycle facilities (i.e.,
byproduct material facilities).
C.28
Appendix C l')i t
V C.9 Regulatory % pact Analysis of Final Environmental Standards for Uranium Mill Tailings at Active Sites (EPA 520/1-83-010)
In EPA 520/1-83-010 (EPA 1983), the EPA performed a regulatory impact analysis covering uranium mills. Specifically, EPA addressed the disposal of uranium mill tailings at active sites by evaluating the impact of final environe-l.;J standards for this disposal. The standards considered were ones which addressed only the disposal of mill tailings; releases during the operations phase of a uranium mill were not included. The study contained the six steps required in a regulatory analysis, following Executive Order 12291 (see Section 1).
The statement of the problem was essentially to investigate final environmental standards for disposal of uranium mill tail-ings in both the short and long term. Uranium mill tailings pose an environmental hazard through the release of radon, a radioactive gas. Four methods of controlling these releases were identified:
- 1. discourage misuse (e.g., use of tailings in constmetion of homes) 2 provide barriers to radon emission
- 3. prevent the spread of tailings
- 4. protect the tailings from water intrusion.
7 The objective was to determine which of many alternative standards proposed to limit emissions from uranium mill tailings would be optimal from a health and cost perspective.
The identification and preliminary analysis of altemative approaches to the problem addressed 13 proposed standards for disposal. These standards were defined according to the ability to control radon release after disposal (in terms of radon release rates) and the length of time for which such control would be required. The spectrum of alternatives is displayed in Table C.87, ranging from a baseline case of no controls (Alternative A) to the most stringent case limiting radon release to 2 pCi/m'-s using passive control for 1,000 years, with improved radon control during operations for new piles (Alterna-tive DS). Both existing and new tailings piles (at both existing and future facilities) were considered.
As part of the preliminary analysis, the status of licensed conventional U.S. mill sites as of 1/1/83 was ascertained and tabulated in EPA 520/1-83-010 (EPA 1983) Chapter 2. Characteristics of the control methods for both existing and new piles were specified for the 13 alternative standards in Thbles C.88 and C.89, respectively.
EPA next proceeded to the estimation and evaluation of values and impacts. The value was quantified in terms of health effects averted through control of radon emissions. This was accomplished in two steps. First, each alternative was
- characterized in terms of how well it provided for the following three items
- 1. stability of the tailings pile
- 2. control of radon emissions from the pile 3 protection of the pile against water intrusion.
i b
C.29 NUREG/BR-0184
Appendix C O
These are summarized in Table C.90. Next, the values were quantified on a comparative basis through the definition of an
' effectiveness index" for the four release control methods previously identified. Each alternative was rated in terms of this index using a scale from 1 to 10, considering the factors shown in Table C.90. A weighted average effectiveness was calculated for each alternative.
Costs for disposal of existing and new mill tailings piles were estimated in 1983 dollars for the control method associated with each alternative based on selected model pile sizes (2, 7, and 22 metric tons (MT) for existing piles; 8.4 MT for new piles). The average cost per effectiveness index was calculated for each alternative as the ratio of the model pile disposal costs to the previously estimated effectiveness index. These were then converted to the incremental cost per alternative i as follows:
(Disposal Cost, - Disposal Costa)/(Effectiveness Index, - Effectiveness Indexa)
These calculations are summanzed in Table C.91 for both existing (all three sizes) and new tailings piles.
The incr-mental costs were plotted against the effectiveness indices for the various alternatives for each model pile size (see Figure C.3). The alternatives exhibiting negative or small positive slopes in the plot were the desirable ones. Sensi-tivity analyses were conducted by varying the weighing factors for the effectiveness index and considering the cost per effectiveness index for 100 rather than 1,000 years.
The analysis of industry cost and economic impact was the next item. Thirty-seven economic impact cases for the 13 alternative standards were identified by considering the following three categories for each of the 12 non-baseline alter-natives (i.e., all but Alternative A):
- 1. existing mill tailings
- 2. new mill tailings at existing mills
- 3. new mill tailings at new mills.
For existing tailings, disposal costs were assumed to be incurred from 1983 through 1987. For new tailings, disposal costs were assumed to be incurred from 1983 through 2000. Present wonh calculations were performed for three discount rates (0,5, and 10%). The cost estimates for all 13 altemative standards are summarized in Table C.92.
The presentation of results consisted of the various tables and figures produced during the value-impact analysis, especially the summary Tables C.90 and C.92. The decision rationale for selection of a recommended disposal standard was as follows. The standards were based on current population data, with no " relaxation" for " remote" sites. Passive controls were preferred over institutional ones because of the need to provide long-term protection. The radon emission limit of 20 2
pCi/m -s was selected since both the cost-effectiveness and practicality of providing additional radon control dropped rapidly below this threshold. As a result, Al'ernatiw C3 was recommended since it best met these criteria while minimiz-ing economic impact and providing high, although not maximum, values.
The implementation step of the regulatory analysis was briefly addressed when EPA considered the relationship of the pro-posed standards to the Regulatory Flexibility Act (see Guidelines Section 5.2). An analysis of compliance with this Act was cited as unnecessary because the standards would not significantly impact a substantial number of small entities.
C.30
Appendix C
~'s (V
C.10 Value-Impact Analysis of Accident Preventive and Mitigative Options for Spent Fuel Pbols (NUREG/CR-9281)
In NUREG/CR-5281, Jo et al. (1989) conducted what essentially amounted to a regulatory analysis of a non-reactor
- nuclear fuel cycle facility using the 1983 Handbook (Heaberlin et al.1983) as guidance. It included the six steps required in a regulatory analysis. In the statement of the problem, Jo et al. observed that spent fuel pools at power reactor sites were being required to store more fuel than originally anticipated because of the lack of a waste reprocessing plant or repository. The objective of the analysis was to assess possible preventive and mitigative strategies for spent fuel pool accidents in light of the pools being used to store more spent fuel than originally anticipated.
In the idenuncation and prelimmary analysis of alternative approaches to the problem, Jo et al. proposed three main alter-natives for spent fuel pool accident prevention and mitigation:
- 1. reduction of pool inventory
. 2. improvement of reliability of pool makeup water
- 3. implementation of one or more ' representative" mitigative options.
i Under the first alternative (inventory reduction), limited low-density fuel storage would be permitted in the pool. Essen-tially, fuel discharged from the reactor within the past two years would be stored in a low-density configuration, promot-ing air cooling of the fuel in the event of a loss of pool water inventory. This alternative would require that a utility O replace its current high-density storage racks with Icw-density ones, increasing the need for added storage capacity. Five 1
h options were considered:
- 1. supplemental wet pool storage 4. storage in a cask 1 '
- 2. drywell storage 5. storage in a silo.
- 3. storage in a vault The prelimmary analysis consisted of collecting spent fuel and fuel pool data for all U.S. plants through 1986 (presented in NUREG/CR-5281 Cl: apter 3).
The analysis proceeded to the estimation and evaluation of values and impacts (Alternative 1), using the 1983 Handbook
- as a guide. Risk-dommant sequences for a spent fuel pool were identified. They consisted of structural failure due to an
- carthquake and a compromise of structural integrity through impact of a heavy object, such as a storage cask. For this lat-ter accident, the conditional probability of pool structural failure was taken to be one. Public health and offsite property damage were estimated using the MACCS computer code (Chanin e' al.1990), specifying both a best-estimate and worst-case radiological source term. Accidental occupational exposure was assumed to be similar to that from TMI-2 (i.e.,
< 4580 person-rem). Onsite property damage was assumed to result from loss of pool inventory followed by a zircaloy fire which spread throughout the pool. This resulted in the melting of 1/2 of the fuel cladding and contammation of containment, with a subsequent loss of contamment integrity. The accident frequencies, offsite consequences (public health and property damage), and onsite property damage are tabulated in Tables C93-C95, respectively. The costs (in 1983 dollars) given in Tables C94 and C95 were expanded on a plant-by-plant basis in NUREG/CR-5281 Appendix A, serving as input to the industry cost estimates provided in Table C%.
m C.31 NUREG/BR-0184
Appendix C O
The presentation of results (Alternatiw 1) consisted of two summary tables. The first (Table C.97) listed all parameters affecting the attributes considered in the value-impact analysis, including data references. The second (Table C.98) was the standard value-impact analysis summary table in the 1983 IIandbook, including the net value and ratio calculations for both the best-estimate and worst cases. Additional value-impact measures were indicated in the second table (i.e., the ratio of benefits (in dollars) to cost and the cost of implementation per averted person-rem).
Sensitivity studies were performed by varying the following:
- pool failure probability a site economics e discount rate
- meteorology.
- monetary conversion factor for health effects Only the first item (increase in failure probability) could shift the net value to the positive side. Based on the analysis results, the decision rationale for Alternative 1 concluded that it was not justified due to the negative net value and low ratios, indicative of an action whose overall effect is undesirable.
Alternative 2 (improvement of pool makeup water reliability) addressed the problem of interruption of the circulation of pool cooling water. Such interruption could result in a pool temperature rise until boiling would occur. Thermal-hydraulic analyses from FSARs indicated a considerable time lag between loss of circulation and uncovering of fuel assemblics. Therefore, much time would be available to restore normal cooling or implement a standby cooling option.
In the estimation and evaluation of values and impacts (Alternative,2), it was decided to examine four " generic" pool cool- ,
ing and makeup systems, ranging from the minimum Standard Review Plan (SRP) requirement to crediting three makeup trains, including the fire system. Scoping calculations were performed to estimate failure frequencies. These are quantified in Table C.99. Radiological impacts were found to be negligible. Further quantification was conducted only for averted cost (resulting from replacement power until pool cooling is restored) and industry implementation costs (dis-counted at 10%), with the costs in 1983 dollars. Table C.100 is essentially the presentation of results (Alternative 2) and indicates very small ratios of averted to implementation cost for each of the four systems. Thus, the decision rationale was that Alternative 2 would not be justified.
- Alternative 3 consisted of the following three representative mitigative options for spent fuel pool accidents
l l 1. M1 = covering fuel debris with solid materials
- 2. M2 = installing a water spray system above the pool
- 3. M3 = installing a building ventilation gas treatment system to reduce the airborne concentration of radionuclides prior to their release.
l l Two representative accident sequences were postulated. The first (A1) consisted of a complete loss of pool water inven-I tory, followed by a zircaloy fire, representing an upper bound in terms of radiological release. The second (A2) consisted of a complete loss of pool water inventory, followed only by cladding failures (i.e., no zircaloy fire). This represented a best estimate in terms of radiological release.
The estimation and evaluation of values and impacts (Alternative 3) considered the six possible pairings of accident and mitigation scenarios (i.e., Al/MI, A1/M2, A1/M3 [ dismissed since M3 could not cope with Al], A2/M1, A2/M2 (judged to be the same as Al/M2] and A2/M3). These reduced to four cases, for which a crude value-impact assessment was NUREG/BR-0184 C.32
. ~. - - _ - _ . - _ ,- - - . - __ - - - ,- -
4 Appendix C 3
\
l performed, similar to what was termed a "first approximation" in Chapter 2 of the 1983 Handbook. Offsite consequences were estimated using M ACCS for both a worst case (high population density and worst source term) and an average case (average population density and average source term). Costs (in 1983 dollars) were generated by assuming a Category I storage tank of 200,000-gal capacity and a complete spray system would need to be installed. The calculation results for each of the four cases are presented in Table C.101. {
i The presentation of results (Alternative 3) consisted of the value-impact summary (Table C.102), which indicated that installation of pool sprays was not cost effective, based on the best-estimate measures provided in the table [ net benefit, ratio, ratio of benefits (in dollars) to cost, and cost of implementation per averted person-rem]. The decision rationale (Alternative 3) was the same as that for the other alternatives, namely not to recommend the alternative based on the value-impact results. However, the possibility ofimplementing Alternative 3 on a plant-by-plant basis was mentioned,
, since the high-estimate measures indicated marginal cost effectiveness. At plants where the conservative assumptions used
} in NUREG/CR-5281 might be approached, Alternative 3 might warrant implementation.
i C.11 Nuclear Fuel Cycle Facility Accident Analysis Handbook (NUREG-1320)
In NUREG-1320, Ayer et al. (1988) provided methods to determine the release of radioactive material to the atmosphere
- and within a plant resulting from potential accidents at the following types of nuclear fuel cycle facilities: fuel fabrication, i fuel reprocessing, high-level waste storage / solidification, and spent fuel storage. Six types of accidents were addressed:
fires, explosions, spills, tornadoes, criticalities, and equipment failures. These were chosen as being the major contribu-tors to the radiological accident risk from the operations of fuel cycle facilities. While NUREG-1320 provided methods for calculating consequences from these accidents, it did not provide methods for determining the accident probabilities.
Ayer et al. assembled accident descriptors for both the facilities and their processes. For simplicity, a representative facility was developed containing common descriptors from each of the four types. These descriptors are shown in Thble C.103. For each type of fuel cycle facility, Ayer et al. assembled process accident descriptors, listed in Tables C.104-C.107. These descriptors were based on the following process parameters:
- quantity, chemical, and physical form of radionuclides e quantity and characteristics of flammable and combustible materials
- radionuclide content of materials with high fissile material content 1
- characteristics of process equipment providing airborne containment or confinement
- others that could enhance or mitigate airborne release (e.g., pressurized syste'ns).
Source terms for each of the six types of accidents were discussed. Behavioral mechanisms for airborne particles were summarized, as shown in Table C.108. Following these were the detailed descriptions of the calculational methods for estimating the source terms from each type of accident. Both hand and computer calculations were presented. All necessary reference tables and figures for conducting a " standard" analysis were provided, along with additional references for " specialized" assessments.
To illustrate the use of the analytic procedures, Ayer et al. identihed four " primary" and seven " secondary" sample prob-lems, as follows:
s C.33 NUREG/BR-0184
Appendix C Primary:
O
- 1. Slug Press Fire (MOX Fuel Manufacturing)
- 2. Solvent Extraction Fire (Fuel P.cprocessia)
- 3. Glove Box Explosion
- 4. Pcwder Spill During Tornado Secondary:
- 5. Flashing Spray (Fuel Reprocessing)
- 6. Pressurized Release of Powder
- 7. Radioactive Powder Spill
- 8. Liquid Spill of Plutonium Nitrate
- 9. Aerodynamic Entrainment of Powders from Thick Beds During Tbrnado
- 10. Fragmentation of Brittle Solids by Crush Impact During Tornado
- 11. Inadvertent Criticality in a Fuel Reprocessing System For each, Ayer et al. conducted a sample source term calculation, showmg use of both hand calculations and computer tools. The main computer codes were as follows:
- 1. TORAC - for analysis of tornado-induced gas dynamics and material transpon (Andrae et al.1985)
- 2. EXPAC - for analysis of explosion-induced gas dynamics and material transport (Nichols and Gregory 1988)
- 3. FIRAC - for analysis of fire-induced gas dynamics, thermal, and material transport (Nichols and Gregory 1986)
Although designed mainly for analysis of the ventilation system (the primary airborne release pathww), these codes can be used for other airflow pathways as well. The codes, especially TORAC, can be extended to model accidents associated with criticality, spills, and equipment failure. Limitations involve the gas dynamics models, which are based strictly on lumped-parameter formulations, and the material transport capability, which is very basic and relies on information found in the literature.
For each of the primary sarrple problems, the authors of NUREG-1320 carried through a complete radioactive airborne release calculation. The results were presented through a series of tables and figures, too numerous to reproduce here.
C.12 Endnotes for Appendix C
- 1. The 1990 BEIR V report updated the radiation exposure coefficient to SE-4 fatal cancer / person-rem, or inversely 2,000 person-rem / fatal cancer (National Research Council 1990).
- 2. For consistency when using Tables C.42-C.47, or values derived from them, the analyst should employ 5,000 person-rem / health effect, the conversion factor assumed by Daling et al. (1990), from whom these tables have been extracted.
However, the analyst should be aware that BEIR V updated the radiation exposure coefficient to SE-4 fatal cancer /
person-rem, or inversely 2,000 person-rem / fatal cancer (National Research Council 1990).
> 3. Rec:nt experience at the DOE Savannah River Site suggests frequencies of glove failure as much as 10 times higher.
O NUREG/BR-0184 C.34
Appendix C O l
! l
- 4. Recent experience at the DOE Savannah River site suggests frequencies of dissolver seal failure as much as 1,000 times higher. j
- 5. Recent experience at the DOE Savannah River Site suggests frequencies of fire in low level waste and fuel assembly drop as much as 100 times higher.
- 6. The iodine-129 part of Table C.81 is suspect.1-129 has a half-life of 17 million years and, correspondingly, specific activity of 1.8E-4 Ci/g.1-129 emits a 150 key beta and,9% of the time, a 40 key gamma, both significantly lower !
energies than the corresponding values for 1-131. The biological half-life of I-129 in the thyroid is 120 days. The !
dose conversion factor for 1-129 would be approximately 0.5 rem / micro-Ci administered to the thyroid. The values i given in the table for I-129 releases and the corresponding thyroid doses seem inconsistent with each other and with l the properties of I-129 given above. The thytoid is relatively radio-resistant and thyreid cancer relatively treatable; the mortality risk factor for the thyroid is 5.0E-6/ person-rem (i.e., one fatality per 2.0E+5 person-rem exposure to the thyroid).
l l
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G C.35 NUREG/BR-0184
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Figure C.1 Uranium process flow among fuel cycle facilities O _ _ _ _ _ _ _ - . _ ___.-____.-_____---____.__T--
O _ _ - - - - _ - _ _ _ _ _ . _ _ _ . _ m___ - - . _ _ . - - - - _ _ .-
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f C.37 NUREG/BR-0184 i
a
Appendix C O
10
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O NUREG/BR-0184 C.38
_ . _ _ _ _ _ __ __4 . - _ . .- . . - - _ _ . . . _ . . - _- - _. _ .. _. .. .. .
a 1
Table C.S.1 Summary description of representative urar,ium fuel cycle facilities (Schneider et al.1982, Table 2,2) .
Fuel Cycle ilament Conversion Er ttement Easeous Eas Fuel Mining Milling Agueous Dry . Blffusion Centrifuge Fabrication [
lace (Sectiony (section 10) (Section ll.Al (Section 11.8) (Sectlen 12.A) (section 12.5) (Section IJ) f acility Based on Ambros t a L d e Highland Sequoyah stetropells Stand-alone, com- Conceptual uestinghousef [
hination of 3 U5 stand-alone Celvable SC plants 5
Major Process Underground rous-and- Acid-leach, solvent Solvent estraction H ydr ofluorlaation, Gaseous diffusion, Gas centrif uge, ADLl process, calcl-piller, cutting, esta., precipitation hydrufluorlaatton, fbuerlaallon, cold trapping, cold trapping, co blasting fluorlaatton fractional dis- waste recovery waste recovery set ton lng,nqpec sinter maste tion, tillation recovery L apac It y f eed/ flyt yr Ore Velafvaries Oref6.6E5 Tellouc dell .2E 4 Telloucdef 7400 Well M4 WF 6 II Mi # tIII" produc t /Mylyrla) Oref t.3th felloucd ef930 UFn f9100 if6 fteC0 IF6fl400 tF 6f1400 Fuel asse=618est!460 3300 1600 15,4uo 11,500 15,500 15,500 16,000 GWJ, Equivalent;yr(b) e operating hrfe and styr 16f312 24t hS 24f365 241300 24f365 24/36b 24f350 letal Statt 1100 92 155 IWL I400 2150 1850 Lunt ac t og=r at ions -All; most is not -Alta ==st is not -All; omst is not -All; most is not -All salmtenance All malatenance neceivsag, red and direct contact direct contact direct contact direct contact element asMlage, malatemance j i
liceute Operattons None hone llone home Most operations 8 tost operatlant Chemical processing, scrap recovery (not shielding)
Alternative Concepth Open-pit, in-site Alkaline teach. Ilume liene W Laser, ur g u taser, IF6 Fluidland bed, ;
(5ol alon) len enchange Laser W plasma Laser, e plasma peuder front-end ten ten I
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oo k Table C.S.1 (Continued)
Iwel Cycle E lesent Weste Storage MOR f uel f uel Eeelegic hef abricat ion Reproc essing 5 pent f uel Shallow iand High-Level Weste IRU Waste itaste Disposal Maste Otsposal item ( Set t ien 84) (Settlen 15) (sec tion 16.A) fransportation (Section 16.8) (Section 16.C) (5ection IP) (Section 18) (Sect 6cn It)
F et t lit y 4.unteptual West- gernwell eith tun- Lonceptual, stand- Conceptual, stand- Cenceptual, stand- Conceptwal NW15 Loateptual stand- S t a t e-et - the - ar t ;
inghouse Secycle ceptual additions alone, mater basta alone, dry-well alone, vault and feels Plent disposal rePost- alone specific to each outside pad tory le salt material formation Me sw t' rue ess Powder blenotsy, PUNE R, UF and Wet untweding and Wet unloadtag, Solids handling computton, sin- Pe converston, Soltes handling, Surial in below- truct and rail storage, ton entapsu)ation, dry- (shleided and undergreend grade trenthes ter lag, waste tilu vitrification esthange, heat well storage unshielded), above transport tress-retever y blasting, machine country estthange grade storage excavat ion L epea 6 t y f eed/M9f yr (Kly ; Pnej f ah; 18 spent futillS00 spent fuell5UU HM Solldtfled HtW1320 lau-westel50,000 5 pent fuel, Mlu ttu, itW/50,000 m3 led 6vidual shipp6ng TRU wastef3900 MM capacity / container equiv.
for eash material g Fr edat t tMglyel e) MDI asseabiles/ 011410; Pull) map hap rap rap 400 tot rap bud, t quival=1st f ye 4400 15,500 S500 15,500 21,600 4 3,0u0 29,000 --
tipee et seg 6.rtd end dtyr 4413S0 2.sJoo 24f 36S 24thb 20/300 241365 51250 varles I lonel 5telt 260 SUD -50 -100 28 2b1 70 1-2lshtsment t w.t x t Upev et tuns -All; most is not Itetelving, name Rese tving, Retelving, All En-Ir1 petelving, -All; most is not direct centatt maintenanc e maletenance maintenance Strect tuntact
-l/2 Att-IRU -All CH-IAU direct contact with conta6ners
! -l/2 an-tRU Nteute Uper at tons Pellet prepara- Moat eperations f uel unloading and Most operations -112 set-IRU -112 pH-Iku Done Weepte unleading L ion, ur ap handtlag, waste-recovery treatment
-All spent fuel, for most materials litu A licenat eve 4 ann ept s t o- prec lp it at ton, M parlations of Dry well, cask. Dry well, cash, Belous-grade, aime Sasalt, salte, Onsite processtag. Tarlat tuns of hard-l remote maintenante P I, Others tunnel rath, vault tunnel ratt, vaelt storage, beres consolidat ten tuff; se - verlous burial were for most shielded pathages earlations cont ainers l - --- --
hA - not eveilebte rap . not applicable (e) As 18 and/wr Po estept fevas men 6ng, (h) eased on produs t rate to f uel f ebr stat eun end 11,t400 mwd elMWIM.
9 9 9
d Appendix C hble C.1 Frequency of contamination incidents for non-reactor nuclear material licensees (Ostmeyer and
) Skinner 1987, Able 3.1)
- i l
i Number of Number of Frequency (incidents !
Application /use class incidents
- Licenses licensed-activity-yr)
Research/ teaching & 7 5100 0.00023 I)
. Diagnostic / therapeutic II) Measurement / calibration 6 5715 0.00018 )
I & irradiation III) Manufacture / distribution 8 510 0.0026 IV) Service organizations / 0 49 -
waste processing / storage Source and Special Nuclear 6 72 0.014 V)
Material Fuel cycle r
(a) For a six year reporting penod.
