ASTM E509-97

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-cooled nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service heat treatment is to improve the mechanical properties (especially fracture toughness) of the reactor vessel materials previously degraded by neutron embrittlement. The measure of the improvement in mechanical properties is generally assessed using Charpy V-notch impact test results or, alternatively, fracture toughness results.  
1.2 This guide is designed to accommodate the variable response of reactor-vessel materials to post-irradiation heat treatment at various temperatures and different time periods. This guide describes certain inherent limiting factors which 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, and ( ) the mechanical and thermal stresses in the components and the system as a whole.  
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 designed based on the requirements and guidance described in Practice E185. The primary factors to be considered in developing an effective annealing procedure include: ( ) the determination of the feasibility of annealing the specific reactor vessel, ( ) the availability of the required information on vessel mechanical properties and fracture toughness properties prior to annealing, ( ) evaluation of the particular plant to determine the practical annealing temperature, and ( ) the procedure to be used for verification of the degree of recovery. Guidelines are provided to determine the post-anneal reference temperature ( RTNDT ), the Charpy V-notch upper shelf energy level, fracture toughness properties, and the predicted re-embrittlement rate 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 re-embrittlement data is to be available to assess life extension of individual reactor pressure vessels.  
1.4 The values stated in inch-pound units are to be regarded as the standard.  
1.5 This standard may involve hazardous materials, operations, and equipment. This standard does not purport to address all of the safety problems 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-Jun-1997
• 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

• Replaced By: ASTM E509-03 - Standard Guide for In-Service Annealing of Light-Water Cooled Nuclear Reactor Vessels
• 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-Jun-1997
• Referred By: ASTM E2956-23 - Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels
• Effective Date: 10-Jun-1997
• Referred By: ASTM E853-23 - Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results
• Effective Date: 10-Jun-1997

Standards Content (Sample)

NOTICE: This standard has either been superceded and replaced by a new version or discontinued.
Contact ASTM International (www.astm.org) for the latest information.
Designation: E 509 – 97
Standard Guide for
In-Service Annealing of Light-Water Cooled 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 (e) indicates an editorial change since the last revision or reapproval.
1. Scope ), the Charpy V-notch upper shelf energy level, fracture
NDT
toughness properties, and the predicted reembrittlement trend
1.1 This guide covers the general procedures to be consid-
for these properties for reactor vessel beltline materials. This
ered for conducting an in-service thermal anneal of a light-
guide emphasizes the need to plan well ahead in anticipation of
water-cooled nuclear reactor vessel and demonstrating the
annealing if an optimum amount of post-anneal reembrittle-
effectiveness of the procedure. The purpose of this in-service
ment data is to be available for use in assessing the ability of
heat treatment is to improve the mechanical properties, espe-
a nuclear reactor vessel to operate for the duration of its present
cially fracture toughness, of the reactor vessel materials previ-
license, or qualify for a license extension, or both.
ously degraded by neutron embrittlement. The improvement in
1.4 The values stated in inch-pound or SI units are to be
mechanical properties generally is assessed using Charpy
regarded separately as the standard.
V-notch impact test results, or alternatively, fracture toughness
1.5 This standard does not purport to address all of the
test results or inferred toughness property changes from tensile,
safety concerns, if any, associated with its use. It is the
hardness, indentation, or other miniature specimen testing (1).
responsibility of the user of this standard to establish appro-
1.2 This guide is designed to accommodate the variable
priate safety and health practices and determine the applica-
response of reactor-vessel materials in post-irradiation heat
bility of regulatory limitations prior to use.
treatment at various temperatures and different time periods.
Certain inherent limiting factors must be considered in devel-
2. Referenced Documents
oping an annealing procedure. These factors include system-
2.1 ASTM Standards:
design limitations; physical constraints resulting from attached
E 184 Practice for Effects of High-Energy Neutron Radia-
piping, support structures, and the primary system shielding;
tion on the Mechanical Properties of Metallic Materials
the mechanical and thermal stresses in the components and the
E 706 (1B)
system as a whole; and, material condition changes that may
E 185 Practice for Conducting Surveillance Tests for Light-
limit the annealing temperature.
