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ASTM E900-02(2007)

(Guide)

Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)

Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)

SIGNIFICANCE AND USE
Operation of commercial power reactors must conform to pressure-temperature limits during heatup and cooldown to prevent over-pressurization at temperatures that might cause nonductile behavior in the presence of a flaw. Radiation damage to the reactor vessel beltline region is compensated for by adjusting the pressure-temperature limits to higher temperature as the neutron damage accumulates. The present practice is to base that adjustment on the increase in transition temperature produced by neutron irradiation as measured at the Charpy V-notch 30-ft·lbf (41-J) energy level. To establish pressure temperature operating limits during the operating life of the plant, a prediction of adjustment in transition temperature must be made.
4.1.1 In the absence of surveillance data for a given reactor (see Practice E 185), the use of calculative procedures will be necessary to make the prediction. Even when credible surveillance data are available, it will usually be necessary to extrapolate the data to obtain an adjustment in transition temperature for a specific time in the plant operating life. The embrittlement correlation presented herein has been developed for those purposes.
Research has established that certain elements, notably copper and nickel, cause a variation in radiation sensitivity of steels. The importance of other elements, such as phosphorus (P), remains a subject of additional research. Copper and nickel are the key chemistry parameters used in developing the calculative procedures described here.
Only power reactor surveillance data were used in the derivation of these procedures. The measure of fast neutron fluence used in the procedure is n/cm2  (E > 1 MeV). Differences in the neutron fluence rate and neutron energy spectra experienced in power reactors and test reactors have not been applied in these procedures. The manner in which these factors were considered is addressed elsewhere.3
SCOPE
1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 30-ftlbf (41-J) data. Radiation damage calculative procedures have been developed from a statistical analysis of an irradiated material database that was available as of May 2000. The embrittlement correlation used in this guide was developed using the following variables: copper and nickel contents, irradiation temperature, and neutron fluence. The form of the model was based on current understanding for two mechanisms of embrittlement: stable matrix damage (SMD) and copper-rich precipitation (CRP); saturation of copper effects (for different weld materials) was included. This guide is applicable for the following specific materials, copper, nickel, and phosphorus contents, range of irradiation temperature, and neutron fluence based on the overall database:
1.1.1 MaterialsA 533 Type B Class 1 and 2, A302 Grade B, A302 Grade B (modified), A508 Class 2 and 3.
Submerged arc welds, shielded arc welds, and electroslag welds for materials in .
1.1.2 Copper contents within the range from 0 to 0.50 wt %.
1.1.3 Nickel content within the range from 0 to 1.3 wt %.
1.1.4 Phosphorus content within the range 0 to 0.025 wt %.
1.1.5 Irradiation exposure temperature within the range from 500 to 570F (260 to 299C).
1.1.6 Neutron fluence within the range from 1 1016 to 8 1019 n/cm2  (E > 1 MeV).
1.1.7 Neutron energy spectra within the range expected at the reactor vessel core beltline region of light water cooled reactors and fluence rate within the range from 2 108 to 1 1012 n/cm2s (E > 1 MeV).
1.2 The basis for the method of adjusting the reference temperature is discussed in a separate report.
1.3 This guide is Part IIF of Master Matrix E 706 which coordinates several standards used for irradiation surveillance of light-water reactor vessel materials. Methods of determining the applicable fluence for ...

General Information

Status
Historical
Publication Date
14-Jul-2007
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

Relations

Replaces
ASTM E900-02 - Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)
Effective Date
15-Jul-2007
Replaced By
ASTM E900-15 - Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
Effective Date
01-Feb-2015
Referred By
ASTM E853-23 - Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results
Effective Date
15-Jul-2007
Referred By
ASTM E1005-21 - Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
Effective Date
15-Jul-2007
Referred By
ASTM E185-21 - Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
Effective Date
15-Jul-2007
Referred By
ASTM E1214-11(2023) - Standard Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance
Effective Date
15-Jul-2007
Referred By
ASTM E2215-19 - Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
Effective Date
15-Jul-2007
Referred By
ASTM E910-18 - Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance
Effective Date
15-Jul-2007
Referred By
ASTM E1006-21 - Standard Practice for Analysis and Interpretation of Physics Dosimetry Results from Test Reactor Experiments
Effective Date
15-Jul-2007
Referred By
ASTM E509/E509M-21 - Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels
Effective Date
15-Jul-2007
Referred By
ASTM E2956-23 - Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels
Effective Date
15-Jul-2007

