1、ANSI/ANS-g25g17g20g17g21-2013criteria for modeling and calculatingatmospheric dispersion of routine radiologicalreleases from nuclear facilitiesANSI/ANS-g25g17g20g172-2013group-averaged neutron and gamma-ray cross sectionsfor radiation protection and shielding calculationsfor nuclear power plantsANS
2、I/ANS-6.1.2-2013 American National Standard Group-Averaged Neutron and Gamma-Ray Cross Sections for Radiation Protection and Shielding Calculations for Nuclear Power PlantsSecretariatAmerican Nuclear SocietyPrepared by theAmerican Nuclear SocietyStandards CommitteeWorking Group ANS-6.1.2Published by
3、 theAmerican Nuclear Society555 North Kensington AvenueLa Grange Park, Illinois 60526 USAApproved August 28, 2013by theAmerican National Standards Institute, Inc.Designation of this document as an American National Standard attests that the principles of openness and due process have been followed i
4、n the approval procedure 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 In
5、stitute, 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 interests have had an opportunity to participate.An American National Standard is intended to aid industry, consumer
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13、rdThe American Nuclear Society (ANS) Standards Committee will provide responses to inquiries about requirements, recommendations, and/or permissive statements (i.e., “shall,” “should,” and/or “may,” respectively) in American National Standards that are developed and approved by ANS. Responses to inq
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16、iry requests must include the following:(1) the name, company name if applicable, mailing address, and telephone number of the inquirer;(2) reference to the applicable standard edition, section, paragraph, figure and/or table;(3) the purposes of the inquiry; (4) the inquiry stated in a clear, concis
17、e manner; (5) a proposed reply, if the inquirer is in a position to offer one.Inquiries should be addressed to: American Nuclear SocietyATTN: Standards Administrator 555 N. Kensington AvenueLa Grange Park, IL 60526or standardsans.orgInquiry RequestsInquiry Format(This Foreword is not a part of Ameri
18、can National Standard “Group-Averaged Neutron and Gamma-Ray Cross Sections for Radiation Protection and Shielding Calculations for Nuclear Power Plants,” ANSI/ANS-6.1.2-2013.) A need for computer-readable standard reference neutron and gamma-ray cross-section data was identified by American Nuclear
19、Society Standards Subcommittee ANS-6 in 1975. These cross sections are required for materials and energy ranges of importance in radiation protection and shielding calculations for nuclear power plants. It was observed at that time that data sets that did not meet desired standards of documentation
20、and verification were nonetheless becoming de facto standards. This standard provides guidance in the preparation and verification of neutron and gamma-ray cross-section sets and identifies several sets of standard reference data that meet the procedures specified herein. The identification of stand
21、ard neutron and gamma-ray data is expected to improve the efficiency of shielding and radiation protection computations by reducing redundant verification and processing operations by each user. In addition, shielding computations are expected to become more accurate as a result of effort having bee
22、n focused on the development and testing of nuclear data to be used as a standard. A coupled neutron-gamma multigroup cross-section set, referred to as BUGLE, was developed and tested for this purpose. A revised data set, BUGLE-80, was developed in 1980 on the basis of the BUGLE test results, and th
23、e BUGLE-80 data set was recommended for shielding and radiation protection calculations in ANSI/ANS-6.1.2-1983. A more detailed coupled neutron-gamma multigroup data set, VITAMIN-C, also was identified as meeting the requirements of ANSI/ANS-6.1.2-1983. The SAILOR broad-group averaged data set was a
24、dded in ANSI/ANS-6.1.2-1989 as another recommended data set. ANSI/ANS-6.1.2-1999 cited the BUGLE-96 broad-group cross-section library as the recommended set, replacing both the BUGLE-80 and the SAILOR sets. The more detailed VITAMIN-B6 set was also cited as a replacement for the VITAMIN-C set. Both
25、multigroup cross-section sets were based on the most recent version of the evaluated point cross-section library available at the time the standard was prepared, i.e., ENDF/B-VI, Release 3. ENDF/B-VI contained numerous significant changes to nuclear data relative to earlier versions of ENDF/B. Impro
26、ved experimental data and model predictions are included, and several format changes were made to provide for better representation of the underlying physics and the extension to higher energies. The present edition of this standard cites the BUGLE-B7 broad-group cross-section library, which is an u
27、pdate of the BUGLE-96 library, and the corresponding fine-group cross-section library VITAMIN-B7, which is an update of the VITAMIN-B6 library, as good examples of data sets for use in radiation protection and shielding calculations. Both VITAMIN-B7 and BUGLE-B7 are based on ENDF/B-VII, Release 0. T
28、his edition of the standard also recommends generating problem-specific libraries if the group structure of the broad-group cross-section library BUGLE-B7 will not provide results with sufficient accuracy. An example of where this recommendation should be applied is for cases where thermal neutrons
29、play an important role in calculations.This standard is related to ANSI/ANS-19.1-2002 (R2011), “Nuclear Data Sets for Reactor Design Calculations.” The scope of that standard includes data of importance for reactor core design, while ANS-6.1.2 covers radiation transport and shielding applications, e
30、specially for nuclear power plants. This standard is also related to ANSI/ANS-19.3-2011, “Steady-State Neutronics Methods for Power Reactor Analysis.”Foreword-i-ii-This standard is intended to prescribe recommended practices. The data sets identified are those a novice may use with some confidence a
