U.S. NUCLEAR REGULATORY COMMISSION 

Revision 2 

May 1988 

REGULATORY GUIDE 

OFFICE OF NUCLEAR REGULATORY RESEARCH 

REGULATORY GUIDE 1.99 

(Task ME 305-41 

RADIATION EMBRITTLEMENT OF REACTOR VESSEL MATERIALS 

A. INTRODUCTION 

General Design Criterion 31, "Fracture Prevention of Reactor 

Coolant Pressure Boundary,'' of Appendix A, ''General Design 

Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Domestic 

Licensing of Production and Utilization Facilities," requires, in 

part, that the reactor coolant pressure boundary be designed with 

sufficient margin to ensure that, when stressed under operating, 

maintenance, testing, and postulated accident conditions, (1) the 

boundary behaves in a nonbrittle manner and (2) the probability 

of rapidly propagating fracture is minimized. General Design 

Criterion 31 also requires that the design reflect the uncertainties 

in determining the effects of irradiation on material properties. 

Appendix G, "Fracture Toughness Requirements," and Appendix 

H, "Reactor Vessel Material Surveillance Program Requirements," 

which implement, in part, Criterion 31, necessitate the calculation 

of changes in fracture toughness of reactor vessel materials caused 

by neutron radiation throughout the service life. This guide describes 

general procedures acceptable to the NRC staff for calculating the 

effects of neutron radiation embrittlement of the low-alloy steels 

currently used for light-water-cooled reactor vessels. 

The calculative procedures given in Regulatory Position 1.1 of 

this guide are not the same as those given in the Pressurized Thermal 

Shock rule(§ 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," of 10 CFR 

Part 50) for calculating RTPTS, the reference temperature that is 

to be compared to the screening criterion given in the rule. The 

information on which this Revision 2 is based may also affect the 

basis for the PTS rule. The staff is presently considering whether 

to propose a change to § 50.61. 

The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory 

position. 

Any information collection activities mentioned in this regulatory 

guide are contained as requirements in 10 CFR Part 50, which provides the regulatory basis for this guide. The information collecUSN RC REGULATORY GUIDES 

Regulatory Guides are Issued to describe and make available to the 

public methods acceptable to the N RC staff of implementing specific parts of the Commission's regulations, to delineate techniques used bY the staff In evaluating specific problems or postu- lated accidents, or to provide guidance to applicants. Regulatory Guides are not substitutes for regulations, and compliance with 

them Is not required. Methods and solutions different from those set 

out in the guides will be acceptable if they provide a basis for the 

findings requisite to the Issuance or continuance of a permit or license by the commission. 

This guide was issued after consideration of comments received from 

the public. Comments and suggestions for Improvements In these 

guides are encouraged at all times, and guides will be revised, as appropriate, to accommodate comments and to reflect new Information or experience. 

Written comments may be submitted to the Rules and Procedures 

.Branch, DRR, ADM, U.S. Nuclear Regulatory Commission, 

Washington, DC 20555. 

tion requirements in 10 CFR Part 50 have been cleared under 0MB 

Clearance No. 3150-0011. 

B. DISCUSSION 

Some NRC requirements that necessitate calculation of radiation embrittlement are: 

1. Paragraph V .A of Appendix G requires the effects of neutron 

radiation to be predicted from the results of pertinent radiation effects 

studies. This guide provides such results in the form of calculative 

procedures that are acceptable. to the ·NRc. 

2. Paragraph V .B of Appendix G describes the basis for setting 

the upper limit for pressure as a function of temperature during 

heatup and cooldown for a given service period in terms of the 

predicted value of the adjusted reference temperature at the end 

of the service period. 

3. The definition of reactor vessel beltline given in Paragraph 

II. F of Appendix G requires identification of regions of the reactor 

vessel that are predicted to experience sufficient neutron radiation 

embrittlement to be considered in the selection of the most limiting 

material. Paragraphs ill.A and N.A.1 specify the additional test 

requirements for beltline materials that supplement the requirements 

for reactor vessel materials generally. 

4. Paragraph 11.B of Appendix H incorporates ASTM E 185 

by reference. Paragraph 5.1 of ASTM E 185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled 

Nuclear Power Reactor Vessels'' (Ref. 1), requires that the materials 

to be placed in surveillance be those that may limit operation of 

the reactor during its lifetime, i.e., those expected to have the highest 

adjusted reference temperature or the lowest Charpy upper-shelf 

energy at end of life. Both measures of radiation embrittlement must 

be considered. In Paragraph 7.6 of ASTM E 185-82, the requirements for the number of capsules and the withdrawal schedule are 

based on the calculated amount of radiation embrittlement at end 

of life. 

