ASTM E844-2003 Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance E 706(IIC)《反应堆监视用传感器装置设计和辐照的标准指南E-706(ⅡC)》.pdf

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1、Designation: E 844 03Standard 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 (e) 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-simete

3、rs (sensors), thermal neutron shields, and capsules forreactor surveillance neutron dosimetry.1.2 The values stated in inch-pound units are to be regardedas the standard. The values given in parentheses are forinformation only.1.3 This standard does not purport to address all of thesafety problems,

4、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:E 170 Terminology Relating to Radiation Measur

5、ementsand Dosimetry2E 261 Practice for Determining Neutron Fluence Rate, Flu-ence, and Spectra by Radioactivation Techniques2E 854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for Reactor Sur-veillance, E 706(IIIB)2E 910 Test Method for Application and Analys

6、is of HeliumAccumulation Fluence Monitors for Reactor Vessel Sur-veillance, E 706(IIIC)2E 1005 Test Method for Application and Analysis of Radio-metric Monitors for Reactor Vessel Surveillance,E 706(IIIA)2E 706(IIID) Analysis of Damage Monitors for Reactor Ves-sel Surveillance3E 706(IIIE) Analysis o

7、f Temperature Monitors for ReactorVessel Surveillance3E 706(IIE) Benchmark Testing of Reactor Vessel Dosim-etry33. Terminology3.1 Definitions:3.1.1 neutron dosimeter, sensor, monitora substance irra-diated in a neutron environment for the determination ofneutron fluence rate, fluence, or spectrum, f

8、or 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 shielda substance (that is, cad-mium, boron, gadolinium) that filters or absorbs thermalneutrons.3.2 For definitions

9、 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 used for determining afluence-rate, fluence, or neutron spectrum, or both, in nuclearreactors. Each dosimeter is sens

10、itive to a specific energy range,and, if desired, increased accuracy in a flux-spectrum can beachieved by the use of several dosimeters each coveringspecific neutron energy ranges.4.2 A wide variety of detector materials is used for variouspurposes. Many of these substances overlap in the energy oft

11、he neutrons which they will detect, but many differentmaterials are used for a variety of reasons. These reasonsinclude available analysis equipment, different cross sectionsfor different flux levels and spectra, preferred chemical orphysical properties, and, in the case of radiometric dosimeters,va

12、rying requirements for different half-life isotopes, possibleinterfering activities, and chemical separation requirements.5. Selection of Neutron Dosimeters and Thermal NeutronShields5.1 Neutron Dosimeters:5.1.1 The choice of dosimeter material depends largely onthe dosimetry technique employed, for

13、 example, radiometricmonitors, helium accumulation monitors, track recorders, anddamage monitors. At the present time, there is a wide variety of1This guide is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility of SubcommitteeE10.05 on

14、 Nuclear Radiation Metrology.Current edition approved Feb. 10, 2003. Published February 2003. Originallyapproved in 1981. Last previous edition approved in 1997 as E 844 97.2Annual Book of ASTM Standards, Vol 12.02.3For standards that are in the draft stage and have not received an ASTMdesignation,

15、see Section 5 as well as Figures 1 and 2 of Matrix E 706.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.detector materials used to perform neutron dosimetry measure-ments. These are generally in the form of foils, wires, powders,and

16、 salts. The use of alloys is valuable for certain applicationssuch as (1) dilution of high cross-section elements, (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 neutr

17、on dosimeters, the reaction rates are usuallydeduced from the absolute gamma-ray radioanalysis (thereexist exceptions, 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

18、must be long enough to allowfor time differences between the end of the irradiation and thesubsequent counting. Refer 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 ove

19、r which accurate measurements can be per-formed is restricted to several decades of activity levels (5 to 7decades for 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 powe

20、r and high power experiments, the propersizing of dosimeter materials is essential to assure accurate andeconomical 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 rat

21、e of productnuclide buildup during irradiation. The applicable equation forsuch calculations is (in the absence of flux perturbations) asfollows:A 5 Nos fa1 2 elt1!e2lt2! (1)where:A = expected disintegration rate (dps) for the prod-uct nuclide at the time of counting,No= number of target element ato

22、ms,f = estimated flux density level,s = spectral averaged cross section,a = product of the nuclide fraction 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 t

23、he product nuclide.5.1.5 For SSTRs and HAFMs, the same type of informationas for radiometric monitors (that 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 fo

24、r long-term irradiations.5.1.6 Fission detectors shall be chosen that have accuratelyknown fission yields. 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 tha