31、nd should be seriously considered by the expert. In cases where experts select to use other data sets, they would be expected to have strong reasons why the reference data sets provided herein were not used. This revision references documents and other standards that may have been superseded or with
32、drawn at the time this standard is applied. A statement has been included in the references section that provides guidance on the use of references. This standard does not incorporate the concepts of generating risk-informed insights, performance-based requirements, or a graded approach to quality a
33、ssurance. The user is advised that one or more of these techniques could enhance the application of this standard.The membership of Working Group ANS-6.1.2 at the time of the preparation and approval of this revision of the standard included the following: F. A. Alpan (Chair), Westinghouse Electric
34、CompanyJ. M. Adams, Corvus Integration, Inc.S. L. Anderson, Westinghouse Electric Company J. F. Carew, Brookhaven National Laboratory J. L. Francois, Universidad Nacional Autonoma de MexicoP. J. Griffin, Sandia National Laboratories A. Haghighat, Virginia TechR. C. Little, Los Alamos National Labora
35、tory Y. Orechwa, U.S. Nuclear Regulatory CommissionJ. C. Ryman, Individual M. L. Williams, Oak Ridge National LaboratorySubcommittee ANS-6, Radiation Protection and Shielding, had the following membership at the time of its approval of this standard:C. E. Sanders (Chair), University of Nevada, Las V
36、egas/Sanders EngineeringF. A. Alpan, Westinghouse Electric CompanyR. Amato, IndividualP. Bergstrom, National Institute of Standards and TechnologyR. Faw, IndividualN. Hertel, Georgia Institute of TechnologyS. J. Nathan, Savannah River Nuclear SolutionsJ. C. Ryman, Individual A. A. Simpkins, Dade Moe
37、ller the word “should” is used to denote a recommendation; and the word “may” is used to denote permission, neither a requirement nor a recommendation.2.3 DefinitionsThe following definitions apply for purposes of this standard. Other specialized terms conform to Glossary of Terms in Nuclear Science
38、 and Technology 1.1)cross-section processing code: A computer code that converts data in ENDF-6 2 format to a form that is appropriate for use in applications. A cross-section processing code performs calculations such as resonance reconstruction, Doppler broadening, and multigroup averaging. evalua
39、ted nuclear cross-section data: Microscopic cross-section representations derived from basic experimental data, from nuclear models and systematics, and may include the consideration of integral measurements. Evaluated Nuclear Data File: Evaluated nuclear data file stored using a specified format an
40、d procedure. Examples are ENDF/B-VII.1 3, JEFF-3.1.2 4, and JENDL-4.0 5, 6, 7, 8.Evaluated Nuclear Data File/B (ENDF/B): A U.S.-evaluated nuclear data file prepared and reviewed by subject matter experts that is coordinated and maintained by the Cross Section Evaluation Working Group (CSEWG) and the
41、 National Nuclear Data Center (NNDC) at Brookhaven National Laboratory. experimental benchmark: An experiment for which conclusions can be drawn as to the accuracies of calculational models and the underlying nuclear data. An experimental benchmark contains the following: a complete description of t
42、he conditions under which the experiment took place, including input data such as reactor geometry, material compositions, core power distribution, relevant material temperatures, and experimental conditions specified in sufficient detail to model or to replicate the experiment; measured data and th
43、eir associated uncertainties along with a complete specification of data correlations.1)Numbers in brackets refer to corresponding numbers in Sec. 8, “References.”2An experimental benchmark can provide “integral” or “differential” metrics. ”Integral” pertains to integral quantities such as reaction
44、rates, while “differential” provides energy-dependent spectral information such as time-of-flight measurements.group-averaged nuclear data: Evaluated nuclear data averaged over energy groups (intervals) as weighted by specified functions.neutron and gamma-ray cross sections: Cross sections for the i
45、nteractions of neutrons and gamma rays with matter, including cross sections for the secondary emission of neutrons and gamma rays as well as cross sections for the effects of neutrons and gamma rays on materials (e.g., heating or helium generation). The cross sections may be averaged over energy gr
46、oups for use in radiation protection analyses. nuclear cross-section covariance matrix: A matrix representation providing the uncertainty of the nuclear cross sections and the correlation between the evaluated nuclear cross sections for a set of incident particle/neutron energies.numerical benchmark
47、: Specification of a set of input quantities (e.g., composition and geometry of bulk material and radiation sources) and of reference calculated output quantities relevant to the benchmark (e.g., spatial and energy dependence of neutron or gamma-ray fluence profiles) in detail sufficient to determin
48、e the accuracies of a specified calculational method when applied to modeling of the same input specifications. Such determinations are usually made by comparison to a reference output that is an analytic solution. reference data: Published, peer-reviewed, and readily available tabulations of values
49、 and associated uncertainties of physical constants as determined by state-of-the-art methods. In order to minimize transcription errors, these data should be made available in electronic form. standard reference data: Reference data that have been reviewed by a standards development organization and found to meet minimum requirements set forth by the standards development organization for specified purposes. 3 Preparation and verification of neutron and gamma-ray cross sections 3.1 Evaluated microscopic cross sections Evaluated microscopic cross sections shall be derived
