ASTM E509/E509M-21

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...

General Information

• Status: Published
• Publication Date: 31-Jan-2021
• ICS: 27.120.10 - Reactor engineering
• Technical Committee: E10 - Nuclear Technology and Applications
• Drafting Committee: E10.02 - Behavior and Use of Nuclear Structural Materials

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: E509/E509M − 21
Standard Guide for
In-Service Annealing of Light-Water Moderated Nuclear
1
Reactor Vessels
This standard is issued under the fixed designation E509/E509M; 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 annealing time and temperature; and, the procedure to be used
for verification of the degree of recovery and the trend for
1.1 This guide covers the general procedures for conducting
reembrittlement. Guidelines are provided to determine the
anin-servicethermalannealofalight-watermoderatednuclear
post-anneal reference nil-ductility transition temperature
reactor vessel and demonstrating the effectiveness of the
(RT ), the Charpy V-notch upper shelf energy level, fracture
NDT
procedure. The purpose of this in-service annealing (heat
toughness properties, and the predicted reembrittlement trend
treatment) is to improve the mechanical properties, especially
for these properties for reactor vessel beltline materials. This
fracture toughness, of the reactor vessel materials previously
guideemphasizestheneedtoplanwellaheadinanticipationof
degraded by neutron embrittlement. The improvement in
annealing if an optimum amount of post-anneal reembrittle-
mechanical properties generally is assessed using Charpy
ment data is to be available for use in assessing the ability of
V-notch impact test results, or alternatively, fracture toughness
anuclearreactorvesseltooperateforthedurationofitspresent
testresultsorinferredtoughnesspropertychangesfromtensile,
2 license, or qualify for a license extension, or both.
hardness, indentation, or other miniature specimen testing (1).
1.4 The values stated in either SI units or inch-pound units
1.2 This guide is designed to accommodate the variable
are to be regarded separately as standard. The values stated in
response of reactor-vessel materials in post-irradiation anneal-
each system are not necessarily exact equivalents; therefore, to
ing at various temperatures and different time periods. Certain
ensure conformance with the standard, each system shall be
inherent limiting factors must be considered in developing an
used independently of the other, and values from the two
annealing procedure. These factors include system-design
systems shall not be combined.
limitations;physicalconstraintsresultingfromattachedpiping,
1.5 This standard does not purport to address all of the
support structures, and the primary system shielding; the
safety concerns, if any, associated with its use. It is the
mechanical and thermal stresses in the components and the
responsibility of the user of this standard to establish appro-
system as a whole; and, material condition changes that may
priate safety, health, and environmental practices and deter-
limit the annealing temperature.
mine the applicability of regulatory limitations prior to use.
1.3 This guide provides direction for development of the
1.6 This international standard was developed in accor-
vessel annealing procedure and a post-annealing vessel radia-
dance with internationally recognized principles on standard-
tion surveillance program. The development of a surveillance
ization established in the Decision on Principles for the
program to monitor the effects of subsequent irradiation of the
Development of International Standards, Guides and Recom-
annealed-vessel beltline materials should be based on the
mendations issued by the World Trade Organization Technical
requirements and guidance described in Practices E185 and
Barriers to Trade (TBT) Committee.
E2215. The primary factors to be considered in developing an
effective annealing program include the determination of the
2. Referenced Documents
feasibility of annealing the specific reactor vessel; the avail-
3
2.1 ASTM Standards:
ability of the required information on vessel mechanical and
E185 Practice for Design of Surveillance Programs for
fracture properties prior to annealing; evaluation of the par-
Light-Water Moderated Nuclear Power Reactor Vessels
ticular vessel materials, design, and operation to determine the
E636 Guide for Conducting Supplemental Surveillance
Tests for Nuclear Power Reactor Vessels
1
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear
E900 Guide for Predicting Radiation-Induced Transition
Technology and Applications and is the direct responsibility of Subcommittee
E10.02 on Behavior and Use of Nuclear Structural Materials.
Current edition app
...

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: E509/E509M − 14 E509/E509M − 21
Standard Guide for
In-Service Annealing of Light-Water Moderated Nuclear
1
Reactor Vessels
This standard is issued under the fixed designation E509/E509M; 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 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
2
miniature specimen testing (1).
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
procedure. These factors include system-design limitations; physical constraints resulting from attached piping, support structures,
and the primary system shielding; the mechanical and thermal stresses in the components and the system as a whole; and, material
condition changes that may limit the annealing temperature.
1.3 This guide provides direction for development of the vessel annealing procedure and a post-annealing vessel radiation
surveillance program. The development of a surveillance program to monitor the effects of subsequent irradiation of the
annealed-vessel beltline materials should be based on the requirements and guidance described in Practices E185 and E2215. The
primary factors to be considered in developing an effective annealing program include the determination of the feasibility of
annealing the specific reactor vessel; the availability of the required information on vessel mechanical and fracture properties prior
to annealing; evaluation of the particular vessel materials, design, and operation to determine the annealing time and temperature;
and, the procedure to be used for verification of the degree of recovery and the trend for reembrittlement. Guidelines are provided
to determine the post-anneal reference nil-ductility transition temperature (RT ), the Charpy V-notch upper shelf energy level,
NDT
fracture toughness properties, and the predicted reembrittlement trend for these properties for reactor vessel beltline materials. This
guide emphasizes the need to plan well ahead in anticipation of annealing if an optimum amount of post-anneal reembrittlement
data is to be available for use in assessing the ability of a nuclear reactor vessel to operate for the duration of its present license,
or qualify for a license extension, or both.
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 are not necessarily exact equivalents; therefore, to ensure conformance with the standard, each system shall be used
independently of the other, and values from the two systems shall not be combined.
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 Jan. 1, 2014Feb. 1, 2021. Published February 2014February 2021. Originally approved in 1997. Last previous edition approved in 20082014
as E509–03 (2008). –14. DOI: 10.1520/E0509_E0509M-14.10.1520/E0509_E0509M-21.
2
The boldface numbers in parentheses refer to the list of references at the end of this standard.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
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E509/E509M − 21
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 safety, health, and healthenvironmental practices and deter
...