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ASTM E531-76(2007)

(Practice)

Standard Practice for Surveillance Testing of High-Temperature Nuclear Component Materials

Standard Practice for Surveillance Testing of High-Temperature Nuclear Component Materials

SIGNIFICANCE AND USE
The requirements contained herein can be used as a basis for establishing conditions for safe operation of critical components. The requirements provide for general plant assessment and verification that materials meet design criteria. The test specimens and procedures presented in this practice are for guidance when establishing a surveillance program.
This practice for high-temperature materials surveillance programs is used when nuclear reactor component materials are monitored by specimen testing. Periodic testing is performed through the service life of the components to assess changes in selected material properties that are caused by neutron irradiation and thermal effects. The properties are those used as design criteria for the respective nuclear components. The extent of material property change caused by neutron irradiation depends on the composition and structure of the initial material, its conditioning in component fabrication, as well as the nature of the irradiation exposure. The need for surveillance arises from a concern of specific material behavior under all irradiation conditions including spectrum and rate effects on material properties.
SCOPE
1.1 This practice covers procedures for specimen testing to establish changes occurring in the mechanical properties due to irradiation and thermal effects of nuclear component metallic materials where these materials are used for high temperature applications above 370°C (700°F).

General Information

Status
Historical
Publication Date
31-May-2007
ICS
27.120.10 - Reactor engineering
Technical Committee
E10 - Nuclear Technology and Applications
Drafting Committee
E10.02 - Behavior and Use of Nuclear Structural Materials
Current Stage
Ref Project

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Replaces
ASTM E531-76(2001) - Standard Practice for Surveillance Testing of High-Temperature Nuclear Component Materials
Effective Date
01-Jun-2007