Water Cooled Nuclear Power Reactor Vessels E 706 (IF)
1.3 This guide provides direction for development of the
E 636 Practice for Conducting Supplemental Surveillance
vessel annealing procedure and a post-annealing vessel radia-
Tests for Nuclear Power Reactor Vessels E 706 (IH)
tion surveillance program. The development of a surveillance
E 900 Guide for Predicting Neutron Radiation Damage to
program to monitor the effects of subsequent irradiation of the
Reactor Vessel Materials E 706 (IIF)
annealed-vessel beltline materials should be based on the
E 1253 Guide for Reconstitution of Irradiated Charpy
requirements and guidance described in Practice E 185. The
Specimens
primary factors to be considered in developing an effective
2.2 ASME Standards:
annealing program include the determination of the feasibility
Boiler and Pressure Vessel Code, Section III, Rules for
of annealing the specific reactor vessel; the availability of the
Construction of Nuclear Power Plant Components
required information on vessel mechanical and fracture prop-
Code Case N-557, In-Place Dry Annealing of a PWR
erties prior to annealing; evaluation of the particular vessel
Nuclear Reactor Vessel (Section XI, Division 1)
materials, design, and operation to determine the annealing
2.3 Nuclear Regulatory Commission Documents:
time and temperature; and, the procedure to be used for
NRC Regulatory Guide 1.99, Revision 2, Effects of Re-
verification of the degree of recovery and the trend for
sidual Elements on Predicted Radiation Damage on Re-
reembrittlement. Guidelines are provided to determine the
actor Vessel Materials
post-anneal reference nil-ductility transition temperature (RT-
NRC Regulatory Guide 1.162, Format and Content of
This guide is under the jurisdiction of ASTM Committee E-10 on Nuclear
Technology and Applications and is the direct responsibility of Subcommittee Annual Book of ASTM Standards, Vol 12.02.
E10.02 on Behavior and Use of Metallic Materials in Nuclear Systems. Available from the American Society of Mechanical Engineers, 345 E. 47th
Current edition approved June 10, 1997. Published May 1998. Street, New York, NY 10017.
2 5
The boldface numbers in parentheses refer to the list of references at the end of Available from Superintendent of Documents, U.S. Government Printing
this standard. Office, Washington, DC 20402.
Copyright © ASTM, 100 Barr Harbor Drive, West Conshohocken, PA 19428-2959, United States.
NOTICE: This standard has either been superceded and replaced by a new version or discontinued.
Contact ASTM International (www.astm.org) for the latest information.
E 509
Report for Thermal Annealing of Reactor Pressure Ves- future irradiation. Furthermore, if a vessel is annealed, the
sels same information is needed as the basis for establishing
pressure-temperature limits for the period immediately follow-
3. Significance and Use
ing the anneal and demonstrating compliance with other design
3.1 Reactor vessels made of ferritic steels are designed with
requirements and the PTS screening criteria. The effects on
the expectation of progressive changes in material properties
upper shelf toughness similarly must be addressed. This guide
resulting from in-service neutron exposure. In the operation of
primarily addresses RT changes. Handling of the upper
NDT
light-water-cooled nuclear power reactors, changes in
shelf is possible using a similar approach as indicated in NRC
pressure-temperature (P– T) limits are made periodically
Regulatory Guide 1.162. Appendix X1 provides a bibliography
during service life to account for the effects of neutron
of existing literature for estimating annealing recovery and
radiation on the ductile-to-brittle transition temperature mate-
reembrittlement trends for these quantities as related to U.S.
rial properties. If the degree of neutron embrittlement becomes
and other country pressure-vessel steels, with primary empha-
large, the restrictions on operation during normal heat-up and
sis on U.S. steels.
cool down may become severe. Additional consideration
3.3.2 A key source of test material for determining the
should be given to postulated events, such as pressurized
post-anneal RT , upper shelf energy level, and the reem-
NDT
thermal shock (PTS). A reduction in the upper shelf toughness
brittlement trend is the original surveillance program, provided
also occurs from neutron exposure, and this decrease may
it represents the critical materials in the reactor vessel.
reduce the margin of safety against ductile fracture. When it
Appendix X2 describes an approach to estimate changes in
appears that these situations could develop, certain alternatives
RT both due to the anneal and after reirradiation. The first
NDT
are available that reduce the problem or postpone the time at
purpose of Appendix X2 is to suggest ways to use available
which plant restrictions must be considered. One of these
materials most efficiently to determine the post-anneal RT
NDT
alternatives is to thermally anneal the reactor vessel beltline
and to predict the reembrittlement trend, yet leave sufficient
region, that is, to heat the beltline region to a temperature
material for surveillance of the actual reembrittlement for the
sufficiently above the normal operating temperature to recover
remaining service life. The second purpose is to describe
a significant portion of the original fracture toughness and
alternative analysis approaches to be used to assess test results
other material properties that were lost as a result of neutron
of archive (or similar) materials to obtain the essential post-
embrittlement.
anneal and reirradiation RT , upper shelf energy level, or
NDT
3.2 Preparation and planning for an in-service anneal should
fracture toughness, or a combination thereof.
