1、Designation: E266 11E266 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 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 test method covers procedures measuring reaction rates by the activation reaction 27Al(n,)24 Na.1.2 This ac
3、tivation reaction is useful for measuring neutrons with energies above approximately 6.5 MeV and for irradiationtimes up to about 2 days (for longer irradiations, or when there are significant variations in reactor power during the irradiation,see Practice E261).1.3 With suitable techniques, fission
4、-neutron fluence rates above 106 cm2s1 can be determined.1.4 Detailed procedures for other fast neutron detectors are referenced in Practice E261.1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibilityof the user of this sta
5、ndard to establish appropriate safety safety, health, and healthenvironmental practices and determine theapplicability of regulatory limitations prior to use.1.6 This international standard was developed in accordance with internationally recognized principles on standardizationestablished in the De
6、cision on Principles for the Development of International Standards, Guides and Recommendations issuedby the World Trade Organization Technical Barriers to Trade (TBT) Committee.2. Referenced Documents2.1 ASTM Standards:2E170 Terminology Relating to Radiation Measurements and DosimetryE177 Practice
7、for Use of the Terms Precision and Bias in ASTM Test MethodsE181 Test Methods for Detector Calibration and Analysis of RadionuclidesE261 Practice for Determining Neutron Fluence, Fluence Rate, and Spectra by Radioactivation TechniquesE456 Terminology Relating to Quality and StatisticsE844 Guide for
8、Sensor Set Design and Irradiation for Reactor SurveillanceE944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor SurveillanceE1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel SurveillanceE1018 Guide for Application of ASTM Evaluated Cross
9、 Section Data File3. Terminology3.1 Definitions:3.1.1 Refer to TerminologyTerminologies E170 and E456.4. Summary of Test Method4.1 High-purity aluminum is irradiated in a neutron field, thereby producing radioactive 24Na from the 27Al(n,)24Na activationreaction.4.2 The gamma rays emitted by the radi
10、oactive decay of 24Na are counted (see Test Methods E181) and the reaction rate, asdefined by Practice E261, is calculated from the decay rate and irradiation conditions.1 This test method is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applicationsand is the direct respons
11、ibility of Subcommittee E10.05 onNuclear Radiation Metrology.Current edition approved June 1, 2011Aug. 1, 2017. Published June 2011October 2017. Originally approved in 1965. Last previous edition approved in 20072011 asE266 07.E266 11. DOI: 10.1520/E0266-11.10.1520/E0266-17.2 For referencedASTM stan
12、dards, visit theASTM website, www.astm.org, or contactASTM Customer Service at serviceastm.org. For Annual Book of ASTM Standardsvolume information, refer to the standards Document Summary page on the ASTM website.This document is not an ASTM standard and is intended only to provide the user of an A
13、STM standard an indication of what changes have been made to the previous version. Becauseit may not be technically possible to adequately depict all changes accurately, ASTM recommends that users consult prior editions as appropriate. In all cases only the current versionof the standard as publishe
14、d by ASTM is to be considered the official document.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States14.3 The neutron fluence rate above about 6.5 MeV can then be calculated from the spectral-weighted neutron activation crosssection as
15、defined by Practice E261.5. Significance and Use5.1 Refer to Guide E844 for the selection, irradiation, and quality control of neutron dosimeters.5.2 Refer to Practice E261 for a general discussion of the determination of fast-neutron fluence rate with threshold detectors.5.3 Pure aluminum in the fo
16、rm of foil or wire is readily available and easily handled. 27Al has an abundance of 100 % (1)3.5.4 24Na has a half-life of 14.9574 14.958 (2)4 h (2) and emits gamma rays with energies of 1.368626 and 2.754007 1.368630(5) and 2.754049 (5) MeV(2).5.5 Fig. 1 shows a plot of the Russian Reactor Dosimet
17、ry File (RRDF) cross section (3, 4)versus neutron energy for thefast-neutron reaction 27Al(n,)24Na (3) along with a comparison to the current experimental database (45, 6). This RRDF-2008cross section is identical to what is found in the latest International Atomic Energy Agency (IAEA) International
