ASTM E509/E509M-21

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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/E509M − 14 E509/E509M − 21
Standard Guide for
In-Service Annealing of Light-Water Moderated Nuclear
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
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).
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,
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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 either SI units or inch-pound units are to be regarded separately as standard. The values stated in each
system are not necessarily exact equivalents; therefore, to ensure conformance with the standard, each system shall be used
independently of the other, and values from the two systems shall not be combined.
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 Jan. 1, 2014Feb. 1, 2021. Published February 2014February 2021. Originally approved in 1997. Last previous edition approved in 20082014
as E509–03 (2008). –14. DOI: 10.1520/E0509_E0509M-14.10.1520/E0509_E0509M-21.
The boldface numbers in parentheses refer to the list of references at the end of this standard.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E509/E509M − 21
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 safety, health, and healthenvironmental practices and determine the
applicability of regulatory limitations prior to use.
1.6 This international standard was developed in accordance with internationally recognized principles on standardization
established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued
by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
2. Referenced Documents
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
E900 Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
E1253 Guide for Reconstitution of Irradiated Charpy-Sized Specimens
E1820 Test Method for Measurement of Fracture Toughness
E1921 Test Method for Determination of Reference Temperature, T , for Ferritic Steels in the Transition Range
o
E2215 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
Boiler and Pressure Vessel Code, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components
Code Case N-557, In-Place Dry Annealing of a PWR Nuclear Reactor Vessel (Section XI, Division 1)
Code Case N-830, Direct Use of Master Curve Fracture Toughness Curve Pressure-Retaining Materials of Class 1 Vessels
(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). 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 temperature requires a balance of opposing conditions. Higher annealing temperatures, and longer
annealing times, can produce greater recovery of fracture toughness and other material properties and thereby increase the
post-anneal lifetime. The annealing temperature also can have an impact on the reembrittlement trend after the anneal. On the other
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,
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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
number of years of additional service life that annealing of the vessel will provide. Two pieces of information are needed to answer
the question: the post-anneal adjusted RT and upper shelf energy level, and their subsequent changes during future irradiation.
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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. Handling
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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 trend
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is the original surveillance program, provided it represents the critical materials in the reactor vessel. Appendix X2 describes an
approach to estimate changes in RT both due to the anneal and reirradiation. The first purpose of Appendix X2 is to suggest
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ways to use available materials most efficiently to determine the post-anneal RT and to predict the reembrittlement trend, yet
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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 representative) materials to obtain the
essential post-anneal and reirradiation RT , upper shelf energy level, or fracture toughness, or a combination thereof.
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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 within ASME 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 the ASME Boiler and Pressure Vessel Code, Section III, should
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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. used as allowed in the ASME Boiler and Pressure Vessel Code, Sections
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III and XI. Consideration should be given to the technical justification for alternate methodologies and the data, which form the
basis for the RT determination. Initial RT values should be available or estimated for all materials located in these areas.
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4.1.1.2 The initial Charpy upper shelf energy as defined by Practices E185 and E2215 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.
Consideration can be given to the reevaluation of broken Charpy specimens from capsules withdrawn earlier which can be reconstituted using Guide E1253 or from
material obtained (sampled) from the actual pressure-vessel wall.
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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 E185 and E2215. Previously tested specimens should be retained as an additional
source of material.
4.1.1.4 A record of the actual fabrication history, including heat treatment and welding procedure, of the materials in the beltline
region of the vessel should be maintained.
4.1.1.5 The chemical composition should be determined for base metal(s) and deposited weld metal(s) and should include all
elements potentially relevant to irradiation, annealing, and reirradiation behavior, for example, copper, nickel, phosphorus,
manganese and sulfur. The variability in chemical composition should be determined when possible.
4.1.2 The anticipated remaining operating lifetime of the reactor vessel without annealing should be established using neutron
embrittlement projections for the reactor vessel materials.
4.1.2.1 A surveillance program conducted in accordance with the requirements of Practices E185 and E2215 will provide
information from which to evaluate vessel condition. Attention should be given to assuring that variations in the fluence-rate,
neutron energy spectrum, and irradiation temperature for all different reactor neutron environments utilized are taken into account.
4.1.2.2 Transition temperature and upper-shelf Charpy energy data have been compiled and used to develop correlations of
ΔRT and upper shelf drop versus fluence, for example, Guide E900 or NRC Regulatory Guide 1.99, Revision 2. These
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approaches, or other class-specific correlations, should be used to estimate ΔRT and upper shelf energy drop for the specific
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heats of materials in the vessel beltline.
