1、 Probabilistic Risk Assessment Standard for Advanced Non-LWR Nuclear Power Plants 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. Distribution of this standard for trial use
2、and comment shall not continue beyond 60 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 60-month period, this draft standard, revised as necessary, will be submitted to the American N
3、ational 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 approval as an American National Standard.
4、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 10016-5990ASME/ANS RA-S-1.4-2013 Date of Issuance: De
5、cember 9, 2013 NOTE: The original trial use period of 36 months was extended to 60 months by the Joint Committee on Nuclear Risk Management. ASME is the registered trademark of The American Society of Mechanical Engineers. This code or standard was developed under procedures accredited as meeting th
6、e criteria for American National Standards. The standards committee that approved the code or standard was balanced to assure that individuals from competent and concerned interests have had an opportunity to participate. The proposed code or standard was made available for public review and comment
7、 that provides an opportunity for additional public input from industry, academia, regulatory agencies, and the public at large. ASME does not “approve,” “rate,” or “endorse” any item, construction, proprietary device, or activity. ASME does not take any position with respect to the validity of any
8、patent rights asserted in connection with any items mentioned in this document and does not undertake to insure anyone utilizing a standard against liability for infringement of any applicable letters patent nor assumes any such liability. Users of a code or standard are expressly advised that deter
9、mination of the validity of any such patent rights, and the risk of infringement of such rights, is entirely their own responsibility. Participation by federal agency representative(s) or person(s) affiliated with industry is not to be interpreted as government or industry endorsement of this code o
10、r standard. ASME accepts responsibility for only those interpretations of this document issued in accordance with the established ASME procedures and policies, which precludes the issuance of interpretations by individuals. No part of this document may be reproduced in any form, in an electronic ret
11、rieval system or otherwise, without the prior written permission of the publisher. The American Society of Mechanical Engineers Two Park Avenue, New York, NY 10016-5990 Copyright 2013 by THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS All rights reserved Published in U.S.A. i CONTENTS (A detailed conte
12、nts precedes each section.) Foreword ii Preparation of Technical Inquiries to the Joint Committee on Nuclear Risk Management . iv Contributors to the Probabilistic Risk Assessment Standard for Advanced Non-LWR Nuclear Power Plants . vi Section 1 Introduction 2 Section 2 Acronyms and Definitions . 16
13、 Section 3 Risk Assessment Application Process 43 Section 4 Risk Assessment Technical Requirements . 58 Section 5 PRA Configuration Control 471 Section 6 Peer Review 474 (All references are distributed within the above sections.) ii FOREWORD The American Society of Mechanical Engineers (ASME) Board
14、on Nuclear Codes and Standards (BNCS) and the American Nuclear Society (ANS) Standards Board mutually agreed in 2004 to form the Nuclear Risk Management Coordinating Committee (NRMCC). NRMCC was chartered to coordinate and harmonize standards activities related to probabilistic risk assessment (PRA)
15、 between ASME and ANS. A key activity resulting from NRMCC was the development of PRA standards structured around the Levels of PRA (i.e., Level 1, Level 2, Level 3) to be jointly issued by ASME and ANS. In 2011, ASME and ANS decided to combine their respective PRA standards committees to form the A
16、SME/ANS Joint Committee on Nuclear Risk Management (JCNRM). In 2006, ASME BNCS established the New Reactor Task Group under the Committee on Nuclear Risk Management (CNRM) to evaluate the need for codes and standards to support the design, construction, licensing, and operation of advanced nonlight
17、water reactor (non-LWR) nuclear power plants (NPPs). Following the formation of JCNRM, the New Reactor Task Group is now known as the ASME/ANS JCNRM Advanced Non-LWR PRA Standard Writing Group (Non-LWR WG). The charter of the Non-LWR WG is to develop recommendations to JCNRM on requirements for the
18、performance of PRAs for advanced non-LWRs. The expected applications of such PRAs include input to licensing and design decisions such as selection of licensing-basis events and safety classification of equipment, satisfaction of U.S. Nuclear Regulatory Commission PRA requirements for advanced non-L
19、WRs, and support of risk-informed applications for advanced non-LWR NPPs. With the concurrence of JCNRM, the Non-LWR WG decided early on that a new PRA standard was needed to support a broad range of applications for advanced reactor designs. To support a diverse mixture of reactor concepts, includi
20、ng high-temperature gas-cooled reactors, liquid metalcooled fast reactors, and small modular reactors, CNRM decided early on to develop this new PRA standard on a reactor-technology-neutral basis using established technology-neutral risk metrics common to existing light water reactor (LWR) Level 3 P