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ASTM E531-76(2007) - Standard Practice for Surveillance Testing of High-Temperature Nuclear Component MaterialsASTM E531-76(2007) - Standard Practice for Surveillance Testing of High-Temperature Nuclear Component Materials
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ASTM E531-76(2007) - Standard Practice for Surveillance Testing of High-Temperature Nuclear Component Materials
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NOTICE: This standard has either been superseded and replaced by a new version or withdrawn.
Contact ASTM International (www.astm.org) for the latest information
Designation: E531 − 76(Reapproved 2007)
Standard Practice for
Surveillance Testing of High-Temperature Nuclear
Component Materials
This standard is issued under the fixed designation E531; 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 E399 Test Method for Linear-Elastic Plane-Strain Fracture
Toughness K of Metallic Materials
Ic
1.1 This practice covers procedures for specimen testing to
E453 Practice for Examination of Fuel Element Cladding
establishchangesoccurringinthemechanicalpropertiesdueto
Including the Determination of the Mechanical Properties
irradiation and thermal effects of nuclear component metallic
E482 Guide for Application of Neutron Transport Methods
materials where these materials are used for high temperature
for Reactor Vessel Surveillance, E706 (IID)
applications above 370°C (700°F).
E844 Guide for Sensor Set Design and Irradiation for
Reactor Surveillance, E 706 (IIC)
2. Referenced Documents
2.1 ASTM Standards:
3. Significance and Use
A370 Test Methods and Definitions for Mechanical Testing
3.1 The requirements contained herein can be used as a
of Steel Products
E3 Guide for Preparation of Metallographic Specimens basis for establishing conditions for safe operation of critical
components. The requirements provide for general plant as-
E8 Test Methods for Tension Testing of Metallic Materials
E21 TestMethodsforElevatedTemperatureTensionTestsof sessment and verification that materials meet design criteria.
The test specimens and procedures presented in this practice
Metallic Materials
E23 Test Methods for Notched Bar Impact Testing of Me- are for guidance when establishing a surveillance program.
tallic Materials
3.2 This practice for high-temperature materials surveil-
E29 Practice for Using Significant Digits in Test Data to
lance programs is used when nuclear reactor component
Determine Conformance with Specifications
materialsaremonitoredbyspecimentesting.Periodictestingis
E45 Test Methods for Determining the Inclusion Content of
performed through the service life of the components to assess
Steel
changes in selected material properties that are caused by
E112 Test Methods for Determining Average Grain Size
neutronirradiationandthermaleffects.Thepropertiesarethose
E139 Test Methods for Conducting Creep, Creep-Rupture,
used as design criteria for the respective nuclear components.
and Stress-Rupture Tests of Metallic Materials
The extent of material property change caused by neutron
E185 Practice for Design of Surveillance Programs for
irradiation depends on the composition and structure of the
Light-Water Moderated Nuclear Power Reactor Vessels
initial material, its conditioning in component fabrication, as
E206 Definitions of Terms Relating to Fatigue Testing and
well as the nature of the irradiation exposure. The need for
the Statistical Analysis of Fatigue Data; Replaced by
surveillancearisesfromaconcernofspecificmaterialbehavior
E 1150 (Withdrawn 1988)
under all irradiation conditions including spectrum and rate
E261 Practice for Determining Neutron Fluence, Fluence
effects on material properties.
Rate, and Spectra by Radioactivation Techniques
4. Description of Term
This recommended practice is under the jurisdiction ofASTM Committee E10
4.1 test specimen—a coupon or a piece of metal cut from a
on Nuclear Technology andApplicationsand is the direct responsibility of Subcom-
larger metal piece which is then formed to final size for testing
mittee E10.02 on Behavior and Use of Nuclear Structural Materials.
to determine physical or mechanical properties.
CurrenteditionapprovedJune1,2007.PublishedJuly2007.Originallyapproved
in 1975. Last previous edition approved in 2001 as E531–76(2001). DOI: 10.1520/
E0531-76R07.
5. Test Specimens
For referenced ASTM standards, visit the ASTM website, www.astm.org, or
contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
5.1 Pre-Exposure Material Characterization—It is impor-
Standards volume information, refer to the standard’s Document Summary page on
tant that test specimen materials be characterized prior to
the ASTM website.
exposure and that the following should be considered as a
The last approved version of this historical standard is referenced on
www.astm.org. minimum:
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E531 − 76 (2007)
5.1.1 Process history, material designation, manufacturer, for surveillance specimens, the archive, base line, and thermal
heat number, weld and fabrication procedures used, and heat control specimens shall be identical with the surveillance
treatment, specimens.
5.1.2 Original location and orientation in the parent 5.3.3 Surface Condition—Test specimens where surface
material, condition is critical to the test results should not be finish
machined in such critical areas (Charpy notch, fatigue speci-
5.1.3 Specimen weight and dimensions,
mentestarea,surfaceofdensitychangesample)untiljustprior
5.1.4 Metallographic characteristics (grain size,
to test. Specimens should be oversized to allow for removal of
microstructure,andinclusioncontentestablishedinaccordance
at least 0.1 mm of surface prior to test. Where possible, test
with Test Methods E45 and E112),
specimens with the exception of weight change specimens
5.1.5 Chemical analysis results,
shall be encapsulated in an inert environment so as to deter-
5.1.6 All specimens shall be taken from the specified
mine only the effect of neutron irradiation and temperature on
location and orientation specified in Test Methods and Defini-
mechanical or physical properties. It is recognized the integrity
tions A370 and Test Methods E8, and
of the encapsulation may be breached in some cases during
5.1.7 Mechanical properties including yield strength, tensile
longexposureandanallowanceforfinalmachiningevenofthe
strength, stress rupture life, creep strength, fatigue strength,
encapsulated specimens should be considered.This will ensure
and impact strength as a function of test temperature.
a meaningful comparison between baseline and exposed speci-
5.1.8 The information described in 5.1.1-5.1.7 should be
mens.
reported in a single document.
5.3.4 Number of Specimens—The number of specimens
5.2 Post Exposure Material Characterization—After
employedformechanicalpropertytestingshouldbeselectedso
exposure, the following should be reported:
astoincludeeachcriticalcomponentthatvariessignificantlyin
5.2.1 Observations from visual examination,
composition, processing, or in exposure conditions from simi-
5.2.2 Changes in specimen weight and dimensions,
lar components. Specific recommendations as to number of
5.2.3 Metallographic characteristics (grain size,
specimens will be found in the respective specimen sections.
microstructure, and inclusion content), At least four sets of specimens shall be included in each
5.2.4 Results of chemical analysis, surveillance program.
