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ASTM E266-92(1996)

(Test Method)

Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Aluminum

Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Aluminum

SCOPE
1.1 This test method describes procedures measuring reaction rates by the activation reaction  27 Al(n,[alpha]) 24 Na.
1.2 This activation reaction is useful for measuring neutrons with energies above approximately 6.5 MeV and for irradiation times up to about 2 days (for longer irradiations, see Test Method E261).
1.3 With suitable techniques, fission-neutron fluence rates above 10 [dot]cm -2 [dot]s -1  can be determined.
1.4 Detailed procedures for other fast neutron detectors are referenced in Test Method E261.
1.5 This standard does not purport to address all of the safety problems, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

General Information

Status
Historical
Publication Date
31-Dec-1995
ICS
17.240 - Radiation measurements
27.120.30 - Fissile materials and nuclear fuel technology
Technical Committee
E10 - Nuclear Technology and Applications
Drafting Committee
E10.05 - Nuclear Radiation Metrology
Current Stage
Ref Project

Relations

Replaced By
ASTM E266-02 - Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Aluminum
Effective Date
10-Jun-2002

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ASTM E266-92(1996) - Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of AluminumASTM E266-92(1996) - Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Aluminum
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ASTM E266-92(1996) - Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Aluminum
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Standards Content (Sample)


Designation: E 266 – 92 (Reapproved 1996)
Standard Test Method for
Measuring Fast-Neutron Reaction Rates by Radioactivation
of Aluminum
This standard is issued under the fixed designation E 266; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (e) indicates an editorial change since the last revision or reapproval.
1. Scope 4. Summary of Test Method
1.1 This test method covers procedures measuring reaction 4.1 High-purity aluminum is irradiated in a neutron field,
7 24 24 27 24
rates by the activation reaction Al(n,a) Na. thereby producing radioactive Na from the Al(n,a) Na
1.2 This activation reaction is useful for measuring neutrons activation reaction.
with energies above approximately 6.5 MeV and for irradiation 4.2 The gamma rays emitted by the radioactive decay of
times up to about 2 days (for longer irradiations, see Practice Na are counted (see Test Methods E 181) and the reaction
E 261). rate, as defined by Practice E 261, is calculated from the decay
1.3 With suitable techniques, fission-neutron fluence rates rate and irradiation conditions.
6 −2 −1
above 10 ·cm ·s can be determined. 4.3 The neutron fluence rate above about 6.5 MeV can then
1.4 Detailed procedures for other fast neutron detectors are be calculated from the spectral-weighted neutron activation
referenced in Practice E 261. cross section as defined by Practice E 261.
1.5 This standard does not purport to address all of the
5. Significance and Use
safety concerns, if any, associated with its use. It is the
responsibility of the user of this standard to establish appro- 5.1 Refer to Guide E 844 for the selection, irradiation, and
quality control of neutron dosimeters.
priate safety and health practices and determine the applica-
bility of regulatory limitations prior to use. 5.2 Refer to Practice E 261 for a general discussion of the
determination of fast-neutron fluence rate with threshold de-
2. Referenced Documents
tectors.
2.1 ASTM Standards: 5.3 Pure aluminum in the form of foil or wire is readily
E 170 Terminology Relating to Radiation Measurements available and easily handled.
2 24
and Dosimetry 5.4 Na has a half-life of 15.0 h and emits gamma rays with
E 181 Test Methods for Detector Calibration and Analysis energies of 1.369 and 2.754 MeV.
of Radionuclides 5.5 Fig. 1 shows a plot of cross section versus neutron
27 24
E 261 Practice for Determining Neutron Fluence Rate, Flu- energy for the fast-neutron reaction Al(n,a) Na. This figure
ence, and Spectra by Radioactivation Techniques is for illustrative purposes only to indicate the range of
E 844 Guide for Sensor Set Design and Irradiation for response of the Al(n,a) reaction. Refer to Guide E 1018 for
Reactor Surveillance, E 706(IIC) descriptions of recommended tabulated dosimetry cross sec-
E 944 Guide for Application of Neutron Spectrum Adjust- tions.
2 28 27
ment Methods in Reactor Surveillance, (IIA) 5.6 Two competing activities, Al and Mg, are formed in
27 28 27 27
E 1005 Test Method for Application and Analysis of Radio- the reactions Al(n,g) Al and Al(n,p) Mg, respectively,
metric Monitors for Reactor Vessel Surveillance, but these can be eliminated by waiting 2 h before counting.
E706(IIIA)
6. Apparatus
E 1018 Guide for Application of ASTM Evaluated Cross
Section Data File, Matrix E 706(IIB) 6.1 NaI(T1) or High Resolution Gamma-Ray Spectrometer.
Because of its high resolution, the germanium detector is
3. Terminology
useful when contaminant activities are present (see Test Meth-
3.1 Definitions: ods E 181 and E 1005).
3.1.1 Refer to Terminology E 170. 6.2 Precision Balance, able to achieve the required accu-
racy.
This test method is under the jurisdiction of ASTM Committee E-10 on Nuclear
Technology and Applications and is the direct responsibility of Subcommittee
E10.05 on Nuclear Radiation Metrology. Kocher, D. C., “Radioactive Decay Data Tables” DOE/TIC-11026, April 1981.
Current edition approved Oct. 15, 1992. Published May 1993. Originally “International Reactor Dosimetry File (IRDF–90),” assembled by N.P.
published as E 266 – 65 T. Last previous edition E 266 – 87. Kocherov, et al., International Atomic Energy Agency, Nuclear Data Section,
Annual Book of ASTM Standards, Vol 12.02. IAEA-NDS-141, Rev. 0, August 1990.
Copyright © ASTM, 100 Barr Harbor Drive, West Conshohocken, PA 19428-2959, United States.
E 266
27 24
FIG. 1 Al(n,a) Na Cross Section
6.3 Digital Computer, useful for data analysis (optional). 8.5 After irradiation, the sample should be thoroughly
rinsed in warm water. This will remove any Na surface
7. Materials
contamination produced during irradiation.
7.1 The purity of the aluminum is important. No impurities 8.6 Check the sample for activity from cross-contamination
should be present that produce long-lived gamma-ray-emitting
by other irradiated materials. Clean, if necessary, and reweigh.
radionuclides
...

