ANSI ANS 19.4-2017 A Guide for Acquisition and Documentation of Reference Power Reactor Physics Measurements for Nuclear Analysis Verification.pdf

1、An American National StandardPublished by the American Nuclear Society 555 N. Kensington AveLa Grange Park, IL 60526ANSI/ANS-19.4-2017A Guide for Acquisition and Documentation of Reference Power Reactor Physics Measurements for Nuclear Analysis VerificationANSI/ANS-19.4-2017ANSI/ANS-19.4-2017 Americ

2、an National Standard a guide for acquisition and documentation of reference power reactor physics measurements for nuclear analysis verification Secretariat American Nuclear Society Prepared by the American Nuclear Society Standards Committee Working Group ANS-19.4 Published by the American Nuclear

3、Society 555 North Kensington Avenue La Grange Park, Illinois 60526 USA Approved August 24, 2017 by the American National Standards Institute, Inc. American National Standard ANSI/ANS-19.4-2017 American National Standard Designation of this document as an American National Standard attests that the p

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7、rocesses, or procedures not conforming to the standard. By publication of this standard, the American Nuclear Society does not insure anyone utilizing the standard against liability allegedly arising from or after its use. The content of this standard reflects acceptable practice at the time of its

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11、 ensure consensus. Comments on this standard are encouraged and should be sent to Society Headquarters. Published by American Nuclear Society 555 North Kensington Avenue La Grange Park, Illinois 60526 USA This document is copyright protected. Copyright 2017 by American Nuclear Society. All rights re

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13、ican Nuclear Society. Printed in the United States of America American National Standard ANSI/ANS-19.4-2017 Inquiry Requests The American Nuclear Society (ANS) Standards Committee will provide responses to inquiries about requirements, recommendations, and/or permissive statements (i.e., “shall,” “s

14、hould,” and “may,” respectively) in American National Standards that are developed and approved by ANS. Responses to inquiries will be provided according to the Policy Manual for the ANS Standards Committee. Nonrelevant inquiries or those concerning unrelated subjects will be returned with appropria

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16、ine, Nuclear News, and are available publicly on the ANS website or by contacting the ANS Scientific Publications and Standards Department. Inquiry Format Inquiry requests shall include the following: (1) the name, company name if applicable, mailing address, and telephone number of the inquirer; (2

17、) reference to the applicable standard edition, section, paragraph, figure, and/or table; (3) the purpose(s) of the inquiry; (4) the inquiry stated in a clear, concise manner; (5) a proposed reply, if the inquirer is in a position to offer one. Inquiries should be addressed to: American Nuclear Soci

18、ety Scientific Publications and Standards Department 555 N. Kensington Avenue La Grange Park, IL 60526 or standardsans.org American National Standard ANSI/ANS-19.4-2017 American National Standard ANSI/ANS-19.4-2017 i Foreword (This foreword is not a part of American National Standard “A Guide for Ac

19、quisition and Documentation ofReference Power Reactor Physics Measurements for Nuclear Analysis Verification,” ANSI/ANS-19.4-2017, but is included for informational purposes.) It is the purpose of this standard to specify criteria for performing and documenting measurements on light water power reac

20、tors that are to be used as reference measurements in the validation of reactor physics computational methods. Considerably more confidence is placed in nuclear analysis methods when they have been successfully used to calculate performance characteristics that have been carefully measured in an act

21、ual operating system. The existence of well-documented measurements made in a number of operating power reactors will fill a need on the part of the nuclear designer and reactor operator and will permit the development of increased confidence in the design and performance analysis methods used to pr

22、edict reactor performance. This standard is not a guide for routine measurement of reactor physics parameters in an operating reactor. The objective of routine measurements carried out on an operating reactor is to satisfy specific operational, licensing, and contractual requirements. In many cases,

23、 however, measurements made on a routine basis are of sufficient quality to merit their use as reference measurements, and reporting these measurements in accordance with this guide is encouraged. In addition, if time, personnel, and instrumentation are available during the course of operation to pe

24、rform additional measurements not normally required, reactor designers and operators are encouraged to specify, perform, document, and report such measurements for use as reference power reactor physics measurements. This standard was developed primarily for application to measurements on reactors w

25、hose pertinent descriptions are available, or can be made available, to the technical community. This does not preclude its use on reactors for which some reactor information required to simulate the measurement is proprietary. Since performance of reference measurements is not required on any syste

26、m, application of this standard is not related to the question of what reactor design information should, or should not, be publicly disseminated. This standard provides criteria for reference reactor physics measurements. As such it specifically considers only those types of reactor physics measure

27、ments that experience has shown to be practical and reproducible when carried out in large power reactors. The intended current application of this standard is confined to light watermoderated and light watercooled power reactors. This standard is intended primarily for measurements that can be perf

28、ormed at the reactor site; destructive analysis of the spent or partially spent fuel to determine isotopic composition, for example, is not covered. In the event such destructive analysis is carried out, however, its usefulness is increased if it is preceded by a series of reference-quality measurem

29、ents carried out in accordance with this standard. As revised, this standard reflects current power reactor terminology and general practices. Guidance on periodic updates of spatial in-core power distributions and reactor core operating history is given in terms of effective full-power months rathe

30、r than specific megawatt days per metric ton to provide for a common and flexible terminology. Proper axial alignment relative to the core midplane is identified as an important consideration in adjusting raw spatial in-core measurements. Control rod worth measurement using the control group exchang

31、e (rod swap) method that is now commonly used in the industry has been added. References to ANSI/ANS-19.6.1-2011 (R2016), “Reload Startup Physics Tests for Pressurized Water Reactors,” and ANSI/ANS-19.11-2017, “Calculation and Measurement of the Moderator Temperature Coefficient of Reactivity for Pr

