ANSI ASME RA-S-1.3-2017 Standard for Radiological Accident Offsite Consequence Analysis (Level 3 PRA) to Support Nuclear Installation Applications.pdf

1、Standard for Radiological Accident Osite Consequence Analysis (Level 3 PRA) to Support Nuclear Installation Applications TRIAL USE AND PILOT APPLICATIONPublication of this Standard for trial use has been approved by The American Society of Mechanical Engineers and the American Nuclear Society. Distr

2、ibution of this Standard for trial use and comment shall not continue beyond 24 months from the date of publication, unless this period is extended by action of the Joint Committee on Nuclear Risk Management. It is expected that following this 24-month period, this draft Standard, revised as necessa

3、ry, will be submitted to the American National Standards Institute (ANSI) for approval as an American National Standard. A public review in accordance with established ANSI procedures is required at the end of the trial-use period and before a Standard for trial use may be submitted to ANSI for appr

4、oval as an American National Standard. This trial-use Standard is not an American National Standard. Comments and suggestions for revision should be submitted to:Secretary, Joint Committee on Nuclear Risk ManagementThe American Society of Mechanical EngineersTwo Park AvenueNew York, NY 10016-5990ASM

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12、re with the estaRA-S-1.3-20copyright conational ls from s made dustry, y items lity for pressly ghts, is eted as blished 17 with vention i CONTENTS Foreword . iii Preparation of Technical Inquiries to the Joint Committee on Nuclear Risk Management . vi Committee Rosters . viii SECTION 1 INTRODUCTION

13、 1 Section 1.1 Objective . 1 Section 1.2 Coordination With Other Probabilistic Risk Assessment Standards 1 Section 1.3 Purpose and Scope 1 Section 1.4 Structure for Level 3 Requirements 3 Section 1.5 The Nature of the Requirements . 4 Section 1.6 Risk Assessment Application Process: Section 3 . 5 Se

14、ction 1.7 Level 3 Consequence Analysis Technical Requirements: Section 4 5 Section 1.8 Risk Estimation (RI): Section 5 5 Section 1.9 Configuration Control: Section 6 5 Section 1.10 Peer Review: Section 7 . 5 Section 1.11 Documentation Requirements . 5 Section 1.12 Use of Expert Judgment 5 Section 1.

15、13 Process Check . 7 Section 1.14 Computer Codes: Appendix A 7 SECTION 2 ACRONYMS AND DEFINITIONS 9 Section 2.1 Acronyms and Abreviations . 9 Section 2.2 Definition of Terms . 11 SECTION 3 RISK ASSESSMENT APPLICATION PROCESS . 16 Section 3.1 Purpose 16 Section 3.2 Identification of Application and D

16、etermination of Capability Categories (Stage A) . 17 Section 3.3 Assessment of PRA for Necessary Scope, Results, and Models (Stage B) 17 Section 3.4 Determination of the Standards Scope and Level of Detail (Stage C) 18 Section 3.5 Comparison of Level 3 Model to Standard (Stage D). 18 Section 3.6 Acc

17、essing the Risk Implications (Stage E) . 18 SECTION 4 LEVEL 3 CONSEQUENCE ANALYSIS TECHNICAL REQUIREMENTS . 20 Section 4.1 Scope . 20 Section 4.2 Level 3 Consequence Model . 20 Section 4.3 Technical Requirements: General . 20 Section 4.4 Probabilistic Framework for Consequence Analyses . 21 Section

18、4.5 Radionuclide Release Characterization for Level 3 (RE) . 21 Section 4.6 Protective Action Parameters and Other Site Data (PA) 25 Section 4.7 Meteorological Data (ME) 31 Section 4.8 Atmospheric Transport and Dispersion (AD) . 35 Section 4.9 Dosimetry (DO) 41 Section 4.10 Health Effects (HE) . 45

19、Section 4.11 Economic Factors (EC) . 48 Section 4.12 Conditional Consequence Quantification and Reporting (QT) . 52 ii SECTION 5 RISK ESTIMATION (RI) . 55 Section 5.1 Introduction . 55 Section 5.2 Objective . 55 Section 5.3 High Level Requirements . 55 SECTION 6 CONFIGURATION CONTROL . 58 Section 6.