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ASTM E481-23(Practice)
Standard Practice for Measuring Neutron Fluence Rates by Radioactivation of Cobalt and Silver

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3.1 This practice uses one monitor (cobalt) with a nearly 1/v absorption cross-section curve and a second monitor (silver) with a large resonance peak so that its resonance integral is large compared to its thermal cross section. The pertinent data for these two reactions are given in Table 1. The equations are based on the Westcott formalism ((2, 3) and Practice E261) and determine a Westcott 2200 m/s neutron fluence rate nv0 and the Westcott epithermal index parameter  . References (4-6) contain a general discussion of the two-reaction test method. In this practice, the absolute activities of both cobalt and silver monitors are determined. This differs from the test method in the references wherein only one absolute activity is determined.  (A) The numbers in parentheses following given values are the uncertainty in the last digit(s) of the value; 0.729 (8) means 0.729 ± 0.008, 70.8(1) means 70.8 ± 0.1.(B) The decay constant, λ, is defined as ln(2) / t1/2 with units of sec–1, where t1/2 is the nuclide half-life in seconds.(C) Calculated using Eq 10.(D) In Fig. 1, Θ = 4ErkT/AΓ2 = 0.2 corresponds to the value for 109Ag for T = 293 K, ∑r = N0σr,max,T=0Kσr,max,T=0K = 31138.03 barn at 5.19 eV (13). The value of σr,max,T=0K = 31138.03 barns is calculated using the Breit-Wigner single-level resonance formula  where the 109Ag atomic mass is A = 108.9047558 amu (14), the ENDF/B-VIII.0 (MAT = 4731) (13) resonance parameters are: resonance total width Γ = 0.1427333 eV, formation neutron width Γn = 0.0127333 eV, and radiative/decay width Γγ = 0.13 eV, with a resonance spin J=1, and the statistical spin factor  where s1 = 1/2 and s2 = 1/2 are the spins of the two particles (neutron and 109Ag ground state (15)) forming resonance.  
3.2 The advantages of this approach are the elimination of four difficulties associated with the use of cadmium: (1) the perturbation of the field by the cadmium; (2) the inexact cadmium cut-off energy; (3) the low melting temperature of cadmium; a...
SCOPE
1.1 This practice covers a suitable means of obtaining the thermal neutron fluence rate, or fluence, in nuclear reactor environments where the use of cadmium, as a thermal neutron shield as described in Test Method E262, is undesirable for reasons such as potential spectrum perturbations or due to temperatures above the melting point of cadmium.  
1.2 The reaction 59Co(n,γ )60Co results in a well-defined gamma emitter having a half-life of 5.2711 years2 (8)3 (1).4 The reaction 109Ag(n,γ)110mAg results in a nuclide with a well-known, complex decay scheme with a half-life of 249.78 (2) days (1). Both cobalt and silver are available either in very pure form or alloyed with other metals such as aluminum. A reference source of cobalt in aluminum alloy to serve as a neutron fluence rate monitor wire standard is available from the National Institute of Standards and Technology (NIST) as Standard Reference Material (SRM) 953.5 The competing activities from neutron activation of other isotopes are eliminated, for the most part, by waiting for the short-lived products to die out before counting. With suitable techniques, thermal neutron fluence rate in the range