ASTM E900-87(1994)

Standard Guide for Predicting Neutron Radiation Damage to Reactor Vessel Materials, E 706 (IIF)

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 41-J (30-ftlbf) data. Radiation damage calculative procedures have been developed from a statistical analysis of an irradiated material data base that was available as of June 1982, and checked against readily available data up to August 1983. In the procedure, a chemistry factor given in tabular form as a function of copper and nickel contents, is multiplied by a fluence factor read from a graph or calculated from a formula. A difference between this guide and the earlier edition is the addition of nickel content in the chemistry factor. This guide is applicable for the following specific materials, range of irradiation temperature, neutron fluence, and fluence rate:
1.1.1 Materials
1.1.1.1 A 533 Type B Class 1 and 2, A302 Grade B, A302 Grade B (modified), A508 Class 2 and 3.
1.1.1.2 Submerged arc welds, shielded arc welds, and electroslag welds for materials in 1.1.1.1.
1.1.1.3 Weld heat-affected zones of the materials in 1.1.1.1 and 1.1.1.2.
1.1.2 Copper contents within the range from 0.01 to 0.40 weight %.
1.1.3 Nickel content within the range from 0 to 1.2 weight %.
1.1.4 Irradiation exposure temperature within the range from 530 to 590°F (277 to 310°C).
1.1.5 Neutron fluence within the range from 1 by 10 17 to 1 by 1020 n/cm2 (E > 1 MeV).
1.1.6 Neutron fluence rate and energy spectra within the range expected at the reactor vessel core beltline region of light-water cooled reactors.
1.2 The basis for the method of adjusting the reference temperature is a report describing the basis for Regulatory Guide 1.99. The report is based on the reactor vessel surveillance data and analyses described by Guthrie and Odetle and Lombrozo; the extent of that data base is indicated by the dashed lines in .
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 use in this guide are addressed in Master Matrix E 706, Practices E 560 (IC) and E944 (IIA), and Method E 1005 (IIIA). The overall application of these separate guides and practices is described in Practice E 853 (IA).
1.4 The values given in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only.
1.5 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

General Information

• Status: Historical
• Publication Date: 08-Jul-1987
• Technical Committee: E10 - Nuclear Technology and Applications
• Drafting Committee: E10.02 - Behavior and Use of Nuclear Structural Materials

Relations

• Replaced By: ASTM E900-87(2001) - Standard Guide for Predicting Neutron Radiation Damage to Reactor Vessel Materials, E 706 (IIF)
• Effective Date: 09-Jul-1987
• Referred By: ASTM E2956-23 - Standard Guide for Monitoring the Neutron Exposure of LWR Reactor Pressure Vessels
• Effective Date: 09-Jul-1987
• Referred By: ASTM E853-23 - Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results
• Effective Date: 09-Jul-1987
• Referred By: ASTM E185-21 - Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 09-Jul-1987
• Referred By: ASTM E509/E509M-21 - Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels
• Effective Date: 09-Jul-1987
• Referred By: ASTM E2215-19 - Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels
• Effective Date: 09-Jul-1987
• Referred By: ASTM E1214-11(2023) - Standard Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance
• Effective Date: 09-Jul-1987
• Referred By: ASTM E1005-21 - Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance
• Effective Date: 09-Jul-1987
• Referred By: ASTM E1006-21 - Standard Practice for Analysis and Interpretation of Physics Dosimetry Results from Test Reactor Experiments
• Effective Date: 09-Jul-1987
• Referred By: ASTM E910-18 - Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance
• Effective Date: 09-Jul-1987

Standards Content (Sample)