21、RAs. Such risk metrics include frequency of radiological consequences, e.g., dose, health effects, and property damage impacts. In order to support a wide range of applications defined by the non-LWR stakeholders, the scope of this standard is very broad and is comparable to a full-scope Level 3 PRA
22、 for an LWR with a full range of plant operating states (POSs) and hazards. Because some of the advanced non-LWR designs supported by this standard include modular reactor concepts, this standard includes requirements that support an integrated risk of multireactor facilities including accidents on
23、two or more reactor units concurrently. In preparing the technical requirements in this standard, the Non-LWR WG made use of source material from the existing ASME/ANS PRA standard ASME/ANS RA-Sa-2009 as revised in 2013 in ASME/ANS RA-Sb-2013 (Addendum B) as well as draft PRA standards under develop
24、ment by ANS for Low-Power-and-Shutdown PRA, Level 2 PRA, and Level 3 PRA. JCNRM has approved the use of draft ANS standards with a requirement to follow up with changes to reflect changes in the supporting standards. Such changes could necessitate a need for revisions to this standard. The use of so
25、urce material from not-yet-approved PRA standards and the relative lack of experience in performing PRAs on non-LWR NPPs have shaped the decision by JCNRM to issue this standard for trial use. It is expected that changes that may be required to account for changes to the supporting standards will be
26、 accomplished as part of the effort to upgrade this trial-use standard to the requirements of the American National Standards Institute. In preparing this draft standard, the non-LWR WG has worked closely with the Advanced Light Water Reactor Writing Group (ALWR WG) to ensure consistency in approach
27、 and language to address requirements for PRAs on plants in preoperational stages of the plant life cycle. The approach to iii Capability Categories and supporting requirements for preoperational plant PRAs in this standard is consistent with the approach being taken by ALWR WG. iv PREPARATION OF TE
28、CHNICAL INQUIRIES TO THE JOINT COMMITTEE ON NUCLEAR RISK MANAGEMENT INTRODUCTION The ASME/ANS Joint Committee on Nuclear Risk Management (JCNRM) will consider written requests for interpretations and revisions to risk management standards and development of new requirements as dictated by technologi
29、cal development. JCNRMs activities in this latter regard are limited strictly to interpretations of the requirements or to the consideration of revisions to the requirements on the basis of new data or technology. As a matter of published policy, The American Society of Mechanical Engineers (ASME) d
30、oes not “approve,” “certify,” “rate,” or “endorse” any item, construction, proprietary device, or activity, and accordingly, inquiries requiring such consideration will be returned. Moreover, ASME does not act as a consultant on specific engineering problems or on the general application or understa
31、nding of the standards requirements. If based on the inquiry information submitted, it is the opinion of JCNRM that the inquirer should seek assistance, the inquiry will be returned with the recommendation that such assistance be obtained. To be considered, inquiries will require sufficient informat
32、ion for JCNRM to fully understand the request. INQUIRY FORMAT Inquiries shall be limited strictly to interpretations of the requirements or to the consideration of revisions to the present requirements on the basis of new data or technology. Inquiries shall be submitted in the following format: (a)
33、Scope. The inquiry shall involve a single requirement or closely related requirements. An inquiry letter concerning unrelated subjects will be returned; (b) Background. State the purpose of the inquiry, which would be either to obtain an interpretation of the standards requirement or to propose cons
34、ideration of a revision to the present requirements. Provide concisely the information needed for JCNRMs understanding of the inquiry (with sketches as necessary), being sure to include references to the applicable standard edition, addenda, part, appendix, paragraph, figure, or table; (c) Inquiry S
35、tructure. The inquiry shall be stated in a condensed and precise question format, omitting superfluous background information and, where appropriate, composed in such a way that “yes” or “no” (perhaps with provisos) would be an acceptable reply. This inquiry statement should be technically and edito
36、rially correct; (d) Proposed Reply. State what it is believed that the standard requires. If in the inquirers opinion a revision to the standard is needed, recommended wording shall be provided; (e) Typewritten/Handwritten. The inquiry shall be submitted in typewritten form; however, legible, handwr
37、itten inquiries will be considered; (f) Inquirer Information. The inquiry shall include name, telephone number, and mailing address of the inquirer; v (g) Submission. The inquiry shall be submitted to the following address: Secretary, Joint Committee on Nuclear Risk Management, The American Society
38、of Mechanical Engineers, Two Park Avenue, New York, NY 10016-5990. USER RESPONSIBILITY Users of this standard are cautioned that they are responsible for all technical assumptions inherent in the use of PRA models, computer programs, and analysis performed to meet the requirements of this standard.
