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ASTM E665-94

(Practice)

Practice for Using Absorbed Dose Versus Depth in Materials to Verify the X-ray Output of Flash X-ray Machines (Withdrawn 2000)

Practice for Using Absorbed Dose Versus Depth in Materials to Verify the X-ray Output of Flash X-ray Machines (Withdrawn 2000)

SCOPE
1.1 This practice covers a procedure for determining absorbed dose versus depth in materials exposed to the X-ray output of flash X-ray machines. It is applicable to all machines whose photon energy spectra have maximum energies ranging from 100 keV to 20 MeV. The determination provides information related to the photon spectrum from the flash X-ray machine. Specifically this measurement can be used for verifying expected X-ray spectra obtained from machine diagnostics and radiation transport calculations and for determining the appropriate equilibrator thickness to be used for dosimetric measurements.
1.2 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 and health practices and determine the applicability of regulatory limitations prior to use.

General Information

Status
Withdrawn
Publication Date
31-Dec-1993
Withdrawal Date
09-Jun-2000
ICS
17.240 - Radiation measurements
Technical Committee
E10 - Nuclear Technology and Applications
Drafting Committee
E10.07 - Radiation Dosimetry for Radiation Effects on Materials and Devices
Current Stage
Ref Project

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ASTM E665-94 - Practice for Using Absorbed Dose Versus Depth in Materials to Verify the X-ray Output of Flash X-ray Machines (Withdrawn 2000)ASTM E665-94 - Practice for Using Absorbed Dose Versus Depth in Materials to Verify the X-ray Output of Flash X-ray Machines (Withdrawn 2000)
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ASTM E665-94 - Practice for Using Absorbed Dose Versus Depth in Materials to Verify the X-ray Output of Flash X-ray Machines (Withdrawn 2000)
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Standards Content (Sample)

Designation: E 665 – 94
Standard Practice for
Using Absorbed Dose Versus Depth in Materials to Verify
1
the X-ray Output of Flash X-ray Machines
This standard is issued under the fixed designation E 665; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (e) indicates an editorial change since the last revision or reapproval.
1. Scope 4. Summary of Practice
1.1 This practice covers a procedure for determining ab- 4.1 The first step is to expose a stack consisting of layers of
sorbed dose versus depth in materials exposed to the X-ray material interspersed with thin dosimeters such as thermolu-
output of flash X-ray machines. It is applicable to all machines minescence dosimeters (TLDs). This is done at a large enough
whose photon energy spectra have maximum energies ranging distance from the machine so that the direction of the incident
from 100 keV to 20 MeV. The determination provides infor- radiation can be well defined.
mation related to the photon spectrum from the flash X-ray 4.2 A set of small spheres containing TLDs is then con-
machine. Specifically this measurement can be used for veri- structed and used to check the absorbed dose versus depth
fying expected X-ray spectra obtained from machine diagnos- closer to the source where no single direction of the incident
tics and radiation transport calculations and for determining the radiation determines the absorbed dose.
appropriate equilibrator thickness to be used for dosimetric 4.3 Compare measured dose versus depth to calculated
measurements. profile using predicted X-ray output spectrum.
1.2 This standard does not purport to address all of the
NOTE 1—For a comprehensive discussion of various dosimetry meth-
safety concerns, if any, associated with its use. It is the
ods applicable to pulsed radiation fields, see ICRU Report 34.
responsibility of the user of this standard to establish appro-
priate safety and health practices and determine the applica- 5. Significance and Use
bility of regulatory limitations prior to use.
5.1 The X-ray output of flash X-ray machines is often very
difficult to measure directly. Hence, experimenters must often
2. Referenced Documents
rely on predicted spectra that are calculated using the specific
2.1 ASTM Standards:
bremsstrahlung converter and machine current and voltage
4
E 170 Terminology Relating to Radiation Measurements
diagnostics as input to radiation transport codes (1-5). Al-
2
and Dosimetry
though these techniques work fairly well when the diagnostics
E 666 Practice for Calculating Absorbed Dose from Gamma
are well calibrated and the machine is performing normally, it
2
or X Radiation
is prudent to check these predictions with some type of
E 668 Practice for Application of Thermoluminescence-
measured data.
Dosimetry (TLD) Systems for Determining Absorbed Dose
5.2 Because the dose as a function of depth within a
2
in Radiation-Hardness Testing of Electronic Devices
material is a strong function of the photon energy, such a
2.2 International Commission on Radiation Units and
measurement is a convenient means of verifying the predicted
Measurements (ICRU) Reports:
output of the machine.
3
ICRU Report 33—Radiation Quantities and Units
5.3 The radiation transport codes can again be used to
3
ICRU Report 34—The Dosimetry of Pulsed Radiation
predict the depth-dose profile in the material using the pre-
dicted spectra and can be compared with the measurements (6
3. Terminology
and 7). Agreement between these two indicates the accuracy of
3.1 Terminology used in this practice can be found in
the predicted spectra.
Terminology E 170 and ICRU Report 33
5.4 If there is a disagreement, the machine parameters used
as input to the code can be varied to produce agreement with
the measurements and indicate the magnitude of the discrep-
ancy. This will not only demonstrate a difference in the
1
This practice is under the jurisdiction of ASTM Committee E-10 on Nuclear
machine output from the expected, but also suggest the
Technology and Applications and is the direct responsibility of Subcommittee
E10.07 on Radiation Effects on Electronic Materials, Components, and Devices. possible source of the performance change.
Current edition approved Sept. 15, 1994. Published November 1994. Originally
published as E 665 – 78. Last previous edition E 665 – 78 (1984).
2
Annual Book of ASTM Standards, Vol 12.02.
3
4
Available for the International Commission on Radiation Units and Measure-
The boldface numbers in parentheses refer to a list of references at the end of
ments, 7910 Woodmont Ave., Suite 800, Bethesda, MD 20814.
this practice.
Copyright © ASTM, 100 Barr Harbor Drive, West Conshohocken, PA 19428-2959, United States.
1

---------------------- Page: 1 ----------------------
E 665
6. Interferences low-energy machines and thicker layers for high-energy ma-
c
...

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5.1 Refer to Guide E844 for the selection, irradiation, and quality control of neutron dosimeters.  
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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.
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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.  
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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.  
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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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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.
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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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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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ASTM E2450-23(Practice)
Standard Practice for Application of CaF2(Mn) Thermoluminescence Dosimeters in Mixed Neutron-Photon Environments

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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.  
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  
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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.  
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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.  
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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.  
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Standard Practice for Determining Neutron Fluence, Fluence Rate, and Spectra by Radioactivation Techniques

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

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ASTM
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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