ASTM E900-15e2
(Guide)Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
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 (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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´2
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
ε NOTE—Old Ref 1 was editorially removed and the adjunct information was updated and added to Section 2, Refer-
enced Documents, in April 2017.
2
ε NOTE—In 3.1.2 and 3.1.3, “neutrons per square centimeter” was corrected editorially to “neutrons per square meter” to
reflect the usage in Section 5 in July 2020.
1. Scope 1.1.1 The range of material and irradiation conditions in
the database for variables used in the embrittlement correla-
1.1 This guide presents a method for predicting values of
tion:
reference transition temperature shift (TTS) for irradiated
1.1.1.1 Copper content up to 0.4 %.
pressure vessel materials. The method is based on the TTS
1.1.1.2 Nickel content up to 1.7 %.
exhibited by Charpy V-notch data at 41-J (30-ft·lbf) obtained
1.1.1.3 Phosphorus content up to 0.03 %.
from surveillance programs conducted in several countries for
1.1.1.4 Manganesecontentwithintherangefrom0.55to2.0
commercialpressurized(PWR)andboiling(BWR)light-water
%.
cooled (LWR) power reactors. An embrittlement correlation
1.1.1.5 Irradiationtemperaturewithintherangefrom255to
has been developed from a statistical analysis of the large
300°C (491 to 572°F).
surveillance database consisting of radiation-induced TTS and
21 2
1.1.1.6 Neutronfluencewithintherangefrom1×10 n/m
related information compiled and analyzed by Subcommittee
24 2
to2×10 n/m (E> 1 MeV).
E10.02. The details of the database and analysis are described
2,3
1.1.1.7 A categorical variable describing the product form
in a separate report (ADJE090015-EA). This embrittlement
(that is, weld, plate, forging).
correlation was developed using the variables copper, nickel,
1.1.2 The range of material and irradiation conditions in
phosphorus, manganese, irradiation temperature, neutron
the database for variables not included in the embrittlement
fluence,andproductform.Datarangesandconditionsforthese
variables are listed in 1.1.1. Section 1.1.2 lists the materials correlation:
includedinthedatabaseandthedomainsofexposurevariables 1.1.2.1 A533 Type B Class 1 and 2, A302 Grade B, A302
that may influence TTS but are not used in the embrittlement Grade B (modified), and A508 Class 2 and 3. Also, European
correlation. and Japanese steel grades that are equivalent to these ASTM
Grades.
1.1.2.2 Submerged arc welds, shielded arc welds, and elec-
1
This guide is under the jurisdiction of ASTM Committee E10 on Nuclear
troslag welds having compositions consistent with those of the
Technology and Applications and is the direct responsibility of Subcommittee
welds used to join the base materials described in 1.1.2.1.
E10.02 on Behavior and Use of Nuclear Structural Materials.
12
1.1.2.3 Neutron fluence rate within the range from3×10
Current edition approved Feb. 1, 2015. Published April 2015. Originally
2 16 2
approvedin1983.Lastpreviouseditionapprovedin2007asE900–02(2007).DOI: n/m/sto5×10 n/m /s (E > 1 MeV).
10.1520/E0900-15E02.
1.1.2.4 Neutron energy spectra within the range expected at
2
Available from ASTM International Headquarters. Order Adjunct No.
the reactor vessel region adjacent to the core of commercial
ADJE090015-EA.
3
To inform the TTS prediction of Section 5 of this guide, the E10.02 PWRs and BWRs (greater than approximately 500MW elec-
Subcommittee decided to limit the data considered to Charpy shift values (∆T )
41J tric).
measured from irradiations conducted in PWRs and BWRs. A database of 1,878
1.1.2.5 Irradiation exposure times of up to 25 years in
Charpy TTSmeasurementswascompiledfromsurveillancereportsonoperatingand
boiling water reactors and 31 years in pressurized water
decommissioned light water reactors of Western design from 13 countries (Brazil,
Belgium, France, Germany, Italy, Japan, Mexico, The Netherlands, South Korea,
reactors.
Sweden, Switzlerland, Taiwan, and the United States), and from the technical
1.2 It is the responsibility of the user to show that the
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.
conditions of interest in their application of this guide are
Reports and technical papers documenting the results of research programs
addressed adequately by the technical information on which
conductedinmaterialtestreactorswerealsoreviewed.Datafromthesesourceswas
the guide is based. It should be
...
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.
´2 ´2
Designation: E900 − 15 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. A number in parentheses indicates the year of last reapproval. A
superscript epsilon (´) indicates an editorial change since the last revision or reapproval.
1
ε NOTE—Old Ref 1 was editorially removed and the adjunct information was updated and added to Section 2, Refer-
enced Documents, in April 2017.
2
ε NOTE—In 3.1.2 and 3.1.3, “neutrons per square centimeter” was corrected editorially to “neutrons per square meter” to
reflect the usage in Section 5.
1. 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
2,3
analysis are described in a separate report (ADJE090015-EA). 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).
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 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, and electroslag welds having compositions consistent with those of the welds
used to join the base materials described in 1.1.2.1.
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).
1.1.2.4 Neutron energy spectra within the range expected at the reactor vessel region adjacent to the core of commercial PWRs
and BWRs (greater than approximately 500MW electric).
1.1.2.5 Irradiation exposure times of up to 25 years in boiling water reactors and 31 years in pressurized water reactors.
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. 1, 2015. Published April 2015. Originally approved in 1983. Last previous edition approved in 2007 as E900 – 02(2007). DOI:
10.1520/E0900-15E01.10.1520/E0900-15E02.
2
Available from ASTM International Headquarters. Order Adjunct No. ADJE090015-EA.
3
To inform the TTS prediction of Section 5 of this guide, the E10.02 Subcommittee decided to limit the data considered to Charpy shift values (ΔT ) measured from
41J
irradiations conducted in PWRs and BWRs. A database of 1,878 Charpy TTS measurements was compiled from surveillance reports on operating and decommissioned light
water reactors of Western design from 13 countries (Brazil, Belgium, France, Germany, Italy, Japan, Mexico, The Netherlands, South Korea, Sweden, Switzlerland, Taiwan,
and the United States), and from the technical literature. For each data record, the following information had to be available: fluence, fluence rate, irradiation temperature,
and % content of
...
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