ANS 8 24-2007 validation of neutron transport methods for nuclear criticality safety calculations《核临界安全计算用中子输运法的复核》.pdf

《ANS 8 24-2007 validation of neutron transport methods for nuclear criticality safety calculations《核临界安全计算用中子输运法的复核》.pdf》由会员分享，可在线阅读，更多相关《ANS 8 24-2007 validation of neutron transport methods for nuclear criticality safety calculations《核临界安全计算用中子输运法的复核》.pdf（28页珍藏版）》请在麦多课文档分享上搜索。

1、ANSI/ANS-8.24-2007 validation of neutron transport methods for nuclear criticality safety calculations ANSI/ANS-8.24-2007ANSI/ANS-8.24-2007 American National Standard Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations Secretariat American Nuclear Society Prepared by

2、the American Nuclear Society Standards Committee Working Group ANS-8.24 Published by the American Nuclear Society 555 North Kensington Avenue La Grange Park, Illinois 60526 USA Approved March 16, 2007 by the American National Standards Institute, Inc.American National Standard Designation of this do

3、cument as an American National Standard attests that the principles of openness and due process have been followed in the approval procedure and that a consensus of those directly and materially affected by the standard has been achieved. This standard was developed under procedures of the Standards

4、 Committee of the American Nuclear Society; these procedures are accredited by the Amer- ican National Standards Institute, Inc., as meeting the criteria for American National Standards. The consensus committee that approved the standard was balanced to ensure that competent, concerned, and varied i

5、nterests have had an opportunity to participate. An American National Standard is intended to aid industry, consumers, gov- ernmental agencies, and general interest groups. Its use is entirely voluntary. The existence of an American National Standard, in and of itself, does not preclude anyone from

6、manufacturing, marketing, purchasing, or using prod- ucts, processes, or procedures not conforming to the standard. By publication of this standard, theAmerican Nuclear Society does not insure anyone utilizing the standard against liability allegedly arising from or after its use. The content of thi

7、s standard reflects acceptable practice at the time of its approval and publication. Changes, if any, occurring through developments in the state of the art, may be considered at the time that the standard is subjected to periodic review. It may be reaffirmed, revised, or withdrawn at any time in ac

8、cordance with established procedures. Users of this standard are cautioned to determine the validity of copies in their possession and to establish that they are of the latest issue. The American Nuclear Society accepts no responsibility for interpretations of this standard made by any individual or

9、 by any ad hoc group of individuals. Requests for interpretation should be sent to the Standards Department at Society Headquarters.Action will be taken to provide appropriate response in accordance with established procedures that ensure consensus on the interpretation. Comments on this standard ar

10、e encouraged and should be sent to Society Headquarters. Published by AmericanNuclearSociety 555NorthKensingtonAvenue LaGrangePark,Illinois60526USA Copyright 2007 by American Nuclear Society. All rights reserved. Any part of this standard may be quoted. Credit lines should read “Extracted from Ameri

11、can National Standard ANSI0ANS-8.24-2007 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 AmericaForeword This Foreword is not a pa

12、rt of American National Standard “Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations,” ANSI0ANS-8.24-2007.! ThisstandardgoesbeyondANSI0ANS-8.1-1998;R2007,“NuclearCriticalitySafety in Operations with Fissionable Materials Outside Reactors,” to provide addi- tional det

13、ail about processes and techniques for the validation of computer-based neutron transport calculational methods used in nuclear criticality safety analy- ses. TheANS-8.24 working group has used its experience, results of conferences on area of applicability and validation, and outside experts to exp

14、and on the concepts identified in ANSI0ANS-8.1-1998; R2007. More detail and method de- scriptions are provided here. Section 4.3 of ANSI0ANS-8.1-1998; R2007 estab- lishes the basic criteria for performing validation of calculational methods. This section contains material that was originally in a se

15、parate standard,ANSI0ANS- 8.11-1975Withdrawn 1983!, “Validation of Calculational Methods for Nuclear CriticalitySafety,”butthatwassubsumedintoANSI0ANS-8.1-1983;R1988With- drawn in 1998!, “Nuclear Criticality Safety in Operations with Fissionable Ma- terials Outside Reactors.” As there is currently a

16、 greater reliance on computer calculations in criticality safety applications, it was felt that a separate standard describing the requirements for the validation of computer-based neutron trans- port methods was again needed. Criticality safety analysts have indicated the need for additional guidan

17、ce be- yond that provided byANSI0ANS-8.1-1998; R2007. For example,ANSI0ANS-8.1- 1998; R2007 indicates validation shall be performed by comparison to “critical and exponential experiments” and that the area of applicability for the valida- tion should be established from this comparison. However, cri