, i hble C.2 Incident cleanup cost by material quantity class for non-reactor nuclear material licensees (Ostmeyer and Skinner 1987, hble 4.1) a Licensed Incident Cleanup Cost ($)
Material Quantity LQR Case Average 5
10 mci - 0.1 Ci 70,000 15,000 0.1 Ci- 1.0 Ci 200,000 75,000 )
1.0 Ci - 10 Ci 450,000 230,000 10 Ci - 100 Ci 800,000 500,000 100 Ci - 1000 Ci 1,500,000 900,000 ll G
C.41 NUREG/BR-0184 .
1
Appendix C O
Table C.3 Economic risk as a function of material application /use and licensed curie quantity for non-reactor nuclear material licensees (Ostmeyer and Skinner 1987, Thble 5.1)
Economic Risk ($/ licensed activitv/vr) by Licensed Ouantity'd Application /Use Class 0.01 Cl- 0.1 Ci- 1.0 Ci- 10 Cl- 100 Cl-0.1 Cl 1.0 Cl 10 Cl 100 Cl 1000 Ci I) Research/ Teaching / 4 29 50 120 200 Experimentation and Diagnostic / Therapeutic II) Measurement / Calibration 3 20 40 90 160 Irradiation l III) Manufacture / Distribution 40 230 520 1,300 2.300 (a) Risk is given by the product of incident frequency and average incident cost.
O NUREG/BR-0184 C.42 O
l l
Appendix C V i l
Dble C.4 Sumsaary of economic risk at a reference uranium mill (Philbin et al.1990, hble 4.1)
Frequency Economio Risk l Consequence Cleanup Cast per year (per year) fnetdant Scanarin haerierf an funearcainevi f unea rtai nevi funeartainevi ,
Minor facility Hundreds of g to tena $1100 0.0077 $4 j releases of kg U released. ($900 81,400] [0.0048 0.014] ($5 . $15] .
Confined to small arena in plant.
Selvent Entraction Up to several kg U $370,000 0.0031 $1100 l Fire released. Cleanup [$300,000.$460,000] [0.0014-f,.0082] [$460 $2900) I limited to process )
area. I l
Fire /taplesien in Up to moveral Kg U $500,000 0.0031 $1600 Yellecake Dryer released. Cleanup ($400,000 8630,000] [0.0014 0.0082) ($620 $3900]
limited to process area.
l Major facility Fire Cleanup of main $1.5M 0.00020 $300 ]
process area and ($1.2M 41.fM] [0.00013-0.00040] ($160 $550]
downwind facility area (22.5* sector).
Rotention Pond 8 s 108 the solida $2.5M 0.023 $58,000 Failure with Slurry released. Stabilise {$2M.$3.1M] [0.017 0.033) [$39,000-886,000) .j Release pond and spill areas i and clean up spill.
Sturry selease from 2.2 x 105 the s.11de $69,000 0.0062 $430
\j Distribution pipe released on site. [$55,000.$56,000) (0.0037 0.012] [$230 $400) stabilise spiti eres.
t
\ / Clean up spill area.
Tornade Thousands of kg U $3M 0.000080 $240 released . Clean up [$2.4M $3.8M) [0.000025 0.00025] ($70 8780]
buildings and downwind site area (45' sector).
Transportation Entire lead of ere spilled er 1/3 $300,000 0.0031 $t30 yellowceke drums ($225,000 8375,000) [0.0014 0.0082) ($370 82300) spill. Area cleanup TOTAL TACILITT $63.000 ECONOMIC RISK [$43.000 891,000]
i 1
, I I
J 4
(
C.43 NUREG/BR-0184
Appendix C O
%ble C.5 Summary of economic risk at a reference uranium hexafluoride conversion plant (Philbin et al.1990, hble 4.2)
Frequency Economic Risk Consequence Cloenup Cost per year (per year)
M ant S c e na r ia beacrietten f unca rt a i nt v l funcertafervl funcertainev{
Minor facility Release of buridreds $1,100 0.13 $160 release of grane to tens of [$900 81,400] [0.081 0.22] [$80-$250) kg U. Cleanup limited to immediets ares of the release.
Uranyl Mitrate Release of severst $730,000 0.00032 $230 Evaporater (spiesten kg of U. Cleanup of [$580.000 $910,000] [0.00010 0.0010] [$70-8750]
preessa butiding.
Hydrogen empleston Release of several $730,000 0.0070 $5,100 during raduction kg of U. Cleanup of [$580,000 8910,000) 10.0010-0.050] [$710 837,000]
process area, solvent entraction several hundred kg U ,$81,000 0.00040 $30 fire released Clean up [$65,000.$100,000) [0.00013 0.0013) [$10 8100) solvent entraction butiding.
Release from UFe Release of up to $1.2M 0.021 $25,000 cyttnder 2500 kg of U. Clean ( 40. 96th 81. 5H ) [0.011 0.081] !$9,100 870,000]
up immediate area.
Distillation Valve Release of tena of kg $130,000 0.050 $6.500 tupture of U. Clean up [$100,000 $160,000) [0.016 0.16] [$2,000 $21,000]
lamediate area.
Waata pond Release 7 s 108 lba sellde $230,000 0.056 $13,000 reisesed. Stabilire [$180,000 $290,000) [0.029 0.22) l$4,600 $36,000l pond and epill area and clean up spill.
Traneportation Small rupture of ufo $600.000 0.0031 $1,200 cylinder. Hundred [$320,000 8500,000] [0.0014-0.0082) [$500 83,100) of kg of U ratessed.
Cleanup of eres.
Tornede Thousands of kg U $1.9M 0.0023 $6.600 dispersed. Cleanup l$1.5M $2.6M) l0.00076 0.0076) {$1,400 814.000) of 65' sector of downwind alte area.
l TOTAL FACILITY $56.000 '
ECONOMIC R15K [$20,000 8109,000) j j
l l
1 NUREG/BR-0184 C,44 9
Appendix C l O
%ble C.6 Summary of econontic risk at a reference uranium fuel fabrication facility (Philbin et al.1990, hble4.3) i troMue.cy scene.t. stok c.na. u.nc. cl. ,c.et p.r y..r (per yo.o Incident scenarie h- ariatian funcertaintv1 Juntartaingv1 fummartaintel Minor Fact!!ry Release of hundrede of $3,500 0.21 $740 Release saa to tens of bg U. [$2.800 . $4,400) l0.15 0.32) ($470 $1,100)
Confined to small areas in plant, Lar58 $ Pills due 800e3 waste eslution, $1.ON .
0.024 $24,000 to accidents or 24 Ci solids, 40000 s8 l$0.80M $1.3M) l0.015 0.044] ($13,000 $43,000]
natural phenomena surface contaminated.
Trenapertation Trailer overturns $10,000 0.0028 $20 accident We centamination l$7.500 13,000] l0.0026 - 0.0030] ($22 $35) outside trailer.
Empleaten Retary K!!n. Satch of $3.9M 0.01 $39,000 100 kg U, Ikg rolessed ($3.1N * $4.9M] l0.002 0.05] [$7.700 $200,000) to environment (outside), 1/3 of main building contastnated.
Major Fire Decentastnation of 11M 0.00021 $2,300 entire sein butiding l$8.8N . $14M) [0.00012 0.00051] l$1,100 64,900) is required.
Criticality 10 e itselene; 8 hr $3.9N 0.0033 $13,000 duration. 1/3 el main l$2.9N . $4.9M] [0.00050 0.011) [$2,700 861,000]
building contaminated.
(N Major UFe Release Furture of one er two eylindere. Thousande
$1.2M
[$0.96M . $1.5M) [0.011 0.021 0.C21]
$25,000
[$9,100-870,000) of kg of U released.
Major site !
conteelnation. 6 l ecres. Offsite ca..nup to not ]
empacted.
TOTAL FACILITY $104.000 ECONOMIC R15K l$43.000 $250,0001 l
l O
C.45 NUREG/BR-0184
Appendix C O
hble C.7 Summary of economic risk at a reference byproduct material manufacture / distribution facility (Philbin et al.1990, hble 4.4)
Frequency Eeeneste Risk Consequence Cleanup Cost per year (per year)
MQg benettetten funearta!nevt fa- artainevt f ~ =etaineet Minor Fsettity small decontaminatten $6500 0.0022 $14 pelesaae incident limited to the ($5,200 . $8,100] [0.0015 0.0033) [$9 . $22) immediate area of the release.
Iodine 12b Spill Millleurie spill of $30,000 0.0022 $66 Outside a Filtered Nal 125 en unfiltered [$24,000 . $38,000) [0.0015 0.0033) [$42 $100l Enclosure area of laboratory.
Laboratory decentaminatten required. No offsite cleanup required, tire in a Tune small fire involving ,$44,000 0.00059 $26 Heed melybdenum 99 genere- l$35,000 $55,000) (0.00034 0.0013]
tore in fume hood.
[$13 $53]
Laboratory decentamina-tien required. Ne off-site cleanup coquired.
Major Fine in Fire in todine 125 $290,000 0.00059 $170 an ledine process laboratory. [$230,000.$360,000) (0.00034 0.0013) {$84 . $350]
laboratory Four curies velatilized and dispersed into two laboratories. 0.4 curies released to environment.
Waste Warehouse Single vaste drum fire. $300.000 0.0081 $2.400 Fire (single Severst stilleurtea {$240,000 8380,000} (0.0074 0.0088) ($1,900 . $3,100) drum) volet111 sed. Entire varehouse decontaatne-tien required, Weste Warehouse lot of waste inventory $1.1M 0.0081 $8,900 Fire (multiple released in fire. [$0.9M $1.4M) (0.0074 0.00881 l$7,000 $11,000]
drums) Offette decentastnation required.
Tornade Ss11 din 5 200 er 250 $2M 0.000010 $60 seeerely dassged or ($1.6M $2.$M) (0.000009 0.00009] [$19 $190)
Bit *g. 32 destroyed. It of in process material reliessed. 754 of waste inventory released.
Earthquake several bu11dinge $1.3M 0.0040 $5,200 severely damaged, it {$1.0M $1.6M) [0.0010 0.020l ($1,100 $24,000) of in. process setorial released.
total TACILITY $17,000 ECONOMIC RISK [$8,600 . $31,000)
C.46
Appendix C m
a hble C8 Summary of economic risk at a reference waste warehouse (Philbin et al.1990, Dble 4.5)
Frequency Economit .a sk Consequence Cleanup Cost per year (per year)
Incident t e e r.a r i a Descrintion funcertainevi funcartafntM f unc e rt aint vl Minor facility Failure of one SLSV $4000 0.0041 $16 Releases waste drum. Emcel [$3,200 . $5,000) l0.6022 0.016) ($6 . $45) decentamination.
Waste Compacter Fire involving one $62,000 . 0.0081 $500 Fire drum of DAW waste. [$50,000 878,000) [0.0074 0.0088) l$400 . $640)
Local area decantamination.
Waste Drum Fire Fire consumes one SLSV $410.000 0.0081 $3,300 (single drum) waste drum. Entire ($330,000 8510.000) [0.0074-0.0088) ($2,600 - $4,200) warehouse decontamination required. No offsite cleanup required.
Transportation liighway accident $40,000 0.0011 $44 Accident (without fire 0.2 {$32,000 . $50.000) {0.00035 0.0035) ($14 - $140]
curies released, with fire .. I curte released) into two $53,000 0.00024 $13 taboratories. 0.4 {$42,000 . $66,000) (0.000076 0.00076) [$4 441) curies released to environmwnt.
O Facility Fire Fire eeneuses ten per.
cent of redlelegical
$1.2M l$0.9 M (0.0074 0.0081 $9,700
$1.5M) 0.0088] [$7.700 812.000) k' inventory. Offette
\ decentaatnatten required.
Tornado building destroyed. $1.5M 0.00020 $300 levanty ftve percent {$1.2M . $1.9M1 [0.00006 0.0006) ($93 $970) of weste inventory rol.a..d.
TOTAL FACILITY $14.000 Ec0MOMic Risk [$t1.000 816.000)
BLSV = bulk liquids and scintillation vials DAW = dry radioactive waste hble C9 Estimated 70-year population and worker exposures for repository construction (Daling et al.1990, hble 4.2)
Maximum 80 km iderker Indtytaual Population Geologic Exposures Esposures Exposures Mediur inersof* ram) fras) f oerson-read salt 1.8E 1 2.8E 8 6.8E 3 Grantte 5.0Ee3 4.lE 4 1.CE+2 Basalt 6.!!+3 5.9E-5 1.5 Eel shale 1.9E+3 1.5E-4 3.8E+1 1
C.47 NUREG/BR-0184
I Appendix C Ol ,
J hble C10 Radiation exposure from normal construction and operation for repository preclosure period j (Daling et al.1990, hble 4.13) i Estimated 50 yr Excesure Cateocrv Dose Cemitment Construction Maximally Exposed Individual Annual 0.044 mrem yr 0.42 mres l 80-km Population 50 yr 2.0E+4 person-aren j
Operation Maximally Exposed Individual
-Annual 0.17 mrem 50 yr 5.6 mres 80-km Population 50-yr 3.9E+5 man-area hble Cll %tal radiological worker fatalities from construction and emplacement periods of three alternative Repository Sites (Daling et al.1990, hble 4.20)
Radioloaical Fatalities (a)
Wasta Geologic underground Underground Handling Medium Construction Ocerations Onerations Total Salt 1.4E-2 4.4E 2 1.5E00 1.6E00 Tuff 7.7E 1 4.0E00 1.0E00 5.8E00 Basalt 1.5E00 5.4E00 1.9E00 8.9E00 (a) Based on 5 year construction and 26-year emplacement operations period.
%ble C12 Occupational dose during normal operation and from a shaft drop accident for repository preclosure period (Daling et al.1990, %ble 4.5)
Number of Persons Average Annual Total Dose scenario Irvolved Dese free /vri foersen ree/vr1 Reference Case
- Normal Operation 1.000 0.9 902
- Accident 300 1.5 454 Case 1
- hormal Operation 1.068 1.2 1.295 Accident 352 1.6 569 l 1
Case 2 l
- Norwal Operation 1.045 1.1 1.188 l
- Accident 332 1.6 532 Case 3
- hormal Operation 1.985 1.2 2.301
- Accident 603 1.6 978 91 NUREG/BR-0184 C.48
. - - . - -_ _. ----- -.~~-.-.._-.-.-.-- ..-- .__-. --
i J
i 4
Appendix C 1
2 J
l hble C.13 Public dose during normal operation and from a shaft drop accident for repository preclosure period (Dallag et al.1990, Dble 4.6) l Whole body Dose
- Public Dose 1
Scenario foerson-res/ vel Refererce Case 1 j
- homal Operation 2.5E-5 l
- Accident 6.5E 2 j
'1 Case 1
\
a - hermal Operation 5.0E 6 2 - Accident 5.6E 2
=
Case 2
. hormal Operation 7.7E 6 4
- Accident 5.6E 2 j
i I
Case 3
- Normal Operation 1.1E 5 Accident 5.6E 2 4 j i ;
i l
1
. Case 1. Simple encapsulation and disposal of spent fuel af ter !
storage at an away from reactor storage facility (AFR) for 9 years. '
. Case 2. Encapsulation of fuel, end fittings, and secondary wastes '
after chopping the fuel bundle and removal of volatile materials.
.i Case .. Incapsulation of fuel, end fittings, and secondary wastes '
after cho *ing, removal of volatile materials, calcination, and I vitrificatib.'
} r
> l i 1
.i e
f I
a 4
8 W
s 1
)
l C.49 NUREG/BR-0184
Appendix C l
O l hble C.14 Summary of repository accident releases, frequencies, consequences, and risk values for repository preclosure period, operations phase (Daling et al.1990, %ble 4.11)
Accident Release Frecuency Consecuences(a) Risk Value Descrittica Duaatity IC1) feer vr) foerson-ene) foernon-ree/vr)
Fuel truck H-3; 3 2.0E-6 2.0E+3 4.0E 3 crash into Cs 134; 300 HLW area Cs 137; 70 Fuel truck FPID);400 2.0E-6 2.0E00 4.0E-6 crash into Actinides 0.1 cladding waste area l Fuel truck Actinides; 100 2.0E 6 4 CE+1 8.0E 5 i crash into i NHLW area Aircrash into H-3; 3 1.0E-7 4.0E*3 4.0E 4 receiving Cs 134; 200 area Cs 137; 70 FP; 400 l Actinides; 200 Elevator drop H 3; 4E-3 4.0E-8 5.0E-2 2.0E 9 FP; lE 2 Actinides; 4E-3 ,
hon-HLW Actinides; C.02 5.0E-2 8.0E-1 4.0E-4 pallet drop I Final filter Actinides; 0.2 3.0E-3 2.0E00 6.0E 3 failure ,
Total Preclosure Risk 1.0E 2 (a) Population doses are 50-year whole-body dose commitments.
(b) FP = Various fisston products.
Dble C.15 Radiation exposure from accidents for repository preclosure period, operations phase I (Daling et al.1990, %ble 4.14) )
Population 50-yr Maximally Exposed Dose Commitment seeicent ledivicual t wer) (eersea-mere)
Scent Fuel Drop 4.68E+1 2.99E+3 Comercial HLW Drop 2.74E00 1.75[+2 Spent Fuel Handlin9 3.98E 2 1.29E+3 kemete TRU Drop 3.10E 3 1.98! l Contact TRU Puncture 2.07E-9 6.70E 5 TRU = transuranic HLW = high level waste NHLW = non-HLW 1
l l
1 O NUREG/BR-0184 C.50
\
l l
Appendix C -
i hble C16 Occupational dose during repository operation (Daling et al.1990, 'Ihble 4.15) humber of Collective Dose activitv Workers fPertne-ree/ vel j Receiving 35 44.8 ;
Handling and Packaging 16 6.9 l Surface Storage to )
Emplacement Mor12on 14 6.0 J Emplacement :
Vertical 18 12.4 Horizontal 7 8.7 ,
l hble C17 Summary of annual occupational exposures for spent fuel and HLW operation at a tuff repository l (Daling et al.1990, Dble 4.16) ,
f Total Number Total Annual Dose Gneration of Workers faertop rea/ vel Receiving 35 44.6 !
Handling and Packaging 22 12.3 Transfer to Underground Facilities Shaft Access g 3,35 Ramp Access 7 2.68 i Emplacement in Boreholes Vertical 18 12.4 Horizontal 7 9,$g Retrieval from Boreholes vertical 22 12.6 Horizontal 6 8.86 l Return to Surface (Ramp) 5 2.68 Handling. Packaging. Shipping 17 g Totals (a) l Shaft Access / Vert. Emp1. 72.68 Shaft Access /Horit. Emp1. 69.84 Ramp Access / Vert. Emp1. 71.98 Ramp Access /horig, Emp1. 69.17 (a) Totals do not include retrieval and loadout operations.
I j
C.51 NUREG/BR-0184 1
_l
Appendix C O
DLie C.18 Estimated 50-year whole-body dose commitment to the public, maximally exposed individual workers from accidents for repository preclosure period, operations phase (Daling et al.1990, Dble 4.17) maximally Exposed 80 km Popu-Individual lation Dose Worker Accideat Scenarie Date free) foerton-ree) feeetoa-reet Natural Phenomena Flood 2$8E!! 1.2E 9 5.0E-10 Earthquake 2.4E 4 3.1E-3 0.37 Tornado 2.4E 4 3.lf 3 0.37 Man-made Events Aircraft Impact 6.8E-2 110 5.5 Nuclear Test 2.4I 4 3.1E-3 0.37 Dperational Accidents Fuel Assembly Drop 5.3E-6 8.0E 5 8.1E 3 Loading Dock Fire
$ pent Fuel 2.!E 2 6.8E-3 8.9E-3 3.5I ")
Commerc8a1 HLW 3.6E 3 9.2E-4 1.5E 3 - 0.6(a)
Waste Handling Ramp Fire 1.8E 7 3.6E-7 3.8E 64 IbI Emplacement Drift fire 1.8E-7 3.6E-1 3.8E-8 180(b)
(a) The first value represents the estimated dose to workers at the site surface and subsurface facilities; the second value is for the worker exposures at the loading dock.
(b) The first value is for the doses to worsers in the surface facilitiest the second value is for underground waste emplacement workers.
I I
I NUREG/BR-0184 C.52 O.
l
Appendix C Able C.19 Preliminary risk estimates for postulated accidents at a repository in tuff for operations phase (Daling et al.1990, Dble 4.18)
Estimated 50 yr Dose Frequency Commitment Population Risk Accident iconarie fevents/vr1 (nerson-real faerten ram / vel Natural Phenomena Flood 1.0E-2 1.2E 9 1.2E-11 Earthquake <1.3t-3 3.1I 3 <4.0E-6 Tornado <9.lf-11 3.!E-3 <2.8E-13 Man-made Events Aircraft impact <2.0E 10 1.lt+2 <2.2E-B Nuclear Test <!.0E-3 3.1E 3 <3.!E-6 Operational accidents Fuel Assembly Drop 1.0E 1 8.0E 5 8.0E 6 Loading Dock Fire Spent fuel <1.0f-7 6.8f 3 (6.8t 10 Commercial HLW <1.0I-7 9.2I 4 (9.2I 11 Waste Handling Ramp Fire <1.0( 7 4.8E 7 (4.8E 14 Emplacement Drift i Fire <1.0I-7 4.8E 7 <4.8I 14 Total 1.5f 5 k
1 l
l i
l C.53 NUREG/BR-0184
Appendix C O
Able C20 Frequencies and consequences of accident scenarios projected to result in offsite doses greater than i 0.05 rem for repository preclosure period, operations phase (Daling et al.1990, Dble 4.23)
Frequency. Conseguence Accident teenarie Deterietion mm war Internall y initiated fvents Crane drops shipping cask, cask breached 5E-6 340 Crane drops fuel assembly in hot cell. IE-8 170 HVAC fails Crane drops open consolidated fuel container. IE 9 1100 HYAC fails Container dropped in storage vault, filtration 3E 8 230 system fails to activate friernally initiated Events fall caused by earthaunkel Crane fails, falls on or drops cask in 5E-8 340 receiving area Train falls on cask 5E-8 290 Structural object falls on fuel in cask SE-7 110 unloading cell Crane fails, falls on or drops fuel in IE-6 110 cask unloading cell Structural ceject falls on fuel in SE 7 110 consolidation cell Crane fails, falls on or drops fuel in IE 6 110 consolidation cell Structural object falls on fuel in 5E-7 330 packaging cell Crane fails, falls on or drops fuel in 1[ 6 1100 packaging cell. HVAC fails Structural object falls on fuel in 5E-7 200 transfer tunnel HVAC = heating, ventilation, air conditioning O
NUREG/BR-0184 C.54
Appendix C
< p i
(
- k a
hble C21 Occupationt.1 dose during rormal operation and from accidents during decommissioning and retrieval phases of a repository (Daling et al.1990, %ble 4.7) 1 4
traapia Aranal haea faarnan-ren/ve) }
Deramtsutanian RetrievalL87 Reference Case
. hermal Operation 6 163
- Accident 5 89
' Case 1
. hermal Operation 23 588 Accident 16 254 +
1 Case t
- hormal Operation 22 487
- Accident 15 215 ,
I Case 3
- Normal Operation 40 1,116
- Accident 28 491 a
4 (a) Represents sum of doses from maste removal, offgas recovery and release, and sining and drilling activities.