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ASTM E900-02(2007) - Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)ASTM E900-02(2007) - Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)
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ASTM E900-02(2007) - Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)
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NOTICE: This standard has either been superseded and replaced by a new version or withdrawn.
Contact ASTM International (www.astm.org) for the latest information
Designation: E900 − 02(Reapproved 2007)
Standard Guide for
Predicting Radiation-Induced Transition Temperature Shift
1
in Reactor Vessel Materials, E706 (IIF)
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. Scope 1.2 The basis for the method of adjusting the reference
3
temperature is discussed in a separate report.
1.1 This guide presents a method for predicting reference
transition temperature adjustments for irradiated light-water 1.3 This guide is Part IIF of Master Matrix E706 which
cooled power reactor pressure vessel materials based on coordinates several standards used for irradiation surveillance
Charpy V-notch 30-ft·lbf (41-J) data. Radiation damage calcu- oflight-waterreactorvesselmaterials.Methodsofdetermining
lative procedures have been developed from a statistical the applicable fluence for use in this guide are addressed in
analysisofanirradiatedmaterialdatabasethatwasavailableas Master Matrix E706, Practices E560 (IC) and Guide E944
2
ofMay2000. Theembrittlementcorrelationusedinthisguide (IIA), and Test Method E1005 (IIIA). The overall application
wasdevelopedusingthefollowingvariables:copperandnickel of these separate guides and practices is described in Practice
contents, irradiation temperature, and neutron fluence. The E853 (IA).
form of the model was based on current understanding for two
1.4 The values given in customary U.S. units are to be
mechanisms of embrittlement: stable matrix damage (SMD)
regarded as the standard. The SI values given in parentheses
and copper-rich precipitation (CRP); saturation of copper
are for information only.
effects (for different weld materials) was included. This guide
1.5 This standard guide does not define how the shift in
is applicable for the following specific materials, copper,
transition temperature should be used to determine the final
nickel, and phosphorus contents, range of irradiation
adjusted reference temperature. (That would typically include
temperature,andneutronfluencebasedontheoveralldatabase:
consideration of the initial starting point, the predicted shift,
1.1.1 Materials:
and the uncertainty in the shift estimation method.)
1.1.1.1 A533 Type B Class 1 and 2, A302 Grade B, A302
Grade B (modified), A508 Class 2 and 3. 1.6 This standard does not purport to address all of the
1.1.1.2 Submerged arc welds, shielded arc welds, and elec- safety concerns, if any, associated with its use. It is the
responsibility of the user of this standard to establish appro-
troslag welds for materials in 1.1.1.1.
1.1.2 Coppercontentswithintherangefrom0to0.50wt%. priate safety and health practices and determine the applica-
bility of regulatory limitations prior to use.
1.1.3 Nickel content within the range from 0 to 1.3 wt%.
1.1.4 Phosphorus content within the range 0 to 0.025 wt%.
2. Referenced Documents
1.1.5 Irradiation exposure temperature within the range
4
from 500 to 570°F (260 to 299°C).
2.1 ASTM Standards:
16
1.1.6 Neutron fluence within the range from 1 × 10 to8×
E185Practice for Design of Surveillance Programs for
19 2
10 n/cm (E > 1 MeV).
Light-Water Moderated Nuclear Power Reactor Vessels
1.1.7 Neutron energy spectra within the range expected at
E560Practice for Extrapolating Reactor Vessel Surveillance
5
the reactor vessel core beltline region of light water cooled
Dosimetry Results, E 706(IC) (Withdrawn 2009)
8
reactors and fluence rate within the range from2×10 to1×
E693Practice for Characterizing Neutron Exposures in Iron
12 2
10 n/cm s (E > 1 MeV).
and Low Alloy Steels in Terms of Displacements Per
1 3
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear Charpy Embrittlement Correlations—Status of Combined Mechanistic and
Technology and Applications and is the direct responsibility of Subcommittee Statistical Bases for U.S. Pressure Vessel Steels (MRP-45), PWR Materials
E10.02 on Behavior and Use of Nuclear Structural Materials. Reliability Program (PWRMRP), EPRI, Palo Alto, CA, 2001, 1000705.
4
Current edition approved July 15, 2007. Published August 2007. Originally For referenced ASTM standards, visit the ASTM website, www.astm.org, or
approved in 1983. Last previous edition approved in 2002 as E900–02. DOI: contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
10.1520/E0900-02R07. Standards volume information, refer to the standard’s Document Summary page on
2
The Charpy surveillance data were originally obtained from the Oak Ridge the ASTM website.
5
National Laboratory Power Reactor-Embrittlement Database (PR-EDB) and subse- The last
...