begin early so that pertinent information can be obtained to
3.3.3 An evaluation must be conducted of the engineering
guide the annealing operation. Sufficient time should be
problems posed by annealing at the highest practical tempera-
allocated to evaluate the expected benefits in operating life to
ture. Factors required to be investigated to reduce the risk of
be gained by annealing; to evaluate the annealing method to be
distortion and damage caused by mechanical and thermal
employed; to perform the necessary system studies and stress
stresses at elevated temperatures to relevant system compo-
evaluations; to evaluate the expected annealing recovery and
nents, structures, and control instrumentation are described in
reembrittlement behavior; to develop such equipment as may
5.1.3 and 5.1.4.
be required to do the in-service annealing; and, to train
3.4 Throughout the annealing operation, accurate measure-
personnel to perform the annealing treatment.
ment of the annealing temperature at key defined locations
3.3 Selection of the annealing temperature requires a bal-
must be made and recorded for later engineering evaluation.
ance of opposing conditions. Higher annealing temperatures,
3.5 After the annealing operation has been carried out,
and longer annealing times, can produce greater recovery of
several steps should be taken. The predicted improvement in
fracture toughness and other material properties and thereby
fracture toughness properties must be verified, and it must be
increase the post-anneal lifetime. The annealing temperature
demonstrated that there is no damage to key components and
also can have an impact on the reembrittlement trend after the
structures.
anneal. On the other hand, higher temperatures can create other
3.6 Further action may be required to demonstrate that
undesirable property effects such as permanent creep deforma-
reactor vessel integrity is maintained within ASME Code
tion or temper embrittlement. These higher temperatures also
requirements such as indicated in the referenced ASME Code
can cause engineering difficulties, that is, core and coolant
Case N-557 (2). Such action is beyond the scope of this guide.
removal and storage, localized heating effects, etc., in prevent-
ing the annealing operation from distorting the vessel or 4. General Considerations
damaging vessel supports, primary coolant piping, adjacent
4.1 Successful use of in-service annealing requires a thor-
concrete, insulation, etc. See ASME Code Case N-557 for
ough knowledge of the irradiation behavior of the specific
further guidance on annealing conditions and thermal-stress
reactor-vessel materials, their annealing response and reirra-
evaluations (2).
diation embrittlement trend, the vessel design, fabrication
3.3.1 When a reactor vessel approaches a state of embrittle-
history, and operating history. Some of these items may not be
ment such that annealing is considered, the major criterion is
the number of years of additional service life that annealing of
the vessel will provide. Two pieces of information are needed
Consideration can be given to the reevaluation of broken Charpy specimens
to answer the question: the post-anneal adjusted RT and
NDT from capsules withdrawn earlier which can be reconstituted using Guide E 1253 or
upper shelf energy level, and their subsequent changes during from material obtained (sampled) from the actual pressure-vessel wall.
NOTICE: This standard has either been superceded and replaced by a new version or discontinued.
Contact ASTM International (www.astm.org) for the latest information.
E 509
available for specific older vessels, and documented engineer- 4.1.2.3 The results of surveillance specimen tests required
ing judgment may be required to conservatively estimate the by Practice E 185 should be compared to the data developed
missing information. for 4.1.2.2 to ascertain whether the materials are performing in
4.1.1 To ascertain the design operating life-knowledge of the manner expected. If not, an evaluation should be made to
the following items is needed: reactor vessel material compo- establish the extent of the remaining service life before
sition, mechanical properties, fabrication techniques, nonde- restoration of properties is necessary.
structive test results, anticipated stress levels in the vessel, 4.1.3 Available data should be compiled for the annealing
neutron fluence, neutron energy spectrum, operating tempera- and post-anneal reirradiation responses of each class of mate-
ture, and power history. rial, and if available, for the specific heats of materials in the
4.1.1.1 The initial RT as specified in subarticle NB-2300 vessel. The bibliography (3-67) in Appendix X1 provides
NDT
of the ASME Boiler and Pressure Vessel Code, Section III, references for data compilation. Data collected should include
should be determined or estimated for those materials of transition temperature shift and upper shelf Charpy energy
concern in the high fluence regions of the reactor pressure changes. Actual fracture toughness data also should be com-
vessel. Alternative methods for the determination of RT piled, as well as other supplemental information or data such as
NDT
also may be used. Consideration should be given to the instrumented Charpy, indentation/hardness, tensile, and other
technical justification for alternate methodologies and the data, miniature specimen test results (see Practice E 636 for addi-
which form the basis for the RT determination. Initial tional testing that can be utilized in assessing annealing
NDT
RT values should be available or estimated for all materials behavior). The extent of the increased service life after
NDT
located in these areas. annealing should be estimated using the guidance provided in
4.1.1.2 The initial Charpy upper shelf energy as defined by Appendix X2.
Practice E 185 should be determined for materials of concern 4.1.4 Irradiated material from the vessel surveillance pro-
in the beltline region of the reactor pressure vessel. Initial gram should be retain
...