1、Designation: E496 141Standard Test Method forMeasuring Neutron Fluence and Average Energyfrom3H(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 adoption or
2、, 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.1NOTEThe figures were updated editorially in February 2014.1. Scope1.1 This test method cov
3、ers a general procedure for themeasurement of the fast-neutron fluence rate produced byneutron generators utilizing the3H(d,n)4He reaction. Neutronsso produced are usually referred to as 14-MeV neutrons, butrange in energy depending on a number of factors. This testmethod does not adequately cover f
4、usion sources where thevelocity of the plasma may be an important consideration.1.2 This test method uses threshold activation reactions todetermine the average energy of the neutrons and the neutronfluence at that energy. At least three activities, chosen from anappropriate set of dosimetry reactio
5、ns, are required to charac-terize the average energy and fluence. The required activitiesare typically measured by gamma ray spectroscopy.1.3 The values stated in SI units are to be regarded asstandard. No other units of measurement are included in thisstandard.1.4 This standard does not purport to
6、address all of thesafety concerns, if any, associated with its use. It is theresponsibility of the user of this standard to establish appro-priate safety and health practices and determine the applica-bility of regulatory limitations prior to use.2. Referenced Documents2.1 ASTM Standards:2E170 Termi
7、nology Relating to Radiation Measurements andDosimetryE181 Test Methods for Detector Calibration and Analysis ofRadionuclidesE261 Practice for Determining Neutron Fluence, FluenceRate, and Spectra by Radioactivation TechniquesE265 Test Method for Measuring Reaction Rates and Fast-Neutron Fluences by
8、 Radioactivation of Sulfur-32E720 Guide for Selection and Use of Neutron Sensors forDetermining Neutron Spectra Employed in Radiation-Hardness Testing of Electronics2.2 International Commission on Radiation Units and Mea-surements (ICRU) Reports:3ICRU Report 13 Neutron Fluence, Neutron Spectra andKe
9、rmaICRU Report 26 Neutron Dosimetry for Biology andMedicine2.3 ISO Standard:4Guide to the Expression of Uncertainty in Measurement2.4 NIST Document:5Technical Note 1297 Guidelines for Evaluating and Express-ing the Uncertainty of NIST Measurement Results3. Terminology3.1 DefinitionsRefer to Terminol
10、ogy E170.4. Summary of Test Method4.1 This test method describes the determination of theaverage neutron energy and fluence by use of three activitiesfrom a select list of dosimetry reactions. Three dosimetryreactions are chosen based on a number of factors including theintensity of the neutron fiel
11、d, the reaction half-lives, the slopeof the dosimetry reaction cross section near 14-MeV, and theminimum time between sensor irradiation and the gammacounting. The activities from these selected reactions aremeasured. Two of the activities are used, in conjunction withthe nuclear data for the dosime
12、try reactions, to determine the1This test method is under the jurisdiction ofASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility of SubcommitteeE10.07 on Radiation Dosimetry for Radiation Effects on Materials and Devices.Current edition approved Jan. 1, 2014. Pu
13、blished February 2014. Originallyapproved in 1973. Last previous edition approved in 2009 as E496 09. DOI:10.1520/E0496-14E01.2For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Book of ASTMStandards volume information,
14、 refer to the standards Document Summary page onthe ASTM website.3Available from the International Commission on Radiation Units, 7910Woodmont Ave., Washington, DC 20014.4Available from American National Standards Institute (ANSI), 25 W. 43rd St.,4th Floor, New York, NY 10036, http:/www.ansi.org.5Av
15、ailable from National Institute of Standards and Technology (NIST), 100Bureau Dr., Stop 1070, Gaithersburg, MD 20899-1070, http:/www.nist.gov.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States1average neutron energy. The third activity i
16、s used, along withthe neutron energy and nuclear data for the selected reaction, todetermine the neutron fluence. The uncertainty of the neutronenergy and the neutron fluence is determined from the activitymeasurement uncertainty and from the nuclear data.5. Significance and Use5.1 Refer to Practice
17、 E261 for a general discussion of themeasurement of fast-neutron fluence rates with thresholddetectors.5.2 Refer to Test Method E265 for a general discussion ofthe measurement of fast-neutron fluence rates by radioactiva-tion of sulfur-32.5.3 Reactions used for the activity measurements can bechosen
18、 to provide a convenient means for determining theabsolute fluence rates of 14-MeV neutrons obtained with3H(d,n)4He neutron generators over a range of irradiation times fromseconds to approximately 100 days. High purity thresholdsensors referenced in this test method are readily available.5.4 The ne