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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.  
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.  
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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  
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Specimen Preparation  
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Irradiation Techniques (including Helium Injection)  
11 – 12  
Damage Calculations  
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Postirradiation Examination  
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Reporting of Results  
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Correlation and Interpretation  
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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
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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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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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1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.
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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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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).  
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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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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
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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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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 E2005-21(Guide)
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.

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ASTM
ASTM E636-20(Guide)
Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels

SIGNIFICANCE AND USE
3.1 Practices E185 and E2215 describe a minimum program for the surveillance of reactor vessel materials, specifically mechanical property changes that occur in service. This guide may be applied to generate additional information on radiation-induced property changes to better assist the determination of the optimum reactor vessel operation schemes.
SCOPE
1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required by Practices E185 and E2215 for the surveillance of nuclear reactor vessels. The supplemental mechanical property tests outlined permit the acquisition of additional information on radiation-induced changes in mechanical properties of the reactor vessel steels.  
1.2 This guide provides recommendations for the preparation of test specimens for irradiation, and identifies special precautions and requirements for reactor surveillance operations and post-irradiation test planning. Guidance on data reduction and computational procedures is also given. Reference is made to other ASTM test methods for the physical conduct of specimen tests and for raw data acquisition.  
1.3 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.  
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 E2215-19(Practice)
Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels

SIGNIFICANCE AND USE
4.1 Neutron radiation effects are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters may be made throughout the service life of the reactor to account for these effects. A surveillance program is used to measure changes in the properties of actual vessel materials due to the irradiation environment. This practice describes the criteria that should be considered in evaluating surveillance program test capsules.  
4.2 Prior to the first issue date of this standard, the design of surveillance programs and the testing of surveillance capsules were both covered in a single standard, Practice E185. Between its provisional adoption in 1961 and its replacement linked to this standard, Practice E185 was revised many times (1966, 1970, 1973, 1979, 1982, 1993 and 1998). Therefore, capsules from surveillance programs that were designed and implemented under early versions of the standard were often tested after substantial changes to the standard had been adopted. For clarity, the standard practice for surveillance programs has been divided into the new Practice E185 that covers the design of new surveillance programs and this standard practice that covers the testing and evaluation of surveillance capsules. Modifications to the standard test program and supplemental tests are described in Guide E636.  
4.3 This practice is intended to cover testing and evaluation of all light-water moderated reactor pressure vessel surveillance capsules. The practice is applicable to testing of capsules from surveillance programs designed and implemented under all previous versions of Practice E185.  
4.4 The radiation-induced changes in the properties of the reactor pressure vessel are generally monitored by measuring the index temperatures, the upper-shelf energy and the tensile properties of specimens from the surveillance program capsules. The significance of these radiation-induced changes is described in Practice E185.  
4...
SCOPE
1.1 This practice covers the evaluation of test specimens and dosimetry from light water moderated nuclear power reactor pressure vessel surveillance capsules.  
1.2 Additionally, this practice provides guidance on reassessing withdrawal schedule for design life and operation beyond design life.  
1.3 This practice is one of a series of standard practices that outline the surveillance program required for nuclear reactor pressure vessels. The surveillance program monitors the irradiation-induced changes in the ferritic steels that comprise the beltline of a light-water moderated nuclear reactor pressure vessel.  
1.4 This practice along with its companion surveillance program practice, Practice E185, is intended for application in monitoring the properties of beltline materials in any light-water moderated nuclear reactor.2  
1.5 Modifications to the standard test program and supplemental tests are described in Guide E636.  
1.6 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.  
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.

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