E10.05 - Nuclear Radiation Metrology

SIGNIFICANCE AND USE
5.1 High levels of antimony are commonly used in flame retardant formulations for various materials. NAA is a test method that can be useful for verifying these levels and, for other materials, NAA can also be useful in establishing the amount of low level contamination, if any, with high sensitivity and high precision.  
5.2 Neutron activation analysis provides a rapid, highly sensitive, nondestructive procedure for antimony determination in a wide range of matrices. This test method is independent of the chemical form of the antimony.  
5.3 This test method can be used for quality and process control in the petrochemical and other manufacturing industries, and for research purposes in a broad spectrum of applications.
SCOPE
1.1 This test method covers the measurement of antimony concentration in plastics or other hydrocarbon or organic matrix by using neutron activation analysis (NAA). The sample is activated by irradiation with neutrons from a research reactor and the subsequently emitted gamma-rays are detected with a germanium semiconductor detector. The same system may be used to determine antimony concentrations ranging from 1 ng/g to 10 000 μg/g with the lower end of the range limited by numerous interferences and the upper limit established by the demonstrated practical application of NAA.  
1.2 This test method may be used on either solid or liquid samples, provided that they can be made to conform in size and shape during irradiation and counting to a standard sample of known antimony content using very simple sample preparation. Several variants of this method have been described in the technical literature. A monograph is available which provides a comprehensive description of the principles of neutron activation analysis using reactor neutrons (1).2  
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. Specific precautions are given in Section 9.  
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.

SIGNIFICANCE AND USE
3.1 The objectives of a reactor vessel surveillance program are twofold. The first requirement of the program is to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region resulting from exposure to neutron irradiation and the thermal environment. The second requirement is to make use of the data obtained from the surveillance program to determine the conditions under which the vessel can be operated throughout its service life.  
3.1.1 To satisfy the first requirement of 3.1, the tasks to be carried out are straightforward. Each of the irradiation capsules that comprise the surveillance program may be treated as a separate experiment. The goal is to define and carry to completion a dosimetry program that will, a posteriori, describe the neutron field to which the material test specimens were exposed. The resultant information will then become part of a database applicable in a stricter sense to the specific plant from which the capsule was removed, but also in a broader sense to the industry as a whole.  
3.1.2 To satisfy the second requirement of 3.1, the tasks to be carried out are somewhat complex. The objective is to describe accurately the neutron field to which the pressure vessel itself will be exposed over its service life. This description of the neutron field must include spatial gradients within the vessel wall. Therefore, heavy emphasis must be placed on the use of neutron transport techniques as well as on the choice of a design basis for the computations. Since a given surveillance capsule measurement, particularly one obtained early in plant life, is not necessarily representative of long-term reactor operation, a simple normalization of neutron transport calculations to dosimetry data from a given capsule may not be appropriate (1-67).  
3.2 The objectives and requirements of a reactor vessel's support structure's surveillance program are much less stringent, and at present, are limited to ...
SCOPE
1.1 This practice covers the methodology, summarized in Annex A1, to be used in the analysis and interpretation of neutron exposure data obtained from LWR pressure vessel surveillance programs and, based on the results of that analysis, establishes a formalism to be used to evaluate present and future condition of the pressure vessel and its support structures2 (1-74).3  
1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods (see Master Matrix E706) (1, 5, 13, 48, 49).2 In order to make this practice at least partially self-contained, a moderate amount of discussion is provided in areas relating to ASTM and other documents. Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, and exposure units.  
1.3 This practice is restricted to direct applications related to surveillance programs that are established in support of the operation, licensing, and regulation of LWR nuclear power plants. Procedures and data related to the analysis, interpretation, and application of test reactor results are addressed in Practice E1006, Guide E900, and Practice E1035.  
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.

SIGNIFICANCE AND USE
4.1 Regulatory Requirements—The USA Code of Federal Regulations (10CFR Part 50, Appendix H) requires the implementation of a reactor vessel materials surveillance program for all operating LWRs. Other countries have similar regulations. The purpose of the program is to (1) monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline6 resulting from exposure to neutron irradiation and the thermal environment, and (2) make use of the data obtained from surveillance programs to determine the conditions under which the vessel can be operated with adequate margins of safety throughout its service life. Practice E185, derived mechanical property data, and (r, θ, z) physics-dosimetry data (derived from the calculations and reactor cavity and surveillance capsule measurements (1) using physics-dosimetry standards) can be used together with information in Guide E900 and Refs. 4, 11-18 to provide a relation between property degradation and neutron exposure, commonly called a “trend curve.” To obtain this trend curve at all points in the pressure vessel wall requires that the selected trend curve be used together with the appropriate (r, θ, z) neutron field information derived by use of this guide to accomplish the necessary interpolations and extrapolations in space and time.  
4.2 Neutron Field Characterization—The tasks required to satisfy the second part of the objective of 4.1 are complex and are summarized in Practice E853. In doing this, it is necessary to describe the neutron field at selected (r, θ, z) points within the pressure vessel wall. The description can be either time dependent or time averaged over the reactor service period of interest. This description can best be obtained by combining neutron transport calculations with plant measurements such as reactor cavity (ex-vessel) and surveillance capsule or RPV cladding (in-vessel) measurements, benchmark irradiations of dosimeter sensor materials, and knowledge of th...
SCOPE
1.1 This guide establishes the means and frequency of monitoring the neutron exposure of the LWR reactor pressure vessel throughout its operating life.  
1.2 The physics-dosimetry relationships determined from this guide may be used to estimate reactor pressure vessel damage through the application of Practice E693 and Guide E900, using fast neutron fluence (E > 1.0 MeV and E > 0.1 MeV), displacements per atom (dpa), or damage-function-correlated exposure parameters as independent exposure variables. Supporting the application of these standards are the E853, E944, E1005, and E1018 standards, identified in 2.1.  
1.3 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.  
1.4 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.

SIGNIFICANCE AND USE
4.1 A characteristic advantage of charged-particle irradiation experiments is the precise, individual control over most of the important irradiation conditions such as dose, dose rate, temperature, and quantity of gases present. Additional attributes are the lack of induced radioactivation of specimens and, in general, a substantial compression of irradiation time, from years to hours, to achieve comparable damage as measured in displacements per atom (dpa). An important application of such experiments is the investigation of radiation effects that may occur in materials exposed to environments which do not currently exist, such as in first wall materials used in fusion reactors.  
4.2 The primary shortcoming of ion bombardments stems from the damage rate, or temperature dependences of the microstructural evolutionary processes in complex alloys, or both. It cannot be assumed that the time scale for damage evolution can be comparably compressed for all processes by increasing the displacement rate, even with a corresponding shift in irradiation temperature. In addition, the confinement of damage production to a thin layer just (often ∼1 μm) below the irradiated surface can present substantial complications. It must be emphasized, therefore, that these experiments and this practice are intended for research purposes and not for the certification or the qualification of materials.  
4.3 This practice relates to the generation of irradiation-induced changes in the microstructure of metals and alloys using charged particles. The investigation of mechanical behavior using charged particles is covered in Practice E821.
SCOPE
1.1 This practice provides guidance on performing charged-particle irradiations of metals and alloys, although many of the methods may also be applied to ceramic materials. It is generally confined to studies of microstructural and microchemical changes induced by ions of low-penetrating power that come to rest in the specimen. Density changes can be measured directly and changes in other properties can be inferred. This information can be used to estimate similar changes that would result from neutron irradiation. More generally, this information is of value in deducing the fundamental mechanisms of radiation damage for a wide range of materials and irradiation conditions.  
1.2 Where it appears, the word “simulation” should be understood to imply an approximation of the relevant neutron irradiation environment for the purpose of elucidating damage mechanisms. The degree of conformity can range from poor to nearly exact. The intent is to produce a correspondence between one or more aspects of the neutron and charged-particle irradiations such that fundamental relationships are established between irradiation or material parameters and the material response.  
1.3 The practice appears as follows:    
Section  
Apparatus  
4  
Specimen Preparation  
5 – 10  
Irradiation Techniques (including Helium Injection)  
11 – 12  
Damage Calculations  
13  
Postirradiation Examination  
14 – 16  
Reporting of Results  
17  
Correlation and Interpretation  
18 – 22  
1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
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.

ABSTRACT
This guide presents the simulation procedure which would provide advice for conducting experiments to investigate the effects of helium on the properties of irradiated metals where the technique for introducing the helium differs in someway from the actual mechanism of introduction of helium in service. Simulation techniques considered for introducing helium shall include charged particle implantation, exposure to α-emitting radioisotopes, and tritium decay techniques. Procedures for the analysis of helium content and helium distribution within the specimen are also recommended. The two other methods for introducing helium into irradiated materials namely, the enhancement of helium production in nickel-bearing alloys by spectral tailoring in mixed-spectrum fission reactors, and the isotopic tailoring in both fast and mixed-spectrum fission reactors, are not covered in this guide. Dual ion beam techniques for simultaneously implanting helium and generating displacement damage are also not included here.
SIGNIFICANCE AND USE
4.1 Helium is introduced into metals as a consequence of nuclear reactions, such as (n, α), or by the injection of helium into metals from the plasma in fusion reactors. The characterization of the effect of helium on the properties of metals using direct irradiation methods may be impractical because of the time required to perform the irradiation or the lack of a radiation facility, as in the case of the fusion reactor. Simulation techniques can accelerate the research by identifying and isolating major effects caused by the presence of helium. The word ‘simulation’ is used here in a broad sense to imply an approximation of the relevant irradiation environment. There are many complex interactions between the helium produced during irradiation and other irradiation effects, so care must be exercised to ensure that the effects being studied are a suitable approximation of the real effect. By way of illustration, details of helium introduction, especially the implantation temperature, may determine the subsequent distribution of the helium (that is, dispersed atomistically, in small clusters in bubbles, etc.).
SCOPE
1.1 This guide provides advice for conducting experiments to investigate the effects of helium on the properties of metals where the technique for introducing the helium differs in some way from the actual mechanism of introduction of helium in service. Techniques considered for introducing helium may include charged particle implantation, exposure to α-emitting radioisotopes, and tritium decay techniques. Procedures for the analysis of helium content and helium distribution within the specimen are also recommended.  
1.2 Three other methods for introducing helium into irradiated materials are not covered in this guide. They are: (1) the enhancement of helium production in nickel-bearing alloys by spectral tailoring in mixed-spectrum fission reactors, (2) a related technique that uses a thin layer of NiAl on the specimen surface to inject helium, and (3) isotopic tailoring in both fast and mixed-spectrum fission reactors. These techniques are described in Refs (1-6).2 Dual ion beam techniques (7) for simultaneously implanting helium and generating displacement damage are also not included here. This latter method is discussed in Practice E521.  
1.3 In addition to helium, hydrogen is also produced in many materials by nuclear transmutation. In some cases it appears to act synergistically with helium (8-10). The specific impact of hydrogen is not addressed in this guide.  
1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
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 regulat...

