ASTM E900-02(2007)

Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)

SIGNIFICANCE AND USE
Operation of commercial power reactors must conform to pressure-temperature limits during heatup and cooldown to prevent over-pressurization at temperatures that might cause nonductile behavior in the presence of a flaw. Radiation damage to the reactor vessel beltline region is compensated for by adjusting the pressure-temperature limits to higher temperature as the neutron damage accumulates. The present practice is to base that adjustment on the increase in transition temperature produced by neutron irradiation as measured at the Charpy V-notch 30-ft·lbf (41-J) energy level. To establish pressure temperature operating limits during the operating life of the plant, a prediction of adjustment in transition temperature must be made.
4.1.1 In the absence of surveillance data for a given reactor (see Practice E 185), the use of calculative procedures will be necessary to make the prediction. Even when credible surveillance data are available, it will usually be necessary to extrapolate the data to obtain an adjustment in transition temperature for a specific time in the plant operating life. The embrittlement correlation presented herein has been developed for those purposes.
Research has established that certain elements, notably copper and nickel, cause a variation in radiation sensitivity of steels. The importance of other elements, such as phosphorus (P), remains a subject of additional research. Copper and nickel are the key chemistry parameters used in developing the calculative procedures described here.
Only power reactor surveillance data were used in the derivation of these procedures. The measure of fast neutron fluence used in the procedure is n/cm2  (E > 1 MeV). Differences in the neutron fluence rate and neutron energy spectra experienced in power reactors and test reactors have not been applied in these procedures. The manner in which these factors were considered is addressed elsewhere.3
SCOPE
1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 30-ftlbf (41-J) data. Radiation damage calculative procedures have been developed from a statistical analysis of an irradiated material database that was available as of May 2000. The embrittlement correlation used in this guide was developed using the following variables: copper and nickel contents, irradiation temperature, and neutron fluence. The form of the model was based on current understanding for two mechanisms of embrittlement: stable matrix damage (SMD) and copper-rich precipitation (CRP); saturation of copper effects (for different weld materials) was included. This guide is applicable for the following specific materials, copper, nickel, and phosphorus contents, range of irradiation temperature, and neutron fluence based on the overall database:
1.1.1 MaterialsA 533 Type B Class 1 and 2, A302 Grade B, A302 Grade B (modified), A508 Class 2 and 3.
Submerged arc welds, shielded arc welds, and electroslag welds for materials in .
1.1.2 Copper contents within the range from 0 to 0.50 wt %.
1.1.3 Nickel content within the range from 0 to 1.3 wt %.
1.1.4 Phosphorus content within the range 0 to 0.025 wt %.
1.1.5 Irradiation exposure temperature within the range from 500 to 570F (260 to 299C).
1.1.6 Neutron fluence within the range from 1 1016 to 8 1019 n/cm2  (E > 1 MeV).
1.1.7 Neutron energy spectra within the range expected at the reactor vessel core beltline region of light water cooled reactors and fluence rate within the range from 2 108 to 1 1012 n/cm2s (E > 1 MeV).
1.2 The basis for the method of adjusting the reference temperature is discussed in a separate report.
1.3 This guide is Part IIF of Master Matrix E 706 which coordinates several standards used for irradiation surveillance of light-water reactor vessel materials. Methods of determining the applicable fluence for ...

General Information

• Status: Historical
• Publication Date: 14-Jul-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

Relations

• Replaces: ASTM E900-02 - Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)
• Effective Date: 15-Jul-2007
• Replaced By: ASTM E900-15 - Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
• Effective Date: 01-Feb-2015
• Referred By: ASTM E853-23 - Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results
• Effective Date: 15-Jul-2007
• Referred By: ASTM E1005-21 - Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
• Effective Date: 15-Jul-2007
• Referred By: ASTM E185-21 - Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 15-Jul-2007
• Referred By: ASTM E1214-11(2023) - Standard Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance
• Effective Date: 15-Jul-2007
• Referred By: ASTM E2215-19 - Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 15-Jul-2007
• Referred By: ASTM E910-18 - Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance
• Effective Date: 15-Jul-2007
• Referred By: ASTM E1006-21 - Standard Practice for Analysis and Interpretation of Physics Dosimetry Results from Test Reactor Experiments
• Effective Date: 15-Jul-2007
• Referred By: ASTM E509/E509M-21 - Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels
• Effective Date: 15-Jul-2007
• Referred By: ASTM E2956-23 - Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels
• Effective Date: 15-Jul-2007

