ASTM E1018-2001 Standard Guide for Application of ASTM Evaluated Cross Section Data File Matrix E 706 (IIB)《ASTM评价横切面数据文件、矩阵E 706 (IIB)的标准指南》.pdf

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1、Designation: E 1018 01Standard Guide forApplication of ASTM Evaluated Cross Section Data File,Matrix E 706 (IIB)1This standard is issued under the fixed designation E 1018; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision, the year o

2、f last revision. A number in parentheses indicates the year of last reapproval. Asuperscript epsilon (e) 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 uncertainty filefor

3、analysis of single or multiple sensor measurements inneutron fields related to LWR-Pressure Vessel Surveillance (PVS). Thesefields include in- and ex-vessel surveillance positions inoperating power reactors, benchmark fields, and reactor testregions.1.2 Requirements for establishment of ASTM-approve

4、dcross section files address data format, evaluation require-ments, validation in benchmark fields, evaluation of errorestimates (covariance file), and documentation. A furtherrequirement for components of the ASTM-approved crosssection file is their internal consistency when combined withsensor mea

5、surements and used to determine a neutron spec-trum.1.3 Specifications for use include energy region of applica-bility, data processing requirements, and application of uncer-tainties.1.4 This guide is directly related to and should be usedprimarily in conjunction with Guides E 482 and E 944, andPra

6、ctices E 560, E 185, and E 693.1.5 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-etary.1.6 This standard

7、does not purport to address all of thesafety concerns, if any, associated with its use. It is 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 S

8、tandards:E 170 Terminology Relating to Radiation Measurementsand Dosimetry2E 185 Practice for Conducting Surveillance Tests for LightWater-Cooled Nuclear Power Reactor Vessels, E706 (IF)3,2E 482 Guide for Application of Neutron Transport Methodsfor Reactor Vessel Surveillance, E706 (IID)3,2E 560 Pra

9、ctice for Extrapolating Reactor Vessel Surveil-lance Dosimetry Results, E706 (IC)3,2E 693 Practice for Characterizing Neutron Exposures inIron and Low Alloy Steels in Terms of Displacements PerAtom (DPA), E706 (ID)3,2E 706 Master Matrix for Light-Water Reactor PressureVessel Surveillance Standards,

10、E706 (O)3,2E 844 Guide for Sensor Set Design and Irradiation forReactor Surveillance, E706 (IIC)3,2E 853 Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, E706 (IA)3,2E 854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for R

11、eactor Sur-veillance, E706 (IIIB)3,2E 910 Test Method for Application and Analysis of HeliumAccumulation Fluence Monitors for Reactor Vessel Sur-veillance, E706 (IIIC)3,2E 944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, (IIA)3,2E 1005 Test Method for Applic

12、ation and Analysis of Radio-metric Monitors for Reactor Vessel Surveillance, E706(IIIA)3,2E 2005 Guide for the Benchmark Testing of Reactor Do-simetry in Standard and Reference Neutron Fields E706(IIE-1)3,2E 2006 Guide for the Benchmark Testing of LWR Calcula-tions E706 (IIE-2)3,23. Terminology3.1 D

13、efinitions of Terms Specific to This Standard:3.1.1 benchmark fielda limited number of neutron fieldshave been identified as benchmark fields for the purpose ofdosimetry sensor calibration and dosimetry cross section data1This guide is under the jurisdiction of ASTM Committee E10 on NuclearTechnolog

14、y and Applicationsand is the direct responsibility of SubcommitteeE10.05on Nuclear Radiation Metrology.Current edition approved June 10, 2001. Published September 2001. Originallypublished as E 1018 84. Last previous edition E 1018 95.2Annual Book of ASTM Standards, Vol. 12.02.3The reference in pare

15、ntheses refers to Section 5 as well as Figs. 1 and 2 ofMatrix E 706.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.development and testing (1, 2, ).4These fields are permanentfacilities in which experiments can be repeated. In addit

16、ion,differential neutron spectrum measurements have been per-formed in many of the fields to provide, together with transportcalculations and integral measurements, the best state-of-the-art neutron spectrum evaluation. To supplement the dataavailable from benchmark fields, most of which are limited

17、 influx intensity, reactor test regions for dosimetry method vali-dation have also been defined, including both in-reactor andex-vessel dosimetry positions. Table 1 lists some of the neutronfields that have been used for data development, testing, andevaluation. Other benchmark fields used for testi

18、ng LWRcalculations are described in E 2005 Guide for the BenchmarkTesting of Reactor Dosimetry in Standard and ReferenceNeutron Fields, E706 (IIE-1).3.1.1.1 standard fieldthese fields are produced by facili-ties and apparatus that are stable, permanent, and whose fieldsare reproducible with neutron

19、flux intensity, energy spectra,and angular flux distributions characterized to state-of-the-artaccuracy. Important standard field quantities must be verifiedby interlaboratory measurements. These fields exist at theNational Institute of Standards and Technology (NIST) andother laboratories.3.1.1.2 r

20、eference 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 controlled environmentthese environments arewell-defined

21、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 fordosimetry application and which provide the total cross sectionfor production of particular (measurable) reaction product

22、s.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 datavalues of physical quantities repre-senting a current best

23、 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 Evaluation Working Group (CSEWG)(see also section 3.1.5.2).3.1.4

24、 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 approved byCSEWG see ENDF documents 102 (3), 201 (4), and 216 (5