4.1.2.3 The results of surveillance specimen tests required by Practice E2215 should be compared to the data developed in 4.1.2.2
to ascertain whether the materials are performing as expected. If not, an evaluation should be made to establish the extent of the
remaining service life before restoration of properties is necessary.
4.1.3 Available data should be compiled for the annealing and post-anneal reirradiation responses of each class of material, and
if available, for the specific heats of materials in the vessel. The bibliography (3-7885) in Appendix X1 provides references for
data compilation. Data collected should include transition temperature shifts and upper shelf Charpy energy changes. Actual
fracture toughness data representing reference temperature, T , following Test Method E1921 and upper shelf fracture toughness
following Test Method E1820 also should be compiled, as well as other supplemental information or data such as instrumented
Charpy, indentation/hardness, tensile, and other miniature specimen test results (see Practice E636 for additional testing that can
be utilized in assessing annealing behavior). Use of ASME Code Case N-830 in defining the transition temperature fracture
toughness of vessel materials is allowed when T is determined following Test Method E1921. The extent of the increased service
life after annealing should be estimated using the guidance provided in Appendix X2.
4.1.4 Irradiated material from the vessel surveillance program should be retained as a source of material for future vessel condition
assessments.
5. Annealing Method
5.1 The annealing method selected should consider the magnitude of the recovery needed to extend the lifetime, the predicted
annealing response, the reirradiation response, the accessibility of the reactor vessel to allow inspection and temperature
monitoring, the constraints resulting from the design of the reactor, and the structural relationship of the reactor vessel to the
primary system and supports. A detailed annealing procedure should be prepared, for example, see ASME Code Case N-577(2)
and NRC Regulatory Guide 1.162. This written procedure should include all quality assurance measures and training to be
conducted to assure an effective annealing operation.
5.1.1 The annealing method employed must not degrade the original design of the system. The parameters for a dry anneal may
exceed the original design limits of the reactor vessel. In this case, the primary coolant water has been removed and a heating
device is employed to raise the vessel temperature locally in the affected beltline region above the original design temperature.
ASME Code Case N-557 (2) provides a framework for assuring design conformance for an in-service thermal anneal in air. A lower
temperature wet anneal, in which the heating medium is the primary coolant water, should not exceed the original design pressure
and temperature for the reactor vessel.
Consideration should be given to the possibility of thermal embrittlement of beltline materials, including heat-affected-zones, as a result of the annealing heat-treatment.
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5.1.2 A review of all reactor components likely to be impacted by the anneal should be completed prior to the initiation of the
anneal.
5.1.3 Consideration should be given to the effects of mechanical and thermal stresses and temperature on all system components,
structures, and control instrumentation. Specific material properties should be justified by the analyst evaluating these effects.
Examples of such effects are as follows:
5.1.3.1 Changes in the properties of friction reducing materials in sliding or articulating connections.
5.1.3.2 Reduction in neutron and gamma absorption capacity of supplementary shielding materials.
5.1.3.3 Effect of thermal growth on closely machined articulated or sliding interfaces.
5.1.3.4 Changes in mechanical and thermal properties of the reactor vessel insulation.
5.1.3.5 Effect of elevated temperatures on low melting point alloys, if applicable.
5.1.4 A detailed thermal and stress evaluation should be performed to demonstrate that localized temperatures, thermal stresses,
and subsequent residual stresses are acceptable. This evaluation will help to establish the heating system design and
heat-up/cool-down rates for the anneal procedure.
5.1.4.1 Vessel distortion should be considered both analytically and physically. Measurement of dimensions prior to and after
annealing should be considered to assess dimensional stability.
5.1.4.2 Adequate analytical estimation and actual measurement of concrete temperatures in the region near the reactor vessel are
needed to avoid concrete degradation. The properties of the concrete should be known or estimated in order to demonstrate that
no damage will occur during the annealing.
5.1.5 The annealing method selected must assure adequate recovery of the reactor vessel materials. An experimental program may
be undertaken prior to the in-service anneal to establish the degree of material properties recovery for the specific materials in the
beltline of the vessel (see Appendix X2). This program shall use materials that are representative of reactor vessel materials in
accordance with the criteria set forth in Practice E185 for material selection and irradiation conditions. For example, the program
may use existing broken irradiated Charpy halves from the current surveillance program that have been reconstituted following
Guide E1253, or samples taken from the actual pressure vessel. Other miniature or small specimen testing techniques also can be
considered if properly validated. The program also may assess the adequacy of selected heat treatment conditions for achieving
the minimum required recovery. The results from the experimental program should be compared with the data compiled for 4.1.3.
