1、ANSI/ANS-19.10-2009methods for determiningneutron fluence in BWR and PWRpressure vessel and reactor internalsANSI/ANS-19.10-2009ANSI/ANS-19.10-2009American National StandardMethods for DeterminingNeutron Fluence in BWR and PWRPressure Vessel and Reactor InternalsSecretariatAmerican Nuclear SocietyPr
2、epared by theAmerican Nuclear SocietyStandards CommitteeWorking Group ANS-19.10Published by theAmerican Nuclear Society555 North Kensington AvenueLa Grange Park, Illinois 60526 USAApproved February 24, 2009by theAmerican National Standards Institute, Inc.AmericanNationalStandardDesignation of this d
3、ocument as an American National Standard attests thatthe principles of openness 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 Co
4、mmittee ofthe American Nuclear Society; these procedures are accredited 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 interest
5、s havehad 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 notpreclude anyone from manufacturin
6、g, marketing, purchasing, or using prod-ucts, processes, or procedures not 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 refl
7、ects acceptable practice at the time ofits approval and publication. Changes, 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 It may be reaffirmed, revised, or withdrawn atany time in accordance with estab
8、lished procedures. Users of this standardare cautioned to determine the 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
9、 individuals.Requests for interpretation should be sent to the Standards 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
10、sent to SocietyHeadquarters.Published byAmerican Nuclear Society555 North Kensington AvenueLa Grange Park, Illinois 60526 USACopyright 2009 by American Nuclear Society. All rights reserved.Any part of this standard may be quoted. Credit lines should read “Extracted fromAmerican National Standard ANS
11、I0ANS-19.10-2009 with permission of the publisher,the American Nuclear Society.” Reproduction prohibited under copyright conventionunless written permission is granted by the American Nuclear Society.Printed in the United States of AmericaForewordThis Foreword is not part of American National Standa
12、rd “Methods for DeterminingNeutron Fluence in BWR and PWR Pressure Vessel and Reactor Internals,” ANSI0ANS-19.10-2009.!It is the intent of this American National Standard to provide guidance for theevaluation of pressurized water reactor PWR! and boiling water reactor BWR!pressure vessel and reactor
13、 internals fast E . 1.0 MeV! neutron fluence. Thisstandard outlines the attributes of the methods!, the necessary types of data,the required benchmarking of the method, and the necessary steps in performingthe calculations. The methods! described herein require both experimentallymeasured vessel dos
14、imetry data and corresponding fast neutron fluence calcu-lations to perform the benchmark. This also allows the user to determine theexistence of a bias in the calculated values and to quantify its magnitude.Likewise, the information needed for the benchmark allows the quantification ofuncertainties
15、. The method or methods described in this standard calculates abest-estimate value that is suitable and acceptable! for use in applicationsrelated to Code of Federal Regulations, Title 10, “Energy,” Part 50, “DomesticLicensing of Production and Utilization Facilities,” Section 61, “Fracture Tough-ne
16、ss Requirements for Protection Against Pressurized Thermal Shock Events,”Appendix G, “Fracture Toughness Requirements,” and Appendix H, “ReactorVessel Material Surveillance Program Requirements.” The intended applicationsare for American-made PWRs and BWRs.Compliance with the intent of this standard
17、 can be demonstrated by meeting thefollowing two requirements:1! The calculation must be validated as described in Sec. 5 of this standard;2! the validation must be based on a qualified data set from dosimetrymeasurements performed as described in Sec. 4 of this standard.This standard might referenc
18、e documents and other standards that have beensuperseded or withdrawn at the time the standard is applied. A statement hasbeen included in the reference section that provides guidance on the use ofreferences.This standard was developed by Working Group ANS-19.10 of the AmericanNuclear Society. At th
19、e time of the standards completion, the following membersparticipated in the current version:L. Lois Chair!, U.S. Nuclear Regulatory CommissionJ. F. Carew Secretary!, Brookhaven National LaboratoryJ. M. Adams, National Institute of Standards and TechnologyS. Anderson, Westinghouse Electric Company,
