1、ANSI/ANS-19.3-2011steady-state neutronics methodsfor power reactor analysisANSI/ANS-19.3-2011ANSI/ANS-19.3-2011American National StandardSteady-State Neutronics Methodsfor Power Reactor AnalysisSecretariatAmerican Nuclear SocietyPrepared by theAmerican Nuclear SocietyStandards CommitteeWorking Group
2、 ANS-19.3Published by theAmerican Nuclear Society555 North Kensington AvenueLa Grange Park, Illinois 60526 USAApproved August 26, 2011by theAmerican National Standards Institute, Inc.AmericanNationalStandardDesignation of this document as an American National Standard attests thatthe principles of o
3、penness and due process have been followed in the approvalprocedure and that a consensus of those directly and materially affected bythe standard has been achieved.This standard was developed under procedures of the Standards Committee ofthe American Nuclear Society; these procedures are accredited
4、by the Amer-ican National Standards Institute, Inc., as meeting the criteria for AmericanNational Standards. The consensus committee that approved the standardwas balanced to ensure that competent, concerned, and varied interests havehad an opportunity to participate.An American National Standard is
5、 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 notpreclude anyone from manufacturing, marketing, purchasing, or using prod-ucts, processes, or procedures n
6、ot conforming to the standard.By publication of this standard, the American Nuclear Society does not insureanyone utilizing the standard against liability allegedly arising from or afterits use. The content of this standard reflects acceptable practice at the time ofits approval and publication. Cha
7、nges, if any, occurring through developmentsin the state of the art, may be considered at the time that the standard issubjected to periodic review. It may be reaffirmed, revised, or withdrawn atany time in accordance with established procedures. Users of this standardare cautioned to determine the
8、validity of copies in their possession and toestablish that they are of the latest issue.The American Nuclear Society accepts no responsibility for interpretations ofthis standard made by any individual or by any ad hoc group of individuals.Requests for interpretation should be sent to the Standards
9、 Department atSociety Headquarters. Action will be taken to provide appropriate response inaccordance with established procedures that ensure consensus on theinterpretation.Comments on this standard are encouraged and should be sent to SocietyHeadquarters.Published byAmerican Nuclear Society555 Nort
10、h Kensington AvenueLa Grange Park, Illinois 60526 USACopyright 2011 by American Nuclear Society. All rights reserved.Any part of this standard may be quoted. Credit lines should read “Extracted fromAmerican National Standard ANSI0ANS-19.3-2011 with permission of the publisher,the American Nuclear So
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12、dations, and0or permissivestatementsi.e., “shall,” “should,” and “may,” respectively!in American NationalStandards that are developed and approved by ANS. Responses to inquiries willbe provided according to the Policy Manual for the ANS Standards Committee.Nonrelevant inquiries or those concerning u
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14、iries on standards are published in ANSs magazine, NuclearNews, and are available publicly on the ANS Web site or by contacting the ANSstandards administrator.InquiryFormatInquiry requests must include the following:1! the name, company name if applicable, mailing address, and telephonenumber of the
15、 inquirer;2! reference to the applicable standard edition, section, paragraph, figure,and0or table;3! the purposes 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 toAmerican Nuclear Socie
16、tyATTN: Standards Administrator555 N. Kensington AvenueLa Grange Park, IL 60526or standardsans.orgForewordThis Foreword is not a part of American National Standard “Steady-State NeutronicsMethods for Power Reactor Analysis,” ANSI0ANS-19.3-2011.!The intent of this American National Standard is to pro
17、vide guidance for devel-oping, validating, and utilizing steady-state neutronics methods to calculateneutron reaction-rate spatial distributions, power distributions, and effectiveneutron multiplication constants of nuclear power reactors and to provide guide-lines by which the adequacy of design ca
18、lculations may be demonstrated. Thisstandard recognizes the diversity of the calculation procedures employed inreactor design. Consequently, the major thrusts of this standard are in the areasof methodology, verification, validation, and documentation. This standard isintended to cover reactor physi
