ASTM E496-2009 6250 Standard Test Method for Measuring Neutron Fluence and Average Energy from 3H(d n)4He Neutron Generators by Radioactivation Techniques 1《用放射性技术测定H3(d n)He4中子发生器.pdf

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1、Designation: E 496 09Standard Test Method forMeasuring Neutron Fluence and Average Energyfrom3H(d,n)4He Neutron Generators by RadioactivationTechniques1This standard is issued under the fixed designation E 496; the number immediately following the designation indicates the year oforiginal adoption o

2、r, 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 themeasurement of the fast-neu

3、tron 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 fusion sources where thevelocity of the plasma may be an im

4、portant 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 reactions, are required to charac-terize the average energy and f

5、luence. The required activitiesare typically measured by gamma ray spectroscopy.1.3 The measurement of reaction products in their meta-stable states is not covered. If the metastable state decays to theground state, the ground state reaction may be used.1.4 The values stated in SI units are to be re

6、garded asstandard. No other units of measurement are included in thisstandard.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 establish appro-priate safety and health practices and deter

7、mine the applica-bility of regulatory limitations prior to use.2. Referenced Documents2.1 ASTM Standards:2E 170 Terminology Relating to Radiation Measurementsand DosimetryE 181 Test Methods for Detector Calibration and Analysisof RadionuclidesE 261 Practice for Determining Neutron Fluence, FluenceRa

8、te, and Spectra by Radioactivation TechniquesE 265 Test Method for Measuring Reaction Rates andFast-Neutron Fluences by Radioactivation of Sulfur-32E 720 Guide for Selection and Use of Neutron Sensors forDetermining Neutron Spectra Employed in Radiation-Hardness Testing of Electronics2.2 Internation

9、al Commission on Radiation Units andMeasurements (ICRU) Reports:3ICRU Report 13Neutron Fluence, Neutron Spectra andKermaICRU Report 26Neutron Dosimetry for Biology andMedicine2.3 ISO Standard:4Guide to the Expression of Uncertainty in Measurement2.4 NIST Document:5Technical Note 1297Guidelines for E

10、valuating and Ex-pressing the Uncertainty of NIST Measurement Results3. Terminology3.1 DefinitionsRefer to Terminology E 170.4. 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 re

11、actions. Three dosimetryreactions are chosen based on a number of factors including theintensity of the neutron field, 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 thes

12、e selected reactions aremeasured. Two of the activities are used, in conjunction withthe nuclear data for the dosimetry reactions, to determine theaverage neutron energy. The third activity is used, along withthe neutron energy and nuclear data for the selected reaction, todetermine the neutron flue

13、nce. The uncertainty of the neutron1This 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 June 15, 200

14、9. Published August 2009. Originallyapproved in 1973. Last previous edition approved in 2002 as E 496 02.2For 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 standar

15、ds 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.5Available from National

16、 Institute of Standards and Technology (NIST), 100Bureau Dr., Stop 1070, Gaithersburg, MD 20899-1070, http:/www.nist.gov.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.energy and the neutron fluence is determined from the activityme

17、asurement uncertainty and from the nuclear data.5. Significance and Use5.1 Refer to Practice E 261 for a general discussion of themeasurement of fast-neutron fluence rates with thresholddetectors.5.2 Refer to Test Method E 265 for a general discussion ofthe measurement of fast-neutron fluence rates

18、by radioactiva-tion of sulfur-32.5.3 Reactions used for the activity measurements can bechosen to provide a convenient means for determining theabsolute fluence rates of 14-MeV neutrons obtainedwith3H(d,n)4He neutron generators over a range of irradiationtimes from seconds to approximately 100 days.

19、 High puritythreshold sensors referenced in this test method are readilyavailable.5.4 The neutron-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-Me

20、V neutrons;however, they can have a range of energies depending on: (1)the angle of neutron 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 obta

21、in high-neutron yields. As deuterons penetrate through the surface andmove into the bulk of 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

22、 commercially available neutron generators ofthe Cockroft-Walton type. Figs. 1 and 2 (1)6show the variationof the zero degree3H(d,n)4He neutron production cross sectionwith energy, and clearly indicate that maximum neutron yieldis obtained with deuterons having energies near the 107 keVresonance. Si

23、nce 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-mass cross section is typically parameter-ized as a summation of Legendre polynomials. Figs. 3 and 4(1,2) show

24、how the neutron yield varies with the emissionangle in the laboratory system. The insert in Fig. 4 shows howthe magnitude,A1, of the P1(u) term, and hence the asymmetryin the differential cross section grows with increasing energy ofthe incident deuteron. The nonrelativistic kinematics (valid forEd3

