ASTM E509-03(2008)

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
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 lost as a result of neutron embrittlement.
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.
Selection of the annealing temperature requires...
SCOPE
1.1 This guide covers the general procedures to be considered 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).  
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 develop...

General Information

• Status: Historical
• Publication Date: 30-Jun-2008
• ICS: 27.120.10 - Reactor engineering
• Technical Committee: E10 - Nuclear Technology and Applications
• Drafting Committee: E10.02 - Behavior and Use of Nuclear Structural Materials

Relations

• Replaces: ASTM E509-03 - Standard Guide for In-Service Annealing of Light-Water Cooled Nuclear Reactor Vessels
• Effective Date: 01-Jul-2008
• Referred By: ASTM E2956-23 - Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels
• Effective Date: 01-Jul-2008
• Referred By: ASTM E2215-19 - Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 01-Jul-2008
• Referred By: ASTM E853-23 - Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results
• Effective Date: 01-Jul-2008

Standards Content (Sample)

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: E509 − 03(Reapproved 2008)
Standard Guide for
In-Service Annealing of Light-Water Moderated Nuclear
1
Reactor Vessels
This standard is issued under the fixed designation E509; 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 ability of the required information on vessel mechanical and
fracture properties prior to annealing; evaluation of the par-
1.1 This guide covers the general procedures to be consid-
ticular vessel materials, design, and operation to determine the
ered for conducting an in-service thermal anneal of a light-
annealing time and temperature; and, the procedure to be used
water moderated nuclear reactor vessel and demonstrating the
for verification of the degree of recovery and the trend for
effectiveness of the procedure. The purpose of this in-service
reembrittlement. Guidelines are provided to determine the
annealing (heat treatment) is to improve the mechanical
post-anneal reference nil-ductility transition temperature
properties, especially fracture toughness, of the reactor vessel
(RT ), the Charpy V-notch upper shelf energy level, fracture
NDT
materials previously degraded by neutron embrittlement. The
toughness properties, and the predicted reembrittlement trend
improvement in mechanical properties generally is assessed
for these properties for reactor vessel beltline materials. This
using Charpy V-notch impact test results, or alternatively,
guideemphasizestheneedtoplanwellaheadinanticipationof
fracture toughness test results or inferred toughness property
annealing if an optimum amount of post-anneal reembrittle-
changes from tensile, hardness, indentation, or other miniature
2
ment data is to be available for use in assessing the ability of
specimen testing (1).
anuclearreactorvesseltooperateforthedurationofitspresent
1.2 This guide is designed to accommodate the variable
license, or qualify for a license extension, or both.
response of reactor-vessel materials in post-irradiation anneal-
1.4 The values stated in inch-pound or SI units are to be
ing at various temperatures and different time periods. Certain
regarded separately as the standard.
inherent limiting factors must be considered in developing an
1.5 This standard does not purport to address all of the
annealing procedure. These factors include system-design
safety concerns, if any, associated with its use. It is the
limitations;physicalconstraintsresultingfromattachedpiping,
responsibility of the user of this standard to establish appro-
support structures, and the primary system shielding; the
priate safety and health practices and determine the applica-
mechanical and thermal stresses in the components and the
bility of regulatory limitations prior to use.
system as a whole; and, material condition changes that may
limit the annealing temperature.
2. Referenced Documents
1.3 This guide provides direction for development of the
3
2.1 ASTM Standards:
vessel annealing procedure and a post-annealing vessel radia-
E185 Practice for Design of Surveillance Programs for
tion surveillance program. The development of a surveillance
Light-Water Moderated Nuclear Power Reactor Vessels
program to monitor the effects of subsequent irradiation of the
E636 Guide for Conducting Supplemental Surveillance
annealed-vessel beltline materials should be based on the
Tests for Nuclear Power Reactor Vessels, E 706 (IH)
requirements and guidance described in Practices E185 and
E900 Guide for Predicting Radiation-Induced Transition
E2215. The primary factors to be considered in developing an
Temperature Shift in Reactor Vessel Materials, E706 (IIF)
effective annealing program include the determination of the
E1253 Guide for Reconstitution of Irradiated Charpy-Sized
feasibility of annealing the specific reactor vessel; the avail-
Specimens
E2215 Practice for Evaluation of Surveillance Capsules
1
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear from Light-Water Moderated Nuclear Power Reactor Ves-
Technology and Applicationsand is the direct responsibility of Subcommittee
sels
E10.02 on Behavior and Use of Nuclear Structural Materials.
Current edition approved July 1, 2008. Published September 2008. Originally
3
approved in 1997. Last previous edition approved in 2003 as E509–03. DOI: For referenced ASTM standards, visit the ASTM website, www.astm.org, or
10.1520/E0509-03R08. contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM
2
The boldface numbers in parentheses refer to the list of references at the end of Standards volume information, refer to the standard’s Document Summ
...

This document is not anASTM standard and is intended only to provide the user of anASTM 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:E509–97 Designation:E509–03 (Reapproved 2008)
Standard Guide for
In-Service Annealing of Light-Water CooledModerated
1
Nuclear Reactor Vessels
This standard is issued under the fixed designation E 509; 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 to be considered for conducting an in-service thermal anneal of a light-water-
cooled 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
2
tensile, hardness, indentation, or other miniature specimen testing (1).
1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation heat treatment
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 E 185 and E 2215.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
fracturetoughnessproperties,andthepredictedreembrittlementtrendforthesepropertiesforreactorvesselbeltlinematerials.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 inch-pound or SI units are to be regarded separately as 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 and health practices and determine the applicability of regulatory
limitations prior to use.
2. Referenced Documents
2.1 ASTM Standards: E184Practice for Effects of High-Energy Neutron Radiation on the Mechanical Properties of Metallic
3
Materials E706 (1B)
3
E 185Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels E706 (IF) Practice
for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
E 636Practice Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)
3
E 900Guide for Predicting Neutron Radiation Damage to Reactor Vessel Materials E706 (IIF) Guide for Predicting
1
This guide is under the jurisdiction of ASTM Committee E-10 on Nuclear Technology and Applications and is the direct responsibility of Subcommittee E10.02 on
Behavior and Use of Metallic Materials in Nuclear Systems.
Current edition approved June 10, 1997. Published May 1998.
1
This guide i
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