Regulatory Guide 1.99

(Task Me 305-4) Revision 2 Radiation Embrittlement of Reactor Vessel Materials

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Issue date:	05/31/1988
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RG-1.99, Rev 2
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Revlalon 2 May 1988 U.S. NUCLEAR REGULATORY COMMISSION

iREGULATORY GUIDE

OFFICE OF NUCLEAR REGULATORY RESEARCH

REGULATORY GUIDE 1.99 (Task ME 3054)

RADIATION EMBRITrLEMENT 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, (l) 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 0, "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 Pressurizod Thermal Shock rule (§ 50.61, "Fracture Toughness Requirements for Pro tection 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 con sulted 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 pro vides the regulatory basis for this guide. The information collec- tien requirements in 10 CFR Part 50 have been cleared under OMB

Clearance No. 3150-0011.

B. DISCUSSION

Some NRC requirements that necessitate calculation of radia tion embrittlement are:

1. Paragraph V.A of Appendix 0 requires the effews of neutron radiation to be predicted from the results ofperdnent radiation effcts studies. This guide provides such results in the form of calculative procedures that are acceptable to the NRC.

2. Paragraph V.B of Appendix 0 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 betline given in Paragraph

11.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 M.A and IV.A.1 specify the additional test requirements for beitlie materials tha supplement the requirements for reactor vessel materials generally.

4. Paragraph n.B of Appendix H incorporates ASTM E 185 by reference. Paragraph 5.1 of ASTM E 18542, "Standard Prac tice 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, ie., 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 ASIM E 185-32, the require ments for the number of capsules and the withdrawal schedule are based on the calculated amount of radikton embrittlement at end of life.

USNRC REGULATORY GUIDES

The guides are Issued In the following ten broad divisions:

Regulatory Guides are issued to describe and make available to the public methods acceptable to the NRC staff of impieenting

1. Power Reactors

6. Products specific parts of the Commission's regulations, to delineate tech.

2. Research and Test Reactors

7. Transportation niques used by the staff In evaluating specific problems or postu-

3. Fuels and Materials Facilities 8. Occupational Health lated accidents or to provide guidance to applicant

s. Regulatory

4. Environmental and Siting

9. Antitrust and Financial Review Guides are not substitutes for regulations, and compliance with

5. Materlals and Plant Protection 10. General 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 be purchased from the Government licenseInting Office at the current GPO price. information on current GPO prices may be obtained by contacting the Superintendent-of This guide was issued after consideration of comments received from Documents, U.S. Government Printing Office, Post Office Box the public. Comments and suggestions for improvements in these

37082, Washington, DC 20013-7082, telephone (202)275-2060 or guides are encouraged at all times, and guides will be revised, as

(202)275-2171.

arpropriate, to accommodate comments and to reflect new informa tion or experience.

issued guides may also be purchased from the National Technical Written comments may be submitted to the Rules and Procedures Information Service on a standing order basis. Details on this Branch, ORR

ADM,

U.S.

Nuclear Regulatory Commission, service may be obtained by writing NTIS, 5285 Port Royal Road.

Washington, DC 20555.

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 U 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, &RTNDT, is defined in Appendix 0 as the tempera ture 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 B 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 0. 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 correla tion functions should be used for weld and base metal, (2) the func tion should be the product of a chemistry factor and a fluence factor,

(3) the parameters in the chemistry factor should be the elements copper and nickel, and (4) the fluence factor should provide a trend curve slope of about 0.25 to 0.30 on log-log paper at 10"9 n/cm2 (E > 1 MeV), steeper at low fluences 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 low copper 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 neutron per square centimet having energies greater than I million electron volts (E > I MeM). 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 func tion such as displacements per atom (dpa) (Ref. 5) in the analysis of the surveillance data base (Ref. 6).

Howeve, te neutron energ spectrum does change significantly with location in the vessel wall; hence for calculating the attenua tion of radiation embrittlement through the vessel wall, it is necessary to use a damage function to determine ARTNDT versus radial distance into the wall. Te most widely accepted damage flnc tion 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 embrittlemetnt may be affected by elements other than copper and nickel. The original version and Revision I of this guide had a phosphorus term in the chemistry factor, but the studies on which this revision was based ftond other elements such as phosphorus to be of secondary importance, i.e.,

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

Scatter in'the data base used for this guide is relatively signifi cant, as evidenced by the fact that the standard deviations for Guthrie's derived formulas (Ref. 2) are 287OF for welds and 17OF

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 credibil ity 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 Regulatory Position 2.1 normally should be less than 287F for welds and 17oF for base metal. Evenifthe 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 deter mined, following the definition given in ASTM E 185-82 (Ref, I).

4. The irradiation temperature of the Charpy specimens in the capsule should match vessel wall temperature at the cladding/base metal interface within +/-25OF.

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 fac tor 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 error 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 ARTNDT; this seems reasonable for fitting surveillance data, because generally the higher data point will be the more recent and therefore will rePre sent more moderm proced- e.

