ML20154J925

Draft Rev 2 to Reg Guide 1.99, Radiation Damage to Reactor Vessel Matls

ML20154J925
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Issue date:	08/14/1985
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bre /p.w re l W rking P:p3r G August 14, 1985 Office of Nuclear Regulatory Research DRAFT REGULATORY GUIDE 1.99, REVISION 2 RADIATION DAMAGE TO 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, " Licensing of Production and Utilization Facili-ties," requires, in part, that the reactor coolant pressure boundary be designed with sufficient margin to ensure that, when stressed under operating, mainte-nance, testing, and postulated accident. conditions:

(1) the boundary behaves in a nonbrittle manner, and (2) the probability of rapidly propagating fracture is minimized and "...the design shall reflect...the uncertainties in deter-mining...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 damage to the low-alloy steels currently used for light-water-cooled reactor vessels.

The calculative procedures given in paragraph C.I.a of this draf t Guide are not the same as those given in the Pressurized Thermal Shock rule

• for calculating RTPTS, the reference temperature that is to be compared to the screening criterion given in the rule.

Issuance of Regulatory Guide 1.99, Revision 2 for public comment in no way effects the recently promulgated PTS rule.

Licensees and the technical community are requested to comment on the technical merits of this proposal, including its effect on their plants for non

- PTS purposes, chiefly as the basis for calculation of pressure-temperature limits as required by Appendix G, 10 CFR Part 50.

Licensees may also consider

• " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register, July 23, 1935, pp. 29937-29945.

1 RG 1.99 REV 2 08/21/85

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I 96031'10120 050906 PDR REV9P NROCROR NEETIN9001 PDR

P-l and comment on the proposed change's effect on the calculated PTS risk at their plant, assuming the Rev. 2 correlation, if justified, would at some future time I

l l

correlation in the PTS rule.

Following resolution of com-replace the RTPTS ments,and once general agreement is reached regarding the best way to calculate NOT, then it will be appropriate to re-evaluate the overall conservatism of RT the PTS rule.

The Advisory Committee on Reactor Safeguards will be consulted concerning this guide.

B.

DISCUSSION The principal examples of NRC requirements that necessitate calculation of radiation damage are:

1.

Paragraph V.A. of Appendix G requires:

"The effects of neutron radiation...are to be predicted from the results of pertinent radiation effect l

studies...."

This guide provides such results in the form of calculative procedures that are acceptable to the NhC.

I 2.

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

l upper limit for pressure as a function of temperature during heatup and cooldown for a given service pr.rsua 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 damage to be considered in the selection of the most limiting material...." Paragraphs III.A.

and IV.A.1. specify the additional test requirements for beltline materials that supplement the requirements for reactor vessel materials generally.

l 4.

Paragraph II.B. of Appendix H incorporates ASTM E185 by reference, i

Paragraph 5.1 of ASTM E185-82 requires that the materials to be placed in sur-veillance 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 damage must be considered.

In Paragraph 7.6 of ASTM E185-82 the requirements for number of capsules and withdrawal schedule are based on the calculated i

amount of radiation damage at end of Iffe.

The two measures of radiation damage used in this guide are obtained from the results of the Charpy V-notch impact test. Appendix G to 10 CFR Part 50 08/21/85 2

RG 1.99 REV 2 L

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 G 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 damage is the decrease in the Charpy upper-shelf energy level, which is defined in ASTM E185-82.

Revision 2 of this guide 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 Revision 2.

surface, given in Position C.1.a.2.

The basis for equation (2) for ARTNDT of this Guide, is contained in publications by G. L. Guthrie1 and G. R. Odette.2 Both authors used as their data base surveillance data from commercial power reactors, but their analysis techniques were different.

Both authors recommended the following:

(1) separate correlation functions for weld and base metal, (2) the function 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 1018 n/cm8 (E>l MeV), steeper at low fluences and flatter at high fluences.

Position C.1.a. is a blend of the correlation functions presented by the two 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 Position C.I.b. is that given by Spencer H. Bush.8 IG. L. Guthrie, "Charpy Trend Curves Based in 177 PWR Data Points " from LWR Pressure Vessel Surveillance Dosimetry Improvement Program, Quarterly Progress Report April 1983 - June 1983, Hanford Engineering Development Laboratory, NUREG/CR-3391, Vol. 2, HEDL-TME 83-22.

8G. R. Odette and P. M. Lombrozo, " Physically Based Regression Correlations of Embrittlement Data From Reactor Pressure Vessel Surveillance Programs,"

EPRI NP-3319 Final Report, January 1984, Prepared for Electric Power Research Institute.

8 Spencer H. Bush, " Structural Materials for Nuclear Power Plants," 1974 ASTM Gillett Memorial Lecture, published in ASTM Journal of Testing and Evaluation, November 1974, and its addendum, " Radiation Damage in Pressure Vessel Steels for Commercial Light-Water Reactors."

08/14/85 3

RG 1.99 REV 2

B.

The measure of fluence used herein is the number of neutrons per square centimeter having energies greater than 1 million electron volts (E>l MeV).

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)* in the analysis of the surveillance data base.5 However, the neutron energy spectrum does change significantly with location in the vessel wall; hence for calculation of attenuation of radiation damage through the vessel wall, a damage function should be used to determine ART versus radial distance into the wall. The most widely accepted damage NOT function at this time is dpa and the attenuation formula (3) given in Position C.1.a.(2), is based on the attenuation of dpa through the vessel wall.

Sensitivity to neutron radiation damage may be affected by elements other than copper and nickel.

Revisions 0 and 1 of this guide had a phosphorus term in the chemistry factor, but the studies upon 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 are 28*F for welds and 17"F for base metal, despite extensive statis-tical analysis.

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 should 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 damage according to the provisions of this guide.

• ASTM E 693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements Per Atom (dpa)."

sW.N. McElroy, Editor, " LWR Power Reactor Surveillance Physics - Dosimetry Data Base Compendium," NUREG/CR 3319 HEDL TME 84-2 March 1984.

08/14/85 4

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l **

2.

Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be sna11 enough to permit the determination of the 30 ft-lb temperature and the upper shelf energy unambig-uously.

3.

When there are two or more surveillance data from one reactor, the scatter of ART values about a best fit line drawn as described in Posi-NDT tion C.2.a. normally should be less than 28*F for welds and 17*F for base metal.

Evea 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.

4.

The irradiation temperature of the Charpy specimens in the capsule should match vessel wall temperature at the cladding-base metal interface within 125*F.

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.

In using plant surveillance data to develop a plant-specific relationship of ART to fluence, it was deemed advisable (because of scatter) to determine NOT the slope, i.e., the fluence factor, from other than the plant data.

Instead, Equation 2, paragraph C.1.a.(2), is to be fitted to 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; which seems reasonable for fitting surveillance data, because generally that datum will be the more recent one and therefore will represent more modern 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 damage to the beltline of reactor vessels of light water reactors should be based on the following procedures, within the limitations in Paragraph C.1.c.:

08/14/85 5

RG 1.99 REV 2

The adjusted reference temperature (ART) for each material in a.

the beltline is given by the following expression:

1 (1)

ART = Initial RTNDT + ARTNDT + Margin (1) " Initial RT

" is the reference temperature for the NDT unirradiated material as defined in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code.

In cases where measured values of Initial RT for the material in question are not available, generic mean values NDT of material may be used if there are sufficient test results e

for that class to establish a mean and standard deviation for the class.

(2) " ART

" is the mean value of the adjustment in reference NDT temperature caused by irradiation and should be calculated as follows:

surface = [CF]f(0.28-0.10 log f)

(2)

ARTNDT The chemistry factor, "CF," "F, a function of copper and nickel content, is given in Table I for welds and Table II for base metal (plates and forgings).

Linear interpolation is permitted.

In Tables I and II, " Percent Copper" and " 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, sFor the welds with which this guide is concerned, for estimating Initial RT

, class is generally determined by whether the welding flux is Linde 80 or N her; for base mets), by the ASTM Standard Specification.

N 08/14/85 6

RG 1.99 REV 2

The fluence, "f," is the calculated value of the neutron fluence at the inner wetted surface of the vessel at the location of the postulated defect, n/cas (E>l MeV) divided by 10to, The fluence factor, f.28 - 0.10 log f, is determined by calculation or 0

from Figure 1.

To calculate ART at any depth, (e.g., at 1/4T or 3/4T), the following NOT attenuation formula should be used:

ARTNDT = [ARTNDT.surfacele (3) where "x" (in inches) is the depth into the vessel wall measured from the vessel inner (wetted) surface.

(3) " 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, 10 CFR Part 50.

Margin =24oj+og (4)

If a measured value of Initial RT f r the material in question is NOT is used, used, o may be taken as zerc.

If a generic value of Initial-RTNDT g

should be obtained from the same set of data (see paragraph C.1.a.(1)).

og The standard deviations for ART

""A, are 28'F for welds and 17'F for base NOT' metal, except o need not exceed 0.50 times the mean value of ART surface.

g NOT i

b.

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, Application of the foregoing procedures should be subject to c.

the following limitations:

(1) The procedures apply to those grades of SA-302, 336, 533, and 508 steels having minimum specified yield strer.gths of 50,000 psi and under and to their welds and heat effected zones.

(2) The procedures are valid for a nominal irradiation tempera-ture of 550*F.

Irradiation below 525'F should be considered to produce greater j

damage, and irradiation above 590'F may be considered to produce less damage.

The correction factor used should be justified by reference to actual data.

08/14/85 7

RG 1.99 REV 2

(3) Application of these procedures to fluence levels or to

~

copper or nickel content beyond the ranges.given in Figure 1 and Tables I and II 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 as defined in the Discussion, Section B, 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 the following Paragraphs a. and b.,

respectively.

The adjusted reference temperature should be obtained by first a.

fitting the surveillance data using Equation 2, paragraph C.1.a.(2), to obtain the relationship of ART surface to fluence. To do so, calculate the chem-NDT istry factor, "CF," for the best fit as follows. Multiply each measured ART by its corresponding fluence factor, sum the products and divide by the NDT sum of the squares of the fluence factors. The resulting value of CF when surface to H uence entered in Equation 2 will give the relationship of ARTNDT that fits the plant surveillance data in such a way as to minimize the sums of the squares of the errors.

To calculate the Margin in this case, use the procedure given in paragraph may be cut in half.

C.1.a.(3), except the values given there for og If this procedure gives a higher value of adjusted reference temperature than that given by using the procedures of paragraph C.I.a. the surveillance data should be used.

If this procedure gives a lower value, either may be used.

b.

The decrease in upper-shelf energy may be obtained as follows.

Plot the reduced plant surveillance data on Figure 2 of this Guide.

Fit 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.

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 08/14/85 8

RG 1.99 REV 2

Y',

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.

c, D.

IMPLEMENTATION sThe purpose of this section is to provide information to applicants and licensees regarding the NRC staff's plans for utilizing this regulatory guide.

Except in those cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the positions described in this guide will be used by the NRC staff as follows:

1.

The method described in regulatory positions C.1 and C.2 of this guide will be used in evaluating all predictions of radiation damage needed to implement Appendices G and H to 10 CFR Part 50 submitted on or after (60 dajs af ter publication); however, if an applicant wishes to use the recommendations of regulatory position C.1 and C.2 in developing submittals before (60 days after publication), the pertinent portions of the submittal will be evaluated on the basis of this guide.

2.

Following publication of this guide in final form, the owners of all operating reactors and all applicants for an operating license should review the basis for the pressure-temperature limits in their Technical Specifications for consistency with Position C.1.

Those for whom the allowable operating period has been reduced or has already expired, when judged by the criteria of Revision 2, should revise their operating procedures, as appropriate, to conform with the criteria of Revision 2 of this guide and submit the appropriate revi-sion to their Technical Specifications within three years of the date of publi-cation of Revision 2 of this guide in final form.

Those fcr whcm the allowable operating period has been extended, when judged by the criteria of Revision 2, should submit the appropriate revision to their. Technical Specifications no later than 90 days prior to the expiration of their current operating period.

j

• For example, see the Appendix to ASTM Standard Specification A 533.

08/15/85 9

RG 1.99 REV 2

i.

j 3.

The recommendations of regulatory position C.3 are unchanged from

~

those us.cri to evaluate construction permit applications docketed on or after June 1, 1977.

l L

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08/14/85 10 RG 1.99 REV 2

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TA8LE I 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 l

0.19 83 100 126 157 191 220 238 O.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 j

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 08/14/85 11 RG 1.99 REV 2

TABLE II CHEMISTRY FACTOR FOR 8ASE METAL, *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 20 20 20 20 20 20 20 0.03 20 20 20 20 20 20 20 0.04 22 26 26 26 26 26 26 0.05 25 31 31 31 31 31 31

~0.06 28 37 37 37 37 37 37 0.07 31 43 44 44 44 44 44 0.08 34 48 51 51 51 51 51 0.09 37 53 58 58 58 58 58 0.10 41 58 65 65 67 67 67 0.11 45 62 72 74 77 77 77 0.12 49 67 79 83 86 86 86 0.13 53 71 85 91 96 96 96 0.14 57 75 91 100 105 106 106 0.15 61 80 99 110 115 117 117 0.16 65 84 104 118 123 125 125 0.17 69 88 110 127 132 135 135 0.18 73 92 115 134 141 144 144 0.19 78 97 120 142 150 154 154 0.20 82 102 125 149 159 164 165 0.21 86 107 129 155 167 172 174 0.22 91 112 134 161 176 181 184 0.23 95 117 138 167 184 190 194 0.24 100-121 143 172 191 199 204 0.25 104 126 148 176 199 208 214 0.26 109 130 151 180 205 216 221 0.27 114 134 155 184 211 225 230 0.28 119 138 160 187 216 233 239 0.29 124 142 164 191 221 241 248 0.30 129 146 167 194 225 249.

257 0.31 134 151 172 198 228 255-266 0.32 139 155 175 202 231 260 274 0.33 144 160 180 205 234 264 282 0.34 149 164 184 209 238 268 290 0.35 153 168 187 212 241 272 298 0.36 158 173 191 216 245 275 303 1

0.37 162 177 196 220 248 278 308 0.38 166 182 200 223 250 281 313 0.39 171 385 203 227 254 285 317 O.40 175 189 207 231 257 288 320 08/14/85 12 RG 1.99 REV 2

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sueosed'A52eu3 peg ut oses23eo 12/07/84 ity

- RG 1.99 REV 2

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BACKGROUND INFORMATION FOR CRGR REVIEW OF REGULATORY GUIDE 1.99, REVISION 2 The following information is provided in the format required by the CRGR charter.

1.

The proposed generic action is issuance of Revision 2 of Regulatory Guide 1.99, " Radiation Damage to Reactor Vessel Materials," for public comment.

It is Enclosure 1 to this memorandum.

The proposed implementa-tion schedule is Section D of the Guide.

2.

A staff paper giving the technical basis for the procedures for calculating the extent of radiation damage given in the Guide is Enclosure 3.

Copies of the references in Enclosure 3 and in the Guide are available.

3.

A brief description of each of the steps anticipated that licensees must carry out in order to comply with the recommendations of the Guide is as follows.

3.1 Are there separate short-term and long-term requirements?

When Revision 2 becomes effective, it will be used in our review of all submittals of P-T limit updates by owners of operating reactors and appli-cants for OLs.

Revision 2 will also be used in analyses of transients that threaten the integrity of the reactor vessel beltline and in the evaluation of flaws found in the reactor vessel beltline.

Licensees may continue wit,h schedulesfor review of the P-T limits presently given in their Technical Specifications, for a maximum of three years. Within that period, those for whom the allowable operating period has been reduced or has expired when judged by the criteria of Revision 2, should revise their operating procedures and submit the appropriate revision to their Technical Specifications.

05/22/85

~1 CRGR BACK INFO RG 1.99 REV 2 l

I s

1 3.2 Is it the definitive, comprehensive position on the subject or is it the first of a series of requirements to be issued in the future?

R; vision 2 is an update of the Revision 1 procedures for calculating the adjustment of the reference temperature caused by radiation damage.

Another revision will be required to update the procedures for calculating the decrease in upper-shelf energy when the technical basis for that because available in a year or two.

That change will not affect plant P-T limits. The technology for prediction of radiation damage is in such a state that further revision of the calculative procedures probably will be necessary in a few years time.

+

3.3 How does this requirement affect other requirements? Does this requirement mean that other items or systems or prior analyses need to be reassessed?

Paragraph 5.3.2 of the Standard Review Plan should be changed to refer to Revision 2 of Regulatory Guide 1.99.

As an example of the editorial changes, the references to " copper and phosphorus" should be changed to " copper and nickel." The changes of most significance involve the use of mean values and the separate calculation of margin described in Revision 2.

The pro-posed amendments to the Standard Review Plan are given in Enclosure 4.

Issuance of Revision 2 and the associated changes in the Standard' Review Plan will eventually affect the Tech Spec P-T limits for most plants and near-term OLs.

Issuance of Revision 2 for public comment will not affect the procedure for calculating RT relative to:the screening criteria given in the proposed NDT final rule on pressurized thermal shock, SECY 85-60, February 20, 1985.

That procedure is associated with screening criteria (270*F for base metal and axial welds, 300*F for circumferential welds), which were justified by a probabilistic analysis that considered all identifiable uncertainties including those in the calculation of RT Only if a licensee expects NDT.

a screening criterion to be exceeded 3 years hence will Revision 2 (or the then-current revision) be applied in the reanalysis of susceptibility to 4

05/22/85 2

CRGR BACK INFO RG 1.99 REV 2

I 4

pressurized thermal shock for that particular reactor vessel. Note, how-ever, the action recommended in paragraph 4.b.1 of Enclosure 5.

3.4 Is it only computation? Or, does it require or may it entail engineering design of a new system or modification of any existing system?

It entails the computation of new P-T limits, processing a Tech Spec change, and making the resulting changes in operating procedures during heatup and cooldown.

Such changes occur every few years anyway.

It will also be used in the analysis of transients that affect the reactor vessel beltline and the evaluation of flaws found in inservice inspection of the beltline.

3.5 There are no plant hardware changes involved in issuance of Revision 2.

4.

Revision 2 will apply to all owners of operating reactors and to all applicants for an operating license.

5.

In the. regulatory analysis (Enclosure 5), Table 1 gives a breakdown of the benefits and impacts to operating reactors and applicants for OLs, and to PWRs and BWRs.

5.1. Risk reduction assessments and cost assessments given in Enclosure 5 are based on a value/ impact analysis made by Pacific Northwest Laboratories.

(Enclosure 6) and a review of the costs by the NRC's Cost Analysis Group (Enclosures 7 and 8).

5.2. The basis for requiring implementation as described in Section D of the Guide isgivenintheRegulatoryAnalysis(Enclosure 5), Sections 5and6.

Section D states in effect that Revision 2 will be used in evaluating all predictions of radiation damage submitted after Revision 2 becomes effective.

However, as given in S 3.1, the implementation schedule will require that every utility review its pressure-temperature limits by 3 years after Revision 2 becomes effective.

The reasons for doing so are discussed in the Regulatory Analysis, Enclosure 5.

05/22/85 3

CRGR BACK INFO RG 1.99 REV 2

5.3 The schedule for staff actions to get Revision 2 published in final form will involve:

ACRS review, final editing, a 2-month public comment period, resolution of comments and redrafting of the final version, resubmittal to CRGR, and publication in final form.