5.2.5 Appropriate mechanical properties being surveyed 5.3.5 Material—Test specimens shall be taken from the
material used in component fabrication. The material shall be
including considerations of changes in tensile strength, stress-
to-rupture strength, creep strength, fatigue strength, impact processed at the same time as the component or processed in a
fashion identical to the component surveyed. Weld and heat-
strength (control tests are recommended to be performed
affected zone test specimens shall be taken from equivalent
simultaneously with the tests of exposed specimens to ensure
material welded at the same time as the particular component
that deviations in test results can be attributed to the exposed
or equivalent material welded with the same welding param-
specimen’s environment and not to variations in testing
eters. It is not necessary to include each heat or minor
methods), and
variation, but only to select those receiving the highest expo-
5.2.6 Optional quantitative examination of surface chemis-
sure or those previously found to be most sensitive to neutron
try and subsequent changes.
irradiation and temperature or those that can restrict the
5.2.7 Exposed test specimens should be cleaned in accor-
operation of the reactor. Test specimens may be taken from
dance with accepted cleaning procedures. (Refer to Subcom-
components periodically removed from the reactor for other
mittee G01.08 for practices for preparing, cleaning, and evalu-
reasons. These specimens can be used to provide supplemental
ating test specimens.)
surveillanceinformation.Forthisinformationtobemeaningful
5.3 Specimen Preparation—Test specimens shall be stan-
a full characterization of the pre-exposure condition must be
dard recommended specimens where possible as described in
available along with measured exposure conditions.
Test Methods E8, E21 and E23 and Practice E139. The use of
5.4 Tension Test Specimens—The type and size of specimen
the word specimen or words test specimen as used in this
to be used shall conform to the smaller sizes as recommended
practice is described in Section 4.
in Test Methods E8 and E21. Either threaded or button-head
5.3.1 The test area of a specimen (for example, Charpy
ends will be acceptable. For plate or sheet specimens, pin ends
notch, reduced section of a fatigue specimen) may be left
as described in Test Methods E8 are recommended. The
unfinished if further environmental exposure prior to testing is
location and orientation of test specimens shall be as defined in
anticipated.
Test Methods E8 or Test Methods and Definitions A370,orin
5.3.2 Size—In general, due to the limited space available in
Practice E185. Both base metal and weld metal specimens will
surveillance capsules the smaller sizes of test specimens are
be taken.Aset of tension specimens shall consist of three each
recommended.Where it is not possible to use specimens of the
base metal and weld metal.
recommended size, the least deviation possible from recom-
mended sizes should be adhered to. Non-standard specimens 5.5 Creep and Stress-Rupture in Specimens—The type and
shall be evaluated prior to use as surveillance specimens to size of specimen to be used shall be the same as those used for
ensure that test results from the use of non-standard specimens tension specimens except that button-head or pinned-end
can be correlated with test results from specimens of recom- specimens are recommended for high-temperature testing.
mendedsize.Intheeventthatnon-standardspecimensareused Practice E453 describes the attention that must be paid to
E531 − 76 (2007)
specimen alignment and dimensional tolerances. One set of to be made, the test specimen may be electropolished in
tests shall be conducted at the operating temperature of the accordance with Methods E3. Test specimens that are suscep-
component of interest. A set shall comprise a minimum of six tible to corrosion in room-temperature air shall be stored as
stress rupture tests at six different stress levels. The stress soon as practicable after preparation in an inert dry gas or
levels should be selected so that the time-to-rupture ranges vacuum. A set of specimens shall consist of ten each of base
from not less than 100 h to at least 3000 h. Creep strain metal and weld metal.
measurements may be made if desired.
5.7 Swelling Specimens—The swelling specimens shall be
right circular cylinders 5.0 mm diameter and 10.0 mm long.
5.6 Low-Cycle Fatigue Specimens—For base metal the type
The sharp-cornered specimens shall be finished by turning or
and size of specimen to be used may be the “hourglass” type
grinding and have a surface finish of 0.2 µm R (8 µin. AA).
with threaded or button-head ends as shown in Fig. 1. For weld
a
Thespecimensshallbedegreasedinsuitablesolventandstored
metal specimens the uniform gage type may be used. Machin-
in an inert gas or vacuum.
ing and polishing of the test specimens shall be performed with
care so as to minimize the effects of specimen preparation on
5.8 Charpy Impact Specimens—The specimens to be used
fatigue life. In the final stages of machining, material shall be
shall conform to the Charpy V-notch specimens recommended
removed in the radial thickness amounts of 0.2 mm until 0.1
in Test Methods E23. The notch shall not be formed prior to
mm remains. After exposure the final 0.075 mm shall be
exposure as a surveillance specimen. The location and orien-
removed by cylindrical grinding at no more than 0.005 mm per
tation of test specimens shall be as defined in Practice E185
pass. The final 0.025 mm shall be buffed and finished with an
and Test Methods and Definitions A370. A set of specimens
0.2 µm R (8 µin. AA) surface roughness. After polishing, all
a shallbemadeupoftwelveeachbasemetal,heat-affectedzone,
remaining grinding and polishing marks shall be longitudinal
and weld metal.
and any circumferential grinding marks must be removed. The
6. Irradiation Conditions
finished specimens shall be degreased in suitable solvent.
Specimens to be exposed to liquid sodium shall not be 6.1 Introduction—The intent of the section on irradiation
degreased in halogenated solvents. If surface observations are conditions is to provide guidance on how to place surveillance
NOTE 1—A = 0.2 µm R (8 min AA).
a
Metric Conversion
mm in.
0.3 0.01
6.35 0.250
8.4 0.33
12.68 0.499
12.70 0.500
19.1 0.75
38.1 1.5
82.6 3.25
FIG. 1 Standard Hour-Glass Low Cycle Fatigue Specimen (Threaded Ends (a) and Button Head (b))
E531 − 76 (2007)
samples to obtain the desired irradiation conditions in terms of 6.5 Specimen Withdrawal Schedule—The specimen with-
temperature, neutron flux, and neutron spectrum to ensure a drawal schedule shall be as specified in Practice E185, Case B,
based on the percentage of component life as specified inTable
realistic evaluation of the component that the surveillance
1.
specimen is representing.
6.2 Irradiation Temperature—It is very important that ad-
7. Measurement of Neutron Exposure
equate consideration be given to test specimens to ensure that
7.1 The neutron flux and neutron energy spectrum at the
they experience the correct temperature during irradiation.
surveillance location shall be given as well as the method used
Temperature must be controlled for the surveillance specimens
for determination. All assumptions should be clearly stated,
to match as nearly as possible the temperature of the compo-
and all physical constants, such as cross sections, half lives,
nent being surveyed.
...