19、utron-energy spectrum must be known in order tomeasure fast-neutron fluence using a single threshold detector.Neutrons produced by bombarding a tritiated target withdeuterons are commonly referred to as 14-MeV neutrons;however, they can have a range of energies depending on: (1)the angle of neutron
20、emission with respect to the deuteronbeam, (2) the kinetic energy of the deuterons, and (3) the targetthickness. In most available neutron generators of theCockroft-Walton type, a thick target is used to obtain high-neutron yields. As deuterons penetrate through the surface andmove into the bulk of
21、the thick target, they lose energy, andinteractions occurring deeper within the target produce neu-trons with correspondingly lower energy.5.5 Wide variations in neutron energy are not generallyencountered in commercially available neutron generators ofthe Cockroft-Walton type. Figs. 1 and 2 (1)6sho
22、w the variation6The boldface numbers in parentheses refer to the list of references at the end ofthis standard.FIG. 1 Variation of 0 Degree3H(d,n)4He Differential Cross Section with Incident Deuteron Energy (1)E496 1412of the zero degree3H(d,n)4He neutron production cross sectionwith energy, and cle
23、arly indicate that maximum neutron yieldis obtained with deuterons having energies near the 107 keVresonance. Since most generators are designed for high yield,the deuteron energy is typically about 200 keV, giving a rangeof neutron energies from approximately 14 to 15 MeV. Thedifferential center-of
24、-mass cross section is typically parameter-ized as a summation of Legendre polynomials. Figs. 3 and 4(1,2) show how the neutron yield varies with the emissionangle in the laboratory system. The insert in Fig. 4 shows howthe magnitude,A1, of the P1() term, and hence the asymmetryin the differential c
25、ross section grows with increasing energy ofthe incident deuteron. The nonrelativistic kinematics (valid forEd3.71 MeV) this reaction is nolonger monoenergetic. Monoenergetic neutron beams withenergies from about 14.8 to 20.4 MeV can be produced by thisreaction at forward laboratory angles (7).5.7 I
26、t is recommended that the dosimetry sensors be fieldedin the exact positions where the dosimetry results are wanted.There are a number of factors that can affect the monochro-maticity or energy spread of the neutron beam (7,8). Thesefactors include the energy regulation of the incident deuteronenerg
27、y, energy loss in retaining windows if a gas target is usedor energy loss within the target if a solid tritiated target is used,the irradiation geometry, and background neutrons from scat-tering with the walls and floors within the irradiation chamber.6. Apparatus6.1 Either a NaI(Tl) or a Ge semicon
28、ductor gamma-rayspectrometer, incorporating a multichannel pulse-height ana-lyzer is required. See Test Methods E181 for a discussion ofspectrometer systems and their use.6.2 If sulfur is used as a sensor, then a beta particle detectoris required. The apparatus required for beta counting of sulfuris
29、 described in Test Methods E181 and E265.6.3 A precision balance for determining foil masses isrequired.7. Materials and Manufacture7.1 High purity threshold foils are available in a largevariety of thicknesses. Foils of suitable diameter can bepunched from stock material. Small diameter wire may al
30、so beused. Prepunched and weighed high purity foils are alsoavailable commercially. Guide E720 provides some details ontypical foil masses and purity. Foils of 12.7 and 25.4 mm (0.50and 1.00 in.) diameter and 0.13 and 0.25 mm (0.005 and 0.010in.) thickness are typical.7.2 See Test Method E265 for de
31、tails on the availability andpreparation of sulfur sensors.FIG. 3 Energy and Angle Dependence of the3H(d,n)4He Differen-tial Cross Section (1)FIG. 4 Change in Neutron Energy from3H(d,n)4He Reaction withLaboratory Emission Angle (2)E496 14148. Calibration8.1 See Test Methods E181 for general detector
32、 calibrationmethods. Test Methods E181 addresses both gamma-ray spec-trometers 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 forthe determination of the neutron energy. The activity ratiomethod h
33、as been described in Ref (9). This test method hasbeen validated for ENDF/B-VI cross sections (10) in Ref (11).9.1.2 Sensor selection depends upon the length of theirradiation, the cross section for the relevant sensor reaction,the reaction half-life, and the expected fluence rate. Table 1lists some
34、 dosimetry-quality reactions that are useful in the14-MeV energy region. The short half-lives of some of thesereaction products, such as27Mg and62Cu, generally limit theuse of these activation products to irradiation times of less thanabout 15 min. Table 2 and Fig. 6 show the recommended crosssectio
