ASTM E2005-2010 Standard Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields《标准和参考中子场中反应堆剂量测定用基准试验的标准指南》.pdf

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1、Designation: E2005 10Standard Guide forBenchmark Testing of Reactor Dosimetry in Standard andReference Neutron Fields1This standard is issued under the fixed designation E2005; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision, the ye

2、ar 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 facilities and procedures for bench-marking neutron measurements and calculations. Particulars

3、ections of the guide discuss: the use of well-characterizedbenchmark neutron fields to calibrate integral neutron sensors;the use of certified-neutron-fluence standards to calibrateradiometric counting equipment or to determine interlaboratorymeasurement consistency; development of special benchmark

4、fields to test neutron transport calculations; use of well-knownfission spectra to benchmark spectrum-averaged cross sections;and the use of benchmarked data and calculations to determinethe uncertainties in derived neutron dosimetry results.1.2 The values stated in SI units are to be regarded assta

5、ndard. No other units of measurement are included in thisstandard.2. Referenced Documents2.1 ASTM Standards:2E170 Terminology Relating to Radiation Measurements andDosimetryE261 Practice for Determining Neutron Fluence, FluenceRate, and Spectra by Radioactivation TechniquesE263 Test Method for Measu

6、ring Fast-Neutron ReactionRates by Radioactivation of IronE264 Test Method for Measuring Fast-Neutron ReactionRates by Radioactivation of NickelE265 Test Method for Measuring Reaction Rates and Fast-Neutron Fluences by Radioactivation of Sulfur-32E266 Test Method for Measuring Fast-Neutron ReactionR

7、ates by Radioactivation of AluminumE343 Test Method for Measuring Reaction Rates by Analy-sis of Molybdenum-99 Radioactivity From Fission Dosim-eters3E393 Test Method for Measuring Reaction Rates by Analy-sis of Barium-140 From Fission DosimetersE482 Guide for Application of Neutron Transport Method

8、sfor Reactor Vessel Surveillance, E706 (IID)E523 Test Method for Measuring Fast-Neutron ReactionRates by Radioactivation of CopperE526 Test Method for Measuring Fast-Neutron ReactionRates by Radioactivation of TitaniumE704 Test Method for Measuring Reaction Rates by Radio-activation of Uranium-238E7

9、05 Test Method for Measuring Reaction Rates by Radio-activation of Neptunium-237E854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for Reactor Sur-veillance, E706(IIIB)E910 Test Method for Application and Analysis of HeliumAccumulation Fluence Monitors for Rea

10、ctor Vessel Sur-veillance, E706 (IIIC)E1297 Test Method for Measuring Fast-Neutron ReactionRates by Radioactivation of NiobiumE2006 Guide for Benchmark Testing of Light Water Reac-tor Calculations3. Significance and Use3.1 This guide describes approaches for using neutron fieldswith well known chara

11、cteristics to perform calibrations ofneutron sensors, to intercompare different methods of dosim-etry, and to corroborate procedures used to derive neutron fieldinformation from measurements of neutron sensor response.3.2 This guide discusses only selected standard and refer-ence neutron fields whic

12、h are appropriate for benchmarktesting of light-water reactor dosimetry. The Standard Fieldsconsidered are neutron source environments that closely ap-proximate the unscattered neutron spectra from252Cf sponta-neous fission and235U thermal neutron induced fission. These1This guide is under the juris

13、diction of ASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility of SubcommitteeE10.05 on Nuclear Radiation Metrology.Current edition approved June 1, 2010. Published August 2010. Originallyapproved in 1999. Last previous edition approved in 2005 as E2005 - 051. D

14、OI:10.1520/E2005-10.2For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Book of ASTMStandards volume information, refer to the standards Document Summary page onthe ASTM website.3Withdrawn. The last approved version of

15、this historical standard is referencedon www.astm.org.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.standard fields were chosen for their spectral similarity to thehigh energy region (E 2 MeV) of reactor spectra. The variouscategor

16、ies of benchmark fields are defined in TerminologyE170.3.3 There are other well known neutron fields that havebeen designed to mockup special environments, such aspressure vessel mockups in which it is possible to makedosimetry measurements inside of the steel volume of the“vessel.” When such mockup

17、s are suitably characterized theyare also referred to as benchmark fields. A variety of theseengineering benchmark fields have been developed, or pressedinto service, to improve the accuracy of neutron dosimetrymeasurement techniques. These special benchmark experi-ments are discussed in Guide E2006

18、, and in Refs (1)4and (2).4. Neutron Field Benchmarking4.1 To accomplish neutron field “benchmarking,” one mustperform irradiations in a well-characterized neutron environ-ment, with the required level of accuracy established by asufficient quantity and quality of results supported by arigorous unce

