ASTM E261-2003 Standard Practice for Determining Neutron Fluence Fluence Rate and Spectra by Radioactivation Techniques《用放射性技术测定中子的积分通量、积分通量率和中子波谱的标准实施规范》.pdf

《ASTM E261-2003 Standard Practice for Determining Neutron Fluence Fluence Rate and Spectra by Radioactivation Techniques《用放射性技术测定中子的积分通量、积分通量率和中子波谱的标准实施规范》.pdf》由会员分享，可在线阅读，更多相关《ASTM E261-2003 Standard Practice for Determining Neutron Fluence Fluence Rate and Spectra by Radioactivation Techniques《用放射性技术测定中子的积分通量、积分通量率和中子波谱的标准实施规范》.pdf（10页珍藏版）》请在麦多课文档分享上搜索。

1、Designation: E 261 03Standard Practice forDetermining Neutron Fluence, Fluence Rate, and Spectra byRadioactivation Techniques1This standard is issued under the fixed designation E 261; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision

2、, the year of last 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 practice describes the general procedures for thedetermination of neutron fluence rate, fluence, a

3、nd energyspectra from the radioactivity that is induced in a detectorspecimen.1.2 The practice is directed toward the determination ofthese quantities in connection with radiation effects on mate-rials.1.3 For application of these techniques to reactor vesselsurveillance, see also Test Methods E 100

4、5.1.4 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 determine the applica-bility of regulatory limitations prior to use.NOTE 1Detaile

5、d methods for individual detectors are given in thefollowing ASTM test methods: E 262, E 263, E 264, E 265, E 266, E 343,E 393, E 481, E 523, E 526, E 704, E 705, and E 854.2. Referenced Documents2.1 ASTM Standards:E 170 Terminology Relating to Radiation Measurementsand Dosimetry2E 181 Test Methods

6、for Detector Calibration and Analysisof Radionuclides2E 262 Test Method for Determining Thermal Neutron Re-action and Fluence Rates by Radioactivation Techniques2E 263 Test Method for Measuring Fast-Neutron ReactionRates by Radioactivation of Iron2E 264 Test Method for Measuring Fast-Neutron Reactio

7、nRates by Radioactivation of Nickel2E 265 Test Method for Measuring Reaction Rates andFast-Neutron Fluences by Radioactivation of Sulfur-322E 266 Test Method for Measuring Fast-Neutron ReactionRates by Radioactivation of Aluminum2E 343 Test Method for Measuring Reaction Rates by Analy-sis of Molybde

8、num-99 Radioactivity from Fission Dosim-eters2E 393 Test Method for Measuring Reaction Rates by Analy-sis of Barium-140 from Fission Dosimeters2E 481 Test Method for Measuring Neutron Fluence Rate byRadioactivation of Cobalt and Silver2E 523 Test Method for Measuring Fast-Neutron ReactionRates by Ra

9、dioactivation of Copper2E 526 Test Method for Measuring Fast-Neutron ReactionRates by Radioactivation of Titanium2E 704 Test Method for Measuring Reaction Rates by Ra-dioactivation of Uranium-2382E 705 Test Method for Measuring Reaction Rates by Ra-dioactivation of Neptunium-2372E 844 Guide for Sens

10、or Set Design and Irradiation forReactor Surveillance, E 706(IIC)2E 854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for Reactor Sur-veillance, E 706(IIIB)2E 944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, (IIA)2E 10

11、05 Test Method for Application and Analysis of Radio-metric Monitors for Reactor Vessel Surveillance, E 706(IIIA)E 1018 Guide for Application of ASTM Evaluated CrossSection Data File, Matrix E 706 (IIB)22.2 ISO Standard:Guide in the Expression of Uncertainty in Measurement3. Terminology3.1 Descripti

12、ons of terms relating to dosimetry are found inTerminology E 170.1This practice is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility of SubcommitteeE10.05 on Nuclear Radiation Metrology.Current edition approved July 10, 2003. Publishe

13、d August 2003. Originallyapproved in 1965 as E 261 65 T. Last previous edition approved in 1998 asE 261 98.2Annual Book of ASTM Standards, Vol 12.02.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.4. Summary of Practice4.1 A sample c

14、ontaining a known amount of the nuclide tobe activated is placed in the neutron field. The sample isremoved after a measured period of time and the inducedactivity is determined.5. Significance and Use5.1 Transmutation ProcessesThe effect on materials ofbombardment by neutrons depends on the energy

15、of theneutrons; therefore, it is important that the energy distributionof the neutron fluence, as well as the total fluence, bedetermined.6. Counting Apparatus6.1 A number of instruments are used to determine thedisintegration rate of the radioactive product of the neutron-induced reaction. These in

16、clude the scintillation counters,ionization chambers, proportional counters, Geiger tubes, andsolid state detectors. Recommendations of counters for particu-lar applications are given in General Methods E 181.7. Requirements for Activation-Detector Materials7.1 The general considerations concerning

