ASTM E1018-2009(2013) 5000 Standard Guide for Application of ASTM Evaluated Cross Section Data File Matrix E706 (IIB)《ASTM经评定横截面数据文件 矩阵E706 (IIB) 应用的标准指南》.pdf

《ASTM E1018-2009(2013) 5000 Standard Guide for Application of ASTM Evaluated Cross Section Data File Matrix E706 (IIB)《ASTM经评定横截面数据文件 矩阵E706 (IIB) 应用的标准指南》.pdf》由会员分享，可在线阅读，更多相关《ASTM E1018-2009(2013) 5000 Standard Guide for Application of ASTM Evaluated Cross Section Data File Matrix E706 (IIB)《ASTM经评定横截面数据文件 矩阵E706 (IIB) 应用的标准指南》.pdf（8页珍藏版）》请在麦多课文档分享上搜索。

1、Designation: E1018 09 (Reapproved 2013)Standard Guide forApplication of ASTM Evaluated Cross Section Data File,Matrix E706 (IIB)1This standard is issued under the fixed designation E1018; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revis

2、ion, the year of last revision. A number in parentheses indicates the year of last reapproval. Asuperscript epsilon () indicates an editorial change since the last revision or reapproval.1. Scope1.1 This guide covers the establishment and use of anASTM evaluated nuclear data cross section and uncert

3、ainty filefor analysis of single or multiple sensor measurements inneutron fields related to light water reactor LWR-PressureVessel Surveillance (PVS). These fields include in- and ex-vessel surveillance positions in operating power reactors,benchmark fields, and reactor test regions.1.2 Requirement

4、s for establishment of ASTM-approvedcross section files address data format, evaluationrequirements, validation in benchmark fields, evaluation oferror estimates (covariance file), and documentation. A furtherrequirement for components of the ASTM-approved crosssection file is their internal consist

5、ency when combined withsensor measurements and used to determine a neutron spec-trum.1.3 Specifications for use include energy region ofapplicability, data processing requirements, and application ofuncertainties.1.4 This guide is directly related to and should be usedprimarily in conjunction with G

6、uides E482 and E944, andPractices E560, E185, and E693.1.5 The ASTM cross section and uncertainty file representsa generally available data set for use in sensor set analysis.However, the availability of this data set does not preclude theuse of other validated data, either proprietary or nonpropri-

7、etary. When alternate cross section files are used that deviatefrom the requirements laid out in this standard, the deviationsshould be noted to the customer ofr the dosimetry application.1.6 This standard does not purport to address all of thesafety concerns, if any, associated with its use. It is

8、theresponsibility of the user of this standard to establish appro-priate safety and health practices and determine the applica-bility of regulatory limitations prior to use.2. Referenced Documents2.1 ASTM Standards:2E170 Terminology Relating to Radiation Measurements andDosimetryE185 Practice for De

9、sign of Surveillance Programs forLight-Water Moderated Nuclear Power Reactor VesselsE482 Guide for Application of Neutron Transport Methodsfor Reactor Vessel Surveillance, E706 (IID)E560 Practice for Extrapolating Reactor Vessel SurveillanceDosimetry Results, E 706(IC) (Withdrawn 2009)3E693 Practice

10、 for Characterizing Neutron Exposures in Ironand Low Alloy Steels in Terms of Displacements PerAtom (DPA), E 706(ID)E706 Master Matrix for Light-Water Reactor PressureVesselSurveillance Standards, E 706(0) (Withdrawn 2011)3E844 Guide for Sensor Set Design and Irradiation forReactor Surveillance, E 7

11、06 (IIC)E853 Practice forAnalysis and Interpretation of Light-WaterReactor Surveillance Results, E706(IA)E854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for ReactorSurveillance, E706(IIIB)E910 Test Method for Application and Analysis of HeliumAccumulation F

12、luence Monitors for Reactor VesselSurveillance, E706 (IIIC)E944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, E 706 (IIA)E1005 Test Method for Application and Analysis of Radio-metric Monitors for Reactor Vessel Surveillance, E 706(IIIA)E2005 Guide for Benchm

13、ark Testing of Reactor Dosimetryin Standard and Reference Neutron Fields3. Terminology3.1 Definitions of Terms Specific to This Standard:1This guide is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applicationsand is the direct responsibility of SubcommitteeE10.05 on Nuclear

14、Radiation Metrology.Current edition approved June 1, 2013. Published July 2013. Originallypublished as E1018 84. Last previous edition approved in 2009 as E1018-09. DOI:10.1520/E1018-09R13.2For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at servic

15、eastm.org. For Annual Book of ASTMStandards volume information, refer to the standards Document Summary page onthe ASTM website.3The last approved version of this historical standard is referenced onwww.astm.org.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA