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 beldine 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 beitline is given by the following expression:

ART W Initial RTNDT + ARTNDT + Margin

(1)

Initial RTNDT Is the reference temperature for the unirradiated material as defined in Paragraph NB-2331 of Section M of the ASME Boile and Pressure Vessel Code Of. 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 devia tion 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(O.28 - 0.10 log f)

(2)

CF (OF) is the chemistry factor, a function of copper and nickel content. CF is given in Table I 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 ifjustifi on 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 dMe vessel wall, f(1019 n/cm, E > 1 MeV), is determined as follows:

f - fsurf (e -0.24x)

(3)

where fsurf (10"9 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 die vessel wall measured from the vessel inner (weted) 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, fO.28 - 0.10 log f, is determined by calcula tion or from Figure 1.

"Margin" is the quantity, OF, that is to be added to obtain con servative, upper-bound values of adjusted reference temperature for the calculations required by Appendix G to 10 CFR Part 50.

Margint.*

2 A

Sandr

(4)

• Th das f" eawtimang Iniia FLT,~ iB genealy deemIlned, fibr the welds withwhihdI

d xz a

g oncered by d~ii of eli flux (Unde 90 or other);

kr uaemtl h

SMStnadSeiiain Here, oI is the standard deviation for the initial RTNDT. H a meured value of initial RTNDT for the material in question is available, ol is to be estimated from the precision of the test method.

If not, and generic mean values for that class of material are used, oI is the standard deviation obtained from the set of data used to establish the mean.

The standard deviation for hRTNDT, vA, is 28 *F for welds and

17OF 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 lnmtations 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 stngh of 50,000

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

2. The procedures are valid for a nominal irradiation tmpertr of 550OF. Irradiation below 525 OF should be considered to pro duce greater embrittlement, and irradiation above 590"F may be considered to produce lass embrittlement. The correction factor used should be justified by reference to actual data.

3. Application of these procedures to fluence levels or to cop per or nick content beyond the ranges given in Figure I 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 Rrgdatory Positions 2.1 and 2.2, respectively.

2.1 Adjusted Reference Temperature The adjusted referce temperature should be obtained as follows. First, if there is dear 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 surveilance 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 fa&Ltrs. The resulting value of CF when entered in Equation 2 will give the relationship of ARTýT to

1.99-3

TABLE I

CHEMISTRY FACTOR FOR WELDS, OF

fluence that fits the plant surveillance data in such a way as to minimize the sum of the squares of the errors.

To calculate the margin in this case, use Equation 4; the values given there for OA may be cut in half.

If this procedure gives a higher value of adjusted reference temperature than that given by using the procedures of Regulatory Position 1.1, the surveillance data should be used. If this procedur gives a lower value, either may be used.

For plants having surveillance data that are credible in all respects except that the material does not represen the critical material in the vessel, the calculative procedures in this guide should be used to obtain mean values of sA, ARTNDT. In calculatng the margin, the value of OA may be reduced from the values given in the last paragraph of Regulatory Position 1.1 by an amount to be decided on a case-by-case basis, depending on where the measured values fall relative to the mean calculated for the surveillance materials.

2.2 Charpy Upper-Shelf Energy The decrease in upper-shelf energy may be obtained by plot ting the reduced plant surveillance data on Figure 2 of this guide and fitting the data with a line drawn parallel to the existing lines as the upper bound of all the data. This line should be used in preference to the existing graph.

1.99-4

0 ~t.*e Nickel, Wt-%

r,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-

7ý

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

TABLE2

CHEMSTRY FACTOR FOR

.ASE.METAL,

-F

Coper, Nickel, Wt-%

f

0

0.20

-

0.40

0.(Dr

0-

o.o ...

i.06 .D1.2 u

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.10

0.11

0.12

0.13

0.14

0.15

0.16

0.17

0.18

0.19

20

20

20

20

22

25

28

31

34

37

41

45

49

53

57

61

65

69

73

78

0.20

82

0.21

86

0.22

91

0.23

95

0.24

100

0.25

104

0.26

109

0.27

114

0.28

119

0.29

124

0.30

129

0.31

134

0.32

139

0.33

144

0.34

149

0.35

153

0.36

158

0.37

162

0.38"

166

0.39

171

0.40

175

20

20

20

20

26

31

37

43

48

53

58

62

67

71

75

80

84

88

92

97

102

107

112

117

121

126

130

134

138

142

146

151

155

160

164

168

173

.177

182

185

189

20

20

20

20

26

31

37

44

51

58

65

72

79

85

91

99

104

110

115

120

125

129

134

138

143

148

151

155

160

164

167

172

175

180

184

187

191

196

200

203

207

207

20

20

20

20

26

31

37

44

51

58

65

74

83

91

100

110

118

127

134

142

149

155

161

167

172

176

180

184

187

191

194

198

202

205

209

212

216

220

223

227

231

231

257

20

20

20

20

26

31

37

44

51

58

67

77

86

96

105

115

123

132

141

150

159

167

176

184

191

199

205

211

216

221

225

228

231

234

238

241

245

248

250

254

257.

