ASTM E636-09

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

SIGNIFICANCE AND USE
Practices E 185 and E 2215 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 E 185 and E 2215 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 and ISO test methods for the physical conduct of specimen tests and for raw data acquisition.

General Information

• Status: Historical
• Publication Date: 31-Jan-2009
• 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 E636-02 - Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)
• Effective Date: 01-Feb-2009
• Replaced By: ASTM E636-10 - Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)
• Effective Date: 01-Mar-2010
• Referred By: ASTM E1005-21 - Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
• Effective Date: 01-Feb-2009
• Referred By: ASTM E509/E509M-21 - Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels
• Effective Date: 01-Feb-2009
• Referred By: ASTM E185-21 - Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 01-Feb-2009
• Referred By: ASTM E2215-19 - Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 01-Feb-2009

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 – 09
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. 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 2.2 Other Standards:
ASME Boiler and Pressure Vessel Code, Section III Sub-
1.1 This guide discusses test procedures that can be used in
3
section NB (Class 1 Components)
conjunction with, but not as alternatives to, those required by
ISO 14556 Steel Charpy V-notch Pendulum Impact Test-
Practices E185 and E2215 for the surveillance of nuclear
4
Instrumented Test Method
reactor vessels. The supplemental mechanical property tests
outlined permit the acquisition of additional information on
3. Significance and Use
radiation-induced changes in fracture toughness, notch ductil-
3.1 Practices E185 and E2215 describe a minimum program
ity, and yield strength properties of the reactor vessel steels.
for the surveillance of reactor vessel materials, specifically
1.2 This guide provides recommendations for the prepara-
mechanical property changes that occur in service. This guide
tion of test specimens for irradiation, and identifies special
may be applied in order to generate additional specific fracture
precautions and requirements for reactor surveillance opera-
toughness property information on radiation-induced property
tions and postirradiation test planning. Guidance on data
changes to better assist the determination of the optimum
reduction and computational procedures is also given. Refer-
reactor vessel operation schemes.
ence is made to other ASTM and ISO test methods for the
physical conduct of specimen tests and for raw data acquisi-
4. Supplemental Mechanical Property Test
tion.
4.1 Fracture Toughness Test—This test involves the dy-
namic or static testing of a fatigue-precracked specimen during
2. Referenced Documents
2
which a record of force versus displacement is used to
2.1 ASTM Standards:
determine material fracture toughness properties such as the
E23 Test Methods for Notched Bar Impact Testing of
plane strain fracture toughness (K ), the J-integral fracture
Ic
Metallic Materials
toughness (J ), the J-R curve, and the reference temperature
Ic
E185 Practice for Design of Surveillance Programs for
(T ) (seeTest Methods E399, E1820, and E1921, respectively).
0
Light-Water Moderated Nuclear Power Reactor Vessels
These test methods generally apply to elastic, ductile-to-brittle
E399 Test Method for Linear-Elastic Plane-Strain Fracture
transition, or fully plastic behavior. The rate of specimen
Toughness K of Metallic Materials
Ic
loading or stress intensity increase required for test classifica-
E1253 Guide for Reconstitution of Irradiated Charpy-Sized
tion as quasi-static or dynamic is indicated by the referenced
Specimens
test methods. All three test methods specify a lower limit on
E1820 Test Method for Measurement of Fracture Tough-
loading rate for dynamic tests.
ness
4.2 FractureToughnessTestatImpactLoadingRates—This
E1921 Test Method for Determination of Reference Tem-
test involves impact testing of Charpy V-notch specimens that
perature, T , for Ferritic Steels in the Transition Range
o
have been fatigue precracked.Aforce versus deflection or time
E2215 Practice for Evaluation of Surveillance Capsules
record, or both, is obtained during the test to determine an
from Light-Water Moderated Nuclear Power Reactor Ves-
estimate of material dynamic fracture toughness properties.
sels
Currently, no standard test method is available for performing
andanalyzingthistest;detailsontherecommendedprocedures
1
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear
are given in 7.1– 7.4 and Appendix Appendix X1.
Technology and Applications and is the direct responsibility of Subcommittee
4.3 Instrumented Charpy V-Notch Test—This test involves
E10.02 on Behavior and Use of Nuclear Structural Materials.
Current edition approved Feb. 1, 2009. Published March 2009. Originally the impact testing of standard CharpyV-notch specimens using
approved in 1983. Last previous edition approved in 2002 as E636 – 02. DOI:
10.1520/E0636-09.
2 3
For referenced ASTM standards, visit the ASTM website, www.astm.org, or Available from American Society of Mechanical Engineers, 345 E. 47th St.,
contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM New York, NY 10017.
4
Standards volume information, refer to the standard’s Document Summary page on Available from International Organization for Standardization (ISO), 1 rue de
the ASTM website. Varembe,
...

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:E636–02 Designation: E 636 – 09
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
1.1This1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required
byPracticesE185andE2215forthesurveillanceofnuclearreactorvessels.Thesupplementalmechanicalpropertytestsoutlined
permit the acquisition of additional information on radiation-induced changes in fracture toughness, notch ductility, and
tensileyield 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
2.1 ASTM Standards:
2
E8Test Methods for Tension Testing of Metallic Materials
2
E23 Test Methods for Notched Bar Impact Testing of Metallic Materials
E184Practice for Effects of High-Energy Neutron Radiation on the Mechanical Properties of Metallic Materials, E706 (IB)Test
Methods for Notched Bar Impact Testing of Metallic Materials
3
E185 Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels, E706 (IF) Practice
for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
2
E399 Test Method for Plane-Strain Fracture Toughness of Metallic Materials
3
E482Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)
3
E560Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, E706 (IC)
2
E616Terminology Relating to Fracture Testing
3
E706Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E706 (0)
2
E1221Test Method for Determining Plane-Strain Crack-Arrest Fracture Toughness, K , of Ferritic Steels Test Method for
1a
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
3
E1921 Test Method for Determination of Reference Temperature, T , for Ferritic Steels in the Transition Range Test Method
0
for Determination of Reference Temperature, T , for Ferritic Steels in the Transition Range
o
E2215 Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
2.2 Other Standards:
3
ASME Boiler and Pressure Vessel Code, Section III, Subsection NB (Class 1 Components)
4
ISO 14556 Steel Charpy V-notch Pendulum Impact Test- Instrumented Test Method
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 June 10, 2002. Published September 2002. Originally published as E636–83. Last previous edition E636–95.
Current edition approved Feb. 1, 2009. Published March 2009. Originally approved in 1983. Last previous edition approved in 2002 as E636–02.
2
ForreferencedASTMstandards,visittheASTMwebsite,www.astm.org,[e-mail]ards
, Vol 03.01.volume information, refer to the standard’s Document Summary page on the ASTM website.
3
Annual Book of ASTM Standards, Vol 12.02.
3
Available from American Society of Mechanical Engineers, 345 E. 47th St., New York, NY 10017.
4
Available from American Society of Mechanical Engineers, 345 E. 47th St., New York, NY 10017.
4
Available from International Organization for Standardization (ISO), 1 rue de Varembe, Case postale 56, CH-1211, Geneva 20, Switzerland.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.
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E636–09
3. Significance and Use
3.1 Practices E185 describes and E2215 describe a minimum program for the surveillance of
...