ASTM E2005-2005e1 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的基准检测的标准指南》.pdf

《ASTM E2005-2005e1 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的基准检测的标准指南》.pdf》由会员分享，可在线阅读，更多相关《ASTM E2005-2005e1 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的基准检测的标准指南》.pdf（6页珍藏版）》请在麦多课文档分享上搜索。

1、Designation: E 2005 05e1Standard Guide forBenchmark Testing of Reactor Dosimetry in Standard andReference Neutron Fields1This standard is issued under the fixed designation E 2005; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision, th

2、e year of 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.e1NOTEEditorial changes were made in June 2005.1. Scope1.1 This guide covers facilities and procedures for bench-markin

3、g neutron measurements and calculations. 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 interlaboratorymeasu

4、rement consistency; development of 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.2. Ref

5、erenced Documents2.1 ASTM Standards:2E 170 Terminology Relating to Radiation Measurementsand DosimetryE 261 Practice for Determining Neutron Fluence Rate, andSpectra by Radioactivation TechniquesE 263 Test Methods for Measuring Fast-Neutron ReactionRates by Radioactivation of IronE 264 Test Methods

6、for Measuring Fast-Neutron ReactionRates by Radioactivation of NickelE 265 Test Methods for Measuring Fast-Neutron ReactionRates by Radioactivation of Sulfur-32E 266 Test Methods for Measuring Fast-Neutron ReactionRates by Radioactivation of AluminumE 343 Test Methods for Measuring Reaction Rates by

7、Analysis of Molybdenum 99 Activity from Fission Dosim-etersE 393 Test Methods for Measuring Reaction Rates byAnalysis of Barium-140 from Fission DosimetersE 482 Guide for Application of Neutron Transport Methodsfor Reactor Vessel Surveillance, E 706 (IID)E 523 Test Methods for Measuring Fast-Neutron

8、 ReactionRates by Radioactivation of CopperE 526 Test Methods for Measuring Fast-Neutron ReactionRates by Radioactivation of TitaniumE 704 Test Methods for Measuring Fast-Neutron ReactionRates by Radioactivation of Uranium-238E 705 Test Methods for Measuring Fast-Neutron ReactionRates by Radioactiva

9、tion of Neptunium-237E 854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for Reactor Sur-veillance, E 706 (IIIB)E 910 Test Method for Application and Analysis of HeliumAccumulation Fluence Monitors for Reactor Vessel Sur-veillance, E 706 (IIIC)E 1297 Test Meth

10、od for Measuring Fast-Neutron ReactionRates by Radioactivation of NiobiumE 2006 Guide for Benchmark Testing of Light Water Reac-tor Calculations3. Significance and Use3.1 This guide describes approaches for using neutron fieldswith well known characteristics to perform calibrations ofneutron sensors

11、, 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 which are appropriate for benchmarktesting of light-wate

12、r 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. Thesestandard fields were chosen for their spectral similarity to thehigh energy region

13、 (E 2 MeV) of reactor spectra. The variouscategories of benchmark fields are defined in TerminologyE 170.3.3 There are other well known neutron fields that havebeen designed to mockup special environments, such aspressure vessel mockups in which it is possible to make1This guide is under the jurisdi

14、ction of ASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility of SubcommitteeE10.05 on Nuclear Radiation Metrology .Current edition approved Jan. 1, 2005. Published February 2005. Originallyapproved in 1999. Last previous edition approved in 1999 as E 2005 - 99.2

15、For 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.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C

16、700, West Conshohocken, PA 19428-2959, United States.dosimetry measurements inside of the steel volume of the“vessel.” 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

17、, to improve the accuracy of neutron dosimetrymeasurement techniques. These special benchmark experi-ments are discussed in Guide E 2006, and in Refs (1)3and (2).4. Neutron Field Benchmarking4.1 To accomplish neutron field “benchmarking,” one mustperform irradiations in a well-characterized neutron

18、environ-ment, with the required level of accuracy established by asufficient quantity and quality of results supported 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 tes

19、t the capabilities of a new dosim-eter;4.1.2 Benchmarking to ensure long-term stability, or conti-nuity, of procedures that are influenced 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.

20、4 Benchmarking to determine the accuracy of newlyestablished benchmark fields; and4.1.5 Benchmarking to validate certain ASTM standardmethods or practices which derive exposure parameters (forexample, fluence 1 MeV or dpa) from dosimetry measure-ments and calculations.5. Description of Standard and

21、Reference Fields5.1 There are a few facilities which can provide certified“free field” fluence irradiations. The following provides a listof such facilities. The emphasis is on facilities that have along-lived commitment to development, maintenance, re-search, and international interlaboratory compa

22、rison calibra-tions. As such, discussion is limited to recently existingfacilities.5.2252Cf Fission SpectrumStandard Neutron 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 irradia

23、tion time.In the U.S., neutron source emission rate calibrations are allreferenced to source calibrations at the National Institute ofStandards and Technology (NIST) accomplished by theMnSO4technique (3). Corrections for neutron absorption,scattering, and other than point-geometry conditions may, by

24、careful experimental design, be held to less than 3 %. Associ-ated uncertainties for the NIST252Cf irradiation facility arediscussed in Ref (4). The principal uncertainties, which onlytotal about 2.5 %, come from the source strength determina-tion, scattering corrections, and distance measurements.

