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ASTM E1893-15

(Guide)

Standard Guide for Selection and Use of Portable Radiological Survey Instruments for Performing In Situ Radiological Assessments to Support Unrestricted Release from Further Regulatory Controls

Standard Guide for Selection and Use of Portable Radiological Survey Instruments for Performing In Situ Radiological Assessments to Support Unrestricted Release from Further Regulatory Controls

SIGNIFICANCE AND USE
4.1 The purpose of this standard is to provide the user information and guidance for selecting and using instrumentation that will provide measurement results that can be compared to criteria for unrestricted use.  
4.2 Use of this standard will provide greater assurance that the measurements obtained will be technically and administratively sufficient for making decisions regarding completion of decontamination and/or demolition/removal activities.  
4.3 Use of this standard will provide greater assurance that the measurements obtained will be technically and administratively sufficient to meet all applicable regulatory requirements for unrestricted release of a component for recycle or reuse, or for unrestricted release of a remaining surface or area.
SCOPE
1.1 This standard provides recommendations on the selection and use of portable instrumentation that is responsive to levels of radiation that are close to natural background. These instruments are employed to detect the presence of residual radioactivity that is at, or below, the criteria for release from further regulatory control of a component to be salvaged or reused, or a surface remaining at the conclusion of decontamination and/or decommissioning.  
1.2 The choice of these instruments, their operating characteristics and the protocols by which they are calibrated and used will provide adequate assurance that the measurements of the residual radioactivity meet the requirements established for release from further regulatory control.  
1.3 This standard is applicable to the in situ measurement of radioactive emissions that include:  
1.3.1 alpha  
1.3.2 beta (electrons)  
1.3.3 gamma  
1.3.4 characteristic x-rays  
1.3.5 The measurement of neutron emissions is not included as part of this standard.  
1.4 This standard dose not address instrumentation used to assess residual radioactivity levels contained in air samples, surface contamination smears, bulk material removals, or half/whole body personnel monitors.  
1.5 This standard does not address records retention requirements for calibration, maintenance, etc. as these topics are considered in several of the referenced documents.  
1.6 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

General Information

Status
Historical
Publication Date
31-Jan-2015
ICS
17.240 - Radiation measurements
Technical Committee
E10 - Nuclear Technology and Applications
Drafting Committee
E10.03 - Radiological Protection for Decontamination and Decommissioning of Nuclear Facilities and Components
Current Stage
Ref Project

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ASTM E1893-15 - Standard Guide for Selection and Use of Portable Radiological Survey Instruments for Performing In Situ Radiological Assessments to Support Unrestricted Release from Further Regulatory ControlsASTM E1893-15 - Standard Guide for Selection and Use of Portable Radiological Survey Instruments for Performing In Situ Radiological Assessments to Support Unrestricted Release from Further Regulatory Controls
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ASTM E1893-15 - Standard Guide for Selection and Use of Portable Radiological Survey Instruments for Performing In Situ Radiological Assessments to Support Unrestricted Release from Further Regulatory Controls
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REDLINE ASTM E1893-15 - Standard Guide for Selection and Use of Portable Radiological Survey Instruments for Performing In Situ Radiological Assessments to Support Unrestricted Release from Further Regulatory ControlsREDLINE ASTM E1893-15 - Standard Guide for Selection and Use of Portable Radiological Survey Instruments for Performing In Situ Radiological Assessments to Support Unrestricted Release from Further Regulatory Controls
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Standards Content (Sample)