. Case 1. Simple encapsulation and disposal of spent fuel after storage at an away from reactor storage facility (AFR) for 9 years.
d
. Case 2. Encapsulation of fuel, end fittings, and econdary wastes after chopping the fuel bundle and removal of volatile materials. c
- . Case 3. [ncapsulation of fuel, end fittings, a a secondary wastes af ter chopping, removal of volatile materials, calcination, and vitrification.
N j 4
hble C22 Comparison of normalized public accident risk values from various studies for repository -
preclosure period (Daling et al.1990, %ble 4.27)
Risk '
Document f aernan. ram /MTU) fement GEIS 8.4E 9 One accident Bechtel (1979) 1.1E-10 One accident Waite et al. (1986) 1.7[.8 Five accidents Jackson et al. (1984) 5,7E 9 Ten accidents Eromann et al (1979) 1.8t-6 Seven accidents Pepping et al, (1981) 6.3[ 10 One accident F
C.55 NUREG/BR-0184 l
1
Appendix C Thble C.23 1985 Revised EPA estimates of 10,000-year health effects fos 100,00041TIIM repositories in basalt, O
bedded salt, tuff, and granite (Daling et al.1990, Table 4.29) teenacio Basalt Bedded salt Igff CraMte Undisturbed 97 0 0 184 Drilling (misses 2.30 3.16 0 0.92 canister)
Drilling (hits 1,73 3.41 0.44 0.44 canister)
Faulting h L g y Total health Effects 125 6.57 3.44 194 (a) Palo Duro Basin Table C.24 70-year cumulative maximally exposed individual and regional population doses for the two peak dose periods for a tuff repository (Daling et al.1990, Thble 4.35)
Accumulated Accumulated Dose at the Dose at the Drean 27.000-bear Peak 250.000-Year Peak Total Body 0.2 0.2 Bone 0.6 3.0 Thyroid 2.0 2.0 Gastro intestinal 4.0 2.0 Lifetima Peculatice Deses from tne Drientne watee 5cera*1e for Two f utuee Times feersea-rem)
Accumulated Accumulated Dese at Dose at Orca
- 2L000..itEl 250.000 Years Total Body 2.0 200 Bone 4.0 4,000 Thyroid 600 600 Gastro-intestinal 200 400 NUREG/BR-0184 C.56 0
N Appendix C 4
4 4
hble C25 Peak conditional cancer risks due to ingestion for the 100,000-year postclosure period for a j 90,000-MTU spent fuel repository in bedded salt (Daling et al.1990, Able 4.38) i j Zone 1: Area from Zone 2: Area Repository to River Bounded by a 40 km Scenario (Number) 40 km Away, Plus 6 km Stretch of River and And Desertation Alena River 2 km Alona Rath Sides (1) Borehole (s) with 8.0E-2 8.0E-7 lower Aquifer
- Wells (2) U-Tube with Upper 2.0E-1 4.0E 6 Aquifer Wells (3) Dissolution 3.0E-1 7.0E 6 Cavity with Wells (4) torehole(s) 1.0E 6 1.0E 6
] (5) U-Tube 2.0E-6 1.0E 6 l (6) Sorehele(s) inter- 3.0E-6 2.0E-6 secting a Eanister
\
l 4
Dble C26 Radiation exposures from routine operations at the MRS facility (Daling et al.1990, Dble 4.42) k $0-Year Dose Eommitment from Annual Release Pathway and LocatNn Maximally Exposed Population in the Body Individual (real foerton-ren)
Total Body 2.4 x 10"* 2 x 10 I Bone 3.0 x 10-6 I x 10"I Lungs 2.4 x 10 2 x 10I Thyroid 1.3 x 10'3 1 x 102 a
1 9
O C.57 NUREG/BR-0184
Appendix C hble C27 Radiological impacts of potential MRS facility accidents for sealed storage cask at the Clinch River Site O
for operations phase (Daling et al.1990, Able 4.43) 50-Year Dose Commitment to the Public Location Maximally [xposed Population Accident in the body Individual frael gerson-res)
Fuel Assembly Drop Total Body Bone 4.4x10*f 1.4 x 10' 3 x 10]'
7x10 Lungs 4.6 x 10'3 3 x 10-2 Thyroid 2.9 x 10 2 2 x 10'3 Shipping Cask Drop Total Body 9.1x10j' 6 x 10'3 Bone 3.0 x 10 1 x 10'I Lungs 9.6 x 10'4 6 x 10'I Thyroid 6.0 x 10'3 3 n 10 2 Storage Cask Drop Total Body 6 .* 10'I Bone 8.9x10*f 1 x 10*II Lungs 2.9 x 10'"
9.3 5.9 xx 10*3 6 Thyroid 10- 3 xx 10*2 10-Table C28 Occupational dose from MRS facility operations (Daling et al.1990, %ble 4.44)
Unit Occupational Oceration foerton rea/1.000 MTU)
Receipt and Unicading 58 '
Lonsolidation 6 Loading Consolidated fuel Rods 9 Maintenance / Monitoring 2 Emplacement and Petrieval 29 Total 95 l
1 hble C.29 Summary of occupational doses from MRS facility operations (Daling et al.1990, hble 4.49) l Doeration foerson-ren/vr1 Receipt. Inspection, Unloading 148.0 Transfer to Storage Casks 6.2 Emplacement in Storage Area 7.2 Surveillance in Storage Area 5.3 Retrieval from Storage Area 7.1 Transfer to Prccess Cells 4.0
$hipment to Repository li2J j Total 318.7 '
NUREG/BR-0184 C.58 O
l
Appendix C
' N
\
O hble C.30 Occupational dose estimates for selected MRS operations (Daling et al.1990, Dble 4.50)
Occupational Dose i
, omaratten fearson-aram/1.00mITU) )
. Conselldate and package fuel 3.5 l
8 Conselldate and package 1.1 i non fuel components Receiving and unloading . Truck 135 Ratl 25 l
1 1
%ble C.31 Summary of MRS drywell risk analysis for operations phase (Daling et al.1990, Dbles 4.45 and 4.46)
Latent frequency telease Cancer h tatsaarv f atal:t tes A Transporter selitsten during empla:ement
- ne f tre 1.7E-8 !!! 3.4t 5 8.81 !!
. fire 6.1E 7 IV 1.9t 3 1.21 9 fransporter collisten during retrieval
(
. as ein fattures as fire 0.9t 3 !! 5.91 7 8.3t-t
. pie failures ne fire 3.6E 3 !!! 3.H l I.lf 6
. se pie fatters; fire 1.4t 4 IV ' 3.8t 4 3.6t 13
- pte failures fire 1.4t 4 IV 2.St 4 3.8t 8 e
rw Transporter setten etth seatster partially to place l
. emplacement 4.4E 3 y ..st . 1.st t g . retrievals ne ple fallere 4.9( 3 II 8.M 7 8.3E 9 i
. retrievali pte fatlure 1.4t 1 V 1.41 3 3.ft 4 tanteter drop emplacesset 3.Ff 0 8 3.M 4 4.8E 14 Cantster drop retrieval I.!( t i S.st 7 3.l!.s Plane crashi ne fire 4.St 10 V 2.6E l 1.0f 10 t Plane crashi fire F.4t 9 VI 1.3E+4 9.4t t tarthemates se pie fatture 4.SE-9 11 6.18-8 2.91 10 tarthgnhet pie fatture 4.31 8 !! 3.M+e 1.tL,!
I total 1.FE 3 4 !
neiease neiene free Assumed Dan.ees Per fract.
aedi ice Re. lease.of r av.ucii t ra u.,t Catamary tr.anarie Events suister in r..i nt v.ived j I filtered gas release $as teventory free Gases:I*I 3.0E -3
} (centster taeact to 185 pins released is 3.0E 4 s the Interface areas) through filters 11 Lietted gas release Gap inventary free $ases: 3.0E.3 (castster 1est) la stes (assumed te I: 5.01 4 esveles feats while to storage) released via lens and east channels fit Imlietted pas release toeplete gap 6ases: 3.01 2 (cantater tapact la teventory free 105 12 3.08 2 storage areas) plas it (levated toeparature cumplete inventory Gases: 1.Ot+0 release (toeparary of gases and I sne 1: 1.7E 1 less of coeltag) 11 of volettles (s. Au: 1.0E-4 released eis leets and estt channels V tapened fuel release lot of fuel esposed Gates: 3.0E 1 i (severe centster releastag gas is 5.OE 1 tapact) teventory, velatiles. (s. av: 1.0E 3 and particulates. particles: 1.lt 6 i tematacer releases I as laventory via l
feats and estt caannels vi tapesed heated fwel as is W. with Esses: 1.0t.4 release (severe tacreased releases ti 1.0E !
canister tapact ts, av: 5. lf-3 with fire) particles: 3.0E 4 (a) Eases tec1 wee C 14.16 3, and Kr-85.
C.59 NUREG/BR-0184
Appendix C O
Thble C.32 Summary of results of MRS operations phase (Daling et al.1990, Table 4.48)
I frapy hadier of teteese Conseipence elsk l Accident teensele f eventtM) 4aamdH Ies CattapfY flCP) flCF M I !
Fuel Assamtdy Drop During Loading 1E 1 1 1 4E 5 4E-6 Drop of Transport Cask During Loading Cast 4E 3 10 1 4E 4 2E 6 Drywett 7E 2 10 1 4E 4 3E-5 Venting of Cask Durire fransport !
Cask 2E 3 24 2 1E 1 2E-4 Drywell 3E 2 1 2 4E 3 1E 4 Collision Durfre fransport Cask 25 4 24 3 1E-1 2E-5 Drywell 2E-5 1 3 4E 3 8E 8 Collisten with Fire Durlig fransport Cask 2E 6 24 5 SE-1 1E-6 Dryntt 2E 7 1 5 2E 2 4E 9 Canister Drop During Esplacement Drywell 1E 6 1 3 4E 3 4E 9 Centster Sheer DurIng Esplacement Drywlt 2E-6 1 3 4E-3 8E-9 Cask Drop Durins Esplacement Cask 1E-5 24 3 1E 1 1E-6 fornado Missite Penetration Cask 6E 6 10 3 4E 2 2E 7 Drywell 1E-4 10 3 4E 2 4E 6 Plane Crash toppies Cask with Fire Cask 6E 9 24 5 5E 1 3E 9 Plane Crash Plus Fire Cask 9E 9 24 5 5E-1 4E-9 Drywell 2E 7 1 5 2E 2 4E 9 2E 8 10 5 2E 1 4E 9 EarthiNoke task 4E 6 24 3 1E 1 4E 7 4E 8 2400 3 1E+1 4E-7 Drywell SE 6 1 3 4E 3 3E-8 8E 7 10 3 4E 2 3E 8 ZE 8 2400 3 2.4 .E 8 total alsk: Cast 2.3E 4 Dryweti 1.4E-4 LCF = latent cancer fatality NUREG/BR-0184 C.60 O
i l
f Appendix C 1
s l hble C.33 Projected maximuun individual exposures from nonnal spent fuel transport by a
truck cask * (Daling et al.1990, Dble 4.61)
.isterne to %e .Meatsue ese .ete 4
tiervice er Activltv1 Center of Cask f(ma and 'otal Dune Gacare0 Passengers in vehicles trevellrg 10 m 30 min 60 prem/ min in adjacent tones in the same 1 mram directten se cask vehicle Traffic teatruction Pesseroers in stopped vehicles in 5a 30 min 100 pren/ min 4 less adjacent to the cast vehicle 3 mram i edtish have storped dJe to traffic ekstructic.
i
- M 4e and Pedestrfens h Stew transit (e m to traffic 6m 6 min 70 stem / min i contret devices through eres ulth 0.4 arme pedestriere) 4 Truck step for driver's rest. En- 40 a 8 heure 6 g en/ min pesures to residente and passers tyr. (assumes overnight) 3 aren a
stou trerelt throuen eroe with 15 a 6 min 20 prem/ min residents (heess, tnainesses, etc.) 0.1 mrom fruck Servicino O Aefuelfre (100 setten cepecity) 7m 60 prem/ min (at tank) 5 1 noa:Le from 1 p g 40 min 2 arme 2 nozzles from 1 gamp 20 min 1 mram ,
I Load inspection /enf w coment 3m 12 min 160 prem/ min (near personnel 2 arma ,
barrier) !
j fire change or repair to cask 5m 50 min 100 aren/ min tretter (inside tire 5 mrom ,
- nearest cask)
State welght scales 5a 2 min 80 sram/ min I 0.2 mran 1
(e) these esposures should not be euttiplied try the espected rammer of shigunants to e rappettory in en ettempt to calculate total exposures to en individael; the same
- person usutd probably not to esposed for every shigunent, nor would these ensimum
- esposure circumsterres necessarily erlee chring overy shipment.
1 V i C.61 NUREG/BR-0184 1
l
Appendix C O
%ble C.34 Piojected maximum individual exposures from normal spent fuel transport by rail cask" (Daling et al.1990, %ble 4.62)
Maalam Distance to Emposure Dese Rate (service er Activity) tenter of task Time EELI.gpl. Dean Lat.nzaa Possengers in relI care or high- 20 m 10 min 30 arem/ min wey vehistes traveting in same 0.3 mram direction ord vicinity as cask vehicle Traffic Otutruction Exposures to persons in vicinity 6a 25 min 100 aren/ min of stopped / stowed cask vehicle 2 eram d6m to raf t traffic otstruction Residents end Podestriane slow transit (through station or 8m 10 min 70 aren/ min che to traffic control devices) 0.7 mram through area with pedestriers Slow transit thro 6qph area with 20 m 10 min 30 gram / min residents (homes, businesses, 0.3 arem etc.)
Train stop for crew's personst 50 m 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 5 arew ein needs (food, crew change, first 0.6 mram sid, etc.)
frein servietna Engine refueling, car chargpos, 10 m 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 50 gram / min train maintmonce, etc. 6 eram task inspection / enforcement by 3m 10 min 200 erme train, state or federet officials 2 eran Cask car co6pter inspection / 9m 20 min 70 grav nin maintenance 1 arme Aale, h l or brake inspection / 7m 30 min 90 gram / min ierication/msIntenance on cask 3 eren car (a) Thest exposures should not be suttiplied try the expected ramber of shipeonts to e i
i repository in an ettempt to calculate total exposures te an indivichat; the same person would probably not be esposed for every shipment, nor would these analsun exposure circumstances necesserity arise chring every shipment.
l l bble C.35 Summary of results from the NRC for spent fuel shipments (Daling et al.1990, Dble 4.54) l Normal Population Accident Risk, Shtpoents Dose, Latent Cancer ItAr !!ggg_ Per Year igerson.ren/vr1 ifatalities/vr) 1975 Truck 254 93.80 0.047 Rail 17 7.78 0.021 1985 Truck 1,530 565.0 0.29 Rail 652 298.0 0.8 NUREG/BR-0184 C.62 O
l l
- - _ _ _ _ _ _m - _ _ _ _ _ - - - - _ _ _ - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
Appendix C l 1
- O J
hble C36 Maximum individual radiation done estimates for rail cask accidents during spent fuel
! transportation (Daling et sit.1990, hble 4.63)
Deze farael(a)
, Plume ed [
Accident Class Inhalation M m.
I impact 179 10.7 12.3 3
Impact and Burst 6,130 71.1 90.9 Impact, Surst and 0midation 8.950 547 707 A
(a) The maximally exposed individual dose occurs about 70 meters downwind of the release point and 5 assumes that the individual remains at this location for the duration of the passage of the plume of nuclides that are released.
hble C37 50-year population dose estimates for spent fuel rail cask accidents with no cleanup of deposited nuclides" (Deling et al.1990, hble 4.64)
Urtaan Area (3.840 - 'a/km 1 twal Area (6 -8e/km )
Ptuse Ground Pt une Grsund Accident class trhalation h h total Inhalation . Eama h total lapact Dose (person-rem) 3.09 0.33 936 939 0.005 0.0005 1.45 1.45 Letont g ith 0.19 0.00029 E f f ec ts Ispect and burst Dose (person ree) 106 2.23 13,400 13,500 0.16 0.0034 20.8 21 Latent g tth 2.7 0.0042 Effects 1spect, Surst ord Datestion 154 17.2 112,000 112,000 0.24 0.27 174 174 Dosgperson-ree) tME 22 (a) the gro6ed game dose is nAst would be received if each souter of the population stayed at the same location for 50 years. the inhaistion dose is e 50 year dose commitment from inhalation of the possine plume. Doses are for the population within 80 kilometers of the release point. It is assumed that there is no clears @ of deposited nuttides and that no other possures are used to redLce radiation esposures.
(b) Based on 1 person-rom a 2.0E LNEs. An LME is defined here as an early cancer death try en exposed person or a serious genetts heetth problem in the two generations after those esposed. About half of the LMEs are espected to be cancers and the rest genetic heetth problems.
LHE = latent health effect i
i C.63 NUREG/BR-0184
Appendix C O
%ble C.38 Pbpulation radiation exposure from water ingestion for severe but credible spent fuel rail cask accidents (Daling et al.1990, 'Ihble 4.65)
Total Release fa Population Dose Acetdent Class frem Rail Cask (C11 ) fffects from Water Incestion impact 8.07 182 person-res 0.036 LHE(b)
!apact and Burst 153 6870 person-rea 1.4 LHE(b)
Impact Burst 1379 63,000 person-res 12.6 LHE(b)
(a) Tn. Soole gas Kr-85 is omitted because of its negligible uptake by a sur' ace water body.
(b) LW4 estleates are based upon 1 person res . 2.0E-4 LHE.
Thble C.39 Summary of spent fuel truck and rail transportation risks (Daling et al.1990, hble 4.58)
Annual Average Ouantity shipping Probability of Shipped, Distance, Namber of One or More Model/ Fuel Ace INTU/rri Ikal ( stit amerit s /vr1 ftHE/vr1 Truck 180 day 380 690 885 2.2E-5 4 yr 380 690 885 3.6E-6 Rail 180-day 1,474 912 471 5.5E 5 4 yr 1,474 912 471 8.3E 7 O
NUREG/BR-0184 C.64
Appendix C
\
(~
x
%ble C.40 Summary of the routine transportation dsks for the waste management system without an MRS facility (Daling et al.1990, hble 4.59) -
"-- eitory tacation Deaf enada 3ailh Yucca ht. W 1005 Track from origin SF to Repository RadiologicalI *I 6.2 9.2 10 Nonrsdiologicallb) 18 29 31 HLW to Repository Radiological 1.7 2.1 2.1 Nonradiological 6.2 7.4 7.4 100% Rati from ortgte 5f to Repository Radiological 0.18 0.24 0.25 Nonradiological' l.0 1.6 1.6 HLW to Repository Radiological 0.063 0.079 0.074 Nonradiological 0.64 0.84 0.79 TOTALS Truck from origin Radiological 7.9 11 12
./
\
Nonradiological 24 36 38 Rail from origin Radiological 0.24 0.32 0.32 Nonradiological I.6 2.4 2.4 I
fa) Radiological health effects include lethal cancer f atalities and genetic effects in all generations. l (b) Nonradiological fatalttles. I SF = spent fuel C.65 NUREG/BR 9184
Appendix C Dble C.41 Summary of the reatine transportation risks for the waste management system with an I MRS facility (Daling r4 al.1990, Able 4.60) i l
^^- attery tacatian Deaf i Moda jggh Tucca Mt. jiggf,ggd j 1005 Truck from origin l SF to MRS i Radiological (a) 3.6 3.6 3.6 l Nonradiologica1IbI g.) 9.1 9.1 !
HLW to Repository by Truck I Radiological 1.7 2.1 2.1 Nonradiological 6.2 7.4 7.4 1005 Rail from origin SF to MR$
Radiological 0.14 0.14 0.14 Nonradiological 0.92 0.92 0.92 HLW to Repository by Rail l
Radiological 0.063 0.079 0.074 Nonradiological 0.64 0.84 0.79 i 150T Rail from MR3 1
Radiological 0.035 0.054 0.042 Nonradiological 3.8 1.0 6.1 1 Truck from ortgtn, 150T Rail from MR$
Radiological 5.3 50 O!'
i 5.7 Nonradiological 19 18IC) 23 Rail from origin,150T Rati from MR$
Radiological 0.24 0.27 0.26 Nonradiological 5.3 12 7.8 (a) Radiological health effects include lethal cancer fatalttles and genetic effects in all generations.
(b) honradiological fatalities (c) An error was found in the source document. The value in this !
table is believed to be correct. '
l l
O NUREG/BR-0184 C.66
I l
Appendix C I
1 l
%ble C.42 Aggregated public risks for the preclosure phases of the waste management system without I an MRS Facility"' (Daling et al.1990, hble 5.11) l Radiological Risk:(b) Nonradiological Risks itHE/vr1 Accidents Routine System Element Routine (health Oneratina Phase Accidents Ocarations Ifatalities/vrl effects /vr1 Repository Proclosure i Construction N/A lE-5 (c Neglipble Operations 6E 9 9E 4 (c Negitgthie Decommissioning Information 2E-II (c Negingibi. i Not l Available j Transportation System (d)
Operations IE 3 9E-2 3E 1 IE 2 T6tal Aggregated Risk. IE 3 9E-2 3E 1 1E 2 (f or Facility Operating Phases Only)
(a) Risks for the facility operations phase are annual risks for a fully functioning waste management system :perating at a 3,000 MTU/yr throughput rate. Risks for other f acility phases are leveltred annual risks prorated over the number of years required for the specific phase.
(b) Nealth effects include latent cancer fatalities plus first and second generation genetic effects.
(c) There are not expected to be site related public nonradiological fatalities. Traffic related public fatalttles are included with traffic related worker f atalttles in Table 5.12.
(d) Shipping modes are as follows: spent fuel, 30% truck and 70% rail; HLW, 100% rail.
t l
l C.67 NUREG/BR-0184
Appendix C O
Hble C43 Aggregated occupational risks for the preclosure phases of the waste management system without an MRS facility (Daling et al.1990, %ble 5.12)
Radiological Risks (b) Nonradiological Risks f1HE/vri Accidents Doerations System Element Routine J oeratino Phase (nealth tecidents Oceratiens ffatalities/vr; effects /vrl Repository Preclosure Construction N/A 11-1 2f+0 No Signiftsant lapact Operations 6( 5 2[-2 3f+0 No Significant lapact Decommissioning Information ?E-2 8I-l No Not Significant Available Japact Transportation System (C) Included 2E-2 8E-2 Operations Information With Pubitc Not Risks Available Total Aggregated Risks 6t-5 4E-2 3f+0 (for facilit Information PhasesOnly)yQperating tti hot Available (a) Risks fcr the facility operations phase are annual risks for a fully functioning waste management system operating at a 3,000 MTV/yr throughput rate. Risks for other facility phases are levellied annual risks prorated over the number of years required for the specific phase.
(b) Health effects include latent cancer fatalities plus first and second generation genetic effects.
(c) 5htpping modes are as follows: spent fuel. 30% truck and 70% rati; HLW, 200% rail, i
l NUREG/BR-0184 9
C.68
l Appendix C l
O \
'thble C.44 Aggregated public deks for the preclosure phases of the weste managesnent systent with an MRS facility" (Daling et al.1990, 'Ibble 5.13)
Radiological Risks (b) Nonradiological Risks j fLHf/vr1 Aceidonti kautine System Element .
