ASTM E900-15

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.  
4.2 Research has established that certain elements, notably copper (Cu), nickel (Ni), phosphorus (P), and manganese (Mn), cause a variation in radiation sensitivity of reactor pressure vessel steels. The importance of other elements, such as silicon (Si), and carbon (C), remains a subject of additional research. Copper, nickel, phosphorus, and manganese are the key chemistry parameters used in developing the calculative procedures described here.  
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 (1).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 arc welds, a...

General Information

• Status: Historical
• Publication Date: 31-Jan-2015
• 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(2007) - Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)
• Effective Date: 01-Feb-2015
• Replaced By: ASTM E900-15e1 - Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
• Effective Date: 01-Feb-2015
• Referred By: ASTM E910-18 - Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance
• Effective Date: 01-Feb-2015
• Referred By: ASTM E1214-11(2023) - Standard Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance
• Effective Date: 01-Feb-2015
• Referred By: ASTM E185-21 - Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 01-Feb-2015
• Referred By: ASTM E1006-21 - Standard Practice for Analysis and Interpretation of Physics Dosimetry Results from Test Reactor Experiments
• Effective Date: 01-Feb-2015
• Referred By: ASTM E2956-23 - Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels
• Effective Date: 01-Feb-2015
• Referred By: ASTM E1005-21 - Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
• Effective Date: 01-Feb-2015
• Referred By: ASTM E509/E509M-21 - Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels
• 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: 01-Feb-2015
• Referred By: ASTM E2215-19 - Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 01-Feb-2015

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 − 15
Standard Guide for
Predicting Radiation-Induced Transition Temperature Shift
1
in Reactor Vessel Materials
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.Anumber in parentheses indicates the year of last reapproval.A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1. Scope 1.1.1.3 Phosphorus content up to 0.03 %.
1.1.1.4 Manganesecontentwithintherangefrom0.55to2.0
1.1 This guide presents a method for predicting values of
%.
reference transition temperature shift (TTS) for irradiated
1.1.1.5 Irradiationtemperaturewithintherangefrom255to
pressure vessel materials. The method is based on the TTS
300°C (491 to 572°F).
exhibited by Charpy V-notch data at 41-J (30-ft·lbf) obtained
21 2
1.1.1.6 Neutronfluencewithintherangefrom1×10 n/m
from surveillance programs conducted in several countries for
24 2
to2×10 n/m (E> 1 MeV).
commercialpressurized(PWR)andboiling(BWR)light-water
1.1.1.7 A categorical variable describing the product form
cooled (LWR) power reactors. An embrittlement correlation
(that is, weld, plate, forging).
has been developed from a statistical analysis of the large
1.1.2 The range of material and irradiation conditions in
surveillance database consisting of radiation-induced TTS and
the database for variables not included in the embrittlement
related information compiled and analyzed by Subcommittee
correlation:
E10.02. The details of the database and analysis are described
2,3,
1.1.2.1 A533 Type B Class 1 and 2, A302 Grade B, A302
in a separate report (1). This embrittlement correlation was
developed using the variables copper, nickel, phosphorus, Grade B (modified), and A508 Class 2 and 3. Also, European
and Japanese steel grades that are equivalent to these ASTM
manganese,irradiationtemperature,neutronfluence,andprod-
uct form. Data ranges and conditions for these variables are Grades.
listed in 1.1.1. Section 1.1.2 lists the materials included in the 1.1.2.2 Submerged arc welds, shielded arc welds, and elec-
database and the domains of exposure variables that may troslag welds having compositions consistent with those of the
influence TTSbutarenotusedintheembrittlementcorrelation. welds used to join the base materials described in 1.1.2.1.
12
1.1.1 The range of material and irradiation conditions in
1.1.2.3 Neutron fluence rate within the range from3×10
2 16 2
the database for variables used in the embrittlement correla-
n/m/sto5×10 n/m /s (E > 1 MeV).
tion:
1.1.2.4 Neutron energy spectra within the range expected at
1.1.1.1 Copper content up to 0.4 %.
the reactor vessel region adjacent to the core of commercial
1.1.1.2 Nickel content up to 1.7 %.
PWRs and BWRs (greater than approximately 500MW elec-
tric).
1.1.2.5 Irradiation exposure times of up to 25 years in
1
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear
boiling water reactors and 31 years in pressurized water
Technology and Applications and is the direct responsibility of Subcommittee
reactors.
E10.02 on Behavior and Use of Nuclear Structural Materials.
Current edition approved Feb. 1, 2015. Published April 2015. Originally
1.2 It is the responsibility of the user to show that the
approvedin1983.Lastpreviouseditionapprovedin2007asE900–02(2007).DOI:
conditions of interest in their application of this guide are
10.1520/E0900-15.
2
The boldface numbers in parentheses refer to a list of references at the end of addressed adequately by the technical information on which
this standard.
the guide is based. It should be noted that the conditions
3
To inform the TTS prediction of Section 5 of this guide, the E10.02
quantified by the database are not distributed evenly over the
Subcommittee decided to limit the data considered to Charpy shift values (∆T )
41J
range of materials and irradiation conditions described in 1.1,
measured from irradiations conducted in PWRs and BWRs. A database of 1,878
Charpy TTSmeasurementswascompiledfromsurveillancereportsonoperatingand
and that some combination of variables, particularly at the
decommissioned light water reactors of Western design from 13 countries (Brazil,
extremes of the data range are under-represented. Particular
Belgium, France, Germany, Italy, Japan, Mexico, The Netherlands, South Korea,
attention is warranted when the guide is applied to conditions
Sweden, Switzlerland, Taiwan, and the United States), and from the technical
near the extremes of the data range used to develop the TTS
literature. For each data record, the following information had to be available:
fluence, fluence rate, irradiation temperature, and % content of Cu, Ni, P, and Mn.
equationandwhentheapplicationinvolvesaregionofthedata
Repo
...