from 108 cm−2·s−1 to 3 × 1015 cm−2·s−1 can be measured. Two calculational practices are described in Section 9 for the determination of neutron fluence rates. The practice described in 9.3 may be used in all cases. This practice describes a means of measuring a Westcott neutron fluence rate in 9.2 (Note 1) by activation of cobalt- and silver-foil monitors (see Terminology E170). For the Wescott Neutron Fluence Convention method to be applicable, the measurement location must be well moderated and be well represented by a Maxwellian low-energy distribution and an (1/E) epithermal distribution. These conditions are usually only met in positions surrounded by hydrogenous moderator without nearby strongly multiplying or absorbing materials.
Note 1: Westcott fluence rate    ...

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ASTM E266-23(Test Method)
Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Aluminum

SIGNIFICANCE AND USE
5.1 Refer to Guide E844 for the selection, irradiation, and quality control of neutron dosimeters.  
5.2 Refer to Practice E261 for a general discussion of the determination of fast-neutron fluence rate with threshold detectors.  
5.3 Pure aluminum in the form of foil or wire is readily available and easily handled. 27Al has an abundance of 100 % (1).3  
5.4 24Na has a half-life of 14.958 (2)4 h (2) and emits gamma rays with energies of 1.368630 (5) and 2.754049 (13) MeV (2).  
5.5 Fig. 1 shows a plot of the International Reactor Dosimetry and Fusion File (IRDFF-II) cross section (3, 4) versus neutron energy for the fast-neutron reaction  27Al(n,α) 24Na (3) along with a comparison to the current experimental database (5, 6). While the RRDF-2008 and IRDFF-1.05 cross sections extend from threshold up to 60 MeV, due to considerations of the available validation data, the energy region over which this standard recommends use of this cross section for reactor dosimetry applications only extends from threshold at ~4.25 MeV up to 20 MeV. This figure is for illustrative purposes and is used to indicate the range of response of the  27Al(n,α) reaction. Refer to Guide E1018 for recommended sources for the tabulated dosimetry cross sections.
FIG. 1 27Al(n,α)24Na Cross Section, from IRDFF-II Library, with EXFOR Experimental Data  
5.6 Two competing activities, 28Al (2.25 (2) minute half-life) and 27Mg (9.458 (12) minute half-life), are formed in the reactions 27Al(n,γ)28Al and 27Al(n,p)27Mg, respectively, but these can be eliminated by waiting 2 h before counting.
SCOPE
1.1 This test method covers procedures measuring reaction rates by the activation reaction  27Al(n,α)24Na.  
1.2 This activation reaction is useful for measuring neutrons with energies above approximately 6.5 MeV and for irradiation times up to about two days (for longer irradiations, or when there are significant variations in reactor power during the irradiation, see Practice E261).  
1.3 With suitable techniques, fission-neutron fluence rates above 106 cm−2·s−1 can be determined.  
1.4 Detailed procedures for other fast neutron detectors are referenced in Practice E261.  
1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.6 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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ASTM E526-22(Test Method)
Standard Test Method for Measuring Fast-Neutron Reaction Rates By Radioactivation of Titanium

