ASTM E509-03(2008)
(Guide)Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels
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...
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Designation: E509 − 03(Reapproved 2008)
Standard Guide for
In-Service Annealing of Light-Water Moderated Nuclear
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
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
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
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
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 service@astm.org. For Annual Book of ASTM
The boldface numbers in parentheses refer to the list of references at the end of Standards volume information, refer to the standard’s Document Summary page on
this standard. the ASTM website.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E509 − 03 (2008)
2.2 ASME Standards: undesirable property effects such as permanent creep deforma-
Boiler and Pressure Vessel Code, Section III, Rules for tion or temper embrittlement. These higher temperatures also
Construction of Nuclear Power Plant Components can cause engineering difficulties, that is, core and coolant
Code Case N-557, In-Place Dry Annealing of a PWR removal and storage, localized heating effects, etc., in prevent-
Nuclear Reactor Vessel (Section XI, Division 1) ing the annealing operation from distorting the vessel or
damaging vessel supports, primary coolant piping, adjacent
2.3 Nuclear Regulatory Commission Documents:
concrete, insulation, etc. See ASME Code Case N-557 for
NRCRegulatoryGuide1.99,Revision2, EffectsofResidual
further guidance on annealing conditions and thermal-stress
Elements on Predicted Radiation Damage on Reactor
evaluations (2).
Vessel Materials
3.3.1 When a reactor vessel approaches a state of embrittle-
NRC Regulatory Guide 1.162, Format and Content of Re-
ment such that annealing is considered, the major criterion is
port for Thermal Annealing of Reactor Pressure Vessels
the number of years of additional service life that annealing of
3. Significance and Use
the vessel will provide. Two pieces of information are needed
to answer the question: the post-anneal adjusted RT and
3.1 Reactor vessels made of ferritic steels are designed with NDT
upper shelf energy level, and their subsequent changes during
the expectation of progressive changes in material properties
future irradiation. Furthermore, if a vessel is annealed, the
resulting from in-service neutron exposure. In the operation of
same information is needed as the basis for establishing
light-water-cooled nuclear power reactors, changes in
pressure-temperature limits for the period immediately follow-
pressure-temperature (P– T) limits are made periodically
ingtheannealanddemonstratingcompliancewithotherdesign
during service life to account for the effects of neutron
requirements and the PTS screening criteria. The effects on
radiation on the ductile-to-brittle transition temperature mate-
upper shelf toughness similarly must be addressed. This guide
rial properties. If the degree of neutron embrittlement becomes
primarily addresses RT changes. Handling of the upper
large, the restrictions on operation during normal heat-up and NDT
shelf is possible using a similar approach as indicated in NRC
cool down may become severe. Additional consideration
Regulatory Guide 1.162.Appendix X1 provides a bibliography
should be given to postulated events, such as pressurized
of existing literature for estimating annealing recovery and
thermal shock (PTS).Areduction in the upper shelf toughness
reembrittlement trends for these quantities as related to U.S.
also occurs from neutron exposure, and this decrease may
and other country pressure-vessel steels, with primary empha-
reduce the margin of safety against ductile fracture. When it
sis on U.S. steels.
appears that these situations could develop, certain alternatives
3.3.2 A key source of test material for determining the
are available that reduce the problem or postpone the time at
post-anneal RT , upper shelf energy level, and the reem-
which plant restrictions must be considered. One of these NDT
brittlement trend is the original surveillance program, provided
alternatives is to thermally anneal the reactor vessel beltline
it represents the critical materials in the reactor
region, that is, to heat the beltline region to a temperature
vessel. Appendix X2 describes an approach to estimate
sufficiently above the normal operating temperature to recover
changesin RT bothduetotheannealandafterreirradiation.
a significant portion of the original fracture toughness and NDT
The first purpose of Appendix X2 is to suggest ways to use
other material properties that were lost as a result of neutron
available materials most efficiently to determine the post-
embrittlement.