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.
Note 1: Westcott fluence rate    ...

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.

SIGNIFICANCE AND USE
3.1 Prediction of neutron radiation effects to pressure vessel steels has long been a part of the design and operation of light water reactor power plants. Both the federal regulatory agencies (see 2.3) and national standards groups (see 2.1 and 2.2) have promulgated regulations and standards to ensure safe operation of these vessels. The support structures for pressurized water reactor vessels may also be subject to similar neutron radiation effects (1, 3-6).2 The objective of this practice is to provide guidelines for determining the neutron radiation exposures experienced by individual vessel supports.  
3.2 It is known that high-energy photons can also produce displacement damage effects that may be similar to those produced by neutrons. These effects are known to be much less at the belt line of a light water reactor pressure vessel than those induced by neutrons. The same has not been proven for all locations within vessel support structures. Therefore, it may be prudent to apply coupled neutron-photon transport methods and photon-induced displacement cross sections to determine whether gamma-induced dpa exceeds the screening level of 3.0 × 10–4 used in this practice for neutron exposures. (See 1.3.)
SCOPE
1.1 This practice covers procedures for monitoring the neutron radiation exposures experienced by ferritic materials in nuclear reactor vessel support structures located in the vicinity of the active core. This practice includes guidelines for:  
1.1.1 Selecting appropriate dosimetric sensor sets and their proper installation in reactor cavities.  
1.1.2 Making appropriate neutronics calculations to predict neutron radiation exposures.  
1.2 The values stated in SI units are to be regarded as standard; units that are not SI can be found in Terminology E170 and are to be regarded as standard. Any values in parentheses are for information only.  
1.3 This practice is applicable to all pressurized water reactors whose vessel supports will experience a lifetime neutron fluence (E > 1 MeV) that exceeds 1 × 1017 neutrons/cm2 or exceeds 3.0 × 10−4 dpa (1).2 (See Terminology E170.)  
1.4 Exposure of vessel support structures by gamma radiation is not included in the scope of this practice, but see the brief discussion of this issue in 3.2.  
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. (For example, (2).)  
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.

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.

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.  
1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

SIGNIFICANCE AND USE
4.1 A pressure vessel surveillance program requires a methodology for relating radiation-induced changes in materials exposed in accelerated surveillance locations to the condition of the pressure vessel (see Practice E853). An important consideration is that the irradiation exposures be expressed in a unit that is physically related to the damage mechanisms.  
4.2 A major source of neutron radiation damage in metals is the displacement of atoms from their normal lattice sites. Hence, an appropriate damage exposure index is the number of times, on the average, that an atom has been displaced during an irradiation. This can be expressed as the total number of displaced atoms per unit volume, per unit mass, or per atom of the material. Displacements per atom is the most common way of expressing this quantity. The number of dpa associated with a particular irradiation depends on the amount of energy deposited in the material by the neutrons, and hence, depends on the neutron spectrum. (For a more extended discussion, see Practice E521.)  
4.3 No simple correspondence exists in general between dpa and a particular change in a material property. A reasonable starting point, however, for relative correlations of property changes produced in different neutron spectra is the dpa value associated with each environment. That is, the dpa values themselves provide a spectrum-sensitive index that may be a useful correlation parameter, or some function of the dpa values may affect correlation.  
4.4 Since dpa is a construct that depends on a model of the neutron interaction processes in the material lattice, as well as the cross section (probability) for each of these processes, the value of dpa would be different if improved models or cross sections are used. The calculated displacement cross section for ferritic iron, as given in this practice, is determined by the procedure given in 6.3. The currently recommended iron displacement cross section in this practice (Table 1) wa...
SCOPE
1.1 This practice describes a standard procedure for characterizing neutron irradiations of iron (and low alloy steels) in terms of the exposure index displacements per atom (dpa) for iron.  
1.2 Although the methods of this practice apply to any material for which a displacement cross section σd(E) is known (see Practice E521), this practice is written specifically for iron.  
1.3 It is assumed that the displacement cross section for iron is an adequate approximation for calculating displacements in steels that are mostly iron (95 to 100 %) in radiation fields for which secondary damage processes are not important.  
1.4 Procedures analogous to this one can be formulated for calculating dpa in charged particle irradiations. (See Practice E521.)  
1.5 The application of this practice requires knowledge of the total neutron fluence and flux spectrum. Refer to Practice E521 for determining these quantities.  
1.6 The correlation of radiation effects data is beyond the scope of this practice.  
1.7 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.  
1.8 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.

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

SIGNIFICANCE AND USE
3.1 General:  
3.1.1 The methodology recommended in this guide specifies criteria for validating computational methods and outlines procedures applicable to pressure vessel related neutronics calculations for test and power reactors. The material presented herein is useful for validating computational methodology and for performing neutronics calculations that accompany reactor vessel surveillance dosimetry measurements (see Master Matrix E706 and Practice E853). Briefly, the overall methodology involves: (1) methods-validation calculations based on at least one well-documented benchmark problem, and (2) neutronics calculations for the facility of interest. The neutronics calculations of the facility of interest and of the benchmark problem should be performed consistently, with important modeling parameters kept the same or as similar as is feasible. In particular, the same energy group structure and common broad-group microscopic cross sections should be used for both problems. Further, the benchmark problem should be characteristically similar to the facility of interest. For example, a power reactor benchmark should be utilized for power reactor calculations. Non-power reactors may have special features that may affect pressure vessel fluence and require consideration when developing a benchmark, such as beam tubes, irradiation facilities, and non-core neutron sources. The neutronics calculations involve two tasks: (1) determination of the neutron source distribution in the reactor core by utilizing diffusion theory (or transport theory) calculations in conjunction with reactor power distribution measurements, and (2) performance of a fixed fission rate neutron source (fixed-source) transport theory calculation to determine the neutron fluence rate distribution in the reactor core, through the internals and in the pressure vessel. Some neutronics modeling details for the benchmark, test reactor, or the power reactor calculation will differ; therefore, the proc...
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1.1 Need for Neutronics Calculations—An accurate calculation of the neutron fluence and fluence rate at several locations is essential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposure parameter values in the pressure vessel. Exposure parameter values may be obtained directly from calculations or indirectly from calculations that are adjusted with dosimetry measurements; Guide E944 and Practice E853 define appropriate computational procedures.  
1.2 Methodology—Neutronics calculations for application to reactor vessel surveillance encompass three essential areas: (1) validation of methods by comparison of calculations with dosimetry measurements in a benchmark experiment, (2) determination of the neutron source distribution in the reactor core, and (3) calculation of neutron fluence rate at the surveillance position and in the pressure vessel.  
1.3 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.  
1.4 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.

SIGNIFICANCE AND USE
5.1 The conventional determination of oxygen content in liquid or solid samples is a relatively difficult chemical procedure. It is slow and usually of limited sensitivity. The 14-MeV neutron activation and direct counting technique provides a rapid, highly sensitive, nondestructive procedure for oxygen determination in a wide range of matrices. This test method is independent of the chemical form of the oxygen.  
5.2 This test method can be used for quality and process control in the metals, coal, and petroleum industries, and for research purposes in a broad spectrum of applications.
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1.1 This test method covers the measurement of oxygen concentration in almost any matrix by using a 14-MeV neutron activation and direct-counting technique. Essentially, the same system may be used to determine oxygen concentrations ranging from under 10 μg/g to over 500 mg/g, depending on the sample size and available 14-MeV neutron fluence rates.  
Note 1: The range of analysis may be extended by using higher neutron fluence rates, larger samples, and higher counting efficiency detectors.  
1.2 This test method may be used on either solid or liquid samples, provided that they can be made to conform in size, shape, and macroscopic density during irradiation and counting to a standard sample of known oxygen content. Several variants of this method have been described in the technical literature. A monograph is available which provides a comprehensive description of the principles of activation analysis using a neutron generator (1).2  
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.Specific precautions are given in Section 8.  
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.