SIGNIFICANCE AND USE
5.1 Refer to Guide E844 for the selection, irradiation, and quality control of neutron dosimeters.  
5.2 Refer to Practice E261 for a general discussion of the determination of fast-neutron fluence rate with threshold detectors.  
5.3 Titanium has good physical strength, is easily fabricated, has excellent corrosion resistance, has a melting temperature of 1668 °C, and can be obtained with satisfactory purity.  
5.4 46Sc has a half-life of 83.787 (16)4 days (2). The 46Sc decay emits a 0.889271 (2) MeV gamma 99.98374 (35) % of the time and a second gamma with an energy of 1.120537 (3) MeV 99.97 (2) % of the time.  
5.5 The recommended “representative isotopic abundances” for natural titanium (3) are:    
8.25 (3) % 46Ti  
7.44 (2) % 47Ti  
73.72 (2) % 48Ti  
5.41 (2) % 49Ti  
5.18 (2) % 50Ti  
5.6 The radioactive products of the neutron reactions 47Ti(n,p)47Sc (τ1/2 = 3.3485 (9) d) (2) and 48Ti(n,p)48Sc (τ1/2 = 43.67 h), (3) might interfere with the analysis of 46Sc.  
5.7 Contaminant activities (for example, 65Zn and 182Ta) might interfere with the analysis of 46Sc. See 7.1.2 and 7.1.3 for more details on the 182Ta and 65Zn interference.  
5.8 46Ti and 46Sc have cross sections for thermal neutrons of 0.59 ± 0.18 and 8.0 ± 1.0 barns, respectively (4); therefore, when an irradiation exceeds a thermal-neutron fluence greater than about 2 × 1021 cm–2, provisions should be made to either use a thermal-neutron shield to prevent burn-up of 46Sc or measure the thermal-neutron fluence rate and calculate the burn-up.  
5.9 Fig. 1 shows a plot of the International Reactor Dosimetry and Fusion File, IRDFF-II cross section (5) versus neutron energy for the fast-neutron reactions of titanium which produce 46Sc (that is, natTi(n,X)46Sc). Included in the plot is the 46Ti(n,p) reaction and the 47Ti(n,np:d) contributions to the 46Sc production, normalized per natTi atom with the individual isotopic contributions weighted using the natural abundances (3). This figure ...
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1.1 This test method covers procedures for measuring reaction rates by the activation reaction natTi(n,X)46Sc. The “X” designation represents any combination of light particles associated with the production of the residual 46Sc product. Within the applicable neutron energy range for fission reactor applications, this reaction is a properly normalized combination of three different reaction channels: 46Ti(n,p)46Sc; 47Ti(n, np)46Sc; and 47Ti(n,d)46Sc.
Note 1: The 47Ti(n,np)46Sc reaction, ENDF-6 format file/reaction identifier MF=3, MT=28, is distinguished from the 47Ti(n,d)46Sc reaction, ENDF-6 format file/reaction identifier MF=3/MT=104, even though it leads to the same residual product (1).2 The combined reaction, in the IRDFF-II library, has the file/reaction identifier MF=10/MT=5.
Note 2: The cross section for the combined 47Ti(n,np:d) reaction is relatively small for energies less than 12 MeV and, in fission reactor spectra, the production of the residual 46Sc is not easily distinguished from that due to the 46Ti(n,p) reaction.  
1.2 The reaction is useful for measuring neutrons with energies above approximately 4.4 MeV and for irradiation times, under uniform power, up to about 250 days (for longer irradiations, or for varying power levels, see Practice E261).  
1.3 With suitable techniques, fission-neutron fluence rates above 109 cm–2·s–1 can be determined. However, in the presence of a high thermal-neutron fluence rate, 46Sc depletion should be investigated.  
1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice E261.  
1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices an...

