ASTM E636-10

Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)

SIGNIFICANCE AND USE
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 in order to generate additional specific fracture toughness property 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 fracture toughness, notch ductility, and yield strength 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 postirradiation 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.

General Information

• Status: Historical
• Publication Date: 28-Feb-2010
• ICS: 27.120.10 - Reactor engineering
• Technical Committee: E10 - Nuclear Technology and Applications
• Drafting Committee: E10.02 - Behavior and Use of Nuclear Structural Materials

Relations

• Replaces: ASTM E636-09 - Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)
• Effective Date: 01-Mar-2010
• Replaced By: ASTM E636-14 - Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)
• Effective Date: 01-Jan-2014
• Referred By: ASTM E1005-21 - Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
• Effective Date: 01-Mar-2010
• Referred By: ASTM E2215-19 - Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 01-Mar-2010
• Referred By: ASTM E185-21 - Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 01-Mar-2010
• Referred By: ASTM E509/E509M-21 - Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels
• Effective Date: 01-Mar-2010

Standards Content (Sample)

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: E636 − 10
StandardGuide for
Conducting Supplemental Surveillance Tests for Nuclear
1
Power Reactor Vessels, E 706 (IH)
This standard is issued under the fixed designation E636; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision.Anumber in parentheses indicates the year of last reapproval.A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope from Light-Water Moderated Nuclear Power ReactorVes-
sels
1.1 This guide discusses test procedures that can be used in
E2298Test Method for Instrumented Impact Testing of
conjunction with, but not as alternatives to, those required by
Metallic Materials
Practices E185 and E2215 for the surveillance of nuclear
3
2.2 ASME Standards:
reactor vessels. The supplemental mechanical property tests
ASME Boiler and PressureVessel Code, Section IIISubsec-
outlined permit the acquisition of additional information on
tion NB (Class 1 Components)
radiation-induced changes in fracture toughness, notch
ductility, and yield strength properties of the reactor vessel
3. Significance and Use
steels.
3.1 PracticesE185andE2215describeaminimumprogram
1.2 This guide provides recommendations for the prepara-
for the surveillance of reactor vessel materials, specifically
tion of test specimens for irradiation, and identifies special
mechanical property changes that occur in service. This guide
precautions and requirements for reactor surveillance opera-
may be applied in order to generate additional specific fracture
tions and postirradiation test planning. Guidance on data
toughness property information on radiation-induced property
reduction and computational procedures is also given. Refer-
changes to better assist the determination of the optimum
ence is made to other ASTM test methods for the physical
reactor vessel operation schemes.
conduct of specimen tests and for raw data acquisition.
4. Supplemental Mechanical Property Test
2. Referenced Documents
4.1 Fracture Toughness Test—This test involves the dy-
2
2.1 ASTM Standards:
namicorstatictestingofafatigue-precrackedspecimenduring
E23Test Methods for Notched Bar Impact Testing of Me-
which a record of force versus displacement is used to
tallic Materials
determine material fracture toughness properties such as the
E185Practice for Design of Surveillance Programs for
plane strain fracture toughness (K ), the J-integral fracture
Ic
Light-Water Moderated Nuclear Power Reactor Vessels
toughness (J ), the J-R curve, and the reference temperature
Ic
E399Test Method for Linear-Elastic Plane-Strain Fracture
(T )(seeTestMethodsE399,E1820,andE1921,respectively).
0
Toughness K of Metallic Materials
Ic
These test methods generally apply to elastic, ductile-to-brittle
E1253Guide for Reconstitution of Irradiated Charpy-Sized
transition, or fully plastic behavior. The rate of specimen
Specimens
loading or stress intensity increase required for test classifica-
E1820Test Method for Measurement of FractureToughness
tion as quasi-static or dynamic is indicated by the referenced
E1921 Test Method for Determination of Reference
test methods. All three test methods specify a lower limit on
Temperature, T , for Ferritic Steels in the Transition
o
loading rate for dynamic tests.
Range
4.2 Fracture Toughness Test at Impact Loading Rates—This
E2215Practice for Evaluation of Surveillance Capsules
test involves impact testing of Charpy V-notch specimens that
havebeenfatigueprecracked.Aforceversusdeflectionortime
record, or both, is obtained during the test to determine an
1
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear
Technology and Applicationsand is the direct responsibility of Subcommittee estimate of material dynamic fracture toughness properties.
E10.02 on Behavior and Use of Nuclear Structural Materials.
Currently, no standard test method is available for performing
Current edition approved March 1, 2010. Published April 2010. Originally
andanalyzingthistest;detailsontherecommendedprocedures
approved in 1983. Last previous edition approved in 2009 as E636–09. DOI:
are given in 7.1–7.4 and Appendix Appendix X1.
10.1520/E0636-10.
2
For referenced ASTM standards, visit the ASTM website, www.astm.org, or
contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM
3
Standards volume information, refer to the standard’s Document Summary page on Available from American Society of Mechanical Engineers, 345 E. 47th St.,
the ASTM website. New York, NY 10017.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
1