This document is not an ASTM standard and is intended only to provide the user of an ASTM 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: E900 − 02 (Reapproved 2007) E900 − 15
Standard Guide for
Predicting Radiation-Induced Transition Temperature Shift
1
in Reactor Vessel Materials, E706 (IIF)Materials
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
1.1 This guide presents a method for predicting values of reference transition temperature adjustments shift (TTS) for irradiated
light-water cooled power reactor pressure vessel materials based on pressure vessel materials. The method is based on the
TTSCharpy V-notch 30-ft·lbf (41-J) data. Radiation damage calculative procedures have 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
an irradiated material database that was available as of Maythe large surveillance database consisting of radiation-induced
2000.TTS and related information compiled and analyzed by Subcommittee E10.02. The details of the database and analysis are
2,3,
described in a separate report (1). TheThis embrittlement correlation used in this guide was developed using the following
variables: copper and nickel contents, variables copper, nickel, phosphorus, manganese, irradiation temperature, and neutron
fluence. The form of the model was based on current understanding for two mechanisms of embrittlement: stableneutron fluence,
and product form. Data ranges and conditions for these variables are listed in 1.1.1matrix damage (SMD). Section 1.1.2 and
copper-rich precipitation (CRP); saturation of copper effects (for different weldlists the materials included in the database and the
domains of exposure variables that may influence materials)TTS 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: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).
21 2 24 2
1.1.1.6 Neutron fluence within the range from 1 × 10 n/m to 2 × 10 n/m (E> 1 MeV).
1.1.1.7 A categorical variable describing the product form (that is, weld, plate, forging).
1.1.2 Materials: 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, A302A302 Grade B, A302A302 Grade B (modified), A508and 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 arc welds, and electroslag welds for materials having compositions consistent with those
of the welds used to join the base materials described in 1.1.1.11.1.2.1.
12 2 16 2
1.1.2.3 Neutron fluence rate within the range from 3 × 10 n/m /s to 5 × 10 n/m /s (E > 1 MeV).
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 July 15, 2007Feb. 1, 2015. Published August 2007April 2015. Originally approved in 1983. Last previous edition approved in 20022007 as
E900 – 02.E900 – 02(2007). DOI: 10.1520/E0900-02R07.10.1520/E0900-15.
2
The Charpy surveillance data were originally obtained from the Oak Ridge National Laboratory Power Reactor-Embrittlement Database (PR-EDB) and subsequently
updated by ASTM Subcommittee E10.02, May 2000.boldface numbers in parentheses refer to a list of references at the end of this standard.
3
To inform the CharpyTTS Embrittlement Correlations—Status of Combinedprediction of Section 5 Mechanistic and Statistical Bases for U.S. Pressure Vessel Steels
(MRP-45), PWR Materials Reliability Program (PWRMRP),of this guide, the E10.02 Subco
...