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ASTM E 509 Document Information:
Title
Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels
ASTM International
Publication Date:
Mar 10, 2003
Scope:
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).
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.
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-NDT), 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.
The values stated in inch-pound or SI units are to be regarded
separately as the standard.
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.
Keywords:
- fracture toughness
- irradiation
- nuclear reactor vessels (light-water moderated)
- radiation exposure
- surveillance (of nuclear reactor vessels)
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