Routine (health Oneratina Phase Accidents Omarations ifatalitian/vri affacit/vr1
( Repository Preclosure Construction N/A IE 5 (c) Negligible Operations if g 8( 7 (c) Negligible Deconstssioning Infomation 2[ 11 (c) Negligible Not Available MR5 factitty Construction No Radteactive Materials Onstte (c) No Significant Operations Of 7 SE 3 Impacts Decommissioning Not 2[ 11 (valuated Transportatiglysten 2f 3 3[ ! 4E 1 BI 3 I Operations Total Aggregated Risks 2[ 3 4E 2 4[ l 8f 3 )
(for facility O .
Phases Only)lC)perating >
(a) Risks for the factitty operations phase are annual risks for a fully functiontag easte management system operating at a 3.000 MTU/yr i throughput rate. Risks for other factltty phases are leveltzed annual i
' v risks prorated over the number of years required for the specific phase. ;
(b) Health effetts include latent cancer fatal 1 ties plus first and second !
generation genetic effects.
(c) There are not expected to be site related public nonradiological l fatalttles. Traffic related pubitc f atalttles are included with 4 traffic related worker fatallites in Table 5.14.
L (d) Shipping modes are as follows: spent fuel from reactors to MRS. 307, truck and 705 raill HtW.1005 ratin all wastes from MRS facility to repository. 100E rall. '
1
{
l 1 i
L l 1
C.69 NUREG/BR-0184 i
Appendix C O
hble C45 Aggregated occupational risks for the preclosure phases of the waste management system with an MRS facility'* (Daling et al.1990, hble 5.14)
Radiological Risks (b) honradiological Risks f D8f /vrl Acetdents peuttne System [lement koutine (health Goeratina Phase Accidents Deerations ffatalities/vri effects /vri Repository Preclosure Construction N/A 1[-] 2[+0 No
$tgnificant lapacts Operations 5[ 5 2[ 2 2[e0 No Significant Impacts l Decomissioning Information 3[ 2 7[ l No Not $tgnificant !
Available Impacts MR$ factitty Construction No Radioactive Materials Onsite 2[eo No Significant impacts i Operations 1[ 4 6[ 2 2[+0 No Significant lapacts l
Decomissioning 3[-3 $E 3 IE 1 No l $1 ntficant mpacts TransportationSystem(Cl included 8[ 3 4[ 2 Information litth Public Not Risks Available
[ Total Aggregated Risks 2[ 4 g[ 2 4f+0 Information l (For Factitt Not PhasesOnly){Cgperating Available (a) Risks for the factltty operations phase are annual risks for a fully functioning waste management system operating at a 3.000 MTU/yr throughput rate. Risks for other facility phases are leveltzed annual risks prorated over the number of years reQutred for the specific phase.
(b) Health effects include latent cancer fatalttles plus first and second generation genetic effects.
(c) Ehtpping modes are as follows: spent fuel from reactors to MRS, 35 truck and 70% ratls HLif.100% rati; all wastes from the HR$ to the repository,100% rail.
l l
hble C46 Tbtal preclosure life-cycle riskid estimates for the waste management system"
! (Daling et al.1990, hble $ 15) l Radiological Risks (LHf) Nonradiologgl l
Peculatton treun Accidents Routine Fatalities Public Risks 0.04 2 10 Occupattonal Risks 0.004 3 100 (a) Sum of risks during construction, operation, ano decomissioning phases of the waste management system.
(b) Average life cycle risks with respect to system configurations with and without an MR$ factitty.
(c) Sum of nonradiological accident and routine risks.
NUREG/BR-0184 C.70
l Appendix C ]
, v ,
'Ihble C.47 Summary of annual and total life-cycle risk estimates for the waste management system" l (Deling et al.1990, hble S.2) :
Operating Phase (b.c) Total Life.(c.d)
Risk fataanrv Annual Riiks fvele Riikt Pubite Risks
. Radiological Accidents (') 0.001 0.04
. Radiological Routine (8) 0.06 2
- Nonradiological(f) 0.4 10 ,
. Postclosure RadiologicalI9) 0.001 ..Not calculated..
Occupational Risks
. Radiological Accidents (') 0.0001 0.004 l Radiological Routine (') 0.06 3
.Nonradiological(f) 0.4 300 Risk Perspective
. Natural Background Radiation (h) 60 2000 (a) Average for waste management system configurations with and without an MR$ factitty.
(b) Annual risks from facility operating phases only. Does not include i construction, decommissioning, and repository retrieval risks.
(c) Based on 30% truck /70% rail shipments from reactors, 100% rati from the MR$ facility (where applicatle), and 100% rail shipments from high level waste (HLW) generators.
(d) Risks associated with spent fuel storage at reactor and other commercial sites are not included on the total life cycle risk estimates. ;
.(e) Annual radiological risks are given in units of latent health effects
( per year (LHE/yr): total life. cycle risks are given in units of LH[s.
5 v (f) Annual nonradiological risks are given in units of fatalttles/yr total Itfe cycle nonradiological risks are given in units of fataltties.
(g) Peak annual radiological health effects from routine releases and selected disruptive events.
(h) Based on the estimated latent health effects from the population dose from natural background radiation within 80 km of the reposttery and MR$
sites and within 0.5 km of a highway or ratiroad, i T
)
C.71 NUREG/BR-0184
Appendix C O
Thble C.48 Accident frequencies and population doses for milling in the nuclear fuel cycle (Cohen and Dance 1975)
Population Dose Frecuency for Reference Plant Accident {per plant year) (person. rem total body) fire in scIvent estraction circuit af.4 to 3t.3 1.0t.1 8elease of taillegs slurry from 4[.2 1.9(.1 tallings pond Release of tatitags slurry from II 2 8.3E.3 tainings distributton pipeltne A tey assumption is that it of the solvent estraction inventory is dis.
persed curing a fire. Study iteitations include the small number of accident
'thble C.49 Accident frequencies and population doses for conversion in the nuclear fuel cycle (Cohen and Dance 1975)
Population Dose Frequency for Reference Plant accideat (poe plant year) (person rem total bocy)
Uranyl nitrate evaporator 1[.4 to 1[.3 4.0 explosion Hydrogen esplosion in reduction II.3 to 5t.2 4.0 Fire in solvent estraction 4[.4 3.9t.1 operation Release from a hot UF 6 CIiI"#'I II*I 4*II*I Valve rupture in distillation step 5t.2 1.6E.1 Release of raffinate from maste FE.2 3.1[.1 retention pond
'Ihble C.50 Accident frequencies and population doses for enrichment in the nuclear fuel cycle (Cohen and Dance 1975)
Population Dose F requency for Reference Plant Acci dent (per pleet year) (person. rem total tiedy)
Catastrophic fire 4[.a to 3E.! 4.9 Release from a hot UF6 cylinder at.1 7.5f.) l Leaks or fatture of valves and 1.8 7.7f.3 piping Criticality BE.5 1.2E.2 O
NUREG/BR-0184 C,72 ,
1 1
\
,. _ . _ _ _ _ . . _ ._ . _ . _ . - .. _.. _ _ . _ _ . _ . _ _ _ _ _ . _ . _ . . _ _ _ _ _ - _ . ~ _ . _ . _ . _ _ ,
I i
l l l l
1 Appendix C ;
1 t
i 1 1
I Thble C.51 Accident frequencies and population doses for fuel fabrication in the nuclear fuel cycle (Cohen and Dance 1975)
Poesistion Deee Frequency for Reference Plant occident (see olent year) (serson rom total body)
Hydrogen emplosion in reduction 2E.3 to SE.2 7.4E.5 to 7.4E.2 furnace l
Mejor facility fire FE.4 7.4E.2 to 7.4E1 Fire in a roughing filter IE.! 1.0E.5 to 1.8E.2 Release from a hot UF6 cyltader 3E.! 7.8E.3 to 7.8 Failure of valves and piping 4E.3 2.2E.3 to 2.2 Criticality SE.4 1.1 Weste Retention Pond Failure 2E.3 to 2E.! 3.5E.2 1hble C.52 MOX fuel refabrication radiological accident risk Espected Pooulation Dominant Dose Risk 1 Study (person rem /GW.. year) Cont ribut or ,
Cohen and Dance (1976) 1.2E.2 to 1.9E.! (total body) Disolver fire in scrap (
recovery combined with MEPA failure. <
l Eremen et al. (1979) 4.0E.2 (total body) Sreater than design l basis earthquake. j Fv11=ood and Jackson 4.0E.7 (total body) Criticality in wet scrap.
(1960)
I l I
l C.73 NUREG/BR 0184 1
t 1
m - - . . - - , . . , _
Appendix C O
bble C.53 Accident frequencies and population doses for MOX fuel refabrication in the nuclear fuel cycle (Cohen and Dance 1975)*
Population Dose Freovency for Reference Plant Accident fper plant year) (persoa.com total body)
Explosion in oxidation-reduction scrap furnace Normal HEPA filtration 2[.3 to 5[.2 3.1[.2 NEPA filter failure 2[.6 to 5[.5 3.1[3 Major facility fire hermal HEPA filtratton 2[.4 1.6 HEPA filter failure 2[.7 1.4[5 Fire is warte concoction glove boa hermal MrPA filtratian 1[.2 3.1[.3 M!PA filter failure 1[.5 3.1[2 len-eschange resin fire Normal MEPA filtration 1[.4 to 1[.1 9.![.3 HIPA filter failure 1[.7 to 1[.4 9.!!!
Dissolver fire in scrap recovery hermal M!PA filtration 1[.2 1.6[ 1 NEPA filter fatture 1[.3 1.6[4 Glove failure hermal H[PA filtration 1 1,3[.$
MEPA filter failure 1[.3 1.3 Severe giove bon damage Noemal HEPA filtration 1[.2 6.1[.2 NEPA filter fatture 1[.5 6.1[3 Criticality hormal HEPA filtration 3[.5 to 8[.3 3.8[.1 ptPA filter failure 3[.8 to 8[.6 4.![2 HEPA = high efHeleng paniculate air hble C.54 Accident frequencies and population doses for MOX fuel refabrication in the nuclear fuel cycle (Erdmann et al.1979)
Population Dose Frequency for Reference Plant Accioent (per cient year) (person. rem total body)
Greater than design basis 5[.6 1[5 ea rthquak e Aircraft crash 3[.7 3[4 Hydrogen empleston in ROR reactor 1[-3 5[.9 Mydrogen explosion in sintering 1[.3 2[.7 furnace ton enchange resin fire 5[.a !!.9 Dissolver esplosion wet scrap 5[.3 !!.6 recovery Loaded final f11ter failure 2[.4 3[*1 Criticality 6[.5 5 NUREG/BR-0184 O
C.74
i Appendix C
' l S
G l i
i hble C55 Accident frequencies and population doses for MOX fuel refabrication in the nuclear fuel cycle (Fullwood and Jackson 1980) p wuistion osse
! er..vency for atter.nc. piant Accident (oer plant year) (oteson. rem total teody}
Aircraft crash 1.5(.9 5t!
Myerogen saplosion in ROR 5E 3 1.1t.11 Myerogen emplosion in sintering 5E.3 4t.10 Myerogen explosion in wet scrap St.4 1.11 11 1
Criticality in wet scrap 6E.5 2 powder shipping container set 11 3E.5 1.lt.11
[nothermic reactions in powder 1.5t.6 1E.10 storage
%ble C56 Fuel reprocessing radiological accident risk tapectes population nominant '
nose nis 5tutty (oersen-res/GW..yearl Cont ribut or Cohen and Dance (1975) 2.BE.3 to 4.3t.3 (total body) Fuel assembly rupture combined with H[pA fatiure.
Ichn et al. (1979) 2.0E.4(totalbody) Krypton cylinser failure; esplosion ih HLW calciner.
Futtwood and Jackson 7.0E.5 (total body) Krypton cylineer f ailure. ,
(1980)
ROR = reduction-oxidation reactor l
l l
k C.75 NUREG/BR-0164 i
l Appendix C O
Thble C.57 Accident frequencies and population doses for reprocessing in the nuclear fuel cycle (Cohen und Dance 1975)")
Population Dose I Frecuency for Reference Plant Acetoeat fper plaat year) (pertoa. rem total body)
Explosion in MAW Concentration hermal HEPA 1E.5 4.312 Failed MEPA 1E.8 9.5E3 l Explosion in LAW concentratton l hormal HEPA 1E.a 2.8E1 i Failed M[PA II.7 4.8t1
{
Explosion in MAW feed tant normal HIPA lt.5 1.6E3 Failed MEPA 11 7 1.7t3 Emplosion in waste esiciner hormal MIPA l tt.6 4.3t3 l Failed MEPA 1[.9 1.3ta i Explosion in iodine absorter 2[.4 a.8 Solvent fire in codeep cycle hermal MEPA 1E.6 to it.4 2.3t1 Failed MIPA 1E.9 to it.7 5.6tl l
Solvent fire in Pu entraction cycle Normal MEPA 1E.6 to 114 3.1E.4 Failed HIPA It.11 to IE
- 5.2E2 ton exchange resin fire hermal MIPA 1E.4 to 1E.1 3.6E.1 Falled HEPA If.9 to 1E.6 1.813 Fuel assembly rupture in fuel l,
receiving and storage i Normal MEPA II.2 to IE.1 1.3E.2 i
Failed MEPA 1E.5 to it.4 1.313 Dissolver seal failure i
hermal MEPA II.5 2.3E.2 l
Failed NEPA II.8 2.3t3 Release free hot LIF6 cylinoer 5t.2 1.5 I Criticality horsal HEPA 3E.5 to 8E.3 3.OE.2 Failed MEPA 3t.8 to 8f.6 3.5E.2 HAW = high activity waste LAW = low activity waste j 1
O l NUREG/BR.0184 C.76 l
i l
I Appendix C !
h G '
Mie C.58 Accident frequencies and ;:, ' = doses for reprocessing in the nuclear fuel cycle l
(Ersknamn et al.1979)')
l I peoulatten Dese Frequency for Geference plant Aceteent (ser plant year) foersen-ree total body)
Less of fuel storage peel water M-4 50 Ion enchange bed fire and eglesien $(.4 PE.1 Criticality M.5 5 Mydrogen eglesten in MAF tank 71 5 7E.!
Fire to tem level weste IE.! It 1 Fuel asumbly drop 21.J 11 1 Espleston in high-level weste St.10 SE6 calciner cambined with NEpA filter failure Krypten cyltneer rupture IE.4 60 HAF = high aqueous feed i
I l
l 1
1 f
C.77 NUREG/BR-0184
l Appendix C i hble C59 Accident frequencies and population doses for wprocessing in the nuclear fuel cycle e l l
(Fullwood and Jackson 1980) j Pooulatten Dese Freevancy for Reference Plant ,
Accident (per plaat year) (person.res total IMHty) i H fire an explosion in MAF tank 3t.6 9[.4 e mbined with one MEPA filter f tied ;
Solvent flee in the My concen. 21 6 7t.4 tration combined with one MEPA filter failed Wed oil emplosion in MLW concen. 4E.8 St.3 tretion combined with one MEPA filter failed Explosion in the MLW calciner 2[.7 2E.1 combined with one MEPA filter failed Red oil explosion in the fuel 4[.8 6[.4 product concentration combined with one MEPA failed IP-plosion in fuel product 4[.g 1.!!.2 deltrator combined with one MEPA failed Criticality in a process cell !!.5 I f ailure of Krypton storage 1.3t.4 4[1 cylinger Hydrogen esplosion in urentum 9t.6 1.4[.4 reduction combined with one M[PA filter failed Fuel assembly drop 1.21 3 SE.2 Hydrogen emplosion in fuel 3t.6 1.2t.2 product centtrator fusel tank combinee w11th one MEPt filter fetied hble C60 Accident frequencies and population doses for reprocessing in the nuclear fuel cycle (Cooperstein et al.)
Popuistion Dose Frequency for Reference Plant Acci dent {per plant year) (person.rea total body)
MAW concentration amplosion 1E.5 57 Cocecontaminetton solvent fire 1[.6 2.6 LAW concentrator explosion it.4 3.3 MAF tant emplosion 1[.5 4.9t2 Weste calciner explosion it.6 $.112 Fuel receivieg and storage it.t 2,og.3 accident O
NUREG/BR-0184 C.78
Appendix C Dble C.61 Accident frequencies and populatlan doses for spent fuel storage in the nuclear fuel cycle (Karn.Bransle.Sakerhat 1977)
Population oese for Frequency Reference Plant Acci dent (per plant year) (person. rem tot al body) l Fuel transfer basket is deopped PW IE.4 I be 2.5f.4 1.8 Fuel assemblies a 9(.4 ff.1 pm 6t.3 31 1 Dble C.62 Accident frcquencies and population doses for solidified HLW storage in the nuclear fuel cycle (Smith and Kastenberg 1976)
Population Oose Frecuency for Reference Plant i l
Accident (per plant rese) (person ree total body)
Major rupture of a waste canister 1.0E.4 7.2 dropped curin handling. Vent s,st . offect .e Major rupture of a weste canister 1.0E.6 7.!!!
with an ineependent failure of one HEPA filter 0.1 1 ton meteor impact in storate 4.1E.9 1.0E5 area 10100 ton meteor impact in 2.0E.10 n.1E6 stora9e area 0.1 1 ton meteor impact in 4.8t.10 3.1[5 l receivin9 area l 1 10 ton meteor ispect in 1.25t.11 2.6E7 receivin9 area i
hble C.63 Preciosure geologic waste disposal radiological accident risk ,
l tapected Population Dominant Dose Risa Study ( pe rson.com /GW.. yea r i Comt eibut oe USDOE (1979) Spent Fuel Waste package dropped l 2.lt.9 (whole body) down shaft l Glass HLSW 9.6E.12 (whole body)
Erenan et al. (1979) Class ML5W Final Filter Fatture i 4.0f.5(wholebody)
C.79 NUREG/BR-0184
Appendix C O
%ble C.64 Transportation radiological accident risk'd ;
Study Plutonium Oside Spent Fuel kigh Level Waste Cohen and Dance 1.2f.3 to 1.7t.2 3.5t.3 to 1.6 (1975) (total body) (total body) frihan et al. (1979) 1.0E.3 3.0t.5 3.0t.3 (total body)
(total booy) (total body)
Fullwood and Jackson 3.01 5 1.0f.5 (tetal body)
(1980) (total body) 05001 (1979)* 5.01 5 1.117 (total body)
(total body)
U$NRC (1977)* 1.41 1 (total body)
Berman et al. (1978)* 9.4t.3 (total body)
USAEC (1972): USNRC* 8.3t.3 (1975); U$NRC (1976) (total bsdy)
Hodge and Jarrett* 1.21 2 5.114 (total body)
(1974) (total body)
UshRC (19?6)* 2.3t.6 5.417 (total body)
(total body)
(a) Measured in person-rem /GWe-year
%ble C.65 Accident frequencies and population doses for transportation of spent fuel by rail and PuO2 by truck in the nuclear fuel cycle (Cohen and Dance 1975)
Po'pulation Dose F requency for Generic Shipment Ace t dent _
(per shipment ) (person-rem total body)
Spent Fuel Leake9e of coolant from spent 31 4 5.8f.4 fuel cask Release from a colliston 2f 8 to 9t.6 1.9(4 involving spent fuel Release from a colliston involv. 21 10 to 9t.8 2.714 '
in9 spent fuel followed by '
release of fuel from the cask Plutoalum Ost de Improperly closed plutonium at.4 to it.1 1.1 l eside container Releese from a colliston 2t.9 to 3!.6 1.4E3 involving plutonium oxide Criticality of plutontum 2t 11 to 3!.8 2.5E4 calde O
NUREG/BR-0184 C.80 l 1
_ _ . ._ _ _ . _ _ _ . _ _ _ _ . . ._ _ . . . . _ _ - _ _ _ _ _ __ _. _ ~
i Appendix C f
\
'Ihble C.66 Accident frequencies and population doses for transportation in the nuclear fuel cycle (Ert imann et al.1979)
Population Dese Freguency for Generic Shipment ~
j Aceteent (per shipment) (person res total body)
Spent Fuel by Ratt l Less of pses from inner cavity 9E.6 ft.6 from raft accident !
Less of confinement and $01 4t.7 !!.1 fuel dame y Loss of confinement, 505 fuel 2t.9 !!3 damage, extensive fire Speat Fuel by Truck l
Loss of gas from inner cavity 2f.5 l!.9 i from truck accident Loss of confinement and 601 !!.7 1(2 l fuel damap l Less of confinement. 501 fuel 21 9 612 dame y , extensive fire Plutonium Oride by Truck Truck accteent it.6 release 1E.6 2 fraction !
Truck accident It.4 release 41 11 til l t fraction s
Truct .ccident 112 release 6t.8 214 fract:en ,
Mish-level Was e by Ret)
Release to atmosphere and one it.l y!!
centster breakage from rail l accident Release to atmosphere and 6E.8 6(3 significant overheating C.81 NUREG/BR-0184
Appendix C Thble C67 Accident frequencies and population doses for ral! transportation la the nuclear fuel cycle O.
(Fullwood and Jackson 1980)
Population Dese F recuency for Generic Shipment Accident _ (per shionent) (person. rem total bog 5 pent Fuel loss of centron shielding from tt.l St.7 a rail accident taposure of the inner spent fuel 91 6 1.71 6 containing cavity Esposure of the inner spent fuel 4t.7 0.5 containing cavity and 505 fuel damage taposure of spent fuel with 3t.8 1.7t3 severe damage and fire Nigh level Weste Loss of neutron shielding from it.8 St.5 a rail accident telease and extensive centster 3t.10 30 dama ge telease, entensive canister St.12 3t3 damage and fire
'Ihble C68 Accident frequencies and population doses for rail transportation in the nuclear fuel cycle (PSE 1981)
Population Dese F requency for Generic Shipment Accident (per year) (person. rem total body) j 25 40 m fa11 ft.6 2.8t.1 l 9 25 m fall !!.5 2.8t.1 ,
50 80 km/hr co111ston 2t.5 2.8t.1 80-100 km/hr colitston 3t.4 2.8[.1 C >l hr 8t.5 1.712 Co111ston tellision and fireand 800"fire 100[C > 2 !E.5hr 1.7t!