18、ticality safety analysts would benefit from requirements and recommendations on establish- ment of the area of applicability as well as criteria that should be considered in theextensionoftheareaofapplicability,andtheuseofbiasandbiasuncertainty based on comparison to experiments. The existing databa

19、se of critical experi- mentswasdevelopedlargelyinaperiodwhenthefissilematerialoperationsand technical criteria were different from many of the current and planned opera- tions involving fissile material. However, as the number of experiments that focus on current and planned operations has decreased

20、, the industry need to optimize operations and reduce unnecessary conservatism has increased. Thus, the scrutiny and importance placed on validation has increased in recent years. This standard provides requirements and recommendations on proper validation processes and techniques for computer-based

21、 neutron transport calculational methods to expand on the basic criteria established in ANSI0ANS-8.1-1998; R2007. This version of the standard was drafted by Working Group ANS-8.24 of Sub- committee 8 of the American Nuclear Society. The membership of the working group at the time of issuance was as

22、 follows: R. D. BuschChair!, University of New Mexico J. S. Bullington, Washington Safety Management Solutions, LLC C. D. Harmon, Los Alamos National Laboratory J. E. Hicks, U.S. Department of Energy K. D. Kimball, NISYS Corporation D. C. Morey, U.S. Nuclear Regulatory Commission C. V. Parks, Oak Ri

23、dge National Laboratory A. W. Prichard, Pacific Northwest National Laboratory B. M. Rothleder, U.S. Department of Energy N. R. Smith, Serco Assurance, United Kingdom R. W. Tayloe, Individual C. S. Tripp, U.S. Nuclear Regulatory Commission iF. E. Trumble, Washington Safety Management Solutions, LLC L

24、. L. Wetzel, BWX Technologies, Inc. This standard was prepared under the guidance of ANS Subcommittee 8, Fis- sionable Materials Outside Reactors, which had the following membership at the time of its approval: T. P. McLaughlinChair!, Individual J. A. SchlesserSecretary!, Washington Safety Managemen

25、t Solutions, LLC F. M. Alcorn, Individual H. D. Felsher, U.S. Nuclear Regulatory Commission A. S. Garcia, U.S. Department of Energy N. Harris, British Nuclear Fuel, PLC C. M. Hopper, Oak Ridge National Laboratory B. O. Kidd, BWX Technologies R. A. Libby, Pacific Northwest National Laboratory D. A. R

26、eed, Oak Ridge National Laboratory T. A. Reilly, Individual H. Toffer, Fluor Federal Services G. E. Whitesides, Individual Consensus Committee N16, Nuclear Criticality Safety, had the following mem- bership at the time of its approval of this standard: C. M. HopperChair!, Oak Ridge National Laborato

27、ry R. KniefVice-Chair!, XE Corporation G. H. Bidinger, Individual R. D. Busch, University of New Mexico R. S. Eby, American Institute of Chemical Engineers M. A. Galloway, U.S. Nuclear Regulatory Commission C. D. Manning, AREVA NP B. McLeod, Institute of Nuclear Materials Management S. P. Murray, He

28、alth Physics Society R. E. Pevey, University of Tennessee R. L. Reed, Washington Safety Management Solutions, LLC B. M. Rothleder, U.S. Department of Energy W. R. Shackelford, Nuclear Fuel Services, Inc. R. G. Taylor, Individual R. M. Westfall, Oak Ridge National Laboratory L. L. Wetzel, BWX Technol

29、ogies, Inc. R. E. Wilson, U.S. Department of Energy iiContents Section Page 1 Introduction . 1 2 Scope . 1 3 Definitions 1 4 Computer code system . 2 5 Selectionandmodelingofbenchmarks 2 6 Establishment of bias, bias uncertainty, and margins 3 7 Adequacy of the validation . 3 8 Documentation and ind

30、ependent technical review 4 9 References 4 Appendices Appendix A Examples of Physical and Derived Parameters . 5 Appendix B Sources of Information on Experiments 7 Appendix C Annotated Bibliography for Use in Validation of Computational Methods for Criticality Safety 9 Appendix D Validation Example

31、. 12 Tables TableD.1 Processparametersanddata 13 Table D.2 Parameters of benchmark experiments . 14 Table D.3 Calculation results for the benchmark experiments . 15 Table D.4 Comparison of benchmark applicability and process parameters 17 Figures Figure D.1 Benchmark results plotted against average

32、lethargy energy causing fission . 18 FigureD.2 Benchmarkresultsplottedagainsturaniumconcentration 18 Figure D.3 Benchmark results plotted against moderation ratio H0 235 U! . 18 Figure D.4 Validation results showing the calculational margin and the margin of subcriticality . 20 iiiValidation of Neut

