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ASTM E900-21

(Guide)

Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials

Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials

SIGNIFICANCE AND USE
4.1 Operation of commercial power reactors must conform to pressure-temperature limits during heatup and cooldown to prevent over-pressurization at temperatures that might cause non-ductile behavior in the presence of a flaw. Radiation damage to the reactor vessel is compensated for by adjusting the pressure-temperature limits to higher temperatures as the neutron damage accumulates. The present practice is to base that adjustment on the TTS  produced by neutron irradiation as measured at the Charpy V-notch 41-J (30-ft·lbf) energy level. To establish pressure temperature operating limits during the operating life of the plant, a prediction of TTS  must be made.  
4.1.1 In the absence of surveillance data for a given reactor material (see Practice E185 and E2215), the use of calculative procedures are necessary to make the prediction. Even when credible surveillance data are available, it will usually be necessary to interpolate or extrapolate the data to obtain a TTS  for a specific time in the plant operating life. The embrittlement correlation presented herein has been developed for those purposes.  
4.2 Research has established that certain elements, notably copper (Cu), nickel (Ni), phosphorus (P), and manganese (Mn), cause a variation in radiation sensitivity of reactor pressure vessel steels. The importance of other elements, such as silicon (Si), and carbon (C), remains a subject of additional research. Copper, nickel, phosphorus, and manganese are the key chemistry parameters used in developing the calculative procedures described here.  
4.3 Only power reactor (PWR and BWR) surveillance data were used in the derivation of these procedures. The measure of fast neutron fluence used in the procedure is n/m2  (E > 1 MeV). Differences in fluence rate and neutron energy spectra experienced in power reactors and test reactors have not been accounted for in these procedures.
SCOPE
1.1 This guide presents a method for predicting values of reference transition temperature shift (TTS) for irradiated pressure vessel materials. The method is based on the TTS exhibited by Charpy V-notch data at 41-J (30-ft·lbf) obtained from surveillance programs conducted in several countries for commercial pressurized (PWR) and boiling (BWR) light-water cooled (LWR) power reactors. An embrittlement correlation has been developed from a statistical analysis of the large surveillance database consisting of radiation-induced TTS and related information compiled and analyzed by Subcommittee E10.02. The details of the database and analysis are described in a separate report (ADJE090015-EA).2,3 This embrittlement correlation was developed using the variables copper, nickel, phosphorus, manganese, irradiation temperature, neutron fluence, and product form. Data ranges and conditions for these variables are listed in 1.1.1. Section 1.1.2 lists the materials included in the database and the domains of exposure variables that may influence TTS but are not used in the embrittlement correlation.  
1.1.1 The range of material and irradiation conditions in the database for variables used in the embrittlement correlation:  
1.1.1.1 Copper content up to 0.4 %.
1.1.1.2 Nickel content up to 1.7 %.
1.1.1.3 Phosphorus content up to 0.03 %.
1.1.1.4 Manganese content within the range from 0.55 to 2.0 %.
1.1.1.5 Irradiation temperature within the range from 255 to 300°C (491 to 572°F).
1.1.1.6 Neutron fluence within the range from 1 × 1021 n/m2 to 2 × 1024 n/m2 (E> 1 MeV).
1.1.1.7 A categorical variable describing the product form (that is, weld, plate, forging).  
1.1.2 The range of material and irradiation conditions in the database for variables not included in the embrittlement correlation:  
1.1.2.1 A533 Type B Class 1 and 2, A302 Grade B, A302 Grade B (modified), and A508 Class 2 and 3. Also, European and Japanese steel grades that are equivalent to these ASTM Grades.
1.1.2.2 Submerged arc welds, shielded a...