SIGNIFICANCE AND USE
4.1 This guide deals with the difficult problem of benchmarking neutron transport calculations carried out to determine fluences for plant specific reactor geometries. The calculations are necessary for fluence determination in locations important for material radiation damage estimation and which are not accessible to measurement. Typically, the most important application of such calculations is the estimation of fluence within the reactor vessel of operating light water reactors (LWR) to provide accurate estimates of the irradiation embrittlement of the base and weld metal in the vessel. The benchmark procedure must not only prove that calculations give reasonable results but that their uncertainties are propagated with due regard to the sensitivities of the different input parameters used in the transport calculations. Benchmarking is achieved by building up data bases of benchmark experiments that have different influences on uncertainty propagation. For example, in simple vessel wall mockups where measurements are made within a simulated reactor vessel wall, the integral effect of uncertainties in iron cross sections (absorption and elastic and inelastic scattering) are dominant and have been bounded by the agreement between calculation and measurement. For more complicated integral benchmarks, other factors such as: uncertainties in the distribution of fission sources, geometry, the energy-dependent cross sections, and the angular scattering distribution for elemental components of major materials in the neutron field (such as water and iron) may all be important uncertainty contributors. This guide describes general procedures for using neutron fields with known characteristics to corroborate the calculational methodology and nuclear data used to derive neutron field information from measurements of neutron sensor response.  
4.2 The bases for benchmark field referencing are usually irradiations performed in standard neutron fields with well-known energy spe...
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1.1 This guide covers general approaches for benchmarking neutron transport calculations for pressure vessel surveillance programs in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor pressure vessel surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with a higher degree of confidence.  
1.2 Although this guide and the companion guide, Guide E2005, are focused on power reactors, the principle of this guide is also applicable to non-power light water reactor pressure vessel surveillance programs.  
1.3 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.  
1.4 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.

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.
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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.

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

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

SIGNIFICANCE AND USE
3.1 This guide describes approaches for using neutron fields with well known characteristics to perform calibrations of neutron sensors, to intercompare different methods of dosimetry, and to corroborate procedures used to derive neutron field information from measurements of neutron sensor response.  
3.2 This guide discusses only selected standard and reference neutron fields which are appropriate for benchmark testing of light-water reactor dosimetry. The Standard Fields considered here include neutron source environments that closely approximate: a) the unscattered neutron spectra from  252Cf spontaneous fission; and b) the  235U thermal neutron induced fission. These standard fields were chosen for their spectral similarity to the high energy region (E > 2 MeV) of reactor spectra. The various categories of benchmark fields are defined in Terminology E170.  
3.3 There are other well known neutron fields that have been designed to mockup special environments, such as pressure vessel mockups in which it is possible to make dosimetry measurements inside of the steel volume of the “vessel.” When such mockups are suitably characterized, they are also referred to as benchmark fields. A variety of these engineering benchmark fields have been developed, or pressed into service, to improve the accuracy of neutron dosimetry measurement techniques. These special benchmark experiments are discussed in Guide E2006, and in Refs (1)4 and (2).
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1.1 This guide covers facilities and procedures for benchmarking neutron measurements and calculations. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to calibrate integral neutron sensors; the use of certified-neutron-fluence standards to calibrate radiometric counting equipment or to determine interlaboratory measurement consistency; development of special benchmark fields to test neutron transport calculations; use of well-known fission spectra to benchmark spectrum-averaged cross sections; and the use of benchmarked data and calculations to determine the uncertainties in derived neutron dosimetry results.  
1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.3 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.  
1.4 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.

SIGNIFICANCE AND USE
3.1 The mechanical properties of steels and other metals are altered by exposure to neutron radiation. These property changes are assumed to be a function of chemical composition, metallurgical condition, temperature, fluence (perhaps also fluence rate), and neutron spectrum. The influence of these variables is not completely understood. The functional dependency between property changes and neutron radiation is summarized in the form of damage exposure parameters that are weighted integrals over the neutron fluence spectrum.  
3.2 The evaluation of neutron radiation effects on pressure vessel steels and the determination of safety limits requires the knowledge of uncertainties in the prediction of radiation exposure parameters (for example, dpa (Practice E693), neutron fluence greater than 1.0 MeV, neutron fluence greater than 0.1 MeV, thermal neutron fluence, etc.). This practice describes recommended procedures and data for determining these exposure parameters (and the associated uncertainties) for test reactor experiments.  
3.3 The nuclear industry draws much of its information from databases that come from test reactor experiments. Therefore, it is essential that reliable databases are obtained from test reactors to assess safety issues in Light Water Reactor (LWR) nuclear power plants.
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1.1 This practice covers the methodology summarized in Annex A1 to be used in the analysis and interpretation of physics-dosimetry results from test reactors.  
1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods.  
1.3 Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, exposure units, and neutron spectrum adjustment methods.  
1.4 This practice is directed towards the development and application of physics-dosimetry-metallurgical data obtained from test reactor irradiation experiments that are performed in support of the operation, licensing, and regulation of LWR nuclear power plants. It specifically addresses the physics-dosimetry aspects of the problem. Procedures related to the analysis, interpretation, and application of both test and power reactor physics-dosimetry-metallurgy results are addressed in Practices E185, E853, and E1035, Guides E900, E2005, E2006 and Test Method E646. See also E706.  
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.

SIGNIFICANCE AND USE
4.1 Integral Mode Dosimetry—As shown in 3.2, two different integral relationships can be established using proton-recoil emulsion data. These two integral reactions can be obtained with roughly an order of magnitude reduction in scanning effort. Consequently, this integral mode is an important complementary alternative to the customary differential mode of NRE spectrometry. The integral mode can be applied over extended spatial regions, for example, perhaps up to as many as ten in-situ locations can be covered for the same scanning effort that is expended for a single differential measurement. Hence the integral mode is especially advantageous for dosimetry applications which require extensive spatial mapping, such as exist in Light Water Reactor-Pressure Vessel (LWR-PV) benchmark fields (see Test Method E1005). In low power benchmark fields, NRE can be used as integral dosimeters in a manner similar to RM, solid state track recorders (SSTR) and helium accumulation monitors (HAFM) neutron dosimeters (see Test Methods E854 and E910). In addition to spatial mapping advantages of these other dosimetry methods, NRE offer fine spatial resolution and can therefore be used in-situ for fine structure measurements. In integral mode scanning, both absolute reaction rates, that is I(ET) and J(Emin), are determined simultaneously. Separate software codes need to be used to permit operation of a computer based interactive system in the integral mode (see Section 9). It should be noted that the integrals I(ET) and J(Emin) possess different units, namely proton-recoil tracks/MeV per hydrogen atom and proton-recoil tracks per hydrogen atom, respectively.  
4.2 Applicability for Spectral Adjustment Codes—In the integral mode, NRE provide absolute integral reaction rates that can be used in neutron spectrum least squares adjustment codes (see Guide E944). In the past, such adjustment codes could not utilize NRE integral reaction rates because of the non-existence of NRE data. NRE in...
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1.1 Nuclear Research Emulsions (NRE) have a long and illustrious history of applications in the physical sciences, earth sciences and biological sciences (1, 2)2. In the physical sciences, NRE experiments have led to many fundamental discoveries in such diverse disciplines as nuclear physics, cosmic ray physics and high energy physics. In the applied physical sciences, NRE have been used in neutron physics experiments in both fission and fusion reactor environments (3-6). Numerous NRE neutron experiments can be found in other applied disciplines, such as nuclear engineering, environmental monitoring and health physics. Given the breadth of NRE applications, there exist many textbooks and handbooks that provide considerable detail on the techniques used in the NRE method (1-4, 6). As a consequence, this practice will be restricted to the application of the NRE method for neutron measurements in reactor physics and nuclear engineering with particular emphasis on neutron dosimetry in benchmark fields (see Matrix E706).  
1.2 NRE are passive detectors and provide time integrated reaction rates. As a consequence, NRE provide fluence measurements without the need for time-dependent corrections, such as arise with radiometric (RM) dosimeters (see Test Method E1005). NRE provide permanent records, so that optical microscopy observations can be carried out any time after exposure. If necessary, NRE measurements can be repeated at any time to examine questionable data or to obtain refined results.  
1.3 Since NRE measurements are conducted with optical microscopes, high spatial resolution is afforded for fine structure experiments. The attribute of high spatial resolution can also be used to determine information on the angular anisotropy of the in-situ neutron field (4, 5, 7). It is not possible for active detectors to provide such data because of in-situ perturbations and finite-size effects (see Section 11).  
1.4 The existe...

SIGNIFICANCE AND USE
4.1 The ENDF/B library in the United States and similar libraries elsewhere, such as JEFF (23), JENDL (21), and BROND (22), provide a compilation of neutron cross section and other nuclear data for use by the nuclear community. The availability of these excellent evaluations makes possible standardized usage, thereby allowing easy referencing and intercomparisons of calculations. However, as the first ENDF/B files were developed it became apparent that they were not adequate for all applications. This need resulted in the development of the specialized ENDF/B Dosimetry File (17, 25), consisting of activation cross sections important for dosimetry applications. This file was made available worldwide. Later, other “Special Purpose” files were introduced (26). In the ENDF/B-VI compilation (27), dosimetry files no longer appeared as separate evaluation files. The ENDF/B-VII.0 compilation (28) removed most of the reaction-specific covariance files used by the dosimetry community. It kept the covariance files for the “standard cross sections” in a special sub-library, but the covariance data in this sub-library are only provided over the energy range in which each reaction is considered to be a “standard”, and does not include the full energy range required for LWR PVS dosimetry applications. Later updates to the ENDF/B releases added covariance files for some reaction channels but these covariance files were often based solely on calculations and were not representative of the methodology used to derive the underlying ENDF/B cross section. In response to the need for a dosimetry-specific library, the International Atomic Energy Agency convened a Coordinated Research Project (CRP) that drew upon the set of international experts to provide a recommended set of dosimetry cross sections and to compile a set of validation evidence that supported the use of this recommended dataset. This file, the International Reactor Dosimetry and Fusion File (IRDFF) (19, 20), draws upon oth...
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1.1 This guide covers the establishment and use of an ASTM evaluated nuclear data cross section and uncertainty file for analysis of single or multiple sensor measurements in neutron fields related to light water reactor LWR-Pressure Vessel Surveillance (PVS). These fields include in- and ex-vessel surveillance positions in operating power reactors, benchmark fields, and reactor test regions.  
1.2 Requirements for establishment of ASTM-recommended cross section files address data format, evaluation requirements, validation in benchmark fields, evaluation of error estimates (covariance file), and documentation. A further requirement for components of the ASTM-recommended cross section file is their internal consistency when combined with sensor measurements and used to determine a neutron spectrum.  
1.3 Specifications for use include energy region of applicability, data processing requirements, and application of uncertainties.  
1.4 This guide is directly related to and should be used primarily in conjunction with Guides E482 and E944, and Practices E560, E185, and E693.  
1.5 The ASTM cross section and uncertainty file represents a generally available data set for use in sensor set analysis. However, the availability of this data set does not preclude the use of other validated data, either proprietary or nonproprietary. When alternate cross section files that deviate from the requirements laid out in this standard are used, the deviations should be noted to the customer of the dosimetry application.  
1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.7 This international standard was developed in accordance with internationally recognized principles on standar...