20、1 Purpose 58 Section 6.2 PRA Configuration Control Program . 58 Section 6.3 Monitoring Inputs and Collecting New Information 58 Section 6.4 Maintenance and Upgrades . 58 Section 6.5 Pending Changes . 59 Section 6.6 Use of Computer Codes 59 Section 6.7 Documentation 59 SECTION 7 PEER REVIEW 60 Sectio

21、n 7.1 Purpose 60 Section 7.2 Frequency 60 Section 7.3 Methodology . 60 Section 7.4 Peer Reviewer Team Composition and Personnel Qualifications 61 SECTION 8 REFERENCES 63 Nonmandatory Appendix . 66 Appendix A Computer Codes 66 iii (This Foreword is not part of “Standard for Radiological Accident Offs

22、ite Consequence Analysis (Level 3 PRA) to Support Nuclear Installation Applications,” ASME/ANS RA-S-1.3-2017) FOREWORD The American Nuclear Society (ANS) Standards Board and the American Society of Mechanical Engineers (ASME) Board on Nuclear Codes and Standards (BNCS) mutually agreed in 2004 to for

23、m a Nuclear Risk Management Coordinating Committee (NRMCC). This committee was chartered to coordinate and harmonize standards activities related to probabilistic risk assessment (PRA) between the two standards developing organizations (SDOs). A key activity resulting from the NRMCC was direction to

24、 the ASME/ANS Joint Committee on Nuclear Risk Management (JCNRM) to develop PRA standards structured around the three Levels of PRA (i.e., Level 1, Level 2, Level 3) to be jointly issued by the two societies. This Standard sets forth requirements for determining consequences (i.e., Level 3, also ref

25、erred to as L3 in this Standard) as part of PRAs and related analysis methodologies that can be used to support risk-informed decisions for commercial nuclear power plants. This Standard also prescribes a process for applying these requirements for certain other applications involving release of rad

26、ioactive materials into the atmosphere (e.g., non-light water reactor (LWR) nuclear power plants, research reactors, fuel cycle facilities, and non-reactor nuclear Department of Energy (DOE) facilities). In these cases, supplemental requirements may be needed to ensure technical adequacy. This Stand

27、ard was developed based on the body of knowledge and experience accumulated through the development and application of the ASME/ANS RA-Sb-2013, “Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications,” and L

28、evel 2 PRA Standard ASME/ANS RA-S-1.2-2014, “Severe Accident Progression and Radiological Release (Level 2) PRA Standard for Nuclear Power Plant Applications for Light Water Reactors (LWRs),” which has been approved for trial use and pilot application. This Standard, however, is not dependent upon t

29、hese other PRA standards, although it is noted that the development of the final risk estimation for reactors will be based on combining the results of the Level 1 and Level 2 (Level 1/2) PRA portions (e.g., release frequencies, release characterizations) and the results of the consequence analysis.

30、 Consequences covered within the scope of this Standard include radiation dose and induced health effects, and economic impacts, taking into account atmospheric dispersion, demography, dosimetry, pathways to man, and plant/site characteristics. The radioactive source terms and their frequencies ofte

31、n are passed on from Level 1/2 analyses. The scope of a PRA covered by this Standard is primarily targeted for use to determine the impact of an accident at a nuclear power plant. However, the technology discussed here can be used to determine the impact of a release of radioactive material from any

32、 facility. A Level 3 analysis can use the results of a Level 1 analysis followed by a Level 2 analysis or the results of a combined Level 1/2 analysis (e.g., gas-cooled or other advanced reactors). This Standard describes requirements for calculating the consequences of radionuclide releases into th

33、e environment and how to present the results of such calculations. It is assumed that a computerized con-sequence model will be used. Therefore, emphasis has been placed on the information that is typically required as input and available output. As with any computer code, there are pitfalls associa