16、19428-2959. United States13.1.1 benchmark fielda limited number of neutron fieldshave been identified as benchmark fields for the purpose ofdosimetry sensor calibration and dosimetry cross section datadevelopment and testing (1, 2).4See Terminology E170. Thesefields are permanent facilities in which

17、 experiments can berepeated. In addition, differential neutron spectrum measure-ments have been performed in many of the fields to provide,together with transport calculations and integral measurements,the best state-of-the-art neutron spectrum evaluation. Tosupplement the data available from benchm

18、ark fields, most ofwhich are limited in fluence rate intensity, reactor test regionsfor dosimetry method validation have also been defined,including both in-reactor and ex-vessel dosimetry positions.Table 1 lists some of the neutron fields that have been used fordata development, testing, and evalua

19、tion. Other benchmarkfields used for testing LWR calculations are described in E2005Guide for the Benchmark Testing of Reactor Dosimetry inStandard and Reference Neutron Fields, E706 (IIE-1).3.1.1.1 standard fieldthese fields are produced by facili-ties and apparatus that are stable, permanent, and

20、whose fieldsare reproducible with neutron fluence rate intensity, energyspectra, and angular fluence rate distributions characterized tostate-of-the-art accuracy. Important standard field quantitiesmust be verified by interlaboratory measurements. These fieldsexist at the National Institute of Stand

21、ards and Technology(NIST) and other laboratories.3.1.1.2 reference fieldthese fields are produced by facili-ties and apparatus that are permanent and whose fields arereproducible, less well characterized than a standard field, butacceptable as a measurement reference by the community ofusers.3.1.1.3

22、 controlled environmentthese environments arewell-defined neutron fields with some spectral definitions,employed for a restricted set of validation experiments over arange of energies.3.1.2 dosimetry cross sectionscross sections used for do-simetry application and which provide the total cross secti

23、onfor production of particular (measurable) reaction products.These include fission cross sections for production of fissionproducts, activation cross sections for the production of radio-active nuclei, and cross sections for production of measurablestable products, such as helium.3.1.3 evaluated da

24、tavalues of physical quantities repre-senting a current best estimate. Such estimates are developedby experts considering measurements or calculations of thequantity of interest, or both. Cross section evaluations, forexample, are conducted by teams of scientists such as theENDF/B Cross Section Eval

25、uation Working Group (CSEWG)(see also section 3.1.5.2).3.1.4 Evaluated Nuclear Data File (ENDF)consists ofneutron cross sections and other nuclear data evaluated fromavailable experimental measurements and calculations. Twotypes of ENDF files exist.3.1.4.1 ENDF/B filesevaluated files officially appr

26、oved byCSEWG see ENDF documents 102 (3), 201 (4), and 216 (5)after suitable review and testing.3.1.4.2 ENDF/A filesevaluated files including outdatedversions of ENDF/B, the International Reactor Dosimetry File(IRDF-2002) (6), the Japanese Evaluated Nuclear Data Library(JENDL) (7), BROND (USSR) (8) a

27、nd other evaluated crosssection libraries. These files include partial as well as completeevaluations.3.1.5 integral data/differential dataintegral data are datapoints that represent an integrated sensors response over arange of energy. Examples are measurements of reaction rates4The boldfaced numbe

28、rs in parentheses refer to the list of references at the endof this guide.TABLE 1 Partial List of Neutron Fields for Validating Dosimetry Cross SectionsNeutron FieldSample FacilityLocationEnergyUseful Energy Rangefor Data TestingAReferenceDocumentationMedian AverageStandard FieldsThermal Maxwellian

29、NIST . . 0.51 eV252Cf Fission NIST (24) 1.68 MeV 2.13 MeV 100 keV8 MeV Ref 24Designation XCF-5-N1235U Thermal Fission NIST (24) 1.57 MeV 1.97 MeV 250 keV3 MeV Ref 24Mol-25(25, 26) Designation XU5-5-N1ISNF NIST (27) 0.56 MeV ;1.0 MeV 10 keV3.5 MeV Ref 24NISUS (28) Designation ISNF(5)-1-L1Mol- (29)Ref

30、erence FieldsBIG TEN LANL (30, 31) 0.33 MeV 0.58 MeV 10 keV3 MeV Ref 30Fast Reactor Benchmark20CFRMF EGG-Idaho (30, 32) 0.375 MeV 0.76 MeV 4 keV2.5 MeV Ref 30Dosimetry Benchmark 1Controlled EnvironmentsPCA-PV ORNL (33) . . 100 keV10 MeV Ref 33EBR-II ANL-West (34) . . 1 keV10 MeV Ref 34FFTF HEDL (35)

31、 . . 1 keV10 MeV Ref 35AThe requirements for the data testing energy range are much more strict for reference and standard fields than for controlled fields. These testing energy ranges reflectcomparison with calculations based on published spectra for reference and standard fields, but only address