SIGNIFICANCE AND USE
4.1 The SSTR method provides for the measurement of absolute-fission density per unit mass. Absolute-neutron fluence can then be inferred from these SSTR-based absolute fission rate observations if an appropriate neutron spectrum average fission cross section is known. This method is highly discriminatory against other components of the in-core radiation field. Gamma rays, beta rays, and other lightly ionizing particles do not produce observable tracks in appropriate LWR SSTR candidate materials. However, photofission can contribute to the observed fission track density and should therefore be accounted for when nonnegligible. For a more detailed discussion of photofission effects, see 14.4.  
4.2 In this test method, SSTRs are placed in surface contact with fissionable deposits and record neutron-induced fission fragments. By variation of the surface mass density (μg/cm 2) of the fissionable deposit as well as employing the allowable range of track densities (from roughly 1 event/cm2 up to 105 events/cm2 for manual scanning), a range of total fluence sensitivity covering at least 16 orders of magnitude is possible, from roughly 102 n/cm 2 up to 5 × 10 18 n/cm2. The allowable range of fission track densities is broader than the track density range for high accuracy manual scanning work with optical microscopy cited in 1.2. In particular, automated and semi-automated methods exist that broaden the customary track density range available with manual optical microscopy. In this broader track density region, effects of reduced counting statistics at very low track densities and track pile-up corrections at very high track densities can present inherent limitations for work of high accuracy. Automated scanning techniques are described in Section 11.  
4.3 For dosimetry applications, different energy regions of the neutron spectrum can be selectively emphasized by changing the nuclide used for the fission deposit.  
4.4 It is possible to use SSTRs directly for neutron ...
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1.1 This test method describes the use of solid-state track recorders (SSTRs) for neutron dosimetry in light-water reactor (LWR) applications. These applications extend from low neutron fluence to high neutron fluence, including high power pressure vessel surveillance and test reactor irradiations as well as low power benchmark field measurement. (1)2 Special attention is given to the use of state-of-the-art manual and automated track counting methods to attain high absolute accuracies. In-situ dosimetry in actual high fluence-high temperature LWR applications is emphasized.  
1.2 This test method includes SSTR analysis by both manual and automated methods. To attain a desired accuracy, the track scanning method selected places limits on the allowable track density. Typically, good results are obtained in the range of 5 to 800 000 tracks/cm2 and accurate results at higher track densities have been demonstrated for some cases.  (2) Track density and other factors place limits on the applicability of the SSTR method at high fluences. Special care must be exerted when measuring neutron fluences (E>1MeV) above 1016 n/cm2 (3) .  
1.3 Low fluence and high fluence limitations exist. These limitations are discussed in detail in Sections 13 and 14 and in Refs (3-5).  
1.4 SSTR observations provide time-integrated reaction rates. Therefore, SSTRs are truly passive-fluence detectors. They provide permanent records of dosimetry experiments without the need for time-dependent corrections, such as decay factors that arise with radiometric monitors.  
1.5 Since SSTRs provide a spatial record of the time-integrated reaction rate at a microscopic level, they can be used for “fine-structure” measurements. For example, spatial distributions of isotopic fission rates can be obtained at very high resolution with SSTRs.  
1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the ...

SIGNIFICANCE AND USE
5.1 Refer to Guide E844 for the selection, irradiation, and quality control of neutron dosimeters.  
5.2 Refer to Practice E261 for a general discussion of the measurement of neutron fluence rate and fluence. The neutron spectrum must be known in order to measure neutron fluence rates with a single detector. Also it is noted that cross sections are continuously being reevaluated. The latest recommended cross sections and details on how they can be obtained are discussed in Guide E1018.  
5.3 The reaction rate of a detector nuclide of known cross section, when combined with information about the neutron spectrum, permits the determination of the magnitude of the fluence rate impinging on the detector. Furthermore, if results from other detectors are available, the neutron spectrum can be defined more accurately. The techniques for fluence rate and fluence determinations are explained in Practice E261.  
5.4 140Ba is a radioactive nuclide formed as a result of fission. Although it is formed in fission of any heavy atom, the relative yield will differ. Table 1 gives recommended cumulative fission yields for 140Ba production and direct (independent) fission yields for the daughter product  140La. The independent fission yields for  140La are relatively low compared to the  140Ba cumulative fission yield and will not significantly affect the accuracy of the nondestructive procedure and need not be considered. (A) Thermal = 0.0253 eV19.(B) Fast = 0.4 MeV.(C) From JEF-3.1.1 (Ref (1, 2)), except as noted. Uncertainties in percent of given value.(D) From ENDF/B-VIII.0 (Ref (3)). Not available in JEF-3.1.1. Uncertainties in percent of given value.  
5.5 The half-life of  140Ba is 12.752 days. Its daughter  140La has a half-life of 1.6781 days (4).3 The comparatively long half-life of 140Ba allows the counting to be delayed several weeks after irradiation in a high-neutron field. However, to achieve maximum sensitivity the daughter product  140La should be counted five to...
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1.1 This test method describes two procedures for the measurement of reaction rates by determining the amount of the fission product  140Ba produced by the non-threshold reactions  235U(n,f),  241Am(n,f), and  239Pu(n,f), and by the threshold reactions  238U(n,f),  237Np(n,f), and  232Th(n,f).  
1.2 These reactions produce many fission products, among which is 140Ba, having a half-life of 12.752 days.  140Ba emits gamma rays of several energies; however, these are not easily detected in the presence of other fission products. Competing activity from other fission products requires that a chemical separation be employed or that the  140Ba activity be determined indirectly by counting its daughter product 140La. This test method describes both procedure (a), the nondestructive determination of  140Ba by the direct counting of 140La several days after irradiation, and procedure (b), the chemical separation of  140Ba and the subsequent counting of  140Ba or its daughter  140 La.  
1.3 With suitable techniques, fission neutron fluence rates can be measured in the range from 107  n (neutrons) · cm−2 · s−1  to approximately 1015  n · cm−2  · s−1.  
1.4 The measurement of time-integrated reaction rates with fission dosimeters by  140Ba analysis is limited by the half-life of  140Ba to irradiation times up to about six weeks.  
1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the De...

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

SIGNIFICANCE AND USE
3.1 Adjustment methods provide a means for combining the results of neutron transport calculations with neutron dosimetry measurements (see Test Method E1005 and NUREG/CR-5049) in order to obtain optimal estimates for neutron damage exposure parameters with assigned uncertainties. The inclusion of measurements reduces the uncertainties for these parameter values and provides a test for the consistency between measurements and calculations and between different measurements (see 3.3.3). This does not, however, imply that the standards for measurements and calculations of the input data can be lowered; the results of any adjustment procedure can be only as reliable as are the input data.  
3.2 Input Data and Definitions:  
3.2.1 The symbols introduced in this section will be used throughout the guide.  
3.2.2 Dosimetry measurements are given as a set of reaction rates (or equivalent) denoted by the following symbols:
These data are, at present, obtained primarily from radiometric dosimeters, but other types of sensors may be included (see 4.1).  
3.2.3 The neutron spectrum (see Terminology E170) at the dosimeter location, fluence or fluence rate Φ(E) as a function of neutron energy E , is obtained by appropriate neutronics calculations (neutron transport using the methods of discrete ordinates or Monte Carlo, see Guide E482). The results of the calculation are customarily given in the form of multigroup fluences or fluence rates.
where:
Ej and Ej+1  are the lower and upper bounds for the j-th energy group, respectively, and k is the total number of groups.  
3.2.4 The reaction cross sections of the dosimetry sensors are obtained from an evaluated cross section file. The cross section for the i-th reaction as a function of energy E will be denoted by the following:
Used in connection with the group fluences, Eq 2, are the calculated group-averaged cross sections σij. These values are defined through the following equation:
...
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1.1 This guide covers the analysis and interpretation of the physics dosimetry for Light Water Reactor (LWR) surveillance programs. The main purpose is the application of adjustment methods to determine best estimates of neutron damage exposure parameters and their uncertainties.  
1.2 This guide is also applicable to irradiation damage studies in research reactors.  
1.3 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.  
1.4 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.