5.4 Implementation of Revision 2 will crowd the normal calendar for review of P-T limits somewhat, especially toward the end of the 3 year period described in S 3.1, but the NRC staff schedule will not be impacted severely.

6.

Each proposed requirement implements existing regulations, namely General Design Criterion 31 and Appendices G and H, 10 CFR Part 50.

7.

Section D of Revision 2 describes the implementation. To ensure a proper and timely response, copies of the Guide when issued in final form will be sent to each licensee and applicant for an operating license as part of the distribution of Revision 2.

The concurrence of OELD and DL was obtained when the completed package was sent for Office concurrence to transmit to CRGR.

8.

The OMB clearance package when required under the Paperwork Reduction Act, and the Regulatory Flexibility Statement are not required.

Regulatory Guide 1.99 does not impose reporting requirements.

It describes acceptable procedures for fulfilling certain requirements of Appendices G and H, 10 CFR Part 50 for which OM8 clearance has been obtained and for which the Commission has certified that it will not have a significant impact on a substantial number of small entities (Federal Register Vol. 48, Number 104, May 27, 1983, p. 24008).

05/22/85 4

CRGR BACK INFO RG 1.99 REV 2

phb BASIS FOR REVISION 2 0F U.S. NRC REGULATORY GUIDE 1.99 P. N. Randall, U.S. Nuclear Regulatory Commission Abstract Regulatory Guide 1.99, " Radiation Damage to Reactor Vessel Materials,"

is being updated to reflect recent studies of the physical basis for neutron radiation damage and efforts to correlate damage to chemical composition and fluence.

Revision 2 of the Guide contains several significant changes. 'Jelds and base metal are treated separately.

Nickel content is added as a variable and phosphorus removed. The exponent in the fluence factor is reduced, espe-cially at high fluences. And, guidance is given for calculating attenuation of damage through the vessel wall. This paper describes the basis for these changes in the Guide.

Key Words Neutron irradiation, nuclear reactor materials, low alloy steels, fluence, copper, nickel, fracture toughness.

t s

INTRODUCTION Revision 2 of Regulatory Guide 1.99 " Radiation Damage to Reactor Vessel Materials" (the Guide), is an outgrowth of many activities:

(1) experience with the application of Revision 1 in licensing work since 1977, (2) technical contacts at meetings of ASTM Comr.fttee E-10 and Metal Properties Council Subcommittee 6 Task Groups on radiation damage plus an American Nuclear Society seminar in 1983, (3) accumulation of surveillance data from commercial power reactors, (4) resolution of the pressurized thermal shock issue, which required best-estimate calculative procedures and careful attention to uncertainties, (5) extensive help in data analysis by G. L. Guthrie 1 and (6) interaction with 2

G. R. Odette,

The objective.of the Guide is to provide calculative procedures for the adjusted reference temperature (ART) that are acceptable to the NRC. The purpose of this paper is to describe the basis for those procedures. As given in the Guide, ART

= Initial RTNDT + A RTNDT + Margin where RT

= reference temperature, nil-ductility transition, deg. F.

NDT Initial RTNDT = the reference temperature for the unitradiated material.

ARTNDT = the adjustment of reference temperature, commonly called the Charpy shift, i.e., the temperature shift (measured at the 30 ft lb level) in the average Charpy curve for the irradiated material relative to that for the unirradiated material.

Margin = the quantity, deg. F, that is to'be added to obtain conservative, upperbound values of ART.

1From Hanford Engineering and Development Laboratories, a contractor to NRC.

8From University of California, Santa Barbara, CA, a contractor to the Electric Power Research Institute.

05/21/85 1.99-1 Rev. 2

This paper focuses on the calculation of ART from chemistry and fluence NOT information, which will be a typical procedure for all plants until they have credible surveillance data.

As used in the Guide, " ART surface" is the NDT product of a chemistry factor and a fluence factor. The latter is based on the fluence at the inside surface of the vessel beltline at the location of interest.

However, in the fracture mechanics calculations that are the basis for pressure-temperature limits for reactor heatup and cooldown, the fracture toughness value of interest is that at the tip of the postulated flaw.

(Typically, toughness data are given in terms of temperature relative to RTNDT.) Thus one is also required to calculate the attenuation of ARTNDT through the vessel wall. This paper will also discuss the formula given in the Guide for this purpose.

THE GUTHRIE DATA BASE Commercial power reactor surveillance data were used exclusively (no test reactor data) in the analyses made by Guthrie. There were 51 weld and 126 base metal (plate and forging) data points, taken from surveillance reports by P. N. Randall and rechecked by Guthrie.

Shift values were those read from the hand-drawn Charpy curves at the 30 ft Ib level by the authors of the surveil-lance reports.

In those few cases where the authors. reported only 50 ft lb shift values, Randall and Guthrie made their own estimate of the 30 ft Ib shift from the plotted data.

Fluence values for the surveillance capsules were those reported by Simons(2) when available, otherwise the surveillance report information was used directly.

Copper and nickel content were as reported except in the case of welds made by Babcock and Wilcox for which newer information was available(3).

t The distribution of the data with respect to copper and nickel content is illustrated in Figure 1 for welds and Figure 2 for base metal, taken from one of Guthrie's computer printouts.

To produce the data shown, each measured value of ART was first normalized to a fluence of 102' n/cm2 NDT using Guthrie's 1

fluence function (described below) then placed in the " box" corresponding to its copper and nickel content.

If there were two or more entries in a box, they were averaged. Thus, in the lower left corner of Figure 1 the entry "77(2)" indicates that there were two pieces of data for welds having l

05/21/85 1.99-2 Rev. 2

0.10 1 0.025 percent Cu and 0.15 1 0.025 per cent nickel, and the average of their normalized shift values was 77'F (43'C).

To get a feel for the task of developing correlations, it is instructive to attempt to draw isoshift " contour" lines on Figs. 1 and 2 for several values of shift.

Two characteristics of the data base will become obvious from this exercise:

the degree of scatter is significant, and there are clumps of data and sizeable blank areas where I

there are no data.

In Revision 2 of the Guide, it was necessary to give values of the chemistry factor for copper content ranging from 0 to 0.40 percent and for nickel ranging fror. O to 1.2 percent to provide guidance over the full range of expected compositions.

Admittedly, these values somewhat exceed the ranges of the data base; hence, the Guide violates a restriction placed on the correlation func-tions by Guthrie. The most likely occurrence is for welds having copper content less than 0.12 percent and nickel in the 0.6 perceat range.

Clearly, applica-tion of the Guide at the fringes of the data base should be made with caution and supported by additional data.

Unfortunately, these will likely be test reactor data, and their applicability to a correlation useful for operating reactors is still in doubt.

There is also a dearth of surveillance data for materials having low copper and higher than normal phosphorus contents.

The upper limit on phos-phorus in the data base is 0.020 percent for welds and 0.017 for base metal.

Application of the Guide to cases where the phosphorus content is significantly higher should not be made without supporting data, which again brings up the question of the applicability of test reactor data.

i To observe the distribution of fluence values in the data base, refer to Figure 11, which will be described in the' discussion of residuals (measured minus calculated" values) as a function of fluence.

The range of fluence values was from 7.3 x 1017 to 7.8 x 1018 n/cm2 (E>1MeV), and the distribution within that range was reasonably uniform.

05/21/85

-1.99-3 Rev. 2

THE ODETTE DATA BASE The data base used by Odette (4) was the EPRI data base (5) which contained 65 weld data and 151 base metal data.

It overlapped the Guthrie data base almost completely and the ranges of copper, nickel and fluence were about the sama.

There are more data from boiling water reactors in the Odette data base.

The principle difference.is in the derivation of shift values.

In the EPRI data base, the plots of Charpy energy as a function of temperature were refitted using a hyperbolic tangent function and the 30 ft ib shift values were then recalculated.

A spot check showed the results are about the same as those in the Guthrie data base.

DERIVATION OF THE CHEMISTRY FACTOR FROM THE GUTHRIE AND ODETTE CORRELATION FUNCTIONS Guthrie and Odette reached similar conclusions in several areas:

(1) separate correlations are needed for welds and base metal, (2) the expression should be the product of a chemistry factor and a fluence factor, (3) the elements in the chemistry factor should be copper and nickel, and (4) the fluence factor should provide a trend curve slope when plotted on log-log paper of about 0.25 to 0.30 at 1018 2

n/cm, and it should be steeper at lower fluences and flatter at higher fluences.

For welds, their correlation functions for ART are as follows:

NDT NOT = [624 Cu - 331/CuNi + 251 Ni][f.282-0.0409 in f) 0 Guthrie: ART 1 Standard Deviation = 28'F (16'C) i NDT = 360 Cu [1+1.38(erf{0.3 1-Cu}+1)][1-exp(0 11)] *

~

Odette:

ART

[f *

]

1 Standard Deviation = 27*F (15'C)

For base metal their correlation functions for ARTNDT *I* ** I*II""S*

NDT=[-38+556Cu+480Cutanh0.353ff][f.266-0.0449inf) 0 Guthrie: ART 1 Standard Deviation = 17'F (10'C) 05/21/85 1.99-4 Rev. 2 3

NDT = 389 Cu[1+0.33(erf(0.7 1+1)][f '

]

Odette: ART Cu 1 Standard Deviation = 23'F (13'C)

The units of ART are degrees Fahrenheit.

(The equations by Odette have NDT been converted from degrees Centigrade.) Copper and nickel are given in weight percent, and fluence, "f", is in units of 1018 n/cm2 (E>l MeV).

Values for.the error function, "erf", are given in Table A-3 of Reference 4.

To compare the correlation functions given by Guthrie and Odette, the first step is to compare the chemistry factors, the expressions obtained by setting the fluence equal to 1 x 1018 n/cm.

Figures 3 and 4 compare the 2

Guthrie and Odette chemistry factors for welds for 2 nickel contents.

There is remarkable agreement at the higher copper levels. However, the function chosen by Odette passes through zero at zero copper whereas that by Guthrie intercepts the zero-copper ordinate at increasingly higher values as nickel increases. This difference is understandable in view of the lack of weld data below 0.10 percent copper. Because it has been generally accepted that the effect of nickel is a synergistic copper-nickel effect, if follows that ART NDT should be low when the copper content is low, regardless of the nickel content.

Therefore, the Odette curves were used throughout for welds with the exception of the cutoff at the lower end, discussed below.

For base metal, Figures 5 and 6 compare the Guthrie and Odette chemistry factors with regard to the effects of copper.

In this case, the curves cross.

Those by Guthrie are higher at high copper levels, but those by Odette are higher at low copper levels.

For the Guide, the higher curve was used, with two exceptions.

First, when the Guthrie curves for base metal exceeded those for welds, which they did at high copper jevels (see Figure 6), the latter were used, the justification being that there were no base metal data for copper above 0.25 percent and there is no basis to believe that base metal should be more sensitive to radiation than welds.

Second, at very low copper levels, the curves for both welds and base metal were levelled off at CF = 20*F.

Again, the weld data base is lacking in the range 0-0.10 percent copper, but some guidance is needed in this range because newer plants will have low copper.

05/21/85 1.99-5 Rev. 2

Therefore, test reactor data were used as described below to assist in setting the CF function for very low copper values.

Test reactor data (6,7) for low copper, high nickel (0.7 percent, nominal) materials were normalized to a fluence of 1028 n/cm2 by dividing the measured I

shift by the quantity:

(f)o s, the fluence function favored by the authors of the reports. Most test fluences were in the range 2 to 8 x 1028 2

n/cm, hence the normalized values of shift were felt to be as low as one could justify.

The results are plotted in Figure 7 for welds and Figure 8 for base metal, i

superimposed on the chemistry factor data as given in the Guide.

For base metal, there is adequate surveillance data to support the Odette and Guthrie j

work, but for welds the data are sparse.

Nevertheless, in the light of the test reactor data plotted in Figures 7 and 8, it' seeaed prudent to establish q

a minimum at 20*F (11*C) for the chemistry factor.

In addition, for base metal the curves in Figure 8 were faired in to the minimum at 0.05 percent copper, which meant raising the curves for 0 and 0.2 percent nickel about 15'F (8'C).

I A comparison of the chemistry factc,rs for welds and base metal is given in Figure 9.

The fact that the differences disappear at copper levels above about 0.25 percent is an artifact of the procedure used to draw the curves, described above.

1 DERIVATION OF THE FLUENCE FACTOR 1

Guthrie found only small differences in the constants of the fluence I

factars for welds and base metal.

(See the correlation functions given earlier.)

In Figure 10, the two factors are plotted,over the range 2 x 1027 to 1020 They differ by less than 4 percent. Consequently, in the interests of j

simplicity, the fluence factor used for both was:

f exp (0.28 - 0.0434 in f) i or f exp (0.28 - 0.10 log f) as it is given in the Guide.

For clarity this curve is not shown in Figure 10.

It would fall between the Guthrie curves for weld and base metal.

The fluence factor for welds derived by Odette, also i

shown in Figure 10, gives good agreement with that obtained by Guthrie except at fluences below 1.5 x 101s, where the Odette fluence factor drops off sharply.

j For base metal, Odette used a uniform slope of 0.28, which (happily) agrees l

05/21/85 1.99-6 Rev. 2

with that found by Guthrie.at 101' n/cm.

Therefore, it was an easy decision 2

to use Guthrie's fluence factor with the constants given above.

JUSTIFICATION FOR THE CALCULATIVE PROCEDURES GIVEN IN THE GUIDE To show that the calculative procedures given in the Guide are faithful to the data base, they were used to calculate a shift value based on the copper, nickel and fluence values for each line of data in the Guthrie data base. The residual (observed minus calculated value) is plotted versus fluence, copper and nickel content in Figures 11, 12, and 13, respectively.

Scatter about the zero residual axis is fairly well balanced between overprediction and underprediction.

One exception is seen in Figure 12 where for base metal the perturbation seen at low copper values is a reflection of the adjustment of chemistry factors made to reflect test reactor data and provide a conser-vative minimum.

(

Another purpose in showing these plots of residuals is to demonstrate that the blending of Guthrie's and Odette's results to get the calculative procedures for the Guide has not invalidated the use of twice the standard deviation from Guthrie's regression analysis to provide suitable margin. The "two-sigma" limits, 156*F (131*C) for welds and 134*F (119'C) for base metal, plotted on the Figures, do indeed show that only one weld and two base metal data points will be underpredicted if the margin on ART is made twice the NDT standard deviation.

In considering the requirement for the amount of margin to be t.dded, there was a question about the choice of margin for very low values of ca7culated shift. A more general question was:

should the margin be'some function of the shift? To answer this question, the residuals were plotted against the l

calculated value of shift as shown in Figure 14. There is no clear evidence of a relationship of the residuals to the calculated value.

Consequently, it was decided to add twice the standard deviation across the board except at low values where it was arbitrarily decided to add 100 percent of the calculated value, as shown in Figure 14.

05/22/85 1.99-7 Rev. 2

As given in the Guide, the margin to be added in calculating conservative values of RT f r use in Appendix G (10 CFR Part 50) evaluations includes NDT margin on initial RT as well as margin on ART Following the precedent NDT NDT.

set in the analyses for the pressurized thermal shock problem, the two are combined in the expression:

Margin = 2 4 o,3 +c2 3

where o, is the standard deviation on. initial RT when a generic mean value NDT is used, and o is the standard deviation on ART g

NDT*

ATTENUATION OF RADIATION DAMAGE WITHIN THE VESSEL WALL The changes in neutron energy spectra with depth of penetration in the wall are significant; and to take this into consideration, it was decided to use a "dpa equivalent" attenuation formula. This is one change in the method used previously. Another change is brought about by the change in the fluence factor from a simple power law with an exponent of 0.50 to one with an exponent of (0.28-0.10 log f).

In the face of considerable uncertainty about the various elements in this calculation, we elected to use the following simplified procedure.

The starting point was the attenuation formula used for a number of years:

f = (fsurface)

• where f is the fluence in units of n/cm2 (E> MeV) and "x" is depth in the wall, in inches, measured from the inside surface. This formula came from a staff review of surveillance reports made several years ago.

To convert to a "dpa equivalent" formula we used some calculations reported at the 4th ASTM -

Euratom Symposium (8), which showed that dpa attenuation through an 8.0 in.

vessel wall is less than the attenuation of fluence, n/cm2 (E) 1 MeV) by a factor of 2.06, the average of six calculations made for different reactor 05/21/85 1.99-8 Rev. 2

1 vessels. To achieve this reduction in attenuation, the equation for fluence attenuation becomes, i

f = (fsurface) '

For simplicity, the relationship of ART to fluence is taren to be a simple NDT power function with an exponent of 0.28 with the result:

ARTNDT = [ARTNOT surface3*

This is a best-estimate expression. The uncertainty is assumed to be accounted for in the margin term described earlier.

CONCLUSION The work of two independent investigators, working from separate data bases, yet coming to very similar conclusions, has provided a sound basis for the calculative procedures for adjustment of reference temperature given in Revision 2 of Regulatory Guide 1.99.

The Guide will receive peer review when published for public comment and will be reviewed again within the NRC in response to those comments.

\\

t 05/21/85 1.99-9 Rev. 2

REFERENCES (1)

G. L. Guthrie, "Charpy Trend Curves Based on 177 PWR Data Points," from LWR Pressure Vessel Surveillance Dosimetry Improvement Program, Quarterly Progress Report April 1983 - June 1983, Hanford Engineering Development Laboratory, NUREG/CR-3391, Vol. 2, HEDL-TME 83-22.

(2)

W. N. McElroy, Editor, LWR Pressure Vessel Surveillance Dosimetry Improvement Program.

1983 Annual Report, NUREG/CR-3391, Vol. 3, HEDL-TME 83-23.

(3)

K. E. Moore and A. S. Heller, "B&W 177FA Reactor Vessel Beltline Weld Chemistry Study," Babcock and Wilcox Co., BAW-1799, July 1983.

(4)

G. R. Odette, P. M. Lombrozo, J. F. Perrin and R. A. Wu11aert, " Physically Based Regression Correlations of Embrittlement Data From Reactor Pressure Vessel Surveillance Programs," Electric Power Research Institute, EPRI NP-3319, Jan. 1984.

(5)

W. Oldfield, P. McConnell, W. Server, and F. Oldfield, " Irradiated Nuclear Pressure Vessel Steel Data Base, "EPRI NP2428, June 1982.

(6)

C. Guionnet, B. Houssin, D. Brasseur, A. Lefort, D. Gros and R. Perdreau,

" Radiation Embrittlement of PWR Reactor Vessel Weld Metals:

Nickel and Copper Synergisic Effects," ASTM STP 782, Brager and Perrin, Editors, 1982.

(7)

P. Petrequin, A Review on Activities Jn France on Irradiation Embrittlement Pressure Vessel Steels," Presented at IAEA Specialists Meeting on Irradia-tion Embrittlement, Vienna, October 1981.

I (8)

G. L. Guthrie, W. N. McElroy and S. L. Anderson, "A Preliminary Study of the Use of Fuel Management Techniques for Slowing Pressure Vessel Embrittle-ment," Paper presented at 4th ASTM - Euratom Symposium, March, 1982.