18、 ReactorDosimetry and Fusion File, IRDFF-1.05 (7). While the RRDF-2008 and IRDFF-1.05 cross sections extend from threshold up to60 MeV, due to considerations of the available validation data, the energy region over which this standard recommends use of thiscross section for reactor dosimetry applica
19、tions only extends from threshold at 4.25 MeV up to 20 MeV. This figure is forillustrative purposes only and is used to indicate the range of response of the 27Al(n,) reaction. Refer to Guide E1018 fordescriptions of recommended recommended sources for the tabulated dosimetry cross sections.5.6 Two
20、competing activities, 28Al and 27Mg, are formed in the reactions 27Al(n,) 28Al and 27Al(n,p) 27Mg, respectively, butthese can be eliminated by waiting 2 h before counting.6. Apparatus6.1 NaI(T1) or High Resolution Gamma-Ray Spectrometer. Because of its high resolution, the germanium detector is usef
21、ulwhen contaminant activities are present (see Test Methods E181 and E1005).6.2 Precision Balance, able to achieve the required accuracy.7. Materials7.1 The purity of the aluminum is important. No impurities should be present that produce long-lived gamma-ray-emittingradionuclides having gamma-ray e
22、nergies that interfere with the 24Na determination. Discard aluminum that contains suchimpurities or that contains quantities of 23Na sufficient to interfere, through thermal-neutron capture, with 24Na determination. Thepresence of these impurities should be determined by activation analysis since s
23、pectrographically pure aluminum may contain acontaminant not detectable by the emission spectrograph. If the 24Na content of the irradiated samples is determined from theemission rate of the 2.7540072.754049 MeV gamma ray, the probability of interference from contaminant gamma rays is muchless than
24、if the 1.3686261.368630 MeV gamma ray is used.7.2 Encapsulating MaterialsBrass, stainless steel, copper, aluminum, quartz, or vanadium have been used as primaryencapsulating materials. The container should be constructed in such a manner that it will not create significant flux perturbationand that
25、it may be opened easily, especially if the capsule is to be opened remotely (see Guide E844).3 The boldface numbers in parentheses refer to a list of References at the end of this standard.4 The value of uncertainty, in parenthesis, refers to the corresponding last digits, thus 14.958 (2) correspond
26、s to 14.958 6 0.002.FIG. 1 27Al(n,)24Na Cross Section Section, from RRDF-2008/IRDFF-1.05 Library, with EXFOR Experimental DataE266 1728. Procedure8.1 Decide on the size and shape of aluminum sample to be irradiated. This is influenced by the irradiation space and theexpected production of 24Na. Calc
27、ulate the expected production rate of 24Na from the activation equation described in Section 9,and adjust sample size and irradiation time so that the 24Na may be accurately counted. A trial irradiation is recommended.8.2 Determine a suitable irradiation time (see 8.1).Since 24Na has a 14.957414.958
28、 h half-life, the 24Na activity will approach equilibrium 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, and the r
29、elative position of the monitors in the irradiation facility.8.5 After irradiation, the sample should be thoroughly rinsed in warm water. This will remove any 24Na surface contaminationproduced during irradiation.8.6 Check the sample for activity from cross-contamination by other irradiated material
30、s. Clean, if necessary, and reweigh.8.7 Analyze the sample for 24Na content in disintegrations per second using the gamma-ray spectrometer after the 28Al and27Mg have decayed (1 to 2 h will usually suffice) or until the contaminant activities, if any, have decayed (see Test Methods E181and E1005).8.
31、8 Disintegration of 24Na nuclei produces 1.368626-MeV1.368630-MeV and 2.754007-MeV2.754049-MeV gamma rays withprobabilities per decay of 0.999935 and 0.99872, 0.999934 (5) and 0.99862 (3), respectively(2). When analyzing either gamma-raypeak, a correction for coincidence summing may be required if t
32、he sample is placed close to the detector (10 cm or less) (see TestMethods E181).8.9 If any question exists as to the purity of the gamma ray being counted, the sample should be counted periodically todetermine if the decay follows the 14.9574-h14.958-h half-life of 24Na(2).9. Calculations9.1 Calcul
33、ate the saturation activity As, as follows:As5A/12exp2ti#!exp2tw#!# (1)where:A = 24Na disintegrations per second measured by counting, = decay constant for 24Na = 1.287262 105 s1, = decay constant for 24Na = 1.287210 105 s1,ti = irradiation duration, s, andtw = elapsed time between the end of irradi