anneal RT and to predict the reembrittlement trend, yet
NDT
3.2 Preparationandplanningforanin-serviceannealshould
leave sufficient material for surveillance of the actual reem-
begin early so that pertinent information can be obtained to
brittlement for the remaining service life. The second purpose
guide the annealing operation. Sufficient time should be
is to describe alternative analysis approaches to be used to
allocated to evaluate the expected benefits in operating life to
assess test results of archive (or representative) materials to
be gained by annealing; to evaluate the annealing method to be
obtaintheessentialpost-annealandreirradiation RT ,upper
NDT
employed; to perform the necessary system studies and stress
shelf energy level, or fracture toughness, or a combination
evaluations; to evaluate the expected annealing recovery and
thereof.
reembrittlementbehavior;todevelopandfunctionallytestsuch
3.3.3 An evaluation must be conducted of the engineering
equipment as may be required to do the in-service annealing;
problems posed by annealing at the highest practical tempera-
and, to train personnel to perform the anneal.
ture. Factors required to be investigated to reduce the risk of
3.3 Selection of the annealing temperature requires a bal-
distortion and damage caused by mechanical and thermal
ance of opposing conditions. Higher annealing temperatures,
stresses at elevated temperatures to relevant system
and longer annealing times, can produce greater recovery of
components, structures, and control instrumentation are de-
fracture toughness and other material properties and thereby
scribed in 5.1.3 and 5.1.4.
increase the post-anneal lifetime. The annealing temperature
3.4 Throughout the annealing operation, accurate measure-
also can have an impact on the reembrittlement trend after the
ment of the annealing temperature at key defined locations
anneal.Ontheotherhand,highertemperaturescancreateother
must be made and recorded for later engineering evaluation.
Available from the American Society of Mechanical Engineers, 345 E. 47th
Street, New York, NY 10017. Consideration can be given to the reevaluation of broken Charpy specimens
Available from Superintendent of Documents, U.S. Government Printing from capsules withdrawn earlier which can be reconstituted using Guide E1253 or
Office, Washington, DC 20402. from material obtained (sampled) from the actual pressure-vessel wall.
E509 − 03 (2008)
3.5 After the annealing operation has been carried out, 4.1.2 The anticipated remaining operating lifetime of the
several steps should be taken. The predicted improvement in reactor vessel without annealing should be established using
fracture toughness properties must be verified, and it must be neutron embrittlement projections for the reactor vessel mate-
demonstrated that there is no damage to key components and rials.
structures.
4.1.2.1 A surveillance program conducted in accordance
with the requirements of Practices E185 and E2215 will
3.6 Further action may be required to demonstrate that
provide information from which to evaluate vessel condition.
reactor vessel integrity is maintained within ASME Code
Attention should be given to assuring that variations in the
requirements such as indicated in the referenced ASME Code
fluence-rate,neutronenergyspectrum,andirradiationtempera-
Case N-557 (2). Such action is beyond the scope of this guide.
ture for all different reactor neutron environments utilized are
taken into account.
4. General Considerations
4.1.2.2 Transition temperature and upper-shelf Charpy en-
4.1 Successful use of in-service annealing requires a thor-
ergy level data have been compiled and used to develop
ough knowledge of the irradiation behavior of the specific
correlationsof∆RT anduppershelfdropversusfluence,for
NDT
reactor-vessel materials, their annealing response and reirra-
example, Guide E900 or NRC Regulatory Guide 1.99, Revi-
diation embrittlement trend, the vessel design, fabrication
sion 2. These approaches, or other class-specific correlations,
history, and operating history. Some of these items may not be
should be used to estimate ∆RT and upper shelf energy
NDT
available for specific older vessels, and documented engineer-
drop for the specific heats of materials in the vessel beltline.
ing judgment may be required to conservatively estimate the
4.1.2.3 The results of surveillance specimen tests required
missing information.