19、cs calculations for the entire nuclear industry,from fast to thermal power reactors. Since many different kinds of neutronicsmethods have been utilized for analyzing power reactors, and each has its ownvalidation requirements for accuracy, it is necessary that this standard be of ageneral nature. Fu
20、rthermore, this standard does not endorse or exclude theapplication of any methodology that has been adequately verified, validated,tested, and demonstrated to yield reliable reactor physics parameters.For illustrative purposes, a list of computer codes currently being used through-out the nuclear i
21、ndustry is presented in the Appendix. This Appendix, however,is not part of the standard.Compliance with the intent of this standard can be demonstrated for an intendedarea of applicability of the calculation system used, if the following requirementsare met:1! Selection of models and methodsa! cons
22、ideration of reactor configuration, composition, and all conditionsof operation that significantly affect the calculated quantities and justifi-cation for the resultant model approximations,b! preparation of multigroup cross sections and other parameters, if em-ployed, in conformance with ANSI0ANS-1
23、9.1-2002 R2011!, “Nuclear DataSets for Reactor Design Calculations,” through the use of an application-dependent energy spectrum estimate,c! justification of geometric and neutronic transport approximations uti-lized in the spectrum calculation,d! inclusion of all important space and energy effects
24、in the calculationutilized for the generation of few-group cross sections, if these are employed,e! demonstration of capability, as required by the application, to retrieverequired neutron reaction rates in the physical reactor components fromthe computations and to justify any assumptions that need
25、 to be made inorder to perform this retrieval,f! justification of the spectrum calculation interval used in nuclide deple-tion calculations and justification that the numerical integration time stepis sufficiently small to ensure numerical stability and accuracy appropri-ate to the application;2! Ca
26、lculation system verification and validationEstablish degree of agreement between results obtained with the systembeing verified with results of experiments or of calculations using a moreiaccurate model, over the intended area of applicability for the system beingverified;3! Evaluation of accuracyT
27、he accuracy and range of applicability of data and methods should be evalu-ated by establishment of biases and uncertainties, with degree of confidence,for the calculations that include allowance for uncertainties in the comparisondata;4! Documentation.The intent of this standard is to require the i
28、ndividual to a! give carefulconsideration to those physical and numerical effects that may contribute to thevalidity of results; b! document the reasons for selecting a specific calculationpath; and c! validate the calculation system used over the intended range ofapplicability by testing it against
29、 appropriate experiments, numerical bench-marks, and0or previously validated methods.The requirement for documentation is a crucial part of this standard and willprovide an auditable path. Areas omitted due to proprietary consideration shallbe noted where possible.The most important ways in which th
30、is revision differs from its earlier version,ANSI0ANS-19.3-2005, are as follows:1! The passages on common practices for pressurized water reactors, boilingwater reactors, and liquid metal reactors have been revised to reflect thesignificant advances in reactor physics methods and computer codes made
31、since the last revision. New passages have been added for heavy water reactorand high-temperature gas-cooled reactor methods;2! The Appendix, including the list of commonly used computer codes, hasbeen updated. This revision reflects rapid development in two areas, namely,a! significant advances in
32、reactor physics methods and calculation proce-dures as a result of a rapidly increasing experience base of operating powerreactors,b! computer hardware and operating software developments that havepermitted many traditional approximate methods to be replaced, withincreased performance and user produ
33、ctivity as a result.This standard might reference documents and other standards that have beensuperseded or withdrawn at the time the standard is applied. A statementhas been included in Sec. 9, “References,” that provides guidance on the use ofreferences.This standard does not incorporate the conce
34、pts of generating risk-informedinsights, performance-based requirements, or a graded approach to quality as-surance. The user is advised that one or more of these techniques could enhancethe application of this standard.This standard for reactor physics calculations will undergo review and revisionw