Data generated relative to the actual vessel neutron exposure should be reviewed in relation to temperature, fluence and fluence-rate
effects.
5.1.6 The annealing procedure employed should provide for adequate instrumentation to control and monitor the temperature of
the vessel such that a complete temperature record is available throughout all phases of the annealing operation. Special
consideration should be given to axial, azimuthal, and through-wall thermal gradients in the beltline region and any regions
anticipated to experience high stresses during the anneal, such as the nozzles.
5.1.7 The annealing procedure should include a description of the annealing equipment, an outline of the operational requirements,
and integration of pre-annealing test of the heating equipment. Consideration should be given to storage of the core, internals, and
coolant.
5.1.8 Special precautions to assure the protection of plant personnel and the general public from any release of radioactive
materials should be provided. The annealing operation also should give adequate consideration to the radiation exposure of
personnel, as well as any radioactive waste processing, radioactive-material decontamination, and radioactive-waste shipment.
5.2 The annealing process must be carefully monitored to assure that the conditions outlined in the annealing procedure described
Follow American Concrete Institute guidelines as appropriate. Additional guidance may be available from U.S. annealing demonstration programs.
U.S. annealing demonstrations provide further insight into the degree of instrumentation needed to adequately monitor and control the annealing operation.
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in 5.1 are maintained. The temperature of the reactor vessel must be monitored to assure that the annealing operating conditions
are maintained and to demonstrate that temperature gradients are consistent with the thermal and stress analyses.
6. Annealing Surveillance and Verification
6.1 The effectiveness of the anneal depends upon the degree of property recovery and the reembrittlement trend. The surveillance
specimens, as described in Practice E185, provide a means of assessing the degree of properties recovered from an anneal.
6.1.1 Guidelines for assessing annealing recovery from available materials are given in Appendix X2. A surveillance program
must be established after the anneal to monitor reirradiation embrittlement. Appendix X2 also contains guidelines for such a
surveillance program.
6.1.2 If sufficient materials are not available or if conditions dictate that the approach in 6.1.1 is inapplicable, an alternative
program for demonstrating the effectiveness of the in-service anneal and for monitoring the reirradiation response of the vessel
materials should be established. Appendix X2 again contains guidelines that can be followed. The bibliography (3-7885) given in
Appendix X1 also will be valuable in establishing an alternative program.
7. Documentation
7.1 A description and analysis of the annealing procedures, results, and supporting data should be prepared, for example, see
ASME Code Case N-557 (2) and NRC Regulatory Guide 1.162. This documentation should include, but not be limited to, the
following information and data:
7.1.1 A description should be provided of all data and analyses used to support the justification for performing the anneal. This
should include all irradiation analyses or test program results, as well as all special calculations, related stress analyses, and heating
evaluations.
7.1.2 A description of all materials used in the establishment of the annealing process and the monitoring of the actual annealing
operation should be included. This section should include the reporting requirements of Practices E185 and E2215.
7.1.3 A detailed description of the proposed annealing procedure and a chronology of the proposed versus actual procedure for
the annealing operation should be documented. Special emphasis is to be given to the location of temperature monitors and their
records.
7.1.4 A detailed evaluation of the results of the annealing operation with appropriate technical justification should be reported. Any
limitations regarding material property recovery or future plant operation should be described and documented.
7.1.5 Applicable ASME codes, ASTM standards and guides, NRC regulations and guides, and other technical references should
be provided. All appropriate regulations and standards should be addressed as to the extent to which they were met.
7.1.6 Specific details of the planned new surveillance program for monitoring the reembrittlement trend for the beltline materials
should be described.
8. Keywords
8.1 fracture toughness; irradiation; nuclear reactor vessels (light-water moderated); radiation exposure; surveillance (of nuclear
reactor vessels)
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APPENDIXES
(Nonmandatory Information)
X1. BIBLIOGRAPHY OF MATERIAL PROPERTIES FOR PRESSURE VESSEL STEELS
X1.1 References containing existing material property information for pressure vessel materials are listed to cover annealing
response, changes in RT and upper shelf recovery, and reirradiation embrittlement. Limited fracture toughness data also are
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available. These data are to be used in assessing the anticipated annealing recovery and reembrittlement for similar pressure vessel
steels. These same data may be used to determine a generic response when relevant materials are not available for actual recovery
demonstration and surveillance.