25、.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 It is recommended that the dosimetry sensors be fieldedin the exact positions that will be used for the customer

26、s of the14-MeV neutron source. There are a number of factors that canaffect the monochromaticity or energy spread of the neutronbeam (7,8). These factors include the energy regulation of theincident deuteron energy, energy loss in retaining windows if agas target is used or energy loss within the ta

27、rget if a solidtritiated target is used, the irradiation geometry, and back-ground neutrons from scattering with the walls and floorswithin the irradiation chamber.6. Apparatus6.1 Either a NaI(Tl) or a Ge semiconductor gamma-rayspectrometer, incorporating a multichannel pulse-height ana-lyzer is req

28、uired. See Test Methods E 181 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 described in Test Methods E 181 and E 265.FIG. 3 Energy and Angle Dependence of the3H(

29、d,n)4HeDifferential Cross Section (1)FIG. 4 Change in Neutron Energy from3H(d,n)4He Reaction withLaboratory Emission Angle (2)E4960936.3 A precision balance for determining foil masses isrequired.7. Materials and Manufacture7.1 High purity threshold foils are available in a largevariety of thickness

30、es. Foils of suitable diameter can bepunched from stock material. Small diameter wire may also beused. Prepunched and weighed high purity foils are alsoavailable commercially. Guide E 720 provides some details ontypical foil masses and purity. Foils of 12.7 and 25.4 mm (0.50and 1.00 in.) diameter an

31、d 0.13 and 0.25 mm (0.005 and 0.010in.) thickness are typical.7.2 SeeTest Method E 265 for details on the availability andpreparation of sulfur sensors.8. Calibration8.1 See Test Methods E 181 for general detector calibrationmethods. Test Methods E 181 addresses both gamma-rayspectrometers and beta

32、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 has been described in Ref (9). This test method hasbeen validated for ENDF/B-VI cross sec

33、tions (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 dosimetry-quality reactions that are useful in the14-MeV energy region. The short half-

34、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 crosssections, in the vicinity of 14-MeV, for these reactions. Thecross sections and uncertainties

35、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-native source for the cross sections and uncertainty data for thereactions mentioned in Table

36、 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 determination of the neutron energy.Table 3 and Fig. 7 givethe neutron energy-dependent acti

37、vity 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 of this standard to work, it is necessary for the ratios ofthe cross sections to be monot

38、onic 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 combinations are recommended dueto their steep slope and accurate dosimetry cross sectionevaluation

39、s.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 Irradiation of SensorsIrradiate the sensors, makingcertain that both sensors experience exactly the

40、 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 irradiation, ti, and the time the irradiation ended.Some sensors may have an interference reactio

41、n 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 Cu sensor is the only case where theprimary sensor element may be responsible for an inter

42、ferencereaction. In this case the useful65Cu(n,2n)64Cu reaction activ-ity must be distinguished from the63Cu(n,g)64Cu interferencereaction activity (for example, by using an isotopically puresensor or by experimentally verifying bounds on the maximumpossible level of interference). Other examples of

43、 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 casethe55Mn(n,g)56Mn reaction interferes with the desired sensorresponse from the56Fe(n,p)56Mn reaction. Salt from handlingAl se

44、nsors can result in the23Na(n,g)24Na contaminant reac-tion which affects the use of the27Al(n,a)24Na dosimetryFIG. 5 Dependence of3H(d,n)4He Neutron Energy on Angle (2)E496094sensor. If one is uncertain about the importance of an interfer-ence reaction that has a high thermal neutron cross section,

45、itis recommended that the sensor be irradiated with and withouta cadmium cover to quantify the importance of this interfer-ence term.9.4 Determination of Sensor ActivityGuide E 720 pro-vides details on the calculational procedure for determining theactivity of an irradiated sensor. The results of th

46、is 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 E 720. This decaycorrection is especially important for short half-life reactions.The activity should have units of

47、 Bq per target atom.9.5 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 Sensor:10.1.1 To avoid sensitivity to uncertainty in the exactneutron energy, the

48、 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-nation of Fig. 6 and Table 2 clearly indicates a strongpreference to use the93N

49、b(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. Ifthe93Nb(n,2n)92mNb reaction cannot be used in a specific case,TABLE 1 Cross Section Parameters for Some Useful ReactionsDosimetryReactionsTarget Nucleus Product NucleusReactionNotesElementalAtomic Wei