1、 Severe Accident Progression and Radiological Release (Level 2) PRA Standard for Nuclear Power Plant Applications for Light Water Reactors (LWRs)TRIAL USE AND PILOT APPLICATIONPublication of this standard for trial use has been approved by The American Society of Mechanical Engineers and the America
2、n Nuclear Society. Distribution of this standard for trial use and comment shall not continue beyond 36 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 36-month period, this draft stan
3、dard, revised as necessary, 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 su
4、bmitted to ANSI for approval 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
5、 York, NY 10016-5990ASME/ANS RA-S-1.2-2014 Date of Issuance: January 5, 2015 NOTE: The original trial use period of 24 months was extended to 36 months by the Joint Committee on Nuclear Risk Management. ASME is the registered trademark of The American Society of Mechanical Engineers. This code or st
6、andard was developed under procedures accredited as meeting the 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 cod
7、e or standard was made available for public review and comment 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 d
8、oes not take any position with respect to the validity of any 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
9、. Users of a code or standard are expressly advised that determination 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 in
10、terpreted as government or industry endorsement of this code or 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. The American S
11、ociety of Mechanical Engineers Two Park Avenue, New York, NY 10016-5990 Published by American Nuclear Society 555 North Kensington Avenue La Grange Park, Illinois 60526 USA This document is copyright protected. Copyright 2015 by American Nuclear Society. All rights reserved. Any part of this Standar
12、d may be quoted. Credit lines should read “Extracted from ASME/ANS RA-S-1.2-2014 with permission of the publisher, the American Nuclear Society.” Reproduction prohibited under copyright convention unless written permission is granted by the American Nuclear Society. Printed in the United States of A
13、merica ASME/ANS RA-S-1.2-2014 iii CONTENTS Foreword .v Preparation of Technical Inquiries To The Joint Committee On Nuclear Risk Management . vii Committee Rosters . ix PART 1 INTRODUCTION .1 Section 1.1 Objectives 1 Section 1.2 Coordination with Other Probabilistic Risk Assessment Standards 1 Secti
14、on 1.3 Scope 2 Section 1.4 PRA Capability Categories 3 Section 1.5 Requirements for the PRA Elements .6 Section 1.6 Risk Assessment Application Process .8 Section 1.7 Risk Assessment Technical Requirements 8 Section 1.8 PRA Configuration Control .11 Section 1.9 Peer Review .11 Section 1.10 Addressin
15、g Multiple Hazards 11 PART 2 ACRONYMS AND DEFINITIONS .14 Section 2.1 Acronyms .14 Section 2.2 Definitions .15 PART 3 PRA CONFIGURATION CONTROL 25 Section 3.1 Purpose 25 Section 3.2 PRA Configuration Control Program 25 Section 3.3 Monitoring PRA Inputs and Collecting New Information 25 Section 3.4 P
16、RA Maintenance and Upgrade 25 Section 3.5 Pending Changes .26 Section 3.6 Use of Computer Codes .26 Section 3.7 Documentation .26 PART 4 TECHNICAL REQUIREMENTS .27 Section 4.1 Scope 27 Section 4.2 Level 1/Level 2 PRA InterfaceAccident Sequence Grouping .27 Section 4.3 Containment Capacity Analysis 3
17、5 Section 4.4 Severe Accident Progression Analysis 44 Section 4.5 Probabilistic Treatment of Accident Progression and Source Terms 50 Section 4.6 Source Term Analysis 69 Section 4.7 Evaluation and Presentation of Results .73 Section 4.8 Interface Between Level 2 PRA and Level 3 PRA 75 ASME/ANS RA-S-
18、1.2-2014 iv PART 5 PEER REVIEW .78 Section 5.1 Purpose 78 Section 5.2 Frequency 78 Section 5.3 Methodology 78 Section 5.4 Peer Review Team Composition and Qualifications .78 Section 5.5 Review of PRA Technical Elements to Confirm the Methodology 79 Section 5.6 Expert Judgment 80 Section 5.7 PRA Conf
19、iguration Control .80 Section 5.8 Documentation .80 PART 6 REFERENCES .83 ASME/ANS RA-S-1.2-2014 v (This Foreword is not a part ASME/ANS RA-1.2-2014, “Severe Accident Progression and Radiological Release (Level 2) PRA Standard for Nuclear Power Plant Applications for Light Water Reactor (LWRs)”.) FO
20、REWORD The American Society of Mechanical Engineers (ASME) Board 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). The NRMCC was chartered to coordinate and harmoniz
21、e standards activities related to probabilistic risk assessment (PRA) between ASME and ANS. A key activity resulting from the NRMCC was the development of PRA standards structured around the levels of PRA (i.e., Level 1, Level 2, and Level 3) to be jointly issued by ASME and ANS. In 2011, ASME and A
22、NS decided to combine their respective PRA standards committees to form the ASME/ANS Joint Committee on Nuclear Risk Management (JCNRM). The Severe Accident Progression and Radiological Release (Level 2) PRA Standard for Nuclear Power Plant Applications for Light Water Reactors (LWRs) was initiated
23、by the ANS Risk Informed Standards Committee (RISC) in 2005 and is currently within the responsibility of the JCNRM Subcommittee on Standards Development. The Severe Accident Progression and Radiological Release (Level 2) PRA Standard for Nuclear Power Plant Applications for Light Water Reactors (LW
24、Rs) was developed to provide requirements for the evaluation of containment performance and radiological releases to the environment. The radiological releases considered result from postulated accidents that cause fuel damage. The requirements of this standard apply to the evaluation of risk inform
25、ed applications that use radionuclide release information or as input to the determination of inputs for Level 3 PRA evaluations (e.g., ex-plant consequences). This standard addresses sequences initiated by internal or external events during all modes of operation for operating and evolutionary comm
26、ercial light water reactor (LWR) nuclear plants. This standard is used in conjunction with the ASME/ANS PRA Standard RA-Sa-2009. Specifically, the applicable requirements of the ASME/ANS PRA Standard RA-Sa-2009 are also applicable to those comparable parts of the Level 2 Analysis. In addition, the S
