ASTM E266-17 Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Aluminum.pdf

1、Designation: E266 17Standard Test Method forMeasuring Fast-Neutron Reaction Rates by Radioactivationof Aluminum1This standard is issued under the fixed designation E266; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision, the year of l

2、ast 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 test method covers procedures measuring reactionrates by the activation reaction27Al(n,)24Na.1.2 This activation r

3、eaction is useful for measuring neutronswith energies above approximately 6.5 MeV and for irradiationtimes up to about 2 days (for longer irradiations, or when thereare significant variations in reactor power during theirradiation, see Practice E261).1.3 With suitable techniques, fission-neutron flu

4、ence ratesabove 106cm2s1can be determined.1.4 Detailed procedures for other fast neutron detectors arereferenced in Practice E261.1.5 This standard 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 establis

5、h appro-priate safety, health, and environmental practices and deter-mine the applicability of regulatory limitations prior to use.1.6 This international standard was developed in accor-dance with internationally recognized principles on standard-ization established in the Decision on Principles for

6、 theDevelopment of International Standards, Guides and Recom-mendations issued by the World Trade Organization TechnicalBarriers to Trade (TBT) Committee.2. Referenced Documents2.1 ASTM Standards:2E170 Terminology Relating to Radiation Measurements andDosimetryE177 Practice for Use of the Terms Prec

7、ision and Bias inASTM Test MethodsE181 Test Methods for Detector Calibration and Analysis ofRadionuclidesE261 Practice for Determining Neutron Fluence, FluenceRate, and Spectra by Radioactivation TechniquesE456 Terminology Relating to Quality and StatisticsE844 Guide for Sensor Set Design and Irradi

8、ation forReactor SurveillanceE944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor SurveillanceE1005 Test Method for Application and Analysis of Radio-metric Monitors for Reactor Vessel SurveillanceE1018 Guide for Application of ASTM Evaluated CrossSection Data File3. Termino

9、logy3.1 Definitions:3.1.1 Refer to Terminologies E170 and E456.4. Summary of Test Method4.1 High-purity aluminum is irradiated in a neutron field,thereby producing radioactive24Na from the27Al(n,)24Naactivation reaction.4.2 The gamma rays emitted by the radioactive decay of24Na are counted (see Test

10、 Methods E181) and the reactionrate, as defined by Practice E261, is calculated from the decayrate and irradiation conditions.4.3 The neutron fluence rate above about 6.5 MeV can thenbe calculated from the spectral-weighted neutron activationcross section as defined by Practice E261.5. Significance

11、and Use5.1 Refer to Guide E844 for the selection, irradiation, andquality control of neutron dosimeters.5.2 Refer to Practice E261 for a general discussion of thedetermination of fast-neutron fluence rate with threshold de-tectors.5.3 Pure aluminum in the form of foil or wire is readilyavailable and

12、 easily handled.27Al has an abundance of 100 %(1)3.1This test method is under the jurisdiction ofASTM Committee E10 on NuclearTechnology and Applicationsand is the direct responsibility of SubcommitteeE10.05 on Nuclear Radiation Metrology.Current edition approved Aug. 1, 2017. Published October 2017

13、. Originallyapproved in 1965. Last previous edition approved in 2011 as E266 11. DOI:10.1520/E0266-17.2For 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

14、Document Summary page onthe ASTM website.3The boldface numbers in parentheses refer to a list of References at the end ofthis standard.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United StatesThis international standard was developed in accorda

15、nce with internationally recognized principles on standardization established in the Decision on Principles for theDevelopment of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.146 5.424Na has a half-life of 14.9

16、58 (2)4h (2) and emitsgamma rays with energies of 1.368630 (5) and 2.754049 (5)MeV(2).5.5 Fig. 1 shows a plot of the Russian Reactor DosimetryFile (RRDF) cross section (3, 4) versus neutron energy for thefast-neutron reaction27Al(n,)24Na (3) along with a compari-son to the current experimental datab

17、ase (5, 6). This RRDF-2008 cross section is identical to what is found in the latestInternational Atomic Energy Agency (IAEA) InternationalReactor Dosimetry and Fusion File, IRDFF-1.05 (7). While theRRDF-2008 and IRDFF-1.05 cross sections extend fromthreshold up to 60 MeV, due to considerations of t

18、he availablevalidation data, the energy region over which this standardrecommends use of this cross section for reactor dosimetryapplications only extends from threshold at 4.25 MeV up to20 MeV. This figure is for illustrative purposes and is used toindicate the range of response of the27Al(n,) reac

19、tion. Referto Guide E1018 for recommended sources for the tabulateddosimetry cross sections.5.6 Two competing activities,28Al and27Mg, are formed inthe reactions27Al(n,)28Al and27Al(n,p)27Mg, respectively,but these can be eliminated by waiting 2 h before counting.6. Apparatus6.1 NaI(T1) or High Reso