- TC11hr IE.4 f.01 1
'.;it'C >! hr it.5 2.0E.1
.ston and closure errors it.4 1.1 Thble C69 Accident frequencies and population doses for rail transportation in the nuclear fuel cycle (Elder 1981)
Population Dose Frequency for Generic Shipment Accidont (per shtoment) (person-res total body)
Rett accitlent and impact fails 6.4t.4 6.8t!
cask seal causes loss of coolant and fuel falls Side impact fails pressure rettet 1.2t.6 1.9[3 valve causing loss of coolant and fuel falls End tapact fails pressure rel',ef 6.4t.6 1.9[3 valve causing loss of coolant and fuel fails Side impact fails cask seal 1.ft.6 6.8t2 causing loss of coolant and fuel fatis NUREG/BR-0184 C.82
Appendix C l
'Ihble C.70 Nornealized risk results for nuclear fuel cycle Espected Populaties l Dese (Tesal Body l marn h G We-vaar)
Fuel Cycle Elamment Origleal Norimallmed Reference Milling 1.0E3 2.7E-4 (Cohen and Dance 1975) l Conversios 5.6E 3 1.2E2 (Cohen and Dance 1975) '
EnriA==r 3.7E-3 1.2E 2 (Cohen and Dance 1975) 3 Fuel Fabrication 1.0E-2 5.0E 3 (Cohen and Dance 1975)
MOX Fuel Rafabrication 1.9E 2 1.2E l (Cohen and Dance 1975) 4.052 3.6E-2 (Erdmann et al.1979) 4.0E7 3.3E5 - (Fullmood and Jackson 1980)-
l Fuel R.,a-- -g - 3.1&2 (Wood and Becar 1979) 6.3E-3 3.2E 3 (Cohen and Dance 1975)
- 5.6E-4 (PSE 1981) >
2.0E4 2.2E-4 (Erdmann et al.1979)
- 1.5E-4 (Cooperstein et al.1979)
- 7.0E5 5.4E-5 (Fulls. cod and Jackson 1980)
Spent Fuel Storage - 1.8E 1 (PSE 1f,81)
- 3.1&2 (Wood and Becar 1979) 1.7E-6 3.7E 5 (USDOP.1979) 2.0E-5 2.7E5 (Erdmara et al.1979) 8.9E 5 5.7E-6 (KBS 197Y) l l Solidined High Level Waste 2.3E-4 2.3E-4 (Smith and Kamenberg 1976) l Geologic Waste 4.0E 5 4.0E 5 (Erdmann et al.1979)
Disposal (proclosure) 2.1E-9 2.1E9 (USDOE 1979)
! Transportation l Plutonium Oxide 1.7E-2 6.6E 2 (Cohen and Dance 1975) 1.0E 3 1.3E 3 (Erdmann et al.1979)
Spent Fuel - 1.6E 1 (Elder 1981) 1.4E 1 1.6E-1 (USNRC 1977) 1.6 7.8E 2 (Cohen and Dance 1975) 1.2E-2 1.3E2 (Hodge and Jarrett 1974) l 8.3E 3 9.3E 3 (USAEC 1972) 7.lE-4 (PSE 1981) 5.0E 5 5.6E-5 (USDOE 1970) 3.0E 5 B.4E-6 (Erdmann e al.1979) 3.0E5 8.4E-6 (Fullmored and Jackson 1980) 2.3E-6 2.6E-6 (USNRC 1976)
High 1.svol Waste 9.4E-3 4.2E-2 (Berman et al.1978)
- 5. lE-4 2.3E3 (Hodge and Jarrett 1974) 3.0E-3 8.4E-4 (Erdmann et al.1979) 1.0E-5 2.8E-6 (Fullwood and Jackson 1980) l 5.4E7 2.4E6 (USNRC 1976)
C.83 NUREG/BR-0184 -
Appendix C O
Hble C.71 Capital equipeneet costs for fuel pellet fabrication (MiaMme et al.1983, hble 1)
(quipment/ Procedure Description Manufacturer Cost 2 Glove bones Insth floor dimensions: 5' 3' s Molitar $ 52,000
, , , , , e' 11" Englewood, Colorado 16 glove ports Boa wall 0.25* lead sandwiched between stainlets steel sheets sheets 0.125" Windows: Leaded glass Gloves: Lead. loaded neoprene.
0.040" thick 2 Balances Cat. #3330 04 $ctentech $ 4,100 Load cell with remote controls and Soulder, Colorado resoouts. Qual range: To 3 kg, 0.1 g sensitivity; to 300 g, 0.01 g sensitivity Ory Granulator (RWEKA Granulator Chemical and Pharmaceutical $ 3,600 Orive AA 400 Co., Inc.
Granulator 7G 2/5 225 Broadway, New York Blender "Turbula:" Type T2C Chemical and Pharmaceutical $ 3,000 Co., Inc.
225 Broadway, New York Press 30 Ton Western Statering $110,000 Hydraulic, double acting Richland, Washington Reservoir and pumps reemoto (outside glove boa)
All controls outstde glove boa
. love boa $10,000/ bon $ 20,000 installation [ngineering and Crafts: 425 h at 547/h Equipment Press: 200 n at 146/h $ 14,720 installation Other: 120 m at $46/h TOIAL $ETTM
- 4egistered trademark of Willy A. Bachofer, finnufacturer, 84511. Switserland O
NUREG/BR-0184 C.84
. -= - -. . - , . - . . - . - - - - . -
l Appendix C *
. 0%
l Table C.72 Capital equipment costs for powder reconstitution during fuel fabrication (Mishima et al.1983, 'Ihble 2) l (qvf pment / Procedure Description Menuf act urer Cost 2 GIove bones ins 1de fIoor disensIons: Mol1 tar $52.000 >
l' 3" x 4' 11* Englewood. Colorado 16 Glove ports l Son well 0.75* lead sandwiched between stainless steel sheets 0.125' Windows: Leaded glass Gloves Lead-loaded neoprene, 0.040* thick Balance Cat. #3330-04 Scientech $ 2,100 Load cell with remote controls and Sovider, Colorado and readouts. Dual range To 3 kg, 0.1 g sensitivity; to 300 g.
0.01 g sonsttivity Dry Granulator ERWIKA Granulator Cheetcal Pharmact.aital $ 3.600 Ortve M 400 Co., Inc.
Genaulator TG 2/5 225 troadway New York Furnace Model #51442 Lindberg $ 1.950 Control model #5g344 (remote) Watertown Wisconsin 4000 watts Esterior dimensions: 20* W s 20" M x 24.5" L Mill rack and mills Rack Pbdel #764AV 30 1/4" a E. T. Horn $ 2.310 12 3/4* a 15 3/4" H La Mirada Caltfernia 3 Mills: Rubber. lined steel size 1 Stainless steel balls, 0.5*,100 lbs
. ([
' )
Glove bon installation
$10,000/ box Engineering and Crafts: 425 h 120.000
%/ at $47/hr
[quipment 160 hr at $46/h $ 7.360 installation TOTAL M ]
i I
I
! r f
l C.85 NUREG/BR-0184 i
1 i
i Appendix C l 1
i hble C73 Start.up operation costs for fuel fabrication (Mishima et al.1983, Dble 3)
O Process personnel Job Descriptton Cost Pellet factication Engineer 120 h at $65/h $16.400 ;
i Prepare detailed operating procedures i in conjunction with an operator. l Supervise equipment shakedown.
Operator 120 h as $50/h Operate equipment start-up and 1 shamadown !
... Preodration of criticality spectftcation: i 40 h at 165/h
... Radiation munitortng: Included in labor contract Powder reconstitution Engineer 120 h at $65/hr $16.400 Prepare detailed operating proceduras in conjunction with an operator. Supervise equipment shamadown.
Operator 120 h at $50/h Operate equipment start-up and shakedown hble C74 Process operation costs for fuel fabrication (Mishima et al.1983, Dble 4)
Process Pellet Fabrication Estimate assumes 3 snif ts/ day processing a 100.ng mintmum lot of Pu02 powder.
Two operators / shift at $50/h/ operator Mastam 20 kg powder processed / day Labor cott/kg $120.00 Radiation monitoring: Included in labor overhead.
Supplies /kg: Does not include items reautred for shipping e. i.30 powder. Includes such items as stainless steel cylinders.
neoprene lead.lcaded gloves for replacement, organics.
Only utt11ttes: Electricity /kg 0. 80 k Wh total pellet faorication price /kg $122.00 Powder Reconstitution W vrator/ shift for a h at $50/hr 10 kg peIIets processed to powder in 4 shifts 16 h labor Labor cost /kg 1 80.00 Radiation monitoring: Included in labor overhead.
Supplies /kg $ 0.75 Only utilities: Electricity /kg !!.0 kWh Total powder reconstitution price /kg $ 81.00 NUREG/BR-0184 C.86 0
E i
Appendix C
()
Thble C75 Summary of dose equivalent estimates for febricating PuO2 Powder to un8 red pellets I during fuel Fabrication (Mishima et al.1983, Thble 9) i total Dose fautvalent for Three. Person Crew Processtag 100 kg of Puo, (man. rem L Average of Light idater Reactor Low.Esposure <
l Plutonium Produced in 1985 Plutentum '
Contact or hand exposure 67.0 18.0 (gamma only)
Whole body dose equivalent including roen background Avera ge 0.95 0.14 Range based on vertations in room background (0.87 to 1.1) (0.11 to 0.15) '
%ble C76 Summary of dose equivalent estimates for reconstituting un8 red PuO2 Pellets back to powder during fuel fabrication (Mishima et al.1983, hble 10) i Total Dose Equivalent for two Person j
Crew Processing 100 kg of o u 0., (man. rem) j Average of Lignt mater Reactor Low.E mposur e Plutonium Produced in 1985 Plut onium I
Contact or hand esposure 64.0 17.0 I (gama only) k' Whole.bndy dose equivalent ina.lu'ttnq rorwn background ,
i Average 0.19 0.038 Range based on (
variations in room p bactground (9.14 to .26) (.03 to .06)
N C.87 NUREG/BR.0184
Appendix C l
O Table C77 Accident source terms and doses from uranium mill accidents (McGuire 1988. hble 3) resi.e. .r in air air. i. s.i. t ci..aig 5,it s.r.iae in.
v.=u r.iisas e a.iu.. f.trusi circ.it v.iima. orria, aru are.r. . Co.. on, s.i u. o. . seina oeu n.iuu on.
cris si..oo se u t.t.i .: : . io ' r uno t.u ..iin s ii. cina .n.v . i x m. n as in.iai. u.=
. n 4ao as a t. ione. a soc. u. coo.ooo sei. u. . o.ss is in.ri.- t. .t uusu. i. ioas r... irai. ii in soo . ..., e n .t tooo .
t.a. sui esso is u t.t.: . i : . to ' - s u cris .
. i. is u to . r ua i n. i a i. to a -
Ns . esbo as u .4 ooo. L. n.a. U ..si . ..., t. iung .t t..p t ral. (e.. o.) .t Sooo a 8 but. aooo a
< a.. ru t ca.. m t r.. in= . : m i ac.:
gage . . . .
i.3 te a o.oi t. -
a.p.,t o i r.e (of
. re . .ar . .
Gl is . 'i ta.i Gener ts (a.6r.=at.1 lacus st.t.a.at .a Ur.af. niiling, ens.(C-olo6 v.s i. po f* : t. 7 fo, s.es ner, it.o s.na su. Dit *or.f t in.e r.re.at.1 stat.e.at 8.i.ted t. the op.t.ti .f 5.ad 8.c . mi ei Pr.J.c t.* maiG-ooet. e.ge s-i t.
s-is, m.nm su
.in. ...i. f~ c4isi..a.er., in. ..i..at utrusi.a .a u. t. c.ac.i. u o =a sa so a . ., .at - is. ..i- a ie
.. a u *r **i.*ir f ai. ere=r c. .a ia. .a. t. 6. ...ratiates or . f u t.r et a=ad 50.000 u=.
hble C78 Offsite doses calculated for fuel fabrication plants (McGuire 1988, Table 9)
Criticality UF,-lona enrich. UF,-high enrich.
Key Analysis Ass.mptions Effective DE Thyroid DE Effective LE Bone OE Effective DE NUR(G-1140 Building size: 250 as 0.5 to 1.1 to -
0.2 to Wind: F. 1 m/sec 2.6 rees at 8.2 rees 1.5 res Release height: ground 100 m at 100 e at 100 m (child's thyroid)
Combustion Building site: 0 0.27 res 1.7 rees 0.05 res 0.82 rum -
Engineering Wind: F. I m/sec at 800 m at 800 m at 800 m at 800 m Release height. stack Ermon Buildtr.g site: 0 0.009 res 4.5 rees 0.11 res 1.7 rees -
Wind: F la/sec at 2000 m at 2000 m at 2000 m at 2000 m Release height; ground hf5. Erwin Building size: 0 -
5 reos - -
1 ree Wind: G. 0.5 m/sec at 1000 e at 1000 o Release height: same level as residence l
DE = dose equivalent EDE = effective DE NUREG/BR-0184 C.88 O
J
- - .. -.. - . - . _ _ _ - _~ - - _-. - - .- ._ . .
l Appendix C l
,~
ks l
hble C.79 Dose commitments from plutonium fuel fabrication facility accidents (McGuire 1988) '
l l
Type of accident Oose commitment (r**) i l Criticality 0.36 (thyrold) i Fire 0.02 (bone)
I taplosion 0.02 (bone) i !
Dble C 80 Maxicium offsite individual dose commitments (Rem) from spent fuel reprocessing l facility accidents (McGuire 1988) f
'hartous Off ette Individual Dese Commitment (res)
Accident pwe nor Fuel Criticality 0.056 (thyreld)
Weste Concentrator (splosion 0.0069 (bene)
Pu Evaporater Espiesten 0.019 (bone)
Fire 0.0135 (bone) hble C.81 Calculated releases and doses from spent fuel storage accidents (McGuire 1988, hble 10)* l l
Kr-85 Skin Effective Dose Thyroid Reference Accident Release Dose Equivalent I-129 telease Dose l j 5torage in pools: Tornado driven 19,000 Cl 0.06 rom Not calculated 0.00006 Cl 0.03 ren Generic Environmental missile followed at 275 e at 275 m
' (V Impact Statament, by cale MuaEG-0575 Storage in pools: Drop of a fuel 6,000 Ci Not 0.016 res 0.00008 Cl 0.0004 rem GE-krris SER, storage basket calculated at 150 e at 150 m NUaEG-0709 Dry cask, dryinell, Removal of cask 8,000 Ci Not 0.003 res 0.004 Cl 0.005 to or dry vault Ild with all fuel calculated at 100 s 0.04 rem -
storage: NUAEG-1140 elements ruptured within 100 m (child) l t
l l
1 I
l
\
C.89 NUREG/BR-0184
Appendix C 1hble C.82 Maximum possession limits, release fractians, and doses due to a msdor facility Are O
for radiopharmaceutical manufacturing (McGuire 1988,'Ihble 14)
Maalem licensed l i
nodi .ctlee Messesslen note ss gerectie,isose materlet limit (Cl) Licensee fractlen egulealent, res** j 1
H-3 150,000 MEN
- O.5 0.1 to 10.
C 14 500 MIN-Boston 0.01*** O to 0.01 P-32 500 NEN 0.5 0.04 to 4.
5 35 1.000 MEN 0.5 0.01 to 1.
Co-4 50 NEN 0.01 0 to 0.003 '
Cr-51 100 Mtu 0.01 0 Fe 55 200 NtN 0.01 0 to 0.005 )
Ni 63 1,000 Ntn 0.01 0.001 to 0.06 Se 75 100 MIN 0.01 0 to 0.008 i Kr-85 10,000 NIN 1.0 0 to 0.002 !
tb-86 50 NEN 0.01 0 to 0.003 5r-90 500 Mtu D.01 0.05 to 5.
No-99 2,000 NEN/5quibb 0.01 0.001 to 0.08 Ru-103 25 MEN 0.01 0 to 0.002 Sa-113 100 MIN 0.01 0 to 0.01 1 125 100 ktN/N11 tfu hredt 0.5 0.3 to 30. (calld's thyreld) l 1*131 500 14allinckrodt 0.5 5 to 500. (child's thyreld)
Re-133 1,000 Ntu 1.0 0 to 0.001 Cs-134 25 NEN 0.01 0 to 0.01 Cs-137 500 MEN 0.01 0.002 to 0.2 Co 141 50 NEN 0.01 0 to 0.004 Yb-M9 50 NEN 0.01 0 to 0.004 lo-170 25 NIN 0.01 0 to 0.006 Ar198 200 NIN 0.01 0 to 0.008
- MIN e new Encland huclear, North Ol11 erica, Mass.
- sero in the dose col on indicates a dose of less than one millites.
- Non-carbon dioalde rolesse fraction.
i l
1 1
l 1
l l
C.90
l Appendix C l
l Dble C83 Maximum possession limits, release fractions, and doses due to a mWor facility Are for a radiopharmacy (McGuire 1988, hble 15) l mast tisensed Des.
nadi. acts.: siessi.n cheatcai neiease eevivaitat, material itelt (Cl) foros fraction ree H3 0.05 Cl in vitre test kits 0.5 0 C 14 0.05 In vitre test kits 0.01' 0 Cr 51 0.15 Labeled serum, 0.01 0 sodium chrenate-Co 50 0.15 Cyanecobalenta 0.001 0 (witeeln 312) fe-59 0.15 Chlertee, citrate, 0.01 0 sieltate Se*75 0.1 Labeled compound 0.01 0 Sr-90 0.5 Nitrate, chlertoe 0.01 0 to 0.006 Mo-99/1c-99* ?$. Mo-99/1c-99m 0.01 0 to 0.00e ponerators (Itquid) 1 125 0.15 ha 1. fibrogen. 0.5 0.001 to diagnostic kits 0.1 (cid's thyroid I 131 0.75 Na 1, labeled 0.5 0.007 to organic cog ounds 0.7 (child's thyroid)
Ne-113 1. Cas or saline 1.0 0 Note: sealed sources are not included, neference: Sutter report.
'Non-carben dientee release frer .w.
s l
I i
l J
C.91 NUREG/BR-0184
Appendix C O
Able C.84 Maximum possession limits, release fractions, and doses due to a major facility fire for sealed source manufacturing (McGuire 1988, hble 16)
Manteue Iffective licensed dose Radioactive posession Release equivalen esterial llelt (Ct; fern Licensee fraction reos M-3 100,000 Cl volatile Safety Light 0.5 0.06 to 6 C 14 50 Amershan 0.01' 0 to 0.00 Co-60 20,000 753 metallic Automation 0.0001 0 004 to pellets Ind. 04 251 sealed sources ar-85 1.500 noble gas 3M 1.0 0 tr-to 3,000 1000 Cl in 3M 0.01 0.3 to 33 solutlec in 0.1 liter of 0.1 W MCI aise, sealed sources56-124 50 Monsanto 0.01 0 to 0.01 1-125 100 5 Cl in KOH 3M 0. 5 0.7 to 70 lieute (child's 5 Cl en eesin thyreld) beads ts-137 10,000 Tech / Ops 0.01 0.03 to 3 Po-147 3,500 800 Cl in 3M 0.01 0.008 to solutica in 0 0.1 Iliee of 0.1 N MCI aise, sealed sources Yb-169 100 5 C1 Ileufd 3M 0. 5 0.004 to Yb chelate 0.4 le 170 5,000 1ech/ Ops 0.01 0.01 to 1.
la 182 200 metallic or Tech / Ops 0.01 0 to 0.001 carblee Ta 183 2,000 meta 111e er Tech / Ops 0.01 0 to 0.00) carblee Ir-192 50,000 selld metal Tech / Ops 0.0001 0.001 to or sealed 0.1 source 11 204 50 Monsente 0.01 0 to 0.001 81-210 200 metal slugs
- 0.001 0 to 0.03 Pe-210 4.000 up to 1500 C1 3M 0.01 1. to 100.
In 40 flteri (per of 2M HNO ;
up to 2500 Cl 1500 C1) la neste 0.001 0.2 to 20 (per primarily as elcrespheets 2500 Cl) ha-217 0.1 i Monsante 0.001 0 to 0.04 Pu-234, 235, 199 g 250 C1 as 239, 240 Monsanto 0.001 0.75 to 75.
i unsealed 241. 242 powder enlee (see l 250 Cl)
An-241 6.000 250 C1 as Mensante 0.001 1.2 to 120 unsealed (per powder eside; reestner as 250 Cl) sealed serrces Co-242 600 Mensante 0.001 0.1 to 10.
Co-243 10 Mensante 0.001 0.03 to 3.0 to-244 600 Mensante 0.001 1.5 to 150 Cf-252 10 og solid pellet Monsente 0.001 0.006 to
- 0. 6
- Non-cartpon diestde release f raction.
NUREG/BR-0184 C.92
l Appendix C l
i V l %ble C85 Maximme possession ',lmits, release fractions, and doses due to a major facility Are for l university research laboratories (McGuire 1988, %ble 17) )
i l i sediescil e me is. iscensed a.iesse etfective dos, meterial possession 16elt (Cl) fraction egulveient, rest i M-3 3000 0. 5 0.002 to 0.2 C+14 10 0.01* O P-32 5 0.5 0 to 0.04 5 35 5 0.5 0 to 0.01 Mi 63 1 0.01 0
$r-90 0.5 0.01 0 to 0.005 Me 9fAs 99m 10 0.01 0 1 125 8 0.5 0.06 to 5.5 (chlid's thyreld) 1*131 1 0.5 0.01 to 1. (chlid's thyrefJ)
Re-133 10 1. O Po 210 to 0.01 0.009 to 0.9 As-241 0.5 0.001 0.003 to 0.1 te-244 1 0.001 0.003 to 0.3 Cf 252 0.1 0.001 0 to 0.01
- hon certron disalee rettese f ractlen.
l Dble C86 Waste warebousing airborne releases and doses due to a mWor facility Are (McGuire 1988, hble 18) 1 l
i nodi.ective oventity sei.... Effeett.e dese meterisi present (ci) fraction eevivaieat, re. ,
H-3 6200 0.5 0.004 to 0.4 c 14 160 c.01* O to 0.004 P 32 160 0.5 0.01 to 1.
5 35 120 0.5 0.0c! to 0.2 cr-51 60 0.01 0 1-125 200 0.5 4 to 400. (chlid's thyreld) 1 131 20 0.5 0.4 to 40. (chiid's thyreld)
- n.n-care.n dientee rotes.. frection.
hble C87 Alternative disposal standards for uranium mill tailings (EPA 1983, %ble S.1) longe vit y Radon Control af ter Disposal (pC1/e e)
Requirement ho kacos kequirement ou Ju o 2 l No Controls A 1
Active control 21 12 BJ l
ttre 100 years Passive centrol C1 (2 03 04 C3 for 1000 years Paselve control for 14 D3 D4 DS 1000 years, with teproved recon control durlog operations for new piles
\
, V C.93 NUREG/BR-0184 1
Appendix C O
hble C.88 Alternative standards and control methods for existing uranium mill tallings piles (EPA 1983, hble 4.2)
Control Mett od Charact erist {ce
.5m Pebbly Alternatave Control Method Earth Cover Rock on 4011 Des i gnat ion Thickness (m) Slope Slopes on top Maintenance Landsca ping StandarJ
=
A 31 al-t 0.5 31 100 years X B; g2.g 1.5 3:1 100 years E 33 33.E 2.4 3:1 100 years I Cl Cl-E 0.5 5:1 X X C2 C2-E 1.5 5:1 X X C3 C3-E 2.4 5:1 X X C4 C4-E 3.4 5:1 X X CS C5-1 4.3 5:1 X .X D2 5eme se C2 03 Same as C3 D4 Sue as C4 05 Same as C5 Able C89 Alternative standards and control methods for new uranium mill tallings piles (EPA 1983, hble 4.3)
Control Method Chaterter(st{es
.5m Febbly Alterestive Control Nethod Earth Cover Rock on Soil Put
$tandard Designation Thickness (m) 51 ope Slope s on Top Maintenance Below Crede Liner Landscapint A A-N Construction of initial embankments only 31 31-N .5 3:1 100 yeare X X 52 B2-N 1.5 3:1 100 years I %
B3 53-N 2.4 3:1 100 years I X C1 Cl-N .5 5:1 X X X C2-p 1.5 5:1 I I I C2 C 3-N 2.4 5:1 I I K C3 C4-N 3.4 5:1 1 1 I C4 4.3 5:1 I E I CS CS-N I I D1 D2-M 1.5 I D3 D3-N 2.4 3
- I D4-N 3.4 K 1 I D4 4.3 I I 3 D3 D5-N O
NUREG/BR-0184 C.94
Appendix C t(- l 1
Thble C.90 Summary of values for alternative disposal standards for uranium mill tailings (EPA 1983, 'Ihble S.2) f
(
l l Stab 111:ation Redue Control Water Protection !