35、ns, in the vicinity of 14-MeV, for these reactions. Thecross sections and uncertainties in Table 1 are from theIRDF-2002 (12) cross section compilation. The original sourceof each cross section is listed in the table. The SNLRML crosssection compendium (13) is a single-point-of-reference alter-nativ
36、e source for the cross sections and uncertainty data for thereactions mentioned in Table 1, but somewhat dated, reflectinglarger uncertainties than IRDF 2002. The references for theother nuclear data in Table 1 are given in the table.9.1.3 Longer high fluence irradiations are recommended forthe dete
37、rmination of the neutron energy. Table 3 and Fig. 7 givethe neutron energy-dependent activity ratios for some com-monly used sensor combinations. Fig. 8 displays some slopesfor these ratios. In general, the larger the slope, the moresensitive the method is to the neutron energy. For the proce-dures
38、of this standard to work, it is necessary for the ratios ofthe cross sections to be monotonic in the vicinity of 14 MeV,but the slopes need not be monotonic.9.1.4 Table 4 shows the energy resolutions of some specificsensor combinations for a 14.5 MeV neutron source.The58Ni(n,2n)57Ni-based combinatio
39、ns are recommended dueto their steep slope and accurate dosimetry cross sectionevaluations.9.2 Determine the Sensor MassWeigh each sensor to aprecision of 0.1 %. Nonuniform foil thicknesses can resultfrom the use of dull punches and frequently result in weightvariation of 10 % or more.9.3 Irradiatio
40、n of SensorsIrradiate the sensors, makingcertain that both sensors experience exactly the same fluence.The fluence gradients near a 14-MeV source tend to be highand it may be necessary to stack the sensors together or tomount them on a rotating disk during irradiation. Note thelength of the irradiat
41、ion, ti, and the time the irradiation ended.Some sensors may have an interference reaction that issensitive to low energy neutrons.The interference reaction maybe associated with the primary sensor element or with acontaminant material in the sensor. Of the reactions listed inTable 1, the use of a C
42、u sensor is the only case where theprimary sensor element may be responsible for an interferencereaction. In this case the useful65Cu(n,2n)64Cu reaction activ-ity must be distinguished from the63Cu(n,)64Cu interferencereaction activity (for example, by using an isotopically puresensor or by experime
43、ntally verifying bounds on the maximumpossible level of interference). Other examples of interferencereactions from contaminant materials include trace impuritiesof Mn in Fe sensors and Na in Al sensors. Manganese is afrequent contaminant in Fe foils. In this case the55Mn(n,)56Mn reaction interferes
44、 with the desired sensor responsefrom the56Fe(n,p)56Mn reaction. Salt from handlingAl sensorscan result in the23Na(n,)24Na contaminant reaction whichaffects the use of the27Al(n,)24Na dosimetry sensor. If one isuncertain about the importance of an interference reaction thathas a high thermal neutron
45、 cross section, it is recommendedthat the sensor be irradiated with and without a cadmium coverto quantify the importance of this interference term.9.4 Determination of Sensor ActivityGuide E720 providesdetails on the calculational procedure for determining theactivity of an irradiated sensor. The r
46、esults of this step shouldbe the activities, corrected to a time corresponding to the endof the irradiation. The activity should be corrected for decayduring the irradiation, as explained in Guide E720. This decaycorrection is especially important for short half-life reactions.The activity should ha
47、ve units of Bq per target atom.FIG. 5 Dependence of3H(d,n)4He Neutron Energy on Angle (2)E496 14159.5 CalculationsSection 11 details the calculations thatuse a ratio of two sensor activities to determine the neutronaverage energy.10. Procedure for Determining the Neutron Fluence10.1 Selection of Sen
48、sor:10.1.1 To avoid sensitivity to uncertainty in the exactneutron energy, the 14-MeVneutron fluence sensor is generallychosen to have a flat response in the 13 MeV to 15 MeV energyregion. Fig. 6 and Table 2 show the energy dependence near 14MeV for some frequently used dosimetry sensors. An exami-n
49、ation of Fig. 6 and Table 2 clearly indicates a strongpreference to use the93Nb(n,2n)92mNb reaction. This prefer-ence is based on the flat energy response and the small crosssection uncertainty near 14 MeV. The93Nb(n,2n)92mNb reac-tion has been used as a transfer standard for 14-MeV sourcesby national standards laboratories (18) and in internationalintercomparisons (19). The footnotes in Table 1 list someprecautions about use of some other reactions. If the93Nb(n,2n)92mNb reaction cannot be used in a specific case, theuncertainty of the3H(d,n)4He neutron energy