05/21/85 1.99-10 Rev. 2

t i

1.20 1.15 1.10 231(1) 1.06 1.00 210(1)

O.95 243(1) 0.90 0.05 a 0.00 242(3)

E 0.75 198(2) 134(1) 238(1) 2 0.70 256(1) 284(1) 2 0.05 217(1) 214(2) z 0.40 104(5) 252(3) 276(1)

' j 0.b5 153(4) 188(2) 190(2) 173(2)

E 0.50 100(1)

E 0.45 0.40 0.5 O.30 112(1) 0.25 0.20 126(1) 112(2) i 0.15 77(2) 29(1) 0.10 11115) 0.05 08(1) 116(3) 0.05 0.10 0.15 0.20 0.25 0.30 0.35 PERCENT COPPER l

FIG.1 DISTRIBUTION OF GUTHRIE'S WELD DATA BASE IN TERMS OF COPPER AND NICKEL CONTENT l

l t

i

I l

1.20 1.15 1.10 1.06 1.00 0.95 0.90 0.05 g 0.s0 at 0.75 8(4) 183(1) l 0.70 19(14) 39(6) 136(2) 0.5 70(1) 108(14)

$ 0.00 116(6)

U 0.55 0(1)

E S) 08(1) 214(2) 146(2)

E 0.50 99(6) 91(7) 128(1) 183(2)

E 0.45 0.40 0.36 0.30 0.25 0.20 5?(16) 101(1) 102(25) 0.15 106(1)

~

0.10 0.05 1

0.05 0.10 0.15 0.20 0.25 0.30 0.35 i

PERCENT COPPER i

i FIG. 2 DISTRIBUTION OF GUTHRIE'S BASE METAL DATA BASE IN TERMS OF COPPER AND NICKEL CONTENT I

n

. _.- _. - - -. - - _ _,. ~ _ _ - - - - - _ _. - - -. - -, -,. -.. - - - - _ _, _ _ -

i 300 y

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0.1 0.2 0.3 0.4 PERCENT COPPER FIG.3 COMPARISON OF GUTHRIE AND ODETTE STUDIES ON THE EFFECT OF COPPER ON THE CHEMISTRY FACTOR -- WELDS WITH 0.2 PERCENT NICKEL I

..-----.,__---____n_---,

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FIG.4 COMPARISON OF GUTHRIE AND ODETTE STUDIES ON THE EFFECT OF COPPER ON THE CHEMISTRY l

FACTOR -- WELDS WITH 0.8 PERCidNT NICKEL

300 y

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• 200 I

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O 0.1 0.2 0.3 0.4 PERCENT COPPER FIG. 6 COMPARISON OF GUTHRIE AND ODETTE STUDIES ON THE EFFECT OF COPPER ON THE CHEMISTRY i

FACTOR -- BASE METAL WITH 0.2 PERCENT NICKEL i

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FIG.6 COMPARISON OF GUTHRIE AND ODETTE STUDIES ON i

THE EFFECT OF COPPER ON THE CHEMISTRY FACTOR -- BASE METAL WITH 0.8 PERCENT NICKEL

320 1.2 1.0 280

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FIG. 9 COMPARISON OF CHEMISTRY FACTORS FOR WELDS AND BASE METAL, GIVEN IN REVISION 2 1

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NUREG-0800

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u.s. NUCLEAR REGULATORY COMMISSION QWfi STANDARD REVIEW PLAN N%

OFFICE OF NUCLEAR REACTOR REGULATION e.eee 088 C/?##98.5 Mar [(88 5.3.2 PRESSURE-TEMPERATURE LIMITS

~

W REVIEW RESPONSIBILITIES on page.s S. 3 2 -E

-3 Primary - Materials Engineering Branch (MTEB)

- 5.

Secondary - None

- /7 1.

AREAS OF REVIEW 1.

Pressure-Temperature Limits The regulationsi requiring the imposition of pressure-temperature limits on the reactor coolant pressure boundary are the following:

Paragraph 50.55a of 10 CFR Part 50, " Codes and Standards," requires that structures, systems, and components be designed, fabricated, erected, con-structed, tested, and inspected to quality standards commensurate with the importance of the safety fuoction to be performed.

In addition, General Design Criterion 1 of Appendix A of 10 CFR Part 50, " Quality Standards and Records," requires that the codes and standards used to assure quality pro-ducts in keeping with the safety function be identified and evaluated to determine thei.r adequacy.

General Design Criterion 14 of Appendix A of 10 CFR Part 50, " Reactor Coolant Pressure Boundary," requires that the reactor coolant pressure boundary be designed, fabricated,' erected, and tested in order to have an extremely low probability of abnormal leakage, of rap"id failure, and of gross rupture.

Likewise, General Design Criterion 31, Fracture Prevention of Reactor Coolant Pressure Boundary," requires, in part, that the reactor coolant pressure boundary be designed with sufficient margin to assure that when stressed under operating, maintenance and testing, the boundary behaves in a nonbrittle man-1 ner and the probability of rapidly pr.opagating fracture is minimized.

Further, in order to assess the structural integrity of the reactor vessel, General Design Criterion 32, " Inspection of Reactor Coolant Pressure Boundary," re-quires, in part, an appropriate materials surveillance program for the reactor vessel beltline region.

s Rev. 1 - July 1981 USNRC STANDARD REVIEW PLAN Stenderd toview piene are propered for the guldence of the office of Nucteer Reectet Roguestien etsff nc : l': for the review of Com form industry arte h

pubi ese tory pree pas Stander to P5sn revish sectlene e to the Standard and C tent Se ty leer ew Not en sect 6ene of the Standard Formet have a seeroepending review plan.

Published standard review piens well be rew6aed p':M. es appropriate, to accommodate commente and to reflect new informe-tien and emportense.

Com toand t

forImpt ed and should be sent to the u.S. Nucteer Reguietory Commisolen

,------------,-------------~--------,---e-

-n

~-

~

9.

i The pressure-temperature limits imposed on the reactor coolant pressure boundary during operation and tests are reviewed in this section of the Standard Review Plan (SRP) to assure adequate safety margins of structural integrity for the ferritic components of the reactor coolant pressure boundary.

II.

ACCEPTANCE CRITERIA The requirements of paragraph 50.55a and General Design Criteria 1, 14. 31 and i

32 of Appendix A of 10 CFR Part 50 are met by the assurance that material of the reactor coolant pressure boundary possess adequate fracture toughness 1

properties to resist rapidly propagating failure and act in a nonbrittle manner when stressed under operating, maintenance, testing, and anticipated 1

operational conditions. The requirement in part, of General Design Criterion 32 is met by conducting a surveillance progr,am to monitor the change in fracture toughness properties of the ferritic materials in the reactor vessel.

The fracture toughness requirements for ferritic materials in the pressure-retaining components of the RCPB are specified for testing and operational conditions, including anticipated operational occurrences, in Section IV of Appendix G of 10 CFR Part 50.

This appendix requires the accestance and perfomance criteria of Appendix G of Section III of the ASME Joiler and Pressure Vessel Code. Pressure-temperature calculation procedures are described in Appendix G of the ASME code; while the detailed technical basis for the ASME code requirement is provided by the Welding Research Council (WRC) Bulletin 175, "PVRC Recommendation on Toughness Requirements for Ferritic Materials." Chan beltline region, ges in the fracture. toughness properties of materials in the resulting from neutron irradiation and the thermal environment, are monitored by a surveillance program in compliance to the requirements of 1

Appendix H of 10 CFR Part 50. The effect of neutron fluence on the shift in the nil-ducti,11,ty temperature of pressure vessel steel is, predicted by Regulatory Guide 1.99, Revision 2. "Effeet-of-Residual-Elements-en-pred4eted" Radia' tion

~

l i

Damage to Reactor vessel Materials."

~

1.

Applicable Regulations. Codes, and Basis Documents Appendices G and H of 10 CFR Part 50 describe the conditions that require pressure-temperature limits and provide the general basis for these limits. These appendices specifically require that pressure-temperature limits must provide safety margins et least as great as those recommended t

in the ASME Boiler and Pressure Vessel Ccu (hereinafter "the Code"),

Section III, Appendix G, " Protection Against Monductile Failure," during heatup, cooldown, and test conditions. Appendix G to 10 CFR Part 50 also j

requires additional safety margins whenever the reactor core is critical (except for low-level physics tests).

4 2.

Technical Bases Since many of the fracture toughness requirements for the ferritic materials in the pressure retaining components were not required at the time some of the reactor facilities were designed and constructed, the Materials Engineering Branch Technical Position MTE8 5-2, " Fracture Toughness Requirements," describe procedures for making estimates and assumptions on the fracture toughness properties of materials in the older plants.

Calculations ars required, and an evaluation is made by the reviewer to show compliance with the regulations and to show an adequate margin of 5.3.2-2 Rev. 1 - July 1981

._-.f,

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---

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quality and safety for the facility. When it h3s b2:n determined that certain requirements of Appendices G or H have not been strictly complied with by these older plants, and when it has been detemined that an equivalent level of quality and safety, as required by the regulations exist, then exemption to the specific requirements of these appendices will be granted by the Commission.

The principles of linear elastic fracture mechanics (LEFM) are used a.

to determine safe operational conditions.' The basic parameter of LEFM is the stress intensity factor, K, which is a function of the g

stress state and flaw configuration. An analytical method is used to determine the effects of real or postulated flaws. The minimum that can,cause failure is defined as the critical stress intensity Ky The factor, Kyg, and is the material parameter used in this method.

K of the material is either directly measured as a function of Ic temperature, or is conservatively estimated, using information from other fracture toughness tests, b.

The Code specifies the maximum Kyg, as a function of temperature, that can be assumed for the specific material, based on results of I

tests on the material used. This value is called Kgg, reference stress intensity-factor. The Code also provides rules for calculat-

.ing K, including definitions of postulated flaws, and specifies the g

safety factors to be applied. The acceptance criterion is that the calculated.

K of the material must always be higher,than the Kg gg as a function of temperature is expensive c.

D'irect measurement of the Kgg and time cons'using and requires more sample material than is usually available. Correlations between the K determined directly and gg results of simpler fracture toughness tests are not exact, but may be used if appropriate allowances are made for variations in material as a function behavior and data scatter. The Code gives values of KIR of temperature relative to a conservative determination of the reference temperature of the material. This reference temperature, NDT, is determined for the ferritic' materials of components for RT which operating and testing limit curves must be calculated. The effects cf radiation on the fractfure toughness of the material in the beltline region of the reactor vessel is accounted for by adjusting the RT of the affected material upward. The amount of upward NDT shift depends on the composition of the steel (especially its copper Conservative and phospherous nickel content), and the neutron fluence.

predictions of the effect of radiation on RTGu6de-h99-are-feetered-i eurves are achieved by adding margin to the best estimates as described in Regulatory Guide 1.99, Revision 2.

The continued conservatism of these predictions throughout plant l

life is verified by a mandatory material surveillance program described in Appendix H of 10 CFR Part 50.

l

.i I

d.

The Code specifies the stress components that must be used for the K calculations, and the factors that must be applied to each to l

y l

5.3.2-3 Rev. 1 - July 1981

provide adequate safety margins. The code, by reference to WRC-175, specifies the expression to use for calculating the K, using the g

applied stresses and the postulated flaw geoettry. Although calcu-I l

lations are usually made by a computer, curve: cre provided in the f,

Code to facilitate the use of conservative hand calculations if desired.

3.

Pressure-T m rature Requiroments The requirements for the pressure-temperature limits are as follows:

Pressure-Temperature Limits for Prese-vice Hydrostatic Tests a.

During preservice hydrostatic tests (if fuel is not in the vessel),

caused by pressure. The the K must be greater than the Kg jg expression used is:

K = K (pressure) < Kgg g

g b.

Pressure-Temperature Limits for Inservice Leak and Hydrostatic' Tests During perfomance of inservice leak s.nd hydrostatic tests, the KIR caused by pressure. The sust be greater than 1.5 times the Kg expression used is:

i K = 1.5 K (pressure) < Kgg g

g Pressure-Temperature Limits for Heatup and Cooldown Operations c.

sust be At all times during heatup and cooldown operations, the Egg caused by pressure and the Kg greater than the sum of 2 times the Kg caused by themal gradients. The expression used is:

K =. 2K (pressure) + K (thermal)

<K y

g y

gg d.

Pressure-Tesperature Limits for Core Operation At all times that the reactor core is critical (except for low power physics tests) the temperature must be higher than that required for inservice hydrostatic testing, and in addition, the pressure-temperature relationship shall provide at least a 40?F margin over that required i

for heatup and cooldown operations.

i III. REVIEW PROCEDURES The reviewer will select and emphasize material from the procedures described i

i below, as any be appropriate for a particular case.

1.

Preliminary Safety Analysis Report (PSAR)

Information in the PSAR is reviewed for a commitment that the fracture

. toughness of the ferritic materials in tha reactor coolant pressure boundary will comply with the requirements of Appendix G of 10 CFR Part 50, 4

5.3.2-4 Rev. 1 - July 1981 0

as detailed in Section III of the ASME Boiler and Pressure Vessel Code and that the materials in the beltline region of the reactor vessel will 9typf Appendices G and H of 10 CFR Part 50 and complywiththerequipMeets-of-Residual-Elements-on"PredictedRadiation E

Regulatory Guide 1.993 Damage to Reactor Vessel Materials."

2.

Final Safety Analysis Report (FSAR)

The limits in the plant Technical Specifications will be shown using real temperature. These curves and their bases are reviewed to detemine acceptability in the following areas:

1 a.

The limiting RT has been properly determined, and radiation ET effects are included in a conservative manner.

b.

Limits are shown for all required conditions.

c.

The limits proposed are consistent with the acceptance criteria described in II. above.

d.

The procedures for updating the limit curves, in conjunction with scheduled. tests on material surveillance specimens, are well defined and included in the Technical Specifications.

3.

Acceptability Determination Methods 1

l The redwer evaluates each limit curve for acceptability by performing check calculations using the simplified methods referenced in the Code and WRC Bulletin 175 that have been verified by the Materials Engineering Branch to yield conservative values. These methods are described in I

]

detail by examples below, and the curves necessary to perform the calculations are included herein as Figures 1, 2 and 3.

l a.

Preservice Hydrostatic Tests The preservice hydrotest at 1.25 design pressure corresponds to the standard Code component hydrotest usually perfomed in the shop, but in this case it is the hydrotest for field welds, so it may involve the entire reactor coolant system.

i The Code recommends that component hgdrostatic tests be run at a temperature no lower than RTE T + 60 F, but also recommends that j

system tests should have more stringent requirements. The MTEB position is that the minimum temperature for the preservice test, if fuel is not in the vessel, be determined using the methods of Code Section III, Appendix G, using less stringent factors.

First, the RT of the vessel material must be detersir.ed. This is ET defined by the Code for new plants, and is essentially a conserv'ative value of the ETT as determined by drop weight test. Guidelines for estimating the RT if the prescribed tests have not been run are ET

~

given by Branch Technical Position - MTEB 5-2, " Fracture Toughness

)

Requirements." Technical justification for all estimates of RTET l

must be provided by the applicant.

5.3.2-5 Rev. 1 - July 1981 l

l The t;ughness of the mat: rial is a functitn of the diffGrence between the RT of the material and the temperature of interest. The Code ET provides a curve (Figure G-2210.1) for the allowable calculated stress intensity factor (Kgg) as a function of the temperature relative to RT Refer to Figure 2 herein.

ET.

The Code also provides a recommended basis for calculating K,.

g including recommendations for assumed flaw size and shape, and appropriate front and back surface correction factors. Because the assumed flaw size is proportional to the wall thickness, t (flaw depth = 0.25 t ard length = 1.5 t), the Kg expressions are simpli-fied to multiples that are a function only of wall thickness and stress level. These factors, M, for membrane stresses and M8 I'"

bending stresses, are provided in oraphical form in Figure G-2214.1 of the Code.

Refer to Figure 1 herein.

The criterion recommended by MTEB can be expressed w; I

Kg<KIR for the shell region.

To get K, the stress level and wall thickness must be known. The g

pressure for the hydrostatic test is 1.25 times the design pressure, so that the higher of two simple methods described below to approximate the membrane stress should be accurate enough for this purpose:

stress = 1.25 times the Code allowable (5,)

stress =h where P is the test pressure and r is the vessel radius. As an example, assume a vessel with a design pressure of 2500 psig, made of steel with an S, of 26,700 psi, and a minimum yield strength of 1

50,000 psi. The stress for the preservice hydrotest is then 26,700 x 1.25 = 33,400 psi, or (1.25) 500) (95) = 33,400 psi, for a vessel with a

(

radius of 95 inches and a wall thickness of 9 inches.

1 The next step is to determine the factor to apply to this stress to obtain K. Figure G-2214.1 (reproduced here as Fig. 1) provides g

several curves, depending on the ratio of the stress level to the yield strength of the material.

In this case, the stress level is l

33,400; the yield strength is conservatively assumed to be 50,000 so the curve for a ratio of.? should be used.

(A ratio equal to or higher than the actua! ratio must be used for conservatism.) For a 9-in. thick vessel /,4 = 3),is then-the value of M, from Figure G-2214.1 i5 2.94.

The K for W s case g

i 5.3.2-6 Rev. 1 - July 1981 L

y = (M,) (Membrane Stress)

K y = (2.94) (33,400) = 98,300 psi E K

From Figure G-2210.1 (reprococed here as Fig. 2), a temperature of s

of this level.

at least RTNOT + 120*F is necessary for a Ky of 40'F is assumed, the required If, for exauple, an original RTET temperature is then 40 + 120, or 160'F.

6 Inservice Lea'k and Hydrotest.

b.

The temperatures for the inservice leak and hydrotest, performed at operating pressure and about 1.1 operating pressure, respectively, y

The differences are are calculated in essentially the same way.

that a factor of 1.5 must be applied to the calculated K to provide j

g extra margin, and the stress levels are lower, so the value of M,is taken from a lower ratio curve.

Using'the same vessel as an example, with a normal operating pressure (P,) of 2250 psi, the membrane stress for the leak test can be approximated as:

operating pressure x allowable stress design pressure or h x 26,700 = 24,000 psi This is about half of the minimum yield strength, so the M, is taken that must from the 0.5 ratio curve, and is 2.87. The calculated Ky 1

be assumed is then:

K = (1.5) (M,) (Membrane Stress) y or K = (1.5) (2.87) (24,000) = 103,500 psi 5 y

From the K curve, a temperature of about RTET + 125?F is required.

IR would probably have been As this is an inservice test, the RTET increased from its original value of + 40?F by some shift caused by Assume this shift is 100?F, thus the temperature for the radiation.

leak test must be at least:

40 + 100 + 125 = 265?F The inservice hydrotest temperature (at 1.1 P,) is determined in exactly the same way,*and requires a minimum temperature of about ET + 133?F, or 273 F.

RT

~.

Heatup, Cooldown, and Normal Operation c.

For normal op, ration, which includes upset conditions and startup and shutdown procedures, operating limit curves must be provided Rev. 1 - July 1981 5.3.2-7 I

l

_~ n- - _...

..,n

e that show the maximum permissible pressure at any temperature from cold shutdown conditions to full pressurization conditions.

Reactor vendors have developed computer codes to perform the necessary calculations, because thermal stresses must be included, l

and hand calculations of even moderate sophistication are very time consuming. WRC Bulletin 175 includes a set of curves derived from computer programs that can be used to approximate the Kg caused by thenaal stresses, as a function of wall thickness and rate of ten-perature change.

Pressure-temperature curves developed using these approximations agree fairly well with those determined using much more rigorous procedures, and can be used with confidence to evaluate the proposed operating limits given in Technical Specifications.