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4.1 Regulatory Requirements—The USA Code of Federal Regulations (10CFR Part 50, Appendix H) requires the implementation of a reactor vessel materials surveillance program for all operating LWRs. Other countries have similar regulations. The purpose of the program is to (1) monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline6 resulting from exposure to neutron irradiation and the thermal environment, and (2) make use of the data obtained from surveillance programs to determine the conditions under which the vessel can be operated with adequate margins of safety throughout its service life. Practice E185, derived mechanical property data, and (r, θ, z) physics-dosimetry data (derived from the calculations and reactor cavity and surveillance capsule measurements (1) using physics-dosimetry standards) can be used together with information in Guide E900 and Refs. 4, 11-18 to provide a relation between property degradation and neutron exposure, commonly called a “trend curve.” To obtain this trend curve at all points in the pressure vessel wall requires that the selected trend curve be used together with the appropriate (r, θ, z) neutron field information derived by use of this guide to accomplish the necessary interpolations and extrapolations in space and time.  
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4.1 A characteristic advantage of charged-particle irradiation experiments is the precise, individual control over most of the important irradiation conditions such as dose, dose rate, temperature, and quantity of gases present. Additional attributes are the lack of induced radioactivation of specimens and, in general, a substantial compression of irradiation time, from years to hours, to achieve comparable damage as measured in displacements per atom (dpa). An important application of such experiments is the investigation of radiation effects that may occur in materials exposed to environments which do not currently exist, such as in first wall materials used in fusion reactors.  
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4.1 This guide deals with the difficult problem of benchmarking neutron transport calculations carried out to determine fluences for plant specific reactor geometries. The calculations are necessary for fluence determination in locations important for material radiation damage estimation and which are not accessible to measurement. Typically, the most important application of such calculations is the estimation of fluence within the reactor vessel of operating light water reactors (LWR) to provide accurate estimates of the irradiation embrittlement of the base and weld metal in the vessel. The benchmark procedure must not only prove that calculations give reasonable results but that their uncertainties are propagated with due regard to the sensitivities of the different input parameters used in the transport calculations. Benchmarking is achieved by building up data bases of benchmark experiments that have different influences on uncertainty propagation. For example, in simple vessel wall mockups where measurements are made within a simulated reactor vessel wall, the integral effect of uncertainties in iron cross sections (absorption and elastic and inelastic scattering) are dominant and have been bounded by the agreement between calculation and measurement. For more complicated integral benchmarks, other factors such as: uncertainties in the distribution of fission sources, geometry, the energy-dependent cross sections, and the angular scattering distribution for elemental components of major materials in the neutron field (such as water and iron) may all be important uncertainty contributors. This guide describes general procedures for using neutron fields with known characteristics to corroborate the calculational methodology and nuclear data used to derive neutron field information from measurements of neutron sensor response.  
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4.1 Predictions of neutron radiation effects on pressure vessel steels are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters often are made throughout the service life of the reactor vessel to account for radiation effects. Due to the variability in the behavior of reactor vessel steels, a surveillance program is warranted to monitor changes in the properties of actual vessel materials caused by long-term exposure to the neutron radiation and temperature environment of the reactor vessel. This practice describes the criteria that should be considered in planning and implementing surveillance test programs and points out precautions that should be taken to ensure that: (1) capsule exposures can be related to beltline exposures, (2) materials selected for the surveillance program are samples of those materials most likely to limit the operation of the reactor vessel, and (3) the test specimen types are appropriate for the evaluation of radiation effects on the reactor vessel.  
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4.1 Operation of commercial power reactors must conform to pressure-temperature limits during heatup and cooldown to prevent over-pressurization at temperatures that might cause non-ductile behavior in the presence of a flaw. Radiation damage to the reactor vessel is compensated for by adjusting the pressure-temperature limits to higher temperatures as the neutron damage accumulates. The present practice is to base that adjustment on the TTS  produced by neutron irradiation as measured at the Charpy V-notch 41-J (30-ft·lbf) energy level. To establish pressure temperature operating limits during the operating life of the plant, a prediction of TTS  must be made.  
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4.3 Only power reactor (PWR and BWR) surveillance data were used in the derivation of these procedures. The measure of fast neutron fluence used in the procedure is n/m2  (E > 1 MeV). Differences in fluence rate and neutron energy spectra experienced in power reactors and test reactors have not been accounted for in these procedures.
SCOPE
1.1 This guide presents a method for predicting values of reference transition temperature shift (TTS) for irradiated pressure vessel materials. The method is based on the TTS exhibited by Charpy V-notch data at 41-J (30-ft·lbf) obtained from surveillance programs conducted in several countries for commercial pressurized (PWR) and boiling (BWR) light-water cooled (LWR) power reactors. An embrittlement correlation has been developed from a statistical analysis of the large surveillance database consisting of radiation-induced TTS and related information compiled and analyzed by Subcommittee E10.02. The details of the database and analysis are described in a separate report (ADJE090015-EA).2,3 This embrittlement correlation was developed using the variables copper, nickel, phosphorus, manganese, irradiation temperature, neutron fluence, and product form. Data ranges and conditions for these variables are listed in 1.1.1. Section 1.1.2 lists the materials included in the database and the domains of exposure variables that may influence TTS but are not used in the embrittlement correlation.  
1.1.1 The range of material and irradiation conditions in the database for variables used in the embrittlement correlation:  
1.1.1.1 Copper content up to 0.4 %.
1.1.1.2 Nickel content up to 1.7 %.
1.1.1.3 Phosphorus content up to 0.03 %.
1.1.1.4 Manganese content within the range from 0.55 to 2.0 %.
1.1.1.5 Irradiation temperature within the range from 255 to 300°C (491 to 572°F).
1.1.1.6 Neutron fluence within the range from 1 × 1021 n/m2 to 2 × 1024 n/m2 (E> 1 MeV).
1.1.1.7 A categorical variable describing the product form (that is, weld, plate, forging).  
1.1.2 The range of material and irradiation conditions in the database for variables not included in the embrittlement correlation:  
1.1.2.1 A533 Type B Class 1 and 2, A302 Grade B, A302 Grade B (modified), and A508 Class 2 and 3. Also, European and Japanese steel grades that are equivalent to these ASTM Grades.
1.1.2.2 Submerged arc welds, shielded a...