17、the suitability ofa material for use as an activation detector are found in GuideE 844.7.2 The amounts of fissionable material needed for fissionthreshold detectors are rather small and the availability of thematerial is limited. Licenses from the U.S. Nuclear RegulatoryCommission are required for p

18、ossession.7.3 A detailed description of procedures for the use offission threshold detectors is given in Test Methods E 343,E 393, and E 854, and Guide E 844.8. Irradiation Procedures8.1 The irradiations are carried out in two general waysdepending upon whether the instantaneous fluence rate or thef

19、luence is being determined. For fluence rate, irradiate thedetector for a short period at sufficiently low power thathandling difficulties and shielding requirements are minimized.Then extrapolate the resulting fluence rate value to the valueanticipated for full reactor power. This technique is some

20、timesused for the fluence mapping of reactors (1, 2).8.2 The determination of fluence is most often required inexperiments on radiation effects on materials. Irradiate thedetectors for the same duration as the experiment at a positionin the reactor where, as closely as possible, they will experi-enc

21、e the same fluence, or will bracket the fluence of theposition of interest. When feasible, place the detectors in theexperiment capsule. In this case, long-term irradiations areoften required.8.3 It is desirable, but not required, that the neutron detectorbe irradiated during the entire time period

22、considered and thata measurable part of the activity generated during the initialperiod of irradiation be present in the detector at the end of theirradiation. Therefore, the effective half-life, t81/2- 0.693/l8,ofthe reaction product should not be much less than the totalelapsed time from the initi

23、al exposure to the final shutdown.8.4 As mentioned in 9.11 and 9.12, the use of cadmium-shielded detectors is convenient in separating contributions tothe measured activity from thermal and epithermal neutrons.Also, cadmium-shielding is helpful in reducing activities dueto impurities and the loss of

24、 the activated nuclide by thermal-neutron absorption. The recommended thicknesses of cadmiumis 1 mm. When bare and cadmium-shielded samples are placedin the same vicinity, take care to avoid partial shielding of thebare detectors by the cadmium-shielded ones.9. Calculation9.1 The activity of the sam

25、ple, A, at the end of the exposureperiod is calculated as follows:A 5lD/1 2 exp 2l tc! exp 2l tw!# (1)where:l = decay constant for the radioactive nuclide,tc= time interval for counting,tw= time elapsed between the end of the irradiation periodand the start of the counting period, andD = number of d

26、isintegrations (net number of countscorrected for background, random and true coinci-dence losses, efficiency of the counting system, andfraction of the sample counted) in the interval tc.9.1.1 If, as is often the case, the counting period is shortcompared to the half-life ( = 0.693/l) of the radioa

27、ctive nu-clide, the activity is well approximated as follows:A 5 D/tcexp 2l tw!# (2)9.2 For irradiations at constant fluence rate, the saturationactivity, As, is calculated as follows:As5 A/1 2 exp 2l8ti! (3)where:ti= exposure duration, andl8 = effective decay constant during the irradiation.NOTE 2T

28、he saturation activity corresponds to the number of disinte-grations per foil per unit time for the steady-state condition in which therate of production of the radioactive nuclide is equal to the rate of loss byradioactive decay and transmutation.9.2.1 The effective decay constant, which may be a f

29、unctionof time, is related to the decay constant as follows:l8 5l1*0saE!fE! dE (4)where:sa(E) = neutron absorption cross section for the productnuclide, andf(E) = neutron fluence rate per unit energy.9.2.2 Application of the effective decay constant for irradia-tions under varying fluence rates is d

30、iscussed in this sectionand in the detailed methods for individual detectors.9.3 The reaction rate is calculated as follows:Rs5 Asl8/Nl (5)where:N = number of target nuclei in the detector at time ofirradiation.9.3.1 The number of target nuclei can often be assumed tobe equal to No, the number prior

31、 to irradiation.No5 NAFm/M (6)E261032where:NA= Avogadros number= 6.022 3 1023mole1,F = atom fraction of the target nuclide in the targetelement,m = mass of target element, g, andM = atomic mass of the target element.9.3.2 Calculations of the isotopic concentration after irra-diation is discussed in

32、9.6.6 and in the detailed methods forindividual detectors.9.4 The neutron fluence rate, f, is calculated as follows:f5Rs/s (7)where:s = the spectral weighted neutron activation cross section.9.4.1 Cross sections should be processed from an appropri-ate cross-section library that includes covariance

33、data. GuideE 1018 provides information and recommendations on how toselect the cross section library. The International ReactorDosimetry File (IRDF-90) (38) is one good source for crosssections. The SNLRML cross section compendium (25) pro-vides a processed fine-group representation of recommendeddo

34、simetry cross sections and covariance matrices.9.4.2 If spectral-averaged cross-section or spectrum data arenot available, one of the alternative procedures discussed in9.10 to 9.13 may be used to calculate an approximate neutronfluence rate from the saturation activity.9.5 The neutron fluence, F, i

