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ASTM E844-03

(Guide)

Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706(IIC)

Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706(IIC)

SIGNIFICANCE AND USE
In neutron dosimetry, a fission or non-fission dosimeter, or combination of dosimeters, can be used for determining a fluence-rate, fluence, or neutron spectrum, or both, in nuclear reactors. Each dosimeter is sensitive to a specific energy range, and, if desired, increased accuracy in a flux-spectrum can be achieved by the use of several dosimeters each covering specific neutron energy ranges.
A wide variety of detector materials is used for various purposes. Many of these substances overlap in the energy of the neutrons which they will detect, but many different materials are used for a variety of reasons. These reasons include available analysis equipment, different cross sections for different flux levels and spectra, preferred chemical or physical properties, and, in the case of radiometric dosimeters, varying requirements for different half-life isotopes, possible interfering activities, and chemical separation requirements.
SCOPE
1.1 This guide covers the selection, design, irradiation, post-irradiation handling, and quality control of neutron dosimeters (sensors), thermal neutron shields, and capsules for reactor surveillance neutron dosimetry.  
1.2 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only.  
1.3  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
09-Feb-2003
ICS
17.240 - Radiation measurements
Technical Committee
E10 - Nuclear Technology and Applications
Drafting Committee
E10.05 - Nuclear Radiation Metrology
Current Stage
Ref Project

Relations

Replaces
ASTM E844-97 - Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706(IIC)
Effective Date
10-Feb-2003
Replaced By
ASTM E844-09 - Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706(IIC)
Effective Date
01-Jun-2009
Referred By
ASTM E264-19 - Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Nickel
Effective Date
10-Feb-2003
Referred By
ASTM E2956-23 - Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels
Effective Date
10-Feb-2003
Referred By
ASTM E266-23 - Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Aluminum
Effective Date
10-Feb-2003
Referred By
ASTM E704-19 - Standard Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238
Effective Date
10-Feb-2003
Referred By
ASTM E1855-20 - Standard Test Method for Use of 2N2222A Silicon Bipolar Transistors as Neutron Spectrum Sensors and Displacement Damage Monitors
Effective Date
10-Feb-2003
Referred By
ASTM E2215-19 - Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
Effective Date
10-Feb-2003
Referred By
ASTM E1005-21 - Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
Effective Date
10-Feb-2003
Referred By
ASTM E944-19 - Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance
Effective Date
10-Feb-2003
Referred By
ASTM E721-22 - Standard Guide for Determining Neutron Energy Spectra from Neutron Sensors for Radiation-Hardness Testing of Electronics
Effective Date
10-Feb-2003
Referred By
ASTM E185-21 - Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
Effective Date
10-Feb-2003
Referred By
ASTM E720-23 - Standard Guide for Selection and Use of Neutron Sensors for Determining Neutron Spectra Employed in Radiation-Hardness Testing of Electronics
Effective Date
10-Feb-2003
Referred By
ASTM E1035-18(2023) - Standard Practice for Determining Neutron Exposures for Nuclear Reactor Vessel Support Structures
Effective Date
10-Feb-2003
Referred By
ASTM E705-18 - Standard Test Method for Measuring Reaction Rates by Radioactivation of Neptunium-237
Effective Date
10-Feb-2003

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ASTM E844-03 - Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706(IIC)ASTM E844-03 - Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706(IIC)
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ASTM E844-03 - Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706(IIC)
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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:E844–03
Standard Guide for
Sensor Set Design and Irradiation for Reactor Surveillance,
1
E 706(IIC)
This standard is issued under the fixed designation E844; 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 (e) indicates an editorial change since the last revision or reapproval.
1. Scope E706(IIE) Benchmark Testing of Reactor Vessel Dosim-
3
etry
1.1 This guide covers the selection, design, irradiation,
post-irradiation handling, and quality control of neutron do-
3. Terminology
simeters (sensors), thermal neutron shields, and capsules for
3.1 Definitions:
reactor surveillance neutron dosimetry.
3.1.1 neutron dosimeter, sensor, monitor—a substance irra-
1.2 The values stated in inch-pound units are to be regarded
diated in a neutron environment for the determination of
as the standard. The values given in parentheses are for
neutron fluence rate, fluence, or spectrum, for example: radio-
information only.
metricmonitor(RM),solidstatetrackrecorder(SSTR),helium
1.3 This standard does not purport to address all of the
accumulation fluence monitor (HAFM), damage monitor
safety problems, if any, associated with its use. It is the
(DM), temperature monitor (TM).
responsibility of the user of this standard to establish appro-
3.1.2 thermal neutron shield—a substance (that is, cad-
priate safety and health practices and determine the applica-
mium, boron, gadolinium) that filters or absorbs thermal
bility of regulatory limitations prior to use.
neutrons.
2. Referenced Documents 3.2 Fordefinitionsorothertermsusedinthisguide,referto
Terminology E170.
2.1 ASTM Standards:
E170 Terminology Relating to Radiation Measurements
4. Significance and Use
2
and Dosimetry
4.1 In neutron dosimetry, a fission or non-fission dosimeter,
E261 Practice for Determining Neutron Fluence Rate, Flu-
2 or combination of dosimeters, can be used for determining a
ence, and Spectra by Radioactivation Techniques
fluence-rate, fluence, or neutron spectrum, or both, in nuclear
E854 Test Method for Application and Analysis of Solid
reactors.Eachdosimeterissensitivetoaspecificenergyrange,
State Track Recorder (SSTR) Monitors for Reactor Sur-
2 and, if desired, increased accuracy in a flux-spectrum can be
veillance, E706(IIIB)
achieved by the use of several dosimeters each covering
E910 Test Method forApplication andAnalysis of Helium
specific neutron energy ranges.
Accumulation Fluence Monitors for Reactor Vessel Sur-
2 4.2 Awide variety of detector materials is used for various
veillance, E706(IIIC)
purposes. Many of these substances overlap in the energy of
E1005 TestMethodforApplicationandAnalysisofRadio-
the neutrons which they will detect, but many different
metric Monitors for Reactor Vessel Surveillance,
2 materials are used for a variety of reasons. These reasons
E706(IIIA)
include available analysis equipment, different cross sections
E706(IIID) Analysis of Damage Monitors for ReactorVes-
3 for different flux levels and spectra, preferred chemical or
sel Surveillance
physical properties, and, in the case of radiometric dosimeters,
E706(IIIE) Analysis of Temperature Monitors for Reactor
3 varying requirements for different half-life isotopes, possible
Vessel Surveillance
interfering activities, and chemical separation requirements.
5. Selection of Neutron Dosimeters and Thermal Neutron
1
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear
Shields
Technology and Applications and is the direct responsibility of Subcommittee
5.1 Neutron Dosimeters:
E10.05 on Nuclear Radiation Metrology.
Current edition approved Feb. 10, 2003. Published February 2003. Originally
5.1.1 The choice of dosimeter material depends largely on
approved in 1981. Last previous edition approved in 1997 as E844–97.
the dosimetry technique employed, for example, radiometric
2
Annual Book of ASTM Standards, Vol 12.02.
3 monitors, helium accumulation monitors, track recorders, and
For standards that are in the draft stage and have not received an ASTM
designation, see Section 5 as well as Figures 1 and 2 of Matrix E706. damagemonitors.Atthepresenttime,thereisawidevarietyof
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.
1