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

SIGNIFICANCE AND USE
5.1 Refer to Practice E261 for a general discussion of the determination of fast-neutron fluence rate with fission detectors.  
5.2 237Np is available as metal foil, wire, or oxide powder. For further information, see Guide E844. It is usually encapsulated in a suitable container to prevent loss of, and contamination by, the  237Np and its fission products.4  
5.3 One or more fission products can be assayed. Pertinent data for relevant fission products are given in Table 15 and Table 2. (A) The lightface numbers in parentheses are the magnitude of plus or minus uncertainties in the last digit(s) listed.(B) With 137mBa (2.552 min) in equilibrium.(C) Probability of daughter 140La decay.(D) With 140La (1.67850 d) in transient equilibrium.(E) Primary reference for half-life, gamma energy, and gamma emission probability is Ref (1) when data is available. Note this reference is to the BIPM data that was recommended at the time of the recommended fission yields were set, that is, as of 2009, and not to the latest Vol 8 data that was published in 2016.  (A) The JEFF-3.1/3.1.1 radioactive decay data and fission yields sub-libraries, JEFF Report 20, OECD 2009, Nuclear Energy Agency (2).(B) All yield data given as a %; RC represents a cumulative yield; RI represents an independent yield.(C) The neutron energy represents a generic “fast neutron” spectrum and has been characterized in the JEFF 3.1.1 fission yield library as having an average neutron energy of 0.4 MeV.  
5.3.1 137Cs-137mBa is chosen frequently for long irradiations. Radioactive products 134Cs and 136Cs may be present, which can interfere with the counting of the 0.661657 MeV 137Cs-137mBa gamma ray (see Test Methods E320).  
5.3.2 140Ba-140La is chosen frequently for short irradiations (see Test Method E393).  
5.3.3 95Zr can be counted directly, following chemical separation, or with its daughter 95Nb, using a high-resolution gamma detector system.  
5.3.4 144Ce is a high-yield fission product applicable t...
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1.1 This test method covers procedures for measuring reaction rates by assaying a fission product (F.P.) from the fission reaction 237Np(n,f)F.P.  
1.2 The reaction is useful for measuring neutrons with energies from approximately 0.7 to 6 MeV and for irradiation times up to 90 years, provided that the analysis methods described in Practice E261 are followed. If dosimeters are analyzed after irradiation periods longer than 90 years, the information inferred about the fluence during irradiation periods more than 90 years before the end of the irradiation should not be relied upon without supporting data from dosimeters withdrawn earlier.  
1.3 Equivalent fission neutron fluence rates as defined in Practice E261 can be determined.  
1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice E261.  
1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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

SIGNIFICANCE AND USE
4.1 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 in nuclear reactors. Each dosimeter is sensitive to a specific energy range, and, if desired, increased accuracy in a fluence-rate spectrum can be achieved by the use of several dosimeters each covering specific neutron energy ranges.  
4.2 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 fluence-rate 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.
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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 SI units are to be regarded as standard. Values in parentheses are for information only.  
1.3 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.  
1.4 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.

SIGNIFICANCE AND USE
5.1 Refer to Practice E261 for a general discussion of the determination of decay rates, reaction rates, and neutron fluence rates with threshold detectors (1-29).3 Refer to Practice E1006, Practice E185 and Guide E1018 for the use and application of results obtained by this test method.(30-32)  
5.2 The half-life of  93mNb is 16.1 (2)4 years5(34) and has a K X-ray emission probability of 0.11442 ± 3.356 % per decay (35). The Kα and Kβ X-rays of niobium are at 16.521–16.615 and 18.607–18.9852 keV, respectively (35). The recommended  93Nb(n,n′)93mNb cross section comes from the International Reactor Dosimetry and Fusion File (IRDFF version 1.05, cross section compendium (36), and is shown in Fig. 1. This nuclear data evaluation is part of the Russian Reactor Dosimetry File (RRDF), cross section evaluations (37). The nuclear decay data referenced here are not taken from the latest dosimetry recommended database (33) but are selected to be consistent with the nuclear data used in the recommended IRDFF evaluation.
FIG. 1 RRDF/IRDFF-1.05 Cross Section Versus Energy for the  93Nb(n,n′) 93mNb Reaction  
5.3 Chemical dissolution of the irradiated niobium to produce very low mass-per-unit area sources is an effective way to obtain consistent results. The direct counting of foils or wires can produce satisfactory results provided appropriate methods and interpretations are employed. It is possible to use liquid scintillation methods to measure the niobium activity provided the radioactive material can be kept uniformly in solution and appropriate corrections can be made for interfering activities.  
5.4 The measured reaction rates can be used to correlate neutron exposures, provide comparison with calculated reaction rates, and determine neutron fluences. Reaction rates can be determined with greater accuracy than fluence rates because of the current uncertainty in the cross section versus energy shape.  
5.5 The  93Nb(n,n′)93mNb reaction has the desirable properties o...
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1.1 This test method describes procedures for measuring reaction rates by the activation reaction  93Nb(n,n′) 93mNb.  
1.2 This activation reaction is useful for monitoring neutrons with energies above approximately 0.5 MeV and for irradiation times up to about 48 years (three half-lives), provided that the analysis methods described in Practice E261 are followed.  
1.3 With suitable techniques, fast-neutron reaction rates for neutrons with energy distribution similar to fission neutrons can be determined in fast-neutron fluences above about 1016 cm−2. In the presence of high thermal-neutron fluence rates (>1012cm−2·s−1), the transmutation of  93mNb due to neutron capture should be investigated. In the presence of high-energy neutron spectra such as are associated with fusion and spallation sources, the transmutation of 93mNb by reactions such as (n,2n) may occur and should be investigated.  
1.4 Procedures for other fast-neutron monitors are referenced in Practice E261.  
1.5 Fast-neutron fluence rates can be determined from the reaction rates provided that the appropriate cross section information is available to meet the accuracy requirements.  
1.6 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.7 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.  
1.8 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.

SIGNIFICANCE AND USE
5.1 The HAFM test method is one of several available passive neutron dosimetry techniques (see, for example, Test Methods E854 and E1005). This test method can be used in combination with other dosimetry methods, or, if sufficient data are available from different HAFM sensor materials, as an alternative dosimetry test method. The HAFM method yields a direct measurement of total helium production in an irradiated sample. Absolute neutron fluence can then be inferred from this, assuming the appropriate spectrum integrated total helium production cross section. Alternatively, a calibration of the composite neutron detection efficiency for the HAFM method may be obtained by exposure in a benchmark neutron field where the fluence and spectrum averaged cross section are both known (see Guide E2005).  
5.2 HAFMs have the advantage of producing an end product, helium, which is stable, making the HAFM method very attractive for both short-term and long-term fluence measurements without requiring time-dependent corrections for decay. HAFMs are therefore ideal passive, time-integrating fluence monitors. Additionally, the burnout of the daughter product, helium, is negligible.  
5.2.1 Many of the HAFM materials can be irradiated in the form of unencapsulated wire segments (see 1.1.2). These segments can easily be fabricated by cutting from a standard inventoried material lot. The advantage is that encapsulation, with its associated costs, is not necessary. In several cases, unencapsulated wires such as Fe, Ni, Al/Co, and Cu, which are already included in the standard radiometric (RM) dosimetry sets (Table 1) can be used for both radiometric and helium accumulation dosimetry. After radiometric counting, the samples are later vaporized for helium measurement. (A) Evaluated  235U fission neutron spectrum averaged helium production cross section and energy range in which 90 % of the reactions occur. All values are obtained from ENDF/B-V Gas Production Dosimetry File data. Bracke...
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1.1 This test method describes the concept and use of helium accumulation for neutron fluence dosimetry for reactor vessel surveillance. Although this test method is directed toward applications in vessel surveillance, the concepts and techniques are equally applicable to the general field of neutron dosimetry. The various applications of this test method for reactor vessel surveillance are as follows:  
1.1.1 Helium accumulation fluence monitor (HAFM) capsules,  
1.1.2 Unencapsulated, or cadmium or gadolinium covered, radiometric monitors (RM) and HAFM wires for helium analysis,  
1.1.3 Charpy test block samples for helium accumulation, and  
1.1.4 Reactor vessel (RV) wall samples for helium accumulation.  
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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.3 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.

SIGNIFICANCE AND USE
4.1 This test method can be extended to use any material that has the necessary nuclear and activation properties that suit the experimenter's particular situation. No attempt has been made to fully describe the myriad problems of counting techniques, neutron-fluence depression, and thick-foil self-shielding. It is assumed that the experimenter will refer to existing literature on these subjects. This test method does offer a referee technique (the standard gold foil) to aid the experimenter when they are in doubt of their ability to perform the radiometric technique with sufficient accuracy.  
4.2 The standard comparison technique uses a set of foils that are as nearly identical as possible in shape and mass. The foils are fabricated from any material that activates by an (n, γ) reaction, preferably having a cross section approximately inversely proportional to neutron speed in the thermal energy range. Some of the foils are irradiated in a known neutron field (at NIST) or other standards laboratory). The foils are counted in a fixed geometry on a stable radiation-detecting instrument. The neutron-induced reaction rate of the foils is computed from the counting data, and the ratio of the known neutron fluence rate to the computed reaction rate is determined. For any given foil, neutron energy spectrum, and counting set-up, this ratio is a constant. Other foils from the identical set can now be exposed to an unknown neutron field. The magnitude of the fluence rate in the unknown field can be obtained by comparing the reaction rates as determined from the counting data from the unknown and reference field, with proper corrections to account for spectral differences between the two fields (see Section 5). One important feature of this technique is that it eliminates the need for knowing the detector efficiency.  
4.3 This test method follows the Stoughton and Halperin convention for reporting thermal neutron fluence. Other conventions are the Wescott convention (fol...
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1.1 The purpose of this test method is to define a general procedure for determining an unknown thermal-neutron fluence rate by neutron activation techniques. It is not practicable to describe completely a technique applicable to the large number of experimental situations that require the measurement of a thermal-neutron fluence rate. Therefore, this method is presented so that the user may adapt to their particular situation the fundamental procedures of the following techniques.  
1.1.1 Radiometric counting technique using pure cobalt, pure gold, pure indium, cobalt-aluminum, alloy, gold-aluminum alloy, or indium-aluminum alloy.  
1.1.2 Standard comparison technique using pure gold, or gold-aluminum alloy, and  
1.1.3 Secondary standard comparison techniques using pure indium, indium-aluminum alloy, pure dysprosium, or dysprosium-aluminum alloy.  
1.2 The techniques presented are limited to measurements at room temperatures. However, special problems when making thermal-neutron fluence rate measurements in high-temperature environments are discussed in 9.2. For those circumstances where the use of cadmium as a thermal shield is undesirable because of potential spectrum perturbations or of temperatures above the melting point of cadmium, the method described in Test Method E481 can be used in some cases. Alternatively, gadolinium filters may be used instead of cadmium. For high temperature applications in which aluminum alloys are unsuitable, other alloys such as cobalt-nickel or cobalt-vanadium have been used.  
1.3 This test method may be used to determine the equivalent 2200 m/s fluence rate. The accurate determination of the actual thermal neutron fluence rate requires knowledge of the neutron temperature, and determination of the neutron temperature is not within the scope of the standard.  
1.4 The techniques presented are suitable only for neutron fields having a significant thermal neutron component, in w...