3 These curves require the calculation of only 3 to 5 points.

Either allowable pressure at a given temp' rature, or allowable temperature e

i at a given pressure can be calculated.

It is usually more convenient to calculate allowable minimum temperature, so this j

method will be used in the example.

Using the same reactor vessel as in the previous example, and a rate 1

of temperature change of 50*F per hour, calculations of required temperatures for several pressures are illus+ rated. The curves for thermal offacts given in WRC Bulletin 175 are very conservative.

l thus no additional margin need by applied to the K from thermal g

stress, but a factor of 2.0 is used on primary stresses. The basic expression is then:

l.

K 1 2 K (me h ane) + Kg (thermal)

IR g

l K (membrane) is calculated exactly as in the previous examples'.

g K (thermal) for a 9-in. thick wall, at 50*/hr is about 12,000 psi g

E from Figure 4-5, WRC Bulletin 175 (reproduced here as Fig. 3).

Thus, for a pressure of 2250 psig, a membrane stress of 24,000 psi, and M,of 2.87, the basic expression is given by l T' Kig > (2)(24,000)(2.87) + 12,000 = 150,000 pst E From the Kgg curve, a temperature of RTET + 158'F is required.

)

With an RTET of 140*F (including irradiation effects), the temperature required for operating pressure at a heatup or cooldown rate of 50*/hr is then l

140 + 158 = 298'F l

For a pressure of 1/2 of operating (1125 psig), the membrane stress is 1/2 of that at operating pressure, or 12,000 psi.

{

The M, can be taken from the 0.5 h ratio curve in Figure G-2214.1 t

j (reproduced as Figure 1 herein), so is again 2.87.

i Kgg 1 (2)(12,000)(2.87) + 12,000 = 81,000 psi E I

5.3.2-8 Rev. 1 - July 1981 1

-_,-n.,.----n..__--

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From the K, curve, the minimum temperature is RTET + 100*F, or 3

140 + 100 = 240'F.

1 The same calculation for a pressure of 1/5 operating pressure (450 psig and 4800 psi stress) is similar, but in this case the stress is less than.1 of the yield strength, so the ( (from the

.1 ratio curve) is only 2.82.

K 1 (2)(4800)(2.82) + 12, M = 39,000 psi E IR The K curve shows that the minimum temperature is RTET + 0'F, or gg 140*F.

Three points on a 50'/hr operating limit curve for this vessel at this time in its service lifetime have thus been calculated:

Pressure Min. Temperature (psic)

(Fahrenheit) 450 140 1150 240 2250 298 A smooth curve drawn through these points will very closely approximate the results usingamore rigorous methods.

d.

Core Operation Appendix G,10 CFR Part 50, specifies pressure-temperature limits for core operation to provide additional margin during actual power production.

The pressure-temperature limits for core operation (except for low power physics tests) are that the reacter vessel must be at a tem-perature equal to or higher than the minious temperature required for the inservice hydrostatic test, and at least 40*F higher than the minimum pressure-temperature curve for heatup and cooldown calculated as described in the preceding section. The minimum temperature for the inservice hydrostatic test for the vessel used in the preceding example was 273*F A vertical line at 273*F on the pressure-temperaturecurve,interhe.cting a curve 40*F higher than the pressure-temperature limit curve as determined in the preceding section, constitutes the limit for core operation for this example.

IV.

EVAL.UATION FINDINGS 1

The reviewer verifies that r,ufficient information has been provided to satisfy l

the requirements of this $RP section and that the completeness and technical adequacy of his evaluation will support the following statement in the staff's safety evaluation report:

The pressure-temperature limits imposed on the reactor coolant system for all operating and testing conditions to assure adequate j

safety margins against nonductile or rapidly propagating failure are 5.3.2-9 Rev. 1 - July 1981

in conformance with the fracture toughness criteria of Appendix G of 10 CFR Part 50 and Section III, including Appendix G, " Protection Against Nonductile Failure," of the ASME Boiling and Pressure Vessel Code. The change in fracture toughness requirements of the pressure vessel during operation will be determined by Appendix H of 10 CFR Part 50. The use of operating limits, based upon the criteria defined in Standard Review Plan Section 5.3.2, provides reasonable assurance that nonductile or rapidly propagating failure will not occur, and constitutes an acceptable basis f6r satisfying the requirements of paragraph 50.55a of 10 CFR Part 50 and General Design Criteria 1, 14, 31 and 32 of Appendix A of 10 CFR Part 50.

V.

IMPLEMENTATION The following is intended to provide guidance to applicants and licensees regarding the NRC staff's plan to using this SRP section.

Except in those cases in which the applicant proposed an acceptable alternative method for complying with specific portions of the Commission's regulations, the method described herein will be used by the staff in its evaluation of confor-mance with Commission regulations.

Implementation schedules for confomance to parts of the method discussed here-in are contained in the referenced regulatory guide.

VI. REFERENCES 1.

10 CFR Part 50, Appendix A General Design Criteria 1, 14, 31, and 32.

2.

10 CFR Part 50, Appendix G " Fracture Toughness Requirements."

3.

10 CFR Part 50, Appendix H, " Reactor Vessel Material Surveillance Program Requirements."

4.

ASME Boiler and Pressure Vessel Code,Section III, including Appendix G,

" Protection Againsst Nonductile Failure," American Society of Mechanical Engineers.

5.

WRC Bulletin 175, "PVRC Recommendation on Fracture Toughness," Welding Research Council.

6.

Branch Technical Position - MTES 5-2, " recture Toughness Requirements for Older Plants," attached to this SRP section.

7.

Regulatory Guide 1.99, " Effects of Resi, dual Elements on Predicted Radiation Damage to Reactor Vessel Materials."

8.

10 CFR Part 50, paragraph 50.55a, " Codes and Standards."

l I

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BRANCH TECHNICAL POSITION - MTEB 5-2 FRACTURE TOUGHNESS REQUIREMENTS 3,

A.

Background

l Current requirements regarding fracture toughness, pressure-temperature limits.

and material surveillance are covered by the ASME Code and Appendices A, G, and H to 10 CFR Part 50. The purpose of this branch technical position is to summarize these requirements and provide clarification, as necessary, i

t Since many of these requirements were not in force when some plants were designed and built, this position also provides guidance for applying these requirements to older plants. Also included is a description of acceptable procedures for making the conservative estimates and assumptions for older

(

It plants that may be used to show compliance with the new requiremen cations for any estimates of material properties required by the regulations before exemption to the regulations may be granted.

8.

Branch Technical Position 1.

Preservice Fracture Toughness Test Requirements.

The fracture toughness of all ferritic materials used for pressure-retaining components of the reactor coolant pressure boundary shall be evaluated in accordance with the requirements of Appendix G,The 10 CFR Part 50, as augmented by Section III of the ASME Code.

fracture toughness test requirements for plants with construction permits prior to August 15, 1973 may not comply with the new codes and Regulations in all respe:ts. The fracture toughness of the materials for these plants must be assessed by using the available test data to estimate the fracture toughness in the same teres as This must be done because the operating the new requirements.

limitations imposed on old plante must provide the same safety margins as are required for new plants.

4 1

I'" Y"I N'tI'I' 1.1 Detemination of RTET l

l Temperature limitations are determined in relation to a characteristic j

temperature of the material, RTET, that is established from results of fracture toughness tests. Both drop weight E TT tests and Charpy The NOTT temperature, V-notch tests must be run to determine the RTET.

II' **

as determined by drop weight tests (ASTM E-208) is the RTET 60*F above the NOTT, at least 50 ft-1bs of energy and 35 mils lateral expansion are obtained in Charpy V tests on specimens oriented in j

the weak direction (traverse to the direction of maximum working).

In most cases, the fracture toughnes: testing performed on vessel i

I material for older plants did not include all tests required to l

in this manner. Acceptable estination set, hods determine the RTET for the most common cases, based on correlations of data from 4 large number of heats of vessel material, are provided for guidance /

in determinine RT when measured values are not available._

g7 i

Rev.1 - July 1981 5.3.2-13 i

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e 4

i (1) If dropweight tests were not performed, but full Charpy V-notch curves were obtained, the NDTT for SA-533 Grade B, Class 1 plate and weld material may be assumed to be the temperature at which 30 ft-Ibs was obtained in Charpy V-notch tests, or O'F, whichever was higher.

(2) If dropweight tests were not performed on SA-508, Class II forgings, the NOTT may be estimated as the lowest of the following temperatures:

(a) 60*F.

(b) The temperatures of the Charpy V-notch upper shelf.

(c) The temperature at which 100 ft-lbs was obtained on Charpy V-notch tests if the upper-shelf energy values were above 100 ft-lbs.

(3) If transversely-oriented Charpy V-notch specimens were not tested, the temperature at which 50 ft-lbs and 35 mils LE would have been obtained on traverse specimens may be estimated by one of the following criteria:

(a) Test results from longitudinally-oriented specimens reduced to 65% of their value to provide conservative estimates of values expect,ed from transversely oriented specimens.

(b) Temperatures at which 50 ft-lbs and 35 mils LE were obtained on longitudinally-oriented specimens increased 20*E to provide a conservative estimate of the temperature that would have been required to obtain the same values on transversely-oriented specimens.

i (4) If limited Charpy V-notch tests were performed at a single temperature to confirm that at least 30 ft-lbs was obtained, that temperature may be used as an estimate of the RTNOT P

vided that at least 45 ft-1bs was obtained if the specimens were longitudinally oriented.

If the minimum value obtained may be estimated as 20'F was less than 45 ft-lbs, the RT above the test temperature.4, NOT 1.2 Estimation of Charpy V Upper-Shelf Energies I

For the beltline region of reactor vessels, the upper sholf toughness must be adequate to accommodate degradation by neutron radiation. The original minimum shelf energy must be 75 ft-1bs for vessels with an estimated end of life neutron fluence (> 1 MeV) of 1 x 101' and over. A value of 70 ft-lbs is considered adequate for material for vessels that will be subjected to lower fluences.

If upper-shelf Charpy energy values were not obtained, conservative estimates should be made using results of tests on specimens from the first surveillance capsule removed.

i 5.3.2-14 Rev. 1 - July 1981 j

l

L ;.

If tests were only made en longitudinal specimens, the values should be reduced to 65% of the longitudinal values to estimate the transverse properties.

1.3 Reporting Requirements Fracture toughness information required by the Code and by Appendix G, 10 CFR Part 50, must be reported in the FSAR to provide a basis for evaluating the adequacy of the operating limitations given in the Technical Specifications.

In the case of older plants, the data may be estimated, using the procedures listed above, or other methods that can be shown to be conservative.

i 2.

Operating Limitations for Fracture Toughness 2.1 Required Pressure-Temperature Operating Limitations As required by Appendix G, 10 CFR Part 50, the following operating limitations shall be determined and included in the Technical Specifications. The basis for determination shall be reported, and j

is the' responsibility of the applicant, but in no case shall the i

limitations provide less safety margin than those determined in accordance with Appendix G, 10 CFR Part 50, and Appendix G to i

Section III of the Code.

(1) Minimum temperatures for performing any hydrostatic test involving pressurization of the reactor vessel after installation in the system.

(2) Minimum temperatures far all leak and hydrostatic tests pefformed after the plant is in service.

(3) Maximum pressure-minious temperature curves for operation, including startup, upset, and cooldown conditions.

(4) Faximum pressure-minimum temperature curves for core operation, i

2.2 Recommended Bases for Operating Limitations 2.2.1 Leak and Hydrostatic Tests t

(1) It is recommended that no tests at pressures higher than design pressure be conducts.! with fuel in the vessel.

(2) Tests at pressures less than design pressure should be condud.ed at temperatures calculated according(to Appendix G of Section III of the Code for the beltline region including conservative estimates of radiation damage, see Section 3.0 below) if the maximum calculated primary stress in no other region of the of the vessel exceeds 1.25 S, during the test, and the RTET l

beltline is assumed to be at least 30*F above that of the 1

higher stressed regions.

If primary stresses are calculated to be over 1.25 S,in any region during the test, the RTET of the i

vessel must be assumed to be at least 50*F higher than that of t

5.3.2-15 Rev. 1 - July 1981

any region where the calculated primary stresses are over 1.25 S,.

(3) Alternatively, a fracture mechanics analysis, with technical justification for all assumptions and bases, may be made to determine the minimum test temperature.

In no event shall the minimum temperature be lower than that resulting from calculations for the beltline region in accordance with Appendix G of the Code.

2.2.2 Heatup and Cooldown Limit Curves Heatup and cooldown pressure-temperature limit curves may be determined usingsinglefstresscalculations,usingthemethodgivenin l

Appendix G of the Code. The effect of themal gradients may be conservatively approximated by the procedures in Appendix G of the Code or from Figure 4-5 in WRC Ballatin 175.

Calculations need only be perfomed for the beltline region, if the RT of the beltilne is demonstrated to be adequately higher than NDT the RT for all higher stressed regions.

NOT Alternatively, more rigorous analytical procedures may be used, provided that the intent of the Code is set, and adequate technical justification for all assumptions and bases is provided.

2.2.3 Core Operation Limits To provide added margins during actual core operation, Appendix G, 10 CFR Part 50 requires a minimum temperature during core operation, and a 40*F margin in tagerature over the pressure-temperature limits as detemined for heatup and cooldown in 2.2.2 above. The minimum temperature, regardless of pressure, is the temperature calculated for the inservice hydrostatic test according to 2.2.1 above.

2.2.4 Upset Conditions The pressure-temperature limits despribed in 2.2.2 and 2.2.3 above are applicable to upset conditions.

Normal operating procedures must pemit variations from intended operation, including all upset conditions, without exceeding the limit curves.

4 2.2.5 Emergency and Faulted Conditions It is recognized that the severity of a transient resulting from an emergency or faulted conditici. is not directly related to operating conditions, and resulting temperature-stress relationships in the reactor coolant boundary components are primarily system dependent, i

and therefore not under direct control of the operator.

For these reasons, operating Ilmits for emergency and faulted conditions are not a requirement of the Technical Specifications.

I 5.3.2-16 Rev. 1 - July 1981

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The SAR should present descriptiens of the continued integrity of all vital components of the RCPS during postulated faulted conditions.

It is recommended that such descriptions be made in as realistic a

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manner as possible, avoiding grossly overconservative assumptions and procedures.

2.3 Reporting Requirements The Technical Specifications must include the operating and test limits discussed above, and the basis for their determination. The Technical Specifications must also include information on the intended operating procedures, and justify that adequate margins between the expected conditions and the limit conditions will be provided to protect against unexpected or upset conditions, i

3.

Inservice Surveillance of Fracture Toughness The reactor vessel may be exposed to significant neutron radiation during the service life. This will affect both the tensile and toughness properties. A material surveillance program in conformance with Appendix H, 10 CFR Part 50, must be carried out.

J 3.1 Surveillance Program Requirements f

The minimum requirements for the surveillance program are covered by Appendix H, 10 CFR Part 50. It is strongly recommended that con-sideration by given to the desirability of additional surveillance methods, such as the inclusion of CT, DWT, DT, or other specimens to l

provide the capability of redundant test methods and analytical procedures, particularly if the estimated neutron fluence is over 2 x 1018, or the toughness of the vessel material is marginal.

The selection of mat,erial to be included in the surveillance procram shocid be in accordance with ASTM E-185-73 82, unless the intent of the program is better realized by using more rigorous criteria.

For and upper shelf example, the approach of estimating the actual RTET l

toughness of each plate, forging, or weld in the beltline as a function of service life, and choosing as the surveiilance materials l

those that are expected to be most limiting, may be preferable in some cases. This would include consideration of the initial RTET' I

l the upper. shelf toughness, the expected radiation sensitivity of the material (based on copper and phoopwrees nickel content, for example) and the neutron fluence expected at its location in the vessel.

3.2 SAR Requirements l

The adequacy of the surveillance program cannot be evaluated unless all pertinent information is included in the SAR.

Information requested for beltline materials includes the following:

(1) Tensile properties.

(2) DWT and Charpy V test results used to determine RTET*

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1 i

5.3.2-17 Rev. 1 - July 1981 i

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6 (3) Charpy V test results to determine the upper shelf teughness.

(4) Composition, specifically the copper and phosphorous content.

(5) Estimated maximum fluence for each beltline material.

(6) List of materials included in the surveillance program, with basis used for their selection.

3.3 Surveillance Test Procedures Surveillance capsules must be removed and tested at intervals in accordance with Appendix H, 10 CFR Part 50. The proposed removal and test schedule shall be included in the Technical Specifications.

3.4 Reporting Requirements All information used to evaluate results of the tests on surveillance materials, evaluation methods, and results of the evaluation should be submitted with the evaluation report. This should include:

(1) Original properties and compositions of the materials.

(2) Fluence calculations, including original predictions, for both surveillance specimens and vessel wall.

(3) Test results on surveillance specimens.

(4) Basis for evaluation of changes in RT and upper shelf ET toughness.

(5) Updated prediction of vessel properties.

3.5 Technical Specification Changes Changes in the operating and test limits recommended as a result of evaluating the properties of the surveillance material, together with the basis for these changes, shall be submitted to the Division of Licensing for approval.

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5.3.2-18 Rev. 1 - July 1981 l

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REGULATORY ANALYSIS REGULATORY GUIDE 1.99, REVISION 2 RADIATION DAMAGE TO REACTOR VESSEL MATERIALS 1.

STATEMENT OF THE PROBLEM One obvious constraint on the operation of a reactor is prevention of fracture of the vessel. This is accomplished, in part, by warming it before pressurization, following the pressure-temperature (P-T) limits given in the Technical Specifications (Tech Specs). Neutron radiation damage to the reactor vessel is compensated for by shifting the P-T limits up the temperature scale every few years by an amount corresponding to the shift in the Charpy test transition temperature produced by the accumulated neutron fluence.

The NRC regulates this process on the basis of Appendices G and H,10 CFR Part 50.

Paragraph V.A of Appendix G requires:

"The effects of neutron radiation...are to be predicted from the results of pertinent radiation effects studies...."

Since Revision 1 of Regulatory Guide 1.99 was published eight years ago, there has been a significant accumulation of power reactor surveillance data, which constitutes a much more pertinent basis for the Guide than was available when Revision 1 was written.

Revision 2 is based entirely on the surveillance data, and its issuance will provide a basis for licensing decisions that con-stitutes the most pertintnt results available, in conformance with the regulation.

It may be asked why the Guide is needed if plants now have surveillance data of their own.

Of course, for the newer plants such data are not yet avail-able.

For many older plants, unfortunately, the materials in the survefilance capsules are not the controlling materials for that reactor according to our present day understanding. Thus, instead of using the plants' own surveillance Q 4j 'jlA/l ? A '

rf3 08/22/85 1

REG ANAL RG 1.99 REV 2

results directly, the staff must rely on calculated values based on the chemi-cal composition of the vessel materials and the neutron fluence.

Regulatory Guide 1.99 Revision 2 upgrades and expands the calculative procedures that are acceptable to the NRC, and it describes acceptable proce-dures for using plant-specific surveillance data when they become available.

The Guide is used in any regulatory action that requires knowledge of the fracture toughness of reactor vessel beltline materials.

Three examples of such actions are:

(1) setting pressure - temperature (P-T) limits for heatup and cooldown, (2) evaluating transients that threaten the integrity of the reactor vessel, such as low temperature overpressurization and pressurized thermal shock events, (3) evaluating flaws found during inspection.