20、lution Gamma-Ray Spectrometer.Because of its high resolution, the germanium detector isuseful when contaminant activities are present (see Test Meth-ods E181 and E1005).6.2 Precision Balance, able to achieve the required accu-racy.7. Materials7.1 The purity of the aluminum is important. No impuritie

21、sshould be present that produce long-lived gamma-ray-emittingradionuclides having gamma-ray energies that interfere withthe24Na determination. Discard aluminum that contains suchimpurities or that contains quantities of23Na sufficient tointerfere, through thermal-neutron capture, with24Na determi-na

22、tion. The presence of these impurities should be determinedby activation analysis since spectrographically pure aluminummay contain a contaminant not detectable by the emissionspectrograph. If the24Na content of the irradiated samples isdetermined from the emission rate of the 2.754049 MeVgamma ray,

23、 the probability of interference from contaminantgamma rays is much less than if the 1.368630 MeV gamma rayis used.7.2 Encapsulating MaterialsBrass, stainless steel, copper,aluminum, quartz, or vanadium have been used as primaryencapsulating materials. The container should be constructedin such a ma

24、nner that it will not create significant fluxperturbation and that it may be opened easily, especially if thecapsule is to be opened remotely (see Guide E844).8. Procedure8.1 Decide on the size and shape of aluminum sample to beirradiated. This is influenced by the irradiation space and theexpected

25、production of24Na. Calculate the expected produc-tion rate of24Na from the activation equation described inSection 9, and adjust sample size and irradiation time so thatthe24Na may be accurately counted. A trial irradiation isrecommended.8.2 Determine a suitable irradiation time (see 8.1). Since24Na

26、 has a 14.958 h half-life, the24Na activity will approachequilibrium after a day of irradiation.8.3 Weigh the sample.8.4 Irradiate the sample for the predetermined time period.Record the power level and any changes in power during theirradiation, the time at the beginning and end of the irradiation,

27、and the relative position of the monitors in the irradiationfacility.8.5 After irradiation, the sample should be thoroughlyrinsed in warm water. This will remove any24Na surfacecontamination produced during irradiation.8.6 Check the sample for activity from cross-contaminationby other irradiated mat

28、erials. Clean, if necessary, and reweigh.8.7 Analyze the sample for24Na content in disintegrationsper second using the gamma-ray spectrometer after the28Aland27Mg have decayed (1 to 2 h will usually suffice) or untilthe contaminant activities, if any, have decayed (see TestMethods E181 and E1005).8.

29、8 Disintegration of24Na nuclei produces 1.368630-MeVand 2.754049-MeVgamma rays with probabilities per decay of0.999934 (5) and 0.99862 (3), respectively(2). When analyzingeither gamma-ray peak, a correction for coincidence summingmay be required if the sample is placed close to the detector (10cm or

30、 less) (see Test Methods E181).8.9 If any question exists as to the purity of the gamma raybeing counted, the sample should be counted periodically todetermine if the decay follows the 14.958-h half-life of24Na(2).4The value of uncertainty, in parenthesis, refers to the corresponding last digits,thu

31、s 14.958 (2) corresponds to 14.958 6 0.002.FIG. 127Al(n,)24Na Cross Section, from RRDF-2008/IRDFF-1.05Library, with EXFOR Experimental DataE266 17246 9. Calculations9.1 Calculate the saturation activity As, as follows:As5 A/1 2 exp2ti#!exp2tw#!# (1)where:A =24Na disintegrations per second measured b

32、y counting, = decay constant for24Na = 1.287210 105s1,ti= irradiation duration, s, andtw= elapsed time between the end of irradiation andcounting, s.NOTE 1The equation Asis valid if the reactor operated at essentiallyconstant power and if corrections for other reactions (for example,impurities, burn

33、out, etc.) are negligible. Refer to Practice E261 for moregeneralized treatments.9.2 Calculate the reaction rate, Rs, as follows:R 5 As/No(2)where:As= saturation activity, andNo= number of27Al atoms.9.3 Refer to Practice E261 and Guide E944 for a discussionof fast-neutron fluence rate and fluence.10

34、. Report10.1 Practice E261 describes how data should be reported.11. Precision and BiasNOTE 2Measurement uncertainty is described by a precision and biasstatement in this standard. Another acceptable approach is to use Type Aand B uncertainty components (8, 9). This Type A/B uncertainty specifi-cati

35、on is now used in International Organization for Standardization (ISO)standards and this approach can be expected to play a more prominent rolein future uncertainty analyses.11.1 Precision and bias in this standard are treated inaccordance with the definitions in Practice E177. Generalpractice indic

36、ates that disintegration rates can be determinedwith bias of 6 3 % (1S %) and with a precision of 6 1%(1S %).11.2 The energy-dependent uncertainty, expressed as a per-centage of the baseline cross section, for the27Al(n,)24Nacross section is shown in Fig. 2.(3)11.3 Test results have been reported in