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

SIGNIFICANCE AND USE
4.1 Regulatory Requirements—The USA Code of Federal Regulations (10CFR Part 50, Appendix H) requires the implementation of a reactor vessel materials surveillance program for all operating LWRs. Other countries have similar regulations. The purpose of the program is to (1) monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline6 resulting from exposure to neutron irradiation and the thermal environment, and (2) make use of the data obtained from surveillance programs to determine the conditions under which the vessel can be operated with adequate margins of safety throughout its service life. Practice E185, derived mechanical property data, and (r, θ, z) physics-dosimetry data (derived from the calculations and reactor cavity and surveillance capsule measurements (1) using physics-dosimetry standards) can be used together with information in Guide E900 and Refs. 4, 11-18 to provide a relation between property degradation and neutron exposure, commonly called a “trend curve.” To obtain this trend curve at all points in the pressure vessel wall requires that the selected trend curve be used together with the appropriate (r, θ, z) neutron field information derived by use of this guide to accomplish the necessary interpolations and extrapolations in space and time.  
4.2 Neutron Field Characterization—The tasks required to satisfy the second part of the objective of 4.1 are complex and are summarized in Practice E853. In doing this, it is necessary to describe the neutron field at selected (r, θ, z) points within the pressure vessel wall. The description can be either time dependent or time averaged over the reactor service period of interest. This description can best be obtained by combining neutron transport calculations with plant measurements such as reactor cavity (ex-vessel) and surveillance capsule or RPV cladding (in-vessel) measurements, benchmark irradiations of dosimeter sensor materials, and knowledge of th...
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1.1 This guide establishes the means and frequency of monitoring the neutron exposure of the LWR reactor pressure vessel throughout its operating life.  
1.2 The physics-dosimetry relationships determined from this guide may be used to estimate reactor pressure vessel damage through the application of Practice E693 and Guide E900, using fast neutron fluence (E > 1.0 MeV and E > 0.1 MeV), displacements per atom – dpa, or damage-function-correlated exposure parameters as independent exposure variables. Supporting the application of these standards are the E853, E944, E1005, and E1018 standards, identified in 2.1.  
1.3 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.  
1.4 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.

SIGNIFICANCE AND USE
4.1 The ENDF/B library in the United States and similar libraries elsewhere, such as JEFF (23), JENDL (21), and BROND (22), provide a compilation of neutron cross section and other nuclear data for use by the nuclear community. The availability of these excellent evaluations makes possible standardized usage, thereby allowing easy referencing and intercomparisons of calculations. However, as the first ENDF/B files were developed it became apparent that they were not adequate for all applications. This need resulted in the development of the specialized ENDF/B Dosimetry File (17, 25), consisting of activation cross sections important for dosimetry applications. This file was made available worldwide. Later, other “Special Purpose” files were introduced (26). In the ENDF/B-VI compilation (27), dosimetry files no longer appeared as separate evaluation files. The ENDF/B-VII.0 compilation (28) removed most of the reaction-specific covariance files used by the dosimetry community. It kept the covariance files for the “standard cross sections” in a special sub-library, but the covariance data in this sub-library are only provided over the energy range in which each reaction is considered to be a “standard”, and does not include the full energy range required for LWR PVS dosimetry applications. Later updates to the ENDF/B releases added covariance files for some reaction channels but these covariance files were often based solely on calculations and were not representative of the methodology used to derive the underlying ENDF/B cross section. In response to the need for a dosimetry-specific library, the International Atomic Energy Agency convened a Coordinated Research Project (CRP) that drew upon the set of international experts to provide a recommended set of dosimetry cross sections and to compile a set of validation evidence that supported the use of this recommended dataset. This file, the International Reactor Dosimetry and Fusion File (IRDFF) (19, 20), draws upon oth...
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1.1 This guide covers the establishment and use of an ASTM evaluated nuclear data cross section and uncertainty file for analysis of single or multiple sensor measurements in neutron fields related to light water reactor LWR-Pressure Vessel Surveillance (PVS). These fields include in- and ex-vessel surveillance positions in operating power reactors, benchmark fields, and reactor test regions.  
1.2 Requirements for establishment of ASTM-recommended cross section files address data format, evaluation requirements, validation in benchmark fields, evaluation of error estimates (covariance file), and documentation. A further requirement for components of the ASTM-recommended cross section file is their internal consistency when combined with sensor measurements and used to determine a neutron spectrum.  
1.3 Specifications for use include energy region of applicability, data processing requirements, and application of uncertainties.  
1.4 This guide is directly related to and should be used primarily in conjunction with Guides E482 and E944, and Practices E560, E185, and E693.  
1.5 The ASTM cross section and uncertainty file represents a generally available data set for use in sensor set analysis. However, the availability of this data set does not preclude the use of other validated data, either proprietary or nonproprietary. When alternate cross section files that deviate from the requirements laid out in this standard are used, the deviations should be noted to the customer of the dosimetry application.  
1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.7 This international standard was developed in accordance with internationally recognized principles on standar...

SIGNIFICANCE AND USE
3.1 The objectives of a reactor vessel surveillance program are twofold. The first requirement of the program is to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region resulting from exposure to neutron irradiation and the thermal environment. The second requirement is to make use of the data obtained from the surveillance program to determine the conditions under which the vessel can be operated throughout its service life.  
3.1.1 To satisfy the first requirement of 3.1, the tasks to be carried out are straightforward. Each of the irradiation capsules that comprise the surveillance program may be treated as a separate experiment. The goal is to define and carry to completion a dosimetry program that will, a posteriori, describe the neutron field to which the materials test specimens were exposed. The resultant information will then become part of a data base applicable in a stricter sense to the specific plant from which the capsule was removed, but also in a broader sense to the industry as a whole.  
3.1.2 To satisfy the second requirement of 3.1, the tasks to be carried out are somewhat complex. The objective is to describe accurately the neutron field to which the pressure vessel itself will be exposed over its service life. This description of the neutron field must include spatial gradients within the vessel wall. Therefore, heavy emphasis must be placed on the use of neutron transport techniques as well as on the choice of a design basis for the computations. Since a given surveillance capsule measurement, particularly one obtained early in plant life, is not necessarily representative of long-term reactor operation, a simple normalization of neutron transport calculations to dosimetry data from a given capsule may not be appropriate (1-67).2  
3.2 The objectives and requirements of a reactor vessel's support structure's surveillance program are much less stringent, and at present, are limited ...
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1.1 This practice covers the methodology, summarized in Annex A1, to be used in the analysis and interpretation of neutron exposure data obtained from LWR pressure vessel surveillance programs; and, based on the results of that analysis, establishes a formalism to be used to evaluate present and future condition of the pressure vessel and its support structures2 (1-74).3  
1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods (see Master Matrix E706) (1, 5, 13, 48, 49).2 In order to make this practice at least partially self-contained, a moderate amount of discussion is provided in areas relating to ASTM and other documents. Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, and exposure units.  
1.3 This practice is restricted to direct applications related to surveillance programs that are established in support of the operation, licensing, and regulation of LWR nuclear power plants. Procedures and data related to the analysis, interpretation, and application of test reactor results are addressed in Practice E1006, Guide E900, and Practice E1035.  
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.

