ANS 8 24-2017 Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations.pdf

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1、An American National StandardPublished by the American Nuclear Society 555 N. Kensington AveLa Grange Park, IL 60526ANSI/ANS-8.24-2017Validation of Neutron TranportMethods for Nuclear Criticality Safety CalculationsANSI/ANS-8.24-2017ANSI/ANS-8.24-2017 American National Standard Validation of Neutron

2、 Transport Methods for Nuclear Criticality Safety Calculations Secretariat American Nuclear Society Prepared by 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 De

3、cember 12, 2017 by the American National Standards Institute, Inc. American National Standard ANSI/ANS-8.24-2017 American National Standard Designation of this document as an American National Standard attests that the principles of openness and due process have been followed in the approval procedu

4、re and that a consensus of those directly and materially affected by the standard has been achieved. This standard was developed under the procedures of the Standards Committee of the American Nuclear Society; these procedures are accredited by the American National Standards Institute, Inc., as mee

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11、uld 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 reserved. Any part of this standard may be quoted. Credit lines shoul

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13、can National Standard ANSI/ANS-8.24-2017 Inquiry Requests The American Nuclear Society (ANS) Standards Committee will provide responses to inquiries about requirements, recommendations, and/or permissive statements (i.e., “shall,” “should,” and “may,” respectively) in American National Standards tha

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17、ure, 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 Society Scientific Publications and Standards Department 555 N. Kensington A

18、venue La Grange Park, IL 60526 or standardsans.org American National Standard ANSI/ANS-8.24-2017 American National Standard ANSI/ANS-8.24-2017 i Foreword (This foreword is not a part of American National Standard “Validation of Neutron Transport Methods forNuclear Criticality Safety Calculations,” A

19、NSI/ANS-8.24-2017, but is included for informational purposes.) This standard provides additional details beyond those contained in ANSI/ANS-8.1-1998 (R2007) (W2014), “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors,” concerning validation of computer-based neutr

20、on transport methods used in nuclear criticality safety analyses. A standard on validation for nuclear criticality safety was originally issued as ANSI N16.9/ANS-8.11-1975 (W1983), “Validation of Calculational Methods for Nuclear Criticality Safety.” Upon its withdrawal, the guidance of this standar

21、d was subsumed into ANSI/ANS-8.1-1983 (R1988) (W1998), “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.” Due to a greater reliance on computer calculations in criticality safety applications in recent years, a separate standard describing the requirements for th

22、e validation of computer-based neutron transport methods was again deemed needed. This need led to the development of ANSI/ANS-8.24-2007 (R2012) (W2017), and detailed guidance on validation was removed from ANSI/ANS-8.1 upon its revision as ANSI/ANS-8.1-2014. The existing database of critical experi

23、ments used in validation was developed largely in a period when the fissile material operations and technical criteria were different from many of the current and planned operations involving fissile material. As the number of experiments that focus on current and planned operations has decreased, t

24、he 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 ne

25、utron transport calculational methods. The ANS-8.24 working group has used its experience, results of conferences on area of applicability and validation, and outside experts to expand on the concepts identified in ANSI/ANS-8.1-1998 (R2007) (W2014). The current revision incorporates user feedback re

26、ceived since the initial issuance in 2007 and comments received during its 2012 reaffirmation. Specific changes include addressing the situation where few or no critical benchmark experiments are available and the potential need for the use of other data beyond critical and exponential experiments,

27、clarifying definitions of bias and other terms, the treatment of apparent outliers, the use of sensitivity/uncertainty methodology in justifying subcritical margin, and the use of correlated experiments. The appendices and references have also been updated. ANSI/ANS-8.1-2014, “Nuclear Criticality Sa

28、fety in Operations with Fissionable Materials Outside Reactors,” describes in Sec. 4.3 the necessity for validation. For validation of neutron transport methods, ANSI/ANS-8.1-2014 defers to ANSI/ANS-8.24-2007 (R2012) (W2017). This standard was prepared by the ANS-8.24 Working Group. The following me

29、mbers contributed to this standard: L. L. Wetzel (Chair), BWX Technologies, Inc. R. D. Busch, University of New Mexico S. H. Finfrock, Savannah River Nuclear Solutions C. E. Gross, Paschal Solutions, Inc. J. E. Hicks, Individual K. D. Kimball, Consolidated Nuclear Security, LLC C. V. Parks, Oak Ridg

30、e National Laboratory A. W. Prichard, Pacific Northwest National Laboratory C. S. Tripp, U.S. Nuclear Regulatory Commission E. F. Trumble, URS Professional Solutions, LLC American National Standard ANSI/ANS-8.24-2017 ii Subcommittee ANS-8, Fissionable Materials Outside Reactors, had the following me

31、mbership at the time of its approval of this standard: B. O. Kidd (Chair), Paschal Solutions, Inc. M. J. Crouse (Secretary), Consolidated Nuclear Security, LLC J. S. Baker, Savannah River Nuclear Solutions M. Barnett, URS Professional Solutions, LLC D. G. Bowen, Oak Ridge National Laboratory E. P. E

