1、Designation: E2005 10 (Reapproved 2015)Standard 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 o
2、f revision, 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 facilities and procedures for bench-marking neutron measurements and calcula
3、tions. Particularsections 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
4、 special benchmarkfields 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
5、 be regarded asstandard. 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 Tes
6、t Method for Measuring 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
7、-Neutron ReactionRates by Radioactivation of AluminumE343 Test Method for Measuring Reaction Rates by Analy-sis of Molybdenum-99 Radioactivity From Fission Do-simeters (Withdrawn 2002)3E393 Test Method for Measuring Reaction Rates by Analy-sis of Barium-140 From Fission DosimetersE482 Guide for Appl
8、ication of Neutron Transport Methodsfor 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 b
9、y Radio-activation of Uranium-238E705 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 ReactorSurveillance, E706(IIIB)E910 Test Method for Application and Analysis of HeliumAccu
10、mulation Fluence Monitors for Reactor VesselSurveillance, E706 (IIIC)E1297 Test Method for Measuring Fast-Neutron ReactionRates by Radioactivation of NiobiumE2006 Guide for Benchmark Testing of Light Water ReactorCalculations3. Significance and Use3.1 This guide describes approaches for using neutro
11、n fieldswith well known characteristics to perform calibrations ofneutron sensors, to intercompare different methods ofdosimetry, and to corroborate procedures used to derive neu-tron field information from measurements of neutron sensorresponse.3.2 This guide discusses only selected standard and re
12、fer-ence neutron fields which 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. Thesest
13、andard fields were chosen for their spectral similarity to the1This guide is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility of SubcommitteeE10.05 on Nuclear Radiation Metrology.Current edition approved Oct. 1, 2015. Published Novem
14、ber 2015. Originallyapproved in 1999. Last previous edition approved in 2010 as E2005 - 10. DOI:10.1520/E2005-10R15.2For 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
15、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 19428-2959. United States1high energy region (E 2 MeV) of reactor spectra
16、. The variouscategories 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“vesse
17、l.” When such mockups 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 disc
18、ussed in Guide E2006, and in Refs (1)4and (2).4. Neutron Field Benchmarking4.1 To accomplish neutron field “benchmarking,” one mustperform irradiations in a well-characterized neutronenvironment, with the required level of accuracy established bya sufficient quantity and quality of results supported
19、 by arigorous uncertainty 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, orcontinuity, of procedures t
20、hat are influenced by changes ofpersonnel 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 A
21、STM standardmethods 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
22、 provides a listof such facilities. The emphasis is on facilities that have along-lived commitment to development, maintenance,research, and international interlaboratory comparison calibra-tions. As such, discussion is limited to recently existingfacilities.5.2252Cf Fission SpectrumStandard Neutron
23、 Field:5.2.1 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 In
24、stitute ofStandards 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 ared
25、iscussed in Ref (4). The principal uncertainties, which onlytotal about 2.5 %, come from the source strengthdetermination, scattering corrections, and distance measure-ments. Extensive details of standard field characteristics andvalues of measured and calculated spectrum-averaged crosssections are
26、all given 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 of
27、a special (32 mg)252Cf source employed to measure thespectrum-averaged cross section of the93Nb(n,n) 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 afundamenta
28、l neutron 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) an
29、d one at 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
30、 rate, or 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 corre
31、ctions are 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 associat
32、ed measured 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 Re
33、search Reactor (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 No
34、tationReaction Rate, Fluence Rate, and FluenceThe notation employed in this section will follow that in E261(Standard Practice for Determining Neutron Fluence Rate, and4The boldface numbers given in parentheses refer to a list of references at theend of the text.E2005 10 (2015)2Spectra by Radioactiv
35、ation Techniques) except as noted. Thereaction rate, R, for some neutron-nuclear reaction reactions/(dosimeter target nucleus)(second) is given by:R 5 *oE! E! dE (1)or:R 5 (2)where:(E) = the dosimeter reaction cross section at energy E(typically of the order of 1024cm2),(E) = the differential neutro
36、n fluence rate, that is thefluence per unit time and unit energy for neutronswith energies between E andE+dE(neutronscm2s1MeV1), = the total fluence rate (neutrons cm2s1), the integralof (E) over all E, and = the spectral-averaged value of (E), R/.NOTE 1Neutron fluence and fluence rate are defined f
37、ormally 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 exper
38、imentally 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
39、, there 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 =R/ from Eq 2, then the uncertainty in will be a propagation
40、of the uncertainties in both R and . The uncertainty in isfrequently large, leading to a less accurate determination of than 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 bothi
41、rradiations and writeA5 BRA/RB!B/A! (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 A(E) and B(E) are fairly similar. There mayalso be important ca
42、ncellation of poorly known factors in theratio RA/RB, which will contribute to the better accuracy of Eq3. Whether 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 usefulvalidation of oth
43、er 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 case of a
44、 lightly encapsulated252Cf sintered-oxide bead, which has an emission rate known to about61.5 % by calibration in a manganese bath (MnSO4solution).Further, consider a dosimeter pair irradiated in compensatedbeam geometry (with each member of the pair equidistantfrom, and on opposite sides of, the252
45、Cf source). For such anirradiation in a large room (where very little room returnoccurs), the fluence rate with a252Cf fission spectrum isknown to within 63 % from the source strength, and theaverage distance of the dosimeter pair from the center of thesource. Questions concerning in- and out-scatte
46、ring by sourceencapsulation, source and foil holders, and foil thicknessesmay be accurately investigated by Monte Carlo calculations.There is no other neutron-irradiation situation that can ap-proach this level of accuracy in determination of the fluence orfluence rate.6.2.2 Fluence Transfer Calibra
47、tions 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 disadvantage can be
48、 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 either a
49、252Cf 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 is