by Practice E2215 should be compared to the data developed
4.1.1 To ascertain the design operating life-knowledge of
for 4.1.2.2 to ascertain whether the materials are performing in
the following items is needed: reactor vessel material
the manner expected. If not, an evaluation should be made to
composition, mechanical properties, fabrication techniques,
establish the extent of the remaining service life before
nondestructive test results, anticipated stress levels in the
restoration of properties is necessary.
vessel, neutron fluence, neutron energy spectrum, operating
4.1.3 Available data should be compiled for the annealing
temperature, and power history.
and post-anneal reirradiation responses of each class of
4.1.1.1 The initial RT as specified in subarticle NB-2300
NDT
material, and if available, for the specific heats of materials in
of the ASME Boiler and Pressure Vessel Code, Section III,
the vessel. The bibliography (3-78) in Appendix X1 provides
should be determined or estimated for those materials of
references for data compilation. Data collected should include
concern in the high fluence regions of the reactor pressure
transition temperature shift and upper shelf Charpy energy
vessel. Alternative methods for the determination of RT
NDT
changes. Actual fracture toughness da
...
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
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
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
Materials E706 (1B)
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)
E 900Guide for Predicting Neutron Radiation Damage to Reactor Vessel Materials E706 (IIF) Guide for Predicting
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.
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, 2008. Published September 2008. Originally approved in 1997. Last previous edition approved in 2003 as E 509–03.
The boldface numbers in parentheses refer to the list of references at the end of this standard.
For referencedASTM standards, visit theASTM website, www.astm.org, or contactASTM Customer Service at service@astm.org. For Annual Book of ASTM Standards
, Vol 12.02.volume information, refer to the standard’s Document Summary page on the ASTM website.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.
E509–03 (2008)
Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)
E 1253Guide for Reconstitution of Irradiated Charpy Specimens Guide for Reconstitution of Irradiated Charpy-Sized
Specimens
E 2215 Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
2.2 ASME Standards:
Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Power Plant Components
Code Case N-557, In-Place Dry Annealing of a PWR Nuclear Reactor Vessel (Section XI, Division 1)
2.3 Nuclear Regulatory Commission Documents:
NRC Regulatory Guide 1.99, Revision 2, Effects of Residual Elements on Predicted Radiation Damage on Reactor Vessel
Materials
NRC Regulatory Guide 1.162, Format and Content of Report for Thermal Annealing of Reactor Pressure Vessels
3. 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).Areduction 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.
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 annealing treatment. anneal.
3.3 Selectionoftheannealingtemperaturerequiresabalanceofopposingconditions.Higherannealingtemperatures,andlonger
annealing times, can produce greater recovery of fracture toughness and other material properties and thereby increase the
post-anneallifetime.Theannealingtemperaturealsocanhaveanimpactonthereembrittlementtrendaftertheanneal.Ontheother
hand, higher temperatures can create other undesirable property effects such as permanent creep deformation or temper
embrittlement. These higher temperatures also can cause engineering difficulties, that is, core and coolant removal and storage,
localized heating effects, etc., in preventing the annealing operation from distorting the vessel or damaging vessel supports,
primary coolant piping, adjacent concrete, insulation, etc. See ASME Code Case N-557 for further guidance on annealing
conditions and thermal-stress evaluations (2) .
3.3.1 When a reactor vessel approaches a state of embrittlement such that annealing is considered, the major criterion is the
numberofyearsofadditionalservicelifethatannealingofthevesselwillprovide.Twopiecesofinformationareneededtoanswer
the question: the post-anneal adjusted RT and upper shelf energy level, and their subsequent changes during future irradiation.
NDT
Furthermore, if a vessel is annealed, the same information is needed as the basis for establishing pressure-temperature limits for
the period immediately following the anneal and demonstrating compliance with other design requirements and the PTS screening
criteria. The effects on upper shelf toughness similarly must be addressed. This guide primarily addresses RT changes.