27、evere Accident Progression and Radiological Release (Level 2) PRA Standard for Nuclear Power Plant Applications for Light Water Reactors (LWRs) is structured to provide the requirements for all of the hazards defined in ASME/ANS PRA Standard RA-Sa-2009 and analyzed with a Level 1 PRA. The original d
28、raft of this standard was developed in 2011 and has undergone several revisions prior to the current ballot. This standard sets forth the criteria for the technical adequacy of a Level 2 analysis to support risk-informed decisions for commercial nuclear power plants. Supporting requirements are prov
29、ided for determining the chronology and physical processes governing core damage progression, containment response, and radiological release to the environment as part of PRAs and related analysis methodologies. This standard establishes the requirements to characterize the fission product release f
30、requencies for various containment performance outcomes. Significant input has been received from the JCNRM, specifically the JCNRM Subcommittee on Standards Development (SC-SD). In addition, an SC-SD consensus ballot readiness review team provided a valuable assessment of the proposed Level 2 PRA S
31、tandard prior to its submittal for ballot. Publication for Trial Use The technical requirements in this standard are based on source material from the existing ASME/ANS PRA standard ASME/ANS RA-Sa-2009 as well as the draft PRA standard under development by JCNRM for Level 3 PRA. Although RA-Sa-2009
32、was revised in 2013 in ASME/ANS RA-Sb-2013 ASME/ANS RA-S-1.2-2014 vi (Addendum B), the changes in Addendum B are not fully addressed in this Level 2 PRA trial use standard. JCNRM has approved the use of draft ANS standards with a requirement to follow up with changes to reflect changes in the suppor
33、ting standards. Such changes could necessitate a need for revisions to this standard. The use of source material from not-yet-approved PRA standards and several other considerations have shaped the decision to issue this standard for trial use. It is expected that changes that may be required to acc
34、ount for changes to the supporting standards will be accomplished as part of the effort to upgrade this trial-use standard to the requirements of the American National Standards Institute. This standard is intended to be used together with other PRA standards that cover different aspects of PRA. Spe
35、cifically, this standard is intended to be used directly with ASME/ANS RA-Sa-2009, “Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications.” ASME/ANS RA-Sa-2009 includes Level 1 PRA and large early release frequency (LERF) for internal e
36、vents at-power, external events, internal flood, and internal fire. The Severe Accident Progression and Radiological Release (Level 2) PRA Standard for Nuclear Power Plant Applications for Light Water Reactors (LWRs) cross references supporting requirements related to Systems, Data, Success Criteria
37、, and Human Reliability Analysis to those technical elements of ASME/ANS RA-Sa-2009. This is consistent with the approach used in the LE element in Section 2-2.8 of ASME/ANS RA-Sa-2009 and in other sections of ASME/ANS RA-Sa-2009. The format for this standard was developed in 2005 when no “standard”
38、 format was available. Therefore, it is not consistent with some other published PRA Standards regarding chapter numbers. Following Trial Use, the format of the section numbering will be reevaluated. This standard is issued for Trial Use. Feedback is requested regarding the standard in all areas inc
39、luding the following general areas: Ease of use Clarity of technical supporting requirements (SR) Difficulty in the incorporation of interface requirements Difficulties in interpretation related to: Different hazards Different Plant Operating States Ability to evaluate significance when multiple rel
40、ease categories are involved Adequacy of references to PRA elements in other standards (e.g., Human Reliability, Systems, and Data) Specific areas for which feedback is requested are: The availability of a realistic HRA technique to be used to satisfy SR PT-D2 for Capability Category II The minimum
41、requirements for a peer review team (number of members, total study duration, total on-site presence) Section 5.4.4 A review of the ER HLR and SRs to ensure that the requirements are sufficiently clear and not duplicative. For SR L1-B2, is greater specification on the treatment of failure to run dur
42、ation needed to assess the operation of mitigation equipment during accident progression? ASME/ANS RA-S-1.2-2014 vii PREPARATION OF TECHNICAL INQUIRIES TO THE JOINT COMMITTEE ON NUCLEAR RISK MANAGEMENT INTRODUCTION NOTE FOR TRIAL USE: The text of this section describes the technical inquiry process
43、for approved standards. However, during the trial use period, users are encouraged to provide feedback, ask questions, and interact with the Severe Accident Progression and Radiological Release (Level 2) PRA Standard for Nuclear Power Plant Applications for Light Water Reactors (LWRs) project team.
44、Such feedback may be provided via the Secretary of the Joint Committee on Nuclear Risk Management, as noted below. The ASME/ANS Joint Committee on Nuclear Risk Management (JCNRM) will consider written requests for the interpretation and revision of risk management standards and the development of ne
45、w requirements as dictated by technological development. JCNRMs activities in this latter regard are strictly limited 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
46、Society of Mechanical Engineers (ASME) does 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
47、or on the general application or understanding of the standards requirements. If, based on the inquiry information submitted, it is the opinion of the JCNRM that the inquirer should seek assistance, the inquiry will be returned with the recommendation that such assistance be obtained. To be consider
48、ed, inquiries will require sufficient information 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 sh
49、all be submitted in the following format: (a) 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 consideration of a revision to the present requirements. Concisely provide the information needed for JCNRMs understanding of the inquiry (with sketches as necessary), being sure to include references to the applicable standa