34、ted with its use, and there are uncertainties inherent in the quality and representativeness of the input data and the fidelity of the modeling. This Standard attempts to caution against improper use of consequence analysis tools. iv This Standard contains a brief description of each major requireme

35、nt to perform a consequence analysis, and explains why it is necessary, what information results, and how it is to be used. The technical requirements for the various technical elements of a consequence analysis include (1) transport and dispersion in the atmosphere; (2) deposition processes; (3) pr

36、ocesses that lead to the accumulation of radiation doses; (4) protective measures, such as evacuation, that can reduce radiation doses; (5) the effects of radiation doses on the human body; and (6) economic impacts. A section is also included describing how the combined risk results of a Level 1, 2,

37、 and 3 PRA can be presented. This process is referred to as “risk estimation.” It is acknowledged that some topics are subject to argument and continuing development, since consequence modeling is not a precise science and contains significant inherent uncertainties. Where an understanding of the cu

38、rrent state-of-the-art is deemed necessary for a sensible interpretation of the results, a discussion of this topic is included. Other areas that are described in some depth are those in which the users choice of input data can significantly affect the output. Examples include evacuation and shelter

39、ing, and dry deposition velocity. Appendix A, Computer Codes, has been included in this Standard to provide some history and to illustrate typical input parameters and output reports of the calculation results from an acceptable computer code. This Standard might reference documents and other standa

40、rds that will have been superseded or withdrawn at the time the Standard is applied. A statement has been included in the reference section that provides guidance on the use of references. The format for this Standard was developed in 2005 when no “standard” format was available. Therefore, it is no

41、t consistent with some other published PRA standards regarding chapter numbers. Following trial use, the format of the section numbering will be re-evaluated. This Standard is issued for trial use and pilot application. Feedback is requested regarding the Standard in all areas including the followin

42、g: Were the format changes that vary slightly from other contemporary PRA standardshelpful? This includes descriptors added for each supporting requirement (SR). Were the technical SRs and action verbs clear? Notes have been included for a number of SRs. Do these notes result in lack of clarityregar

43、ding what is required and what is provided as added information? Are these noteshelpful? Is the information provided in Appendix A useful? The bases for Capability Categories (i.e., Table 1-1) in this Standard differ from the otherPRA standards in that two attributes are used (i.e., site specificity

44、 and model realism) ratherthan three attributes (i.e., scope and level of detail, plant specificity, and realism). It isthought that the scope and level of detail attribute is adequately addressed by the modelrealism attribute for Level 3 analyses, and that site specificity is more appropriate than

45、plantspecificity. Comments on this change are of interest. Capability Category III is expected to be deleted from this Standard (consistent withplanned changes to the Level 1 and Level 2 PRA standards) following the trial use and pilotapplication period. Are there requirements in Capability Category

46、 III that should beconsidered for incorporation into Capability Category II rather than deletion? Some SRs contain multiple actions verbs (e.g., PA-B1, ME-A3). Did the inclusion ofmultiple action verbs in a single SR result in complications in meeting the requirements orassessing their completion as

47、 part of a Peer Review? Were uncertainty requirements easily understood and implemented?v Were the minimum requirements for peer review teams reasonable (number of members,composition)? Was Section 5 on risk estimation used in your application, and if so were the requirementsclear? The application p

48、rocess in Section 3 differs slightly from that of other PRA standards. Wasthe application process (e.g., flowchart in Figure 3-1) applicable (including references toLevel 1 and Level 2 PRA scope)? If so did you have trouble applying the process? The ASME/ANS PRA standards have been developed in view

49、 of assessing the capability ofa “base” PRA. It is recognized that nuclear facilities in the past have typically onlydeveloped Level 3 PRAs for specific applications, which may vary considerably, and werenot maintained. Based on this historical usage of Level 3 PRA for specific applications,which may vary, this Standard has included some flexibility in the supporting requirements(e.g., no requirement for economic cost modeling or protective-action modeling forCapability Category I.) Are there areas where more or less specificity would be helpful inthe supporting requirement