20

20

20

20

26

31

37

44

51

58.

67

77

86

96

106

117

125

135

144

154

164

172

181

190

199

208

216

225

233

241

249

257

255

26

26D

274

264

282

268

290

272

298

275

303

278

308

281

313

285

317

288

320

20

20

20

20

26

31

37

44

51

58

67

77

86

96

106

117

125

135

144

154

165

174

184

194

204

214

"221

230

239

248

3. REQUIREMENT FOR NEW PLANTS

For beItline 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.* Tle copper con tent should be such that the calculated adjusted reference temperamure at the 1/4T position in the vessel wall at end of life is less than

200OF. 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 mxe jfomtiua, we &e eAppead to ASTM SundaWd Specifcai A 533 (Rde. ).

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 pro poses 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 predic tions of radiation embrittlement needed to implement Appendices G and H to 10 CFR Par 50.

1.99-5

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 Appen dix.O to 10 CFR Pat 50, unless they can justify the use of dif ferent methods. The use of the Revision 2 methodology may result in a modification of the pressweýrCpratMr limits contained in Technical Specifications in order to continue to satisfy the requirements of Section V of Appendix 0, 10 CFR Part 50.

3. The recommendations of Regulatory Position 3 are essen tially unchanged from those used to evaluate construction permit applications docketed on or after June 1, 1977.

1.99-6

REFERENCES

1. American Society for Testing and Materials, "Standard Prac tice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," ASTM E 185482, July 1982.

2. G. L. Outrie, "'Tupy Tend Curves Based on 177 PWR Data Points," in "LWR Pressure Vessel Surveillance Dosimetry Xm provement Program," NURECICR-3391, Vol.2, prepared by Hanford Engineering Development Laboratory, HEDL-TME

83-22, April 1984.**

3. 0. 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 Eblumlo, American Society for Testing and Materials, November 1974.*

S. American Society for Testing and Materials, "Standard Prac tice 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 DosihnetY

Improvement Program: LWR Power Reactor Surveillance Physics-Dosimetry Data Base Compendium," NUREo/

CR-3319, prepared by Hanford Engineering Development LaborWty, HEDL-TME 85-3, August 1985.**

7. American Society of Mechanical Engineers, Section m,

"Nuclear Power Plant Components," 9f ASAE Boier and Pressure Vessel Code, New York (updated frequently).tt

8. American Society for Testing and Materials, "Standard Specifcatio for Pressure Vessel Plates, Alloy Steel, Quenched and Tempered, Manganese-Molybdenum and Manganese Molybdenum-Nickel," ASTM A 533/A 533M-82, Septemlier

1982.*

• Copies may be obtained foum the American Society for Testing and Mbater'ls, 1916 Race Steet, Pafladdphi, PA 19103.

• Copies may be obtaindum e Superinted of Docm= ,. U.S. Governme Printing Office, Post Office Box 37092, Wasington, DC 200D13-7062.

tclsmay be obtained how the ecrcPower Research Insftitut

3412 Killview Avenue, Palo Alto, CA 9430.

t1o1smay be obtained fromn Se American Society of Mechanical Engineers, 345 E. 471h Sawee, New York, KY 10017.

1.99-7

2

3

4

6

6

7 8 91*

10,,

Fluence, n/cm2 (E > I MOV)

2

3

.4

6

6 7-8.91 w10

FIGURE I Flummo Factor for Use ia Equation 2, the Expression for ARTNDT

0

am

" U"

9

0

-UJ

S "

0I1 S..lost

2'

3

4

6

6 7 8u 1

1014

L u

. . . ...

. ......

.

. ..

.........

1.G

it

........

.....

...a l.

....

.8

....

........

..

11HP

.. . ...

7

.76....

.. . . .".

.. .

I

1 1 ,

i l5 H IM.1

-flfl

.6.

. . .

.

. .

.. . . .. .

1 1 1 1 1 -z

. . ..

-.4

. .....

..

7 -

T

iI

oio

60

50

40

30

20

10

0.10 -

0.05 jHHHHHHill P.PfPIIPPI

I

pl`j-Vý

'A

i huh

-10F

i iii i i

1

1 1 1 1 1 11

1 Illlll

1 m1;1 1 K!1

1 !t!

2 X 1017

4

6

8

1018

2

4

6

8

1019 FLUENCE, n/cm2 (E > 1MeV)

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

-

!11I I I I Ih I

mII I

I

I

IT

Ilf ~~l7 I

% COPPER

BASE

METAL

WELDS

0.35

0.30

0.30

0.25

0,25

0.20

0.20

0,15

0.15

-

0.10--

C

u'p

0)

C

2

4

6

-1 ....

- ....

11.1 ........

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 Comnmission's Public Document Room at 1717 H Stree NW., Washington, DC, under Regulatory Guide 1.99, Revision 2.

1.99-10