ASTM E900 - 21

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Designation: E900 −21
Standard Guide for
Predicting Radiation-Induced Transition Temperature Shift
in Reactor Vessel Materials
1
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 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 correla-
tion:
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 from1×10
21
n/m
2
to2×10
24
n/m
2
(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 elec-
troslag welds having compositions consistent with those of the
welds used to join the base materials described in 1.1.2.1.
1.1.2.3 Neutron ﬂuence rate within the range from3×10
12
n/m
2
/sto5×10
16
n/m
2
/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 elec-
tric).
1.1.2.5 Irradiation exposure times of up to 25 years in
boiling water reactors and 31 years in pressurized water
reactors.
1.2 It is the responsibility of the user to show that the
conditions of interest in their application of this guide are
addressed adequately by the technical information on which
the guide is based. It should be noted that the conditions
quantiﬁed by the database are not distributed evenly over the
range of materials and irradiation conditions described in 1.1,
and that some combination of variables, particularly at the
extremes of the data range are under-represented. Particular
attention is warranted when the guide is applied to conditions
near the extremes of the data range used to develop the TTS
equation and when the application involves a region of the data
space where data is sparse. Although the embrittlement corre-
lation developed for this guide was based on statistical analysis
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 Sept. 1, 2021. Published October 2021. Originally
approved in 1983. Last previous edition approved in 2015 as E900 – 15
ɛ2
. DOI:
10.1520/E0900-21.
2
Available from ASTM International Headquarters. Order Adjunct No.
ADJE090015-EA.
3
To inform the TTS prediction of Section 5of this guide, the E10.02
Subcommittee decided to limit the data considered to Charpy shift values (∆T
41J
)
measured from 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 Cu, Ni, P, and Mn.
Reports and technical papers documenting the results of research programs
conducted in material test reactors were also reviewed. Data from these sources was
included in the database for information, but was not used in the development of the
TTS prediction of Section 5of this guide.
Copyright © ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States
This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the
Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.
1

of a large database, prudence is required for applications that
involve variable values beyond the ranges speciﬁed in 1.1. Due
to strong correlations with other exposure variables within the
database (that is, ﬂuence), and due to the uneven distribution of
data within the database (for example, the irradiation tempera-
ture and ﬂux range of PWR and BWR data show almost no
overlap) neither neutron ﬂuence rate nor irradiation time
sufficiently improved the accuracy of the predictions to merit
their use in the embrittlement correlation in this guide. Future
versions of this guide may incorporate the effect of neutron
ﬂuence rate or irradiation time, or both, on TTS, as such effects
are described in (1).
4
The irradiated material database, the
technical basis for developing the embrittlement correlation,
and issues involved in its application, are discussed in a
separate report (ADJE090015-EA). That report describes the
nine different TTS equations considered in the development of
this guide, some of which were developed using more limited
datasets (for example, national program data (2,3)). If the
material variables or exposure conditions of a particular
application fall within the range of one of these alternate
correlations, it may provide more suitable guidance.
1.3 This guide is expected to be used in coordination with
several standards addressing irradiation surveillance of light-
water reactor vessel materials. Method of determining the
applicable ﬂuence for use in this guide are addressed in Guides
E482,E944, and Test Method E1005. The overall application
of these separate guides and practices is described in Practice
E853.
1.4 The values stated in SI units are to be regarded as
standard. The values given in parentheses after SI units are
provided for information only and are not considered standard.
1.5 This standard guide does not deﬁne how the TTS should
be used to determine the ﬁnal adjusted reference temperature,
which would typically include consideration of the transition
temperature before irradiation, the predicted TTS, and the
uncertainties in the shift estimation method.
1.6 This standard does not purport to address all of the
safety concerns, if any, associated with its use. It is the
responsibility of the user of this standard to establish appro-
priate safety, health, and environmental practices and deter-
mine the applicability of regulatory limitations prior to use.
1.7 This international standard was developed in accor-
dance with internationally recognized principles on standard-
ization established in the Decision on Principles for the
Development of International Standards, Guides and Recom-
mendations issued by the World Trade Organization Technical
Barriers to Trade (TBT) Committee.
2. Referenced Documents
2.1 ASTM Standards:
5
A302 Speciﬁcation for Pressure Vessel Plates, Alloy Steel,
Manganese-Molybdenum and Manganese-Molybdenum-
Nickel
A508 Speciﬁcation for Quenched and Tempered Vacuum-
Treated Carbon and Alloy Steel Forgings for Pressure
Vessels
A533 Speciﬁcation for Pressure Vessel Plates, Alloy Steel,
Quenched and Tempered, Manganese-Molybdenum and
Manganese-Molybdenum-Nickel
E185 Practice for Design of Surveillance Programs for
Light-Water Moderated Nuclear Power Reactor Vessels
E482 Guide for Application of Neutron Transport Methods
for Reactor Vessel Surveillance
E693 Practice for Characterizing Neutron Exposures in Iron
and Low Alloy Steels in Terms of Displacements Per
Atom (DPA)
E853 Practice for Analysis and Interpretation of Light-Water
Reactor Surveillance Neutron Exposure Results
E944 Guide for Application of Neutron Spectrum Adjust-
ment Methods in Reactor Surveillance
E1005 Test Method for Application and Analysis of Radio-
metric Monitors for Reactor Vessel Surveillance
E2215 Practice for Evaluation of Surveillance Capsules
from Light-Water Moderated Nuclear Power Reactor Ves-
sels
2.2 ASTM Adjunct:
2
ADJE090015-EA Technical Basis for the Equation Used to
Predict Radiation-Induced Transition Temperature Shift in
Reactor Vessel Materials
3. Terminology
3.1 Deﬁnitions of Terms Speciﬁc to This Standard:
3.1.1 best-estimate chemical composition—the best-
estimate chemical composition (copper [Cu], nickel [Ni],
phosphorus [P], and manganese [Mn] in %) may be established
using one of the following methods: (1) Use a simple mean for
a small set of uniformly distributed data; that is, sum the
measurements and divide by the number of measurements; (2)
Use a weighting process for a non-uniformly distributed data
set, especially when the number of measurements from one
source are much greater in terms of material volume analyzed.
For a plate, a unique sample could be a set of test specimens
taken from one corner of the plate. For a weldment, a unique
sample would be a set of test specimens taken from a unique
weld deposit made with a speciﬁc electrode heat. A simple
mean is calculated for test specimens comprising each unique
sample, the sample means are then added, and the sum is
divided by the number of unique samples to get the sample
weighted mean; (3) Use an alternative weighting scheme when
other factors have a signiﬁcant inﬂuence and a physical model
can be established. For the preceding, the best estimate for the
sample should be used if evaluating surveillance data from that
sample.
3.1.1.1 Discussion—For cases where no chemical analysis
measurements are available for a heat of material, the upper
limiting values given in the material speciﬁcations to which the
vessel was built may be used. Alternately, generic mean values
for the class of material may be used.
4
The boldface numbers in parentheses refer to a list of references at the end of
this standard.
5
For referenced ASTM standards, visit the ASTM website, www.astm.org, or
contact ASTM Customer Service at service@astm.org. For Annual Book of ASTM
Standards volume information, refer to the standard’s Document Summary page on
the ASTM website.
E900 − 21
2