SIGNIFICANCE AND USE
3.1 Prediction of neutron radiation effects to pressure vessel steels has long been a part of the design and operation of light water reactor power plants. Both the federal regulatory agencies (see 2.3) and national standards groups (see 2.1 and 2.2) have promulgated regulations and standards to ensure safe operation of these vessels. The support structures for pressurized water reactor vessels may also be subject to similar neutron radiation effects (1, 3-6).6 The objective of this practice is to provide guidelines for determining the neutron radiation exposures experienced by individual vessel supports.  
3.2 It is known that high energy photons can also produce displacement damage effects that may be similar to those produced by neutrons. These effects are known to be much less at the belt line of a light water reactor pressure vessel than those induced by neutrons. The same has not been proven for all locations within vessel support structures. Therefore, it may be prudent to apply coupled neutron-photon transport methods and photon induced displacement cross sections to determine whether gamma-induced dpa exceeds the screening level of 3.0 × 10-4, used in this practice for neutron exposures. (See 1.3).
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1.1 This practice covers procedures for monitoring the neutron radiation exposures experienced by ferritic materials in nuclear reactor vessel support structures located in the vicinity of the active core. This practice includes guidelines for:  
1.1.1 Selecting appropriate dosimetric sensor sets and their proper installation in reactor cavities.  
1.1.2 Making appropriate neutronics calculations to predict neutron radiation exposures.  
1.2 The values stated in SI units are to be regarded as standard; units that are not SI can be found in Terminology E170 and are to be regarded as standard. Any values in parentheses are for information only.  
1.3 This practice is applicable to all pressurized water reactors whose vessel supports will experience a lifetime neutron fluence (E > 1 MeV) that exceeds 1 × 1017 neutrons/cm 2 or exceeds 3.0 × 10−4 dpa.(1)2 (See Terminology E170.)  
1.4 Exposure of vessel support structures by gamma radiation is not included in the scope of this practice, but see the brief discussion of this issue in 3.2.  
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. (For example (2))  
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.

SIGNIFICANCE AND USE
5.1 High levels of antimony are commonly used in flame retardant formulations for various materials. NAA is a test method that can be useful for verifying these levels and, for other materials, NAA can also be useful in establishing the amount of low level contamination, if any, with high sensitivity and high precision.  
5.2 Neutron activation analysis provides a rapid, highly sensitive, nondestructive procedure for antimony determination in a wide range of matrices. This test method is independent of the chemical form of the antimony.  
5.3 This test method can be used for quality and process control in the petrochemical and other manufacturing industries, and for research purposes in a broad spectrum of applications.
SCOPE
1.1 This test method covers the measurement of antimony concentration in plastics or other hydrocarbon or organic matrix by using neutron activation analysis (NAA). The sample is activated by irradiation with neutrons from a research reactor and the subsequently emitted gamma-rays are detected with a germanium semiconductor detector. The same system may be used to determine antimony concentrations ranging from 1 ng/g to 10 000 μg/g with the lower end of the range limited by numerous interferences and the upper limit established by the demonstrated practical application of NAA.  
1.2 This test method may be used on either solid or liquid samples, provided that they can be made to conform in size and shape during irradiation and counting to a standard sample of known antimony content using very simple sample preparation. Several variants of this method have been described in the technical literature. A monograph is available which provides a comprehensive description of the principles of neutron activation analysis using reactor neutrons (1).2  
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. Specific precautions are given in Section 9.  
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.

SIGNIFICANCE AND USE
5.1 Refer to Guide E844 for the selection, irradiation, and quality control of neutron dosimeters.  
5.2 Refer to Test Method 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)3 days(1).4 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 isotopic content of natural titanium recommended for 46Ti is 8.25 %. (2)  
5.6 The radioactive products of the neutron reactions  47Ti(n,p)47Sc (τ1/2 = 3.3485 (9) d) (1) and  48Ti(n,p)48Sc (τ1/2  = 43.67 h), (2) 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 (3); 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 Russian Reactor Dosimetry File (RRDF-2002) cross section (4) versus neutron energy for the fast-neutron reactions of titanium which produce  46Sc [that is, NatTi(n,X)46Sc]. This cross section is identical, for energies up to 20 MeV, to what is found in the latest International Atomic Energy Agency (IAEA) International Reactor Dosimetry and Fusion File, IRDFF-1.05 (5). Included in the plot is the 46Ti(n,p) reaction and the 47Ti(n,np) contribution to the 46Sc production, normalized per  46Ti atom using the natural...
SCOPE
1.1 This test method covers procedures for measuring reaction rates by the activation reactions 46Ti(n,p)  46Sc + 47Ti(n, np)46Sc + 47Ti(n,d)46Sc.
Note 1: The cross section for the 47Ti(n,np+d) reaction is relatively small for energies less than 12 MeV and is not easily distinguished from that of the 46Ti(n,p) reaction. This test method will apply to the composite natTi(n,X) 46Sc reaction that is typically used for dosimetry purposes.  
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 and determine the applicability of regulatory limitations prior to use.  
1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

SIGNIFICANCE AND USE
4.1 A pressure vessel surveillance program requires a methodology for relating radiation-induced changes in materials exposed in accelerated surveillance locations to the condition of the pressure vessel (see Practice E853). An important consideration is that the irradiation exposures be expressed in a unit that is physically related to the damage mechanisms.  
4.2 A major source of neutron radiation damage in metals is the displacement of atoms from their normal lattice sites. Hence, an appropriate damage exposure index is the number of times, on the average, that an atom has been displaced during an irradiation. This can be expressed as the total number of displaced atoms per unit volume, per unit mass, or per atom of the material. Displacements per atom is the most common way of expressing this quantity. The number of dpa associated with a particular irradiation depends on the amount of energy deposited in the material by the neutrons, and hence, depends on the neutron spectrum. (For a more extended discussion, see Practice E521.)  
4.3 No simple correspondence exists in general between dpa and a particular change in a material property. A reasonable starting point, however, for relative correlations of property changes produced in different neutron spectra is the dpa value associated with each environment. That is, the dpa values themselves provide a spectrum-sensitive index that may be a useful correlation parameter, or some function of the dpa values may affect correlation.
TABLE 1 ENDF/B-VI-based Iron Displacement Cross Section    
Bin  
EngA (MeV)  
σd  (barns)  
Bin  
EngA (MeV)  
σd  (barns)  
Bin  
EngA (MeV)  
σd  (barns)  
1  
0.100E-09  
158.3543  
2  
0.1050E-09  
154.6209  
3  
0.110E-09  
151.1395  
4  
0.1150E-09  
147.8895  
5  
0.120E-09  
144.1054  
6  
0.1275E-09  
139.9202  
7  
0.1350E-09  
136.0860  
8  
0.1425E-09  
132.5445  
9  
0.150E-09  
128.7502  
10  
0.160E-09  
124.7860  
11  
0.170E...
SCOPE
1.1 This practice describes a standard procedure for characterizing neutron irradiations of iron (and low alloy steels) in terms of the exposure index displacements per atom (dpa) for iron.  
1.2 Although the methods of this practice apply to any material for which a displacement cross section σd(E) is known (see Practice E521), this practice is written specifically for iron.  
1.3 It is assumed that the displacement cross section for iron is an adequate approximation for calculating displacements in steels that are mostly iron (95 to 100 %) in radiation fields for which secondary damage processes are not important.  
1.4 Procedures analogous to this one can be formulated for calculating dpa in charged particle irradiations. (See Practice E521.)  
1.5 The application of this practice requires knowledge of the total neutron fluence and flux spectrum. Refer to Practice E521 for determining these quantities.  
1.6 The correlation of radiation effects data is beyond the scope of this practice.  
1.7 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.  
1.8 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.

SIGNIFICANCE AND USE
5.1 Refer to Guide E844 for the selection, irradiation, and quality control of neutron dosimeters.  
5.2 Refer to Test Method 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)3 days(1).4 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 isotopic content of natural titanium recommended for 46Ti is 8.25 %. (2)  
5.6 The radioactive products of the neutron reactions  47Ti(n,p)47Sc (τ1/2 = 3.3485 (9) d) (1) and  48Ti(n,p)48Sc (τ1/2  = 43.67 h), (2) 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 (3); 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 Russian Reactor Dosimetry File (RRDF-2002) cross section (4) versus neutron energy for the fast-neutron reactions of titanium which produce  46Sc [that is, natTi(n,X)46Sc]. This cross section is identical, for energies up to 20 MeV, to what is found in the latest International Atomic Energy Agency (IAEA) International Reactor Dosimetry and Fusion File, IRDFF-1.05 (5). Included in the plot is the 46Ti(n,p) reaction and the 47Ti(n,np) contribution to the 46Sc production, normalized per  46Ti atom using the natural...
SCOPE
1.1 This test method covers procedures for measuring reaction rates by the activation reactions 46Ti(n,p)46Sc + 47Ti(n, np)46Sc + 47Ti(n,d)46Sc.
Note 1: The cross section for the 47Ti(n,np+d) reaction is relatively small for energies less than 12 MeV and is not easily distinguished from that of the 46Ti(n,p) reaction. This test method will apply to the composite natTi(n,X)46Sc reaction that is typically used for dosimetry purposes.  
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 and determine the applicability of regulatory limitations prior to use.  
1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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 (5) MeV(2).  
5.5 Fig. 1 shows a plot of the Russian Reactor Dosimetry File (RRDF) 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). This RRDF-2008 cross section is identical to what is found in the latest International Atomic Energy Agency (IAEA) International Reactor Dosimetry and Fusion File, IRDFF-1.05 (7). 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 RRDF-2008/IRDFF-1.05 Library, with EXFOR Experimental Data  
5.6 Two competing activities,  28Al and  27Mg, 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)24 Na.  
1.2 This activation reaction is useful for measuring neutrons with energies above approximately 6.5 MeV and for irradiation times up to about 2 days (for longer irradiations, 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.

