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ASTM E531-76(2001)

(Practice)

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

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

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
25-Mar-1976
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

Relations

Replaces
ASTM E531-76(1996)e1 - Standard Practice for Surveillance Testing of High-Temperature Nuclear Component Materials
Effective Date
26-Mar-1976
Replaced By
ASTM E531-76(2007) - Standard Practice for Surveillance Testing of High-Temperature Nuclear Component Materials
Effective Date
01-Jun-2007

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ASTM E531-76(2001) - Standard Practice for Surveillance Testing of High-Temperature Nuclear Component MaterialsASTM E531-76(2001) - Standard Practice for Surveillance Testing of High-Temperature Nuclear Component Materials
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ASTM E531-76(2001) - 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NOTICE: This standard has either been superseded and replaced by a new version or withdrawn.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 2001)
Standard Practice for
Surveillance Testing of High-Temperature Nuclear
Component Materials
This standard is issued under the fixed designation E 531; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (e) indicates an editorial change since the last revision or reapproval.
1. Scope the Statistical Analysis of Fatigue Data
E 261 Practice for Determining Neutron Fluence Rate, Flu-
1.1 This practice covers procedures for specimen testing to
ence, and Spectra by Radioactivation Techniques
establishchangesoccurringinthemechanicalpropertiesdueto
E 399 Test Method for Plane-Strain Fracture Toughness of
irradiation and thermal effects of nuclear component metallic
Metallic Materials
materials where these materials are used for high temperature
E 453 Practice for Examination of Fuel Element Cladding
applications above 370°C (700°F).
Including the Determination of the Mechanical Properties
2. Referenced Documents E 482 Guide forApplication of Neutron Transport Methods
for Reactor Vessel Surveillance, E706 (IID)
2.1 ASTM Standards:
E 844 Guide for Sensor Set Design and Irradiation for
A 370 Test Methods and Definitions for MechanicalTesting
Reactor Surveillance, E706 (IIC)
of Steel Products
E 3 Methods of Preparation of Metallographic Specimens
3. Significance and Use
E 8 Test Methods forTensionTesting of Metallic Materials
3.1 The requirements contained herein can be used as a
E 21 Test Methods for Elevated Temperature Tension Tests
basis for establishing conditions for safe operation of critical
of Metallic Materials
components. The requirements provide for general plant as-
E 23 Test Methods for Notched Bar Impact Testing of
sessment and verification that materials meet design criteria.
Metallic Materials
The test specimens and procedures presented in this practice
E 29 Practice for Using Significant Digits in Test Data to
are for guidance when establishing a surveillance program.
Determine Conformance with Specifications
3.2 This practice for high-temperature materials surveil-
E 45 Test Methods for Determining the Inclusion Content
lance programs is used when nuclear reactor component
of Steel
materialsaremonitoredbyspecimentesting.Periodictestingis
E 112 Test Methods for Determining the Average Grain
performed through the service life of the components to assess
Size
changes in selected material properties that are caused by
E 139 Practice for Conducting Creep, Creep-Rupture, and
neutronirradiationandthermaleffects.Thepropertiesarethose
Stress-Rupture Tests of Metallic Materials
used as design criteria for the respective nuclear components.
E 184 Practice for Effects of High-Energy Neutron Radia-
The extent of material property change caused by neutron
tion on the Mechanical Properties of Metallic Materials,
5 irradiation depends on the composition and structure of the
E706 (IB)
initial material, its conditioning in component fabrication, as
E 185 Practice for Conducting Surveillance Tests for Light
5 well as the nature of the irradiation exposure. The need for
Water-Cooled Nuclear Power Reactor Vessels, E706 (IF)
surveillancearisesfromaconcernofspecificmaterialbehavior
E 206 Definitions of Terms Relating to Fatigue Testing and
under all irradiation conditions including spectrum and rate
effects on material properties.
This recommended practice is under the jurisdiction ofASTM Committee E10
4. Description of Term
on Nuclear Technology andApplicationsand is the direct responsibility of Subcom-
mittee E10.02 on Behavior and Use of Metallic Materials in Nuclear Systems.
4.1 test specimen—a coupon or a piece of metal cut from a
Current edition approved March 26, 1976. Published July 1976. Originally
larger metal piece which is then formed to final size for testing
published as E 531–75. Last previous edition E 531–75.
to determine physical or mechanical properties.
Annual Book of ASTM Standards, Vol 01.03.