In any of these analyses, a key input to the calculation is the fracture toughness of the material as a function of temperature.

The ASME Code gives reference values of toughness as a function of temperature relative to RTNDT, the " reference temperature nil-ductility transition" of the material.

The Code also describes how to measure the initial RT f r the unirradiated material. This Guide gives calculative NDT caused by neutron radiation.

procedures for ARTNDT, the adjustment of RTNDT The Guide also describes how to combine the initial and the " delta" with a suit-NDT able margin to obtain a value of RT that covers the uncertainties in both.

From analysis of the new data base, and from experience in applying the Guide, the need for certain changes became clear.

Nickel has been found to increase the Charpy shift in the presence of copper, and should be a factor in the calculations.

Thus, some reactor vessels with high nickel welds, which were made when nickel was added in the welding process, have more susceptibility to radiation than previously thought.

Conversely, some early reactor vessels that were made with no deliberate alloying additions of nickel have lower sensitivity to radia-tion.

Implementation of Revision 2 will remedy these situations.

08/22/85 2

REG ANAL RG 1.99 REV 2

The effects of copper and nickel content on the sensitivity of welds to radiation damage are different than they are for base metal--so different as to require separate treatment of welds and base metal in this Guide.

The fluence function needs revision.

The calculative procedure needs to be amended to prescribe mean values instead of upper bound values and to state the margin separately.

I Procedures for calculating the attenuation of radiation damage through the vessel wall need to be stated specifically.

Improved knowledge of scatter in the surveillance data base made it necessary to rewrite the criteria for use of plant-specific surveillance data in setting P-T limits for that reactor.

2.

OBJECTIVE 2.a General Objective of Regulatory Guide 1.99 The objective of the guide is best described by reference to the schematic pressure-temperature (P-T) diagram, Figure 1.

The following discussion is mainly applicable to PWRs.

(See paragraph 4.a.1.6) The upper-lef t boundary of the operating zone, the P-T limit, appears in the Technical Specifications The P-T for all plants together with certain limits on heatup/cooldown rates.

limits are based on Appendix G, 10 CFR 50, which incorporates Appendix G and parts of Section III of the ASME Boiler and Pressure Vessel Code.

And, the P-T limits are affected by Regulatory Guide 1.99 as described in the previous Section.

In the upper-left corner of Figure 1 is a region labelled " hazardous to vessel integrity" which is bounded by a set of curves (instead of one curve) to indicate.that higher cooldown rates increase the hazards.

This boundary moves upscale in temperature as radiation damage accumulates during the operating Iffe of the vessel. The objective of the margins added in calculating P-T limits is to place the operating zone for heatup/cooldown far enough from the hazardous 08/22/85 3

REG ANAL RG 1.99 REV 2

region to provide the operator time to diagnose and correct system transients such as low temperature overpressurizations (LTOPS) and rapid cooldown events i

that could threaten vessel integrity.

2.a.1 Sources of Margin in P-T Limit Calculations A discussion of the sources of m3rgin that are present in the calculative procedure for P-T limits given in the ASME Code and NRC Regulations is in order.

First, the postulated flaw is a semiellipse 0.25T deep by 1.5T long (2.2 in. x 13 in., typically, for a PWR).

From the flaw size distribution used in the VISA code, the probability of such a flaw in the critical beltline weld 4

of a reactor vessel is about one in 60,000. The probability may be debatable, but clearly there is margin in the 0.25T flaw assumption.

Nevertheless, the use of a 0.25T flaw is an accepted feature of the Code procedure for calcula-tion of P-T limits, which we endorse.

This is partly because the efficacy of the flaw detection and sizing in non-destructive examination is still debatable, and partly because some flaws are the result of metallurgical conditions that also degrade the toughness of the adjacent material.

Second, there is a safety factor of two on the stress intensity factor due to primary membrane stress (in the beltline, pressure stress) and a factor i

of one on that due to thermal stress.

These factors were chosen by the Code writing bodies with the help of the basic document, WRC Bulletin 175,* but the I

rationale for the factor of 2 on pressure is not expitcitly stated therein.

A third source of margin is the requirement that the toughness-temperature function used in these calculations should be the "K, curve" for crack arrest, g

curve" for static called the "K curve" in Appendix G, rather than the "KIc IR crack initiation.

(See Figure 4 of the Regulatory Analysis).

The KIR curve is the lower bound of dynamic and crack arrest toughness values for specimens some of which were full thickness.

Its adoption for Appendix G of the ASME Code

• PVRC AdHoc Task Group on Toughness Requirements," PRVC Recommendations on Toughness Requirements for Ferritic Materials," Welding Research Council Bulletin No. 175, August, 1972.

08/22/85 4

REG ANAL RG 1.99 REV 2 i

derived from a philosophy that was subscribed to by many, especially the researchers from the Naval Research Laboratory, that many service failures were the result of dynamic loading.

In a reactor vessel beltline, the pressure and thermal stresses are not applied dynamically in the sense intended here, not.

even in a thermal shock situation.

However, one can postulate a case for rapid loading by postulating a small defect that is surrounded by a nugget of brittle metal that carries stress up to some critical level, then cracks open suddenly, presenting the sound metal with a running crack.

Thus, the requirement to use the crack arrest toughness curve instead of the crack initiation curve is not entirely a matter of adding margin - in some unknown percentage of cases, it is the realistic thing to do. Whether based on this scent.rio or on simple conservatism, the calculations required by the ASME Code for P-T limits are to be based on the K curve.

For evaluation of accident conditions, however, IR Section XI uses the K curve.

The temperature difference between the KIR g

curve and the K curve is about 65*F in the region of interest.

jc The fourth source of margin in the calculation of P-T limits is the use of lower bound toughness curves.

In the VISA code, however, toughness is simulated from the mean K curve with a distribution about the mean of i 10 per cent Ic (1 sigma).

The use of a lower bound curve is consistent with Code philosophy in setting allowables, and was believed justified at the time it was drawn, because the data base was for only one heat of plate material.

The temperature margin between the K curve in the Code and the best-estimate curve used in 1c the VISA code is about 40*F in the region of interest.

(See Figure 5.)

The fifth source of margin is that required by Revision 2, paragraph C.1.a.(3).

Toughness values, K and K an ghen in tem of (T-RTNOT)*

IR gc The margin added to RT covers uncertainty in the initial RT for unirra-NDT NDT diated material as well as the uncertainty in ARTNOT, which is specified in Revision 2 to be 56*F for welds and 34*F for base metal.

These values resulted from the regression analyses of the data and represent two-sigma upper bounds.

They are considered to cover the uncertainty caused by scatter of the data about the mean and uncertainties in the copper, nickel, and fluence.

These variables are entered in the calculative procedure as best-estimate or mean values.

The margin also is assumed to cover uncertainty arising from possible differences between the copper and nickel contents of the weld and base metal samples and 08/22/85 5

REG ANAL RG 1.99 REV 2

those of the actual vessel materials at the location of interest in a fracture analysis.

In conclusion, it should be pointed out that the efforts to provide margin in the five areas described above are quite consistent with the requirements of General Design Criterion 31.

It states in part "The design shall reflect consideration of...the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws."

Constraints on the amount of margin that can prudently be provided derive from the need for an operating zone of reasonable' width for efficiency in heatup/

cooldown operations and the presence of a lower limit on pressure at a given temperature based on avoidance of pump cavitation or thermal hydraulic problems caused by an approach to saturation conditions.

2.b Summa ry An attempt has been made to describe the objectives of Regulatory Guide 1.99 by placing it in context with the other documents that provide the basis for procedures to assure prevention of fracture of the reactor vessel and by describing the sources of margin provided in those procedures.

The scope of the Guide is restricted to one part of the procedures.,

to provide an acceptable basis to account for the effects of neutron radiation on the fracture toughness of reactor vessel materials.

This is needed in the calculation of P-T limits, in analysis of transients that threaten vessel integrity and in the analysis of beltline flaws.

The objective of Revision 2 is to upgrade the calculative procedures in Regulatory Guide 1.99 by basing them on the most pertinent radiation data and the best available understanding of radiation damage in reactor vessel mate-rials in accordance with 10 CFR 50, Appendix G.

The specific changes made in preparing Revision 2 are itemized in Section 1, above.

08/22/85 6

REG ANAL RG 1.99 REV 2

3.

ALTERNATIVES I

The alternatives to issuance of Revision 2 are to leave Revision 1 in place or to eliminate the Guide altogether.

The latter can be disposed of the staff reviews several P-T limits per year, plus an occasional quickly:

transient and flaw indication and clearly needs a published basis for its There is at present nothing equivalent to Regulatory Guide 1.99 in reviews.

the ASME Code.

ASTM Standard Guide E-900-83 contains an equation relating the Charpy shift to copper content and fluence., but, it is out of date, and the Standard does not contain guidance on the use of plant-specific surveillance The alternative of continuing to use Revision 1 has a safety impact on data.

many plants and penalizes other plants; a detailed analysis of these conse-quences is given in the following section.

4.

CONSEQUENCES 4.a.

Costs and Benefits of Alternatives 4.a.1.

Application to P-T Limit Calculations 4.a.1.1 Effects on P-T limits for All Plants.

In this Section, it will be shown that implementation of Revision 2 will mean that for about one-half of the operating reactors the P-T limits should be moved upscale to higher temperatures by amounts that depend on fluence level and copper and nickel content of the beltline materials.

Referring again to Figure 1, this means that the region labelled " hazardous to vessel integrity" extends farther upscale in temperature than would be pre-dicted if Revision 1 were continued to be used; hence the P-T limit should be moved also, to maintain the margin.

About one sixth of the plants will be able to operate with their present P-T limits for a longer period than previously determined based on Revision 1.

The remaining third will be assentially unaffected.

To determine the consequences of changing from Revision 1 to Revision 2 in I

our review of P-T limits, the first step was to calculate what differences would REG ANAL RG 1.99 REV 2 08/22/85 7

l

l.

result if all plants should review their P-T limits this year first according to Revision 1 and then according to Revision 2.

The fluence value used for each plant was chosen, assuming that the goal was to have P-T limits that would be good for 4 or 5 years in the case of PWRs and somewhat longer for BWRs.

i l

l The importance of fluence level is shown in Figure 2 which illustrates i

for one material how the " trend curve" from Revision 1 compares to correspond-ing curves from Revision 2.

The trenc curve from Revision 1 is an upper bound I

curve, hence for comparison, the Revision 2 values are mean plus-margin, cal-culated as described in the Guide. Note how the two curves from Revision 2 cross that for Revision 1 at fluences of about 3 x 101s n/cm2 for base metal and somewhat higher for welds, indicating that plants having 0.35 Cu and 0.6 Ni and low fluences will be ratcheted whereas those having higher fluences will l

l get some benefits.

Figure 2 is drawn for 0.35 percent copper and 0.6 percent nickel.

For other compositions, the general appearance of the figure would be l

similar, with the crossovers occurring at different fluences.

In Figures 3 and 4, the numerical differences between shift values calcu-lated for Revision 1 and those for Revision 2 (mean plus margin) are tabulated for eight copper levels, three nickel levels, and seven fluence levels.

In the figures, the boundaries between conditions for which a plant would be ratcheted j

(the positive values in the Table), and those under which plants would benefit (the negative values) are indicated.

Having Figures 3 and 4, the offect on each plant of a change from Revi-sion 1 to Revision 2 as a basis for the P-T limits was readily estimated on 1

the basis of the copper and nickel content of its critical material and the fluence value described above.

For the eight plants ratcheted 50-100*F, actual shift calculations were made following Revisions 1 and 2.

l Our information on the limiting material and its copper and nickel content l

was quite good for PWR operating reactors, fairly complete for operating BWRs, but for many plants under construction it was limited to knowing whether or not the reactor vessel was bought to a low-copper specification.

However, inspec-l tion of Figures 3 and 4 shows that the numbers do not vary radically from box l

08/22/85 8

REG ANAL RG 1.99 REV 2 l

to box in most cases, and we believe the results are sufficiently accurate for discussion parposes.

Table 1 presents a summary of the changes in P-T limits that would result from a change from Revision 1 to Revision 2.

For example, the Table shows that, of the 81 operating light-water reactors (including three that are licensed only for low power testing), about 30 would find little dif ference (120*F) in the use of Revision 2 or Revision 1.

Eight plants (four PWRs and four BWRs) f would find that the use of Revision 2 would raise their P-T limi*.s somewhere between 50*F and 100*F.

The features that made this happen were:

high nickel welds in three cases, and low fluences in five cases.

Some 33 plants would be ratcheted 20-50*F.

A total of 10 plants would be benefitted a significant Most of these have low-nickel material (A302 B plate and welds with amount.

only residual nickel content).

Most owners of plants undergoing licensing would find the use of Revision 2 raises their P-T limits 20-50*F, but the impact of this is small because flux reduction programs, the use of low-copper materials, and better information about initial RT will cause the expected end-of-life reference temperature NDT to be less than 200'F even when calculated using Revision 2.

It is important to note that the fluences used in the calculations on which Table I is based correspond to about one-fourth of plant life on the average.

Inspection of Figures 3 and 4 reveals that for fluences characteristic of later for PWRs, and for copper levels that 8

life, which range from 1 to 7 x 10" n/cm produce high-shift values, the ratchet effects of Revision 2 (as calculated for Table I) disappear and become benefits during the latter half of the lifetime.

4.a.1.2 Risk Avoided by using Revision 2.

One source of risk of I' *IIY continued operation with P-T limits based on Revision 1 when RfNDT higher (as given by Revision 2) depends upon the increased probability that a given transient will threaten the vessel.

To illustrate this situation in Figure 1, consider that the hazardous region is expanded to the right, closer based to the P-T limit, because its extent really depends on a value of RTHDT on Revision 2.

Thus the operator will be misled by erroneous P-T limits as to REG ANAL RG 1.99 REV 2 08/22/85 9

the potential severity of the transient and will have less time to take correc-tive action to avoid the hazardous region.

To quantify the risk that would be avoided if the P-T Ifmits were moved upscale to conform to Revision 2, several factors must be evaluated:

Expected frequency of transients that may violate P-T limits.

a.

b.

Expected severity of transients in terms of pressure and temperature as functions of time, Reduction in severity caused by having proper P-T limits and therefore c.

more concern on the part of the operator and more time to take corrective action.

This factor must be evaluated as a function of the difference in P-T Ifmits based on Revisions 1 and 2.

A quantitative evaluation of these factors was not undertaken, because it appeared that uncertainties, particularly in item c, would be so large that the result might not be defensible.

I Another source of risk is incurred by a heatup-cooldown operation following P-T limits that are lower than they should be.

Consider, for example, one of the plants for which the P-T limits based on Revision 1 are 100*F below those based on Revision 2.

If a plant in that situation continued to operate with a P-T limit based on Revision 1, and if the operator followed the limits closely in a heatup - cooldown sequence, there would be greater probability of fracture of the vessel. A contract was given to Pacific Northwest Laboratories (PNL) to evaluate the increased risk, calculate the pub 1(c exposure to radiation as a consequence of vessel failure, and calculate the costs resulting from a change to Revision 2 as the basis for P-f limits.

The PNL report is Enclosure 6.

The change in the probability of fracture (Revision 1 - Revision 2) was calculated using a Monte Carlo technique and the VISA code, which originated at the NRC and is being further developed at PNL.

In this analysis, initial NOT, toughness, copper content, fluence, and flaw size were treated proba-RT bilistically. Actually, in most runs, the flaw size had to be treated as 08/22/85 10 REG ANAL RG 1.99 REV 2

[

fixed at a large value (a 2 in, deep continuous flaw, probability of one) to get the Monte Carlo procedure to produce failures frequently enough to keep the required total runs to a reasonable number. Then the probability was reduced by a factor of 3500 to account for the more realistic flaw size distribution normally used, as explained in the PNL report.

The conclusion reached by PNL, based on the probabilistic fracture mechanics analysis, was that the best estimate of the increase in probability of vessel failure resulting f rom following P-T' limits that were 100*F too low was 2.5 E-7 per heatup/cooldown cycle.

In the subsequent risk analysis, this result was used for plants ratcheted 50-100*F and half that amount for plants ratcheted 20-50*F.

Using the techniques describ,ed in Enclosure 6, PNL completed the calculation of increased public risk resulting from these vessel failure probabilities.

For the lifetime of the present population of plants, the total man-rem avoided was 1.1 E+4 man-rem.

4.a.1.3 Costs to Industry--PNL Value/ Impact Analysis.

The cost of adopt-ing Revision 2 derives mainly from the cost of purchased power during delays in startup caused by the more restrictive operating limits.

In the PNL analysis,, the extra time involved was estimated to be 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the worst case of a 100*F ratchet, based on the opinions of their contacts among inspec-tors and utilities. When the P-T limit is moved upscale, the offect is to increase time spent at low pressure during heatup/cooldown, and the time was calculated by dividing 100*F by 50*F/hr, a typical heatup rate Ifmit.

The PNL estimate of industry operating cost is based on an average power cost of

$300,000 per EFPD for all plants.

Thus, for the 8 plants that are expected to be ratcheted 50-100*F an estimated delay of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per heatup, and 6 heatup/

cooldown cycles per year, the cost will be $150,000 per year per plant for the next 25 years.

The 77 plants, including those undergoing Ifcensing, that are expected to be ratcheted 20-50*F per year are assumed to incur half that amount.

The best estimate present value of these industry operating costs was given by PNL as $101,000,000 and the not cost to the industry was $63,000,000.

The bottom line, best estimate, cost per man-rem saved reported by PNL is "in the

$3,500-$5,600 range."

08/22/85 11 REG ANAL RG 1.99 REV 2

4.a.1.4 Costs to Industry - NRC Cost Analysis Group.

The NRC's Cost Analysis Group (CAG) made a number of substantive comments about the September, 1984 draft of the PNL report and also did a complete recalculation of the indus-try operating cost.

(See Enclosures 7 and 8).

In this work, CAG used their own plant specific power costs as well as plant specific values of the extra heatup time required as calculated by ME8R from the actual difference in P-T limits per Revision 2, relative to those per Revision 1.

Their industry operating cost figure was $71,000,000.

If this figure is substituted for the PNL figure of $101,000,000 in the PNL cost estimate, the not cost to industry (See Sec-tion 4 of Enclosure 6) of $33,000,000 instead of the PNL figure of $63.000,000.

(Further reduction of this cost figure may be justified as discussed in para-graph 4.a.1.6.)

This use of the CAG cost estimate reduces the bottom line to

$1.800-$2,900 per man-rem-saved.

l There is also a paper work cost to the industry when a Tech. Spec. change is prepared and submitted to the NRC.

PNL's best estimate of this is $2270 per plant--too small to enter into the cost benefit analysis.

4.a.1.5 Costs to Industry--Corrections Based on Comments Received Durina Concurrence.

The PNL estimate of 6 startup/ shutdowns per plant year was questioned by the staff.

It had been based on reactor scram data available to PNL.

However, after a scram, The plant does not necessarily go to cold shut-down.

In the plant operating data given in NUREG-0020, shutdowns greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> are tabulated for each month.

After correcting for the refueling shut-down, which extends over several months, the average over a three year period, 1982-4 was 4.3 shutdowns greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per reactor year.

From an ORNL analysis of plant operation for 1982, there were approximately 2.0 shutdowns per reactor year greater than 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />.