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 These test 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.  
1.2 These test methods 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 The measurement standards appear in the following order:
Sections  
Spectroscopy Methods:  
Calibration and Usage of Germa-  
nium Detectors  
3 – 12  
Calibration and Usage of Scintillation  
Detector Systems:  
13 – 20  
Calibration and Usage of Scintillation  
Detectors for Simple Spectra  
16  
Calibration and Usage of Scintillation  
Detectors for Complex Spectra  
17  
Counting Methods:  
Beta Particle Counting  
25-26  
Aluminum Absorption Curve  
27 – 31  
Alpha Particle Counting  
32 – 39  
Liquid Scintillation Counting  
40 – 48  
1.4 Additional information on the set-up, 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 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.  
1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

SIGNIFICANCE AND USE
5.1 Refer to Guide E844 for the selection, irradiation, and quality control of neutron dosimeters.  
5.2 Refer to Practice E261 for a general discussion of the measurement of fast neutron fluence rate with threshold detectors. The general shape of the  63Cu(n,α) 60Co cross section is also shown in Fig. 1 (3, 4) along with a comparison to the current experimental database (5). This figure is for illustrative purposes only to indicate the range of the response of the  63Cu(n,α) 60Co reaction. Refer to Guide E1018 for descriptions of recommended tabulated dosimetry cross sections.  
Note 1: The cross section appropriate for use under this standard is ENDF/B-VI release 8 library since it contains a covariance matrix. The ENDF/B-VII library has been released, but it does not contain a covariance matrix for this reaction. For dosimetry applications, an uncertainty metric expressed as a covariance matrix is required. See Guide E1018.  
5.3 The major advantages of copper for measuring fast-neutron fluence rate are that it has good strength, is easily fabricated, has excellent corrosion resistance, has a melting temperature of 1083°C, and can be obtained in high purity. The half-life of  60 Co is long and its decay scheme is simple and well known.  
5.4 The disadvantages of copper for measuring fast neutron fluence rate are the high reaction apparent threshold of 5 MeV, the possible interference from cobalt impurity (>1 μg/g), the reported possible thermal component of the (n,α) reaction, and the possibly significant cross sections for thermal neutrons for  63Cu and  60Co (that is 4.50 and 2.0 barns, respectively),(6) which will require burnout corrections at high fluences.
SCOPE
1.1 This test method covers procedures for measuring reaction rates by the activation reaction  63Cu(n,α) 60Co. The cross section for  60Co produced in this reaction increases rapidly with neutrons having energies greater than about 4.5 MeV. 60Co decays with a half-life of 1925.27 days (±0.29 days)(1)2 and emits two gamma rays having energies of 1.1732278 and 1.332492 MeV(1). The isotopic content of natural copper is 69.17 %  63Cu and 30.83 %  65Cu(2). The neutron reaction,  63Cu(n,γ)64Cu, produces a radioactive product that emits gamma rays (1.34577 MeV (E1005)) which might interfere with the counting of the  60Co gamma rays.  
1.2 With suitable techniques, fission-neutron fluence rates above 109 cm−2·s−1  can be determined. The  63Cu(n,α)60Co reaction can be used to determine fast-neutron fluences for irradiation times up to about 15 years, provided that the analysis methods described in Practice E261 are followed. If dosimeters are analyzed after irradiation periods longer than 15 years, the information inferred about the fluence during irradiation periods more than 15 years before the end of the irradiation should not be relied upon without supporting data from dosimeters withdrawn earlier.  
1.3 Detailed procedures for other fast-neutron detectors are referenced in Practice E261.  
1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

ABSTRACT
This guide presents the simulation procedure which would provide advice for conducting experiments to investigate the effects of helium on the properties of irradiated metals where the technique for introducing the helium differs in someway from the actual mechanism of introduction of helium in service. Simulation techniques considered for introducing helium shall include charged particle implantation, exposure to α-emitting radioisotopes, and tritium decay techniques. Procedures for the analysis of helium content and helium distribution within the specimen are also recommended. The two other methods for introducing helium into irradiated materials namely, the enhancement of helium production in nickel-bearing alloys by spectral tailoring in mixed-spectrum fission reactors, and the isotopic tailoring in both fast and mixed-spectrum fission reactors, are not covered in this guide. Dual ion beam techniques for simultaneously implanting helium and generating displacement damage are also not included here.
SIGNIFICANCE AND USE
4.1 Helium is introduced into metals as a consequence of nuclear reactions, such as (n, α), or by the injection of helium into metals from the plasma in fusion reactors. The characterization of the effect of helium on the properties of metals using direct irradiation methods may be impractical because of the time required to perform the irradiation or the lack of a radiation facility, as in the case of the fusion reactor. Simulation techniques can accelerate the research by identifying and isolating major effects caused by the presence of helium. The word ‘simulation’ is used here in a broad sense to imply an approximation of the relevant irradiation environment. There are many complex interactions between the helium produced during irradiation and other irradiation effects, so care must be exercised to ensure that the effects being studied are a suitable approximation of the real effect. By way of illustration, details of helium introduction, especially the implantation temperature, may determine the subsequent distribution of the helium (that is, dispersed atomistically, in small clusters in bubbles, etc.).
SCOPE
1.1 This guide provides advice for conducting experiments to investigate the effects of helium on the properties of metals where the technique for introducing the helium differs in some way from the actual mechanism of introduction of helium in service. Techniques considered for introducing helium may include charged particle implantation, exposure to α-emitting radioisotopes, and tritium decay techniques. Procedures for the analysis of helium content and helium distribution within the specimen are also recommended.  
1.2 Three other methods for introducing helium into irradiated materials are not covered in this guide. They are: (1) the enhancement of helium production in nickel-bearing alloys by spectral tailoring in mixed-spectrum fission reactors, (2) a related technique that uses a thin layer of NiAl on the specimen surface to inject helium, and (3) isotopic tailoring in both fast and mixed-spectrum fission reactors. These techniques are described in Refs (1-6).2 Dual ion beam techniques (7) for simultaneously implanting helium and generating displacement damage are also not included here. This latter method is discussed in Practice E521.  
1.3 In addition to helium, hydrogen is also produced in many materials by nuclear transmutation. In some cases it appears to act synergistically with helium (8-10). The specific impact of hydrogen is not addressed in this guide.  
1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
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 and health practices and determine the applicability of regulatory limitations ...

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 and health 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.

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 (36) 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, 37-40). (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 (36), “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 method is...

SIGNIFICANCE AND USE
4.1 A characteristic advantage of charged-particle irradiation experiments is precise, individual, control over most of the important irradiation conditions such as dose, dose rate, temperature, and quantity of gases present. Additional attributes are the lack of induced radioactivation of specimens and, in general, a substantial compression of irradiation time, from years to hours, to achieve comparable damage as measured in displacements per atom (dpa). An important application of such experiments is the investigation of radiation effects that may be obtained in environments which do not currently exist, such as fusion reactors.  
4.2 The primary shortcoming of ion bombardments stems from the damage rate, or temperature dependences of the microstructural evolutionary processes in complex alloys, or both. It cannot be assumed that the time scale for damage evolution can be comparably compressed for all processes by increasing the displacement rate, even with a corresponding shift in irradiation temperature. In addition, the confinement of damage production to a thin layer just (often ∼1 μm) below the irradiated surface can present substantial complications. It must be emphasized, therefore, that these experiments and this practice are intended for research purposes and not for the certification or the qualification of materials.  
4.3 This practice relates to the generation of irradiation-induced changes in the microstructure of metals and alloys using charged particles. The investigation of mechanical behavior using charged particles is covered in Practice E821.
SCOPE
1.1 This practice provides guidance on performing charged-particle irradiations of metals and alloys, although many of the methods may also be applied to ceramic materials. It is generally confined to studies of microstructural and microchemical changes induced by ions of low-penetrating power that come to rest in the specimen. Density changes can be measured directly and changes in other properties can be inferred. This information can be used to estimate similar changes that would result from neutron irradiation. More generally, this information is of value in deducing the fundamental mechanisms of radiation damage for a wide range of materials and irradiation conditions.  
1.2 Where it appears, the word “simulation” should be understood to imply an approximation of the relevant neutron irradiation environment for the purpose of elucidating damage mechanisms. The degree of conformity can range from poor to nearly exact. The intent is to produce a correspondence between one or more aspects of the neutron and charged particle irradiations such that fundamental relationships are established between irradiation or material parameters and the material response.  
1.3 The practice appears as follows:    
Section  
Apparatus  
4  
Specimen Preparation  
5 – 10  
Irradiation Techniques (including Helium Injection)  
11–12  
Damage Calculations  
13  
Postirradiation Examination  
14 – 16  
Reporting of Results  
17  
Correlation and Interpretation  
18 – 22  
1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
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 and health practices and determine the applicability of regulatory limitations prior to use.

SIGNIFICANCE AND USE
5.1 High levels of antimony are commonly used in flame retardant formulations for various materials. NAA is a test method that can be useful for verifying these levels and, for other materials, NAA can also be useful in establishing the amount of low level contamination, if any, with high sensitivity and high precision.  
5.2 Neutron activation analysis provides a rapid, highly sensitive, nondestructive procedure for antimony determination in a wide range of matrices. This test method is independent of the chemical form of the antimony.  
5.3 This test method can be used for quality and process control in the petrochemical and other manufacturing industries, and for research purposes in a broad spectrum of applications.
SCOPE
1.1 This test method covers the measurement of antimony concentration in plastics or other hydrocarbon or organic matrix by using neutron activation analysis (NAA). The sample is activated by irradiation with neutrons from a research reactor and the subsequently emitted gamma-rays are detected with a germanium semiconductor detector. The same system may be used to determine antimony concentrations ranging from 1 ng/g to 10 000 μg/g with the lower end of the range limited by numerous interferences and the upper limit established by the demonstrated practical application of NAA.  
1.2 This test method may be used on either solid or liquid samples, provided that they can be made to conform in size and shape during irradiation and counting to a standard sample of known antimony content using very simple sample preparation. Several variants of this method have been described in the technical literature. A monograph is available which provides a comprehensive description of the principles of neutron activation analysis using reactor neutrons (1).2  
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 and health practices and determine the applicability of regulatory limitations prior to use. Specific precautions are given in Section 9.