The guides are issued in the following ten broad divisions: 

1. Power Reactors 6. Products 

2. Research and Test Reactors 7. Transportation 3. Fuels and Materials Facllltles 8. Occupational Health 

4. Environmental and Siting 9. Antitrust and Financial Review 

5. Materials and Plant Protection 10. General 

Copies of Issued guides may be purchased from the Government 

Printing Office at the current GPO price. Information on current 

GPO prices may be obtained by contacting the Superlntendent·of Documents, U.S. Government Printing Office, Post Office Box 

37082, Washington, DC 20013-7082, telephone (202)275-2060 or (202)275-2171. 

Issued guides may also be purchased from the National Technical 

Information Service on a standing order basis. Details on this 

service may be obtained by writing NTIS, 5285 P.ort Royal Road, 

Springfield, VA 22161. 
The two measures of radiation embrittlement used in this guide 

are obtained from the results of the Charpy V-notch impact test. 

Appendix G to 10 CFR Part 50 requires that a full curve of absorbed 

energy versus temperature be obtained through the ductile-to-brittle 

transition temperature region. The adjustment of the reference 

temperature, ARTNDT, is defined in Appendix Gas the temperature shift in the Charpy curve for the irradiated material relative 

to that for the unirradiated material measured at the 30-foot-pound 

energy level, and the data that formed the basis for this guide were 

30-foot-pound shift values. The second measure of radiation 

embrittlement is the decrease in the Charpy upper-shelf energy level, 

which is defined in ASTM E 185-82. This Revision 2 updates the 

calculative procedures for the adjustment of reference temperature; 

however, calculative procedures for the decrease in upper-shelf 

energy are unchanged because the preparatory work had not been 

completed in time to include them in this revision. 

The basis for Equation 2 for ARTNDT (in Regulatory Position 

1.1 of this guide) is contained in publications by G. L. Guthrie (Ref. 

2) and G. R. Odette et al. (Ref. 3). Both of these papers used 

surveillance data from commercial power reactors. The bases for 

their regression correlations were different in that Odette made 

greater use of physical models of radiation embrittlement. Yet, the 

two papers contain similar recommendations: (1) separate correlation functions should be used for weld and base metal, (2) the function should be the product of a chemistry factor and a tluence factor, 

(3) the parameters in the chemistry factor should be the elements 

copper and nickel, and (4) the tluence factor should provide a trend 

curve slope of about 0.25 to 0.30 on log-log paper at 1019 n/cm2 

(E > 1 MeV), steeper at low tluences and flatter at high fluences. 

Regulatory Position 1. 1 is a blend of the correlation functions 

presented by these authors. Some test reactor data were used as 

a guide in establishing a cutoff for the chemistry factor for lowcopper materials. The data base for Regulatory Position 1.2 is that 

given by Spencer H. Bush (Ref. 4). 

The measure of fluence used in this guide is the number of 

neutrons per square centimeter having energies greater than 1 million 

electron volts (E > 1 Me V). The differences in energy spectra at 

the surveillance capsule and the vessel inner surface locations do 

not appear to be great enough to warrant the use of a damage function such as displacements per atom (dpa) (Ref. 5) in the analysis 

of the surveillance data base (Ref. 6). 

However, the neutron energy spectrum does change significantly 

with location in the vessel wall; hence for calculating the attenuation of radiation embrittlement through the vessel wall, it is 

necessary to use a damage function to determine ARTNDT versus 

radial distance into the wall. The most widely accepted damage function at this time is dpa, and the attenuation formula (Equation 3) 

given in Regulatory Position 1.1 is based on the attenuation of dpa 

through the vessel wall. 

Sensitivity to neutron radiation embrittlement may be affected 

by elements other than copper and nickel. The original version and 

Revision 1 of this guide had a phosphorus term in the chemistry 

factor, but the studies on which this revision was based found other 

elements such as phosphorus to be of secondary importance, i.e., 

including them in the analysis did not produce a significantly better fit of the data. 