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ASTM E496-14(2022)(Test Method)
Standard Test Method for Measuring Neutron Fluence and Average Energy from 3H(d,n)4He Neutron Generators by Radioactivation Techniques

SIGNIFICANCE AND USE
5.1 Refer to Practice E261 for a general discussion of the measurement of fast-neutron fluence rates with threshold detectors.  
5.5.1 Fig. 5 (2) shows how the neutron energy depends upon the angle of scattering in the laboratory coordinate system when the incident deuteron has an energy of 150 keV and is incident on a thick and a thin tritiated target. For thick targets, the incident deuteron loses energy as it penetrates the target and produces neutrons of lower energy. A thick target is defined as a target thick enough to completely stop the incident deuteron. The two curves in Fig. 5, for both thick and thin targets, come from different sources. The dashed line calculations come from Ref (3); the solid curve calculations come from Ref (4); and the measured data come from Ref (5). The dash-dot curve and the right-hand axis give the difference between the calculated neutron energies for thin and thick targets. Computer codes are available to assist in calculating the expected thick and thin target yield and neutron spectrum for various incident deuteron energies (6).
FIG. 5 Dependence of 3H(d,n)4He Neutron Energy on Angle (2)  
5.6 The Q-value for the primary 3H(d,n)4He reaction is +17.59 MeV. When the incident deuteron energy exceeds 3.71 MeV and 4.92 MeV, the break-up reactions 3H(d,np)3H and 3H(d,2n)3He, respectively, become energetically possible. Thus, at high deuteron energies (>3.71 MeV) this reaction is no longer monoenergetic. Monoenergetic neutron beams with energies from about 14.8 to 20.4 MeV can be produced by this reaction at forward laboratory angles (7).  
5.7 It is recommended that the dosimetry sensors be fielded in the exact positions where the dosimetry results are wanted. There are a number of factors that can affect the monochromaticity or energy spread of the neutron beam (7, 8). These factors include the energy regulation of the incident deuteron energy, energy loss in retaining windows if a gas target is used or energy loss within t...
SCOPE
1.1 This test method covers a general procedure for the measurement of the fast-neutron fluence rate produced by neutron generators utilizing the 3H(d,n)4He reaction. Neutrons so produced are usually referred to as 14-MeV neutrons, but range in energy depending on a number of factors. This test method does not adequately cover fusion sources where the velocity of the plasma may be an important consideration.  
1.2 This test method uses threshold activation reactions to determine the average energy of the neutrons and the neutron fluence at that energy. At least three activities, chosen from an appropriate set of dosimetry reactions, are required to characterize the average energy and fluence. The required activities are typically measured by gamma-ray spectroscopy.  
1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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ASTM E261-16(2021)(Practice)
Standard Practice for Determining Neutron Fluence, Fluence Rate, and Spectra by Radioactivation Techniques

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5.1 Transmutation Processes—The effect on materials of bombardment by neutrons depends on the energy of the neutrons; therefore, it is important that the energy distribution of the neutron fluence, as well as the total fluence, be determined.
SCOPE
1.1 This practice describes procedures for the determination of neutron fluence rate, fluence, and energy spectra from the radioactivity that is induced in a detector specimen.  
1.2 The practice is directed toward the determination of these quantities in connection with radiation effects on materials.  
1.3 For application of these techniques to reactor vessel surveillance, see also Test Methods E1005.  
1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.
Note 1: Detailed methods for individual detectors are given in the following ASTM test methods: E262, E263, E264, E265, E266, E343, E393, E481, E523, E526, E704, E705, and E854.  
1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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ASTM E523-21e1(Test Method)
Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Copper