25、t produce activation or fission products(that are utilized for reaction rate determinations) with half-lives 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

26、little or no change in neutron spectral shape or fluencerate with time.5.1.9 Tables 1-3 present various dosimeter 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 ord

27、er of increasingenergy. In the case of the fast neutron energy region, the 95 %response ranges (an energy 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) fo

28、r the235U neutronthermal fission spectrum are included.5.2 Thermal Neutron Shields:5.2.1 Shield materials 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, gener

29、ally 0.020 to0.050 in. (0.51 to 1.27 mm) thick. However, because elementalcadmium (m.p. = 320C) will melt if placed within the vesselof an operating water reactor, effective thermal neutron filtersmust be chosen that will withstand high temperatures oflight-water reactors. High-temperature filters i

30、nclude cadmiumoxide (or other cadmium compounds or mixtures), boron(enriched in the10B isotope), and gadolinium. 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 t

31、o 3 MeV includes the fissionmonitors238U,237Np, and232Th. These particular dosimetersmust be shielded from thermal neutrons to reduce fissionproduct production from trace quantities of235U,238Pu,and239Pu and to suppress buildup of interfering fissionablenuclides, for example,238Np and238Pu in the237

32、Npdosimeter,239Pu in the238U dosimeter, and233Uinthe232Thdosimeter. Thermal neutron shields are also necessary forepithermal spectrum measurements in the 5 3 107to 0.3-MeVenergy range. Also, nickel dosimeters used for the fast activa-tion reaction58Ni(n,p)58Co must be shielded from thermalTABLE 1 Do

33、simeter 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,g)116mIn In, In-AlFe58Fe(n,g)59Fe FeFe54Fe(n,g)55Fe FeLi6Li(n,a)3H LiF, Li-Al

34、Mn55Mn(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-Al, UO2,U3O8, alloysE844032neutrons in nuclear environments having thermal flu

35、ence ratesabove 3 3 1012ncm2s1to prevent significant loss of58Coand58mCo by thermal neutron burnout (1).46. Design of Neutron Dosimeters, Thermal NeutronShields, and Capsules6.1 Neutron DosimetersProcedures for handling dosim-eter materials during preparation must be developed to ensurepersonnel saf

36、ety and accurate nuclear environment character-ization. During dosimeter fabrication, care must be taken inorder to achieve desired neutron flux results, especially in the4The boldface number in parentheses refers to the list of references at the end ofthe guide.TABLE 2 Dosimeter ElementsIntermediat

37、e Neutron RegionEnergy of PrincipalResonance, 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-Al5.19

38、109Ag(n,g)110mAg Ag Ag, Ag-Al, AgNO321.806232Th(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

39、 Na NaCl, NaF, NaI329545Sc(n,g)46Sc Sc Sc, Sc2O3778854Fe(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 reacti

40、on.TABLE 3 Dosimeter ElementsFast Neutron RegionDosimetryReactionsElement 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

41、.57 5.79 3.01 Nb, Nb2O5115In(n,n8)115mIn In 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 T

42、i58Ni(n,p)58Co Ni 1.98 3.94 7.51 2.44 Ni, Ni-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)5

43、3Cr Fe 5.19 7.53 11.3 Fe63Cu(n,a)60Co CuE4.53 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)60CoFNiE4.72 6.82 10.8 10.3 Ni, Ni-Al55Mn(n,2n)54Mn MnG11.0 12.6 15.8 13.54 alloysAEnergy response ran

44、ge was derived using the ENDF/B-VI235U fission spectrum, reference (17), MT = 9228, MF = 5, MT = 18. The cross section and associatedcovariance sources are identified in standard E 1018 and in Reference (18).BOne half of the detector response occurs below an energy given by E50; 95 % of the detector

45、 response occurs below E95and 5 % below E05.CUncertainty metric 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.FThis reaction does not appear in the ENDF/B-VI evaluation. The JENDL

46、evaluation (19) was used.GLow iron content necessary.E844033case of thermal and resonance-region dosimeters. A number offactors must be considered in the design of a dosimetry set foreach particular application. Some of the principal ones arediscussed individually as follows:6.1.1 Self-Shielding of

47、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 inneratoms of the material. This is especially true of materials withhigh thermal cross sectio

48、ns 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 regionwhere the cross sections are relatively low; therefore, thermaland resonance dete

49、ctors 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 (2-13,20).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-energy gamma ray, an X ray,or a beta particle, the thickness of the dosimeter may be ofapprecia