Alte rna tive Chance of Tallions hasinus $1 sated bearha avoloed W Longevity 5tandarda Misuse Erosion Avoided of Lung Cancer first (years) (3 reduction) 100 1.000 Total (years) years years l
a ser, likely 0 2 to 10 (u) u O u 0 j s1 Lamely lwndred 1 in 10 (50) 300 1200 1200 100 82 inas Likely laundreds 4 in lu (80) 460 1600 1800 100 SJ inas Likely imadreds 1 in 10 (95) 570 ::100 2100 100 61 Limely hundred 1 la 10 (50) 300 300u Thousands 100 C2 1 na Lakely Thousands 4 in 10 (80) 480 46u0 Many 1000's 100's I C3 bb11 ely TWusanda 1 in 10 (95) 570 5700 Tens of 10uo's 10u0 l c4 \ery Lnlikely Many thousands 3 in 20'(9u.h) 590 5900 Tens of 1000's *10u0 C5 nery Onlikely Many thousanda 1 in 104(99.5) 600 6000 Teos of 2000's $1000 l
i D2 Lnlikely Thousands . to 10 3(60) 450 4600 Many 1000's 1000 93 t;nlikely Many thousands 1 in 103(95) 570 5700 Tens of 1000's 1000 ,
D4 tery unlikely Many thousanas 3 in 10*(v8.5) 590 5900 Tens of 1000's > 1000 D5 sery unlikely Many thousands 1 in 10*(99.5) 00u 6000 Tens of 1000's >1D00 l
(8)L11stlee ries of f atal cancer to an indivicual assumed to be living 600 uecers f rom the center of a model '
l tailless pile. The estimates of besetits assume no credit for engtueering f actors required to provide l
"reasonas,le moeuranea" of Jesign compliance for tne specified rados control level and petiod ol longevity.
(WD.ees estimates pertain to the control of 26 esistles piles and 9 pro tected new pile equivalents. Of the
. (j l
[8 \%
approstaately 6uo a.ath which ste estimated to occur in the first 100 yeare under no control conditions, about 390 are ene result of the . 1stins gettins and luu are due to future satitoss.
l l
l i
I l
l 1
l
.. ,f'%
C.95 NUREG/BR-0184
Appendix C O
bble C.91 Cost-effectiveness of control methods for uranium mill tallings (EPA 1983, hble 4.8)
Control Ef f ectiveness TegalCost Average lacremental Met hd inden (10 1983 $1 Cost Coat I million MT Esistint tile A 0 0 -- ---
51 1.0 4.2 Eliminated f rom consideration B2 1.8 4.9 Eliminated f rom consideration 33 3.1 9.2 Eliminated f ram consideration C1 4.3 3.2 .7 .7 C2 6.9 5.9 .9 1.0 C3 7. 9 8.3 1.1 2.4 C4 8.6 10.9 1.3 3.7 C5 9.2 13.3 1.6 4.0 7 million NT Existing Pile A 0 0 -- --
' 31 1.0 6.4 Eliminated f rom consideration B2 1.8 10.4 Eliminated from consideration 83 3.1 I4.0 Eliminated f rom consideration C1 4.3 6.3 1.5 1.5 C2 6.9 10.5 1.5 1.6 C3 7. 9 14.3 1.4 39 C4 8.6 18.5 2. 6.0 C5 9.2 22.2 1.4 6.2 22 million NT Esisting File A 0 0 -- -
81 1.0 10.8 10.8 10.8 32 1.8 17.3 Eliminated from eensideration 83 3.1 23.0 Eliminated f rom consideration Cl 4.3 13.6 3.2 0.8
) C2 6.9 20.6 3.0 2.7 C3 7.9 26.8 34 6.2 C4 8.6 33.8 3.9 10.0 C5 9.2 40.0 4.3 10.3 8.4 million W New Pile A n.0 1.3 -- --
pl 1.0 11.4 Eliminated frem consideration B2 1.8 15.0 Eliminated from consideration 91 3.1 .0 Eliminated f rom consideration C1 4.3 .4 *7 . 2.3 C2 6.9 .o '.3 1.8 D2 7.5 :.3 Fliminatad frna rangidararien C3 7.1 *0.9 *5. 4.n D3 8.3 35.5 F.liminated from consideration C ". 9.1 ' '. 3 2.8 6.1 D4 9.0 39.5 Fliminated f r om consi dar at ian C5 9.2 '9.4 3.1 4.R 0% ?.6 43.1 4.5 36.8 NUREG/BR-0184 9:
C.96 1
l
?
Appendix C A
/
D)
Rble C92 Summary of costs in millions of 1983 dollars for alternative disposal standards for uranium mill tallings (EPA 1983, hble S.3)
Alternative 4ee med Cover Induette Costs. 17mdiscounted present Worth Costs Standard Control Thickness Esisting f ut ure Total (tot discount rate)
Method (meters) Tallinas Taillmas a No control - 0 4 4 1 31 above-grade, 0.5 155 84-474 239-629 141-318 )
32 3:1 store, 1.5 253 96-549 351-802 219-424 i 33 irrigation and 2.4 338 !!4-632 452-970 288-524 maintenance for 100 years C1 above grade, 0.5 152 124-474 276-626 157-316 C2 5:1 slope, 1.5 253 145-570 398-823 240-433 1 314-537 l C3 rock cover on 2.4 343 165-653 508-996 C4 slopes 0.5 a 3.4 443 186-744 629 1187 387-65I C5 of pebbly sett 4.3 532 - 215-829 747-1361 474-755 on top of pile et Sane se C for 1.5 253 184-837 437-1090 249-546 D3 esisting piles 2.4 543 201-906 544-1249 323-644 D4 and staged 3.4 443 221-989 664-1432 406-755 D5 disposal 4.3 532 252-1065 784-1597 483-855 below-grade f or new piles
/%
hble C93 Estimated risks from spent fuel pool fires (Jo et al.1989, %ble 3.1)
Probability (vent PWR Plant BWR Plant Structural failure of Pool Resulting from Seismic Events 1. 8E -6/Py
- 6.7E-6/Ry Probability of a Cask Droo Caused by Human Error 3. !E -4/Ry 3.1E-4/Ry Reduction in Failure Rate for Cask l Drop implerrenting Generic issue A-36 1.0E-3 1.0E-3 j 1
Conditional Probability of Pool l Structural Failure Given a Cask Drop 1.0 1.0 Conditional Prebability of a Clad Fire Given a Pool Structural Failure ** 1.0 0.25 Freauency of $oent Fuel Pool Fire 1 from Setsmic initiator 1.EE-6/Ry 1.6PE-6/Ry )
Freque..cy cf $ cent Fuel Pool fire from a Cast Droo Initiator 3.lE-7/Py 7.75!-8/Py
- v = Reactor year. I
- mmEG/CR.4992 p. 75.
C.97 NUREG/BR-0184
Appendix C O
Dble C94 Offsite consequence calculations for spent fuel pool Ares (Jo et al.1989, Dble 3.2)
Offsite Pubile Proterty Health Dose Damage Case Characterisation $ource Ters' Population (person. rem) ($1983) 1 Average Case Last fuel discharged 340 persons / 7.97a108 3.41 10' 90 days after dis. mile 8 charge 2 Worst Case Entire pool inventory Zion population 2.56x10' 2.62 108 '
30 days after dis. (roughly 660 charge persons / mileI )
- From NUREG/CR.4982.
Dble C95 Onsite property damage costs in dollars per spent fuel pool accident (Jo et al.1989, hble 3.3)
Item test Estimate Worst case Cleanuo and Decontamination t . 'R8 1.65E8
- epair 7.2E7 7.ZE7 Replacement Power B.6?E8 1.66E9 Total Number of Doersting Years Remaining 29.8 years 29.8 years Number of Years Plant is Out of Service 5 years 7 years Esoected Collar Loss 8.24E9 1.29E10 0
NUREG/BR-0184 C.98
Appendix C m -.
I \
V Hble C.96 lucremental storage costs in 1983 dollars associated with limited low-density racking in the primary spent fuel pool (Jo et al.1989, %ble 3.6)
STORAGE PER UNIT ALL PL ANT $
OPflM ~0t* 51 10% Ot* 51 10%
P00L 2.17+7 1.67+7 1.28+7 2.34+9 1.80+9 1.38+9 DRTWELL 9.13+6 8.2a+6 6.25+6 9.86+8 8.90+8 7.40+8 VAULT 2.07+7 1.67+7 1.28+7 2.24+9 1. 80+ 9 1.38+9 l CASK 1.20+7 1.22+7 1.05+1 1.30+9 1.32+9 1.13+9 SILO 1.56+7 1.22+7 9.35+6 1.68+9 1.32+9 1.01+9
+Zero 1 discount rate corresponds to the case where additional storage capacity is built now.
I Notes! 1. These costs include the cost of in* pool feraCking and the j incremental costs associated with the change in additional ;
storage reeutrements resulting f rom the decrease in primary 1 pool capacity. I
- 2. Assuming the extra storage capacity is built when required, j two discount rates are applied.
I h
l C 99 NUREG/BR-0184
. . ~
Appendix C O
Dble C.97 Summary of Parameters affecting attributes for the spent fuel pool inventory reduction option (Jo et al.1989, Dble 3.8)
Factors Affecting Attributes Attributes Description Quantification References Public Health A. Pool Failure Probabiltty Seismic 5tructural failure Dose Reduction Taole 3.1 High PWR 1.8 a 10*' /Ry Ref. 2
- SWR 1.68 a 10-8 Low 0 Failure due to Cast Drop Htgn . PWR 3.1 a 10-7 /Ry Ref. 2
. SWR 7.75 a 10.a Low 0 Others 0
- 8. Number of Pools involved PWR 69 DOC /RL.87.!!
BWR 39 C. Average Remaining Life. PWR 29.8 DOE /RL 87 11 Time of Plant BWR 27.9 D. Radioactive Inventory Worst Case Total Inventory 30 days NUREG/CR 4982 Release After Discharge Best (stimate Last fuel Discharge 90 Days After Discharge E. Meteorology Zion F. Population Worst Case Zion (860 people /sq. al.)
U.S. Average 340 people /sq. at.
G. Risk Reduction 80% 5eouence Frequency 801 NUREG/CR.4982 Reduction Reduction of Considered to be insig.
Occusational nificant compared to Esposure Public Health Impact
.. Accidental Reduction of No significant change Occupational expected Esposure
.. Routine Factors Affecting Attributes Attributes Description Quantif testion References Offsite Property A,S.C,D,(.F,G $ame as those of Public Health Damage Ecomony lion Discount Rate 301 Onsite Property Decontamination. Refur. $ years NUREG/CR.3568 Damage bishment and Replace. [PRI NP.3380 ement Power Time.
Discount Rate 301 Leg. (fficiency unaffected improvement in Unaffected Knowledge Industry !sple. Additional $torage Migu (Pool Option) D0E/RL 8711 mentation and Option and Reracking Low (Drywell Option) [PRI NP 3365 Operation Cost.
Discount Rate 101 NRC Development Unaffected
/ Implementation /
Operation O
NUREG/BR-0184 C.100 J
Appendix C !
O t
( ,
Dhle C.98 Summary ofIndustry-wide value-hnpact analysis of the spent fuel poolinventory t reduction option (s) (Jo et al.1989, hble 3.9) -
- l. I t
Dose toduction (Person-tem) Evaluation (11983) l Best High Best High f
Attributes [ stimate Estimate (b) Estleste Estles'.e(b) f Pubitc Health 4.00 x 10* 1.28 x 10' 4.00 x 10' 1.28 a 10' Occupational Exposure
/ Accidental =0 =0 =0 =0
/ Routine =0 =0 =0 =0 Offsite Property 1.42 a 10' 2.22 x 10' Onsite Property 5.54 a 10' 4.25 a 10' Regulatory Efficiency Unaffected leprovement in Knowledge Unaffected Industry Inglementation and Operation -1.38 x 10' -1.13 s 10' NRC Development. laple-mentation and Operation Unaffected i
het Benefit ($) -1.33 x 10'ICI-9.57 x 10' h Benefit (1)/ Cast ($) Ratto 0.035 ICI 0.15 f
[D j g
Ratio of Pubite Dose Reduc-tion per Million Dollars 29.0(g) i
.g Cost (Person-res/310*) 113.0 ;
Cost of Irelementation per ,
Averted Person-res l (1/ Person-rem) 3.45x10*ICI 8.83a10' (al Based upon a U.S. pool population of 108.
(butigh estimate is based on the ' Worst Case' source term release and Zion (c) Based site population on 1988 (see Table dollars, the3.2).
Best Estimate het Benefit. Senefit/ Cost Retto, Person.
Public Dose rem would beReduction 7er Million
-1.47x10 Dollars. Dollars 0.032. 26.4 Cost and Cost Person-res andper Averte{ Deller/
3.79x10 Person-rem, respectively. Cost escalation during 1983-1988 was assumed to be 9.81 (Reference 17).
I C.101 NUREG/BR-0184
1 Appendix C Hble C.99 Failure frequency for generic spent fuel pool coollag and makeup systems (Jo et al.1989, hble 4.1) l Total Failure Failure Rates Per Demand frequency Cooling 5ystem Makevo 5rstem Per System System Type Description TTa'ta 1** Train 2 Train 1 Irain 2 Fire System fear A. Minimum 5AP I Requirement 0.1 0.05 0.015 0.05 -
3.8 a 10 8 I
- 8. Minimum SRP Requirement With Credit for Fire System 0.1 0.05 0.015 0.05 0.05 1.9 a 10 '
C. Old Esisting Plant with Both Cooling Pures Required 30% of Time tt 0.1 0.3 0.015 0.05 -
2.2 a 10. s D. Old faisting Plant With Credit for Fire System 0.1 0.3 0.015 0.05 0.05 1.1 a 10-'
' Reference 1.
- Units of failure per system year.
SRP = Standard Review Plan O
O NUREG/BR-0184 C.102
Appendix C o
hble C.100 Value-impact for generic imprwements to the spent fuel pool cooling system *
(Jo et al.1989, Able 4.2)
Improvement Espected Averted Senefit/
System Description Improvement Cost (19831) Cost (19831) Cost Ratto A. Mtntmum $RP 1. Adatttonal pump 50.000 None 0.0
- 2. Additional train 1.0t6 545 to 6640 <<0.01-
- 8. Minimum $RP 1. Additional pump 50.000 None 0.0 Requirement With Credit 2. Additional train 1.0t6 27 to 330 0.0 for Fire System C. Old Isisting 1. Additional pump 50.000 2500 to 30.s00 .05 to 0.61 Plant With Both Cooling 2. Additional train 1.0t6 3160 to 38.550 .003 to 0.04 Pumps Required 301 of Time 1
- 0. Old Esisting 1. Additional pump 50.000 125 to 1500 .0025 to 0.03 Plant With Credit for Fire 2. Additional train 1.016 159 to 1940 4.002 System
- 0uanttf tcation reflects a single spent fuel pool.
System A - Minimum cooling and makeup system reavired by the $RP 38 One full capacity cooling train with redundant active components (i.e., re.
dundant valves and puss). One Category 1 makeup system and one backup pum or system (not required to be Category 1) which can be aligned to a Category I water supply.
System 8 - Minimum cooling and makeup system with credit for makeup from fire system (Note that some plants may identify the fire system @s the backup in $ystem A).
System C - Typical older system comparable to current $RP requirements: One cooling train with backup active components (but backup components are reautred to supplement cooling about 30t of time 38); One safety grade makeup train and one non-safety grade makeup system.
System D - Typical older system ($ystem C) with third makeup train available (e.g.. fire system).
ibie C.101 Offsite property damage and health costs per spent fuel pool accident * (Jo et al.1989, Dble 5.1)
Use of Radiological Property Camage Source Term Population Spray System Cose (person rem) Costs $
Case Characterfastion Last fuel discharged 340 persons / No 7.97E6 3. ale 9 1 Average Case 90 days after discharge so. mile Average Case Last fuel discharged 340 persons / Yes 1.25[6 6.16E7 2
90 days after discharge sq. mile Worst Case Entire pool density 21on Population No 2.56E7 2.62t10 3 I 30 days after dischstge (roughly 860 i persons /sq. mile) l
!!on Population 6.78E6 a.48E8 I 4 Worst Case Enttre pool density Ves j
30 days after discnarge (roughly 860 persons /sq. mile) l
- MACC5 Calculations.
l
(
C.103 NUREG/BR-0184
Appendix C O
Thble C.102 Summary of industry-wide value-impact analysis of the spent fuel pool post-accident spray system" (Jo et al.1989, Thble 5.2)
Total Dose Reduction total Monetary Rist (Person-rem) Reduction ($1983)
Best Hf9h OS "I9" Attributes Estimate (b) Estimate (b) Estimate (b) Estimate (b)
Public Health 4.20E4 1.18E 5 4.20E 7 1.18E 8 Occupational Exposure =0 -0 -0 -0 0'fstte Property 6.77E 6 5.20E7 Onsite Property =0 0 Industry Iglementation -1.08E8 -1.08E8 and Operation Net Benefit ($) -5.92ET(C) g,gg7 Benefit ($)/ Cost (1) Ratio 0.45I 'I 1.57 Datto of Public Dose Peduc-tion per Million Dollars ,
Cost (Person-rem /110') 3.89E2 ICI 1.09E 3 Cost of Imolementation per Averted P:rson-rem (1/ Person-res) 2.57E3I 'I 9.15E2 (a) Population of 108 spent fuel pools.
(b)See Table 3.2 for source terms and demoraphic assump'ttons.
(c) Based on 1984 dollars. Best Estimate fut Benefit. Benefit / Cost Ratto.
Pubite Pose Reduction per Million Dollar Cost and Cost per Averted Person-rem would be -6.92E7 dollars, 0.42 354 Person-rem and 2.82E3 dollars /
person-res,respectively. Cost escalation during 1983-1988 was assumed to be 9.8% (Reference 17).
0 NUREG/BR-0184 C.104
s Appendix C i
b '
hble C.103 Facility descrigdors for accident analysis (Ayer et al.1988, hble 2.1) j a
Descriptor Accident Compartment Wall material Colling material
. Floor material Thickness of wall Thickness of ceiling Thickness of floor Length of room Width of room Height of room Volume of room
) Yessels in Accident Compartment
) Type of vessel (pressurized, unpressurized) i Construction material l Height of vessel Exposed width Elevation of vessel Wolght of empty vessel (or well thickness and density)
Fallure pressure Yentilation System Schematic fN Elevation of inlet duct to compartment Filter type Filter efficiency Blower performance curve Duct height Duct equivalent diameter Duct heat transfer area Duct floor area Duct length Duct X-sectional flow ares Duct Well properties outside emissivity Outside absorptivity Density Thermal conductivity Specific heat Thickness Volume of roass, cells, plenues Alternate Flow Paths Time of generation Elevation of path Size of opening (equivalent area circular dieneter)
Pressure on other side C.105 NUREG/BR-0184 i-
=- - - .. c. , o s -.
Appendix C O
Able C.104 Fuel inanufacturing process descriptors (Ayer et al.1988, hble 3.6)
Descriptor R:dioactive Material Inventories Form Containment Location Quantity Properties Radioactivity Radioactive Material in Containers Volume of Powder Moisture Content of Powder Volume of Air in Closed Containers Mass of Liquid Volume of Liquid Hazardous Material Inventories Location Quantity Surface Area Material Type Energy Process Parameters Initial Temperatures Compartment Radioactive Powders in Closed Containers Radioactive Liquids in Closed Containers Radioactive Liquids in Open Containers Outside of Vessels Duct Wall Initial Pressures in Inlet Duct Compartment Exit Duct O
NUREG/BR-0184 C.106
l Appendix C iO)
- (/
i 1hble C.105 Fuel reprocessing process descriptors (Ayer et al.1988, Thble 3.8) 1 Descriptor Radioactive Material Inventories I Fe,rm '
Location j j Containment ,
< Quantity l Properties Radioactivity Radioactivity Containment Radioactive Material in Containers i Volume of Powder Moisture Content of Powder ,
Volume of Air in Closed Containers !
Mass of Liquid I Volume of Liquid
)
,/
Hazardous Material Inventories Energy i
Location Quantity Surface Area Material Type Process Parameters Initial Temperatures Compartment Radioactbe Powders in Closed Containers Radioaca 3 Liqui.
- in Closed Containers 1 Radioactive Liquids in Open Containers Outside of Vessels Duct Wall Solvent Stream Initial Pressures in Inlet Duct Compartment Exit Duct Solvent-Stream i
C.107 NUREG/BR-0184
Appendix C O
Table C106 Waste storage / solidification process descriptors (Ayer et al.1988, Thble 3.10)
Descriptor Radioactive Material Inventories Form Containment Location Quantity Properties Radioactivity Radionuclide Volatility Radioactive Material in Containers Volume of Powder Moisture Content of Powder !
Volume of Air in Closed l Mass of Liquid i Volume of Liquid Containers i Hazardous Material Inventories l Location !
Quantity '
Surface Area Material Type Energy
]
Process Parameters Initial Temperatures Compartment ,
Radioactive Powders in Closed Containers i Radioactive Liquids in Closed Containers Radioactive Liquids in Open Containers Outside of Vessels Glass Surface Duct Wall Initial Pressures in Inlet Duct Compa-tment Exit Duct O
NUREG/BR 0184 C.108
l Appendix C nbie C.107 Spent fuel storage process descriptors upr et al.1988,1hble 3.11)
Descriptor Radioactive Material Inventories*
Form Containment Location.
Quantity i Properties Radioactivity l
Radioactive Material in Containers Volume of Air in Closed Containers
-Mass of Liquid Volume of Liquid Hazardous Material Inventories Location- ,
Quantity l l
Surface Area Material Type Energy l Process Parameters l Initial Temperatures Compartment Radioactive Powders in Closed Containers Radioactive Liquids in Closed Containers Radioactive Liquids in Open Containers Outside of Vessels
! Duct. Wall l
Initial Pressures in Inlet Duct Compartment Exit Duct ,
l l !
i r
C.109 NUREG/BR-0184
Appendix C O
%ble C108 Behavior mechanisms for airborne particle (Ayer et al.1988, hble 4.1)
Influencing Mechanism Description Elements Diffusion Movement of particles due to random gas Particle size molecular collisions and microscopic Temperature eddies in air Settling Effect of gravity upon airborne particles Particle size Turbulence Induced gas flow Coagulation The adherence of a particle'to another Number of upon collision to produce a par.ticle of particles larger size and, for solids, less dense Eddy velocity Particle size Condensation Particle Generation (condensation of Type of vapor vapors upon condensate nuclei), or Local particle growth (condensation of vapors temperature on existing particles) Particle size Agglomeration Same as coagulation (for colloids) and Number of coalescence (for liquids) particles Eddy velocity Particle size
' Scavenging The removal of airborne particles by Particle size materials falling through a fluid volume Diffusiophoresis Movement of particles caused by concen- Vapor condensa-tration gradients in the gas phase tion rate Thermophoresis Movement of particles down a tempera- Temperature ture gradient gradient O
NUREG/BR-0184 C.110
l
)
1 Appendix C j
[ i
\ 1
)
I hble C.109 Unscaled and scaled total accident risks to the public for non-reactor fuel cycle facilities l
Tbtal Accident Risk (person-rem /yr)
)
Scaled -
Fuel Cycle Element Unscaled (1/GWe)". % ble l Uranium Milling -
2.7E-4 C.70
]
UF. Conversion --
0.012 C.70 ;
Enrichment --
0.012 C.70 l
Fuel Fabrication --
0.0050 C.70 l MOX Fuel Refabrication --
0.12' C.70 l 0.036 C.70 3.3E-5 C.70 Fuel Reprocessing -
0.031 C.70 0.0032 C.70 5.6E-4 C.70 2.2E-4 C.70 1.5E-4 C.70 i 5.4E-5 C.70 Spent Fuel Storage --
0.18 C.70 0.031 C.70 3.7E-5 C.70 '
2.7E-5 C.70 5.7E-6 C.70 i Cask Storage 1.2*) -
C.32 l
Drywell Storage 8.5*) -
C.31 l 0.7*) --
C.32 Operations Phase 0.004*) --
C.44 HLW Storage -
2.3E-4 C.70 Geologic Waste Disposal Total Preclosure -
4.0E-5 C.70 Operations Phase 0.010 - C.14 1.5E-5 --
C.19 Without MRS 3 E-5 *) -- C.42 With MRS 3E-56) - C.44 Total Postclosure - 5.0E-11W --
Transportation Without MRS SS) - C.42 With MRS 10*) --
C.44 O
C.111 NUREG/BR-0184
l Appendix C O ;
1 Table C.109 (Continued)
Total Accident Risk (person-rem /yr)
Scaled Fuel Cycle Element Unscaled (1/GWe)" Table Plutonium Oxide Truck --
0.0013 C.70 Rail --
0.066 C.70 Spent Fuel Truck in 1975 240*) --
C.35 in 1985 1500*' -
C.35 Rail --
0.16 C.70 0.16 C.70 0.078 C.70 0.013 C.70 0.0093 C.70 7.1E-4 C.70 5.6E-5 C.70 8.4 E-6 C.70 8.4E-6 C.70 2.6E-6 C.70 in1975 110*) --
C.35 in 1985 4000*' --
C.35 HLW Rail --
0.042 C.70 0.0023 C.70 8.4E-4 C.70 2.8E-6 C.70 2.4E-6 C.70 (a) Measured in terms of the .anual requirements of a 1,000-MWe (1-GWe) LWR (b) Converted to person-rem /yr using 5,000 person-rem / health effect (c) From Erdmann et al. (1979), see Section C.6.
