ASTM E509-03

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

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 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 (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. 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.

General Information

• Status: Historical
• Publication Date: 09-Mar-2003
• 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 E509-97 - Standard Guide for In-Service Annealing of Light-Water Cooled Nuclear Reactor Vessels
• Effective Date: 10-Mar-2003
• Replaced By: ASTM E509-03(2008) - Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear 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: 10-Mar-2003
• Referred By: ASTM E2215-19 - Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 10-Mar-2003
• Referred By: ASTM E2956-23 - Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels
• Effective Date: 10-Mar-2003

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
Standard Guide for
In-Service Annealing of Light-Water Moderated Nuclear
1
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 (e) 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 to be consid-
reembrittlement. Guidelines are provided to determine the
ered for conducting an in-service thermal anneal of a light-
post-anneal reference nil-ductility transition temperature (RT-
water moderated nuclear reactor vessel and demonstrating the
), the Charpy V-notch upper shelf energy level, fracture
NDT
effectiveness of the procedure. The purpose of this in-service
toughness properties, and the predicted reembrittlement trend
annealing (heat treatment) is to improve the mechanical
for these properties for reactor vessel beltline materials. This
properties, especially fracture toughness, of the reactor vessel
guideemphasizestheneedtoplanwellaheadinanticipationof
materials previously degraded by neutron embrittlement. The
annealing if an optimum amount of post-anneal reembrittle-
improvement in mechanical properties generally is assessed
ment data is to be available for use in assessing the ability of
using Charpy V-notch impact test results, or alternatively,
anuclearreactorvesseltooperateforthedurationofitspresent
fracture toughness test results or inferred toughness property
license, or qualify for a license extension, or both.
changes from tensile, hardness, indentation, or other miniature
2
1.4 The values stated in inch-pound or SI units are to be
specimen testing (1).
regarded separately as the standard.
1.2 This guide is designed to accommodate the variable
1.5 This standard does not purport to address all of the
response of reactor-vessel materials in post-irradiation anneal-
safety concerns, if any, associated with its use. It is the
ing at various temperatures and different time periods. Certain
responsibility of the user of this standard to establish appro-
inherent limiting factors must be considered in developing an
priate safety and health practices and determine the applica-
annealing procedure. These factors include system-design
bility of regulatory limitations prior to use.
limitations;physicalconstraintsresultingfromattachedpiping,
support structures, and the primary system shielding; the
2. Referenced Documents
mechanical and thermal stresses in the components and the
2.1 ASTM Standards:
system as a whole; and, material condition changes that may
E 185 Practice for Design of Surveillance Programs Tests
limit the annealing temperature.
for Light-Water Moderated Nuclear Power Reactor Ves-
1.3 This guide provides direction for development of the
3
sels
vessel annealing procedure and a post-annealing vessel radia-
E 636 Practice for Conducting Supplemental Surveillance
tion surveillance program. The development of a surveillance
3
Tests for Nuclear Power Reactor Vessels E 706 (IH)
program to monitor the effects of subsequent irradiation of the
E 900 Guide for Predicting Radiation-Induced Transition
annealed-vessel beltline materials should be based on the
3
Temperature Shift in ReactorVessel Materials E 706 (IIF)
requirements and guidance described in Practices E 185 and
E 1253 Guide for Reconstitution of Irradiated Charpy
E 2215. The primary factors to be considered in developing an
3
Specimens
effective annealing program include the determination of the
E 2215 Practice for the Evaluation of Surveillance Capsules
feasibility of annealing the specific reactor vessel; the avail-
3
from Light-Water Moderated Nuclear Reactor Vessels
ability of the required information on vessel mechanical and
2.2 ASME Standards:
fracture properties prior to annealing; evaluation of the par-
Boiler and Pressure Vessel Code, Section III, Rules for
ticular vessel materials, design, and operation to determine the
4
Construction of Nuclear Power Plant Components
Code Case N-557, In-Place Dry Annealing of a PWR
4
1
Nuclear Reactor Vessel (Section XI, Division 1)
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear
Technology and Applications and is the direct responsibility of Subcommittee
2.3 Nuclear Regulatory Commission Documents:
E10.02 on Behavior and Use of Metallic Materials in Nuclear Systems.
Current edition approved March 10, 2003. Published May 2003. Originally
3
published as E 509–97. Last previous edition E 509–97. Annual Book of ASTM Standards, Vol 12.02.
2 4
The boldface numbers
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