32、 data reproducibility for controlled environments.E1018 09 (2013)2or fission rates in a fission neutron spectrum. Differential dataare measurements at single energy points or over a relativelysmall energy range. Examples are time-of-flight measurements,proton recoil spectrometry, etc. (38).3.1.6 unc

33、ertainty filethe uncertainty in cross section datahas been included with evaluated cross section libraries that areused for dosimetry applications. Because of the correlationsbetween the data points or cross section parameters, theseuncertainties, in general, cannot be expressed as variances, butrat

34、her a covariance matrix must be specified. Through the useof the covariance matrix, uncertainties in derived quantities,such as average cross sections, can be calculated more accu-rately.4. Significance and Use4.1 The ENDF/B library in the United States and similarlibraries elsewhere, such as JEF (9

35、), JENDL (7), and BROND(8), provide a compilation of neutron cross section and othernuclear data for use by the nuclear community. The availabilityof these excellent and consistent evaluations makes possiblestandardized usage, thereby allowing easy referencing andintercomparisons of calculations. Ho

36、wever, as the firstENDF/B files were developed it became apparent that theywere not adequate for all applications. This need resulted in thedevelopment of the ENDF/B Dosimetry File (5, 10), consistingof activation cross sections important for dosimetry applica-tions. This file was made available wor

37、ldwide. Later, other“Special Purpose” files were introduced (23). In the ENDF/B-VIcompilation (41), dosimetry files were identified, but they nolonger appeared as separate evaluation files. The ENDF/V-VIIcompilation (37) removed most of the covariance files used bythe dosimetry community. It kept th

38、e covariance files for the“standard cross sections” in a special sub-library, but thecovariance data in this sub-library is only provided over theenergy range in which each reaction is considered to be a“standard”, and does not include the full energy range requiredfor LWR PVS dosimetry applications

39、.4.2 Another file of evaluated neutron cross section data hasbeen established by the International Atomic Energy Agency(IAEA) for reactor dosimetry applications. This file, theInternational Reactor Dosimetry File (IRDF-2002) (6), drawsupon the ENDF/B files and supplements these evaluations witha set

40、 of reactions evaluated by groups often outside of theUnited States. Some of the IRDF-2002 supplemental reactionsrepresent material evaluations that are currently being exam-ined by the CSEWG. The supplemental IRDF-2002 evalua-tions only include the specific reactions of interest to thedosimetry com

41、munity and not a full material evaluation. TheENDF community requires a complete evaluation beforeincluding it in the main ENDF/B evaluated library.4.3 The application to LWR surveillance dosimetry mayintroduce new data needs that can best be satisfied by thecreation of a dedicated cross section fil

42、e. This file shall be in aform designed for easy application by users (minimal process-ing).The file shall consist of the following types of informationor indicate the sources of the following type of data that shouldbe used to supplement the file contents:4.3.1 Dosimetry cross sections for fission,

43、 activation, he-lium production sensor reactions in LWR environments insupport of radiometric, solid state track recorder, heliumaccumulation dosimetry methods (see Test Methods E853,E854, E910, and E1005).4.3.2 Other cross sections or sensor response functionsuseful for active or passive dosimetry

44、measurements, forexample, the use of neutron absorption cross sections torepresent attenuation corrections due to covers or self-shielding.4.3.3 Cross sections for damage evaluation, such as dis-placements per atom (dpa) in iron.4.3.4 Related nuclear data needed for dosimetry, such asbranching ratio

45、s, fission yields, and atomic abundances.4.4 The ASTM-recommended cross sections and uncertain-ties are based mostly on the ENDF/B-VI and IRDF-2002dosimetry files. Damage cross sections for materials such asiron have been added in order to promote standardization ofreported dpa measurements within t

46、he dosimetry community.Integral measurements from benchmark fields and reactor testregions shall be used to ensure self-consistency and establishcorrelations between cross sections. The total file is intended tobe as self-consistent as possible with respect to both differentialand integral measureme

47、nts as applied in LWR environments.This self-consistency of the data file is mandatory for LWR-pressure vessel surveillance applications, where only verylimited dosimetry data are available.Where modifications to anexisting evaluated cross section have been made to obtain thisself-consistence in LWR

48、 environments, the modifications shallbe detailed in the associated documentation (see 5.6).5. Establishment of Cross Section File5.1 CommitteeThe cross section and uncertainty file shallbe established and maintained under a responsible task groupappointed by Subcommittee E10.05 on Nuclear Radiation

49、Metrology. The task group shall review, and approve all databefore insertion of the file and ensure the adequate testing hasbeen performed on the file contents. The task group shallestablish requirements, data formats, etc.5.2 FormatsFormats shall generally conform to one oftwo types. The first format type is that referred to as theENDF-6 format and is specified in ENDF-201 (4). The secondformat type consists of multigroup data in the 640 groupSAND-II (11,12) energy structure (see Practice E693 forSAND-II energy group structure). The multigroup data format