19、rtainty analysis. What constitutes sufficient re-sults and their required accuracy level frequently depends uponthe situation. For example:4.1.1 Benchmarking to test the capabilities of a new dosim-eter;4.1.2 Benchmarking to ensure long-term stability, or conti-nuity, of procedures that are influenc

20、ed by changes of person-nel and equipment;4.1.3 Benchmarking measurements that will serve as thebasis of intercomparison of results from different laboratories;4.1.4 Benchmarking to determine the accuracy of newlyestablished benchmark fields; and4.1.5 Benchmarking to validate certain ASTM standardme

21、thods or practices which derive exposure parameters (forexample, fluence 1 MeV or dpa) from dosimetry measure-ments and calculations.5. Description of Standard and Reference Fields5.1 There are a few facilities which can provide certified“free field” fluence irradiations. The following provides a li

22、stof such facilities. The emphasis is on facilities that have along-lived commitment to development, maintenance, re-search, and international interlaboratory comparison calibra-tions. As such, discussion is limited to recently existingfacilities.5.2252Cf Fission SpectrumStandard Neutron Field:5.2.1

23、 The standard fission-spectrum fluence from a suitablyencapsulated252Cf source is characterized by its sourcestrength, the distance from the source, and the irradiation time.In the U.S., neutron source emission rate calibrations are allreferenced to source calibrations at the National Institute ofSt

24、andards and Technology (NIST) accomplished by theMnSO4technique (3). Corrections for neutron absorption,scattering, and other than point-geometry conditions may, bycareful experimental design, be held to less than 3 %. Associ-ated uncertainties for the NIST252Cf irradiation facility arediscussed in

25、Ref (4). The principal uncertainties, which onlytotal about 2.5 %, come from the source strength determina-tion, scattering corrections, and distance measurements. Exten-sive details of standard field characteristics and values ofmeasured and calculated spectrum-averaged cross sections areall given

26、in a compendium, see Ref (5).5.2.2 The NIST252Cf sources have a very nearly unper-turbed spontaneous fission spectrum, because of the light-weight encapsulations, fabricated at the Oak Ridge NationalLaboratory (ORNL), see Ref (6).5.2.3 For a comprehensive view of the calibration and use ofa special

27、(32 mg)252Cf source employed to measure thespectrum-averaged cross section of the93Nb(n,n8) reaction, seeRef (7).5.3235U Fission SpectrumStandard Neutron Field:5.3.1 Because235U fission is the principal source of neu-trons in present nuclear reactors, the235U fission spectrum is afundamental neutron

28、 field for benchmark referencing or do-simetry accomplished in reactor environments. This remainstrue even for low-enrichment cores which have up to 30 %burnup.5.3.2 There are currently two235U standard fission spectrumfacilities, one in the thermal column of the NIST ResearchReactor (8) and one at

29、CEN/SCK, Mol, Belgium (9).5.3.3 A standard235U neutron field is obtained by driving(fissioning)235U in a field of thermal neutrons. Therefore, thefluence rate depends upon the power level of the drivingreactor, which is frequently not well known or particularlystable. Time dependent fluence rate, or

30、 total fluence, monitor-ing is necessary in the235U field. Certified fluence irradiationsare monitored with the58Ni(n,p)58Co activation reaction. Thefluence-monitor calibration must be benchmarked.5.3.4 For235U, as for252Cf irradiations, small (nominally 3 %) scattering and absorption corrections ar

31、e necessary. Inaddition, for235U, gradient corrections of the measured fluencewhich do not simply depend upon distance are necessary. Thescattering and gradient corrections are determined by MonteCarlo calculations. Field characteristics of the NIST235UFission Spectrum Facility and associated measur

32、ed and calcu-lated cross sections are given in Ref (5).5.4 There are several additional facilities that can providefree field fluence irradiations that qualify as reference fields.The following is a list of some of the facilities that havecharacterized reference fields:5.4.1 Annular Core Research Re

33、actor (ACRR) Central Cav-ity Reference Neutron Field (10,11),5.4.2 ACRR Lead-Boron Cavity Insert Reference NeutronField (11),5.4.3 YAYOI fast neutron field Reference Neutron Field(12,13),5.4.4 SIGMA-SIGMA neutron field Reference NeutronField (12,13).6. Applications of Benchmark Fields6.1 NotationRea

34、ction Rate, Fluence Rate, and FluenceThe notation employed in this section will follow that in E261(Standard Practice for Determining Neutron Fluence Rate, andSpectra by Radioactivation Techniques) except as noted. The4The boldface numbers given in parentheses refer to a list of references at theend