39、CORRESPONDENCE Suggestions for improvements to the standard or inclusion of additional topics shall be sent to the following address: Secretary, Joint Committee on Nuclear Risk Management, The American Society of Mechanical Engineers, Two Park Avenue, New York, NY 10016-5990. vi CONTRIBUTORS TO THE
40、PROBABILISTIC RISK ASSESSMENT STANDARD FOR ADVANCED NON-LWR NUCLEAR POWER PLANTS (The following is a roster of the Joint Committee on Nuclear Risk Management at the time of the approval of this standard.) ASME/ANS Joint Committee on Nuclear Risk Management (JCNRM) R. J. Budnitz, Cochair, Lawrence Be
41、rkeley National Laboratory C. R. Grantom, Cochair, South Texas Project Nuclear Operating Company D. W. Henneke, Vice Cochair, General Electric P. F. Nelson, Vice Cochair, National Autonomous University of Mexico P. J. Amico, Hughes Associates, Inc. V. K. Anderson, Nuclear Energy Institute R. A. Bari
42、, Brookhaven National Laboratory S. A. Bernsen, Individual J. R. Chapman, Scientech, Inc. M. Drouin, U.S. Nuclear Regulatory Commission D. J. Finnicum, Westinghouse Electric Company K. N. Fleming, KNF Consulting Services, LLC H. A. Hackerott, Omaha Public Power DistrictNuclear Energy Division E. A.
43、Hughes, Etranco, Inc. K. L. Kiper, NextEra Energy S. Kojima, Kojima Risk Institute, Inc. G. A. Krueger, Exelon Corporation J. L. Lachance, Sandia National Laboratories R. H. Lagdon, U.S. Department of Energy S. H. Levinson, AREVA NP, Inc. S. R. Lewis, Electric Power Research Institute M. K. Ravindra
44、, MKRavindra Consulting M. B. Sattison, Idaho National Laboratory R. E. Schneider, Westinghouse Electric Company B. D. Sloane, ERIN Engineering (b) Different plant operating states (POSs) including various levels of power operation and shutdown modes; (c) Initiating events caused by internal hazards
45、, such as internal events, internal fires, and internal floods, and external hazards such as seismic events, high winds, and external flooding; (d) Different event sequence end states, including core or plant damage states (PDSs), and release categories that are sufficient to characterize mechanisti
46、c source terms, including releases from event sequences involving two or more reactor units or modules for PRAs on multireactor or multiunit plants; (e) Evaluation of different risk metrics including the frequencies of modeled core and PDSs, release categories, risks of off-site radiological exposur
47、es and health effects, and the integrated risk of the multiunit plant if that is within the selected PRA scope. The risk metrics supported by this standard are established metrics used in existing light water reactor (LWR) Level 3 PRAs such as frequency of radiological consequences (e.g., dose, heal
48、th effects) that are inherently technology neutral. Surrogate risk metrics used in LWR PRAs such as core damage frequency and large early release frequency are not used as they may not be applicable to non-LWR PRAs; (f) Quantification of the event sequence frequencies, mechanistic source terms, off-
49、site radiological consequences, risk metrics, and associated uncertainties, and using this information in a manner consistent with the scope and applications PRA. 1For pool-type reactors with no RCPB, the scope includes sources within the RCS. ASME/ANS RA-S-1.4-2013 3 It is recognized that for some PRA applications, a full-scope PRA (i.e., a PRA that addresses the full set of requirements covered in this standard) may not be required. In addition, for PRAs performed in various stages of design and licensing, especially those PRAs performed prior to selection of a specific site, t