Scatter in the data base used for this guide is relatively significant, as evidenced by the fact that the standard deviations for 

Guthrie's derived formulas (Ref. 2) are 28 °F for welds and 17 °F 

for base metal despite extensive efforts to find a model that reduced 

the fitting error. Thus the use of surveillance data from a given 

reactor (in place of the calculative procedures given in this guide) 

requires considerable engineering judgment to evaluate the credibility of the data and assign suitable margins. When surveillance data 

from the reactor in question become available, the weight given 

to them relative to the information in this guide will depend on the 

credibility of the surveillance data as judged by the following 

criteria: 

1. Materials in the capsules should be those judged most likely 

to be controlling with regard to radiation embrittlement according 

to the recommendations of this guide. 

2. Scatter in the plots of Charpy energy versus temperature for 

the irradiated and unirradiated conditions should be small enough 

to permit the determination of the 30-foot-pound temperature and 

the upper-shelf energy unambiguously. 

3. When there are two or more sets of surveillance data from 

one reactor, the scatter of ARTNDT values about a best-fit line 

drawn as described in RegulatQry Position 2.1 normally should be 

less than 28 °P for welds and 17 °P for base metal. Even if the fluence 

range is large (two or more orders of magnitude), the scatter should 

not exceed twice those values. Even if the data fail this criterion 

for use in shift calculations, they may be credible for determining 

decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E 185-82 (Ref. 1). 

4. The irradiation temperature of_ the Charpy specimens in the 

capsule should match vessel wall temperature at the cladding/base 

metal interface within ± 25 °P. 

5. The surveillance data for the correlation monitor material in 

the capsule should fall within the scatter band of the data base for 

that material. 

To use the surveillance data from a specific plant instead of 

Regulatory Position 1, one must develop a relationship of ARTNDT 

to fluence for that plant. Because such data are limited in number 

and subject to scatter, Regulatory Position 2 describes a procedure 

-in which the form of Equation 2 is to be used and the fluence factor therein is retained, but the chemistry factor is determined by 

the plant surveillance data. Of several possible ways to fit such data, 

the method that minimizes the sums of the squares of the errors 

was chosen somewhat arbitrarily. Its use is justified in part by the 

fact that "least squares" is a common method for curve fitting. 

Also, when there are only two data points, the least squares method 

gives greater weight to the point with the higher ARTNOT; this 

seems reasonable for fitting surveillance data, because generally 

the higher data point will be the more recent and therefore will represent more modem procedures. 

C. REGULATORY POSITION 

1. SURVEILLANCE DATA NOT AVAILABLE 

When credible surveillance data from the reactor in question 

are not available, calculation of neutron radiation embrittlement of 

the beltline of reactor vessels of light-water reactors should be based 

on the procedures in Regulatory Positions 1.1 and 1.2 within the 

limitations in Regulatory Position 1.3. 

1.99-2 
1.1 Adjusted Reference Temperature 

The adjusted reference temperature (ART) for each material in 

the beltline is given by the following expression: 

ART = Initial RTNDT + ARJNDT + Margin (1) 

Initial RTNDT is the reference temperature for the unirradiated 

material as defined in Paragraph NB-2331 of Section ill of the 

ASME Boiler and Pressure Vessel Code (Ref. 7). If measured values 

of initial RTNDT for the material in question are not available, 

generic mean values for that class* of material may be used if there 

are sufficient test results to establish a mean and standard deviation for the class. 

ARTNDT is the mean value of the adjustment in reference 

temperature caused by irradiation and should be calculated as 

follows: 

ARTNDT = (CF) f(0.28 - 0.10 log f) (2) 

CF ( °F) is the chemistry factor, a function of copper and nickel 

content. CF is given in Table 1 for welds and in Table 2 for base 

metal (plates and forgings). Linear interpolation is permitted. In 

Tables 1 and 2 "weight-percent copper" and "weight-percent 

nickel" are the best-estimate values for the material, which will 

normally be the mean of the measured values for a plate or forging 

or for weld samples made with the weld wire heat number that 

matches the critical vessel weld. If such values are not available, 

the upper limiting values given in the material specifications to which 

the vessel was built may be used. If not available, conservative 

estimates (mean plus one standard deviation) based on generic data 

may be used if justification is provided. If there is no information 

available, 0.35% copper and 1.0% nickel should be assumed. 