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: E900 − 02(Reapproved 2007)
Standard Guide for
Predicting Radiation-Induced Transition Temperature Shift
in Reactor Vessel Materials, E706 (IIF)
This standard is issued under the fixed designation E900; the number immediately following the designation indicates the year of
original adoption or, in the case of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope 1.2 The basis for the method of adjusting the reference
temperature is discussed in a separate report.
1.1 This guide presents a method for predicting reference
transition temperature adjustments for irradiated light-water 1.3 This guide is Part IIF of Master Matrix E706 which
cooled power reactor pressure vessel materials based on coordinates several standards used for irradiation surveillance
Charpy V-notch 30-ft·lbf (41-J) data. Radiation damage calcu- oflight-waterreactorvesselmaterials.Methodsofdetermining
lative procedures have been developed from a statistical the applicable fluence for use in this guide are addressed in
analysisofanirradiatedmaterialdatabasethatwasavailableas Master Matrix E706, Practices E560 (IC) and Guide E944
ofMay2000. Theembrittlementcorrelationusedinthisguide (IIA), and Test Method E1005 (IIIA). The overall application
wasdevelopedusingthefollowingvariables:copperandnickel of these separate guides and practices is described in Practice
contents, irradiation temperature, and neutron fluence. The E853 (IA).
form of the model was based on current understanding for two
1.4 The values given in customary U.S. units are to be
mechanisms of embrittlement: stable matrix damage (SMD)
regarded as the standard. The SI values given in parentheses
and copper-rich precipitation (CRP); saturation of copper
are for information only.
effects (for different weld materials) was included. This guide
1.5 This standard guide does not define how the shift in
is applicable for the following specific materials, copper,
transition temperature should be used to determine the final
nickel, and phosphorus contents, range of irradiation
adjusted reference temperature. (That would typically include
temperature,andneutronfluencebasedontheoveralldatabase:
consideration of the initial starting point, the predicted shift,
1.1.1 Materials:
and the uncertainty in the shift estimation method.)
1.1.1.1 A533 Type B Class 1 and 2, A302 Grade B, A302
Grade B (modified), A508 Class 2 and 3. 1.6 This standard does not purport to address all of the
1.1.1.2 Submerged arc welds, shielded arc welds, and elec- safety concerns, if any, associated with its use. It is the
responsibility of the user of this standard to establish appro-
troslag welds for materials in 1.1.1.1.
1.1.2 Coppercontentswithintherangefrom0to0.50wt%. priate safety and health practices and determine the applica-
bility of regulatory limitations prior to use.
1.1.3 Nickel content within the range from 0 to 1.3 wt%.
1.1.4 Phosphorus content within the range 0 to 0.025 wt%.
2. Referenced Documents
1.1.5 Irradiation exposure temperature within the range
from 500 to 570°F (260 to 299°C).
2.1 ASTM Standards:
1.1.6 Neutron fluence within the range from 1 × 10 to8×
E185Practice for Design of Surveillance Programs for
19 2
10 n/cm (E > 1 MeV).
Light-Water Moderated Nuclear Power Reactor Vessels
1.1.7 Neutron energy spectra within the range expected at
E560Practice for Extrapolating Reactor Vessel Surveillance
the reactor vessel core beltline region of light water cooled
Dosimetry Results, E 706(IC) (Withdrawn 2009)
reactors and fluence rate within the range from2×10 to1×
E693Practice for Characterizing Neutron Exposures in Iron
12 2
10 n/cm s (E > 1 MeV).
and Low Alloy Steels in Terms of Displacements Per
1 3
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear Charpy Embrittlement Correlations—Status of Combined Mechanistic and
Technology and Applications and is the direct responsibility of Subcommittee Statistical Bases for U.S. Pressure Vessel Steels (MRP-45), PWR Materials
E10.02 on Behavior and Use of Nuclear Structural Materials. Reliability Program (PWRMRP), EPRI, Palo Alto, CA, 2001, 1000705.
Current edition approved July 15, 2007. Published August 2007. Originally For referenced ASTM standards, visit the ASTM website, www.astm.org, or
approved in 1983. Last previous edition approved in 2002 as E900–02. DOI: contact ASTM Customer Service at [e-mail]. For Annual Book of ASTM
10.1520/E0900-02R07. Standards volume information, refer to the standard’s Document Summary page on