20、LLCR. J. Cacciapouti, IndividualA. Haghighat, University of FloridaW. C. Hopkins, IndividualJ. R. Worsham, AREVAM. Mahgerefteh, Exelon CorporationS. P. Baker, Transware EnterprisesJ. C. Wagner, Oak Ridge National LaboratoryY. Orechwa, U.S. Nuclear Regulatory CommissionR. C. Little, Los Alamos Nation
21、al LaboratoryThe membership of Subcommittee ANS-19, Reactor Physics Standards, at thetime of its review and approval of this standard was as follows:D. M. Cokinos Chair!, Brookhaven National LaboratoryC. T. Rombough Secretary!, CTR Technical Services, Inc.iW. H. Bell, American Institute of Chemical
22、EngineersM. Brady-Raap, Pacific Northwest National LaboratoryD. J. Diamond, Brookhaven National LaboratoryJ. Katakura, Japan Atomic Energy Research InstituteE. R. Knuckles, Florida Power and LightR. C. Little, Los Alamos National LaboratoryL. Lois, U.S. Nuclear Regulatory CommissionR. D. Mosteller,
23、Los Alamos National LaboratoryB. Rouben, 12 10 CFR 50.61, Appendix G 2#;and10CFR 50.61, Appendix H 3#;2! those involved in the determination ofmaterial properties of irradiated reactor ves-sel and reactor internals;3! regulatory agencies in the evaluation oflicensing actions concerning the material
24、prop-erties of irradiated pressure vessels and ir-radiated reactor internals.1.2 Scope, purpose, and applicationThis standard provides a procedure for the eval-uation of the best-estimate fast E . 1.0 MeV!neutron fluence in the annular region betweenthe core and the inside surface of the vessel,thro
25、ugh the pressure vessel and the reactorcavity, between the top and bottom of the ac-tive fuel given the neutron source in the core.This evaluation employs both fast neutron fluxcomputations and measurement data from in-vessel and cavity dosimetry, as appropriate. Thestandard applies to both U.S.-des
26、igned pressur-ized water reactors PWRs! and boiling waterreactors BWRs!.2 DefinitionsThe following definitions apply for the pur-poses of this standard. Other specialized termsconform to Glossary of Terms in Nuclear Sci-ence and Technology 4#.benchmark experiment: A well-defined setof physical exper
27、iments with results judged tobe sufficiently accurate for use as a calcula-tional reference point. The judgment is madeby a group of experts in the subject area.best-estimate fluence: The most accuratevalue of the fluence based on all available mea-surements, calculated results, and adjust-ments bas
28、ed on bias estimates, least-squaresanalyses, and engineering judgment.calculational methodology: The mathemat-ical equations, approximations, assumptions, as-sociated parameters, and calculational procedurethat yield the calculated results. When morethan one step is involved in the calculation, thee
29、ntire sequence of steps comprises the “calcu-lational methodology.”codebenchmark:Comparisontotheresultsofanother code system that has been previouslyvalidatedagainstthebenchmarkexperiments!.continuous-energycross-sectiondata:Cross-section data that are specified in a dense point-wise manner that com
30、prises the energy range.dosimeter reaction: A neutron-induced nu-clear reaction with a product nuclide havingsufficient activity to be measured and relatedto the incident neutron fluence.least-squares adjustment procedure: Amethodforcombiningtheresultsofneutrontrans-port calculations and the results
31、 of dosimetrymeasurements that provides an optimal esti-mate of the fluence by minimizing, in the least-squares sense, the calculation-to-measurementdifferences.multigroup cross-section data: Cross-sectiondata that have been determined by averagingthe continuous-energy cross-section data overdiscret
32、e energy intervals using specified weight-ing functions.1!Numbers in brackets refer to corresponding numbers in Sec. 7, “References.”1neutron fluence: The time-integrated neu-tronfluenceratei.e.,neutronflux!asexpressedin neutrons per square centimeter.reactor internals: Reactor components thatare wi
33、thin the pressure vessel such as thecore shroud, core baffle, core barrel, thermalshield, jet pump and riser, core plate, and topguide.shall, should, and may: The word “shall” isused to denote a requirement; the word “should”is used to denote a recommendation; and theword “may” is used to denote per
34、mission, nei-ther a requirement nor a recommendation.solution variance: A measure of the randomstatistical variation of the Monte Carlo trans-port solution due to a finite number of particlehistories. Mathematically, it is the second cen-tral moment of the distribution about the meanvalue, which is
35、used to measure the dispersionof the distribution about the mean.3 Transport theory calculationalmethods3.1 GeneralThe goal is to accurately determine flux or flu-ence distributions for the analysis of integraldosimetry measurements and for the predic-tion of irradiation damage to vessel internalsan
