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ASTM E2005-21

(Guide)

Standard Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields

Standard Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields

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.

General Information

Status
Published
Publication Date
31-Jan-2021
ICS
27.120.10 - Reactor engineering
Technical Committee
E10 - Nuclear Technology and Applications
Drafting Committee
E10.05 - Nuclear Radiation Metrology
Current Stage
Ref Project

Relations

Refers
ASTM E265-15(2020) - Standard Test Method for Measuring Reaction Rates and Fast-Neutron Fluences by Radioactivation of Sulfur-32
Effective Date
01-Jul-2020

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ASTM E2005-21 - Standard Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron FieldsASTM E2005-21 - Standard Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields
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ASTM E2005-21 - Standard Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields
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REDLINE ASTM E2005-21 - Standard Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron FieldsREDLINE ASTM E2005-21 - Standard Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields
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Standards Content (Sample)

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: E2005 − 21
Standard Guide for
Benchmark Testing of Reactor Dosimetry in Standard and
1
Reference Neutron Fields
This standard is issued under the fixed designation E2005; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope Rate, and Spectra by Radioactivation Techniques
E263 Test Method for Measuring Fast-Neutron Reaction
1.1 This guide covers facilities and procedures for bench-
Rates by Radioactivation of Iron
marking neutron measurements and calculations. Particular
E264 Test Method for Measuring Fast-Neutron Reaction
sections of the guide discuss: the use of well-characterized
Rates by Radioactivation of Nickel
benchmark neutron fields to calibrate integral neutron sensors;
E265 Test Method for Measuring Reaction Rates and Fast-
the use of certified-neutron-fluence standards to calibrate
Neutron Fluences by Radioactivation of Sulfur-32
radiometriccountingequipmentortodetermineinterlaboratory
E266 Test Method for Measuring Fast-Neutron Reaction
measurement consistency; development of special benchmark
Rates by Radioactivation of Aluminum
fields to test neutron transport calculations; use of well-known
E343 Test Method for Measuring Reaction Rates by Analy-
fissionspectratobenchmarkspectrum-averagedcrosssections;
sis of Molybdenum-99 Radioactivity From Fission Do-
and the use of benchmarked data and calculations to determine
3
simeters (Withdrawn 2002)
the uncertainties in derived neutron dosimetry results.
E393 Test Method for Measuring Reaction Rates by Analy-
1.2 The values stated in SI units are to be regarded as
sis of Barium-140 From Fission Dosimeters
standard. No other units of measurement are included in this
E482 Guide for Application of Neutron Transport Methods
standard.
for Reactor Vessel Surveillance
1.3 This standard does not purport to address all of the
E523 Test Method for Measuring Fast-Neutron Reaction
safety concerns, if any, associated with its use. It is the Rates by Radioactivation of Copper
responsibility of the user of this standard to establish appro-
E526 Test Method for Measuring Fast-Neutron Reaction
priate safety, health, and environmental practices and deter- Rates by Radioactivation of Titanium
mine the applicability of regulatory limitations prior to use.
E704 Test Method for Measuring Reaction Rates by Radio-
1.4 This international standard was developed in accor-
activation of Uranium-238
dance with internationally recognized principles on standard- E705 Test Method for Measuring Reaction Rates by Radio-
ization established in the Decision on Principles for the
activation of Neptunium-237
Development of International Standards, Guides and Recom- E854 Test Method for Application and Analysis of Solid
mendations issued by the World Trade Organization Technical
State Track Recorder (SSTR) Monitors for Reactor Sur-
Barriers to Trade (TBT) Committee. veillance
E910 Test Method for Application and Analysis of Helium
2. Referenced Documents
Accumulation Fluence Monitors for Reactor Vessel Sur-
2
veillance
2.1 ASTM Standards:
E1297 Test Method for Measuring Fast-Neutron Reaction
E170 Terminology Relating to Radiation Measurements and
Rates by Radioactivation of Niobium
Dosimetry
E2006 Guide for Benchmark Testing of Light Water Reactor
E261 Practice for Determining Neutron Fluence, Fluence
Calculations
1
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear 3. Significance and Use
Technology and Applications and is the direct responsibility of Subcommittee
3.1 This guide describes approaches for using neutron fields
E10.05 on Nuclear Radiation Metrology.
Current edition approved Feb. 1, 2021. Published March 2021. Originally with well known characteristics to perform calibrations of
approved in 1999. Last previous edition approved in 2015 as E2005– 10(2015).
neutron sensors, to intercompare different methods of
DOI: 10.1520/E2005-21.
2
For referenced ASTM standards, visit the ASTM website, www.astm.org, or
contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
3
Standards volume information, refer to the standard’s Document Summary page on The last approved version of this historical standard is referenced on
the ASTM website. www.astm.org.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

---------------------- Page: 1 ----------------------
E2005 − 21
252
dosimetry, and to corroborate procedures u
...