SIGNIFICANCE AND USE
5.1 Refer to Guide E844 for the selection, irradiation, and quality control of neutron dosimeters.  
5.2 Refer to Practice E261 for a general discussion of the measurement of fast neutron fluence rate with threshold detectors. The general shape of the  63Cu(n,α) 60Co cross section is also shown in Fig. 1 (3, 4, 5) along with a comparison to the current experimental database (6). This figure is for illustrative purposes only to indicate the range of the response of the  63Cu(n,α)60Co reaction. Refer to Guide E1018 for descriptions of recommended tabulated dosimetry cross sections.
FIG. 1 63Cu(n,α)60Co Cross Section with EXFOR Experimental Data  
Note 1: The cross section appropriate for use under this standard is from the IRDFF-II library (5) which, up to an incident neutron energy of 20 MeV, is drawn from the RRDF-2002 library (3) and is identical to the adopted cross section in the IRDF-2002 library (4). See Guide E1018.  
5.3 The major advantages of copper for measuring fast-neutron fluence rate are that it has good strength, is easily fabricated, has excellent corrosion resistance, has a melting temperature of 1083°C, and can be obtained in high purity. The half-life of  60 Co is long and its decay scheme is simple and well known.  
5.4 The disadvantages of copper for measuring fast neutron fluence rate are the high reaction apparent threshold of 4.5 MeV, the possible interference from cobalt impurity (>1 μg/g), the reported possible thermal component of the (n,α) reaction, and the possibly significant cross sections for thermal neutrons for  63Cu and  60Co [that is, 4.50(2) and 2.0(2) barns, respectively], (7), which will require burnout corrections at high fluences.
SCOPE
1.1 This test method covers procedures for measuring reaction rates by the activation reaction  63Cu(n,α) 60Co. The cross section for  60Co produced in this reaction increases rapidly with neutrons having energies greater than about 4.5 MeV. 60Co decays with a half-life of 5.2711(8)2 years (1)3,4 and emits two gamma rays having energies of 1.173228(3) and 1.332492(4) MeV (1). The isotopic content of natural copper is 69.174(20) %  63Cu and 30.826(20) %  65Cu (2). The neutron reaction,  63Cu(n,γ)64Cu, produces a radioactive product that emits gamma rays [1.34577(6) MeV (E1005)] which might interfere with the counting of the  60Co gamma rays.  
1.2 With suitable techniques, fission-neutron fluence rates above 109 cm−2·s−1  can be determined. The  63Cu(n,α)60Co reaction can be used to determine fast-neutron fluences for irradiation times up to about 15 years, provided that the analysis methods described in Practice E261 are followed. If dosimeters are analyzed after irradiation periods longer than 15 years, the information inferred about the fluence during irradiation periods more than 15 years before the end of the irradiation should not be relied upon without supporting data from dosimeters withdrawn earlier.  
1.3 Detailed procedures for other fast-neutron detectors are referenced in Practice E261.  
1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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ASTM E265-15(2020)(Test Method)
Standard Test Method for Measuring Reaction Rates and Fast-Neutron Fluences by Radioactivation of Sulfur-32

SIGNIFICANCE AND USE
5.1 Refer to Guides E720 and E844 for the selection, irradiation, and quality control of neutron dosimeters.  
5.2 Refer to Practice E261 for a general discussion of the determination of fast-neutron fluence and fluence rate with threshold detectors.  
5.3 The activation reaction produces  32P, which decays by the emission of a single beta particle in 100 % of the decays, and which emits no gamma rays. The half life of  32P is 14.284 (36)3 days (1)  4 and the maximum beta energy is 1710.66 (21) keV (1).  
5.4 Elemental sulfur is readily available in pure form and any trace contaminants present do not produce significant amounts of radioactivity. Natural sulfur, however, is composed of  32S (94.99 % (26)),  34S (4.25 % (24)) (2), and trace amounts of other sulfur isotopes. The presence of these other isotopes leads to several competing reactions that can interfere with the counting of the 1710-keV beta particle. This interference can usually be eliminated by the use of appropriate techniques, as discussed in Section 8.
SCOPE
1.1 This test method describes procedures for measuring reaction rates and fast-neutron fluences by the activation reaction  32S(n,p)32P.  
1.2 This activation reaction is useful for measuring neutrons with energies above approximately 3 MeV.  
1.3 With suitable techniques, fission-neutron fluences from about 5 × 108 to 1016 n/cm 2 can be measured.  
1.4 Detailed procedures for other fast-neutron detectors are described in Practice E261.  
1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.6 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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ASTM E704-19(Test Method)
Standard Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238