36、d to the pressure vessel. In the practice sug-gested in this standard, a source distributionthroughout the core is prepared using the re-sults of core physics calculations; multidimen-sional transport theory calculations then areperformed to propagate the neutrons to regionsoutside the core.This sta
37、ndard uses codes based on trans-port theory to determine multigroup three-dimensional flux distributions and to evaluatethe reaction rates of dosimetry materials.Transport theory calculations should be per-formed using deterministic discrete ordinatesSN!5# or statistical Monte Carlo 6# ap-proaches a
38、s discussed in Secs. 3.2.2 and 3.2.3,respectively. Other transport methods may beused if they are part of a benchmarkedmethodology.3.2 Transport calculation3.2.1 Input dataThe four major types of input required are 1!material composition, 2! geometric model, 3!cross-section data, and 4! core neutron
39、 source:1! Material composition, number densities,initial fuel material composition, and coolantor moderator density are required;2! The geometric model should represent thephysical configuration as closely as practical.“As-built” dimensions of the reactor configu-ration should be used when availabl
40、e;3! Appropriate cross-section data should beused. Earlier cross-section sets may be used ifthey are part of a benchmarked methodology.Major considerations includea!the accuracyof the data evaluatione.g., ENDF0B!,b!theenergy group structure, c! the order of thescattering anisotropyi.e., Pnexpansion!
41、,andd! the method used for group-collapsing;4! Determination of the core multigroup neu-tron source requires that for calculationsin cylindrical geometry, the x,y,z! powerdistribution must be converted to r, theta,z! geometry. Also, the multigroup neutronsource spectrum and the average number ofneut
42、rons produced per fission n! must bedetermined.3.2.2 Discrete ordinates SN) methodIn order to ensure an accurate representationof three-dimensional effects, three-dimensionaldiscrete ordinates transport calculations shouldbe used when practical. When three-dimensionalcalculations are not practical,
43、a synthesis methodmay be used to determine the three-dimensionalflux or fluence distribution. In this approach,the fluence distribution is determined by syn-thesizing the results of one- and two-dimensionaldiscrete ordinates solutions. 7#3.2.3 Monte Carlo transport methodIn addition to the considera
44、tions in Sec. 3.2.1,the Monte Carlo model construction requires atechnique to reduce the solution variance.The geometric model used in the Monte Carloanalyses should reflect the actual physical con-figuration. The great flexibility in typical MonteCarlo codes allows a very detailed representa-tion,
45、and this should be used to represent allAmerican National Standard ANSI0ANS-19.10-20092the important features of the geometry underconsideration.Typically, Monte Carlo codes allow use of eithermultigrouporcontinuous-energycrosssections.The continuous-energy cross sections should beused when either t
46、he neutron transport or theresponse function being calculated has a strongenergy dependence that is not adequately rep-resented by the multigroup library.Variance-reduction techniques 6, 8# that havebeen validated may be used to reduce the vari-ance in the Monte Carlo calculation. Tech-niques that m
47、ay be used to improve the statisticsat locations far from the core include the fol-lowing: a! neutron energy cutoff, b! sourcebiasing,c!geometry splitting with Russian rou-lette, and d! weight windows.3.2.4 Adjoint fluence calculationsBecause the reactor conditions are generallydependent on the fuel
48、 cycle, multiple transportcalculations are required to track the fluenceduring plant operation. However, when the op-erating conditions that affect the transport cal-culation e.g., downcomer and core bypasscoolant temperatures, core mechanical design!remain constant, the multiple transport calcu-lat
49、ions may be replaced by a single adjoint cal-culation9#. The adjoint is calculated for a sourcelocated at the vessel or other! location of in-terest that is taken to be proportional to theenergy-dependent response cross section. Typ-ically, the source is taken to be unity above 1.0MeV and zero below 1.0 MeV. When a dosim-eter reaction rate is required, the source typi-cally is taken to be equal to an energy-dependentdosimeter cross section.The fluence or reaction rate response! at thelocation of interest is then determined for eachcycl
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