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ASTM
ASTM E2005-21(Guide)
Standard Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields

SIGNIFICANCE AND USE
3.1 This guide describes approaches for using neutron fields with well known characteristics to perform calibrations of neutron sensors, to intercompare different methods of dosimetry, and to corroborate procedures used to derive neutron field information from measurements of neutron sensor response.  
3.2 This guide discusses only selected standard and reference neutron fields which are appropriate for benchmark testing of light-water reactor dosimetry. The Standard Fields considered here include neutron source environments that closely approximate: a) the unscattered neutron spectra from  252Cf spontaneous fission; and b) the  235U thermal neutron induced fission. These standard fields were chosen for their spectral similarity to the high energy region (E > 2 MeV) of reactor spectra. The various categories of benchmark fields are defined in Terminology E170.  
3.3 There are other well known neutron fields that have been designed to mockup special environments, such as pressure vessel mockups in which it is possible to make dosimetry measurements inside of the steel volume of the “vessel.” When such mockups are suitably characterized, they are also referred to as benchmark fields. A variety of these engineering benchmark fields have been developed, or pressed into service, to improve the accuracy of neutron dosimetry measurement techniques. These special benchmark experiments are discussed in Guide E2006, and in Refs (1)4 and (2).
SCOPE
1.1 This guide covers facilities and procedures for benchmarking neutron measurements and calculations. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to calibrate integral neutron sensors; the use of certified-neutron-fluence standards to calibrate radiometric counting equipment or to determine interlaboratory measurement consistency; development of special benchmark fields to test neutron transport calculations; use of well-known fission spectra to benchmark spectrum-averaged cross sections; and the use of benchmarked data and calculations to determine the uncertainties in derived neutron dosimetry results.  
1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.  
1.3 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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.4 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.