37、 neutron benchmarkfields.11.3.1 In the252Cf spontaneous fission reference neutronfield, the measured cross section is 1.016 mb 6 1.47 % (10)and the calculated cross section using the RRDF-2008 crosssection is 1.0184 mb with a spectrum integrated cross sectionundertainty of 0.7158 % (3) and a spectru

38、m characterizationuncertainty of 1.609 % (11). This results in a calculated-to-experimental (C/E) ratio of 1.00236 6 1.761 %.11.3.2 In the235U thermal neutron field, as characterized inthe JENDL 4.0 nuclear data file, the measured cross section is0.7007 mb 6 1.28 % (10) and the calculated cross sect

39、ionusing the RRDF-2008 cross section is 0.7000346 mb with aspectrum integrated cross section uncertainty of 0.37504957 %(4) and a spectrum characterization uncertainty of 13.59351 %(11). This results in a calculated-to-experimental (C/E) ratio of0.99905 6 13.65364 %.12. Keywords12.1 activation; acti

40、vation reaction; aluminum; cross sec-tion; dosimetry; fast-neutron monitor; neutron metrology; pres-sure vessel surveillance; reaction rateREFERENCES(1) Nuclear Wallet Cards, compiled by J. K. Tuli, National Nuclear DataCenter, 8th edition, October 2011. This is the last printed edition ofthis datab

41、ase and the natural abundance values here for27Al agreewith the on-line version as of January 2, 2017.(2) Be, M. M., Chiste, V., Dulieu, C., Browne, E., et al, Table ofRadionuclides (Vol 1 A = 1 to 150), Monographie BIPM-5, BureauInternational des Poids et Mesures (BIPM), France, 2004, http:/www.nuc

42、leide.org/DDEP_WG/DDEPdata.htm (data current as of Jan.2, 2017).(3) Zolotarev, K. I., Ignatyuk,A. V., Mahokhin, V. N., Pashchenco,A. B.,“RRDF-98 Russian Reactor Dosimetry File”, report IAEA-NDS-193,March 1999. The last full release was in 1998. Updated versionsreferenced here corresponding to the RR

43、DF-2008 library.(4) Zolotarev, K. I., Evaluation of Cross-Section Data from Threshold to40-60 MeV for Specific Neutron Reactions Important for NuetronDosimetry Applications, Part 1 Evaluation of the excitation functionsfor the27Al(n,a)24Na,55Mn(N,2n)54Mn,59Co(n,p)59Fe,59Co(n,2n)58m+gCo and90Zr(n,2n)

44、89m+gZr reactions, report INDC(NDS)-0546, International Atomic Energy Agency, Vienna, Austria, April2009.(5) “EXFOR Formats Description for Users (EXFOR Basics)”, reportIAEA-NDS-206, International Atomic Energy Agency, Vienna,Austria, June 2008. On-line database available at: http:/www-nds.iaea.org/

45、indg_nexp.html. Data here as present on January 3, 2011.(6) Otuka, N., Dupont, E. Semkove, V., Pritychenko, B., Blokin, A. I.,Aikawa, M., et al, Towards a more Complete and Accurate Experi-mental Nuclear Reaction Data Library (EXFOR): International Col-laboration Between Nuclear Reaction Data Centre

46、s (NRDC), NuclearData Sheets, Vol 120, pp. 272-276, June 2014, URL for EXFORdatabase: https:/www-nds.iaea.org/exfor/exfor.htmFIG. 2 Energy-dependent Uncertainty (%) for the RRDF-2008/IRDFF-1.0527Al(n,)24Na Cross SectionE266 17346 (7) Capote, R., Zolotarev, K., Pronyaev, V., Trkov, A., “Updating andE

47、xtending the IRDF-2002 Dosimetry Library,” Journal of ASTMInternational, Vol 9, Issue 4, April 2012, http:/www.astm.org/DIGITAL_LIBRARY/JOURNALS/JAI/PAGES/JAI104119.htm.Tabulated data available at: https:/www-nds.iaea.org/IRDFF/(8) Guide to the Expression of Uncertainty in Measurement, International

48、Organization for Standardization, 1995, ISBN 92-67-10188-9.(9) Taylor, B. N., Kuyatt, C. E., Guidelines for Evaluating and Expressingthe Uncertainty of NIST Measurement Results, NIST Technical Note1297, National Institute of Standards and Technology, Gaithersbrug,MD, 1994.(10) Mannhart, W., Validati

49、on of Differential Cross Sections with IntegralData, Report INDC(NDS)-435, pp. 59-64, IAEA, Vienna, September2002.(11) Griffin, P. J., Peters, C. D., Vehar, D. W., Recommended NeutronDosimetry Cross Sections for the Characterization of Neutron Fields,IEEE Transactions on Nuclear Science, Vol 59, pp. 1167-1174,August 2012.ASTM International takes no position respecting the validity of any patent rights asserted in connection with any item mentionedin this standard. Users of this standard are expressly advised that determination of th