O NUREG/BR-0184 C.112
l Appendix C
\
hble C110 Preliminary occupational risk estimates for postulated accidents at a repository in tuff for preclosure operations phase of geologic waste disposal (see hbles C18 and C19) (Daling et al.1990)
Frequency Worker Dose Worker Risk Accident Scenario (1/yr) (person-rem) (person-rem /yr)
Natural Phenomena Flood 0.010 5.0E-10 5.0E-12 1 Earthquake < 0.0013 0.37 < 4.8E-4 Tornado < 9.lE-11 0.37 < 3.4E-11
)
Man-made Events l Aircraft Impact < 2.0E-10 5.5 < 1.lE-9 .
Nuclear Test < 0.0010 0.37 < 3.7E-4 !
l Operational Accidents Fuel Assembly Drop 0.10 0.0081 8. lE-4 Loading Dock Fire Spent Fuel < 1.0E-7 3.5 < 3.5E-7 HLW < l.0E-7 0.6 < 6.0E-8 Waste Handling Ramp Fire < l.0E-7 64 < 6.4E-6 V Emplacement Drift Fire < 1.0E-7 180 < 1.8E-5 i
Total .0017 l
l
\
C.113 NUREG/BR-0184
4 i
4 I
l i
Appendix D l
l Safety Goal Policy Statement and Backfit Rule l
s a
4 4
4 Z
J
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i t
4 e
d i
l
A Appendix D
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Appendix D
{
Safety Goal Pblicy Statement and Backat Rule 1
D.1 Safety Goals for the Operations of Nuclear Pbwer Plants (51 FR 30028; August 21,1986) 4
SUMMARY
- This policy statement focuses on the risks to the public from nuclear power plant operation. Its objective is to establish goals that broadly define an acceptable level of radiological risk. In developing the policy statement, the NRC sponsomd two public workshops during 1981, obtained public comments and held four public meetings during 1982, con-ducted a 2-year evaluation during 1983 to 1985, and received the views ofits Advisory Committee on Reactor Safeguards, i
The Commission has established two qualitative safety goals which are supported by two quantitative objectives. Rese 1
two supporting objectives are based on the principle that nuclear risks should not be a significant addition to other societal
, risks. The Committee wants to make clear that no death attributable to nuclear power plant operation will ever be
, " acceptable" in the sense that the Committee would regard it as a routine or permissible event. The Committee is discus- ,
O sing acceptable risks, not acceptable deaths. '
- The qualitative safety goals are as follows:
- Individual members of the public should be provided a level of protection from the consequences of nuclear power plant operation such that individuals bear no significant additional risk to life and health.
Societal risks to life and health from nuclear power plant operation should be comparable to or less than the risks of generating electricity by viable competing technologies and should not be a significant addition to other societal risks.
a e De following quantitative objectives are to be used in deterrnming achievement of the above safety goals:
The risk to an average individual in the vicinity of a nuclear power plant of prompt facilities that might result 4 from reactor accidents should not exceed one-tenth of one percent (0.1 percent) of the sum of prompt fatality l risks resulting from other accidents to which members of the U.S. population are generally exposed.
The risk to the population in the area near a nuclear power plant of cancer fatalities that might result from nuclear power plant operation should not exceed one-tenth of one percent (0.1 percent) of the sum of cancer fatality risks resulting from a!! other causes.
EFFECTIVE DATE: August 4,1986.
I SUPPLEMENTARY INFORMATION: The following presents the Commission's Final Policy Statement on Safety Goals for the Operation of Nuclear Power Plants:
4 4 7%
D.1 NUREG/BR-0184
Appendix D O
- 1. Introduction A. Purpose and Scope In its response to the recommendations of the President's Commission on the Accident at three Mile Island, the Nacicar Regulatory Commission (NRC) stated that it was " prepared to move forward with an explicit policy statement on safety philosophy and the role of safety-cost tradeoffs in the NRC safety decisions." This policy statement is the result.
Cunent regulatory practices are believed to ensure that the basic statutory requirement, adequate protection of the public, is met. Nevertheless, current practices could be improved to provide a better means for testing the adequacy of and need for current and proposed regulatory requirements. The Commission believes that such improvement could lead to a more coherent and consistent regulation of nuclear power plants, a more predictable regulatory process, a public understanding of the regulatory criteria that the NRC applies, and public confidence in the safety of operating plants. His statement of NRC safety policy expresses the Commission's views on the level of risks to public health and safety that the industry should strive for in its nuclear power plant.
This policy statement focuses on the risks to the public from nuclear power plant operation. Rese are the risks from release of radioactive materials from the reactor to the environment from normal operations as well as from accidents.
The Commission will refer to these risks as the risks of nuclear power plant operation. The risks from the nuclear fuel cycle are not included in the safety goals.
These fuel cycle risks have been considered in their own right and determined to be quite small. They will continue to receive careful consideration. The possible effects of sabotage or diversion of nuclear material are also not presently included in the safety goals. At present there is no basis on which to provide a measure of risk on these matters. It is the Commission's intention that everything that is needed will be done to keep these types of risks at thcir present very low level; and it is the Commission's expectation that efforts on this point will continue to be successful. With these excep-tions, it is the Commission's intent that the risks from all the various initiating mechanisms be taken into account to the best of the capability of current evaluation techniques.
In the evaluation of nuclear power plant operation, the staff considers several types of releases. Current NRC practice addresses the risks to the public resulting from operating nuclear power plants. Before a nuclear power plant is licensed to operate, NRC prepares an environmental impact assessment which includes an evaluation of the radiological impacts of routine operation of the plant and accidents on the population in the region around the plant site. The assessment under-goes public comment and may be extensively probed in adjudicatory hearings. For all plants licensed to operate, NRC has found that there will be no measurable radiological impact on any member of the public from routine operation of the plant. (
Reference:
NRC staff calculation of radiological impact on humans contained in Final Environmental Statements for specific nuclear power plants: e.g., NUREG-0779, NUREG-0812, and NUREG-0854.)
The objective of the Commission's policy statement is to establish goals that broadly define an acceptable level of radio-logical risk that might be imposed on the public as a result of nuclear power plant operation. While this policy statement includes the risks of normal operation, as well as accidents, the Commission believes that because of compliance with Federal Radiation Council (FRC) guidance, (40 CFR Part 190), and NRC's regulations (10 CFR Part 20 and Appendix 1 to Part 50), the risks from routine emissions are small compared to the safety goals. Therefore, the Commission believes that these risks need not be routinely analyzed on a case-by-case basis in order to demonstrate conformance with the safety goals.
l l
NUREG/BR-0184 D.2 9 !
l 1
Appendix D A
V) i B. Development of this Statement of Safety Policy In developing the policy statement, the Commission solicited and benefited from the information and suggestions provided by workshop discussions. NRC-sponsored workshops were held in Palo Alto, California, on April 1-3,1981 and in Harpers Ferry, West Virginia, on July 23-24,1981. The first workshop addressed general issues invohed in developing safety goals. The second workshop focused on a discussion paper which presented proposed safety goals. Both work-sheps featured discussions among knowledgeable persons drawn from industry, public interest groups, universities, and elsewhere, who represented a broad range of perspectives and disciplines.
The NRC Office of Policy Evaluation submitted to the Commission for its consideration a Discussion Paper on Safety Goals for Nuclear Power Plants in November 1981 and a revised safety goal report in July 1982.
The Commission also took into consideration the comments and suggestions received from the public in response to the proposed Policy Statement on " Safety Goals for Nuclear Power Plants," published on February 17,1982 (47 FR 7023).
Following public comment, a revised Policy Statement was issued on march 14,1983 (48 FR 10772) and a 2-year evaluation period began.
The Commission used the staff report and its recommendations that resulted from the 2-year evaluation of safety goals in developing this 'inal Policy Statement. Additionally, the Commission had benefit of further comments from its Advisory Committee on R tetor Safeguards (ACRS) and by senior NRC management.
Based on the results of this information, the Commission has determined that the qualitative safety goals will remain
, p unchanged from its March 1983 revised policy statement and the Commission adopts these as its safety goals for the i
operation of nuclear power plants.
II. Qualitative Safety Goals The Commission has decided to adopt qualitative safety goals that are supported by quantitative health effects objectives
, for use in the regulatory decisionmaking process. The Commission's first quantitative safety goal is that risk from nuclear l power plant operation should not be a significant contributor to a person's risk to accidental death or injury. The intent is to require such a level of safety that individuals living or working near nuclear pour plants should be able to go about their daily lives without special concern by virtue of their proximity to these plants. Thus, the Commission's first safety j goal is -
l Ir.dividual members of the public should be pnwided a level ofprotection from the consequences of nuclear powrplant l operation such that individuals bear no sigmpcant additional risk to life and health.
l
! Even though protection of individual members of the public inherently provides substantial societal protection, the Com-mission also decided that a limit should be placed on the societal risks posed by nuclear power plant operation. The Com-i mission also believes that the risks of nuclear power plant operation should be comparable to or less than the risks from I other viable means of generating the same quantity of electrical energy. Thus, the Commission's second safety goal is -
l Societal risk to life and health from nuclear power plant operation should be compamble to or less than the risks ofgener-ating electricity by nable competing technologies and should not be a sigmpcant addition to other societal risks.
The broad spectmm cf expert opinion on the risks posed by electrical generation by coal and the absence of authoritative data make it impractical to calibrate nuclear safety goals by comparing them with coal risks based on what we know today.
However, the Commission has established the quantitative health effects objectives in such a way that nuclear risks are not a signincant addition to other societal risks.
v D.3 NUREG/BR Ol84 l
t l
t
Appendix D Sevem core damage accidents can lead to more serious accidents with the potential for life-threatening offsite release of radiation, for evacuation of members of the public, and for contammation of public property. Apart from their health and l safety consequences, severe core damage accidents can erode public confidence in the safety of nuclear power and can lead to further instability and unpmdictability for the industry. In order to avoid these adverse consequences, the Commission intends to continue to pursue a regulatory program that has as its objective providing reasonable assurance, while giving appropriate consideration to the uncertainties involved, that a severe core damage accident will not occur at a U.S. nuclear power plant.
III. Quantitative Objectives Used to Gauge Achievement of The Safety Goals A. General Considerations The quantitative health effects objectives establish NRC guidance for public protection which nuclear plant designers and operators should strive to achieve. A key element in formulating a qualitative safety goal whose achievement is measured by quantitative health effects objectives is to understand both the strengths and limitations of the techniques by which one judges whether the qualitative safety goal has been met.
A major step forward in the development and refinement of accident risk quantification was taken in the Reactor Safety Study (WASH-1400) completed in 1975. The objective of the Study was "to try to reach some meaningful conclusions about the risk of nuclear accidents." The Study did not directly address the question of what level of risk from nuclear accidents was acceptable.
Since the completion of the Reactor Safety Study, further progress in developing probabilistic risk assessment and in accu-mulating relevant data has led to a recognition that it is feasible to begin to use quantitative safety objectives for limited purposes. However, because of the sizable uncertainties still present in the methods and the gaps in the data base-essential elements needed to gauge whether the objectives have been achieved--the quantitative objectives should be viewed as aiming points or numerical benchmarks of performance. In particular, because of the present limitations in the state of the art of quantitatively estimating risks, the quantitative health effects objectives are not a substitute for existing regulations.
The Commission recognizes the importance of mitigating the consequences of a core-melt asident and continues to emphasize features such as containment, siting in less populated areas, and emergency planning as integral parts of the defense-in-depth concept associated with its accident prevention and mitigation philosophy.
B. Quantitative Risk Objectives The Commission wants to make clear at the beginning of this section that no death attributable to nuclear power plant operation will ever be " acceptable" in the sense that the Commission would regard it as a routine or permissible event.
We are discussing acceptable risks, not acceptable deaths. In any fatal accident, a course of conduct posing an acceptable risk at one moment results in an unacceptable death moments later. This is tme whether one speaks of driving, swim-ming, flying, or generating electricity from coal. Each of these activities poses a calculable risk to society and to individu-als. Some of those who accept the risk (or are part of a society that accepts risk) do not survive it. We intend that no such accidents will occur, but the possibility cannot be entirely eliminated. Furthermore, individual and societal risks from nuclear power plants are generally estimated to be considerably less than the risk that society is now exposed to from each of the other activities mentioned above.
O NUREG/BR4)184 D.4
4 l
Appendix D
\ /
v C. Health Effects-Prompt and Latent Cancer Mortality Risks 4 1
1 he Commission has decided to adopt the following two health effects as the quantitative objectives concerning mortality risks to be used in de'ermining achievement of the qualitative safety goals -
De risk to an average individ d in the vicinity of a nuclearpowerplant ofpromptfatalities that might resultfrom reactor accidents should not exceed o 'e-tenth of onepercent (0.1 percent) of the sum ofpromptfatality risks resultingfrom other accidents to which members of the U.S. population are generally exposed.
The risk to the population the area near a nuclearpowerplant of cancerfatalities that might resultfrom nuclearpower plant operation should not exceed one-tenth of one percent (0.1 percent) of the sum of cancerfatality risks resultingfrom all other causes.
The Commission believes that this ratio of 0.1 percent appropriately reflects both of the qualitative goals-to provide that individuals and society bear no significant additional risk. However, this does not necessarily mean that an additional risk that exceeds 0.1 percent would by itself constitute a significant additional risk. The 0.1 percent ratio to other risks is low enough to support an expectation that people living or working near nuclear power plants would have no special concern due to the plant's proximity.
The average individual in the vicinity of the plant is defined as the average individual biologically (in terms of age and other risk factors) and locationally who resides within a mile from the plant site boundary. This means that the average individual is found by accumulating the estimated individual risks and dividing by the number of individuals residing in the
,Di vicinity of the plant.
In applying the objective for individual risk of prompt fatality, the Commission has defined the vicinity as the area within one (1) mile of the nuclear power plant site boundary, since calculations of the consequences of major reactor accidents suggest that individuals within a mile of the plant site boundary would generally be subject to the greatest risk of prompt death attnbutable to radiological causes. If there are no individuals residing within a mile of the plant boundary, an indi-vidual should, for evaluation purposes, be assumed to reside one (1) mile from the site boundary.
In applying the 6ojective for cancer fatalities as a population guideline for individuals in the area near the plant, the Commission his defined the population generally considered subject to significant risk as the population within ten (10) miles < f the plant site. The bulk of significant exposures of the population to radiation would be concentrated within this disttuce, and thus this is the appropriate population for comparison with cancer fatality risks from all ot!.er causes. This objective would ensure that the estimated increase in the risk of delayed cancer fatalities from all poteatial radiation releases at a typical plant would be no more than a small fraction of the year-to-year normal variation in the l expected cancer deaths from nonnuclear causes. Moreover, the prompt fatality objective for protecting individuals gener-ally provides even greater protection to the population as a whole. That is, if the quantitative objective for prompt fatality is met for individuals in the immediate vicinity of the plant, the estimated risk of delayed cancer fatality to persons within ten (10) miles of the plant and beyond would generally be much lower than the quantitative objective for cancer fatality.
Rus, compliance with the prompt fatality objective applied to individuals close to the plant would generally mean that the aggregate estimated societal risk would be a number of times lower than it would be if compliance with just the objective applied to the population as a whole were involved. The distance foe averaging the cancer fatality risk was taken as 50 miles in the 1983 policy statement. The change to ten (10) miles could be viewed to provide additional protection to O 1 V
D.5 NUREG/BR-0184
Appendix D O
individuals in the vicinity of the plant, although analyses indicate that this objective for cancer fatality will not be the controlling one. It also provides more representative societal protection, since the risk to the people beyond ten (10) miles will be less than the risk to the people within ten (10) miles.
IV. Treatment of Uncertalaties i ne Commission is aware that uncertainties are not caused by use of quantitative methodology in decisionmakmg but are merely highlighted through use of the quantification process. Confidence in the use of pmbabilistic and risk assessment techniques has steadily improved since the time these were used in the Reactor Safety Study. In fact, through use of quan-titative techniques, important uncertainties have been and continue to be brought into better focus and may even be 4
reduced compared to those that would remain with sole reliance on deterministic decisionmaking. To the extent practica-ble, the Commission intends to ensure that the quantitative techniques used for regulatory decisionmakmg take into account the potential uncertainties that exist so that an estimate can be made on the confidence level to be ascribed to the quantita- j tive results.
The Commission has adopted the use of mean estimates for purposes of implementing the quantitative objectives of this j safety goal policy (i.e., the mortality risk objectives). Use of the mean estimates comports with the customary practices 1 for cost-benefit analyses and it is the correct usage for purposes of the mortality risk comparisons. Use of mean estimated does not however resolve the need to quantify (to the eunt reasonable) and understand those important uncertainties involved in the reactor accident risk predictions. A number of uncertainties (e.g., thermal-hydraulic assumptions and the l phenomenology of core-melt progression, fission product release and transport, and containment loads and performance) arise because of a direct lack of severe accident experience or knowledge of accident phenomenology along with data related to probability distributions.
In such a situation, it is necessary that proper attention be given not only to the range of uncertainty surtounding probabil- {
istic estimates, but also to the phenomenology that most influences the uncertainties. For this reason, sensitivity studies l should be performed to determine those uncertainties most important to the probabilistic estimate. The results of sensi- l tivity of studies should be displayed showing, for example, the range of variation together with the underlying science or i engineering assumptions that dominate this variation. Depending on the decision needs, the probabilistic results should also be reasonably balanced and supported through use of deterministic arguments. In this way, judgements can be made by the decisionmaker about the degree of confidence to be given to these estimates and assumptions. His is a key part of the process of determining the degree of regulatory conservatism that may be warranted for particular decisions. His defense-in-depth approach is expected to continue to ensure the protection of public health and safety. ,
I V. Guidelines for Regulatory Implementation The Commission approves use of the qualitative safety goals, including use of the quantitative health effects objectives in .
, the regulatory decisionmakmg process. The Commission recognizes that the safety goal can provide a useful tool by l which the adequacy of regulations or regulatory decisions regarding changes to the regulations can be judged. Likewise, the safety goals could be of benefit in the much more difficult task of assessing whether existing plants, designed, con-structed and operated to comply with past and current regulations, conform adequately with the intent of the safety goal policy.
However, in order to do this, the staff will require specific guidelines to use as a basis foi determining whether a level of safety ascribed to a plant is consistent with the safety goal policy. As a separate matter, the Commission intends to review and approve guidance to the staff regarding such determinations. It is currently envisioned that this guidance would address matters such as plant performance guidelines, indicators for operational performance, and guidelines for conduct O
NUREG/BR-0184 D.6
Appendix D In\
V of cost-beneSt analyses. This guidance would be derived from additional studies conducted by the staff and resulting in recommendations to the Commission. The guidance would be based on the following general performance guideline which is proposed by the commission for further staff exanunation -
Consistent with the tmditional defense-in-depth approach and the accident mitigation philosophy requiring reliable per-formance of con!ainment systems, the overall meanfrequency of a large release of radioactive materials to the environment from a reactor accident should be less than 1 in 1,000,000per year of reactor operation.
To provide adequate protection of the public health and safety, current NRC regulations require conservatism in design, construction, testing, operation, and maintenance of nuclear power plants. A defense-in-depth approach has been man-dated in order to prevent accidents from happening and to mitigate their consequences. Siting in less populated areas is emphasized. Furthermore, emergency response capabilities are mandated to provide additional defense-in-depth protection to the surrounding population.
These safety goals and these implementation guidelines ase not meant as a substitute for NRC's regulations and do not relieve nuclear power plant permittees and licensees from complying with regulations. Nor are the safety goals and these implementation guidelines in and of themselves meant to serve as a sole basis for licensing decisions. However, if pursu-act to these guidelines, information is developed that is applicable to a particular licensing decision, it may be considered as one factor in the licensing decision.
The additional views of Commissioner Asselstine and the separate views of Commissioner Bernthal are attached.
p Dated at Washington, D.C., this 30th day of July 1986.
\ l V For the Nuclear Regulatory Commission. Lando W. Zech, Jr., Chairman.
Additional Views by Commissioner Asselstine on the Safety Goals Pblicy Statement The commercial nuclear power industry started rather slowly and cautiously in the early 1960's. By the late 1960's and early 1970's, the growth of the industry reached a feverish pace. New orders were coming in for regulatory review on almost a weekly basis. The result was the designs of the plants outpaced operational experience and the development of safety standards. As experience was gained in operational characteristics and in safety reviews, safety standards were developed or modified with a general trend toward stricter requirements. Thus, in the early 1970's, the industry demanded to know how safe is safe enough." In this Safety Goal Policy Statement, the Commission is reaching a first attempt at answering the question. Much credit should go to Chairman Palladino's efforts over the past five (5) years to develop this policy statement. I approve this policy statement but believe it needs to go further. There are four additional aspects which should have been addressed by the policy statement.
Containment Performance First, I believe the Commission should have developed a policy on the relative emphasis to be given to accident prevention and accident mitigation. Such guidance is necessary to ensure that the principle of defense-in-depth is maintained. The Commission's Advisory Committee on Reactor Safeguards has repeatedly urged the Commission to do so. As a step in that direction, I offered for Commission consideration the following containment performance criterion:
In onier to assure a proper balance between accident prevention and accident mitigation, the mean frequency of contain-l mentfailure in the event of a severe core damage accident should be less than 1 in 100 severe core damage accidents.
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Appendix D O
Since the Chernobyl accident, the nuclear industry has been trying to distance itself from the Chernobyl accident on the basis of the expected performance of the containments around the U.S. power reactors. Unfortunately, the industry and the Commission are unwilling to commit to a level of performance for the containments.
The argument has been made that we do not know how to develop contamment performance criteria (accident mitigation) because core meltdown phenomena and containment response thereto are very complex and involve substantial uncertain-ties. On the other hand, to measure how close a plant comes to the quantitative guidelines contained in this policy state-ment and to perform analyses required by the Commission's backfit rule, one must perform just those kinds of analyses. I find these positions inconsistent.