32、essurized Water Reactors,” have American National Standard ANSI/ANS-19.4-2017 ii been identified as related standards. Documentation of fuel rod peaking factors as a desirable measurement, identification of radial variation in fuel assembly geometry, and a more detailed organization to the format of

33、 reference measurement documentation has been incorporated into the appendices. This standard was developed by Working Group ANS-19.4 of the American Nuclear Society. The following members contributed to this standard: E. R. Knuckles (Chair), Individual J. D. Bess, Idaho National Laboratory R. T. Ch

34、iang, Individual M. Eckenrode, AREVA Inc. M. Mahgerefteh, Exelon Corporation J. A. Roberts, Kansas State University C. T. Rombough, CTR Technical Services, Inc. B. Rouben, Individual R. W. Sancton, Individual The Reactor Physics Subcommittee had the following membership at the time of its approval o

35、f this standard: D. M. Cokinos (Chair), Brookhaven National Laboratory C. T. Rombough (Secretary), CTR Technical Services, Inc. A. Attard, U.S. Nuclear Regulatory Commission S. P. Baker, Transware Enterprises J. D. Bess, Idaho National Laboratory M. C. Brady Raap, Individual A. Campos, AREVA Inc. R.

36、 T. Chiang, Individual M. DeHart, Idaho National Laboratory D. J. Diamond, Brookhaven National Laboratory M. Eckenrode, AREVA Inc. I. C. Gauld, Oak Ridge National Laboratory A. Haghighat, Virginia Tech Research Center J. I. Katakura, Japan Atomic Energy Agency E. R. Knuckles, Individual R. C. Little

37、, Los Alamos National Laboratory M. Mahgerefteh, Exelon Corporation E. M. Nichita, University of Ontario Institute of Technology B. Rouben, Individual A. Weitzberg, Individual W. Wilson, Individual The Safety and Radiological Analysis Consensus Committee had the following membership at the time of i

38、ts approval of this standard: A. O. Smetana (Chair), Savannah River National Laboratory J. M. Jarvis (Vice Chair), Bechtel Corporation F. A. Alpan, Westinghouse Electric Company, LLC R. S. Amato, Individual American National Standard ANSI/ANS-19.4-2017 iii M. C. Brady Raap, Individual D. M. Cokinos,

39、 Brookhaven National Laboratory D. J. Dudziak, Los Alamos National Laboratory C. C. Graham, Health Physics Society Representative (Employed by Ameren) M. K. Gupta, AECOM-Professional Solutions N. E. Hertel, Georgia Institute of Technology P. Hulse, Sellafield Ltd. D. E. Palmrose, U.S. Nuclear Regula

40、tory Commission C. T. Rombough, CTR Technical Services, Inc. C. E. Sanders, University of Nevada, Las Vegas A. Weitzberg, Individual American National Standard ANSI/ANS-19.4-2017 iv American National Standard ANSI/ANS-19.4-2017 v Contents Section Page 1 Introduction 1 2 Scope 1 3 Definitions 2 3.1 S

41、hall, should, and may . 2 3.2 Definition of terms . 2 4 Relation to other standards . 2 5 Physical description of reactor 3 5.1 Fixed reactor parameters 3 5.1.1 Reactor core description . 3 5.1.2 Fuel assembly description 4 5.1.3 Control element description . 4 5.1.4 Reactor instrumentation . 4 5.2

42、Reactor operating history . 5 6 Test description . 5 6.1 State-point measurements 6 6.1.1 Zero-power critical configurations . 6 6.1.2 At-power critical configurations 6 6.1.3 Spatial in-core detector response maps 7 6.2 Zero-power differential measurements 7 6.2.1 Control rod group worth . 7 6.2.2

43、Reactor temperature coefficient . 8 6.2.3 Soluble poison worth 9 6.3 Gamma ray spectrometry . 9 6.3.1 Specific data requirements . 10 6.3.2 Additional documentation 10 7 Summary of reference measurement documentation 10 7.1 Abstract 11 7.2 Documentation of fixed parameter and operating history 11 7.

44、3 Documentation of methods and techniques 11 7.4 Documentation of test results . 11 7.5 On-site retention of backup information 12 8 References 12 Appendices Appendix A Desirable Power Reactor Measurements 13 Appendix B Core and Component Description . 15 Appendix C Suggested Format for Reference Me

45、asurement Documentation . 17 Table Table A.1 Desirable power reactor measurements 13 American National Standard ANSI/ANS-19.4-2017 vi American National Standard ANSI/ANS-19.4-2017 1 A Guide for Acquisition and Documentation of Reference Power Reactor Physics Measurements for Nuclear Analysis Verific

46、ation 1 Introduction Verification of the calculational system used to predict the nuclear performance characteristics of a power reactor is an essential prerequisite to establishing the reliability of that calculational system. The components of such a system, including the nuclear data sets, neutro

47、n spectra calculations, spatial homogenization procedures, and neutron transport models all involve approximations and uncertainties. Because of the inherent physical and geometric complexities of an operating reactor, the effects of these approximations and uncertainties on the accuracy of the resu

48、lts obtained with the calculational system are difficult to evaluate. Experience has shown that a very useful technique for assessing the reliability of a calculational system in evaluating the performance characteristics of a reactor is to use it to calculate the same or similar characteristics in

49、one or a number of similar reactors for which measurements are available. It is the intent of this standard1)to establish criteria for the acquisition and documentation of measurements made for this purpose. For the most part, power reactor measurements are of an integral nature and as such could be most valuable in testing an overall prediction. There is also value in having measurements of the same type on several reactors in order to minimize systematic errors that might be attributable to a particular reactor. It is recognized that the measurements program carried out in any single r