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ASTM E3376-23

(Practice)

Standard Practice for Calibration and Usage of Germanium Detectors in Radiation Metrology for Reactor Dosimetry

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.  
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.

General Information

Status
Published
Publication Date
14-Feb-2023
ICS
17.240 - Radiation measurements
Technical Committee
E10 - Nuclear Technology and Applications
Drafting Committee
E10.05 - Nuclear Radiation Metrology
Current Stage
Ref Project

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ASTM E3376-23 - Standard Practice for Calibration and Usage of Germanium Detectors in Radiation Metrology for Reactor DosimetryASTM E3376-23 - Standard Practice for Calibration and Usage of Germanium Detectors in Radiation Metrology for Reactor Dosimetry
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ASTM E3376-23 - Standard Practice for Calibration and Usage of Germanium Detectors in Radiation Metrology for Reactor Dosimetry
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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.
Designation: E3376 − 23
Standard Practice for
Calibration and Usage of Germanium Detectors in Radiation
1
Metrology for Reactor Dosimetry
This standard is issued under the fixed designation E3376; 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 2. Referenced Documents
3
1.1 This standard establishes techniques for calibration,
2.1 ASTM Standards:
usage, and performance testing of germanium detectors for the C1402 Guide for High-Resolution Gamma-Ray Spectrom-
measurement of gamma-ray emission rates of radionuclides in
etry of Soil Samples
radiation metrology for reactor dosimetry. The practice is D7282 Practice for Setup, Calibration, and Quality Control
applicable only to samples of small size, approximating to
of Instruments Used for Radioactivity Measurements
point sources. It covers the energy and full-energy peak E170 Terminology Relating to Radiation Measurements and
efficiency calibration as well as the determination of gamma-
Dosimetry
ray energies in the 0.06 MeV to 2 MeV energy region and is 4
2.2 IEEE/ANSI Standard:
designed to yield gamma-ray emission rates with an uncer-
N42.14 Calibration and Usage of Germanium Detectors
tainty of 63 % (see Note 1). This technique applies to
Spectrometers for Measurement of Gamma-Ray Emission
measurements that do not involve overlapping peaks, and in
Rates of Radionuclides
which peak-to-continuum considerations are not important.
NOTE 1—Uncertainty U is given at the 68 % confidence level; that is, 3. Terminology
2 2
=
U5 Σσ 11⁄3Σδ where δ are the estimated maximum systematic
i i i 3.1 Definitions:
uncertainties, and σ are the random uncertainties at the 68 % confidence
i
3.1.1 certified radioactivity standard source—a calibrated
2
level. Other techniques of error analysis are in use (1, 2).
radioactive source, with stated uncertainties, whose calibration
1.2 Additional information on the setup, calibration, and
is certified by the source supplier as traceable to an interna-
quality control for radiometric detectors and measurements is
tional or national standards laboratory.
given in IEEE/ANSI N42.14 and in Guide C1402 and Practice
3.1.2 check source—a radioactivity source, not necessarily
D7282.
calibrated, that is used to confirm the continuing satisfactory
1.3 The values stated in SI units are generally to be regarded
operation of an instrument.
as standard. The rad is an exception.
3.1.3 correlated photon summing—the simultaneous detec-
1.4 This standard does not purport to address all of the
tion of two or more photons originating from a single nuclear
safety concerns, if any, associated with its use. It is the
disintegration.
responsibility of the user of this standard to establish appro-
3.1.4 dead time—the time after a triggering pulse during
priate safety, health, and environmental practices and deter-
which the system is unable to retrigger.
mine the applicability of regulatory limitations prior to use.
1.5 This international standard was developed in accor- 3.1.5 FWHM—(full width at half maximum) the full width
of a gamma-ray peak distribution measured at half the maxi-
dance with internationally recognized principles on standard-
ization established in the Decision on Principles for the mum ordinate above the continuum.
Development of International Standards, Guides and Recom-
3.1.6 FWTM—(full width at tenth maximum) the full width
mendations issued by the World Trade Organization Technical
of a gamma-ray peak distribution measured at one tenth of the
Barriers to Trade (TBT) Committee.
maximum ordinate above the continuum.
1
This practice is under the jurisdiction of ASTM Committee E10 on Nuclear
3
Technology and Applications and is the direct responsibility of Subcommittee For referenced ASTM standards, visit the ASTM website, www.astm.org, or
E10.05 on Nuclear Radiation Metrology. contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
Current edition approved Feb. 15, 2023. Published April 2023. DOI: 10.1520/ Standards volume information, refer to the standard’s Document Summary page on
E3376-23. the ASTM website.
2 4
The boldface numbers in parentheses refer to a list of references at the end of Available from Institute of Electrical and Electronics Engineers, Inc. (IEEE),
this standard. 445 Hoes Ln., Piscataway, NJ 08854-4141, http://www.ieee.org.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United State
...

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5.1 Refer to Guide E844 for the selection, irradiation, and quality control of neutron dosimeters.  
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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.  
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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.
SCOPE
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.  
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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.  
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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  
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ASTM E526-22(Test Method)
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SIGNIFICANCE AND USE
5.1 Refer to Guide E844 for the selection, irradiation, and quality control of neutron dosimeters.  
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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.  
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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...
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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.  
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1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
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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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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 F526-21(Test Method)
Standard Test Method for Using Calorimeters for Total Dose Measurements in Pulsed Linear Accelerator or Flash X-ray Machines

SIGNIFICANCE AND USE
5.1 An accurate measure of the total absorbed dose is necessary to ensure the validity of the data taken, to enable comparison to be made of data taken at different facilities, and to verify that components or circuits are tested to the radiation specification applied to the system for which they are to be used.  
5.2 The primary value of a calorimetric method for measuring dose is that the results are absolute. They are based only on physical properties of materials, that is, the specific heat of the calorimeter-block material and the Seebeck EMF of the thermocouple used or the temperature coefficient of resistance (α) of the thermistor used, all of which can be established with non-radiation measurements.  
5.3 The method permits repeated measurements to be made without requiring entry into the radiation cell between measurements.
SCOPE
1.1 This test method covers a calorimetric measurement of the total absorbed dose delivered by a single pulse of electrons from an electron linear accelerator or a flash X-ray machine (FXR, e-beam mode). The test method is designed for use with pulses of electrons in the energy range from 10 to 50 MeV and is only valid for cases in which both the calorimeter and the test specimen to be irradiated are “thin” compared to the range of these electrons in the materials of which they are constructed.  
1.2 The procedure described can be used in those cases in which (1) the dose delivered in a single pulse is 5 Gy(matl)2 [500 rd (matl)] or greater, or (2) multiple pulses of a lower dose can be delivered in a short time compared to the thermal time constant of the calorimeter. The units for the total absorbed dose delivered to a material require the specification of the material and the notation “matl” refers to the active material of the calorimeter. The minimum dose per pulse that can be acceptably monitored depends on the variables of the particular test, including pulse rate, pulse uniformity, and the thermal time constant of the calorimeter.  
1.3 A determination of the total dose is made directly for the material of which the calorimeter block is made. The total dose in other materials can be calculated from this measured value using Eq 3 presented in this test method. The need for such calculations and the choice of materials for which calculations are to be made shall be subject to agreement by the parties to the test.  
1.4 The values stated in SI units are to be regarded as the standard. The values in parenthesis are provided for information only.  
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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