1、Designation: E 844 09Standard Guide forSensor Set Design and Irradiation for Reactor Surveillance,E 706(IIC)1This standard is issued under the fixed designation E 844; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision, the year of las
2、t 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 guide covers the selection, design, irradiation,post-irradiation handling, and quality control of neutron do-simeter
3、s (sensors), thermal neutron shields, and capsules forreactor surveillance neutron dosimetry.1.2 The values stated in SI units are to be regarded asstandard. Values in parentheses are for information only.1.3 This standard does not purport to address all of thesafety problems, if any, associated wit
4、h 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:2E 170 Terminology Relating to Radiation Measurementsand DosimetryE
5、261 Practice for Determining Neutron Fluence, FluenceRate, and Spectra by Radioactivation TechniquesE 854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for Reactor Sur-veillance, E706(IIIB)E 910 Test Method for Application and Analysis of HeliumAccumulation Fl
6、uence Monitors for Reactor Vessel Sur-veillance, E706 (IIIC)E 1005 Test Method forApplication andAnalysis of Radio-metric Monitors for Reactor Vessel Surveillance, E706(IIIA)E 1018 Guide for Application of ASTM Evaluated CrossSection Data File, Matrix E 706 (IIB)E 1214 Guide for Use of Melt Wire Tem
7、perature Monitorsfor Reactor Vessel Surveillance, E 706 (IIIE)E 2005 Guide for Benchmark Testing of Reactor Dosimetryin Standard and Reference Neutron FieldsE 2006 Guide for Benchmark Testing of Light Water Reac-tor Calculations3. Terminology3.1 Definitions:3.1.1 neutron dosimeter, sensor, monitora
8、substance irra-diated in a neutron environment for the determination ofneutron fluence rate, fluence, or spectrum, for example: radio-metric monitor (RM), solid state track recorder (SSTR), heliumaccumulation fluence monitor (HAFM), damage monitor(DM), temperature monitor (TM).3.1.2 thermal neutron
9、shielda substance (that is, cad-mium, boron, gadolinium) that filters or absorbs thermalneutrons.3.2 For definitions or other terms used in this guide, refer toTerminology E 170.4. Significance and Use4.1 In neutron dosimetry, a fission or non-fission dosimeter,or combination of dosimeters, can be u
10、sed for determining afluence-rate, fluence, or neutron spectrum, or both, in nuclearreactors. Each dosimeter is sensitive to a specific energy range,and, if desired, increased accuracy in a fluence-rate spectrumcan be achieved by the use of several dosimeters each coveringspecific neutron energy ran
11、ges.4.2 A wide variety of detector materials is used for variouspurposes. Many of these substances overlap in the energy ofthe neutrons which they will detect, but many differentmaterials are used for a variety of reasons. These reasonsinclude available analysis equipment, different cross sectionsfo
12、r different fluence-rate levels and spectra, preferred chemicalor physical properties, and, in the case of radiometric dosim-eters, varying requirements for different half-life isotopes,possible interfering activities, and chemical separation require-ments.5. Selection of Neutron Dosimeters and Ther
13、mal NeutronShields5.1 Neutron Dosimeters:5.1.1 The choice of dosimeter material depends largely onthe dosimetry technique employed, for example, radiometricmonitors, helium accumulation monitors, track recorders, and1This guide is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and