Annual Book of ASTM Standards, Vol 03.01.
Annual Book of ASTM Standards, Vol 14.02.
Annual Book of ASTM Standards, Vol 12.02.
Discontinued—see 1986 Annual Book of ASTM Standards, Vol 03.01.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.
E531
5. Test Specimens shall be evaluated prior to use as surveillance specimens to
ensure that test results from the use of non-standard specimens
5.1 Pre-Exposure Material Characterization—It is impor-
can be correlated with test results from specimens of recom-
tant that test specimen materials be characterized prior to
mendedsize.Intheeventthatnon-standardspecimensareused
exposure and that the following should be considered as a
for surveillance specimens, the archive, base line, and thermal
minimum:
control specimens shall be identical with the surveillance
5.1.1 Process history, material designation, manufacturer,
specimens.
heat number, weld and fabrication procedures used, and heat
treatment,
5.3.3 Surface Condition—Test specimens where surface
5.1.2 Original location and orientation in the parent mate- condition is critical to the test results should not be finish
rial,
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, microstruc-
to test. Specimens should be oversized to allow for removal of
ture, and inclusion content established in accordance with Test
at least 0.1 mm of surface prior to test. Where possible, test
Methods E 45 and E 112),
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 A 370 and Test Methods E 8, 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 expo-
employedformechanicalpropertytestingshouldbeselectedso
sure, 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, microstruc-
specimens will be found in the respective specimen sections.
ture, 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
including considerations of changes in tensile strength, stress-
material used in component fabrication. The material shall be
to-rupture strength, creep strength, fatigue strength, impact
processed at the same time as the component or processed in a
strength (control tests are recommended to be performed
fashion identical to the component surveyed. Weld and heat-
simultaneously with the tests of exposed specimens to ensure
affected zone test specimens shall be taken from equivalent
that deviations in test results can be attributed to the exposed
material welded at the same time as the particular component
specimen’s environment and not to variations in testing meth-
or equivalent material welded with the same welding param-
ods), and
eters. It is not necessary to include each heat or minor
5.2.6 Optional quantitative examination of surface chemis-
variation, but only to select those receiving the highest expo-
try and subsequent changes.
sure or those previously found to be most sensitive to neutron
5.2.7 Exposed test specimens should be cleaned in accor-
irradiation and temperature or those that can restrict the
dance with accepted cleaning procedures. (Refer to Subcom-
operation of the reactor. Test specimens may be taken from
mittee G01.08 for practices for preparing, cleaning, and evalu-
components periodically removed from the reactor for other
ating test specimens.)
reasons. These specimens can be used to provide supplemental
5.3 Specimen Preparation—Test specimens shall be stan-
surveillanceinformation.Forthisinformationtobemeaningful
dard recommended specimens where possible as described in
a full characterization of the pre-exposure condition must be
Test Methods E 8, E 21 and E 23 and Practice E 139. The use
available along with measured exposure conditions.
of the word specimen or words test specimen as used in this
5.4 TensionTestSpecimens—The type and size of specimen
practice is described in Section 4.
to be used shall conform to the smaller sizes as recommended
5.3.1 The test area of a specimen (for example, Charpy
in Test Methods E 8 and E 21. Either threaded or button-head
notch, reduced section of a fatigue specimen) may be left
ends will be acceptable. For plate or sheet specimens, pin ends
unfinished if further environmental exposure prior to testing is
anticipated. as described in Test Methods E 8 are recommended. The
location and orientation of test specimens shall be as defined in
5.3.2 Size—In general, due to the limited space available in
Test Methods E 8 or Test Methods and DefinitionsA 370, or in
surveillance capsules the smaller sizes of test specimens are
recommended.Where it is not possible to use specimens of the PracticeE 185.Bothbasemetalandweldmetalspecimenswill
be taken.Aset of tension specimens shall consist of three each
recommended size, the least deviation possible from recom-
mended sizes should be adhered to. Non-standard specimens base metal and weld metal.
E531
5.5 Creep and Stress-Rupture in Specimens—The type and 0.2 µm R (8 µin. AA) surface roughness. After polishing, all
a
size of specimen to be used shall be the same as those used for remaining grinding and polishing marks shall be longitudinal