Based on these data, and assuming that some of the 72-hour shutdowns were not cold shutdowns, it seems reasonable to assume that there are no more than three startup/ shutdown operations per reactor year instead of six as assumed by PNL. This correction reduces the bottom line to $900-1400 per man-rem-saved.

Consideration has also been given to the estimate of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> extra heatup time, caused by the P-T limits being moved upscale by 100*F.

The PML estimate was based on simple division, assuming a hestup rate of 50'F per hour.

The 08/22/85 12 REG ANAL RG 1.99 REV 2 i

rationale was that any restriction to the operating zone (see Figure 1) would From data require more care in heatup/cooldown and therefore take more time.

f located by R. R. Riggs, SPEB, in EPRI NP-1139, Vol. 2 for PWRs, the average The startup time from cold shutdown to hot standby conditions is 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

average shutdown time, hot to cold, is 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.

If it takes 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> for an average startup/ shutdown operation (from cold shutdown to hot standby and back) it appears that 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is almost " lost in the noise."

There must be a number of operational f actors that determine the critical path, and it is doubtful that the restriction in the operating zone is normally a critical path item considering the number of testing and operational proce-dures involved in a startup from cold shutdown. Moreover, the ratchet effect of Revision 2 is at a maximum early in plant life when the operating zone has its greatest breadth.

In effect, adoption of Revision 2 will simply mean that anticipated restrictions in the operating zone will occur sooner than expected.

Consequently, among the list of uncertainties to be itemized later, there is the judgment that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> figure, and the corresponding cost to industry of

$25,000 per startup/ shutdown, is probably an upper bound, and the lower bound is nearly zero.

Conservatism in the PNL analysis that affects both cost and risk is the assumption that the ratchets shown in Table I are constant for the remainder of plant life.

For the majority of those ratcheted 50-100*F, this may be true, but for some plants the ratchet steadily decreases and becomes a benefit as early as mid-life of the plant.

However, the risk reductions also diminish.

It has been assumed that their ratio, the dollars per man-rem " bottom line" remains constant.

Clearly, there is uncertainty in the assumption that the risks and the present value of the costs are the same function of the ratchet, but it has been made for want of better information.

4.a.1.6 Boilina Water Reactors -- Special_ Considerations.

Bolling water reactors, four of which would be ratcheted 50-100*F according to Table 1, should probably be omitted from the calculation of both cost and risk associated with heatup/cooldown operations, because saturation conditions govern the pres-sure at any temperature even during transients if the vessel is not water solid.

This consideration would reduce the $33,000,000 industry cost estimate to about

$21,000,000, or about one-third of the original PNL cost estimate.

08/22/85 13 REG ANAL RG 1.99 REV 2

It is during hydrotests and leak tests that BWRs may be affected by Regula-l tory Guide 1.99.

The margins required for hydrotest are given by the ASME Code and Appendix G, 10 CFR 50 just as they are for normal operation, except the factor of safety on pressure is 1.5 for hydrotest instead of 2.0.

A number of BWRs are now at the level of radiation damage where the required metal tempera-ture at the hydrotest pressure is nearing 200*F.

This causes two problems, according to verbal reports from some BWP owners representatives.

Heatup with pump power apparently is slow in BWRs, compared to PWRs.

Also, prolonged opera-tion of the pumps at low pressure causes extra wear of the pump seals.

In any case, it takes longer to get to the higher temperature.

The other problem arises from a requirement in the Technical SpecIf fcations l

l that the containment drywell be closed when water temperature approaches 212*F.

Inerting is not required for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more, so entry can be made for leak j

test inspections, but the time required is increased. Yet, the hydrotest l

temperature for PWRs exceeds 200'F in many cases; hence, the problem of inspect-ing for leaks must be manageable.

These delays occur after refueling, normally about once every 18 months. Data from R. Riggs (EPRI NP 1136, Vol. 1) show the mean time to conduct a hydrotest of a BWR is 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />, with a standard deviation of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Thus the impact (time / cost) on BWRs likely falls within the stand-l ard deviation of the hydrotest time interval.

Risks incurred during hydrotest are probably small.

It is debatable whether or not the margin of 1.5 on pressure required by the Code, compared to a margin of 2.0 for normal operation, properly weighs the relative risks.

However, any change in these values involves an amendment to Appendix G, 10 CFR 50 and is beyond the scope of Regulatory Guide 1.99.

It is concluded that the BWR hydro-test situation represents little change in the direction of cost per man-rem-saved ($900-51400, as discussed in 4.a.1.5).

The effect is undoubtedly plant specific because the conclusion depends on operational details involved in conducting hydrotests and leak tests.

4.a.1.7 Effects on LTOP Limits.

After a number of low temperature overpressurizations occurred in the 1970's a requirement was placed on PWRs to l

provide automatic pressure relief at low temperatures that would prevent viola-tion of the pressure-temperature limits.

Some plants accomplish this by enabling 08/22/85 14 REG ANAL RG 1.99 REV 2 l

\\

L

a low setpoint setting on the pressure operated relief valves (PORVs) when at Others use the relief valves in the residual heat removal low temperatures.

8 system (RHR) to provide this function.

For PWRs that would have the P-T limits moved up the temperature scale if the limits were based on Revision 2, the allowable pressure at the cold shut-The amount h

down temperature, say 140*F for example, would be slightly reduced.

is small, because the P-T limit curve is flat at those temperatures, especially for vessels with a large radiation shif t.

Unfortunately, the system constraints g

I The reactor coolant pumps (RCP) used in Westinghouse plants require a pressure of about 325 psig minimum for proper function of the seals, f

are tight.

400-500 psig for Typical settings of the PORV for LTOP protection are temperatures up to about 350*F.

This restriction on the operating zone probably accounts for most of the impact on startup time.

In his memorandum to 8. D. Lisw. February 27, 1985 commenting on Revision 2 B. J. Elliot wrote:

Thus, to begin heatup with these limitations, the reac-tor vessel is pressurized to approximately 350 psig us-ing RHR and CVCS pumps prior to initiating flow through the RCPs.

The RCPs are the main source of heat prior to core critical operations.

At temperatures below 350*F, the reactor coolant pressure (RCP) must be maintained below the LTOP pressure set point of 400 psig, since actuation of the LTOP valves wnuld depressurize the The LTOP system and heatup would have to begin again.

set points for pressure and temperature are derived from the pressure-temperature limits.

Hence, a change in the pressure-temperature limit roquires an adjustment in the LTOP set-points.

The narrowing of the operating band that results from RCP minimum pressurize and the LTOP set point, as plants age, has been recognized as NS55 vendors a problem by the staff and NSSS vendors.

are reevaluating critical component linitations and operating procedures to determine whether the low tem-The staff, in perature operating band may be widened.

a letter from B. W. Sheron to W. Minners dated August 1, 1984, has requested that the LTOP criteria for operat-ing plants be prioritized as a new generic issue.

The last sentence in the quotation refers to another aspect of the LTOP At Maine Yankee, the requirements for LTOP protection have been issue.

RLG ANAL RG 1.99 REV 2 15 08/22/85

i r

4 i

broadened to mean there should be automatic protection for the P-T limits at j

l i

l high temperatures as well as low. Until recently, this was provided by the I'

normal high setpoint on the PORVs, because the temperature at which the RHR system was isolated was higher than the P-T limit temperature at 2250 psig.

i (See AE00 Engineering Evaluation Report No. E426, October 24, 1984, by E. V. Imbro).

E i Actually, Maine Yankee is not affected by Revision 2, because there are sufficient credible surveillance data. They show relatively large shift, g

however, and the P-T limits have been moved upscale so far that the operators

['

l elected to add a PORV setpoint at an intermediate pressure.

At present, the NRC does not require automatic protection of the P-T limits at the higher. temperatures.

If that becomes a requirement, Revision 2

'i may have some impact on this new aspect of LTOP requirements as well.

4.a.1.8 Costs to NRC. The inmediate impact of proposed changes on t

NRC staff review time will be minimal, because we plan to implement Revision 2 for P-T limit calculations at the regularly scheduled times, as described in l

However, the 3 year limit on implementation will crowd the calendar Section 6.

I l

for review of P-T limits somewhat.

4.a.1.9 Summary of Uncertainties and Conclusions re: Application of Revision 2 to P-T Limits. The conclusion reached by PNL - that the costs per man-rem were in the range $3500 5600 has been corrected based on more detailed l

power cost analysis by CAG (paragraph 4.a.1.4) and on data collected by the i

staff with regard to the number of cold shutdown per reactor year to yield a The following uncertainties have been identi-l value of $900-1400 per man-rem.

fled in this result the net effect of which is believed to reduce it signiff-cantly, but it has not been possible to further quantify the number, i

The most important uncertainty is in the risk ters.

In paragraph 4.a.1.2 and Figure 1, it was shown that the principal safety function of P-T limits is l

to separate the operating zone far enough from the zone that is hazardous to f

l vessel integrity to give time for corrective action to mitigate a transient.

REG ANAL RG 1.99 REV 2 16 08/22/85

Regulatory Guide 1.99, being part of the bases for calculating P-T limits, affects the margin provided.

For reasons given previously, it was not possible to quantify the risk avoided by a given correction to the P-T limits, but it is believed to be a larger source of risk than the one PNL was asked to evaluate.

The second most important uncertainty is in the cost term.

The principal cost to industry is the cost of power not generated because of delays in startup/ shutdown operations caused by restrictions in the operating zone if l

the P-T limits are moved upscale.

The PNL estimate of two hours per startup 15 believed to be high for most plants, as discussa i in paragraph 4.a.1.5.

The narrowing of the operating zone resulting from adoption of Revision 2 in certain plants is not significant early in life when the operating zone is Moreover, it is no more severe than was expected to occur late in life.

brcad.

Finally, at higher fluences the " ratchet" disappears, because of dif ferences in the fluence functions given in Revisions 1 and 2.

A11 of these uncertainties are in the direction of significantly reducing the PNL cost estimate.

4.a.2 Other Applications of Regulatory Guide 1.99 must be calculated as part of Regulatory Guide 1.99 is used whenever RTHDT an analysis of a transient that has actually occurred.

The analysis provides the basis for deciding if the possibility that the vessel has been damaged is sufficiently high to warrant an inspection before returning it to service.

Such inspections are time consuming and costly in terms of power replacement.

Another application of Regulatory Guide 1.99 is in the analysis of flaws A recent example found by inservice inspection of the reactor vessel beltline.

occurred at Indian Point 2.

First, the draft Revision 2 was used to calculate at the inside surface, based on the fluence and the weld and plate RTNDT Second, because the flaw was near the outside surface, the formula chemistry.

through the vessel wall, (a feature of Revision 2),

for attenuation of ARTNOT at the tip of the flaw.

The evaluation is the was used to calculate ARTNOT basis for deciding if the vessel nust be repaired before being put back in service.

REG ANAL RG 1.99 REV 2 17 08/22/85

In evaluations of transients and flaws there may be significant safety questions the cost impact may be high, and contention over the decisions may surface. Clearly, it is important to have a basis for the calculation of radiation damage effects that has had public review and resolution of out-I standing issues.

l 4.b.

Impact On Other Requirements l

4.b.1 Impact on the Pressurized Thermal Shock Rule In the PTS rule, FEDERAL REGISTER, July 23, 1985 there is a formula for t distinguish that method from others), which calculating RTNDT (called RTPTS is based on an early version of the formulas that are the basis for Revision 2.

(1) weld and base l

There are several differentes.

For the calculation of RTPTS.

metal data were analyzed as one data base yielding one correlation function, l

(2) the fluence function was of simpler form, and (3) there was a second equa-tion which gave bounding values.

It is not intended to change the proposed PTS Rule to incorporate the latest formula for RT (Note, however, the action recommended below.) The NDT.

calculation of RT required by the PTS rule is associated with screening cri-NDT teria (270*F for base metal and axial welds, 300'F for circumferential welds),

i which were justified by a probabilistic analysis that considered all identifi-At the time able uncertainties including those in the calculation of RTNOT.

the Rule was drafted, it was recognized that there would be an evolution in the calculative procedures, but there would be too many actions taken by the utili-ties in the area of flux reduction programs and other measures to permit fre-quent changes in the position of every plant relative to the screening criteria.

However, when it appears that the screening criteria will be exceeded at a specific plant, the PTS Rule requires an evaluation of all aspects of the PTS analysis as they apply to that plant.

That re-evaluation must include the will re-evalua w fracture toughness of the beltline material and thus RTHDT By that time, a number of plants will have credible surveillance data of too.

l their own.

If not, and if Revision 2 of Regulatory Guide 1.99 were used RTNDf values would be higher for a few vessels having high-nickel welds but most would be slightly lower.

08/22/85 18 REG ANAL RG 1.99 REV 2

H the calculative procedures of Revision 2 were substituted for those of the PTS Rule, and M the present information about copper, nickel, and fluence values is confirmed by the submittals required by paragraph (b) of Section 50.61 of the PTS Rule, it appears that eight plants would reach the screening cri-terion before EOL. Of the eight plants, four would reach the screening crf-terion in the years 1993 to 1996 and four in the year 2000 or later.

Based on these findings, it is recommended that the copies of proposed Revision 2 sent to utilities and others be accorr.panied by a letter containing the following I

statement of the relationship of the proposed Guide to the PTS rule.

The calculative procedures given in paragraph C.I.a of this draft Guide are not the same as those given in the Pressurized Thermal Shock rule

• for calculating RTPTS, the reference temperature that is to be compared to the screening criterion given in the rule.

Issuance of Regulatory Guide 1.99, Revision 2 for public comment in no way affects the recently promulgated PTS rule.

Licensees and the technical community are requested to comment on the technical merits of this proposal, including its effect on their plants for non - PTS purposes, chiefly as the basis for calcula-tion of pressure-temperature Ifmits as required by Appendix G, 10 CFR Licensees may also consider and comment on the prcposed change's Part 50.

effect on the calcuisted PTS risk at their plant, assuming the Revision 2 correlation, if justified, would at some future time replace the RTPTS correlation in the PTS rule.

Following resolution of comments, and once late RT general agreement is reached regarding the best way to cair NDT' then it will be appropriate to re-evaluate the overall conservatism of the PTS rule.

4.b.2 Impact on Material Selection For plants in the very early construction phase before the reactor vessel materials have been ordered, for which the provisions of Position C.3 of the may have some Guide are applicable, the new procedure for calculating RTHDT effect on the limit specified for copper content.

(Position C.3 itself is

• " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register, July 23, 1985, pp. 29937-29945.

08/22/85 19 REG ANAL RG 1.99 REV 2

l..

unchanged from Revision 1).

However, sample calculations indicate that there should not be many instances of a negative effect, because they are made for end-of-life fluences.

For plants for which the surveillance materials have not been selected, the changes from Revision 1 to Revision 2 may affect the decision as to which l

beltline materials will be controlling and therefore which weld and base metal should be represented in the surveillance capsules. However, because we are not changing this requirement, there should not be any extra cost involved.

I 4.c.

Constraints I

We have not identified any constraints such as scheduling or enforceabil-ity that affect the implementation of Revision 2.

5.

DECISION RATIONALE Based on the foregoing analyses of the safety issues, system impacts and costs, it is recommended that Revision 2 be issued for public comment.

The analysis has shown that the Guide is needed, because it provides part of the l

basis for ensuring safe operation of reactors during startup and shutdown and for the evaluation of transients and flaws found in service.

Periodic updating t

of the Guide is consistent with the requirements of Appendix G, 10 CFR Part 50.

Adoption of Revision 2 will raise the P-T limits for about half of the operating reactors that now use Revision 1 as the basis for these limits.

In preparing the value/ impact analysis, based on contract work by PNL, one source of avoided risk by going to Revision 2 was quantified. After making corrections based on staff comments, the cost-benefit ratio was in the range 8900-1400 per man-rem avoided. However, there is another, more significant source of risk l

which was not quantifiable. As discussed in Section 4.a.1.2, the principal safety impact of operating with lower P-T limits (i.e., continuing to use limits i

based on Revision 1) will occur during a transient, because the operator (a) will not have accurate information of the potential hazard to vessel integrity, and (b) will have less time to take corrective action.

In paragraphs 4.a.1.4 and 4.a.1.5, reasons have been given to believe that the amount of replacement power, 08/22/85 20 REG ANAL RG 1.99 REV 2 i

c I

which is the principal cost impact of implementing Revision 2, will be lower for In summary, most plants than the values determined in the PNL cost estimates.

'l it is believed that the cost benefit ratio is in the range of a few hundred Because there is a significaat safety benefit, and dollars per man-rem avoided.

because it is cost effective, the implementation of Revision 2 is recommended.

t I

Finally, we believe tist industry will be receptive to the proposed " trend curves," as they are commonly called. Based on industry response to presenta-l tions to ASTM Committee E-10, the Metal Properties Council, and ASME Soller f'

Code Section XI working groups concerned with this subject, there appears to I

At this time we know of no be general agreement on the need for new curves.

There may serious objection to the calculative procedures given in Revision 2.

be disagreements about the chemistry factor or fluence factor near the edges of the data base, where extrapolation is required, and about the margin to be i

We intend to push for adoption in ASTM Standard Guide E900 as well as added.

in Section XI of the ASME Boiler Code, but this will take at least a year or If implementation of Revision 2 is delayed, there will most likely be longer.

more negative ballots by those who expect to be impacted by the new trend curves.

There will continue to be a need for Regulatory Guide 1.99 to provide acceptable treatment of the question of margin, the treatment of plant specific survell-lance data, and the calculation of attenuation of damage through the vessel wall.

In Revision 2, Further revision of Regulatory Guide 1.99 is to be espected.

Position C.1.b. and Figure 2, which presents a " trend curve" for the percent decrease in upper shelf energy as a function of fluence and copper content, That needs to be upgraded when the basis becomes available in a year or so.

effort will not affect P-T limits and it will affect only relatively few l

further adjustment of plants--those having low upper shelf energy initially, the calculative p.rocedure for reference temperature may be indicated in a few years when more survet11ance data have accumulated or when our understanding of embrittlement mchanisms and the role of copper, nickel and other elements improves.

l REG ANAL RG 1.99 REV 2 l

21 08/22/85

6.

IMPLEMENTATION Paragraph D.1. of the Implementation Section of Revistun 2 reads much as it did for Revision 1, the key sentence being:

...the methods described in regulatory position C.1 and C.2 will be used in evaluating all predictions of radiation damage needed to implement General Design Criterion 31 or as called for in Appendices G and H to 10 CFR Part 50 submitted after the effective date of publication of the Guide." This means that plants will be allowed to con-tinue to follow the present schedule for updating their P-T limits, but only within a 3 year period.

Paragraph D.2. requires all operating plants to review the limits within a 3 year period after the Guide becomes effective and to revise them if necessary.

The decision to forego prompt implementation across the board is based on the existence of significant margins, as discussed in Sectior. 2.a.1 and elsewhere in this analysis.

In the staff's judgment, the risk of allowing some plants to operate three more years with present P-T limits was justified by the reduced impact on the industry and the NRC achieved by not requiring a complete review of all P-T limits in a 6-12 month ;arlod.

No staff actions will be required to implement the Guide other than making sure that each facility gets a copy.

7.

OTHER IMPACTS There are no other actions, systems or prior analyses known to need reassessment as a result of publication of Revision 2.