NDT
Handling of the upper shelf is possible using a similar approach as indicated in NRC Regulatory Guide 1.162. Appendix X1
provides a bibliography of existing literature for estimating annealing recovery and reembrittlement trends for these quantities as
related to U.S. and other country pressure-vessel steels, with primary emphasis on U.S. steels.
3.3.2 A key source of test material for determining the post-anneal RT , upper shelf energy level, and the reembrittlement
NDT
trendistheoriginalsurveillanceprogram,provideditrepresentsthecriticalmaterialsinthereactorvessel. AppendixX2describes
an approach to estimate changes in RT both due to the anneal and after reirradiation. The first purpose of Appendix X2 is to
NDT
suggest ways to use available materials most efficiently to determine the post-anneal RT and to predict the reembrittlement
NDT
trend, yet leave sufficient material for surveillance of the actual reembrittlement for the remaining service life.The second purpose
is to describe alternative analysis approaches to be used to assess test results of archive (or similar)representative) materials to
obtain the essential post-anneal and reirradiation RT , upper shelf energy level, or fracture toughness, or a combination thereof.
NDT
Available from the American Society of Mechanical Engineers, 345 E. 47th Street, New York, NY 10017.
Available from Superintendent of Documents, U.S. Government Printing Office, Washington, DC 20402.
Consideration can be given to the reevaluation of broken Charpy specimens from capsules withdrawn earlier which can be reconstituted using Guide E 1253 or from
material obtained (sampled) from the actual pressure-vessel wall.
E509–03 (2008)
3.3.3 An evaluation must be conducted of the engineering problems posed by annealing at the highest practical temperature.
Factors required to be investigated to reduce the risk of distortion and damage caused by mechanical and thermal stresses at
elevated temperatures to relevant system components, structures, and control instrumentation are described in 5.1.3 and 5.1.4.
3.4 Throughout the annealing operation, accurate measurement of the annealing temperature at key defined locations must be
made and recorded for later engineering evaluation.
3.5 After the annealing operation has been carried out, several steps should be taken. The predicted improvement in fracture
toughness properties must be verified, and it must be demonstrated that there is no damage to key components and structures.
3.6 Further action may be required to demonstrate that reactor vessel integrity is maintained withinASME Code requirements
such as indicated in the referenced ASME Code Case N-557 (2). Such action is beyond the scope of this guide.
4. General Considerations
4.1 Successful use of in-service annealing requires a thorough knowledge of the irradiation behavior of the specific
reactor-vessel materials, their annealing response and reirradiation embrittlement trend, the vessel design, fabrication history, and
operating history. Some of these items may not be available for specific older vessels, and documented engineering judgment may
be required to conservatively estimate the missing information.
4.1.1 To ascertain the design operating life-knowledge of the following items is needed: reactor vessel material composition,
mechanical properties, fabrication techniques, nondestructive test results, anticipated stress levels in the vessel, neutron fluence,
neutron energy spectrum, operating temperature, and power history.
4.1.1.1 The initial RT as specified in subarticle NB-2300 of theASME Boiler and Pressure Vessel Code, Section III, should
NDT
be determined or estimated for those materials of concern in the high fluence regions of the reactor pressure vessel. Alternative
methods for the determination of RT also may be used. Consideration should be given to the technical justification for alternate
NDT
methodologies and the data, which form the basis for the RT determination. Initial RT values should be available or
NDT NDT
estimated for all materials located in these areas.
4.1.1.2 The initial Charpy upper shelf energy as defined by Practices E 185 and E 2215 should be determined for materials of
concern in the beltline region of the reactor pressure vessel. Initial upper shelf energy levels should be available or estimated for
all materials located in this area.
4.1.1.3 Unirradiated archive heats of reactor vessel beltline materials should be maintained for preparation of additional
surveillance samples as required by Practices E 185 and E 2215. Previously teste
...
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