ASTM E496-2014 red 4936 Standard Test Method for Measuring Neutron Fluence and Average Energy from&8201 3H&40 d n&41 4He Neutron Generators by Radioactivation Techniques《采用放射性技术测量3.pdf

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1、Designation: E496 09E496 14Standard Test Method forMeasuring Neutron Fluence and Average Energyfrom 3H(d,n)4He Neutron Generators by RadioactivationTechniques1This standard is issued under the fixed designation E496; the number immediately following the designation indicates the year oforiginal adop

2、tion or, in the case of 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 test method covers a general procedure for the measurement of the f

3、ast-neutron fluence rate produced by neutrongenerators utilizing the 3H(d,n)4He reaction. Neutrons so produced are usually referred to as 14-MeV neutrons, but range in energydepending on a number of factors. This test method does not adequately cover fusion sources where the velocity of the plasmama

4、y be an important consideration.1.2 This test method uses threshold activation reactions to determine the average energy of the neutrons and the neutron fluenceat that energy. At least three activities, chosen from an appropriate set of dosimetry reactions, are required to characterize theaverage en

5、ergy and fluence. The required activities are typically measured by gamma ray spectroscopy.1.3 The measurement of reaction products in their metastable states is not covered. If the metastable state decays to the groundstate, the ground state reaction may be used.1.3 The values stated in SI units ar

6、e to be regarded as standard. No other units of measurement are included in this standard.1.4 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 standard to establish appropriate safety and health practic

7、es and determine the applicability of regulatorylimitations prior to use.2. Referenced Documents2.1 ASTM Standards:2E170 Terminology Relating to Radiation Measurements and DosimetryE181 Test Methods for Detector Calibration and Analysis of RadionuclidesE261 Practice for Determining Neutron Fluence,

8、Fluence Rate, and Spectra by Radioactivation TechniquesE265 Test Method for Measuring Reaction Rates and Fast-Neutron Fluences by Radioactivation of Sulfur-32E720 Guide for Selection and Use of Neutron Sensors for Determining Neutron Spectra Employed in Radiation-HardnessTesting of Electronics2.2 In

9、ternational Commission on Radiation Units and Measurements (ICRU) Reports:3ICRU Report 13Neutron13 Neutron Fluence, Neutron Spectra and KermaICRU Report 26Neutron26 Neutron Dosimetry for Biology and Medicine2.3 ISO Standard:4Guide to the Expression of Uncertainty in Measurement2.4 NIST Document:5Tec

10、hnical Note 1297Guidelines1297 Guidelines for Evaluating and Expressing the Uncertainty of NIST Measurement Results1 This test method is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee E10.07on Radiation Dosimetry f

11、or Radiation Effects on Materials and Devices.Current edition approved June 15, 2009Jan. 1, 2014. Published August 2009February 2014. Originally approved in 1973. Last previous edition approved in 20022009 asE496 02.E496 09. DOI: 10.1520/E0496-09.10.1520/E0496-14.2 For referencedASTM standards, visi

12、t 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.3 Available from the International Commission on Radiation Units, 7910 Woodmont Ave., Washington,

13、 DC 20014.4 Available from American National Standards Institute (ANSI), 25 W. 43rd St., 4th Floor, New York, NY 10036, http:/www.ansi.org.5 Available from National Institute of Standards and Technology (NIST), 100 Bureau Dr., Stop 1070, Gaithersburg, MD 20899-1070, http:/www.nist.gov.This document

14、is not an ASTM standard and is intended only to provide the user of an ASTM 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 appro

15、priate. In all cases only the current versionof the standard as published 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 States13. Terminology3.1 DefinitionsRefer to Terminology E170.4. Summ

16、ary of Test Method4.1 This test method describes the determination of the average neutron energy and fluence by use of three activities from aselect list of dosimetry reactions. Three dosimetry reactions are chosen based on a number of factors including the intensity of theneutron field, the reactio

17、n half-lives, the slope of the dosimetry reaction cross section near 14-MeV, and the minimum time betweensensor irradiation and the gamma counting. The activities from these selected reactions are measured. Two of the activities areused, in conjunction with the nuclear data for the dosimetry reactio

18、ns, to determine the average neutron energy. The third activityis used, along with the neutron energy and nuclear data for the selected reaction, to determine the neutron fluence. The uncertaintyof the neutron energy and the neutron fluence is determined from the activity measurement uncertainty and