35、ithin 5 years. Suggestions for the improvement of this standard will bewelcome. They should be sent to the attention of the Standards Department,American Nuclear Society, 555 N. Kensington Avenue, La Grange Park, IL60526.iiThis standard was developed and later revised by Working Group ANS-19.3 ofthe
36、 American Nuclear Society, which at the time of this revision had the partici-pation of the following members:B. Rouben, 12 2! reactivity;3! change of nuclide compositions with time.The standard provides the following:1! guidance for the selection of computa-tional methods;2! criteria for verificati
37、on and validation ofcalculation methods used by reactor coreanalysts;3! criteria for evaluation of accuracy andrange of applicability of data and methods;4! requirements for documentation of thepreceding.Note that the use of mixed uranium-plutoniumoxideMOX!fuel has been taken as out of scopefor this
38、 revision of the standard. It will be takeninto account in the next revision.3 Definitions3.1 LimitationsThe following definitions are of a restrictednature for the purpose of this standard. Otherspecialized terms are defined in Glossary ofTerms in Nuclear Science and Technology 1#1!and in the defin
39、ition sections of standards spec-ified in Sec. 9, “References.”3.2 Glossary of termsapplication-dependent multigroup: Adis-crete energy group structure that is inter-mediate between the application-independentmultigroup structure and a few-group struc-ture. The application-dependent multigroupstruct
40、ure may be such that the group con-stants are dependent on reactor compositionthrough an estimated neutron energy spec-trum.An application-dependent multigroup dataset is one type of averaged data set.application-independent multigroup: Adiscrete energy group structure that is suffi-ciently detailed
41、 that the group constantsmay be considered as being independent ofreactor composition, geometry, or spectrum fora wide range of reactor analysis. The application-independent multigroup structure may beemployed directly in reactor design spectrum1!Numbers in brackets refer to corresponding numbers in
42、 Sec. 9, “References.”1calculations, or it may be employed to generategroup constants in an application-dependentmultigroupstructure.Anapplication-independentmultigroupdatasetisonetypeofaverageddataset.cellandsupercell:Theword“cell”denotesoneormorereactorcomponentswithassociatedcool-antandpossiblyad
43、ditionalmoderatorandstruc-turalmaterial!that,forcomputationalpurposes,are assumed to form a spatially repeating arrayin the reactor. The simplest example of a cell isthe “pin cell” in which a single fuel rod or pin issurrounded by coolant e.g., light water, heavywater, or sodium!. Another example is
44、 a bundleoffuelrodscooledbyheavywaterwithinahous-ing, surrounded by a heavy water moderatorspace. More complex geometric configurationsare also used for some applications. These areoften referred to as “supercells,” or sometimes“fuel! assembly cells,” although the exact defi-nition of the term varie
45、s greatly between reac-tor types and is even somewhat subjectivelydefined for a particular reactor type. Super-cells, in the context of this standard, representmore complex “cell” configurations that involvea collection of contiguous cells forming an as-sumedrepeatingarraywithinthereactor,oraug-ment
46、edcellsincorporatingadditionalregionstoserve as a computational artifice, e.g., to ac-countforsignificantspectrumeffectsduetocom-positions outside the cell, or cell configurationsincluding a reactivity device in addition to fuel,coolant, and moderator.data set: A collection of microscopic cross sec-
47、tions and nuclear constants encompassing therange of materials and reaction processes neededfor the application area of interest.averaged data set:A data set prepared by av-eraging an evaluated data set or a processedcontinuous data set with a specified weightingfunction over a specified energy grou
48、p struc-ture. The group structure and weighting func-tionsmaybeselectedtobeapplicationdependent.Application-independent averaged data sets fora wide range of reactor analysis, e.g., light wa-ter reactors LWRs!, are dealt with in ANSI0ANS-19.1-2002R2011!, “Nuclear Data Sets forReactor Design Calculat
49、ions”2#.evaluated data set: A data set that is com-pletely and uniquely specified over the rangesof energy and angles important to reactor cal-culations. Such a data set is based upon avail-able information experimental measurementresults and nuclear theories! and employs ajudgment as to the best physical description ofthe interaction process. An evaluated data setis intended to be independent of reactor com-position, geometries, energy group structures,and spectra.processed continuous data set: A data setprepared by expansion or compaction of a