NOTICE: This standard has either been superseded and replaced by a new version or withdrawn.
Contact ASTM International (www.astm.org) for the latest information
Designation: E1893 − 15
Standard Guide for
Selection and Use of Portable Radiological Survey
Instruments for Performing In Situ Radiological
Assessments to Support Unrestricted Release from Further
1
Regulatory Controls
This standard is issued under the fixed designation E1893; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision.Anumber in parentheses indicates the year of last reapproval.A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope 1.6 The values stated in SI units are to be regarded as
standard. No other units of measurement are included in this
1.1 This standard provides recommendations on the selec-
standard.
tion and use of portable instrumentation that is responsive to
levels of radiation that are close to natural background. These
2. Referenced Documents
instruments are employed to detect the presence of residual
2
2.1 ASTM Standards:
radioactivity that is at, or below, the criteria for release from
C998Practice for Sampling Surface Soil for Radionuclides
further regulatory control of a component to be salvaged or
C999Practice for Soil Sample Preparation for the Determi-
reused, or a surface remaining at the conclusion of decontami-
nation of Radionuclides
nation and/or decommissioning.
C1000Test Method for Radiochemical Determination of
1.2 The choice of these instruments, their operating charac-
Uranium Isotopes in Soil by Alpha Spectrometry
teristics and the protocols by which they are calibrated and
C1133Test Method for Nondestructive Assay of Special
usedwillprovideadequateassurancethatthemeasurementsof
Nuclear Material in Low-Density Scrap and Waste by
theresidualradioactivitymeettherequirementsestablishedfor
Segmented Passive Gamma-Ray Scanning
release from further regulatory control.
E170Terminology Relating to Radiation Measurements and
Dosimetry
1.3 Thisstandardisapplicabletotheinsitumeasurementof
E181Test Methods for Detector Calibration andAnalysis of
radioactive emissions that include:
Radionuclides
1.3.1 alpha
C1215Guide for Preparing and Interpreting Precision and
1.3.2 beta (electrons)
Bias Statements in Test Method Standards Used in the
1.3.3 gamma
Nuclear Industry
1.3.4 characteristic x-rays
2.2 ANSI Standards:
1.3.5 Themeasurementofneutronemissionsisnotincluded
ANSI N323ABAmerican National Standard for Radiation
as part of this standard.
Protection Instrumentation Test and Calibration, Portable
3
Survey Instrumentation
1.4 This standard dose not address instrumentation used to
ANSI N42.17AAmerican National Standard for Perfor-
assess residual radioactivity levels contained in air samples,
mance Specifications for Health Physics Instrumentation-
surface contamination smears, bulk material removals, or
Portable Instrumentation for Use in Normal Environmen-
half/whole body personnel monitors.
3
tal Conditions
1.5 Thisstandarddoesnotaddressrecordsretentionrequire-
ANSI N42.17CAmerican National Standard for Perfor-
ments for calibration, maintenance, etc. as these topics are
mance Specifications for Health Physics Instrumentation-
considered in several of the referenced documents.
PortableInstrumentationforUseinExtremeEnvironmen-
3
tal Conditions
1
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear
2
Technology and Applications and is the direct responsibility of Subcommittee For referenced ASTM standards, visit the ASTM website, www.astm.org, or
E10.03 on Radiological Protection for Decontamination and Decommissioning of contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
Nuclear Facilities and Components. Standards volume information, refer to the standard’s Document Summary page on
Current edition approved Feb. 1, 2015. Published April 2009. Originally the ASTM website.
3
approved in 1997. Last previous edition approved in 2008 as E1893-08a. DOI: Available fromAmerican National Standards Institute (ANSI), 25 W. 43rd St.,
10.1520/E1893-15. 4th Floor, New York, NY 10036, http://www.ansi.org.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

---------------------- Page: 1 ----------------------
E1893 − 15
2.3 National Council on Radiation Protection and Measure- 3.3 calibration source, n—asusedinthisstandardguide,see
ments: certified reference material.
NCRPReportNo.57InstrumentationandMonitoringMeth-
3.4 certified reference material, n—a material that has been
ods for Radiation Protection, National Council on Radia-
characterizedbyarecognizedstandardortestinglaboratoryfor
4
tion Protection and Measurements, May 1978
someofitschemicalorphysicalorradiologicalproperties,and
NCRP Report No. 58A Handbook of Radioactivity Mea-
that is generally used for calibration of a measur
...