19、 from the nuclear data.5. Significance and Use5.1 Refer to Practice E261 for a general discussion of the measurement of fast-neutron fluence rates with threshold detectors.5.2 Refer to Test Method E265 for a general discussion of the measurement of fast-neutron fluence rates by radioactivation ofsul

20、fur-32.5.3 Reactions used for the activity measurements can be chosen to provide a convenient means for determining the absolutefluence rates of 14-MeV neutrons obtained with 3H(d,n)4He neutron generators over a range of irradiation times from seconds toapproximately 100 days. High purity threshold

21、sensors referenced in this test method are readily available.5.4 The neutron-energy spectrum must be known in order to measure fast-neutron fluence using a single threshold detector.Neutrons produced by bombarding a tritiated target with deuterons are commonly referred to as 14-MeV neutrons; however

22、, theycan have a range of energies depending on: (1) the angle of neutron emission with respect to the deuteron beam, (2) the kineticenergy of the deuterons, and (3) the target thickness. In most available neutron generators of the Cockroft-Walton type, a thicktarget is used to obtain high-neutron y

23、ields. As deuterons penetrate through the surface and move into the bulk of the thick target,they lose energy, and interactions occurring deeper within the target produce neutrons with correspondingly lower energy.FIG. 1 Variation of 0 Degree 3H(d,n)4He Differential Cross Section with Incident Deute

24、ron Energy (1)E496 1425.5 Wide variations in neutron energy are not generally encountered in commercially available neutron generators of theCockroft-Walton type. Figs. 1 and 2 (1)6 show the variation of the zero degree 3H(d,n)4He neutron production cross section withenergy, and clearly indicate tha

25、t maximum neutron yield is obtained with deuterons having energies near the 107 keV resonance.Since most generators are designed for high yield, the deuteron energy is typically about 200 keV, giving a range of neutronenergies from approximately 14 to 15 MeV. The differential center-of-mass cross se

26、ction is typically parameterized as a summationof Legendre polynomials. Figs. 3 and 4 (1,2) show how the neutron yield varies with the emission angle in the laboratory system.The insert in Fig. 4 shows how the magnitude, A1, of the P1() term, and hence the asymmetry in the differential cross section

27、grows with increasing energy of the incident deuteron. The nonrelativistic kinematics (valid for Ed 3.71 MeV) this reaction is no longer monoenergetic. Monoenergetic neutron beams with energies from about 14.8 to20.4 MeV can be produced by this reaction at forward laboratory angles (7).5.7 It is rec

28、ommended that the dosimetry sensors be fielded in the exact positions that will be used for the customers of the14-MeV neutron source. where the dosimetry results are wanted. There are a number of factors that can affect themonochromaticity or energy spread of the neutron beam (7,8). These factors i

29、nclude the energy regulation of the incident deuteronenergy, energy loss in retaining windows if a gas target is used or energy loss within the target if a solid tritiated target is used,the irradiation geometry, and background neutrons from scattering with the walls and floors within the irradiatio

30、n chamber.6. Apparatus6.1 Either a NaI(Tl) or a Ge semiconductor gamma-ray spectrometer, incorporating a multichannel pulse-height analyzer isrequired. See Test Methods E181 for a discussion of spectrometer systems and their use.6.2 If sulfur is used as a sensor, then a beta particle detector is req

31、uired. The apparatus required for beta counting of sulfur isdescribed in Test Methods E181 and E265.6.3 A precision balance for determining foil masses is required.7. Materials and Manufacture7.1 High purity threshold foils are available in a large variety of thicknesses. Foils of suitable diameter

32、can be punched fromstock material. Small diameter wire may also be used. Prepunched and weighed high purity foils are also available commercially.FIG. 3 Energy and Angle Dependence of the 3H(d,n)4He Differential Cross Section (1)E496 144Guide E720 provides some details on typical foil masses and pur

33、ity. Foils of 12.7 and 25.4 mm (0.50 and 1.00 in.) diameter and0.13 and 0.25 mm (0.005 and 0.010 in.) thickness are typical.7.2 See Test Method E265 for details on the availability and preparation of sulfur sensors.8. Calibration8.1 See Test Methods E181 for general detector calibration methods. Tes

34、t Methods E181 addresses both gamma-rayspectrometers and beta counting methods.9. Procedure for Determining the Neutron Energy9.1 Selection of Sensors:9.1.1 Use of an activity ratio method is recommended for the determination of the neutron energy. The activity ratio methodhas been described in Ref