The neutron fluence at any depth in the vessel wall, f (1019 n/cm2, 

E > 1 MeV), is determined as follows: 

f = fsurf (e -0.24x) (3) 

where fsurf (1019 n/cm2 , E > 1 MeV) is the calculated value of 

the neutron fluence at the inner wetted surface of the vessel at the 

location of the postulated defect, and x (in inches) is the depth into 

the vessel wall measured from the vessel inner (wetted) surface. 

Alternatively, if dpa calculations are made as part of the fluence 

analysis, the ratio of dpa at the depth in question to dpa at the inner 

surface may be substituted for the exponential attenuation factor 

in Equation 3. 

The fluence factor, f0.28 - 0.10 log f, is determined by calculation or from Figure 1. 

''Margin'' is the quantity, °F, that is to be added to obtain conservative, upper-bound values of adjusted reference temperature 

for the calculations required by Appendix G to 10 CPR Part 50. 

Margin = 2 V a2 + a 2 

I A (4) 

"The class for estimating initial RTNDT is generally determined, for the welds 

with which this guide is concerned, by the type of welding flux (Linde 80 or other); 

for base metal, by the ASTM Standard Specification. 

Here, 01 is the standard deviation for the initial RTNDT· If a 

measured value of initial RTNDT for the material in question is 

available, OJ is to be estimated from the precision of the test method. 

If not, and generic mean values for that class of material are used, 

OJ is the standard deviation obtained from the set of data used to 

establish the mean. 

The standard deviation for ARTNDT, a A, is 28 °F for welds and 

17 °F for base metal, except that OA, need not exceed 0.50 times 

the mean value of ARTNDT· 

1.2 Charpy Upper-Shelf Energy 

Charpy upper-shelf energy should be assumed to decrease as 

a function of fluence and copper content as indicated in Figure 2. 

Linear interpolation is permitted. 

1.3 Limitations 

Application of the foregoing procedures should be subject to 

the following limitations: 

1. The procedures apply to those grades of SA-302, 336, 533, 

and 508 steels having minimum specified yield strengths of 50,000 

psi and under and to their welds and heat-affected zones. 

2. The procedures are valid for a nominal irradiation temperature 

of 550 °F. Irradiation below 525 °F should be considered to produce greater embrittlement, and irradiation above 590 °F may be 

considered to produce less embrittlement. The correction factor used 

should be justified by reference to actual data. 

3. Application of these procedures to fluence levels or to copper or nickel content beyond the ranges given in Figure 1 and Tables 

1 and 2 or to materials having chemical compositions beyond the 

range found in the data bases used for this guide should be justified 

by submittal of data. 

2. SURVEILLANCE DATA AVAILABLE 

When two or more credible surveillance data sets (as defined 

in the Discussion) become available from the reactor in question, 

they may be used to determine the adjusted reference temperature 

and the Charpy upper-shelf energy of the beltline materials as 

described in Regulatory Positions 2.1 and 2.2, respectively. 

2.1 Adjusted Reference Temperature 

The adjusted reference temperature should be obtained as 

follows. First, if there is clear evidence that the copper or nickel 

content of the surveillance weld differs from that of the vessel weld, 

i.e., differs from the average for the weld wire heat number 

associated with the vessel weld and the surveillance weld, the 

measured values of ARTNDT should be adjusted by multiplying 

them by the ratio of the chemistry factor for the vessel weld to that 

for the surveillance weld. Second, the surveillance data should be 

fitted using Equation 2 to obtain the relationship of ARTNDT to 

fluence. To do so, calculate the chemistry factor, CF, for the best 

fit by multiplying each adjusted ARTNDT by its corresponding 

fluence factor, summing the products, and dividing by the sum of 

the squares of the fluence factors. The resulting value of CF when 

entered in Equation 2 will give the relationship of ARTNDT to 

1.99-3 
TABLE 1 

CHEMISTRY FACTOR FOR WELDS, °F 

Copper, Nickel, Wt-% 

Wt-% 0 0.20 0.40 0.60 0.80 1.00 1.20 

0 20 20 20 20 20 20 20 

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 

fluence that fits the plant surveillance data in such a way as to to obtain mean values of shift, ARTNDT· In calculating the margin, 

minimize the sum of the squares of the errors. the value of a A may be reduced from the values given in the last 

paragraph of Regulatory Position 1.1 by an amount to be decided 

To calculate the margin in this case, use Equation 4; the values on a case-by-case basis, depending on where the measured values 

given there for a}. may be cut in half. fall relative to the mean calculated for the surveillance materials. 