3.1.1.2 Discussion—In all cases where engineering judg-

ment is used to select a best estimate copper, nickel,

phosphorus, or manganese content, the rationale shall be

documented which formed the basis for the selection.

3.1.2 ﬂuence (Φ)—in this guide the term “ﬂuence” refers to

the fast (E > 1MeV) neutron ﬂuence, that is, the number of

neutrons per square meter with energy greater than 1.0 MeV at

the location of interest.

3.1.3 ﬂuence rate(Φ)—in this guide the term “ﬂuence rate”

refers to the fast (E > 1MeV) neutron ﬂuence rate, that is, the

number of neutrons per square meter per unit time with energy

greater than 1.0 MeV at the location of interest. This is also

referred to as fast neutron ﬂux.

3.1.4 SRM—Standard Reference Material. Also known as

correlation monitor material.

3.1.5 T—irradiation temperature at full power, in °C, given

by the estimated time-weighted average (based on the mean

temperature over each fuel cycle) cold leg temperature for

pressurized water reactors (PWRs) and recirculation tempera-

ture for boiling water reactors (BWRs).

3.1.6 TTS—the mean value of transition temperature shift

predicted by the embrittlement correlation.

4. 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 chem-

istry 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/m

(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.

5. Calculative Procedure for Transition Temperature

Shift (TTS )

5.1 The mean value of TTS, in °C, is calculated as follows

using the embrittlement correlation developed in the ASTM

adjunct (see 2.2):

TTS 5TTS

1TTS

(1)

where:

TTS

5A·5

9· 1.8943 310

212

·Φ

0.5695

1.8 · T132

550

25.47

0.09 1P

0.012

0.216

1.66 1Ni

8.54

0.63

0.39

1.36

0.3

(2)

1.011 for forgings

1.080 for plates and SRM plates

0.919 for welds

(3)

and:

TTS

9· max

min

Cu, 0.28

20.053,0

·M(4)

M5B· max

min

113.87

2ln

4.5 310

, 612.6

1.8·T132

550

25.45

0.1 1P

0.012

20.098

0.168 1Ni

0.58

0.63

0.73

(5)

0.738 for forgings

0.819 for plates and SRM plates

0.968 for welds

(6)

5.2 In the equations of 5.1, Cu, Ni, P, and Mn are all

expressed in weight percent, Φis in n/m

(E> 1 MeV), and Tis

in °C.

6. Attenuation Through the Vessel Wall

6.1 The embrittlement correlation listed in 5.1 was derived

from surveillance specimens withdrawn from PWRs and

BWRs after exposure at positions within the region between

the core and the pressure vessel wall. It is applicable to neutron

ﬂuence spectra that are similar to those found in those

locations.

6.2 To calculate the TTS at some location other than the

region adjacent to the active core between the core and the

pressure vessel, it is necessary to account for the variation in

neutron energy spectrum and ﬂuence around the vessel or

through the vessel wall. Due to these variations, the use of the

local ﬂuence (E> 1 MeV) may give a non-conservative

estimate of the neutron damage. Therefore, it is recommended

E900 − 21

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摘要：

ASTM E900 - 21 是一项由美国材料与试验协会（ASTM）发布的标准指南，专门用于核反应堆压力容器材料辐照监测中的Charpy V型缺口冲击能量转变参考温度的计算方法。该标准提供了基于统计分析的公式，用于预测铁素体钢在中子辐照环境下韧性-脆性转变温度的位移，是核电站安全评估和寿命管理中的关键工具。ASTM E900 - 21 更新了先前的版本，整合了更广泛的实验数据和修正系数，适用于反应堆压力容器材料的辐照脆化趋势分析，帮助工程师确定材料在长期运行中的性能变化。此标准广泛应用于核能行业

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