General Information

Status
Published
Publication Date
31-Aug-2021
ICS
27.120.10 - Reactor engineering
Technical Committee
E10 - Nuclear Technology and Applications
Drafting Committee
E10.02 - Behavior and Use of Nuclear Structural Materials
Current Stage
Ref Project

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ASTM E900-21 - Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel MaterialsASTM E900-21 - Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
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This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the
Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
Designation: E900 − 21
Standard Guide for
Predicting Radiation-Induced Transition Temperature Shift
1
in Reactor Vessel Materials
This standard is issued under the fixed designation E900; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision.Anumber in parentheses indicates the year of last reapproval.A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope 1.1.1.2 Nickel content up to 1.7 %.
1.1.1.3 Phosphorus content up to 0.03 %.
1.1 This guide presents a method for predicting values of
1.1.1.4 Manganesecontentwithintherangefrom0.55to2.0
reference transition temperature shift (TTS) for irradiated
%.
pressure vessel materials. The method is based on the TTS
1.1.1.5 Irradiationtemperaturewithintherangefrom255to
exhibited by Charpy V-notch data at 41-J (30-ft·lbf) obtained
300°C (491 to 572°F).
from surveillance programs conducted in several countries for
21 2
1.1.1.6 Neutronfluencewithintherangefrom1×10 n/m
commercialpressurized(PWR)andboiling(BWR)light-water
24 2
to2×10 n/m (E> 1 MeV).
cooled (LWR) power reactors. An embrittlement correlation
1.1.1.7 A categorical variable describing the product form
has been developed from a statistical analysis of the large
(that is, weld, plate, forging).
surveillance database consisting of radiation-induced TTS and
1.1.2 The range of material and irradiation conditions in
related information compiled and analyzed by Subcommittee
the database for variables not included in the embrittlement
E10.02. The details of the database and analysis are described
2,3
correlation:
in a separate report (ADJE090015-EA). This embrittlement
correlation was developed using the variables copper, nickel, 1.1.2.1 A533 Type B Class 1 and 2, A302 Grade B, A302
Grade B (modified), and A508 Class 2 and 3. Also, European
phosphorus, manganese, irradiation temperature, neutron
fluence,andproductform.Datarangesandconditionsforthese and Japanese steel grades that are equivalent to these ASTM
Grades.
variables are listed in 1.1.1. Section 1.1.2 lists the materials
includedinthedatabaseandthedomainsofexposurevariables 1.1.2.2 Submerged arc welds, shielded arc welds, and elec-
that may influence TTS but are not used in the embrittlement troslag welds having compositions consistent with those of the
correlation. welds used to join the base materials described in 1.1.2.1.
12
1.1.1 The range of material and irradiation conditions in
1.1.2.3 Neutron fluence rate within the range from3×10
2 16 2
the database for variables used in the embrittlement correla-
n/m/sto5×10 n/m /s (E > 1 MeV).
tion:
1.1.2.4 Neutron energy spectra within the range expected at
1.1.1.1 Copper content up to 0.4 %.
the reactor vessel region adjacent to the core of commercial
PWRs and BWRs (greater than approximately 500MW elec-
tric).
1
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear
1.1.2.5 Irradiation exposure times of up to 25 years in
Technology and Applications and is the direct responsibility of Subcommittee
boiling water reactors and 31 years in pressurized water
E10.02 on Behavior and Use of Nuclear Structural Materials.
reactors.
Current edition approved Sept. 1, 2021. Published October 2021. Originally
ɛ2
approved in 1983. Last previous edition approved in 2015 as E900–15 . DOI:
1.2 It is the responsibility of the user to show that the
10.1520/E0900-21.
2
Available from ASTM International Headquarters. Order Adjunct No. conditions of interest in their application of this guide are
ADJE090015-EA.
addressed adequately by the technical information on which
3
To inform the TTS prediction of Section 5 of this guide, the E10.02
the guide is based. It should be noted that the conditions
Subcommittee decided to limit the data considered to Charpy shift values (∆T )
41J
quantified by the database are not distributed evenly over the
measured from irradiations conducted in PWRs and BWRs. A database of 1,878
Charpy TTSmeasurementswascompiledfromsurveillancereportsonoperatingand
range of materials and irradiation conditions described in 1.1,
decommissioned light water reactors of Western design from 13 countries (Brazil,
and that some combination of variables, particularly at the
Belgium, France, Germany, Italy, Japan, Mexico, The Netherlands, South Korea,
extremes of the data range are under-represented. Particular
Sweden, Switzlerland, Taiwan, and the United States), and from the technical
attention is warranted when the guide is applied to conditions
literature. For each data record, the following information had to be available:
fluence, fluence rate, irradiation tempera
...