35、(9). This test method has been validated for ENDF/B-VI cross sections (10) in Ref (11).9.1.2 Sensor selection depends upon the length of the irradiation, the cross section for the relevant sensor reaction, the reactionhalf-life, and the expected fluence rate. Table 1 lists some dosimetry-quality rea

36、ctions that are useful in the 14-MeV energy region.The short half-lives of some of these reaction products, such as 27Mg and 62Cu, generally limit the use of these activation productsto irradiation times of less than about 15 min. Table 2 and Fig. 6 show the recommended cross sections, in the vicini

37、ty of 14-MeV,for these reactions. The cross sections and uncertainties in Table 1 are from the IRDF-2002 (12) cross section compilation. Theoriginal source of each cross section is listed in the table. The SNLRML cross section compendium (13) is a single-point-of-reference alternative source for the

38、 cross sections and uncertainty data for the reactions mentioned in Table 1, but somewhat dated,reflecting larger uncertainties than IRDF 2002. The references for the other nuclear data in Table 1 are given in the table.9.1.3 Longer high fluence irradiations are recommended for the determination of

39、the neutron energy. Table 3 and Fig. 7 givethe neutron energy-dependent activity ratios for some commonly used sensor combinations. Fig. 8 displays some slopes for theseratios. In general, the larger the slope, the more sensitive the method is to the neutron energy. For the procedures of this standa

40、rdto work, it is necessary for the ratios of the cross sections to be monotonic in the vicinity of 14 MeV, but the slopes need not bemonotonic.FIG. 4 Change in Neutron Energy from 3H(d,n)4He Reaction with Laboratory Emission Angle (2)E496 1459.1.4 Table 4 shows the energy resolutions of some specifi

41、c sensor combinations for a 14.5 MeV neutron source.The 58Ni(n,2n)57Ni-based combinations are recommended due to their steep slope and accurate dosimetry cross sectionevaluations.9.2 Determine the Sensor MassWeigh each sensor to a precision of 0.1 %. Nonuniform foil thicknesses can result from theus

42、e of dull punches and frequently result in weight variation of 10 % or more.9.3 Irradiation of SensorsIrradiate the sensors, making certain that both sensors experience exactly the same fluence. Thefluence gradients near a 14-MeV source tend to be high and it may be necessary to stack the sensors to

43、gether or to mount themon a rotating disk during irradiation. Note the length of the irradiation, ti, and the time the irradiation ended. Some sensors mayhave an interference reaction that is sensitive to low energy neutrons. The interference reaction may be associated with the primarysensor element

44、 or with a contaminant material in the sensor. Of the reactions listed in Table 1, the use of a Cu sensor is the onlycase where the primary sensor element may be responsible for an interference reaction. In this case the useful 65Cu(n,2n)64Cureaction activity must be distinguished from the 63Cu(n,)6

45、4Cu interference reaction activity (for example, by using an isotopicallypure sensor or by experimentally verifying bounds on the maximum possible level of interference). Other examples of interferencereactions from contaminant materials include trace impurities of Mn in Fe sensors and Na in Al sens

46、ors. Manganese is a frequentcontaminant in Fe foils. In this case the 55Mn(n,)56Mn reaction interferes with the desired sensor response from the 56Fe(n,p)56Mnreaction. Salt from handling Al sensors can result in the 23Na(n,)24Na contaminant reaction which affects the use ofthe 27Al(n,)24Na dosimetry

47、 sensor. If one is uncertain about the importance of an interference reaction that has a high thermalneutron cross section, it is recommended that the sensor be irradiated with and without a cadmium cover to quantify the importanceof this interference term.9.4 Determination of Sensor ActivityGuide E

48、720 provides details on the calculational procedure for determining the activityof an irradiated sensor. The results of this step should be the activities, corrected to a time corresponding to the end of theirradiation. The activity should be corrected for decay during the irradiation, as explained

49、in Guide E720. This decay correctionis especially important for short half-life reactions. The activity should have units of Bq per target atom.9.5 CalculationsSection 11 details the calculations that use a ratio of two sensor activities to determine the neutron averageenergy.FIG. 5 Dependence of 3H(d,n)4He Neutron Energy on Angle (2)E496 146TABLE 1 Cross Section Parameters for Some Useful ReactionsDosimetryReactionsTarget Nucleus Product NucleusReactionNotesElementalAtomicWeight(14)IsotopicAtomicNumberAbundance,%(14)CrossSectionSourceACrossSectionUncertaint