35、s related to the time varyingdifferential neutron fluence rate f(E,t) by the following expres-sion:F5*0*t1t2f E,t! dt dE (8)where:t2t1= duration of the irradiation period.9.5.1 Long irradiations usually involve operation at variouspower levels, and changes in isotopic content of the system;under suc

36、h conditions f(E, t) can show large variations withtime.9.5.2 It is usual to assume, however, that the neutron fluencerate is directly proportional to reactor power; under theseconditions, the fluence can be well approximated by:F5SfPD (i 5 1nPiti(9)where:f/P = average value of the neutron fluence r

37、ate, f,atareference power level, P,ti= duration of the ithoperating period during which thereactor operated at approximately constant power,andPi= reactor power level during that operating period.9.5.2.1 Alternate methods include measuring the powergeneration rate in a fraction of the reactor volume

38、 adjacent tothe volume of interest.9.6 Transmutation Processes:9.6.1 The neutron fluence rate spectrum, f(E), can bedetermined by computer calculations using neutron transportcodes, and adjustment techniques using radioactivation datafrom multiple foil irradiations.9.6.2 The reaction rate is related

39、 to the fluence rate by thefollowing equation:Rs5*0sE!f E!dE (10)where:s(E) = activation cross section at energy E, andf(E) = differential neutron fluence rate, that is the fluenceper unit energy per unit time for neutrons withenergies between E and E +dE.9.6.3 The production rate of a radioactive n

40、uclide is relatedto the reaction rate by the following equation:dn/dt 5 NRs2 nl8 (11)9.6.4 Solution of Eq 11, for the case where Rsand N areconstant, yields the following expression for the activity of afoil:A 5 l/l8! NRs1 2 exp 2l8t! (12)9.6.5 The saturation activity of a foil is defined as theacti

41、vity when dn/dt = 0; thus Eq 11 yields the followingrelationship for the saturation activity:As5 l/l8! NRs(13)9.6.6 The isotopic content of the target nuclide may bereduced during the irradiation by more than one transmutationprocess and it may be increased by transmutation of othernuclides so that

42、the rate of change of the number of targetnuclei with time is described by a number of terms:dN/dt 5 N Rs1(t 5 1nRi! 2(j 5 1mNjRj(14)where:i = discrete transmutation path for removal of the targetisotope, andj = discrete transmutation reaction whereby the target iso-tope is produced from isotope Nja

43、nd each of the RiandRjterms could be calculated from equations similar toEq 10, using the appropriate cross sections.9.6.6.1 The Rsterm may predominate and, if Rsis constant,Eq 14 can be solved as N = Noexp ( Rst). The change in thetarget composition may be negligible and N may be approxi-mated by N

44、o.9.6.7 During irradiation, the effective decay rate is increasedby transmutations of the product isotope (see Eq 4).9.7 Long Term Irradiations:9.7.1 Long irradiations for materials testing programs andreactor pressure vessel surveillance are common. Long irradia-tions usually involve operation at v

45、arious power levels, includ-ing extended zero-power periods; thus, appropriate correctionsmust be made for depletion of the target nuclide, decay andburnout of the radioactive nuclide, and variations in neutronfluence rate. Multiple irradiations and nuclide burnup must alsobe considered in short-irr

46、adiation calculations where reaction-product half-lives are relatively short and nuclide cross sec-tions are high.E2610339.7.2 The total irradiation period can be divided into acontinuous series of periods during each of which f(E)isessentially constant. Then the activity generated during the ithirr

47、adiation period is:Ai5 lNiRs/l8!i#1 2 exp 2l8iti! (15)where:Ni= number of target atoms, andti= duration of the ithperiod.9.7.2.1 The activity remaining from the ithperiod at the endof the nthperiod can be calculated as the following equation:An!i5 Aiexp 2(j 5 i 1 1nl8jtj! (16)9.7.2.2 The total activ

48、ity of the foil at the end of theirradiation duration is thus the sum of all the (An)iterms.9.7.3 If the product of (l8iti) is very small for all irradiationperiods, the values of Aicalculated from Eq 15 are proportionalto (Rs)iand ti.9.7.3.1 If the spectral averaged cross section is also constantov

49、er all irradiation periods, (Rs)iis proportional to the magni-tude of the neutron fluence rate.9.7.3.2 It is normally assumed that the fluence rate isdirectly proportional to the power generation rate in theadjacent fuel.9.7.4 Under the conditions assumed in 9.7.3, Eq 15 can bewritten as:Ai5 AsPi/P! 1 2 exp 2l8ti!, (17)and Eq 16 can be written as:An!i5 As SNiNoDKi1 2 exp 2l 8ti! (18)where:As= the saturation activity corresponding to a referencepower level, P,Pi= actual power generation rate during the irradiationperiod,Ki=Pi/P! expS2l8(