SIGNIFICANCE AND USE
5.1 Refer to Practice E261 for a general discussion of the determination of fast-neutron fluence rate with fission detectors.  
5.2 238U is available as metal foil, wire, or oxide powder (see Guide E844). It is usually encapsulated in a suitable container to prevent loss of, and contamination by, the 238U and its fission products.  
5.3 One or more fission products can be assayed. Pertinent data for relevant fission products are given in Table 1 and Table 2. (A) The lightface numbers in parentheses are the magnitude of plus or minus uncertainties in the last digit(s) listed.(B) With 137mBa (2.552 min) in equilibrium.(C) The recommended half-life and gamma emission probabilities have been taken from the Reference (3) data that was recommended at the time that the recommended fission yields were set.(D) Probability of daughter 140La decay.(E) This is the activity ratio of 140La/140Ba after reached transient equilibrium (t ≥ 19 days).  (A) The JEFF-3.1/3.1.1 radioactive decay data and fission yields sub-libraries, JEFF Report 20, OECD 2009, Nuclear Energy Agency (5).(B) All yield data given as a %; RC represents a cumulative yield; RI represents an independent yield.  
5.3.1 137Cs-137mBa is chosen frequently for long irradiations. Radioactive products 134Cs and 136Cs may be present, which can interfere with the counting of the 0.662 MeV  137Cs-137mBa gamma rays (see Test Method E320).  
5.3.2 140Ba-140La is chosen frequently for short irradiations (see Test Method E393).  
5.3.3 95Zr can be counted directly, following chemical separation, or with its daughter 95Nb using a high-resolution gamma detector system.  
5.3.4 144Ce is a high-yield fission product applicable to 2- to 3-year irradiations.  
5.4 It is necessary to surround the 238U monitor with a thermal neutron absorber to minimize fission product production from a quantity of 235U in the 238U target and from  239Pu from (n,γ) reactions in the 238U material. Assay of the 239Pu concentration when a signif...
SCOPE
1.1 This test method covers procedures for measuring reaction rates by assaying a fission product (F.P.) from the fission reaction 238U(n,f)F.P.  
1.2 The reaction is useful for measuring neutrons with energies from approximately 1.5 to 7 MeV and for irradiation times up to 30 to 40 years, provided that the analysis methods described in Practice E261 are followed.  
1.3 Equivalent fission neutron fluence rates as defined in Practice E261 can be determined.  
1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice E261.  
1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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ASTM
ASTM E264-19(Test Method)
Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Nickel

SIGNIFICANCE AND USE
5.1 Refer to Guide E844 for the selection, irradiation, and quality control of neutron dosimeters.  
5.2 Refer to Practice E261 for a general discussion of the determination of fast-neutron fluence rate with threshold detectors.  
5.3 Pure nickel in the form of foil or wire is readily available, and easily handled.  
5.4 58Co has a half-life of 70.85 (3) days (Refs (1) and (2))3 and emits a gamma ray with an energy of 810.7602 (20) keV (Refs (2) and (3)).  
5.5 Competing activities  65Ni(2.5172 h) and  57Ni(35.9 (3) h (Ref (2)) are formed by the reactions  64Ni(n,γ)  65Ni, and 58Ni(n,2n)57Ni, respectively.  
5.6 A second 9.04 h isomer,  58mCo, is formed that decays to 70.85-day  58Co. Loss of  58Co and  58mCo by thermal-neutron burnout will occur in environments (Refs (4) and (5) having thermal fluence rates of 3 × 1012  cm−2·s −1  and above. Burnout correction factors,  R, are plotted as a function of time for several thermal fluxes in Fig. 1. Tabulated values for a continuous irradiation time are provided in Hogg, et al. (Ref (5))  
5.7 Fig. 2 shows a plot of cross section (Ref (6)) versus energy for the fast-neutron reaction  58Ni(n,p) 58Co. This figure is for illustrative purposes only to indicate the range of response of the  58Ni(n,p) reaction. Refer to Guide E1018 for descriptions of recommended tabulated dosimetry cross sections.
FIG. 2 58Ni(n,p)58Co Cross Section  
Note 1: The data is taken from the Evaluated Nuclear Data File, ENDF/B-VI, rather than the later ENDF/B-VII. This is in accordance with E1018, section 6.1, since the later ENDF/B-VII data files do not include covariance information. For more details see Section H of Ref (7).
SCOPE
1.1 This test method covers procedures for measuring reaction rates by the activation reaction  58Ni(n,p)58Co.
FIG. 1 R Correction Values as a Function of Irradiation Time and Neutron Flux
Note 1: The burnup corrections were computed using effective burn-up cross sections of 1650 b for 58Co(n,γ) and 1.4E5 b for 58mCo(n,γ).  
1.2 This activation reaction is useful for measuring neutrons with energies above approximately 2.1 MeV and for irradiation times up to about 200 days in the absence of high thermal neutron fluence rates, provided that the analysis methods described in Practice E261 are followed. If dosimeters are analyzed after irradiation periods longer than 200 days, the information inferred about the fluence during irradiation periods more than 200 days before the end of the irradiation should not be relied upon without supporting data from dosimeters withdrawn earlier.  
1.3 With suitable techniques fission-neutron fluence rates densities above 107  cm−2·s −1 can be determined.  
1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice E261.  
1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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ASTM
ASTM E263-18(Test Method)
Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Iron