32、lliott, Los Alamos National Laboratory D. G. Erickson, Savannah River Nuclear Solutions K. D. Kimball, Consolidated Nuclear Security, LLC D. N. Kupferer, Defense Nuclear Facilities Safety Board T. P. McLaughlin, Individual S. P. Monahan, Sandia National Laboratories J. A. Morman, Argonne National La

33、boratory L. E. Paulson, GE Hitachi Nuclear Energy H. Toffer, Fluor Enterprises, Inc. C. S. Tripp, U.S. Nuclear Regulatory Commission D. D. Winstanley, Sellafield, Ltd. The Nuclear Criticality Safety Consensus Committee had the following membership at the time of its approval of this standard: L. L.

34、Wetzel (Chair), BWX Technologies, Inc. W. R. Shackelford (Vice Chair), Nuclear Fuel Services, Inc. R. W. Bartholomay, C.S. Engineering, Inc. L. J. Berg, U.S. Department of Energy D. G. Bowen, Oak Ridge National Laboratory R. D. Busch, University of New Mexico W. Doane, AREVA Inc. R. S. Eby, AlChE (E

35、mployed by Navarro Research estimating the bias and bias uncertainty; selecting appropriate margins, both within and beyond the benchmark applicability; and documenting the validation. To satisfy certain requirements like matching of benchmarks to process systems, an iterative approach may be needed

36、. This standard uses k-effective, but other parameters that determine subcriticality can be used. 2 Scope This standard provides requirements and recommendations for validation, including establishing applicability, of neutron transport calculational methods used in determining critical or subcritic

37、al conditions for nuclear criticality safety analyses. 3 Definitions 3.1 Limitations The definitions given below are of a restricted nature for the purpose of this standard. Other specialized terms are defined in Glossary of Terms in Nuclear Science and Technology 2 and in “Glossary of Nuclear Criti

38、cality Terms” 3. 3.2 Shall, should, and may shall, should, and may: The word “shall” is used to denote a requirement; 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: A

39、representation of an experiment evaluated for use in validation. The experiment may be critical or slightly subcritical (also called exponential experiment). 1)The current standard, ANSI/ANS-8.24-2017, is hereinafter referred to as “this standard.”2)Numbers in brackets refer to corresponding numbers

40、 in Sec. 9, “References.”American National Standard ANSI/ANS-8.24-2017 2 benchmark applicability: The range of benchmark parameters (e.g., material compositions, geometry, neutron energy spectra) over which the bias and bias uncertainty of a calculational method are established. bias: The systematic

41、 difference between the calculated k-effective and the benchmark k-effective. A positive bias exists where the mean bias or the value of the fitted function is greater than zero.3)bias uncertainty: The uncertainty that accounts for the combined effects of uncertainties in the benchmarks, the calcula

42、tional models of the benchmarks, and the calculational method. calculational margin: An allowance for bias and bias uncertainty plus considerations of uncertainties related to interpolation, extrapolation, and trending of the bias. calculational method: The mathematical procedures, equations, approx

43、imations, assumptions, and associated numerical parameters (e.g., cross sections) that yield the calculated results (k-effective). This is typically the code package and cross-section data. computer code system: A calculational method, computer hardware and software (e.g., operating system) that imp

44、acts the calculational results. margin of subcriticality: An allowance beyond the calculational margin to ensure subcriticality. upper subcritical limit (USL): A limit on the calculated k-effective value established to ensure that conditions calculated to be subcritical will actually be subcritical.

45、 The USL is established using both the calculational margin and the margin of subcriticality. validation: The process of quantifying (e.g., establishing the appropriate bias and bias uncertainty) the suitability of a computer code system for use in nuclear criticality safety analyses by comparison w

46、ith benchmark results. validation applicability4): A domain, which could be beyond the bounds of the benchmark applicability, within which the margins derived from validation of a calculational method have been applied. verification: The process of confirming that the computer code system correctly

47、performs intended numerical calculations. 4 Computer code system 4.1 The computer code system to be validated shall be placed under an appropriate configuration control program. 4.2 Verification of the computer code system shall be completed prior to validation. 3)The sign of the bias is arbitrary.

48、For the purpose of this standard, it has been defined to be positive when the calculated valuesexceed the experimental values, but it could be defined otherwise. 4)The “area of applicability for the validation” in Sec. 4.3 of ANSI/ANS-8.1-2014 is the same as “validation applicability” in thisstandar

49、d. American National Standard ANSI/ANS-8.24-2017 3 5 Selection and modeling of benchmarks 5.1 Appropriate system or process parameters that correlate the experiments to the system(s) or process(es) under consideration shall be identified. (See Appendix A for example physical and derived parameters.) Automated selection systems that consider the isotopes, their abundances, energy ranges, cross-section uncertainties, or other parameters may be used. 5.2 If the validation is being developed for a specific system(s) or process(es), normal and credible abnormal conditions for the system(s)