25、Exten-sive details of standard field characteristics and values ofmeasured and calculated spectrum-averaged cross sections areall 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, f

26、abricated at the Oak Ridge NationalLaboratory (ORNL), see Ref (6).5.2.3 For a comprehensive view of the calibration and use ofa special (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.

27、1 Because235U fission is the principal source of neu-trons in present nuclear reactors, the235U fission spectrum is afundamental 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.

28、2 There are currently two235U standard fission spectrumfacilities available, one in the thermal column of the NISTResearch Reactor (8) and 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

29、rate depends upon the power level of the drivingreactor, which is frequently not well known or particularlystable. Time dependent fluence rate, or total fluence, monitor-ing is necessary in the235U field. Certified fluence irradiationsare monitored with the58Ni(n,p)58Co activation reaction. Thefluen

30、ce-monitor calibration must be benchmarked.5.3.4 For235U, as for252Cf irradiations, small (nominally1 MeV) was 2.7 31011neutrons cm2s1and the neutron-to-gamma fluence ratiois 0.35 (dimensionless).6. Applications of Benchmark Fields6.1 NotationReaction Rate, Fluence Rate, and FluenceThe notation empl

31、oyed in this section will follow that in E 261(Standard Practice for Determining Neutron Fluence Rate, andSpectra by Radioactivation Techniques) except as noted. Thereaction rate, R, for some neutron-nuclear reaction reactions/(dosimeter target nucleus)(second) is given by:R 5*osE! fE! dE (1)or:R 5s

32、 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 rate, that is thefluence per unit time and unit energy for neutronswith energies between E andE+dE(neutronscm2s1MeV1),f = the total fluence rate (neutrons cm

33、2s1), 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 formally inTerminology E 170 under the listing “particle fluence.” Fluence is just thetime integral of the fluence rate over the time interval of interest. Theflue

34、nce 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 experimentally using anactive instrument such as a fission chamber (see Ref (12)orapassive dosimeter such as a solid state track recorder (see TestMethod E 854), a he

35、lium accumulation fluence monitor (seeTest Method E 910), or a radioactivation dosimeter (see Prac-tice E 261). For the radioactivation method, there are alsoseparate standards for many particularly important dosimetrynuclides, for example, see Test Methods E 263, E 264, E 265,E 266, E 343, E 393, E

36、 523, E 526, E 704, E 705, and E 1297.6.2 Fluence Rate Transfer: Note that if one determines f =R/s from Eq 2, then the uncertainty in f will be a propagationof the uncertainties in both R and s. The uncertainty in sisfrequently large, leading to a less accurate determination of fthan desired. Howev

37、er, 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 bothirradiations and writefA5fBRA/RB!sB/sA! (3)where “A” denotes the field of interest and “B” denotes thestandard neutron field benchmark. In E

38、q 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 important cancellation of poorly known factors in theratio RA/RB, which will contribute to the better accuracy of Eq3. Whether f is better determi

39、ned 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 other dosimetry. The use of a benchmark neutronfield irradiation and Eq 3 is called fluence rate transfer.6.2.1 Certified Fluence or Fl

40、uence Rate IrradiationsTheprimary benefit from carefully-made irradiations in a standardneutron field is that of knowing the neutron fluence rate.Consider the case of a lightly encapsulated252Cf sintered-oxidebead, which has an emission rate known to about 61.5 % bycalibration in a manganese bath (M

41、nSO4solution). Further,consider a dosimeter pair irradiated in compensated beamgeometry (with each member of the pair equidistant from, andon opposite sides of, the252Cf source). For such an irradiationin a large room (where very little room return occurs), thefluence rate with a252Cf fission spectr

42、um 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 out-scattering by source encap-sulation, source and foil holders, and foil thicknesses may beaccurately investigated by Monte Carlo calculation

43、s. There isno other neutron-irradiation situation that can approach thislevel of accuracy in determination of the fluence or fluencerate.6.2.2 Fluence Transfer Calibrations of Reference FieldsThe benefit of irradiating with a source of known emission rateis lost when one must consider reactor cores

44、or, even, thermal-neutron fissioned235U sources. When the latter are carefullyconstructed to provide for an unmoderated235U spectrum, thismentioned disadvantage can be circumvented by a processcalled fluence transfer. As explained briefly in 6.2, this processis basically as follows. A gamma-counter

45、(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 a252Cf or235U field. Then the252Cfirradiation is accomplished and the nickel foil counted. Fromthis, a ratio of the dosimeter response

46、 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

47、 to the recent irradiationyields the average235U fluence. Note, the average fluence ismeasured. The thermal fluence rate at the235U sources may nothave been constant over the time of the irradiation but that timeis assumed to be short relative to the 70 day half-life of theE200505e1358Co, which moni

48、tors the fast neutron fluence through-out theirradiation. The method of calibration is termed fluence ratetransfer because it is fluence rate which is determined, andthere is no need to determine the absolute radioactivity of thedosimeters. Relative response of the same counter geometry isthe only r

49、equirement.6.2.3 Reactor IrradiationsIn principle, the same fluence-transfer procedures can be applied to more complex irradia-tions. However, there are certain other situations which mustbe considered and weighed to determine if fluence transfer orreaction rate determination is the better method. Also remem-ber that error estimation can be examined by using bothmethods.6.2.3.1 If radioactivation dosimeters are employed for longterm irradiations in a power reactor, the fluence at a dosimeterlocation can be determined by the method explained in 9.7,L