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.
Designation: E2005 − 10 (Reapproved 2015) E2005 − 21
Standard Guide for
Benchmark Testing of Reactor Dosimetry in Standard and
1
Reference Neutron Fields
This standard is issued under the fixed designation E2005; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope
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.
2. Referenced Documents
2
2.1 ASTM Standards:
E170 Terminology Relating to Radiation Measurements and Dosimetry
E261 Practice for Determining Neutron Fluence, Fluence Rate, and Spectra by Radioactivation Techniques
E263 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Iron
E264 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Nickel
E265 Test Method for Measuring Reaction Rates and Fast-Neutron Fluences by Radioactivation of Sulfur-32
E266 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Aluminum
E343 Test Method for Measuring Reaction Rates by Analysis of Molybdenum-99 Radioactivity From Fission Dosimeters
3
(Withdrawn 2002)
E393 Test Method for Measuring Reaction Rates by Analysis of Barium-140 From Fission Dosimeters
E482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance
E523 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Copper
1
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee E10.05 on Nuclear
Radiation Metrology.
Current edition approved Oct. 1, 2015Feb. 1, 2021. Published November 2015March 2021. Originally approved in 1999. Last previous edition approved in 20102015 as
E2005 - 10.E2005– 10(2015). DOI: 10.1520/E2005-10R15.10.1520/E2005-21.
2
For referenced ASTM standards, visit the ASTM website, www.astm.org, or contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM Standards
volume information, refer to the standard’s Document Summary page on the ASTM website.
3
The last approved version of this historical standard is referenced on www.astm.org.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

---------------------- Page: 1 ----------------------
E2005 − 21
E526 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Titanium
E704 Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238
E705 Test Method for Measuring Reaction Rates by Radioactivation of Neptunium-237
E854 Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance
E910 Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance
E1297 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Niobium
E2006 Guide for Benchmark Testing of Light Water Reactor Calculations
3. Significance and Use
3.1 This guide describe
...

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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.  
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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.  
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ASTM E2006-22(Guide)
Standard Guide for Benchmark Testing of Light Water Reactor Calculations

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.  
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SCOPE
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.  
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Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

SIGNIFICANCE AND USE
4.1 Predictions of neutron radiation effects on pressure vessel steels are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters often are made throughout the service life of the reactor vessel to account for radiation effects. Due to the variability in the behavior of reactor vessel steels, a surveillance program is warranted to monitor changes in the properties of actual vessel materials caused by long-term exposure to the neutron radiation and temperature environment of the reactor vessel. This practice describes the criteria that should be considered in planning and implementing surveillance test programs and points out precautions that should be taken to ensure that: (1) capsule exposures can be related to beltline exposures, (2) materials selected for the surveillance program are samples of those materials most likely to limit the operation of the reactor vessel, and (3) the test specimen types are appropriate for the evaluation of radiation effects on the reactor vessel.  
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1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. New advanced light-water small modular reactor designs with a nominal design output of 300 MWe or less have not been specifically considered in this practice. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials.  
1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) exceeds 1 × 1021 neutrons/m 2  (1 × 1017 n/cm2) at the inside surface of the ferritic steel reactor vessel.  
1.3 This practice does not provide specific procedures for monitoring the radiation induced changes in properties beyond the design life. Practice E2215 addresses changes to the withdrawal schedule during and beyond the design life.  
1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.
Note 1: The increased complexity of the requirements for a light-water moderated nuclear power reactor vessel surveillance program has necessitated the separation of the requirements into three related standards. Practice E185 describes the minimum requirements for design of a surveillance program. Practice E2215 describes the procedures for testing and evaluation of surveillance capsules removed from a reactor vessel. Guide E636 provides guidance for conducting additional mechanical tests. A summary of the many major revisions to Practice E185 since its original issuance is contained in Appendix X2.
Note 2: This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. See Appendix X2.  
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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.  
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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...

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Standard Guide for Radiation Protection Program for Decommissioning Operations

SIGNIFICANCE AND USE
4.1 A program based on this guide will provide assurance to all concerned that the appropriate elements of radiation safety have been included to protect workers, the general public, and the environment in proximity to the decommissioning activities.  
4.2 Implementation of such a program will provide assurance to those agencies responsible for review or audit of the decommissioning project that the requirements for radiation protection have been addressed.
SCOPE
1.1 This guide provides instruction to the individual charged with the responsibility for developing and implementing the radiation protection program for decommissioning operations.  
1.2 This guide provides a basis for the user to develop radiation protection program documentation that will support both the radiological engineering and radiation safety aspects of the decommissioning project.  
1.3 This guide presents a description of those elements that should be addressed in a specific radiation protection plan for each decommissioning project. The plan would, in turn, form the basis for development of the implementation procedures that execute the intent of the plan.  
1.4 This guide applies to the development of radiation protection programs established to control exposures to radiation and radioactive materials associated with the decommissioning of nuclear facilities. The intent of this guide is to supplement existing radiation protection programs as they may pertain to decommissioning workers, members of the general public and the environment by describing the basic elements of a radiation protection program for decommissioning operations.  
1.5 This guide defines the elements of a radiation protection program that will ensure that the goals and objectives of a decommissioning activity are attained within the radiological limits and restrictions imposed by applicable governing and regulating agencies. The implementation of such a program will provide radiological protection to personnel and the environment. This guide should be used for developing the documentation that defines the intent and implementation of the radiation protection program for a specific decommissioning project.  
1.6 The Radiation Protection Program should address the following elements (see Note 1). This program shall be developed and maintained such that it satisfies all applicable Quality Assurance requirements developed for the decommissioning project.  
Note 1: If the site to be decommissioned is adjacent to an operating site, the radiological impact of the operating site must be considered in the development of the Radiation Protection Program for the decommissioning site.  
1.7 This guide does not address the subjects of emergency preparedness, safeguards, accountability, waste handling, storage, and transportation. Each of these issues has a direct interface with the radiation protection program. However, each constitutes a program in and of itself from program definition through implementation.  
1.8 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.  
1.9 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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