X1.2 The reference bibliography (3-7885) of annealing information is not intended to be totally inclusive. Major emphasis is given
to U.S. commercial pressure vessel steels and welds, particularly those with high copper concentrations that may be critical in older
operating plants. Studies before 1974 (see Refs (3-12)) involved steels that only are typical of a few commercial vessels in
operation today.
X1.3 The work performed on annealing in the 1970s at the Naval Research Laboratory is summarized in Ref (13). For other
sources of information during the 1970s, see Refs (14-18).
X1.4 Data and evaluations reported in the 1980s and 1990s can be found starting with Ref (19). This compilation includes data
for European and Russian steels, for example, see Refs (20-47).
X1.5 More recent studies for pressure vessel steels, primarily focused on the WWER-440 and WWER-1000 steels, are contained
in Refs (68-7885).
X2. GUIDANCE FOR VERIFYING RECOVERY AND RE-IRRADIATION EMBRITTLEMENT
X2.1 The key elements with respect to continued operation of a reactor vessel after annealing are the degree of recovery and the
reembrittlement trend. Ideally, both of these elements should be measured using existing surveillance capsules containing the
limiting reactor beltline materials. Older vessels, however, which may be the first candidates for annealing, may not have enough
surveillance capsules, or the limiting material may not have been included in the surveillance program. Even if there are capsules
that can be used to assess annealing and the subsequent reembrittlement, different lead factors may make future assessments
difficult to directly quantify unless a reembrittlement trend curve can be estimated. The purpose of this appendix is to provide
guidance for defining the post-anneal reference temperature (RT ) and to estimate and measure the reembrittlement trends for
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reactor beltline materials. This guide is general since it is impractical to give specific quantitative directions due to the variety of
materials, irradiation conditions, and other considerations such as future operating plans.
X2.2 Quantification of annealing recovery has been studied in detail, primarily in test reactor environments, while subsequent
reembrittlement trends have less supporting data, and therefore, less definition. Upper shelf Charpy energy changes can be
addressed in a similar manner as the RT approach presented in this appendix.
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X2.3 The approach presented here is to provide guidance in developing an approximate annealing/reembrittlement trend curve
from the existing surveillance irradiation data and several correlations that can be checked with other available capsule results,
post-anneal, and used to project future trends. Test reactor irradiations with archive, or representative, materials may be used in
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special cases to check the trend curve methodology, but uncertainties due to temperature and fluence-rate effects should be
considered.
X2.4 Since the data base of annealing recovery and reembrittlement trend does not cover all materials and annealing conditions,
several assumptions have been made in developing a trend curve approach, and these assumptions should be kept in mind using
the methodology. Mechanistic modeling of the irradiation, annealing, and reirradiation processes for plant specific materials may
provide useful guidance and help reduce uncertainties in using this methodology.
X2.5 A conservative methodology of post-anneal reirradiation trend curve development is schematically shown in Fig. X2.1. This
methodology is termed “lateral shift” since the initial irradiation trend curve merely is translated laterally to project reirradiation
behavior.
X2.5.1 The initial irradiation correlation must be established for the critical material(s). Suggested methods include using Guide
E900 or NRC Regulatory Guide 1.99, Revision 2. From these guides, a mean prediction curve for initial irradiation (I) damage
is used with an approximate variance (σ ). Existing surveillance data, or other appropriately justified data, can be used to adjust
I
the mean curve, similar to the process allowed in NRC Regulatory Guide 1.99, Revision 2.
X2.5.2 The next step is the estimation of the annealing recovery for the irradiated-annealed (IA) condition. An approach, such as
suggested in NRC Regulatory Guide 1.162 and documented in Ref (49), may be used. This approach has been statistically
analyzed, and the corresponding overall variance (σ ) is no greater than that from the original irradiation. The variance associated
IA
2 2
with the anneal, therefore, is encompassed in the irradiation variance: σ = σ . Certain limitations of this approach are
IA I
acknowledged in Ref (49) relative to the range of applicable data and caution should be exercised when approaching these limiting
conditions. The limited extent of data used to develop the predictive equations also should be considered.