14、 Applications and is the direct responsibility of SubcommitteeE10.05 on Nuclear Radiation Metrology.Current edition approved June 1, 2009. Published June 2009. Originallyapproved in 1981. Last previous edition approved in 2003 as E 844 03.2For referenced ASTM standards, visit the ASTM website, www.a
15、stm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Book of ASTMStandards volume information, refer to the standards Document Summary page onthe ASTM website.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.damage
16、monitors.At the present time, there is a wide variety ofdetector materials used to perform neutron dosimetry measure-ments. These are generally in the form of foils, wires, powders,and salts. The use of alloys is valuable for certain applicationssuch as (1) dilution of high cross-section elements, (
17、2) prepa-ration of elements that are not readily available as foils or wiresin the pure state, and (3) preparation to permit analysis of morethan one dosimeter material.5.1.2 For neutron dosimeters, the reaction rates are usuallydeduced from the absolute gamma-ray radioanalysis (thereexist exception
18、s, such as SSTRs, HAFMs, damage monitors).Therefore, the radiometric dosimeters selected must havegamma-ray yields known with good accuracy (98 %). Thehalf-life of the product nuclide must be long enough to allowfor time differences between the end of the irradiation and thesubsequent counting. Refe
19、r to Method E 1005 for nucleardecay and half-life parameters.5.1.3 The neutron dosimeters should be sized to permitaccurate analysis. The range of high efficiency countingequipment over which accurate measurements can be per-formed is restricted to several decades of activity levels (5 to 7decades f
20、or radiometric and SSTR dosimeters, 8 decades forHAFMs). Since flux levels at dosimeter locations can rangeover 2 or 3 decades in a given experiment and over 10 decadesbetween low power and high power experiments, the propersizing of dosimeter materials is essential to assure accurate andeconomical
21、analysis.5.1.4 The estimate of radiometric dosimeter activity levelsat the time of counting include adjustments for the decay of theproduct nuclide after irradiation as well as the rate of productnuclide buildup during irradiation. The applicable equation forsuch calculations is (in the absence of f
22、luence-rate perturba-tions) as follows:A 5 Nosfa1 2 e2lt1!e2lt2! (1)where:A = expected disintegration rate (dps) for the prod-uct nuclide at the time of counting,No= number of target element atoms,f = estimated flux density level,s = spectral averaged cross section,a = product of the nuclide fractio
23、n and (if appli-cable) of the fission yield,1e-lt1= buildup of the nuclide during the irradiationperiod, t1,e-lt2= decay after irradiation to the time of counting,t2, andl = decay constant for the product nuclide.5.1.5 For SSTRs and HAFMs, the same type of informationas for radiometric monitors (tha
24、t is, total number of reactions)is provided. The difference being that the end products (fissiontracks or helium) requires no time-dependent corrections andare therefore particularly valuable for long-term irradiations.5.1.6 Fission detectors shall be chosen that have accuratelyknown fission yields.
25、 Refer to Method E 1005.5.1.7 In thermal reactors the correction for neutron selfshielding can be appreciable for dosimeters that have highlyabsorbing resonances (see 6.1.1).5.1.8 Dosimeters that produce activation or fission products(that are utilized for reaction rate determinations) with half-liv
26、es that are short compared to the irradiation duration shouldnot be used. Generally, radionuclides with half-lives less thanthree times the irradiation duration should be avoided unlessthere is little or no change in neutron spectral shape or fluencerate with time.5.1.9 Tables 1-3 present various do
