ASME/ANS RA-S-1.4-2013

Probabilistic Risk

Assessment Standard for Advanced Non-LWR Nuclear Power Plants

TRIAL USE AND PILOT APPLICATION

Publication 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 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 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 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 Management

The American Society of Mechanical Engineers

Two Park Avenue

New York, NY 10016-5990

TRIAL USE AND PILOT APPLICATION

Publication 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 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 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 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 Management

The American Society of Mechanical Engineers

Two Park Avenue

New York, NY 10016-5990

Date of Issuance: December 9, 2013

NOTE: The trial use period has been extended by the ASME/ANS Joint Committee on Nuclear Risk Management to December 31, 2019.

ASME is the registered trademark of The American Society of Mechanical Engineers.

This code or standard 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 code 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 does 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. 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.

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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 retrieval 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.

CONTENTS

(A detailed contents 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

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.)

FOREWORD

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). NRMCC was chartered to coordinate and harmonize standards activities related to probabilistic risk assessment (PRA) 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 ASME/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 non–light 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 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

1. Nuclear Regulatory Commission PRA requirements for advanced non-LWRs, 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, including high-temperature gas-cooled reactors, liquid metal–cooled 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 PRAs. 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 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 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 development 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 source 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 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 and language to address requirements for PRAs on plants in preoperational stages of the plant life cycle. The approach to

   Capability Categories and supporting requirements for preoperational plant PRAs in this standard is consistent with the approach being taken by ALWR WG.

   PREPARATION OF TECHNICAL 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 technological development. JCNRM’s 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) 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 or on the general application or understanding of the standard’s 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 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 shall be submitted in the following format:

   
   

   1. Scope. The inquiry shall involve a single requirement or closely related requirements. An inquiry letter concerning unrelated subjects will be returned;

      
      

   2. Background. State the purpose of the inquiry, which would be either to obtain an interpretation of the standard’s requirement or to propose consideration of a revision to the present requirements. Provide concisely the information needed for JCNRM’s 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;

      
      

   3. Inquiry Structure. 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 editorially correct;

      
      

   4. Proposed Reply. State what it is believed that the standard requires. If in the inquirer’s opinion a revision to the standard is needed, recommended wording shall be provided;

      
      

   5. Typewritten/Handwritten. The inquiry shall be submitted in typewritten form; however, legible, handwritten inquiries will be considered;

      
      

   6. Inquirer Information. The inquiry shall include name, telephone number, and mailing address of the inquirer;

   7. Submission. The inquiry shall be submitted 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.

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.

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.

CONTRIBUTORS TO THE 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 Berkeley National Laboratory

1. R. Grantom, Cochair, South Texas Project Nuclear Operating Company

2. 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

1. A. Bari, Brookhaven National Laboratory

2. 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 District–Nuclear Energy Division