NOTICE: This standard has either been superseded and replaced by a new version or discontinued.
Contact ASTM International (www.astm.org) for the latest information.
Designation: E 900 – 87 (Reapproved 1994)
Standard Guide for
Predicting Neutron Radiation Damage to Reactor Vessel
Materials, E706 (IIF)
This standard is issued under the fixed designation E 900; 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 (e) indicates an editorial change since the last revision or reapproval.
1. Scope Guide 1.99. The report is based on the reactor vessel surveil-
lance data and analyses described by Guthrie and Odetle and
1.1 This guide presents a method for predicting reference
Lombrozo ; the extent of that data base is indicated by the
transition temperature adjustments for irradiated light-water
dashed lines in Tables 1 and 2.
cooled power reactor pressure vessel materials based on
1.3 This guide is Part IIF of Master Matrix E 706 which
Charpy V-notch 41-J (30-ft·lbf) data. Radiation damage calcu-
coordinates several standards used for irradiation surveillance
lative procedures have been developed from a statistical
of light-water reactor vessel materials. Methods of determining
analysis of an irradiated material data base that was available
the applicable fluence for use in this guide are addressed in
as of June 1982, and checked against readily available data up
Master Matrix E 706, Practices E 560 (IC) and E944 (IIA), and
to August 1983. In the procedure, a chemistry factor given in
Method E 1005 (IIIA). The overall application of these sepa-
tabular form as a function of copper and nickel contents, is
rate guides and practices is described in Practice E 853 (IA).
multiplied by a fluence factor read from a graph or calculated
1.4 The values given in inch-pound units are to be regarded
from a formula. A difference between this guide and the earlier
as the standard. The values given in parentheses are for
edition is the addition of nickel content in the chemistry factor.
information only.
This guide is applicable for the following specific materials,
1.5 This standard does not purport to address all of the
range of irradiation temperature, neutron fluence, and fluence
safety concerns, if any, associated with its use. It is the
rate:
responsibility of the user of this standard to establish appro-
1.1.1 Materials:
priate safety and health practices and determine the applica-
1.1.1.1 A 533 Type B Class 1 and 2, A302 Grade B, A302
bility of regulatory limitations prior to use.
Grade B (modified), A508 Class 2 and 3.
1.1.1.2 Submerged arc welds, shielded arc welds, and elec-
2. Referenced Documents
troslag welds for materials in 1.1.1.1.
2.1 ASTM Standards:
1.1.1.3 Weld heat-affected zones of the materials in 1.1.1.1
E 185 Practice for Conducting Surveillance Tests for Light-
and 1.1.1.2.
Water Cooled Nuclear Power Reactor Vessels, E706 (IF)
1.1.2 Copper contents within the range from 0.01 to 0.40
E 560 Practice for Extrapolating Reactor Vessel Surveil-
weight %.
lance Dosimetry Results, E706 (IC)
1.1.3 Nickel content within the range from 0 to 1.2 weight
E 693 Practice for Characterizing Neutron Exposures in
%.
Iron and Low-Alloy Steels in Terms of Displacements per
1.1.4 Irradiation exposure temperature within the range
Atom (DPA), E706 (ID)
from 530 to 590°F (277 to 310°C).
17 E 706 Master Matrix for Light-Water Reactor Pressure
1.1.5 Neutron fluence within the range from 1 by 10 to 1
20 2
Vessel Surveillance Standards
by 10 n/cm (E > 1 MeV).
E 853 Practice for Analysis and Interpretation of Light-
1.1.6 Neutron fluence rate and energy spectra within the
Water Reactor Surveillance Results, E706 (IA)
range expected at the reactor vessel core beltline region of
light-water cooled reactors.
1.2 The basis for the method of adjusting the reference
Randall, P. N., “Basis for Revision 2 of U.S. NRC Regulatory Guide 1.99,”
temperature is a report describing the basis for Regulatory
Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels, ASTM STP
909, 1986, pp. 149–162.
Guthrie, G. L., “Charpy Trend Curves Based on 177 Data Points,” LWR
Pressure Vessel Surveillance Dosimetry Improvement Program, Quarterly Progress
Report April 1983 through June 1983, Hanford Engineering Development Labora-
This guide is under the jurisdiction of ASTM Committee E-10 on Nuclear