35、 of the text.E2005 102reaction rate, R, for some neutron-nuclear reaction reactions/(dosimeter target nucleus)(second) is given by:R 5*osE! fE! dE (1)or:R 5s f (2)where:s(E) = the dosimeter reaction cross section at energy E(typically of the order of 1024cm2),f(E) = the differential neutron fluence

36、rate, that is thefluence per unit time and unit energy for neutronswith energies between E andE+dE(neutronscm2s1MeV1),f = the total fluence rate (neutrons cm2s1), the integralof f(E) over all E, ands = the spectral-averaged value of s(E), R/f.NOTE 1Neutron fluence and fluence rate are defined formal

37、ly inTerminology E170 under the listing “particle fluence.” Fluence is just thetime integral of the fluence rate over the time interval of interest. Thefluence rate is also called the flux or flux density in many papers and bookson neutron transport theory.6.1.1 The reaction rate is found experiment

38、ally using anactive instrument such as a fission chamber (see Ref (14)orapassive dosimeter such as a solid state track recorder (see TestMethod E854), a helium accumulation fluence monitor (seeTest Method E910), or a radioactivation dosimeter (see Prac-tice E261). For the radioactivation method, the

39、re are alsoseparate standards for many particularly important dosimetrynuclides, for example, see Test Methods E263, E264, E265,E266, E343, E393, E523, E526, E704, E705, and E1297.6.2 Fluence Rate Transfer: Note that if one determines f =R/s from Eq 2, then the uncertainty in f will be a propagation

40、of the uncertainties in both R and s. The uncertainty in sisfrequently large, leading to a less accurate determination of fthan desired. However, if one can make an additional irradia-tion of the same type of dosimeter in a standard neutron fieldwith known fluence rate, then one may apply Eq 2 to bo

41、thirradiations and writefA5fBRA/RB!sB/sA! (3)where “A” denotes the field of interest and “B” denotes thestandard neutron field benchmark. In Eq 3 the ratios of spectralaverage cross section, will have a small uncertainty if thespectral shapes fA(E) and fB(E) are fairly similar. There mayalso be impo

42、rtant cancellation of poorly known factors in theratio RA/RB, which will contribute to the better accuracy of Eq3. Whether f is better determined by Eq 3 or Eq 2 must beevaluated on a case by case basis. Often the fluence rate fromEq 3 is substantially more accurate and provides a very usefulvalidat

43、ion of other dosimetry. The use of a benchmark neutronfield irradiation and Eq 3 is called fluence rate transfer.6.2.1 Certified Fluence or Fluence Rate IrradiationsTheprimary benefit from carefully-made irradiations in a standardneutron field is that of knowing the neutron fluence rate.Consider the

44、 case of a lightly encapsulated252Cf sintered-oxidebead, which has an emission rate known to about 61.5 % bycalibration in a manganese bath (MnSO4solution). Further,consider a dosimeter pair irradiated in compensated beamgeometry (with each member of the pair equidistant from, andon opposite sides o

45、f, the252Cf source). For such an irradiationin a large room (where very little room return occurs), thefluence rate with a252Cf fission spectrum is known towithin 63 % from the source strength, and the averagedistance of the dosimeter pair from the center of the source.Questions concerning in- and o

46、ut-scattering by source encap-sulation, source and foil holders, and foil thicknesses may beaccurately investigated by Monte Carlo calculations. There isno other neutron-irradiation situation that can approach thislevel of accuracy in determination of the fluence or fluencerate.6.2.2 Fluence Transfe

47、r Calibrations of Reference FieldsThe benefit of irradiating with a source of known emission rateis lost when one must consider reactor cores or, even, thermal-neutron fissioned235U sources. When the latter are carefullyconstructed to provide for an unmoderated235U spectrum, thismentioned disadvanta

48、ge can be circumvented by a processcalled fluence transfer. As explained briefly in 6.2, this processis basically as follows. A gamma-counter (spectrometer) ge-ometry is chosen to enable proper counting of the activities ofa particular isotopic reaction for example,58Ni(n,p)58Co, afterirradiation in

49、 either a252Cf or235U field. Then the252Cfirradiation is accomplished and the nickel foil counted. Fromthis, a ratio of the dosimeter response divided by the252Cfcertified fluence is determined. Subsequently, an identicalnickel is irradiated in the235U spectrum and that foil is countedwith the same counter geometry. Within the knowledge of theratio of the spectrum average cross sections in the two spectra,knowledge of the counter response to the recent irradiationyields the average235U fluence. Note, the average fluence ismeasured. The therma