This document is not an ASTM standard and is intended only to provide the user of an ASTM standard an indication of what changes have been made to the previous version. Because
it may not be technically possible to adequately depict all changes accurately, ASTM recommends that users consult prior editions as appropriate. In all cases only the current version
of the standard as published by ASTM is to be considered the official document.
´2
Designation: E900 − 15 E900 − 21
Standard Guide for
Predicting Radiation-Induced Transition Temperature Shift
1
in Reactor Vessel Materials
This standard is issued under the fixed designation E900; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1
ε NOTE—Old Ref 1 was editorially removed and the adjunct information was updated and added to Section 2, Refer-
enced Documents, in April 2017.
2
ε NOTE—In 3.1.2 and 3.1.3, “neutrons per square centimeter” was corrected editorially to “neutrons per square meter” to
reflect the usage in Section 5 in July 2020.
1. Scope
1.1 This guide presents a method for predicting values of reference transition temperature shift (TTS) for irradiated pressure vessel
materials. The method is based on the TTS exhibited by Charpy V-notch data at 41-J (30-ft·lbf) obtained from surveillance
programs conducted in several countries for commercial pressurized (PWR) and boiling (BWR) light-water cooled (LWR) power
reactors. An embrittlement correlation has been developed from a statistical analysis of the large surveillance database consisting
of radiation-induced TTS and related information compiled and analyzed by Subcommittee E10.02. The details of the database and
2,3
analysis are described in a separate report (ADJE090015-EA). This embrittlement correlation was developed using the variables
copper, nickel, phosphorus, manganese, irradiation temperature, neutron fluence, and product form. Data ranges and conditions for
these variables are listed in 1.1.1. Section 1.1.2 lists the materials included in the database and the domains of exposure variables
that may influence TTS but are not used in the embrittlement correlation.
1.1.1 The range of material and irradiation conditions in the database for variables used in the embrittlement correlation:
1.1.1.1 Copper content up to 0.4 %.
1.1.1.2 Nickel content up to 1.7 %.
1.1.1.3 Phosphorus content up to 0.03 %.
1.1.1.4 Manganese content within the range from 0.55 to 2.0 %.
1.1.1.5 Irradiation temperature within the range from 255 to 300°C (491 to 572°F).
21 2 24 2
1.1.1.6 Neutron fluence within the range from 1 × 10 n/m to 2 × 10 n/m (E> 1 MeV).
1
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee E10.02 on
Behavior and Use of Nuclear Structural Materials.
Current edition approved Feb. 1, 2015Sept. 1, 2021. Published April 2015October 2021. Originally approved in 1983. Last previous edition approved in 20072015 as
ɛ2
E900 – 02E900 – 15 (2007). DOI: 10.1520/E0900-15E02.10.1520/E0900-21.
2
Available from ASTM International Headquarters. Order Adjunct No. ADJE090015-EA.
3
To inform the TTS prediction of Section 5 of this guide, the E10.02 Subcommittee decided to limit the data considered to Charpy shift values (ΔT ) measured from
41J
irradiations conducted in PWRs and BWRs. A database of 1,878 Charpy TTS measurements was compiled from surveillance reports on operating and decommissioned light
water reactors of Western design from 13 countries (Brazil, Belgium, France, Germany, Italy, Japan, Mexico, The Netherlands, South Korea, Sweden, Switzlerland, Taiwan,
and the United States), and from the technical literature. For each data record, the following information had to be available: fluence, fluence rate, irradiation temperature,
and % content of Cu, Ni, P, and Mn. Reports and technical papers documenting the results of research programs conducted in material test reactors were also reviewed. Data
from these sources was included in the database for information, but was not used in the development of the TTS prediction of Section 5 of this guide.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

---------------------- Page: 1 ----------------------
E900 − 21
1.1.1.7 A categorical variable describing the product form (that is, weld, plate, forging).
1.1.2 The range of material and irradiation conditions in the database for variables not included in the embrittlement correlation:
1.1.2.1 A533 Type B Class 1 and 2, A302 Grade B, A302 Grade B (modified), and A508 Class 2 and 3. Also, European and
Japanese steel grades that are equivalent to these ASTM Grades.
1.1.2.2 Submerged arc welds, shielded arc weld
...

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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.  
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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.  
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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.  
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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.  
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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).
SCOPE
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.