~

This Guide does not add to the reporting or information collection require-ments of licensees, nur does it affect small entities as defined in the Regula-tory Flexibility Act.

08/22/85 22 REG ANAL RG 1.99 REV 2 i

k i

v.

i TABLE 1.

SUIWARY OF THE CHANGES IN PRESSURE-TEMPERATURE I

LIMITS EXPECTE0 TO RESULT FROM A CHANGE FROM REVISION 1 TO REVISION 2 0F REGULATORY GUIDE 1.99 Effect of Change Operating Reactors Plants Undernoine Licensing from Rev. I to Rev. 2 PWR SWR Total PWR OWR Total

,l l

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4*

4 8

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0 3

i tenefit 20-50' 7

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l Benefit 50-100' 1

0 1

Benefit 100-150' 1

0 1

35 12 47 Totals 52 29 81

' Values in the table are number of plants.

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PROPOSED REVISION 2 OF REGULATORY GUIDE 1.99 VALUE-lMPACT ANALYSIS 9

All S. Tabatabal William B. Andrews November, 1984 Prepared for the U.S. Nuclear Regulatory Cawnission 4

Energy Systems Department Pactfic Northwest Laboratory Richland, Washington a

l i

')

.J Y.

CONTENTS 1.0 I NTRODU CT I ON..........................

1.1 2.0 PROPOSED ACTION AND POTENTI AL ALTERNATIVES...........

2.1 3.0 AFFECTED DECISION FACTORS...................

3.1 J

4.0 VALUE-lWACT ASSESSENT SUM 4ARY.................

4.1 5.0 UNQUANTIFIED RESIDUAL ASSESSENT................

5.1 6.0 DEVELOPENT OF QUANTIFICATIONS.................

6.1 PUB L I C H E ALTH..........................

6.1 OCQJPATIONAL EXPOSURE ( ACCIDENTAL)

..............6.7' PUBLIC PROPERTY...................'.....6.9 ONSITE PROPERTY........................6.9 INDUSTRY IWLEENTATION....................6.10 INDUSTRY OPERATION (X)ST....................6.11 NRC IWLEENTATION COSTS....................

6.12 7.0 CON CL US I ON S...........................

7.1 REFEREN E S............................... R.1 I ~.

a s

h a

iii i

-n

FIGURES 1.

Heat-up P-T Limits Based 'on Rev.1 Guidel ines..........

6.2 Heat-up P-T Limits Based on Rev. 2 Guidelines..........

6.2 2.

3.

Cool-down P-T Limits Based on Rev.1 Guidelines 6.3 4.

Cool-down P-T Limits Based on Rev. 2 Guidelines

....6.3 TABLES 1.

Summary of Changes in P-T Limits Resulting from a Changt from Rev. I to Rev. 2 of Regulatory Guide 1.99.

1.1 2.

V/ I Ana l y s i s Dec i s i on Fact or s................... 3.1 3.

Ef fects on Change in Failure Probability per Transient from Revision 1 to Revision 2................... 6.4 4.

Changes to Release Category Frequency (per event) 6.5 5.

Dose Conversion Factors (man-rem / release).

6.6 6.

Avolded Calculated Public Dose (man-ren/startup-shutdown).

6.6 7.

Summary of Avoided Public Health Risk

.............6.7 8.

Summary of Avolded Occupational Exposure............. 6.8 9.

Summary of Avolded Public (Offsite) Property Damage........ 6.9

10. Summary of Avolded Onsite Property Damage............. 6.10 II i

0 iv

e PROPOSED REVISION 2 0F REGULATORY GUIDE 1.99 VALUE-lMPACT ANALYSIS

1.0 INTRODUCTION

i This report presents a value-impact assessment of implementing Revision 2 of Regulatory Guide 1.99, " Radiation Damage to Reactor Vessel Materials." in practice, neutron radiation damage to the reactor vessel is compensated f or by shif ting the Pressure-Temperature (P-T) limits up the tanperature scale every few years by an enount corresponding to the shif t In the Charpy test transition i

tenperature produced by the accumulated neutron fluence. The NRC regulates this process on the basis of Appendices G and H,10 CFR Part 50. Paragraph V. A of Appendix G requires:

"The ef fects of neutron radiation... are to be predicted from the results of pertinent radiation ef fects studies...". Regulatory Guide 1.99 provides a means f or detennining results in the form of calculational procedur3s that are acceptable to NRC, and it describes acceptable procedures for using plant-specific surveillance data when they become available.

The objectives of Regulatory Guide 1.99 are best described in terms of the transition tenperature approach to fracture prevention.

In this approach, the margin of safety against fracture is given as the dif ference between the operating temperature of the vessel and the temperature *at which brittle fracture could occur, and the measure of radiation damage is a transition temperature shif t.

Thus, the ability to calculate the shif t has a direct impact on the margin of saf ety against f racture. The objective of Revision 2 of Regulatory Guide 1.99 Is tc make the calculational procedures consistent with the present knowledge of radiation damage. Table 1 presents a summary of the limits changes that would result from the changes in Pressure-Temperature (P-T) that are proposed in Revision 2 of Regulatory Guide 1.99.

The values in this table were obtained from the NRC staff (P. N. Randall, 1984).

Summary of Changes in P-T Limits Resulting f rom a Change from TABLE 1.

Rev. I to Rev. 2 of Regulatory Guide 1.99 Number of Effect of Change Number of Plants Undergoing From Rev. I to Rev. 2 Operating Reactors Lfcensing DiB BMS Tota l EWB HB TotaI t.

Ratchet 50'F to 100*F 4

4 8

Ratchet 20'F to 50*F 16 17 33 34 11' 45 No Change (+20*F )

23 7

30 3

to 13 Benefit 20*F to 50*F 7

1' 8

Benef it 50*F to 100'F 1

0 1

Benefit 100*F to 150'F 1

0 1

TOTALS 52 29 81 37 21 58 (a) As used herein, ratcheting means an increase in P-T limits up the tanperature scale.

(b) A: used herein, benefit means a decrease in P-T limits down the tenperature scale.

1.1

The following sections present the results of a value-impact analysis done on the of f acts of implementing Revision 2 of Regulatory Guide 1.99.

The scope of this analysis is limited to analyzing the changes' In f ailure probability of the reactor vessel during normal startup-shutdown procedures from Revision 1 to Revision 2.

The of f acts on the operational transients and the pressurized thermal shock issue (PTS) have not been addressed in this study.

The value-impact assessment uses the methods developed in The Handbook f or Value-impact Assessment (Heaberlin et. al.1983), the data developed for Safety lssue Priortilzation (Andrews et. al.1983) and the results of calculations done using a NRC Vessel Integrity Simulation Analysis (VISA) code (D.L. Stevens et.

- al.

1 983).

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2.0 PROPOSED ACTION AND POTENTIAL ALTERNATIVES It has been proposed that Revision 2 of Regulatory Guide 1.99 be implemented by operating plants as wel1 as those undergoing licensing review.

This proposal is based on considerations of public risk, occupational dose and cost impacts.

Improvements in knowledge about material properties warrant the use of the guidelines given in Revision 2 of Regulatory Guide 1.99 in determining the Nil Ductility Transition Reference Temperture (RT of the majority $kT)he operating nuclear Use of Revision 2 guidelines will ef fect the RT t

[ censing. The change in RT "III reactors and those undergoing subsequently change the P-T limits set in the Technical Specifications of the ND nuclear power plants. This change and the number of plants af fected were presented in Table 1.

The alternatives to implementation of Revision 2 are to leave Revision 1 in place or to eliminate the Guide all together. The solution is not feasible since the NRC staf f reviews more than 10 Pressure-Temperature limits per year and needs a published basis for its review.

Currently, there is no ASME code equivalent to Regulatory Guide 1.99.

ASTM Standard Guide E-900-83 contains an equation relating the Charpy shif t to copper content and fluence, but It is out of date. Also, the ASTM Standard does not contain guidance on the use of plant-specific surveillance data. Revison 1 of Regulatory Guide 1.99 was used extensively from the time it was published in 1977 until late 1982 when the review of radiation dmage resulting from the pressurized thermal shock (PTS)

Issue revealed the need f or change.

Also, the surveillance data base has increased to the point where Revision 2 is based almost entirely on surveillance data, whereas Revision 1 was based primarily on test reactor data.

\\.

2,1

~

3.0 AFFECTED DECISION FACTORS Parameters considered in the value/ impact analysis and those af fected in this study are shown in Table 2.

TABLE 2.

V/I Analysis Decision Factors Causes Causes Quantified Unquantif iega)No Decision Factors Change Change Change Public Health X

~0ccupational Exposure X

(Accidental)

X Occupational Exposure (Routine)

Public Property X

Onsite Property X

X Regulatory. f ficiency E

X Improvementh in Knowledge Industry fr>Iementation C X

i industry Operation Cost X

k.

X l

NRC Development Cost NRC Implementation Cost X

s X

NRC Operation Cost 2

(a Unquantified means not readily estimated in dollars.

3.1 I

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4.0 VALUE-IMPACT ASSESSMENT

SUMMARY

Best Lower Upper Decision Factors Estimate Estimate Estimate VALUE.S (a) (man-rem)

PubiIc Hoaith 1.1 E+ 04 0

2.2E+04 f

Occupational Exposure (Accidental) 6.7E+01 0

2.2E+02 Occupational Exposure N/A l

(Routine)

Regulatory EtfIclency N/A leprovements In Knowledge N/A 0

Total Quantitled Value IWACTS 5) 1.1 E+04 2.2E+ 04 Industry implementation

-2.0+5

-1.2 E+5

-4.0E+5 Costs Industry Operating Cost

-1.01 E+ 08

-5.01 E+ 07

-1.7 E+ 08 NRC Development Cost N/A NRC Implementation Cost

-1.3E+05

-8.6E+ 04

-3.4E+ 05 NRC Operation Cost 0

0 0

O 3.2E+ 07 PubtIc Property 3.1 E+ 06 g; s

Onsite Property 3.5E+07 0

6.7E+ 07 Total QuantIfled Impact

-6.3 E+ 07,

-5.0E+ 07

-7.2E+07 (a) A decision term is a value if it supports NRC goals. Principal among these goals is the regulation of safety.

Impacts are defined as the costs Incurred as a result of the (b) 3:;r impacts indicate cost savings proposed action. :IE87tva (avoided cost).

(c) Assuming 5% discount rate.

N/A = Not Af f acted 4.1

5.0 t]NNIANTIF1ED. REElDtlAL.ARREESMEMT There are no unquantified decision f actors in the assessment of this 4

3 action.

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6.0 DEVELOPMENT OF'00ANTIFICATIONS Development of quantifications included the following f actors: public health, occupational exposure (accidental), public property, onsite property, industry implementation, industry operation cost and NRC implementation cost.

PUBL IC HEALTH A risk analysis was performed to assess the ef fects of implementing Revision 2 of Regulatory Guide 1.99.

It can be seen from Table 1 that implementation of this revision will change the P-T limits of a majority of the operating nuclear reactors and those plants undergoing licensing. The effects of this revision on those plants that are ratcheted by O'F to 100*F were '

analyzed.

The VISA code was used to develop estimates of the f ailure probabilities of reactor vessels. The P-T limits of a representative case were developed based on Revision 1 and Revision 2 values of RT

, following the procedures prescribed in Appendix G to Section til oY kSNE code and Appendix G of 10 CFR

50. Figures 1 through 4 show the heatup-cooldown P-T limit curves developed f or our assumed case and based on Revision 1 and Revision 2 guidelines.

A conservative estimate of the f ailure probability was determined by choosing a plant that would be ratcheted by 100' F If Revision 2 was to be implemented.

Following are the characteristics of the assumed representative case:

CU$ = 0.20 NI% = 1.00 F = 1E+19 (F = fluence)

An estimate of 4.9E-07 per transient f or the change in f ailure probability of the reactor vessel from Revision 1 to Revision 2 was obtained.

This was assumed to be an upper estimate f or this analysis.

Most of the ratcheted plants are those whose P-T limits are sWfected by 20'F to 50*F.

An Intermediate estimate was detennined by considering a plant that would be ratcheted by about 50'F.

The change in failure probability of the reactor vessel from Revision 1 to Revision 2 is determined to be zero. However, for our analysis, a best estimate of half the upper estimate (2.5E-07 per transient) was assumed.

(,

The lower estimate of the change in f ailure probability of the vessel was assumed to be zero, since there are about 30 plants that will see no change or improvement In their P-T limits due to implementation of Revision 2.

Modifications were also made to VISA in order to obtain estimates of the reactor vessel f ailure probability. The main modification was to revise the flaw size distribution given in VISA. A listing of the flaw size distribution follows:

6.1

I 3000i.

P 200o -....._.._.v.

m 6

m Q.

I 48 1000 b

C 000 TEMPEl'00nTURE F O

100 HEAT-UP LIMITS, REV. 1 FIGURE 1.

Heat-up P-T Limits Based on Rev.1 Guidelines 3000:

~

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2000 A

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Heat-up P-T Limits Based on Rev. 2 Guidelines 6

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3000 pt :

a

--~~~.-.g"---:-~~~

2000 - - - - -

m l

m c

i g o o 0 [

.f1.

W u'

g.c W

0 O

200 400 600 TEt1PERNTURE F y

COOL-DOWN LIMITS, REV. 1 J

I FIGURE 3.

Cool-down P-T Limits Based on Rev.1 Guidelines 0000 c

/. _ _

2030.

l c

G

<p w

}QQQ

/

O W.

O 200 400 600 TCf1Py hTUkE F COOL-DOWN LIMITS, REV. 2 FIGURE 4.

Cool-down P-T Limits Based on Rev. 2 Guidelines e

a "4

4 ee 6.3

s t

Flaw Size IJnches)

Probability 0.000 0.91767661 0.125 0.05507015 0.250 0.02256957 0.500 0.00422063 1.000 0.00036718 1.500 0.00007085 2.000 0.00001667 2.500 0.00000500 3.000 0.00000250 3.500 0.00000083 This distribution was used during the preliminary stages of our analysis.

However, due to the small probabilities given in the distribution no reactor vessel failure was simulated by VISA. Theref ore, to get any kind of an estimate, we assumed that the probability of a 1/4 T (Thickness) flew is 1. For our purposes the thickness of the reactor vessel was assumed to be 8 inches. The f ailure probability results were then reduced by a f actor of '3500 based on the frequency of this flew size. This value was obtained by the ratio of the results given by two VISA runst the first with the adjusted fles size distribution and the second with the original flow size distribution. These values were then multiplied by 6 to account f or 6 welds in the reactor vessel belt line.

Table 3 shows the ef fects of change In f ailure probability of vessels from Revision 1 to Revision 2.

The values In Table 3 should be Interpreted as the

" reduced" f ailure probability or the f ailure probability that will be " avoided" by implementing Revision 2 of Regulatory Guide 1.99. These estimates are very conservative due to the conservative nature of the analysis and the assumptions.

TABLE 3.

Ef fects of Change in Failure Probability per Transient from Revision I to Revision 2 Best Estimate Unner Bound tomar Bound 2.5E-07 4.9E-07 0

A conservative estimate of risk was made by assuming that a propagating crack l

would result in breach of the reactor pressure vessel and subsequent core melt.

This conservative assumption was used to perf orm the Pressurized Thermal Shock Value-impact analysis ( Andrews et. al.

1983) and it is believed that core-melt following a PTS event is more probable than during normal heatup-cooldown procedures, theref ore the subsequent risk analysis done for this report Is believed conservative. The containment failure modes, likelihoods, and release categories are assumed to be the same as for sequences S;D (f or PWRs) and SI (for.BWRs) Appendix A of the Guldallnat (Andrews 1983).

Table 4 shows the changes to release category frequency based on the previously specified release categories.

sp+- %

6.4

-i l

TABLE 4.

Changes to Release Category frequency (perevent)#

+

'3 Total Release Category Core Melt (b)

Iteactor Type Estimate 1

2 3

4 5

6 7

4 l

PWR Best Estimate 2.5E-09 "

0

_9.8E-08 0

, 3.6E-09 0

3.9E-07 '. 5.0E-07 ~

5.0E-03 1.8E-09 2.0E-07 2.5E-07 Upper Bound 4.9E-09 Lcuer Bound 0

0 0

0 0

0 0

0 DWR Best Estimate 2.5E-09 2.5E-07 0

O 2.5E-07 ICI 4.98E-07 Upper Bound

~ 4.9E-09 ' 4.9E-07 '

O O

N/A Lower Bound 0

0 0

0 0

(D b,

ta) Event refers to sTartup-shutdown.

(b) The values in table are per (startup-shutdown). For our purposes 6 (startup-shutdown)/ reactor is assumed.

(c) II/A = not affected 9

e 0

Y m

s Application of the dose conversion f actors in Table 5 to the changes in reisese category frequencies given in Table 4 results in the evolde shown in Table 6.

extremes (e.g., high estimate dose conversion times upper bound release category frequency change).

Dose Conversion Factors (man-rem / release)I*I TARtE 5.

Whole Rodv Data canamananca (--s-ra.)

Relames Categorv 5.4E+06 m

PWR 1 4.8E+06 PWR 2 5.4E+06 PWR 3 2.7 E+06 PWR 4 If 1.0E+06 PWR 5 1.5E+05 PWR 6 e

2.3E+05 PWR 7 5.4E+06 BWR 1 7.1E+06 BWR 2 5.1E+06 BWR 3 6.1E+05 BWR 4 (a) From ORAC, with guidelines and quantitles of radioactive isotopes used in WASH-1400. Estimates are based on the meteorology of a typical Midwest site (Byron-Braldwood) with a uniform population density of 340 people / square mile, no evacuation and 50-mile radius model.

Avolded Calculated Public Dose (men-re/startup-shutdown)

TARIE 6.

Raut Estimata unn.c anund in e n,mnd Reactor PWR 3.3E-01 6.5E-01 0

BWR 1.8E+00 3.5E+00 0

To estimate the total public risk averted, the per-startup-shutdown estimates must be multiplied by the number of offacted facilities The number of ef fected f acilities was assumed to be 86, given In The average number of startup-shutdown for a plant is assumed to be 6 lifetines.

The average remaining lifetime for the plants is assumed to be 25 Table 1.

Use of these values yloids the total evolded public dose estimates shown Per year.

year s.

In Table 7.

S 4

6.6

._.~

TABLE 7.

Sunmary of Avolded blic Health Risk (a,b)

Total Avolded Dosa (Person-ram)

Best Estimate Unnar Estimate Lo=ar EstImata 1.1E+04 2.2E+04 0

(a) For alI the af f acted PWRs and BWRs.

(b) These estimates include both the operating plants and those undergoing licensing. Refer to Table I for exact number of affected plants.

OCCUPATIONAL EXPOEURE (ACCIDENTAL)

The evolded occupational exposure from accidents can be estimated as the product of the change in total core-melt frequency and the occupational expos'ure likely to occur In the event of a major accident. The estimated change In core-melt probabilities ars presented in Table 4.

The occupational exposure in the event of a major accident has two cceponents. The first is the "Ismediate" exposure to the personnel on site during the span of the event and its short-term control. The second Is the longer-term exposure associated with the cleanup and recovery from the accident.

