ASTM E509/E509M-14

Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

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 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,...

General Information

• Status: Historical
• Publication Date: 31-Dec-2013
• 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)

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/E509M − 14
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 may not be exact equivalents; therefore, each
ing at various temperatures and different time periods. Certain
system shall be used independently of the other. Combining
inherent limiting factors must be considered in developing an
values from the two systems may result in non-conformance
annealing procedure. These factors include system-design
with the standard.
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 and health practices and determine the applica-
limit the annealing temperature.
bility of regulatory limitations prior to use.
1.3 This guide provides direction for development of the
vessel annealing procedure and a post-annealing vessel radia-
2. Referenced Documents
tion surveillance program. The development of a surveillance
3
2.1 ASTM Standards:
program to monitor the effects of subsequent irradiation of the
E185 Practice for Design of Surveillance Programs for
annealed-vessel beltline materials should be based on the
Light-Water Moderated Nuclear Power Reactor Vessels
requirements and guidance described in Practices E185 and
E636 Guide for Conducting Supplemental Surveillance
E2215. The primary factors to be considered in developing an
Tests for Nuclear Power Reactor Vessels, E 706 (IH)
effective annealing program include the determination of the
E900 Guide for Predicting Radiation-Induced Transition
feasibility of annealing the specific reactor vessel; the avail-
Temperature Shift in Reactor Vessel Materials, E706 (IIF)
ability of the required information on vessel mechanical and
E1253 Guide for Reconstitution of Irradiated Charpy-Sized
fracture properties prior to annealing; evaluation of the par-
Specimens
ticular vessel materials, design, and operation to determine the
E2215 Practice for Evaluation of Surveillance Capsules
from Light-Water Moderated Nuclear Power Reactor Ves-
1
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear
sels
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, 2014. Published February 2014. Originally
3
approved in 1997. Last previous edition approved in 2008 as E509–03 (2008). DOI: For referenced ASTM standards, visit the ASTM website, www.astm.org, or
10.1520/E0509_E0509M-14. contact ASTM Customer Service at
...

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 − 03 (Reapproved 2008) E509/E509M − 14
Standard Guide for
In-Service Annealing of Light-Water Moderated Nuclear
1
Reactor Vessels
This standard is issued under the fixed designation E509;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 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,
2
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 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 inch-pound or either SI units or inch-pound units are to be regarded separately as the 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 and health practices and determine the applicability of regulatory
limitations prior to use.
2. Referenced Documents
3
2.1 ASTM Standards:
E185 Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
E636 Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)
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 July 1, 2008Jan. 1, 2014. Published September 2008February 2014. Originally approved in 1997. Last previous edition approved in 20032008
as E509–03. –03 (2008). DOI: 10.1520/E0509-0
...