E. A. 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

1. H. Lagdon, U.S. Department of Energy

2. H. Levinson, AREVA NP, Inc.

S. R. Lewis, Electric Power Research Institute

M. K. Ravindra, MKRavindra Consulting

M. B. Sattison, Idaho National Laboratory

R. E. Schneider, Westinghouse Electric Company

B. D. Sloane, ERIN Engineering & Research, Inc.

D. E. True, ERIN Engineering & Research, Inc.

D. J. Wakefield, ABS Consulting, Inc.

1. B. Wall, Individual

2. W. Young, GE Hitachi

G. L. Zigler, Enercon Services

ASME/ANS JCNRM Advanced Non-LWR PRA Standard Writing Group

K. N. Fleming, Chair, KNF Consulting Services, LLC

F. Schaaf, Vice Chair, Sterling Refrigeration Corporation

S. A. Bernsen, Individual

R. J. Budnitz, Lawrence Berkeley National Laboratory

M. Drouin, U.S. Nuclear Regulatory Commission

B. A. Erler, Erler Engineering

D. Johnson, ABS Consulting

P. Lowry, Pacific Northwest National Laboratory

L. Lusse, PBMR Proprietary, Ltd.

A. Maioli, Westinghouse Electric Company

H. Matsumiya, Toshiba

1. B. Sattison, Idaho National Laboratory

2. Siu, U.S. Nuclear Regulatory Commission

G. A. Tinsley, Technology Insights

J. Wood, U.S. Nuclear Regulatory Commission

J. Young, GE Hitachi

JCNRM Subcommittee on Standards Development

B. D. Sloane, Chair, ERIN Engineering & Research, Inc.

D. W. Henneke, Vice Chair, General Electric Company

A. Afzali, Southern Nuclear Company

V. K. Anderson, Nuclear Energy Institute

S. Bernsen, Individual

J. R. Chapman, Scientech, Inc.

H. L. Detar, Westinghouse Electric Company

M. Drouin, U.S. Nuclear Regulatory Commission

K. N. Fleming, KNF Consulting Services, LLC

C. Guey, Tennessee Valley Authority

E. A. Hughes, Etranco, Inc.

M. T. Leonard, Dycoda, LLC

S. R. Lewis, Electric Power Research Institute

R. J. Lutz, Westinghouse Electric Company

Z. Ma, Idaho National Laboratory

M. B. Sattison, Idaho National Laboratory

V. Sorel, EDF Group

F. Tanaka, Mitsubishi Heavy Industries, Ltd.

D. E. True, ERIN Engineering & Research, Inc.

D. J. Wakefield, ABS Consulting, Inc.

T. A. Wheeler, Sandia National Laboratories

K. Woodard, ABS Consulting

K. Canavan, Alternate, Electric Power Research Institute

G. W. Kindred, Alternate, Tennessee Valley Authority

JCNRM Subcommittee on Standards Maintenance

P. J. Amico, Chair, Hughes Associates, Inc.

A. Maioli, Vice Chair, Westinghouse Electric Company

G. W. Parry, Vice Chair, ERIN Engineering & Research, Inc.

V. Andersen, ERIN Engineering & Research, Inc.

vii

V. K. Anderson, Nuclear Energy Institute

K. R. Fine, FirstEnergy Nuclear Operating Company

D. Finnicum, Westinghouse Electric Company

H. A. Hackerott, Omaha Public Power District–Nuclear Energy Division

D. C. Hance, Electric Power Research Institute

D. G. Harrison, U.S. Nuclear Regulatory Commission

T. G. Hook, Arizona Public Service

E. A. Hughes, Etranco, Inc.

K. L. Kiper, NextEra Energy

S. Kojima, Kojima Risk Institute, Inc.

E. A. Krantz, Scientech, Inc.

J. L. Lachance, Sandia National Laboratories

S. H. Levinson, AREVA NP, Inc.

D. N. Miskiewicz, Engineering Planning and Management, Inc.

P. F. Nelson, National Autonomous University of Mexico

S. P. Nowlen, Sandia National Laboratories

M. K. Ravindra, MKRavindra Consulting

J. B. Savy, Savy Risk Consulting

R. E. Schneider, Westinghouse Electric Company

I. B. Wall, Individual

R. A. Weston, Westinghouse Electric Company

J. W. Young, GE Hitachi

G. L. Zigler, Enercon Services

JCNRM Subcommittee on Planning, Implementation, and Interpretations

E. A. Hughes, Chair, Etranco, Inc.

G. A. Krueger, Vice Chair, Exelon Corporation

A. Afzali, Southern Nuclear Company

R. A. Bari, Brookhaven National Laboratory

R. L. Black, Individual

R. E. Bradley, Nuclear Energy Institute

R. J. Budnitz, Lawrence Berkeley National Laboratory

A. L. Camp, Individual

S. Gosselin, Lucius Pitkin, Inc.

1. R. Grantom, South Texas Project Nuclear Operating Company

2. G. Harrison, U.S. Nuclear Regulatory Commission

A. Lyubarskiy, International Atomic Energy Agency

A. Maioli, Westinghouse Electric Company

P. F. Nelson, National Autonomous University of Mexico

J. Primet, EDF Group

B. D. Sloane, ERIN Engineering & Research, Inc.

B. Snyder, Westinghouse Electric Company

D. E. True, ERIN Engineering & Research, Inc.

J. W. Young, GE Hitachi

G. L. Zigler, Enercon Services

S. Bernsen, Contributing Member, Individual

I. B. Kouzmina, Alternate, International Atomic Energy Agency

V. Sorel, Alternate, EDF Group

Section 1: Introduction Contents

1. Introduction 2

   1. Objective 2

   2. Scope 2

      1. Treatment of Hazard Groups 3

      2. Hazards and Initiating Events 4

   3. Graded Requirements for Different Design–Life Cycle Stages 4

   4. Structure for PRA Requirements 5

      1. PRA Elements 5

      2. High-Level Requirements 6

      3. Supporting Requirements 6

   5. Risk Assessment Application Process 11

   6. PRA Configuration Control 11

   7. Peer Review Requirements 11

   8. Addressing Different PRA Scopes 11

   9. Interface with Other PRA Standards 13

   10. References 14

ASME/ANS RA-S-1.4-2013

SECTION 1 INTRODUCTION

1. Objective

   
   

   This standard sets forth the requirements for probabilistic risk assessments (PRAs) used to support risk- informed decisions for advanced non–light water reactor (non-LWR) nuclear power plants (NPPs) and prescribes a method for applying these requirements for specific applications. To support application of this standard to PRAs for a diverse set of reactor designs such as modular high-temperature gas-cooled reactors, liquid metal–cooled fast reactors, and small modular reactors, based on non-LWR technology, and other advanced non-LWRs, the requirements in this standard were developed on a reactor- technology-neutral basis.

   
   

2. Scope

   
   

   This standard establishes requirements for a PRA for advanced non-LWR NPPs. The requirements in this standard were developed for a broad range of PRA scopes that may include the following:

   
   

   1. Different sources of radioactive material both within and outside the reactor core but within the boundaries of the plant whose risks are to be determined in the PRA scope selected by the user. The technical requirements in this trial-use version of the standard are limited to sources of radioactive material within the reactor coolant system (RCS) pressure boundary (RCPB).1 Technical requirements for other sources of radioactive material such as the spent fuel system are deferred to future editions of this standard;

      
      

   2. Different plant operating states (POSs) including various levels of power operation and shutdown modes;

      
      

   3. Initiating events caused by internal hazards, such as internal events, internal fires, and internal floods, and external hazards such as seismic events, high winds, and external flooding;

      
      

   4. Different event sequence end states, including core or plant damage states (PDSs), and release categories that are sufficient to characterize mechanistic source terms, including releases from event sequences involving two or more reactor units or modules for PRAs on multireactor or multiunit plants;

      
      

   5. Evaluation of different risk metrics including the frequencies of modeled core and PDSs, release categories, risks of off-site radiological exposures 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, health 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;

      
      

   6. Quantification of the event sequence frequencies, mechanistic source terms, off-site radiological consequences, risk metrics, and associated uncertainties, and using this information in a manner consistent with the scope and applications PRA.

1 For pool-type reactors with no RCPB, the scope includes sources within the RCS.