34、ation and counting, s.NOTE 1The equation As is valid if the reactor operated at essentially constant power and if corrections for other reactions (for example, impurities,burnout, etc.) are negligible. Refer to Practice E261 for more generalized treatments.9.2 Calculate the reaction rate, Rs, as fol
35、lows:Rs5As/No (2)R 5As/No (2)where:As = saturation activity, andNo = number of 27Al atoms.9.3 Refer to Practice E261 and Guide E944 for a discussion of fast-neutron fluence rate and fluence.10. Report10.1 Practice E261 describes how data should be reported.11. Precision and BiasNOTE 2Measurement unc
36、ertainty is described by a precision and bias statement in this standard. Another acceptable approach is to use Type A andB uncertainty components (58, 69). This TypeA/B uncertainty specification is now used in International Organization for Standardization (ISO) standardsand this approach can be ex
37、pected to play a more prominent role in future uncertainty analyses.11.1 Precision and bias in this standard are treated in accordance with the definitions in Practice E177. General practice indicatesthat disintegration rates can be determined with bias of 6 3 % (1S %) and with a precision of 6 1 %
38、(1S %).11.2 The energy-dependent uncertainty, expressed as a percentage of the baseline cross section, for the 27Al(n,)24Na crosssection is shown in Fig. 2.(3)11.3 Test results have been reported in neutron benchmark fields.E266 17311.3.1 In the 252Cf spontaneous fission reference neutron field, the
39、 measured cross section is 1.016 bmb 6 1.47 % (710) andthe calculated cross section using the RRDF-2008 cross section is 1.0182 b1.0184 mb with a spectrum integrated cross sectionundertainty of 0.3360.7158 % (3) and a spectrum characterization uncertainty of 1.609 %.% (11). This results in acalculat
40、ed-to-experimental (C/E) ratio of 1.00221.00236 6 2.211.761 %.11.3.2 In the 235U thermal neutron field, the as characterized in the JENDL 4.0 nuclear data file, the measured cross section is0.7007 bmb 6 1.28 % (710) and the calculated cross section using the RRDF-2008 cross section is 0.7173 b0.7000
41、346 mb witha spectrum integrated cross section uncertainty of 0.2870.37504957 % (34) and a spectrum characterization uncertainty of 6.951%.13.59351 % (11). This results in a calculated-to-experimental (C/E) ratio of 1.0240.99905 6 7.0813.65364 %.12. Keywords12.1 activation; activation reaction; alum
42、inum; cross section; dosimetry; fast-neutron monitor; neutron metrology; pressurevessel surveillance; reaction rateREFERENCES(1) Nuclear Wallet Cards, compiled by J. K. Tuli, National Nuclear Data Center, April 8th edition, October 2005 2011. This is the last printed editionof this database and the
43、natural abundance values here for 27Al agree with the on-line version as of January 2, 2017.(2) BeUpdate of X-ray and Gamma Ray Decay Data Standards for Detector Calibration and Other Applications: Vol 1: Reommended Decay Data, HighEnergy Gamma Ray Standards and Anular Correlation Coefficients, Inte
44、rnational Atomic Energy Agency, Vienna, report STI/PUB/1287, M. M.,Chiste, V., Dulieu, C., Browne, E., et al, Table of Radionuclides (Vol 1 A = 1 to 150), Monographie BIPM-5, Bureau International des Poids etMesures (BIPM), France, 20072004., http:/www.nucleide.org/DDEP_WG/DDEPdata.htm (data current
45、 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, March1999. The last full release was in 1998. Updated versions referenced here corresponding to the RRDF-2008 library.(4) Zolotarev, K. I., E
46、valuation of Cross-Section Data from Threshold to 40-60 MeV for Specific Neutron Reactions Important for Nuetron DosimetryApplications, Part 1 Evaluation of the excitation functions for the 27Al(n,a)24Na, 55Mn(N,2n)54Mn,59Co(n,p)59Fe,59Co(n,2n)58m+gCo and90Zr(n,2n)89m+gZr reactions, report INDC(NDS)
47、-0546, International Atomic Energy Agency, Vienna, Austria, April 2009.(5) “EXFOR Formats Description for Users (EXFOR Basics)”, report IAEA-NDS-206, InternationalAtomic EnergyAgency,Vienna,Austria, June 2008.On-line database available at URL: at: http:/www-nds.iaea.org/indg_nexp.html. Data here as
48、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 Experimental NuclearReaction Data Library (EXFOR): International Collaboration Between Nuclear Reaction Data Centres (NRDC), Nuclear Data Sheets,
49、Vol 120, pp.272-276, June 2014, URL for EXFOR database: https:/www-nds.iaea.org/exfor/exfor.htm(7) Capote, R., Zolotarev, K., Pronyaev, V., Trkov,A., “Updating and Extending the IRDF-2002 Dosimetry Library,” Journal of ASTM International, Vol9, 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 Organization for Standardization, 1995, ISBN 92-67-10188-9.(9) Taylor, B. N.,