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ASTM
ASTM E509/E509M-21(Guide)
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
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 te...
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).2  
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...

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ASTM
ASTM E1167-15(2021)(Guide)
Standard Guide for Radiation Protection Program for Decommissioning Operations

SIGNIFICANCE AND USE
4.1 A program based on this guide will provide assurance to all concerned that the appropriate elements of radiation safety have been included to protect workers, the general public, and the environment in proximity to the decommissioning activities.  
4.2 Implementation of such a program will provide assurance to those agencies responsible for review or audit of the decommissioning project that the requirements for radiation protection have been addressed.
SCOPE
1.1 This guide provides instruction to the individual charged with the responsibility for developing and implementing the radiation protection program for decommissioning operations.  
1.2 This guide provides a basis for the user to develop radiation protection program documentation that will support both the radiological engineering and radiation safety aspects of the decommissioning project.  
1.3 This guide presents a description of those elements that should be addressed in a specific radiation protection plan for each decommissioning project. The plan would, in turn, form the basis for development of the implementation procedures that execute the intent of the plan.  
1.4 This guide applies to the development of radiation protection programs established to control exposures to radiation and radioactive materials associated with the decommissioning of nuclear facilities. The intent of this guide is to supplement existing radiation protection programs as they may pertain to decommissioning workers, members of the general public and the environment by describing the basic elements of a radiation protection program for decommissioning operations.  
1.5 This guide defines the elements of a radiation protection program that will ensure that the goals and objectives of a decommissioning activity are attained within the radiological limits and restrictions imposed by applicable governing and regulating agencies. The implementation of such a program will provide radiological protection to personnel and the environment. This guide should be used for developing the documentation that defines the intent and implementation of the radiation protection program for a specific decommissioning project.  
1.6 The Radiation Protection Program should address the following elements (see Note 1). This program shall be developed and maintained such that it satisfies all applicable Quality Assurance requirements developed for the decommissioning project.  
Note 1: If the site to be decommissioned is adjacent to an operating site, the radiological impact of the operating site must be considered in the development of the Radiation Protection Program for the decommissioning site.  
1.7 This guide does not address the subjects of emergency preparedness, safeguards, accountability, waste handling, storage, and transportation. Each of these issues has a direct interface with the radiation protection program. However, each constitutes a program in and of itself from program definition through implementation.  
1.8 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, health, and environmental practices and determine the applicability of regulatory limitations prior to use.  
1.9 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.