The other argument against a containment performance criterion is that such a standard would overspecify the safety goal.
However, a containment performance objective is an element of ensuring that the principle of defense-in-depth is main-tained. Since we cannot rule out core meltdown accidents in the foreseeable future, given the current level of safety, I believe it unwise not to establish an expectation on the performance of the final barrier to a substantial release of radioac-tive materials to the environment, given a core meltdown.
)
General Performance Guideline While I have previously supported an objective of reducing the risks to an as low as reasonably achievable level, the gen-eral performance guideline articulated in this policy (i.e., "...the overall mean frequency of a large release of radioactive materials to the environment from a reactor accident should be less than 1 in 1,000,000 per year of reactor operation") is a suitable compromise. I believe it is an objective that is consistent with the recommendations of the Commission's chief safety officer and our Director of Research, and past urgings of the Advisory Committee on Reactor Safeguards. Unfortu-nately, the Commission stopped short of adopting this guideline as a performance objective in the policy statement, but I am encouraged that the Commission is willing at least to enmme the possibility of adopting it. Achieving such a standant coupled with the containment performance objective given above would go a long way toward ensuring that the operating reactors successfully complete their useful lives and that the nuclear option remains a viable component of the nation's energy mix.
In addition to preferring adoption of this standard now, I also believe the Commission needs to define a "large release" of radioactive materials. I would have defined it as "a release that would result in a whole body dose of 5 rem to an indi-vidual located at the site boundary." This would be consistent with the EPA's emergency planning Protective Action Guidelines and with the level proposed by the NRC staff for defining an Extraordinary Nuclear Occurrence under the Price-Anderson Act. In adopting such a definition, the Commission would be saying that its objective is to ensure that there is no more than a 1 in 1,000,000 chance per year that the public would have been to be evacuated from the vicinity of a nuclear reactor and that the waiver of defenses provisions of the Price-Anderson Act would be invoked. I believe this to be an appropriate objective in ensuring that there is no undue risk to the public health and safety associated with nuclear power.
Cost-Benefit Analyses
, I believe it is long overdue for the Commission to decide the appropriate way to conduct cost-benefit analyses. The Com-l mission's own regulations require these analyses, which play a substantial role in the decisionmaking on whether to l improve safety. Yet, the commission continues to postpone addressing this fundamental issue.
Future Reactors in my view, this safety goal policy statement has been developed with a steady eye on the apparent level of safety already achieved by most of operating reactors. That level has been arrived at by a piecemeal approach to designing, constructing NUREG/BR-0184 D.8
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and upgrading of the plants over the years as experience was gained with the plants and as the results of required research became available. Given the performance of the current generation of plants. I believe a safety goal for these plants is not good enough for the future. This policy statement should have had a separate goal that would require substantially better
- plants for the next generation. To argue that the level of safety achieved by plant designs that are over 10 years old is good enough for the next generation is to have little faith in the ingenuity of engineers and in the potential for nuclear tech-nology. I wuld have required the next generation of plants to be substantially safer than the currently operating plants.
Separate Views of Commissioner Bernthal on Safety Goals Pblicy I do not disapprove of what has been said in this policy statement, but too much remains unsaid. The public is under-
- standably desimus of reassurance since Chernobyl
- the NRC staff needs clear guidance to carry out its responsibilities to assure public health and safety; the nuclear industry needs to plan for the future. All want and deserve to see clear, unam-biguous, practical safety objectives that provide the Commission's answer to the question, "How safe is safe enough?" at U.S. nuclear power plants. The question remains unanswered.
It is unrealistic for the Commission to expect that society, for the foreseeable future, willjudge nuclear power by the same standard as it does all other risks. The issue today is not so much calculated risk; the issue is public acceptance and, i consistent with the intent of Congress, preservation of the nuclear option.
In these early decades of nuclear power, TMI-style incidents must be rendered so rare that we would expect to recount such an event only to our grandchildren. For today's population of reactors, that implies a probability for severe core l l damage of 104 per reactor year; for the longer term, it implies something better. I see this as a straightforward policy conclusion that every newspaper editor in the country understands only too well. If the Commission fails to set (and realize) this objective, then the nuclear option will cease to credible before the end of the century. In other words, if '
TMI-style events were to occur with 10-15 year regularity, public acceptance of nuclear power would almost certainly fail.
1 And while the Commission's primary charge is to protect public health and safety, it is also the clear intent of Congress that the Commission, if possible, regulate in a way that preserves rather than jeopardizes the nuclear option. So, for example, if the Commission were to fmd 100 percent confidence in some impervious containment design, but ignored what was inside the contamment, the primary mandate would be satisfied, but in all likelihood, the second would not. Con-sistent with the Commission's long-standing defense-in-depth philosophy, both core-melt and containment performance cri-teria should therefore be cicarly stated parts of the Commission's safety goals.
4
, in short, this pudding lacks a theme. Meaningful assumnce to the public; substantive guidance to the NRC staff; the regu-i latory path to the future for the industry-all these should be provided by plainly stating that, consistent with the Commis-sion's " defense-in-depth" philosophy:
, (1) Severe core-damage accidents should not be expected, on average, to occur in the U.S. more than once in 100 years:
(2) Containment performance at nuclear power plants should be such that severe accidents with substantial offsite damages are not expected, on average, to occur in the U.S. more than one in 1,000 years:
4 i G D.9 NUREG/BR-0184 i
Appendix D O
(3) The goal for offsite consequences should be expected to be met after conservative considerations of the uncertainties associated with the estimated frequency of severe core-damage and the estimated mitigation thereof by containment."
The term " substantial offsite damages" would correspond to the Commission's legal definition of "extraonlinary nuclear occurrence." " Conservative consideration of associated uncertainties" should offer at least 90 percent confidence (typical good engineering judgment, I would hope) that the offsite release goal is met.
The bmad core-melt and offsite-release goals should be met "for the average power plant"; i.e., for the aggregate of U.S.
power plants. The decision to fix or not to fix a specific plant would then depend on achieving "the goal for offsite conse-quences." As a practical matter, this offsite societal risk objective would (and should) be significantly dependent on site-specific population density.
The absence of such explicit population density considerations in the Commission's 0.1 percent goals for offsite conse-quences deserves careful thought. Is it reasonable that Zion and Palo Verde, for example, be assigned the same theoretical
" standard person" risk, even though they pose considerably different risks for the U.S. population as a whole? As they stand, these 0.1 percent goals do not explicitly include population density considerations; a power plant could be located in Central Park and still meet the Commission's quantitative offsite release standard.
I believe the Commission's standards should preserve the impar ent principle that the site-specific population density be quantitatively considered in formulating the Commission's sch nl risk objective; e.g., by requiring that for the entire U.S. population, the risk of fatal injury as a consequence of the U.S. nuclear power plant operations should not exceed some appropriate specified fraction of the sum of the expected risk of fatality form all other hazards to which members of the U.S. population are generally exposed.
I am further concerned by the arbitrary nature of the 0.1 percent incremental " societal" health risk standard adopted by the Commission, a concept grounded in a purely subjective assessment of what the public might accept. The Commission should seriously consider a more rational standard, tied statistically to the average variations in natural exposure to radia-tion from all other sources.
Finally, as noted in its intmductory comments, the Commission long ago committed to " move forward with an explicit policy statement on safety philosophy and the role of saft:y cost tradeoffs in NRC safety decisions." While this policy statement may not be very " explicit", as discussed above, it contains nothing at all on the sr.bject of "' safety-cost' tradeoffs in NRC safety decisions." For example, is $1,000 per person-rem an appropriate cost-benefit standard for NRC regula-tor l, action? While I have long argued that such fundamental decisions are more rightly the responsibility of Congress, the NRC staff continues to use its ad-hoc judgment in lieu of either the Commission or the Congress speakmg to the issue.
In summary, while the Commission has produced a document which is not in conflict with my broad philosophy in such matters, I doubt that the public expected a philosophical dissertation, however erudite. It is a tribute to Chairman Palladino's efforts that the Commission has come this far. But the task remains unfinished.
(a) Interestingly enough the Commission has adopted proposed goals similar to the above core-melt and containment performance objectives-without clearly saying so. Taken together, the Commission's: (1) 0.1 percent offsite prompt fatality goals: (2) proposed 10dper-reactor-year "large ofisite release
- criterion: (3) commitment "to provide reasonable assurance...that a severe core-damage accident will not occur at a U.S. nuclear power plant" though they may be ill-defined. can be read to be more stringent than the plainly stated criteria suggested above.
NUREG/BR-0184 D.10
Appendix D i
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D.2 Backfit Rule (10 CFR 50.109)
(a)(1) Backfitting is defined as the modification of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedu:es or organization required to design, con- l struct or operate a facility; any of which may result from a new or amended provision in the Commission rules or the imposition of a regulatory staff position interpreting the Commission rules that is either new or different from a previously applicable staff position after:
(i) The date of issuance of the construction per:rli fr tne facility for facilities having construction permits issued after l October 21,1985; or (ii) Six months before the date of docketing of the operating license application for the facility for facilities having construction permits issued before October 21,1985; or (iii) The date of issuance of the operating license for the facility for facilities having operating license; or (iv) The date of issuance of the design approval under appendix M, N, or O of part $2. l (2) Except as provided in paragraph (a)(4) of this section, the Commission shall require a systematic and documented analysis pursuant to paragraph (c) of this section for backfits which it seeks to impose.
(3) Except as provided in paragraph (a)(4) of this section, the Commission shall require the backfitting of a facility only I
[mT when it determines, based on the analysis described in paragraph (c) of this section, that there is a substantial increase in l
\x / the overall protection of the public health ard safety or the common defence and security to be derived from the backfit and that the direct and indirect costs if implementation for that facility are justined in view of this increased protection.
(4) The provisions of paragraphs (a)(2) and (a)(3) of this section are inapplicable and, therefore, backfit analysis is not required and the standards in paragraph (a)(3) of this section do not apply where the Commission or staff, as appropriate, finds and declares, with appropriated documented evaluation for its finding, either:
1 (i) That a modi 6 cation is necessary to bring a facility into compliance with license or the rules or orders of the Commission, or into conformance with written commitments by the licensee; or (ii) That regulatory action is necessary to ensure that the facility provides adequate protection to the health and safety of the public and is in accord with the common defense and security; or (iii) That the regulatory action involves defining or redefining what level of protection to me public health and safety or common defense and security should be regarded as adequate.
(M he Commission shall always require the backfitting of a facility if it determines that such regu'.atory action is wmary to ensure that the facility provides adequate protection to the health and safety or the common defense and PN'# ;y.
.: the document evalusco equired by paragraph (a)(4) of this section shall include a statement of the objectives of and
,casons for the modiScation and the basis for invoking the exception. If immediately effective regulatory action is S respJred, then the documented evaluation trmy follow rather than precede the regulatory action.
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Appendix D O
(7) If there are two or more ways to achieve compliance with a license or the rules or oniers of the Commission, or with written licensee commitments, or there are two or more ways to reach a level of protection which is adequate, then ordi-narily the applicant or licensee is free to choose the way which best suits its purposes. However, should it be necessary or appropriate for the Commission to prescribe a specific way to comply with its requirements or to achieve adequate protec-tion, then cost may N a factor in selecting the way, provided that the objective of compliance or adequate protection is met.
(b) Paragraph (a)(3) of the section shall not apply to backfits imposed prior to October 21,1985.
(c) In reaching the determination required by paragraph (a)(3) of this section, the Commission will consider how the backfit should be scheduling light of other ongoing regulatory activities at the facility and, in addition, will consider information available concerning any of the following factors as may be appropriate and any other information relevant and material to proposed backfit:
(1) Statement of the specific objectives that the proposed backfit is designed to achieve; (2) General description of the activity that would be required by the licensee or applicant in order to complete the backfit; (3) Potential change in the risk to the public from accidental off-site release of radioactive material; (4) Potential impact on radiclogical exposure of facility employees; (5) Installation and continuing costs associated with the backfit, including the cost of facility downtime or the cost of construction delay; (6) The potential safety impact of changes in plant or operational complexity, including the relationship to proposed and existing regulatory requirements; (7) The estimated resourte burden on the NRC associated with the pmposed backfit and the availability of such resources; (8) The potential impact or differences in facility type, design or age on the relevancy and practicality of the proposed backfit; (9) Whether the proposed backfit is interim or final and, if interim, the justification for imposing the proposed backfit on an interim basis.
(d) No licensing action will be withheld during the pendency of backfit analyses required by the commissions mies. l (e) The Executive Director for Operations shall be responsible for implementation of this section, and all analyses required by this section shall be approved by the Executive Director for Operations or his designee.
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[54 FR 20610, kne 6,1988, as amended 54 FR 15398, AprS 18,1989]
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Appendix E I
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1 A i Accident Frequency . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.6 Section ' )
Accidents -
non-reactor frequency . . . . . . . . . . . . . . . . . . ....................................... .... C.2.1.1 ;
population dose factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . C2.1.2 .
radiologicat risk ranking . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - C 6 I reactor -
frequency . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 . 6 i population dose factors . . . . ............................ ................... 5.7.1.1 'l Agreernent States . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1, 5.7.11 [
Antitrust ...............................................................5.5.15,5.7.15 l Attributes l
\ algebraic signs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2 !
best estirnate/ expected value . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7 ;
identification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.3, 5.5 ,
B ;
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Backfit $
i de finition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2.1 ;
regulatory analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . 2.2, 4.4 l Best estimates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.3 i C
Chernobyl ................., .............. ...... .......................... 5.7.3.1 Classification of facilities fuel cycle . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C 1.1 non-fuel eycle . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C 1.2 Cleanup of materials licensee contamination incidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C.3, C.4 Computer codes ALLDOS....................................................................C6 CAP-88.................................................................5.7.1.1 COM PLY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 .7.1.1 !
CRAC2...................................................................5.7.5 DECON..................................................................5.7.5 d i E.1 NUREG/BR-0184
Appendix E O
Section EXPAC . . . . . . . . .............. .... ..... . ................... ...........C11 FIRAC .................................................................. . Cll FORECAST ................. 5.6.3, 5.7.1, 5.7.2.2, 5.7.3.3, 5.7.5, 5.7.6.4, 5.7.4.1, 5.7.7, 5.7.7.1, 5.7.8,
........... ........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7. 9, 5. 7.10, 5.7.1 1 GENil ......... ... .................. . . .... ... ........ .... .. 5.7.1.1 HESAP . ........ .... ... . .. ............................. ... ....... A.1.1 IRRAS ....................... ...... ...... ... ........ . . ... ...... 5.6.1 MACCS .... ..... ..... ...... ............................... . 5.7.1.1, 5.7.5, C10 NUCLARR ........ ....................................................... 5.6.1 RECAP ..... ................................................... . ... 5.7.7.1 SARA .... ..... .. .......... ....................... ............. .. 5.6.1 TEMAC .. ... . .... .................... ......................... .. 5.4.3.3 TORAC ...... . .. . ............ ......................... .... .... .. C11 CRGR charter ..... .. ... . ..... .. .............. . ......................... 2.3 regulatory analysis ............... ...... .............. .............. 1.2.1, 2.2, 2.3 D
Cumulative Accounting of Safety lmprovements ................. .. ................ ... ... A.2 Decommissioning costs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... . . . . . . . . . . . . . . 5.7.6.1, 5.7.7.2 Definitions . . . . .. . .. ......... ...... ......... .............. . ..... 1.2.1 Delphi technique . . . ....................... .... . .......................... . 5.7.14 Discounting . ........ ..... ...... ...... ... .... ........ ....... . 5.7, 5.7.1.3, B.2 Dollars, conversion to common ycar . . . . .. ............................... ....... ... 5.8 E
Energy Economic Data Base . . . ... ....... . .. ......... ................. .... B.3 Emergeney preparedness / response . . . .... . ..... ....... ....... . . . . . . . . . 5.7.1.1, C 8 Environmental impacts / considerations ... .. .. .... ................. .. . . . . 5.5.17, 5.7.17 Example regulatory analyses . . . . .. .. ... . ... .. .. .. .. .. . . C8, C9, C10 Expected value ..... . . ... ..... . . .. .... .. . .. .... .. . ... 4.3 Expert judgement . . . .. .. . . .. .. ... . . ... ... ..... .. . 5.6.2 G
General public costs . . . . . . . ... .. . .. .. ... ...... .. . . . .... . . . . 5.7.12 Gross domestic product price deflator . . . . ..... ....... . . ....... .. .. .... . .... . 5.8 O
NUREG/BR-0184 E.2
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. Appendix E 4
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Section Handbook i
history .......................................................... ........... 1 .
l uses .. ......... ... ...................................................... 1.1
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Health effects Accident related . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7.1.1 monetary conversion factor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.~/.1.2 l Human factors issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1 i History of regulatory analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 !
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Improvements in knowledge . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7.13 i Individual plant examination reports . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . 3, 5.6.1 i Individual plant exammation reports of external event reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3,5.6.1 l Industry costs !
implementation . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7.7 ;
operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7. 8 i use of industry risk and cost estimates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.3 ,
Interdiction criteria' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7.1.1. 5.7.5 i t
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i Labor rates other government agencies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7.11 ;
NRC................................................................ 5.7.9.5.7.10 ,
License renewal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . 5.7
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M Major regulatory analysis .............. ............................................. 2.4 !
M etric units . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2. 5 Monetary conversion factor for radiation exposure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7.1.2 ;
N NEPA ...............................................................5.5.17,5.7.17 Net-value measure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... . ...... 4.5 Non-reactor facilities fuel cycle E.3 NUREG/BR-0184 1
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Section fuel enrichment .................................. . . . . . . . . C2.1.1, C2.1.2, C2.1.3, C 8 fuel /MOX fabrication . . . . . . . . . . . . . C.2.1.1, C2.1.2, C2.1.3, C2.3, C2.4, C2.5.1, C4, C6, C.8, C11 fuel reprocessing .................. . . . . . . . . . . C2.1.1, C2.1.7, C2.1.3, C.2.3, C6, C8, C 11 geologic waste disposal . . . . . . . . . . . . . . . . . . . C2.1.1, C2.1.2, C2.1.3, C2.2, C.2.3, C2.4, C5, C6 M RS facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C 5 spent fuel /HLW/TRU waste storage . . . . . . . . . . . . C2.1.1, C.2.1.2 C2.1.3, C.2.2, C2.3, C2.4, C2.5.1, C.6 C.8, C 10, C11 transportation ......... . . . . . . . . . . . . . . . . . . . . C2.1.1, C2.1.3, C2.2, C2.4, C2.5.1, C.5, C6 uranium hexafluoride conversion . . . . . , . . . . . . . . . . . . . C2.1.1, C2.1.2, C2.1,3, C2.5.1, C4, C.6, C8 uranium mining / milling . . . . . . . . . . . . . . . . . . . . . C2.1.1, C2.1.2, C2.1.3, C2.5.1, C4, C6, C 8, C9 non-fuel cycle byproduct / source material manufacturing / distribution . . . . . . . . . . . . . . . . . . . . . . C2.1.1, C2.5.2, C4, C8 measurement / calibration / irradiation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C2.1.1 research/ teaching / experimental / diagnostic / therapeutic ........... .... .... . . . . . . . . . . . C2.1.1 service organizations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C 2.1.1, C 2.5 .2 NRC implementation costs . . . . . . . . . . . . . . . . . . ......................... . .. . . . . 5.5.9, 5.7.9 labor rates ..... ........................................... . . ...... 5.7.9, 5.7.10 operation costs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.5.10, 5.7.10 Non-radiologicat injuries .................................................. ..... 5.7.4.3 o
Occupational health / dose / risk . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
O experience ...................................................................B.3 impacts . . . . . . . . ............. .................. ........... . . . . . . . . 5.7.3, 5.7.4 non-reactor fuel cycle facilities ................................... . . . . . . . . . . . C2.3, C2.4 OMB .............................................................1,4.2,5.7,B.2.1 Onsite property damage costs . . . . . . . . . . . . . . .... ...................... ........... .. 5.7.6 Other considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7.18 Other (non-NRC) government costs . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... ................ 5.7.11 P
Plant specific backfit ........... ........ ... ... ..... ....... ........ . . . . 1.2.1, 2.2 Plutoniiun oxide fabrication and reconstitution . .............. ........... ... .. ... . . . . . . . C.7 Premature facility closure . ....... ........ ... ... ................. .... ....... 5.7.7.2 Price deflator conversions ...................... .................... ......... .... 5.8 Power reactors numbers / lifetimes . . . . . . . . . . . . . . . . . . . . . .................................. . . . . . B.1 N UR EG - 1 150 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... .... .... ......... 5.6.1 Property damage / costs offsite . . . . . . . . . . . . . . . .... .... .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.5.5, 5 .7.5, C 2.5 onsite ............ .. . ..... ............. ................. . . 5.5.6, 5.7.6, C2.5 O
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i Section ;
' Public health / dose / risk . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7.1,5.7.2, B.4, C2.1, C2.2 Probabilistic Risk Assessment . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3,. 5.4, . . . 5.6 l
P 1
l R 1 Ratio measure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . '
RECAP......................................................................5.2
. Regulatory analysis
.......................... 5.7.7.1 )
backfit ..................................<...... ]
CRGR....................................................................2.2 cumulative safety improvements . . . . . . . . . . .. . . . . . . . . . . . . . .
. . . . . . . . . . . . . . . . . . . . . . . 2. 2, 2.3
.j defmition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...........................
. . . . . . . . . . . . . . . . . . . . . . . 1.2.1 .........A history . . . . . . . . . . ........... ....................
1 I level of detail . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .........................
. . . . . . . . . . . . . . . . . . . . . . . . . . . .2.4 level of effort . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... .
major'................................................... 2.4 standard analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... . . . . ....
. . . . . . . . 2.4 required elements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . 4 s'.eps . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . ....... 1.2.2 alternative identification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
decision implementation rationale . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... .... . . .. ........
. . .. . . .4.5. . . . . . . .
g V 7 .. ....... . ....... .. .. .. ..... .. .. ... .... .. .. ....... .. .. .. ....
presentation of results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . '. . . 4.4 4.6 ;
problem / objective statement l
........................................ 4.1 '
value-impact evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . ............. . . . . . . . . . . . . 4.3 l i
when required . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1 !
Regulatory e fficiency . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .................
. . . . . . . . . . . . . . . . . .5.5.14
. I
- Relaxation of requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5 Replacement power costs (reactors) long-term................................................................. 5.7.6.2
. short -term . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7.7.1 '
Routine exposure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . a . . . ............................
. . . . . . . . . . . . . . . . . . . . . . . . . . 5.7.2 S
Safety analysis reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.6.1 Safeguards and security . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.5.16, 5.7.16 Safety goal evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3, 4.1, 4.3, 4.4 Sensitivity / uncertainty analysis error factors . . . . . . . . . . . . . . . . . . . . . . . . -. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7. 5.19 generally . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. .. . 4.3, 5.4 suggested approach for value-input analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4.4 Standard regulatory analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4, 5.3 Sr nary of value-impact results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.8 E.5 NUREG/BR-0184
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Section Thxes .. .................................................................5.5.12 Three Mile lsland ........................................................5.7.3.1,5.7.6.1 Transfer payments . . ........................ ......... ..... . ... ... .... . .... .... . 4.3 V
Values / impacts analysis . . . . . . . . . . . . . . .... .. ...... ......................... .............. . 5.2 definitions .......... .... ........... ........................ .... .......... 4.3 distributional effects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . .. 4.4,5.2 summanzation . . . ...... ....................... .................... ........ 5.8 ,
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