tension specimens except that button-head or pinned-end and any circumferential grinding marks must be removed. The
specimens are recommended for high-temperature testing. finished specimens shall be degreased in suitable solvent.
Practice E 453 describes the attention that must be paid to Specimens to be exposed to liquid sodium shall not be
specimen alignment and dimensional tolerances. One set of degreased in halogenated solvents. If surface observations are
tests shall be conducted at the operating temperature of the to be made, the test specimen may be electropolished in
component of interest. A set shall comprise a minimum of six accordance with Methods E 3. Test specimens that are suscep-
stress rupture tests at six different stress levels. The stress tible to corrosion in room-temperature air shall be stored as
levels should be selected so that the time-to-rupture ranges soon as practicable after preparation in an inert dry gas or
from not less than 100 h to at least 3000 h. Creep strain vacuum. A set of specimens shall consist of ten each of base
measurements may be made if desired. metal and weld metal.
5.6 Low-CycleFatigueSpecimens—For base metal the type 5.7 Swelling Specimens—The swelling specimens shall be
and size of specimen to be used may be the “hourglass” type right circular cylinders 5.0 mm diameter and 10.0 mm long.
with threaded or button-head ends as shown in Fig. 1. For weld The sharp-cornered specimens shall be finished by turning or
metal specimens the uniform gage type may be used. Machin- grinding and have a surface finish of 0.2 µm R (8 µin. AA).
a
ing and polishing of the test specimens shall be performed with Thespecimensshallbedegreasedinsuitablesolventandstored
care so as to minimize the effects of specimen preparation on in an inert gas or vacuum.
fatigue life. In the final stages of machining, material shall be 5.8 Charpy Impact Specimens—The specimens to be used
removed in the radial thickness amounts of 0.2 mm until 0.1 shall conform to the Charpy V-notch specimens recommended
mm remains. After exposure the final 0.075 mm shall be in Test Methods E 23. The notch shall not be formed prior to
removed by cylindrical grinding at no more than 0.005 mm per exposure as a surveillance specimen. The location and orien-
pass. The final 0.025 mm shall be buffed and finished with an tation of test specimens shall be as defined in Practice E 185
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
and Test Methods and Definitions A 370. A set of specimens surveillance specimens should be carefully evaluated with
shallbemadeupoftwelveeachbasemetal,heat-affectedzone, respect to the test results and the conclusions drawn therefrom.
and weld metal. 6.5 Specimen Withdrawal Schedule—The specimen with-
drawalscheduleshallbeasspecifiedinPracticeE 185,CaseB,
6. Irradiation Conditions
based on the percentage of component life as specified inTable
6.1 Introduction—The intent of the section on irradiation
1.
conditions is to provide guidance on how to place surveillance
samples to obtain the desired irradiation conditions in terms of
7. Measurement of Neutron Exposure
temperature, neutron flux, and neutron spectrum to ensure a
7.1 The neutron flux and neutron energy spectrum at the
realistic evaluation of the component that the surveillance
surveillance location shall be given as well as the method used
specimen is representing.
for determination. All assumptions should be clearly stated,
6.2 Irradiation Temperature—It is very important that ad-
and all physical constants, such as cross sections, half lives,
equate consideration be given to test specimens to ensure that
fission yields, etc., should be listed. The neutron flux level
they experience the correct temperatu
...

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Standard Practice for Investigating the Effects of Neutron Radiation Damage Using Charged-Particle Irradiation

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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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4.3 This practice relates to the generation of irradiation-induced changes in the microstructure of metals and alloys using charged particles. The investigation of mechanical behavior using charged particles is covered in Practice E821.
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Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials

SIGNIFICANCE AND USE
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.  
4.1.1 In the absence of surveillance data for a given reactor material (see Practice E185 and E2215), the use of calculative procedures are necessary to make the prediction. Even when credible surveillance data are available, it will usually be necessary to interpolate or extrapolate the data to obtain a TTS  for a specific time in the plant operating life. The embrittlement correlation presented herein has been developed for those purposes.  
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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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