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ASTM
ASTM E636-20(Guide)
Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels

SIGNIFICANCE AND USE
3.1 Practices E185 and E2215 describe a minimum program for the surveillance of reactor vessel materials, specifically mechanical property changes that occur in service. This guide may be applied to generate additional information on radiation-induced property changes to better assist the determination of the optimum reactor vessel operation schemes.
SCOPE
1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required by Practices E185 and E2215 for the surveillance of nuclear reactor vessels. The supplemental mechanical property tests outlined permit the acquisition of additional information on radiation-induced changes in mechanical properties of the reactor vessel steels.  
1.2 This guide provides recommendations for the preparation of test specimens for irradiation, and identifies special precautions and requirements for reactor surveillance operations and post-irradiation test planning. Guidance on data reduction and computational procedures is also given. Reference is made to other ASTM test methods for the physical conduct of specimen tests and for raw data acquisition.  
1.3 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.  
1.4 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.

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ASTM
ASTM E2215-19(Practice)
Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels

SIGNIFICANCE AND USE
4.1 Neutron radiation effects are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters may be made throughout the service life of the reactor to account for these effects. A surveillance program is used to measure changes in the properties of actual vessel materials due to the irradiation environment. This practice describes the criteria that should be considered in evaluating surveillance program test capsules.  
4.2 Prior to the first issue date of this standard, the design of surveillance programs and the testing of surveillance capsules were both covered in a single standard, Practice E185. Between its provisional adoption in 1961 and its replacement linked to this standard, Practice E185 was revised many times (1966, 1970, 1973, 1979, 1982, 1993 and 1998). Therefore, capsules from surveillance programs that were designed and implemented under early versions of the standard were often tested after substantial changes to the standard had been adopted. For clarity, the standard practice for surveillance programs has been divided into the new Practice E185 that covers the design of new surveillance programs and this standard practice that covers the testing and evaluation of surveillance capsules. Modifications to the standard test program and supplemental tests are described in Guide E636.  
4.3 This practice is intended to cover testing and evaluation of all light-water moderated reactor pressure vessel surveillance capsules. The practice is applicable to testing of capsules from surveillance programs designed and implemented under all previous versions of Practice E185.  
4.4 The radiation-induced changes in the properties of the reactor pressure vessel are generally monitored by measuring the index temperatures, the upper-shelf energy and the tensile properties of specimens from the surveillance program capsules. The significance of these radiation-induced changes is described in Practice E185.  
4...
SCOPE
1.1 This practice covers the evaluation of test specimens and dosimetry from light water moderated nuclear power reactor pressure vessel surveillance capsules.  
1.2 Additionally, this practice provides guidance on reassessing withdrawal schedule for design life and operation beyond design life.  
1.3 This practice is one of a series of standard practices that outline the surveillance program required for nuclear reactor pressure vessels. The surveillance program monitors the irradiation-induced changes in the ferritic steels that comprise the beltline of a light-water moderated nuclear reactor pressure vessel.  
1.4 This practice along with its companion surveillance program practice, Practice E185, is intended for application in monitoring the properties of beltline materials in any light-water moderated nuclear reactor.2  
1.5 Modifications to the standard test program and supplemental tests are described in Guide E636.  
1.6 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.  
1.7 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.

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