25、)after suitable review and testing. The current recommended setof ENDF/B files is ENDF/B-VI, revision 7, June 2000 (38).3.1.4.2 ENDF/A filesevaluated files including outdatedversions of ENDF/B, the International Reactor Dosimetry File(IRDF-90) (6), the Japanese Evaluated Nuclear Data Library(JENDL)

26、(7), and BROND (USSR) (8) evaluated cross sectionlibraries.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 ratesor fission rates in a fission neutron spectrum. Differential da

27、taare measurements at single energy points or over a relativelysmall energy range. Examples are time-of-flight measurements,proton recoil spectrometry, etc.3.1.6 uncertainty filethe uncertainty in cross section datahas been included with evaluated cross section libraries that areused for dosimetry a

28、pplications. Because of the correlationsbetween the data points or cross section parameters, theseuncertainties, in general, cannot be expressed as variances, but4The boldfaced numbers in parentheses refer to the list of references at the endof this guide.TABLE 1 Partial List of Neutron Fields for V

29、alidating Dosimetry Cross SectionsNeutron FieldSample FacilityLocationEnergyUseful Energy Rangefor Data TestingAReferenceDocumentationMedian AverageStandard FieldsThermal Maxwellian NIST . . 0.51 eV252Cf Fission NIST (24) 1.68 MeV 2.13 MeV 100 keV8 MeV Ref 24Designation XCF-5-N1235U Thermal Fission

30、NIST (24) 1.57 MeV 1.97 MeV 250 keV3 MeV Ref 24Mol-x25(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)Reference FieldsBIG TEN LANL (30, 31) 0.33 MeV 0.58 MeV 10 keV3 MeV Ref 30Fast Reactor Benchmark20CFRMF EGG-Idaho (30,

31、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) . . 1 keV10 MeV Ref 35AThe requirements for the data testing energy range are much more strict for reference and st

32、andard fields than for controlled fields. These testing energy ranges reflectcomparison with calculations based on published spectra for reference and standard fields, but only address data reproducibility for controlled environments.E 10182rather a covariance matrix must be specified. Through the u

33、seof 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), JENDL (7), and BROND(8), provide a compilation of neu

34、tron 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. However, as the firstENDF/B files were developed it became

35、 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 worldwide. Later, other“Special Purpose” files were introdu

36、ced. In the latest ENDF/B-VI compilation, dosimetry files are identified, but do nottypically appear as separate evaluation files.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

37、, theInternational Reactor Dosimetry File (IRDF-90) (6), drawsupon the ENDF/B-VI files and supplements these evaluationswith a set of reactions evaluated by groups often outside of theUnited States. Some of the IRDF-90 supplemental reactionsrepresent material evaluations that are currently being exa

38、m-ined by the CSEWG. The supplemental IRDF-90 evaluationsonly include the specific reactions of interest to the dosimetrycommunity and not a full material evaluation. The ENDFcommunity requires a complete evaluation before including itin the ENDF/B evaluated library.4.3 The application to LWR survei

39、llance dosimetry intro-duces new data needs that can best be satisfied by the creationof a special cross section file. This file shall be in a formdesigned for easy application by users (minimal processing).The file shall consist of the following:4.3.1 Dosimetry cross sections for fission, activatio

40、n, he-lium production, and damage sensor reactions in LWR envi-ronments in support of radiometric, solid state track recorder,helium accumulation, and damage monitor dosimetry methods(see Test Methods E 853, E 854, E 910, and E 1005 and MatrixE 706-IIID).4.3.2 Other cross sections or sensor response

41、 functionsuseful for active or passive dosimetry 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

42、data needed for dosimetry, such asbranching ratios, fission yields, and atomic abundances.4.4 The ASTM-recommended cross sections and uncertain-ties are based mostly on the ENDF/B and IRDF dosimetryfiles. Damage cross sections for materials such as iron may beadded in order to promote standardizatio

43、n of reported dpameasurements within the dosimetry community. Integral mea-surements from benchmark fields and reactor test regions shallbe used to ensure self-consistency and establish correlationsbetween cross sections. The total file is intended to be asself-consistent as possible with respect to

44、 both differential andintegral measurements as applied in LWR environments. Thisself-consistency of the data file is mandatory for LWR-pressurevessel surveillance applications, where only very limited do-simetry data are available. Where modifications to an existingevaluated cross section have been

45、made to obtain this self-consistence in LWR environments, the modifications shall bedetailed 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

46、by Subcommittee E10.05 on Nuclear RadiationMetrology. The task group shall review, test, and approve alldata before insertion of the file. The task group shall establishrequirements, data formats, etc.5.2 FormatsFormats shall generally conform to one oftwo types. The first format type is that referr

47、ed 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 E 693 forSAND-II energy group structure). The multigroup data formatis the preferred form since it is more compatible with the

48、 formstypically used to represent facility neutron spectrum. Thespectrum weighting function used to collapse the point crosssection data onto the multigroup energy grid should be genericin nature and shall be completely specified in the associateddocumentation.5.3 Cross Section EvaluationMost evalua

49、tions generallyshall be based on the IRDF-90 Dosimetry File. Cross sectionsshall be consistent within error bounds for selected benchmarkfields (see 5.4 and Table 1). Dosimetry cross sections presentlynot in ENDF/B or IRDF-90 shall be obtained from othersources or new evaluations. Other cross sections may beobtained from other sources, for example, the dpa cross sectionfor iron may be obtained from Practice E 693.5.4 Cross Section ValidationThe cross section file will bevalidated for LWR applications using dosimetry measurementsmade in benchmark fields. Such v