27、simeter elements. Listedare the element of interest, the nuclear reaction, and theavailable forms. For the intermediate energy region, the ener-gies of the principal resonances are listed in order of increasingenergy. In the case of the fast neutron energy region, the 95 %response ranges (an energy
28、range that includes most of theresponse for each dosimeter is specified by giving the energiesE05below which 5 % of the activity is produced and E95abovewhich 5 % of the activity is produced) for the235U neutronthermal fission spectrum are included.5.2 Thermal Neutron Shields:5.2.1 Shield materials
29、are frequently used to eliminateinterference from thermal neutron reactions when resonanceand fast neutron reactions are being studied. Cadmium iscommonly used as a thermal neutron shield, generally 0.51 to1.27 mm (0.020 to 0.050 in.) thick. However, because elemen-tal cadmium (m.p. = 320C) will mel
30、t if placed within thevessel of an operating water reactor, effective thermal neutronfilters must be chosen that will withstand high temperatures oflight-water reactors. High-temperature filters include cadmiumoxide (or other cadmium compounds or mixtures), boron(enriched in the10B isotope), and gad
31、olinium. The thickness ofthe shield material must be selected to account for burnoutfrom high fluences.5.2.2 In reactors, feasible dosimeters to date whose responserange to neutron energies of 1 to 3 MeV includes the fissionmonitors238U,237Np, and232Th. These particular dosimetersmust be shielded fr
32、om thermal neutrons to reduce fissionproduct production from trace quantities of235U,238Pu,and239Pu and to suppress buildup of interfering fissionablenuclides, for example,238Np and238Pu in the237Npdosimeter,239Pu in the238U dosimeter, and233Uinthe232Thdosimeter. Thermal neutron shields are also nec
33、essary forepithermal spectrum measurements in the 5 3 107to 0.3-MeVTABLE 1 Dosimeter ElementsThermal Neutron RegionElement ofInterestNuclear Reaction Available FormsB10B(n,a)7Li B, B4C, B-Al, B-NbCo59Co(n,g)60Co Co, Co-Al, Co-ZrCu63Cu(n,g)64Cu Cu, Cu-Al, Cu(NO3)2Au197Au(n,g)198Au Au, Au-AlIn115In(n,
34、g)116mIn In, In-AlFe58Fe(n,g)59Fe FeFe54Fe(n,g)55Fe FeLi6Li(n,a)3H LiF, Li-AlMn55Mn(n,g)56Mn alloysNi58Ni(n,g)59Ni(n,a)56Fe NiPu239Pu(n,f)FP PuO2, alloysSc45Sc(n,g)46Sc Sc, Sc2O3Ag109Ag(n,g)110mAg Ag, Ag-Al, AgNO3Na23Na(n,g)24Na NaCl, NaF, NaITa181Ta(n,g)182Ta Ta, Ta2O5U (enriched)235U(n,f)FP U, U-A
35、l, UO2,U3O8, alloysE844092energy range. Also, nickel dosimeters used for the fast activa-tion reaction58Ni(n,p)58Co must be shielded from thermalneutrons in nuclear environments having thermal fluence ratesabove 3 3 1012ncm2s1to prevent significant loss of58Coand58mCo by thermal neutron burnout (4).
36、36. Design of Neutron Dosimeters, Thermal NeutronShields, and Capsules6.1 Neutron DosimetersProcedures for handling dosim-eter materials during preparation must be developed to ensurepersonnel safety and accurate nuclear environment character-ization. During dosimeter fabrication, care must be taken
37、 inorder to achieve desired neutron flux results, especially in thecase of thermal and resonance-region dosimeters. A number of3The boldface number in parentheses refers to the list of references at the end ofthe guide.TABLE 2 Dosimeter ElementsIntermediate Neutron RegionEnergy of PrincipalResonance
38、, eV(17)Dosimetry Reactions Element of Interest Available FormsA 6Li(n,a)3H Li LiF, Li-AlA 10B(n,a)7Li B B, B4C, B-Al, B-NbA 58Ni(n,g)59Ni(n,a)56Fe Ni Ni1.457115In(n,g)116mIn In In, In-Al4.28181Ta(n,g)182Ta Ta Ta, Ta2O54.906197Au(n,g)198Au Au Au, Au-Al5.19109Ag(n,g)110mAg Ag Ag, Ag-Al, AgNO321.80623