The Charpy surveillance data were originally obtained from the Oak Ridge the ASTM website.
National Laboratory Power Reactor-Embrittlement Database (PR-EDB) and subse- The last approved version of this historical standard is referenced on
quently updated by ASTM Subcommittee E10.02, May 2000. www.astm.org.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
E900 − 02 (2007)
Atom (DPA), E 706(ID) 3.1.7 SMD—the stable matrix damage term of the transition
E706MasterMatrixforLight-WaterReactorPressureVessel temperature shift equation and is based on an assumed under-
Surveillance Standards, E 706(0) (Withdrawn 2011) standing of matrix damage mechanisms in RPV steels.
E853PracticeforAnalysisandInterpretationofLight-Water
3.1.8 T —irradiationtemperatureatfullpower,in°F,andis
c
Reactor Surveillance Results, E706(IA)
the estimated time-weighted average (based on the mean
E944Guide for Application of Neutron Spectrum Adjust-
temperature over each fuel cycle) cold leg temperature for
ment Methods in Reactor Surveillance, E 706 (IIA)
PWRs and recirculation temperature for BWRs.
E1005Test Method for Application and Analysis of Radio-
3.1.9 TTS—the predicted mean value of the transition tem-
metric Monitors for Reactor Vessel Surveillance, E 706
perature shift from the correlation.
(IIIA)
4. Significance and Use
3. Terminology
4.1 Operation of commercial power reactors must conform
3.1 Definitions of Terms Specific to This Standard:
to pressure-temperature limits during heatup and cooldown to
3.1.1 A, B—materialfittingcoefficientsthatareafunctionof
prevent over-pressurization at temperatures that might cause
material type.
nonductile behavior in the presence of a flaw. Radiation
damagetothereactorvesselbeltlineregioniscompensatedfor
3.1.2 best-estimate chemical composition—the best-
byadjustingthepressure-temperaturelimitstohighertempera-
estimatechemicalcomposition(copper[Cu]andnickel[Ni],in
ture as the neutron damage accumulates. The present practice
wt%) may be established using one of the following methods:
istobasethatadjustmentontheincreaseintransitiontempera-
(1) Use a simple mean for a small set of uniformly distributed
tureproducedbyneutronirradiationasmeasuredattheCharpy
data; that is, sum the measurements and divide by the number
V-notch 30-ft·lbf (41-J) energy level. To establish pressure
of measurements; (2) Use a weighting process for a non-
temperature operating limits during the operating life of the
uniformly distributed data set, especially when the number of
plant,apredictionofadjustmentintransitiontemperaturemust
measurements from one source are much greater in terms of
be made.
material volume analyzed. For a plate, a unique sample could
4.1.1 In the absence of surveillance data for a given reactor
be a set of test specimens taken from one corner of the plate.
(see Practice E185), the use of calculative procedures will be
For a weldment, a unique sample would be a set of test
necessary to make the prediction. Even when credible surveil-
specimens taken from a unique weld deposit made with a
lance data are available, it will usually be necessary to
specific electrode heat. A simple mean is calculated for test
extrapolate the data to obtain an adjustment in transition
specimens comprising each unique sample, the sample means
temperature for a specific time in the plant operating life. The
are then summed, and the sum is divided by the number of
embrittlement correlation presented herein has been developed
unique samples to get the sample weighted mean; (3) Use an
for those purposes.
alternative weighting scheme when other factors have a sig-
nificant influence and a physical model can be established. For
4.2 Research has established that certain elements, notably
the preceding, the best estimate for the sample should be used
copper and nickel, cause a variation in radiation sensitivity of
if evaluating surveillance data from that sample.
steels. The importance of other elements, such as phosphorus
3.1.2.1 Discussion—For cases where no chemical analysis
(P),remainsasubjectofadditionalresearch.Copperandnickel
measurements are available for a heat of material, the upper
are the key chemistry parameters used in developing the
limitingvaluesgiveninthematerialspecificationstowhichthe
calculative procedures described here.
vessel was built may be used.Alternately, generic mean values
4.3 Only power reactor surveillance data were used in the
for the class of material may be used.
derivation of these procedures. The measure of fast neutron
3.1.2.
...