This document is not an ASTM standard and is intended only to provide the user of an ASTM standard an indication of what changes have been made to the previous version. Because
it may not be technically possible to adequately depict all changes accurately, ASTM recommends that users consult prior editions as appropriate. In all cases only the current version
of the standard as published by ASTM is to be considered the official document.
Designation: E1893 − 08a E1893 − 15
Standard Guide for
Selection and Use of Portable Radiological Survey
Instruments for Performing In Situ Radiological
Assessments to Support Unrestricted Release from Further
1
Regulatory Controls
This standard is issued under the fixed designation E1893; 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 (´) indicates an editorial change since the last revision or reapproval.
1. Scope
1.1 This standard provides recommendations on the selection and use of portable instrumentation that is responsive to levels
of radiation that are close to natural background. These instruments are employed to detect the presence of residual radioactivity
that is at, or below, the criteria for release from further regulatory control of a component to be salvaged or reused, or a surface
remaining at the conclusion of decontamination and/or decommissioning.
1.2 The choice of these instruments, their operating characteristics and the protocols by which they are calibrated and used will
provide adequate assurance that the measurements of the residual radioactivity meet the requirements established for release from
further regulatory control.
1.3 This standard is applicable to the in situ measurement of radioactive emissions that include:
1.3.1 alpha
1.3.2 beta (electrons)
1.3.3 gamma
1.3.4 characteristic x-rays
1.3.5 The measurement of neutron emissions is not included as part of this standard.
1.4 This standard dose not address instrumentation used to assess residual radioactivity levels contained in air samples, surface
contamination smears, bulk material removals, or half/whole body personnel monitors.
1.5 This standard does not address records retention requirements for calibration, maintenance, etc. as these topics are
considered in several of the referenced documents.
1.6 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.
2. Referenced Documents
2
2.1 ASTM Standards:
C998 Practice for Sampling Surface Soil for Radionuclides
C999 Practice for Soil Sample Preparation for the Determination of Radionuclides
C1000 Test Method for Radiochemical Determination of Uranium Isotopes in Soil by Alpha Spectrometry
C1133 Test Method for Nondestructive Assay of Special Nuclear Material in Low-Density Scrap and Waste by Segmented
Passive Gamma-Ray Scanning
E170 Terminology Relating to Radiation Measurements and Dosimetry
E181 Test Methods for Detector Calibration and Analysis of Radionuclides
C1215 Guide for Preparing and Interpreting Precision and Bias Statements in Test Method Standards Used in the Nuclear
Industry
1
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee E10.03 on
Radiological Protection for Decontamination and Decommissioning of Nuclear Facilities and Components.
Current edition approved Nov. 1, 2008Feb. 1, 2015. Published March 2009.April 2009. Originally approved in 1997. Last previous edition approved in 2008 as
E1893-08.-08a. DOI: 10.1520/E1893-08A.10.1520/E1893-15.
2
For referenced ASTM standards, visit the ASTM website, www.astm.org, or contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM Standards
volume information, refer to the standard’s Document Summary page on the ASTM website.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

---------------------- Page: 1 ----------------------
E1893 − 15
2.2 ANSI Standards:
ANSI N323BN323AB American National Standard for Radiation Protection Instrumentation Test and Calibration, Portable
3
Survey Instrumentation for Near Background Operation
ANSI N42.17A American National Standard for Performance Specifications for Health Physics Instrumentation-Portable
3
Instrumentation for Use in Normal Environmental Conditions
ANSI N42.17C American National Standard for Performance Specifications for Health Physics Instrumentation-Portable
3
Instrumentation for Use in Extreme Environmental Conditions
2.3 National Council on Radiation Protection and Measurements:
NCRP Report No. 57 Instrumentation and Monitoring Methods for Radiation Protection, National Council on Radiation
4
Protection and Measurements, May 1978
NCRP Report No. 58 A Handbook of Radioactivity Measurement Procedures, National
...

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

SIGNIFICANCE AND USE
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.  
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ASTM E3376-23(Practice)
Standard Practice for Calibration and Usage of Germanium Detectors in Radiation Metrology for Reactor Dosimetry

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5.1 High-purity germanium detectors are used for precise gamma-ray spectroscopy for the purpose of determining radioactivity in materials. Typical applications include monitoring, mapping, and characterization of neutron energy spectra in nuclear reactors or isotopic fission sources.
SCOPE
1.1 This standard establishes techniques for calibration, usage, and performance testing of germanium detectors for the measurement of gamma-ray emission rates of radionuclides in radiation metrology for reactor dosimetry. The practice is applicable only to samples of small size, approximating to point sources. It covers the energy and full-energy peak efficiency calibration as well as the determination of gamma-ray energies in the 0.06 MeV to 2 MeV energy region and is designed to yield gamma-ray emission rates with an uncertainty of ±3 % (see Note 1). This technique applies to measurements that do not involve overlapping peaks, and in which peak-to-continuum considerations are not important.
Note 1: Uncertainty U is given at the 68 % confidence level; that is,  where δi are the estimated maximum systematic uncertainties, and σi are the random uncertainties at the 68 % confidence level. Other techniques of error analysis are in use (1, 2).2  
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ASTM E2450-23(Practice)
Standard Practice for Application of CaF2(Mn) Thermoluminescence Dosimeters in Mixed Neutron-Photon Environments