X2.5.3 Next is a lateral shift of the initial irradiation embrittlement path to become the post-anneal reirradiation trend curve for
the irradiated-annealed-reirradiated (IAR ) condition. This step has some technical uncertainty, but appears to be a logical first
2 2
approximation. A variance σ (assumed equivalent to σ ) for reembrittlement may be used to project a reirradiation trend curve
IAR I
and approximate statistical bound.
X2.5.4 The predicted trend curve should be checked by experimental results. This verification should be planned well before the
actual annealing takes place. The suggested procedures that may be followed in evaluating the estimated trend curve are described
in X2.7.
FIG. X2.1 Lateral Shift Method for Estimating Reirradiation Embrittlement
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X2.6 The “vertical shift” trend curve approach is similar to that of the “lateral shift,” except the portion of the initial irradiation
trend, projected as reirradiation behavior, is translated down vertically as shown in Fig. X2.2. The use of this estimated trend curve
should be justified with actual post-anneal reirradiation data since the vertical shift method predicts significantly lower changes
in RT after thermal annealing. Limited data show that reembrittlement trends for anneals near 850°F (454°C) for one week lie
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between the lateral and vertical shift approaches.
X2.7 The following procedures provide guidance for assessing recovery and reembrittlement prior to making the decision to
anneal, as well as developing the post-vessel anneal surveillance program once the decision to anneal has been made. The new
surveillance program will provide a check on the recovery and reembrittlement estimation methodology just described and provide
actual data for making adjustments when appropriate.
FIG. X2.2 Vertical Shift Method for Estimating Reirradiation Embrittlement
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X2.7.1 First, withdraw all, or nearly all, capsules from the reactor and follow the diagram in Fig. X2.3. The entry point into the
flow diagram is to answer the question in the top diamond-shaped box as to whether or not there is adequate material available
FIG. X2.3 Procedure for Evaluating Annealing Feasibility and Reembrittlement Trend
One capsule may be kept in place in case the decision to anneal is later reversed, or for contingency purposes.
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to perform testing on the vessel materials or other representative materials. If the answer is yes, then the following sequence of
steps beginning in X2.7.2 should be followed. If the answer is no, the steps beginning in X2.7.3 should be followed. Note that the
boxes in Fig. X2.3 are identified with the appropriate paragraph number as used throughout X2.7.
X2.7.2 For Use When Adequate Vessel or Representative Material Exists—The material can come from the existing surveillance
program (tested or untested specimens,) samples taken from the vessel wall, archived material from the original vessel construction
or the surveillance program, which can be irradiated, or from materials available from other sources that can be justified as
representative.
X2.7.2.1 An evaluation of how much material is available should be made to determine the extent of testing that can be performed
for IA and IAR condition assessment. If sufficient material is available to perform both pre-vessel and post-vessel annealing
evaluations, proceed to X2.7.2.2. If there is very limited material available, then emphasis should be placed upon the post-vessel
anneal evaluations in the new surveillance program, and the user should proceed to X2.7.2.3.
X2.7.2.2 Pre-vessel annealing testing of irradiated materials for evaluating the IA response, using the projected annealing time and
temperature, should be performed and compared to the existing IA correlation for the same time-temperature and material
condition. If additional material is available, as well as time to perform further irradiations prior to vessel annealing, IAR
experiments, again using the projected annealing time and temperature, can be performed to test the lateral shift model and provide
further insight prior to making the final decision to anneal. Test reactor irradiations can be considered for IA and IAR conditioning.
Proceed to X2.7.2.4.
NOTE X2.1—Actual testing is indicated in Fig. X2.3 by use of a parallelogram-shaped box.
X2.7.2.3 Since there is not adequate material to develop IA data prior to vessel annealing, the IA correlation approach should be
used to assess the degree of anticipated recovery for the vessel materials. The rectangular-shaped box indicates an evaluation or
analytical step. Proceed to the next step.
X2.7.2.4 The lateral shift model for IAR behavior should be used to assess the benefits to be realized once the vessel is placed
back in service and operated to some future point in time.
X2.7.2.5 Utilizing the recovery IA measured data, if available, and the IA correlations for the vessel materials, coupled with the
lateral shift IAR model, the actual decision to anneal the vessel can be made. If inadequate recovery, or reembrittlement trends,
or both, suggest poor performance from annealing, move to X2.7.4. If the decision to anneal is affirmative, then proceed as follows
for the post-vessel anneal surveillance program.
X2.7.2.6 The new post-vessel anneal surveillance program should be planned to utilize the best combination of available vessel
and representative materials for IA and IAR assessment.