SIGNIFICANCE AND USE
5.1 Refer to Guide E844 for guidance on the selection, irradiation, and quality control of neutron dosimeters.  
5.2 Refer to Practice E261 for a general discussion of the determination of fast-neutron fluence rate with threshold detectors.  
5.3 Pure iron in the form of foil or wire is readily available and easily handled.  
5.4 Fig. 1 shows a plot of cross section as a function of neutron energy for the fast-neutron reaction  54Fe(n,p)54Mn (1).3 This figure is for illustrative purposes only to indicate the range of response of the  54Fe(n,p)54Mn reaction. Refer to Guide E1018 for recommended tabulated dosimetry cross sections.
FIG. 1 54Fe(n,p)54Mn Cross Section  
5.5 54Mn has a half-life of 312.19 (3) days4 (2) and emits a gamma ray with an energy of 834.855 (3) keV (2).  
5.6 Interfering activities generated by neutron activation arising from thermal or fast neutron interactions are 2.57878 (46)-h  56Mn, 44.494 (12) days  59Fe, and 5.2711 (8) years  60Co (2,3). (Consult the latest version of Ref (2) for more precise values currently accepted for the half-lives.) Interference from  56Mn can be eliminated by waiting 48 h before counting. Although chemical separation of  54Mn from the irradiated iron is the most effective method for eliminating  59Fe and  60Co, direct counting of iron for  54Mn is possible using high-resolution detector systems or unfolding or stripping techniques, especially if the dosimeter was covered with cadmium or boron during irradiation. Altering the isotopic composition of the iron dosimeter is another useful technique for eliminating interference from extraneous activities when direct sample counting is to be employed.  
5.7 The vapor pressures of manganese and iron are such that manganese diffusion losses from iron can become significant at temperatures above about 700°C. Therefore, precautions must be taken to avoid the diffusion loss of  54Mn from iron dosimeters at high temperature. Encapsulating the iron dosimeter in quartz o...
SCOPE
1.1 This test method describes procedures for measuring reaction rates by the activation reaction 54Fe(n,p)54Mn.  
1.2 This activation reaction is useful for measuring neutrons with energies above approximately 2.2 MeV and for irradiation times up to about three years, provided that the analysis methods described in Practice E261 are followed. If dosimeters are analyzed after irradiation periods longer than three years, the information inferred about the fluence during irradiation periods more than three years before the end of the irradiation should not be relied upon without supporting data from dosimeters withdrawn earlier.  
1.3 With suitable techniques, fission-neutron fluence rates above 108  cm−2·s−1  can be determined. However, in the presence of a high thermal-neutron fluence rate (for example, >2 × 1014  cm−2·s −1)  54Mn depletion should be investigated.  
1.4 Detailed procedures describing the use of other fast-neutron detectors are referenced in Practice E261.  
1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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