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E636 − 10
4.3 Instrumented Charpy V-Notch Test—This test involves
theimpacttestingofstandardCh
...

This document is not anASTM standard and is intended only to provide the user of anASTM standard an indication of what changes have been made to the previous version. Because
it may not be technically possible to adequately depict all changes accurately, ASTM recommends that users consult prior editions as appropriate. In all cases only the current version
of the standard as published by ASTM is to be considered the official document.
Designation:E636–09 Designation: E636 – 10
Standard Guide for
Conducting Supplemental Surveillance Tests for Nuclear
1
Power Reactor Vessels, E 706 (IH)
This standard is issued under the fixed designation E636; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision.Anumber 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 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 fracture toughness, notch ductility, and yield
strength properties of the reactor vessel steels.
1.2 Thisguideprovidesrecommendationsforthepreparationoftestspecimensforirradiation,andidentifiesspecialprecautions
and requirements for reactor surveillance operations and postirradiation test planning. Guidance on data reduction and
computational procedures is also given. Reference is made to other ASTM and ISO test methods for the physical conduct of
specimen tests and for raw data acquisition.
2. Referenced Documents
2
2.1 ASTM Standards:
E23 Test Methods for Notched Bar Impact Testing of Metallic Materials
E185 Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
E399 Test Method for Linear-Elastic Plane-Strain Fracture Toughness K of Metallic Materials
Ic
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 Practice
for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
E2298 Test Method for Instrumented Impact Testing of Metallic Materials
3
2.2 Other Standards: ASME Standards:
ASME Boiler and Pressure Vessel Code, Section IIISubsection NB (Class 1 Components)
ISO 14556 Steel Charpy V-notch Pendulum Impact Test- Instrumented Test Method Subsection NB (Class 1 Components)
3. 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 in order to generate additional specific fracture
toughness property information on radiation-induced property changes to better assist the determination of the optimum reactor
vessel operation schemes.
4. Supplemental Mechanical Property Test
4.1 Fracture Toughness Test—This test involves the dynamic or static testing of a fatigue-precracked specimen during which
a record of force versus displacement is used to determine material fracture toughness properties such as the plane strain fracture
toughness (K ), the J-integral fracture toughness (J ), the J-R curve, and the reference temperature (T ) (see Test Methods E399,
Ic Ic 0
E1820, and E1921, respectively). These test methods generally apply to elastic, ductile-to-brittle transition, or fully plastic
behavior. The rate of specimen loading or stress intensity increase required for test classification as quasi-static or dynamic is
indicated by the referenced test methods. All three test methods specify a lower limit on loading rate for dynamic tests.
1
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 Feb.March 1, 2009.2010. Published March 2009.April 2010. Originally approved in 1983. Last previous edition approved in 20022009 as
E636–029. DOI: 10.1520/E0636-109.
2
ForreferencedASTMstandards,visittheASTMwebsite,www.astm.org,[e-mail]ards
volume information, refer to the standard’s Document Summary page on the ASTM website.
3
Available from American Society of Mechanical Engineers, 345 E. 47th St., New York, NY 10017.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.
1

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E6
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