X2.7.2.7 Perform post-vessel annealing on material irradiated to approximately the same fluence as the vessel, preferably in the
reactor which is to be annealed, to determine the vessel IA condition. The lower bound time and temperature from the actual vessel
anneal should be used for the test material annealing conditions. Another consideration would be the upper bound temperature and
time if temper-type embrittlement is expected to be relevant.
“Representative” materials should match the critical base and weld materials in the vessel with regard to ASTM specification, material heat, vintage, and chemistry
(copper and nickel content) in that order for base materials and weld wire specification, material heat, weld flux, vintage, and chemistry (copper, phosphorus, sulfur, and nickel
content) in that order for weld metals, to the extent practical. In some cases, representative materials also can be equated to bounding materials when shown that the expected
embrittlement trends should be greater than the actual critical materials. Practices E185 and E2215 provide details on original surveillance program design which can yield
guidance in developing a post-anneal surveillance program.
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X2.7.2.8 The IA data generated from the new surveillance program should be compared with projections from existing
correlation(s). If the results are within appropriate statistical limits, there are advantages in being able to directly use the IA
correlation(s) for complete vessel calculations, so proceed to the next step. If the results are different statistically, proceed to
X2.7.2.10 to make appropriate adjustments.
X2.7.2.9 Since the test results closely match and validate the correlation, continue to use existing IA correlation(s) for the vessel
materials. Proceed to X2.7.2.11.
X2.7.2.10 When the tested material results from the new surveillance program do not match statistically the predictions from IA
correlation, the differences between measured and predicted should be used to adjust the predicted response for the vessel materials
using a simple proportionality approach, for example, see NRC Regulatory Guide 1.162. Proceed to the next step.
X2.7.2.11 The lateral shift method provides a prediction approach for reembrittlement trend and should be used initially unless
there are data available to support a different trend methodology, that is, results from the optional IAR testing in X2.7.2.2.
X2.7.2.12 Post-vessel annealing reembrittlement data should be generated for the new surveillance program. At least one IAR
measurement should be made to correspond to the targeted end-of-license (EOL) fluence for the vessel after annealing. Additional
IAR measurements are encouraged since additional data can better define the reembrittlement trend. If only one additional IAR
measurement can be made, it should be conducted at an intermediate fluence between the time of annealing and the targeted EOL.
X2.7.2.13 Once two IAR measurements are made, a new trend curve can be developed by calculating a “new” chemistry factor
for reembrittlement, similar to the preanneal embrittlement approach in NRC Regulatory Guide 1.99, Revision 2 or using Guide
E900 and continuing to use the lateral shift method. Another approach would be to fit the data to a model falling between the lateral
shift and the vertical shift methods or using other predictive methods that can be justified technically. Once the data have been used
to the maximum degree possible, accurate assessment for P – T limits and compliance with Regulatory PTS screening criteria can
be made. Proceed to X2.7.5.
X2.7.3 For Use When Adequate Vessel or Representative Material Does Not Exist—When there is not adequate or representative
material available for developing a program to measure recovery and reembrittlement trends, more emphasis must be placed on
the correlation processes.
X2.7.3.1 The IA recovery correlations and the lateral shift model for IAR trending must be used to assess the benefits to be realized
from thermal annealing. This provides the key information for deciding upon the actual annealing of the embrittled vessel.
X2.7.3.2 This question is the same as in X2.7.2.4. If inadequate recovery or reembrittlement, or both, suggest poor performance
from annealing, move to X2.7.4. If the decision to anneal is affirmative, then proceed as follows for a post-vessel anneal
surveillance program.
X2.7.3.3 A new surveillance program will be needed to monitor the effects of continued embrittlement on vessel steels. Since there
are no materials that can be judged as representative, material(s) from the class of steels in the vessel should be selected in a
conservative manner to develop a new surveillance program. Proceed to X2.7.2.7.
X2.7.4 This step represents the outcome if annealing is judged to be inadequate as a mitigative measure for radiation
embrittlement for the projected EOL fluence of the vessel. Other options for embrittlement management should be pursued, some
of which already may be part of the overall embrittlement management program.
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X2.7.5 This is the completion step for all of paths in Fig. X2.3. Sufficient time must be planned to develop an acceptable
surveillance program for assuring verification and monitoring of the annealing recovery and reembrittlement. The exact degree of
timing depends upon many factors, including capsule lead factors and the condition of available materials. Test reactor experiments
can be a viable option in circumstances in which immediate answers are needed, al
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