tory, NUREG/CR-3391, Vol 2, HEDL-TME 83-22.
Technology and Applicationsand is the direct responsibility of Subcommittee
Odette, G. R., and Lombrozo, P. M., “Physical Based Regression Correlations
E10.02on Nuclear Materials, Components, and Environmental Effects.
of Embrittlement Data From Reactor Pressure Vessel Surveillance Programs,” EPRI
Current edition approved July 9, 1987. Published September 1987. Originally
NP-3319 Final Report, January 1984, Prepared for Electric Power Research
published as E 900 – 83. Last previous edition E 900 – 83.
Institute.
Annual Book of ASTM Standards, Vol 12.02.
Copyright © ASTM, 100 Barr Harbor Drive, West Conshohocken, PA 19428-2959, United States.
E 900
A
TABLE 1 Chemistry Factor for Welds, °F
Nickel, Weight, %
Copper,
Weight, %
0 0.20 0.40 0.60 0.80 1.00 1.20
0 20202020202020
0.01 20 20 20 20 20 20 20
0.02 21 26 27 27 27 27 27
0.03 22 35 41 41 41 41 41
0.04 24 43 54 54 54 54 54
0.05 26 49 67 68 68 68 68
0.06 29 52 77 82 82 82 82
0.07 32 55 85 95 95 95 95
0.08 36 58 90 106 108 108 108
0.09 40 61 94 115 122 122 122
0.10 44 65 97 122 133 135 135
0.11 49 68 101 130 144 148 148
0.12 52 72 103 135 153 161 161
0.13 58 76 106 139 162 172 176
0.14 61 79 109 142 168 182 188
0.15 66 84 112 146 175 191 200
0.16 70 88 115 149 178 199 211
0.17 75 92 119 151 184 207 221
0.18 79 95 122 154 187 214 230
0.19 83 100 126 157 191 220 238
0.20 88 104 129 160 194 223 245
0.21 92 108 133 164 197 229 252
0.22 97 112 137 167 200 232 257
0.23 101 117 140 169 203 236 263
0.24 105 121 144 173 206 239 268
0.25 110 126 148 176 209 243 272
0.26 113 130 151 180 212 246 276
0.27 119 134 155 184 216 249 280
0.28 122 138 160 187 218 251 284
0.29 128 142 164 191 222 254 287
0.30 131 146 167 194 225 257 290
0.31 136 151 172 198 228 260 293
0.32 140 155 175 202 231 263 296
0.33 144 160 180 205 234 266 299
0.34 149 164 184 209 238 269 302
0.35 153 168 187 212 241 272 305
0.36 158 172 191 216 245 275 308
0.37 162 177 196 220 248 278 311
0.38 166 182 200 223 250 281 314
0.39 171 185 203 227 254 285 317
0.40 175 189 207 231 257 288 320
A
t°C 5 t°F/1.8.
E 944 Guide for Application of Neutron Spectrum Adjust- 3.1.1 In the absence of surveillance data for a given reactor
ment Methods in Reactor Surveillance (IIA) (see Practice E 185), the use of calculative procedures will be
E 1005 Test Method for Application and Analysis of Radio- necessary to make the prediction. Even when credible surveil-
metric Monitors for Reactor Vessel Surveillance, E706 lance data are available, it will usually be necessary to
(IIIA) extrapolate the data to obtain an adjustment in transition
temperature for a specific time in the plant operating life. The
3. Significance and Use
fluence function presented herein has been developed for those
purposes.
3.1 Operation of commercial power reactors must conform
3.2 Research has established that certain elements, notably
to pressure-temperature limits during heatup and cooldown to
prevent over-pressurization at temperatures that might cause copper and nickel, cause a variation in radiation sensitivity of
nonductile behavior in the presence of a flaw. Radiation steels. The importance of other suspect elements remains a
damage to the reactor vessel beltline region is compensated for subject of additional research. Copper and nickel are the
by adjusting the pressure-temperature limits to higher tempera- parameters used in developing the radiation damaged calcula-
ture as the neutron damage accumulates. The present practice tive procedures.
is to base that adjustment on the increase in transition tempera- 3.3 Only power reactor surveillance data were used in the
ture produced by neutron irradiation as measured at the Charpy derivation of these procedures. The measure of fluence used in
V-notch 41-J (30-ft·lbf) energy level. To establish pressure- the procedure is n/cm (E > 1 MeV). Differences in the neutron
temperature operating limits during the operating life of the fluence rate and in neutron energy spectra experienced in
plant, a prediction of adjustment in transition temperature must power reactors and test reactors have not been considered in
be made. the development of these procedures because the technology is
E 900
A
TABLE 2 Chemistry Factor for Welds, °F
Nickel, Weight, %
Copper,
Weight, %
0 0.
...