SIGNIFICANCE AND USE
4.1 Electronic devices are typically tested for device response to gamma radiation in pure gamma-ray fields. Testing electronic device response against neutrons is more complex since there is invariably a gamma-ray component in addition to the neutron field. The gamma-ray response of the electronic device is typically subtracted from the overall response to find the device response to neutrons. This approach to the testing requires a determination of the gamma-ray exposure in the mixed field. To enhance the neutron effects, the radiation field is sometimes selected to have as large a neutron component as possible.  
4.2 CaF2(Mn) TLDs are often used to monitor the gamma-ray dose in mixed neutron/gamma radiation fields. Since the dosimeters are exposed along with the device under test to the mixed field, their response must be corrected for neutrons. In a field rich in neutrons, the uncertainty in the interpretation of the TLD response grows. In fields with relatively few neutrons, the total TLD response may be used to make a correction for gamma response of the device under test. Under this condition, the relative uncertainty in the TLD neutron response is not likely to drive the overall uncertainty in the correction to the electronic device response.  
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1.1 This practice describes a procedure for correcting a CaF2(Mn) thermoluminescence dosimeter (TLD) reading for its response to neutrons during the irradiation. The neutron response may be subtracted from the total TLD response to give the gamma-ray response. In fields with a large neutron contribution to the total response, this procedure may result in large uncertainties.  
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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 ...
SCOPE
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
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...
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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 E1005-21(Test Method)
Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance

SIGNIFICANCE AND USE
5.1 Radiometric monitors shall provide a proven passive dosimetry technique for the determination of neutron fluence rate (flux density), fluence, and spectrum in a diverse variety of neutron fields. These data are required to evaluate and estimate probable long-term radiation-induced damage to nuclear reactor structural materials such as the steel used in reactor pressure vessels and their support structures.  
5.2 A number of radiometric monitors, their corresponding neutron activation reactions, and radioactive reaction products and some of the pertinent nuclear parameters of these RMs and products are listed in Table 1. Table 2 provides data (37) on the cumulative and independent fission yields of the important fission monitors. Not included in these tables are contributions to the yields from photo-fission, which can be especially significant for non-fissile nuclides (2-5, 27-29, 38-41). (A) All yield data are given as a percentage with associated uncertainties given as percentages of the percentage at the 1σ level.(B) For this fission yield evaluation (37), “Fast” indicates that the data was extracted from a wide range of reactor-based fission neutron spectra that can be characterized as having an average energy of ~0.4 MeV. Almost all of the fission reactions for U-238 and Th-232 occur above an effective threshold energy of ~1 MeV and, for Np-237, above ~0.2 MeV.
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1.1 This test method describes procedures for measuring the specific activities of radioactive nuclides produced in radiometric monitors (RMs) by nuclear reactions induced during surveillance exposures for reactor vessels and support structures. More detailed procedures for individual RMs are provided in separate standards identified in 2.1 and in Refs (1-5).2 The measurement results can be used to define corresponding neutron induced reaction rates that can in turn be used to characterize the irradiation environment of the reactor vessel and support structure. The principal measurement technique is high resolution gamma-ray spectrometry, although X-ray photon spectrometry and Beta particle counting are used to a lesser degree for specific RMs (1-29).  
1.1.1 The measurement procedures include corrections for detector background radiation, random and true coincidence summing losses, differences in geometry between calibration source standards and the RMs, self absorption of radiation by the RM, other absorption effects, radioactive decay corrections, and burn out of the nuclide of interest (6-26).  
1.1.2 Specific activities are calculated by taking into account the time duration of the count, the elapsed time between start of count and the end of the irradiation, the half life, the mass of the target nuclide in the RM, and the branching intensities of the radiation of interest. Using the appropriate half life and known conditions of the irradiation, the specific activities may be converted into corresponding reaction rates (2-5, 28-30).  
1.1.3 Procedures for calculation of reaction rates from the radioactivity measurements and the irradiation power time history are included. A reaction rate can be converted to neutron fluence rate and fluence using the appropriate integral cross section and effective irradiation time values, and, with other reaction rates can be used to define the neutron spectrum through the use of suitable computer programs (2-5, 28-30).  
1.1.4 The use of benchmark neutron fields for calibration of RMs can reduce significantly or eliminate systematic errors since many parameters, and their respective uncertainties, required for calculation of absolute reaction rates are common to both the benchmark and test measurements and therefore are self canceling. The benchmark equivalent fluence rates, for the environment tested, can be calculated from a direct ratio of the measured saturated activities in the two environments and the certified benchmark fluence rate (2-5, 28-30).  
1.2 This test meth...

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

SIGNIFICANCE AND USE
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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