39、2Th(n,g)233Th Th Th, ThO2, Th(NO3)4B 235U(n,f)FP U U, U-Al, UO2,U3O8, alloys13259Co(n,g)60Co Co Co, Co-Al, Co-Zr103858Fe(n,g)59Fe Fe Fe337.355Mn(n,g)56Mn Mn alloys57963Cu(n,g)64Cu Cu Cu, Cu-Al, Cu(NO3)20.2956243239Pu(n,f)FP Pu PuO2, alloys281023Na(n,g)24Na Na NaCl, NaF, NaI329545Sc(n,g)46Sc Sc Sc, S
40、c2O3778854Fe(n,g)55Fe Fe FeAThis reaction has no resonance that contributes in the intermediate energy region and the principle resonance has negative energy (i.e. the cross section is 1/v).BMany resonances contribute in the 1 100 eV region for this reaction.TABLE 3 Dosimeter ElementsFast Neutron Re
41、gionDosimetryReactionsElement ofInterestEnergy Response Range (MeV)A,BCross SectionUncertainty(%)A,CAvailableFormsLowE05MedianE50HighE95237Np(n,f)FP Np 0.684 1.96 5.61 9.34 Np2O3, alloys103Rh(n,n8)103mRh Rh 0.731 2.25 5.73 3.10 Rh93Nb(n,n8)93mNb Nb 0.951 2.57 5.79 3.01 Nb, Nb2O5115In(n,n8)115mIn In
42、1.12 2.55 5.86 2.16 In, In-Al14N(n,a)11B N 1.75 3.39 5.86 TiN, ZrN, NbN238U(n,f)FP U (depleted) 1.44 2.61 6.69 0.319 U, U-Al, UO3,U3O8, alloys232Th(n,f)FP Th 1.45 2.79 7.21 5.11 Th, ThO29Be(n,a)6Li Be 1.59 2.83 5.26 Be47Ti(n,p)47Sc Ti 1.70 3.63 7.67 3.77 Ti58Ni(n,p)58Co Ni 1.98 3.94 7.51 2.44 Ni, Ni
43、-Al54Fe(n,p)54Mn Fe 2.27 4.09 7.54 2.12 Fe32S(n,p)32P S 2.28 3.94 7.33 3.63 CaSO4,Li2SO432S(n,a)29Si S 1.65 3.12 6.06 Cu2S, PbS58Ni(n,a)55Fe Ni 2.74 5.16 8.72 Ni, Ni-Al46Ti(n,p)46Sc Ti 3.70 5.72 9.43 2.48 Ti56Fe(n,p)56Mn FeD5.45 7.27 11.3 2.26 Fe56Fe(n,a)53Cr Fe 5.19 7.53 11.3 Fe63Cu(n,a)60Co CuE4.5
44、3 6.99 11.0 2.36 Cu, Cu-Al27Al(n,a)24Na Al 6.45 8.40 11.9 1.19 Al, Al2O348Ti(n,p)48Sc Ti 5.92 8.06 12.3 2.56 Ti47Ti(n,a)44Ca Ti 2.80 5.10 9.12 Ti60Ni(n,p)60Co NiE4.72 6.82 10.8 10.3 Ni, Ni-Al55Mn(n,2n)54Mn MnF11.0 12.6 15.8 13.54 alloysAEnergy response range was derived using the ENDF/B-VI235U fissi
45、on spectrum, Ref (1), MT = 9228, MF = 5, MT = 18. The cross section and associated covariancesources are identified in Guide E 1018 and in Refs (2,3).BOne half of the detector response occurs below an energy given by E50; 95 % of the detector response occurs below E95and 5 % below E05.CUncertainty m
46、etric only reflects that component due to the knowledge of the cross section and is reported at the 1s level.DLow manganese content necessary.ELow cobalt content necessary.FLow iron content necessary.E844093factors must be considered in the design of a dosimetry set foreach particular application. S
47、ome of the principal ones arediscussed individually as follows:6.1.1 Self-Shielding of NeutronsThe neutron self-shielding phenomenon occurs when high cross-section atomsin the outer layers of a dosimeter reduce the neutron flux to thepoint where it significantly affects the activation of the innerat
48、oms of the material. This is especially true of materials withhigh thermal cross sections and essentially all resonancedetectors. This can be minimized by using low weight percent-age alloys of high-cross-section material, for example, Co-Al,Ag-Al, B-Al, Li-Al. It is not as significant for the fast
49、regionwhere the cross sections are relatively low; therefore, thermaland resonance detectors shall be as thin as possible. Math-ematical corrections can also be made to bring the material to“zero thickness” but, in general, the smaller the correction, themore accurate will be the results. Both theoretical treatments ofthe complex corrections and experimental determinations arepublished (5-16,17).6.1.2 Self-Absorption of Emitted RadiationThis effectmay be observed during counting of the radiometric dosimeter.If the radiation of interest is a low-energ