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ASTM
ASTM E509/E509M-21(Guide)
Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

ABSTRACT
This guide covers the general procedures to be considered (as well as the demonstration of their effectiveness) for conducting in-service thermal anneals of light-water moderated nuclear reactor vessels. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods, as well as to provide direction for the development of a vessel annealing procedure and a post-annealing vessel radiation surveillance program. Guidelines are provided to determine the post-anneal reference nil-ductility transition temperature (RTNDT), the Charpy V-notch upper shelf energy level, fracture toughness properties, and the predicted reembrittlement trend for these properties for reactor vessel beltline materials.
SIGNIFICANCE AND USE
3.1 Reactor vessels made of ferritic steels are designed with the expectation of progressive changes in material properties resulting from in-service neutron exposure. In the operation of light-water-cooled nuclear power reactors, changes in pressure-temperature (P – T) limits are made periodically during service life to account for the effects of neutron radiation on the ductile-to-brittle transition temperature material properties. If the degree of neutron embrittlement becomes large, the restrictions on operation during normal heat-up and cool down may become severe. Additional consideration should be given to postulated events, such as pressurized thermal shock (PTS). A reduction in the upper shelf toughness also occurs from neutron exposure, and this decrease may reduce the margin of safety against ductile fracture. When it appears that these situations could develop, certain alternatives are available that reduce the problem or postpone the time at which plant restrictions must be considered. One of these alternatives is to thermally anneal the reactor vessel beltline region, that is, to heat the beltline region to a temperature sufficiently above the normal operating temperature to recover a significant portion of the original fracture toughness and other material properties that were degraded as a result of neutron embrittlement.  
3.2 Preparation and planning for an in-service anneal should begin early so that pertinent information can be obtained to guide the annealing operation. Sufficient time should be allocated to evaluate the expected benefits in operating life to be gained by annealing; to evaluate the annealing method to be employed; to perform the necessary system studies and stress evaluations; to evaluate the expected annealing recovery and reembrittlement behavior; to develop and functionally test such equipment as may be required to do the in-service annealing; and, to train personnel to perform the anneal.  
3.3 Selection of the annealing te...
SCOPE
1.1 This guide covers the general procedures for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1).2  
1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing...

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ASTM
ASTM E2956-23(Guide)
Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels

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.

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ASTM
ASTM E521-23(Practice)
Standard Practice for Investigating the Effects of Neutron Radiation Damage Using Charged-Particle Irradiation

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.

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ASTM
ASTM D7013/D7013M-23(Guide)
Standard Guide for Calibration Facility Setup for Nuclear Surface Gauges

SIGNIFICANCE AND USE
5.1 To establish a proper calibration area for nuclear surface gauges.  
5.2 To reduce the chance of improper calibration.
Note 1: The quality of the results produced by this standard is dependent on the competence of the personnel performing it, and the suitability of the equipment and facilities used. Agencies that meet the criteria of practice D3740 are generally considered capable of competent and objective testing/inspection/etc. Users of this standard are cautioned that compliance with practice D3740 does not in itself assure a means of evaluating some of those factors.
SCOPE
1.1 This guide outlines procedures for setup of a nuclear gauge calibration facility in either a shielded bay or an unshielded area—Guide A and Guide B, respectively.  
1.2 This guide does not attempt to describe the calibration techniques or methods. It is assumed that this guide will be used by persons familiar with the operations of the gauge and in performing proper calibration, service and maintenance.  
1.3 This guide does not attempt to address maintenance or service procedures related to the gauge.  
1.4 The values stated in either SI units or inch-pound units are to be regarded separately as standard. The values stated in each system may not be exact equivalents; therefore, each system shall be used independently of the other. Combining values from the two systems may result in non-conformance with the 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 guide offers an organized collection of information or a series of options and does not recommend a specific course of action. This document cannot replace education or experience and should be used in conjunction with professional judgment. Not all aspects of this guide may be applicable in all circumstances. This ASTM standard is not intended to represent or replace the standard of care by which the adequacy of a given professional service must be judged, nor should this document be applied without consideration of a project’s many unique aspects. The word “Standard” in the title of this document has been approved through ASTM consensus process.  
1.7 All observed and calculated values shall conform to the guidelines for significant digits and rounding established in practice D6026.  
1.7.1 The method used to specify how data are collected, calculated, or recorded in this standard is not directly related to the accuracy to which the data can be applied in the design or other uses, or both. How one applies the results obtained using this standard is beyond its scope.  
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.

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