---------------------- Page: 1 ----------------------
E844–03
detectormaterialsusedtoperformneutrondosimetrymeasure- lives that are short compared to the irradiation duration should
ments.Thesearegenerallyintheformoffoils,wires,powders, not be used. Generally, radionuclides with half-lives less than
and salts. The use of alloys is valuable for certain applications three times the irradiation duration should be avoided unless
such as (1) d
...

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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.  
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.
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ASTM E181-23(Guide)
Standard Guide for Detector Calibration and Analysis of Radionuclides in Radiation Metrology for Reactor Dosimetry

ABSTRACT
These methods cover general procedures for the calibration of radiation detectors and the analysis of radionuclides. For each individual radionuclide, one or more of these methods may apply. These methods are concerned only with specific radionuclide measurements. The chemical and physical properties of the radionuclides are beyond the scope of this standard. Among the measurement standards discussed are: the calibration and usage of germanium detectors, scintillation detector systems, scintillation detectors for simple and complex spectra, and counting methods such as beta particle counting, aluminum absorption curve, alpha particle counting, and liquid scintillation counting. For each of the methods, the scope, apparatus used, summary of methods, preparation of apparatus, calibration procedure, measurement of radionuclide, performance testing, sources of uncertainty, precautions and tests, and calculations are detailed.
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1.1 This guide covers general procedures for the calibration of radiation detectors and measurement for radiation metrology for reactor dosimetry. For any particular radionuclide, one or more of these methods may apply.  
1.2 These techniques are concerned only with specific radionuclide measurements. The chemical and physical properties of the radionuclides are not within the scope of this standard.  
1.3 E3376, Standard Practice for Calibration and Usage of Germanium Detectors in Radiation Metrology for Reactor Dosimetry, was previously in Guide E181 and is now found in Volume 12.02 of the Annual Book of ASTM Standards. The discussion herein is not a sufficient substitute for the full standard. This guide is specifically NOT to be used as a direct reference to Practice E3376. Only the standard listed provides sufficient information to serve as a reference.  
1.4 Additional information on the setup, calibration, and quality control for radiometric detectors and measurements is given in Guide C1402 and Practice D7282.  
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.  
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ASTM E3376-23(Practice)
Standard Practice for Calibration and Usage of Germanium Detectors in Radiation Metrology for Reactor Dosimetry

SIGNIFICANCE AND USE
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  
1.2 Additional information on the setup, calibration, and quality control for radiometric detectors and measurements is given in IEEE/ANSI N42.14 and in Guide C1402 and Practice D7282.  
1.3 The values stated in SI units are generally to be regarded as standard. The rad is an exception.  
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Standard Terminology Relating to Radiation Measurements and Dosimetry
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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.  
4.3 This practice gives a means of estimating the response of CaF2(Mn) TLDs to neutrons. This neutron response is then subtracted from the measured response to determine the TLD response due to gamma rays. The procedure has relatively high uncertainty because the neutron response of CaF2(Mn) TLDs may vary depending on the source of the material, and this procedure is a generic calculation applicable to CaF2(Mn) TLDs independent of their manufacturer/source. The neutron response given in this practice is a summary of CaF2(Mn) TLD responses reported in the literature. The associated uncertainty envelops the range of results reported and includes the variety of CaF2(Mn) TLDs used as well as the uncertainties in the det...
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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.  
1.2 More precise experimental techniques may be applied if the uncertainty derived from this practice is larger than the level that the user can accept. These more precise techniques are not discussed here. The references in Section 8 describe some of these techniques.  
1.3 This practice does not discuss effects on the TLD reading from neutron interactions with the material surrounding the TLD and used to ensure a charged particle equilibrium. These effects will depend on the isotopic composition of the surrounding material and its thickness, and on the incident neutron spectrum  (1).2  
1.4 The values stated in SI units are to be regarded as standard.  
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 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 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 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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ASTM
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.
SCOPE
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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