If this procedure gives a higher value of adjusted reference 

temperature than that given by using the procedures of Regulatory 2.2 Charpy Upper-Shelf Energy 

Position 1.1, the surveillance data should be used. If this procedure 

gives a lower value, either may be used. The decrease in upper-shelf energy may be obtained by plotting the reduced plant surveillance data on Figure 2 of this guide 

For plants having surveillance data that are credible in all respects and fitting the data with a line drawn parallel to the existing lines 

except that the material does not represent the critical material in as the upper bound of all the data. This line should be used in 

the vessel, the calculative procedures in this guide should be used preference to the existing graph. 

1.99-4 
TABLE 2 

CHEMISTRY FACTOR FOR BASE METAL, °F 

Copper, 

Wt-% 0 0.20 0.40 

0 20 20 20 

0.01 20 20 20 

0.02 20 20 20 

0 .03 20 20 20 

0.04 22 26 26 

0.05 25 31 31 

0.06 28 37 37 

0.07 31 43 44 

0.08 34 48 51 

0.09 37 53 58 

0.10 41 58 65 

0.11 45 62 72 

0.12 49 67 79 

0.13 53 71 85 

0.14 57 75 91 

0.15 61 80 99 

0.16 65 84 104 

0.17 69 88 110 

0.18 73 92 115 

0.19 78 97 120 

0.20 82 102 125 

0.21 86 107 129 

0.22 91 112 134 

0.23 95 117 138 

0.24 100 121 143 

0.25 104 126 148 

0.26 109 130 151 

0.27 114 134 155 

0.28 119 138 160 

0.29 124 142 164 

0.30 129 146 167 

0.31 134 151 172 

0.32 139 155 175 

0.33 144 160 180 

0.34 149 164 184 

0.35 153 168 187 

0.36 158 173 191 

0.37 162 177 196 

0.38 166 182 200 

0.39 171 185 203 

0.40 175 189 207 

3. REQUIREMENT FOR NEW PLANTS 

For beltline materials in the reactor vessel for a new plant, the 

content of residual elements such as copper, phosphorus, sulfur, 

and vanadium should be controlled to low levels.* The copper content should be such that the calculated adjusted reference temperature 

at the 1/4T position in the vessel wall at end of life is less than 

200 °F. In selecting the optimum amount of nickel to be used, its 

deleterious effect on radiation embrittlement should be balanced 

against its beneficial metallurgical effects and its tendency to lower 

the initial RTNDT· 

*For more infonnation, see the Appendix to ASTM Standard Specification A 533 

(Ref. 8). 

1.99-5 

Nickel, Wt-% 

0.60 0.80 1.00 1.20 

20 20 20 20 

20 20 20 20 

20 20 20 20 

20 20 20 20 

26 26 26 26 

31 31 31 31 

37 37 37 37 

44 44 44 44 

51 51 51 51 

58 58 58 58 

65 67 67 67 

74 77 77 77 

83 86 86 86 

91 96 96 96 

100 105 106 106 

110 115 117 117 

118 123 125 125 

127 132 135 135 

134 14i 144 144 

142 150 154 154 

149 159 164 165 

155 167 172 174 

161 176 181 184 

167 184 190 194 

172 191 199 204 

176 199 208 214 

180 205 216 221 

184 211 225 230 

187 216 233 239 

191 221 241 248 

194 225 249 257 

198 228 255 266 

202 231 260 274 

205 234 264 282 

209 238 268 290 

212 241 272 298 

216 245 275 303 

220 248 278 308 

223 250 281 313 

227 254 285 317 

231 257 288 320 

D. IMPLEMENTATION 

The purpose of this section is to provide information to applicants 

and licensees regarding the NRC staff's plans for using this 

regulatory guide. Except in those cases in which an applicant proposes an acceptable alternative method for complying with specified 

portions of the Commission's regulations, the methods described 

in this guide will be used as follows: 

1. The methods described in Regulatory Positions 1 and 2 of 

this guide will be used by the NRC staff in evaluating all predictions of radiation embrittlement needed to implement Appendices 

G and H to 10 CFR Part 50. 
2. Holders of licenses and permits should use the methods 

described in this guide to predict the effect of neutron radiation on 

reactor vessel materials as required by Paragraph V .A of Appendix G to 10 CFR Part 50, unless they can justify the use of different methods. The use of the Revision 2 methodology may result 

in a modification of the pressure-temperature limits contained in 

Technical Specifications in order to continue to satisfy the 

requirements of Section V of Appendix G, 10 CFR Part 50. 