33、ron Transport Methods for Nuclear Criticality Safety Calculations 1 Introduction This standard amplifies the basic require- ments and recommendations for validation as described in ANSI0ANS-8.1-1998; R2007, “Nu- clear Criticality Safety in Operations with Fis- sionable Materials Outside Reactors”1#,

34、 1! as applied to computer-based neutron transport calculational methods. Requirements and rec- ommendations for the validation of neutron transport calculational methods applied to nu- clear criticality safety analyses are provided in this standard. In particular, this standard pro- vides requireme

35、nts and recommendations for selecting benchmarks; estimating the bias and bias uncertainty; selecting appropriate mar- gins, both within and beyond the benchmark applicability; and documenting the validation. To conform with this standard, all operations shall be performed in accordance with its req

36、uirements. 2 Scope This standard provides requirements and rec- ommendations for validation, including es- tablishing applicability, of neutron transport calculational methods used in determining crit- ical or subcritical conditions for nuclear criti- cality safety analyses. 3 Definitions 3.1 Limita

37、tions The definitions given below are of a restricted nature for the purpose of this standard. Other specialized terms are defined in Glossary of TermsinNuclearScienceandTechnology2#and in Glossary of Nuclear Criticality Terms3#. 3.2 Shall, Should, May The word “shall” is used to denote a require- m

38、ent; the word “should” is used to denote a recommendation; and the word “may” is used to denote permission, neither a requirement nor a recommendation. 3.3 Glossary of terms benchmark:Anexperimentusedforvalidation. benchmark applicability 2! : The benchmark parameterse.g., material compositions, geo

39、m- etry, neutron energy spectra! and their bound- ing values from which the bias and bias uncertainty of a calculational method are established. bias: The systematic difference between calcu- lated results and experimental data. Positive bias is where the calculated results are greater than the expe

40、rimental data. 3! bias uncertainty: The uncertainty that ac- countsforthecombinedeffectsofuncertainties in the benchmarks, the calculational models of the benchmarks, and the calculational method. calculational margin: An allowance for bias and bias uncertainty plus considerations of un- certainties

41、 related to interpolation, extrapola- tion, and trending. calculationalmethod:Themathematicalpro- cedures, equations, approximations, assump- tions,andassociatednumericalparameterse.g., crosssections!thatyieldthecalculatedresults. 1! Numbers in brackets refer to corresponding numbers in Section 9, “

42、References.” 2! Benchmark applicability embodies the same concept as area of applicability as defined in ANSI0ANS-8.1- 1998; R20071#. 3! The sign of the bias is arbitrary. For the purpose of this standard, it has been defined to be positive when the calculated values exceed the experimental values,

43、but it could be defined otherwise. 1computercodesystem:Acalculationalmethod, computerhardware,andcomputersoftwarein- cluding the operating system!. margin of subcriticality: An allowance be- yond the calculational margin to ensure sub- criticality. upper subcritical limit: Alimit on the calcu- lated

44、k-effectivevalueestablishedtoensurethat conditions calculated to be subcritical will ac- tually be subcritical. validation:Theprocessofquantifyinge.g.,es- tablishing the appropriate bias and bias uncer- tainty!thesuitabilityofacomputercodesystem for use in nuclear criticality safety analyses. valida

45、tion applicability 4! : A domain, which could be beyond the bounds of the benchmark applicability, within which the margins de- rived from validation of a calculational method have been applied. verification: The process of confirming that the computer code system correctly performs intended numeric

46、al calculations. 4 Computer code system 4.1 Verification of the computer code system shall becompletedpriortovalidation.Correctinstal- lation and operation of the computer code sys- tem should be documented. 4.2 Thecomputercodesystemtobevalidatedshall be placed under an appropriate configuration con

47、trol program. Any change to the computer codesystemshallbeevaluatedtodetermineits effect on the validation. 5 Selection and modeling of benchmarks 5.1 Appropriate system or process parameters that correlate the experiments to the system or pro- cess under consideration shall be identified. See Appen

48、dix A for example physical and de- rived parameters.! 5.2 Normal and credible abnormal conditions for the system or process shall be identified when determining the appropriate parameters and their range of values. 5.3 Experiments shall be reviewed for complete- ness and accuracy of information prio

49、r to use as benchmarks. See Appendix B for several sources of information on experiments.! 5.4 Selectedbenchmarksshouldencompasstheap- propriateparametervaluesspanningtherange ofnormalandcredibleabnormalconditionsan- ticipatedforthesystemorprocesstowhichthe validation will be applied. 5.5 Benchmarks selected should be consistent with the modeling capabilities of the calculational method. 5.6 To minimize systematic error, benchmarks should be drawn from multiple, independent experimental series and sources. 5.7 The calculational methods and analysis tech- niquese.g.,