3. The recommendations of Regulatory Position 3 are essentially unchanged from those used to evaluate construction permit 

applications docketed on or after June 1, 1977. 

1.99-6 
j 

J 

REFERENCES 

1. American Society for Testing and Materials, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled 

Nuclear Power Reactor Vessels," ASTM E 185-82, July 1982. * 

2. G. L. Guthrie, "Charpy Trend Curves Based on 177 PWR Data 

Points," in "L WR Pressure Vessel Surveillance Dosimetry Improvement Program," NUREG/CR-3391, Vol. 2, prepared by 

Hanford Engineering Development Laboratory, HEDL-TME 

83-22, April 1984. "'* 

3. G. R. Odette et al., "Physically Based Regression Correlations 

of Embrittlement Data from Reactor Pressure Vessel 

Surveillance Programs," Electric Power Research Institute, 

NP-3319, January 1984. t 

4. S. H. Bush, "Structural Materials for Nuclear Power Plants," 

in Journal of Testing and Evaluation, American Society for 

Testing and Materials, November 1974. * 

5. American Society for Testing and Materials, "Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in 

Terms of Displacements per Atom (DPA)," ASTM E 693-79, 

August 1979. * 

6. W. N. McElroy, "LWR Pressure Vessel Surveillance Dosimetry 

Improvement Program: LWR Power Reactor Surveillance 

Physics-Dosimetry Data Base Compendium,'' NUREG/ 

CR-3319, prepared by Hanford Engineering Development 

Laboratory, HEDL-TME 85-3, August 1985.** 

7. American Society of Mechanical Engineers, Section m, 

"Nuclear Power Plant Components," of ASME Boiler and 

Pressure Vessel Code, New York (updated frequently). tt 

8. American Society for Testing and Materials, "Standard 

Specification for Pressure Vessel Plates, Alloy Steel, Quenched 

and Tempered, Manganese-Molybdenum and ManganeseMolybdenum-Nickel,'' ASTM A 533/A 533M-82, September 

1982.* 

*Copies may be oblained from the American Society for Testing and Materials, 1916 Race Street, Philadelphia, PA 19103. 

**Copies may be obtained from the Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Wasington, DC 20013-7082. 

tcopies may be obtained from the Electric Power Research Institute, 3412 Hillview Avenue, Palo Alto, CA 94304. 

ttcopies may be obtained from the American Society of Mechanical Engineers, 345 E. 47th Street, New York, NY 10017. 

1.99-7 
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2 3 4 5 6 7 8 91 

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40 

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C: 

,, I I I 

=!' ,c-.=-J::±:, ·::r- +j :··,_-t-l-+ 

-t-f- 'I ►- -- _.._ __ •t-- N -- - -- --- 1·:=: " H 

-

- + 

- Q) 

en iD cg '° Q) \0 ... 

(.) 

...... --

- £ i ~~ ,__,_ ' ... ' i 

I+ I=++~- :-,.-

- -- - t- 1-- c,:: ~ r:;: 1--1-- ..,.. 

a, 

0 

+- r-r-l+f+r + ,_ --- ' CL 

' t:±: t::::J.·1--"- ' 

- . ' ' I ·t--i--- t-+ • -!- H- , ~ ·'.~jl~B~ ,-1- 4 - - ' [H± ci IT 11 h, ~~ 1-t-+ ➔:- . I I ~ ' • - I--+---,- -~ ~ 

-.L+--H-- :+ '. , ... - oc:::a . -

r .- ,_ + ,---~ - . ., : 11-::::' ~ = , ; . --+-+- ....J - ~tf --~~ ,_ i--H It -, -- ' H ~: -~ 

I 

- l 

1----H-r It-I- H 

- ~ 

h-t-+1 tT 

2X1017 4 6 8 1018 2 4 6 8 1019 2 4 6 

FLUENCE, n/cm2 (E > 1MeV) 

FIGURE 2 Predicted Decrease in Shelf Energy as a Function of Copper Content and Fluence 
REGULATORY ANALYSIS 

A copy of the regulatory analysis prepared for this Regulatory 

Guide 1. 99, Revision 2, is available for inspection and copying for 

a fee at the Commission's Public Document Room at 1717 H Street 

NW., Washington, DC, under Regulatory Guide 1.99, Revision 2. 

1.99-10