The final data required are the number of affected facilities and their remaining lif etimes. The number of plants af f acted is given in Table 1 and their average remaining lifetimes Is assumed to be 25 years. The total avoided occupational exposure is then calculated as follows:

UTOA " OOA = AF (DIO + DLTOI where DTOA = total evolded occupational dose N = number of af facted f acilities T = average remaining lifetime DOA = avoided occupational dose per reactor-year 4F = change in core-melt probabjilty DIO = "Immediate" occupational dose DLTO = long-term occupational dose

' Table 8 shows the values taken as best estimates and bounds for these parameters.

Uncertainties are conservatively propagated by use of extremes (e.g., high estimate D10 + high estimate DLTO '

I 6.7

_ ~ -. -

TABLE 8.

Summary of Avolded Occupational Exposure (a)

I I

Total Avolded Change in Core-Melt lamediate(b) long-Term (c)

Occupational Probability Occupational Dose Occupational Dose Exposure (events / reactor-yr)

(man-rem / event)

(man-rem / event)

(man-rem)

PWR BWR Best Estimate 2.5E-07 2.5E-07 1.0E+ 03 2.0E+ 04 6.7E+01

?

~

Upper Estimate 5.0E-07 4.50E-07 4.2E+ 03 3.0E+04 _

2.2E+02

~

Lower Estimate 0

0 0

1.0E+ 04 0

Forboththeophatingplantsandthoseundergoinglicensing.

(a)

For the exact number of plants affacted refer to Table 1.

(b) Based on Initial (4 month) occupational exposure following the accident at TMI.

l (c) Based on cleanup and decommissioning estimates.

MJREG/CR-2601 (Murphy.1982).

S

% =

s.

PUBLIC PROPERTY The ef fect of the proposed action upon reducing risk to public (i.e.,

of fsite) property is calculated by multiplying the change in accident probability by a generic of f site property damage estimate. This estimate was derived f rom the mean value of results of CRAC2 calculations, assuming an SSTI release (i.e., major accident) for 154 reactors (Strip 1982). The damage estimate is converted to present value by discounting at 105.

The following discounting f ormula was used:

i 4

-0.f ot _,.10tf g

D=V e

0.10 where D = discount value V = damage estimate t, = years before reactor begins operating 0 for operating reactor t, = years remaining until end of life.

The average number of years of renalning life is 25. Theref ore, the discount D/V = 9.

This must be multiplied by the number of affected facilities.to yield the total ef f ect of the action. Table 9 sunmarizes these results. The high and 4he low estimates are values f or Indian Point No. 2 and Palo Verde No. 3 calculated f rom Strip (1982).

The oost estimates have also been calculated using a 5% discount rate.

This was done as a sensitivity analysis to determine the impact of discount rate on the overall value-impact ratlo.

TABLE 9.

Stenmary of Avolded Public (Of fsite) Property Damage Offsite Value of Avolded Pronerty D-- ga (1/avant)

Offstta Pranarty D -- ga (1) 101 51 Best Estimate 1.7E+09 1,

2.0E+06 3.lE+06 Upper Bound 9.2E+09 2.1 E+07 3.2E+07 i

Lower Bound 8.3E+08 0

0 t

ONSITE PROPERTY The ef f ect of the proposed action on reducing' the risk to onsite property is estimated by multiplying the change in accident probability by a discounted onsite property cost. This discounted property cost was developed f rom the generic onsite property cost taken from Andrews et. al. (1983).

It includes an estimate f or replacement power.

6.9

s This valua is disc::untcd st 10% using tho following form.le:

~

(,,,-0.10(t,-t,))

l,,,-0.10M )

/

/

r,-0.10, I

D

=V l

I

( 0.01M /

\\

/

\\

/

where D = discounted value V = damage estimate g = years before reactor begins operation, O for operating t

reactors tg = years remaining until end of life M = period of time over which damage cost is paid out (recovery period in years)

Assuming that the remaining reactor Ilfe is 25 years and that the recovery period is 10 years, the discount D/V = 5.8.

To obtain the total effect of the actions, the per-reactor results are multipIIed by the number of af fected f acilities (86). The results are summarized in Table.10. The uncertainty bounds given in the table reflect a

+505 spread in the generic property cost coupled with the bounds on core-melt probability. This was estimated to be Indicative of the uncertainty level.

TABl_E 10.

Summary of Avolded Onsite Property Damage Onsite Value of Avolded Pronertv Dammee f1/ event)

OnsIte Pronerty Damaon (11 101 51 Best Estimate 1.65E+09 \\

1.2E+06 3.5E+07 Upper Bound 2.5E+ 09 '

3.6E+06 6.7 E+07 Lower Bound 8.2E+08 0

0 INDUSTRY lMPLEMENTATlON The primary cost element associated hith impler..enting Revision 2 of Regulatory Guide 1.99 consists of changing the P-T limits given in a plant's technical specification. The time required to make this change is estimated to be one man-week per plant. This cost is, taken to be $2270/ week (Andrews et. al.

1983). The number of af fected plants is assumed to be 86 (from Table 1).

This includes both the ope.ating plants and those undergoing licensing. Therefore, the total cost of industry implementation is estimated to be:

(86 plants) ($2270/ plant) = $2.0E+05 6.10

This value is taken to be our best estimate. For our purposes an upper estimate of 2 weeks and a lower estimate of 3 days have been assumed f or the time required to make the changes in the Technical Specifications.

l INDUSTRY OPERATION COST The main ef fect of Revision 2 of Regulatory Guide 1.99 Is to shif t the P-T limits up the temperature scale. This will require the plants to warm up the vessel to a higher temperature bef ore reaching to their normal operating conditions.

It is estimated that the maximum " lost time" f or the ratcheted plants would be 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> assuming 50'F/hr heat up rate. This estimate is derived by studying the revised P-T Ilmits and discussions with selected Industry representatives. This estimate is converted to a monetary value as follows:

(2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> lost time)(1 day /24 hours)($300,000/ day) = $2.5E+04 The $300,000/ day estimate represents the cost of one day delay in startup (full power operation) to the industry. Multiplying $2.5E+04/ plant by the number of af fected plants, number of assumed startup-shutdowns /yr, and the remaining lifetime of the plants (assumed to be 25 years) yields the total industry operation cost. Theref ore, the total Industry operation cost is:

(8 plants)($2.5E+4/ plant)(6 startup-shutdown /yr) (25 yr) = $3.0E+07 (1)

This value represents the industry operation costs for those plants that are ratcheted by 100' F. There are about 78 plants that are operating and undergoing licensing and that are ratcheted by up to 50' F. Their " lost time" will theref ore be half of the ones that are ratcheted by 100* F.

The industry operating cost for these plants is:

(2)

(78 plants)($1.25E+4/ plant)(6 startup-shutdown /yr)(25 yr) = $1.5E+08 Adding these values (i.e., $3.0E+07 and $1.5E+08) gives a total industry cost of $1.8E+08.

Assuming 5% discount over the next 2 years, the present value of total industry operation cost becomes $1.01E+08{. ' At 10% d!scount this estimate is

$6.54E+07.

Assuming a 100'F/hr heatup rate, can reduce this estimate by 50 percent.

An estimate of 30*F/hr heatup rate is chosen f or the upper bound cost calculations.

Some of the older plants (i.e., those that are ratcheted significantly),

ti.ere is concern that their P-T limits are already close to the saturation curve and implementation of Revision 2 will cause their P-T limits to get closer e

6.11

i to the saturation curve and thereby f urther limit their startup procedures.

Operations to cover heatup-cooldown rates may be needed to avoid this problem.

Additional costs would be incurred if this approach is taken.

p NRC lMPt_EENTATf ON COST The Impact of proposed changes with respect to staf f review time will be It will be limited to reviewing the revised P-T limits given in theIt is minimal.

Technical Specifications.

The cost review time would be required (estimated cost = $1500/ plant).For 86 plants, this per-plant might range from SICOO/ plant to $4000/ plant.

yields a total NRC Impact of $1.3E+05 with bonds ranging from $8.6E+04 tp

$3.4 E+05.

I t

?

e s

9 6.12

~

f

7.0 CONCLUSION

S The summary results f or the value-impact assessment are shown below.

l L==n ar v ef Yalue-lanact Antaasmant lanact*I (1)

A Value (man-rem)

Best Upper Lower Best Upper Lower Fatimata Estimate Estimate Estimate Ettfasta Estlanta sf 101 51 to 51

01 1.1E+04 2.2E+04 0

-6.3E+07 -6.2E+07 -7.2E+07 -7.7E+07

-5.0E+07 -3.2E+07 (a) Assuming a 5% and 10% discount rate.

The best estimates for cost and dose reductions Indicate that the cost of one man-ram avoided is in the $3500 - $5600 range. This high cost estimate, therefore, prevents us from recommending the implementation of this revision.

This conclusion is not sensitive to the assumed discount rate.

There is no doubt that the Revision 2 guidelines are more up-to-date and more accurate than Revision 1.

However, there are several drawbacks associated The main one is the fact that the implementation of with the proposed revision.

this revision will slow down the startup time of the majority of the operating plants and those undergoing licensing. The other drawback is that the P-T limits of some of the older plants is already close to the saturation curve.

Implementation of this revision will further limit their operational procedures.

t.

l 7.1

a

_ REFERENCES Guldattnes for Nuclear Power Plant safety NUREG/CR-2800 (PNL-4297),

Andrews, W. 8.

et. al.1983.

inf ormation Deveinnment.

issue Priorfttration Pacific Northwest Laboratory, Richland, Washington.

ASE code.

Appendix G to the Section ill of the REGULATORY GUfDE - Effects of Draf t Regulatory Guide 1.99, Revision 2,1984.

i l Residual Elements on Predicted Radiation Damage to Reactor Vessel United States Nuclear Regulatory Commission, Washington, D.C.

e Accac ema nf.

_A Handbnnk f or val ue-f =nact Heaberli n, S. W., et. al. 1983.PNL-4646, Pacific Northwest Laborato E

safatv and En=ts of

[

1982. Technologv.

Decnemtssioning Ref erence Light Water Reactors Following Postulated

Murphy, E., and G. Holter.

7-NUREG/CR-2601, Pacific Northwest Laboratory, Richand, t-Accidents.

Washington, 2EGULATORY GulDE - Ef fects of Regulatory Guide 1.99, Revisto,1,1977.

Residual Elements on Predicted Radiation Damag cnde for Predleting the YlSA - A namnuter NUREG/CR-3384, (PNL-4774),

Stevens, D.L. et. al.1983.of Remeter Pressure Vessel Faflure.

Probab i l i ty Pacific Northwest Laboratory, Richland, Washington Estimates of the Finanefal Risks of Nuclear Power Reac NUREG/CR-2723, Sandla National Laboratories, Albuquerque, Ne 1982.

Strip, D.R.

Accidents.

Mexico.

\\

4 I

e

fxe/c.wr 7

po s

_us.m W

REVIEW OF COST ANALYSIS CONTAINED IN VALUE IMPACT ANALYSIS FOR REVISION 2 0F REGULATORY GUIDE 1.99 OCTOBER 1984

+.sae n<>/ psje.

3repared by:

COST ANALYSIS GROUP OFFICE OF RESOURCE MANAGEMENT J.S. Nuclear Regulatory L.

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ADDENDUM TO ENCLOSURE 7:

" REVIEW OF COST ANALYSIS CONTAINED IN VALUE IMPACT ANALYSIS FOR REVISION 2 0F REGULATORY GUIDE 1.99" r

I This review provides comments on the September 1984 draft version of PNL's

" Proposed Revision 2 of Regulatory Guide 1.99 Value Impact Analysis." The reader should be alerted that subsequent revisions, as reflected in PNL's current version of the value impact analysis (Enclosure 6 of this regulatory package), have negated a number of the specific comments contained in this report.

However, issues we view as still being relevant are:

1.

PNL's failure to use reactor specific replacement energy costs, and 2.

PNL's failure to include savings in replacement energy costs from those plants having start-up and shut-down times which will be positively impacted.

( '.

9

REGULATORY CUIDE 1.99 REVISION 2

Purpose:

This report is in response to a request from the Materials Engineering Branch, Division of Engineering Technology, RES to review the cost analysis contained in Regulatory Guide 1.99 Revision 2 Value Impact Analysis prepared by the Battelle-Pacific Northwest Laboratory (PNL).

Background:

PNL's analysis concludes.that the costs associated with adopting revision 2 of Regulatory guide 1.99 far outweigh the benefits, and that the proposed revision should not be implemented. This value impact analysis is dominated by an estimate of industry operating costs which attempts to capture the replacement energy cost penalty resulting from incremental reactor downtime due to the requirements of revision 2.

Because other impacts are basically at the

" noise" level relative to this consideration, our review focuses predominately on this cost estimate. Additional comments, of a more minor nature, are also provided on several of the other costs estimates presented in the PNL analysis.

Industry Operating Costs:

Industry operating costs dominate the overall value impact analysis. The contractor derives a cost of abcut $180 million by assuming that 86 reactors will experience incremental downtime of betsveen 1.o 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during each reactor's startup. This estimate is based on an average daily replacement energy cost penalty of $300,000 perreactof,anaverageof6startupsper reactor per year, and an average remaining lifetime for reactors of 25 years.

Our review indicates that the resulting cost estimate is flawed in two respects.

In addition, a procedure is readily available that could significantly refine the estimate; a procedure we believe to be warranted given its dominance on the overall value impact equation.

A-The current estimate captures costs that will occur over the next 25 years, and for a large subset of impacted reactors (45) where construction is still on-going, impacts will be experienced well beyond

p the 25 year timeframe.

In either case, the cost is clearly future oriented yet no discounting has been perfonned. Use of a 10% real l

discount rate, as is applied to other dollar costs included'in the value impact analysis, would effectively reduce the industry operating cost of $180 million by approximately a factor of three.

8-The CAG notes that if revision 2 is adopted, 10 operating reactors would benefit (See Table 1, page 1.1 of Value Impact Analysis). This benefit 4

is manifested in a decrease in P-T limits down the temperature scale.

If this in turn results in a reduction in startup time without any incremental risk of vessel fracture then a quantified benefit is realized, i.e., part of the replacement energy currently required during startup for these ten reactors could be effectively reduced. This savings should be deducted from the cost attributed to the 86 ratcheted plants to produce a net industry operating cost that is somewhat less than the current estimate.

C Although $300,000 per day is probably a reasonable representation of a daily replacement energy cost, it must be noted that significant variation exists between reactors. For example, a recent study completed for the CAG (Replacement Energy Costs for Nuclear Electricity-Generating Units in The United States -- NUREG/CR-4012) estimates replacement energy costs varying by reactor from less than $10,000 per day to in excess of $900,000 per day. Depending on the actual reactors impacted by this regulatory change, a significant difference in the bottom line estimate could result. Given the importance of this cost on the overall value impact analysis, a more in depth, case specific, analysis of this cost may be warranted. For example if you could identify the impacted reactors, the CAG could provide you with a more precise measurement of the replacement energy cost penalty based on reactor specific cost estimates.

In addition, we could provide this in a present worth context to make it consistent with the 10% discount rate used in evaluating all the other future dollar impacts already included in the value impact analysis.

g f

- o 4

3 Offsite (Public) Property PNL's best estimate and upper bound for avoided offsite property damage are

$3.3E+4, and $3.6E+5 respectively. Our interpretation of the methodology employed by PNL to obtain these results is to multiply the estimate of offsite property damage (V), by the change in failure probability per transient (W),

by the number of impacted reactors (X), by the number of transients per year (Y), by the present worth discount factor (Z). Applying this logic, we obtain estimates that are approximately one order of magnitude greater than those appearing in the contractor's report.

Best Estimate = (V)-(W)-(X)-(Y)-(Z) = $1.7E+9 X 4.1E-8 X 86 X 6 X 9

=

$3.2E+5 Upper Bound = (V)-(W)-(X)-(Y)-(Z) = $9.2E+9 X 8.2E-8 X 86 X 6 X 9

=

$3.5E+6 In addition, we note that PNL's estimate of offsite property damage (V) is obtained from Strip (Estimates of the Financial Consequences of Nuclear Power ReactorAccidents-NUREG/CR-2723),andtheheforethesedollarvaluesare expressed in 1980 dollars. Strip's estimates should be adjusted by the GNP implicit price deflator to express these avoided costs in 1984 dollars.

Onsite Property:

Given the level of detail provided in the contractor's report, the CAG was unable to verify the methodology or final estimates being offered by PNL with respect to its estimate of avoided onsite property damage. However, there is some concern on our part that the analysis may contain two errors.

First it appears that PNL adopts a present worth discount factor (multiplier) of 5.8 to

s, i

4

(

capture the present value of a cost stream spread over a 10 year period.

a However, in addition to capturing the distribution over a 10 year period, one f

must also account for the fact that the 10 year stream could comence any time i

over the remaining 25 year period. This phenomenen adds significantly to the multiplier. For example, in Strip, the multiplier used to capture a 10 year cost stream over a 32 year remaining life at 4% discount rate was 149. Given that PNL's analysis is predicated on a 10% discount rate and 10 year cost i

streams over a 25 year remaining life the multiplier should be 58 instead of 5.8.

Second, CAG's attempt to replicate PNL'i values of $3.5 E+4 and $1.1 E+5 for the value of avoided offsite property damage suggests that even thef 5.8 multiplier was omitt'ed from their calculations.

Editorial Coments and Error Identification:

s P 2.1 Second paragraph, third line should be " Temperature" not "Temperture" 4

,3 4

P 6.5 Table 4: The total core melt frequencies are presented as the sum of the frequencies of the individual release categories. However, the PWR upper bound estimate for the total core melt (6.8E-8) is not equal l

to the sum of the release categories'(8.3E-8).

It is not possible to deterinine if the error occurred in one of the addends or in the If the sum or total core melt frequency of the PWR upper bound sum.

l estimateisincorrect,thereareNsultingerrorsintables6,7,8 f

9, and 10 as well as in the Sussnary of Value-Impact Assessment Table I

on page 7.1.

P 6.7 Equation: DT0A = NTDOA = F (DIO + DLTO) is not correct given P variable definitions provided on that page and is not consisten, with the results contained in Table 8.

l 1

l L

l

'i 5

l p

P. 6.10 Table 10: Headings "Offsite Property Damage" and "Value of Avoided Offsite Property Damage" should both read "Onsite".

P. 6.11 First equation: (2hourslosttime)($300,000/ day)= $2.5E+4 needs to be restated with a conversion factor of (1 day /24 hours) so that the units are correct and the consistent answer may be calculated.

P. 7. I First paragraph, last sentence: "the cost of one man-rem avoided is estimated to be in the $80,000 - $90,000 range." Should be "...in the $80,000 - $100,000 range." Using the data from the "Sunnary of Value - Impact Assessment" Table, the "Best - Estimate" Cost is 1.8E+8/1.81E+3 or $99,000.

\\

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PLANT-SPECIFIC REPLACEMENT ENERGY COST ESTIMATES FOR THE REGULATORY GUIDE 1.99, REVISION 2 VALUE-IMPACT ANALYSIS November 1984

'repared by* Cost Analysis Group Office of Resource Management J.S. Nuclear Regulatory

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PLANT-SPECIFIC REPLACEMENT ENERGY COST ESTIMATES FOR THE REGULATORY GUIDE 1.99, REVISION 2 VALUE-IMPACT ANALYSIS 1

l purpose:

This report is in response to a request from the Materials Engineering Branch (MEBR), Division of Engineering Technology, RES to modify the replacement energy cost estimate contained in the Value-Impact Analysis of Regulatory Specifically, MEBR requested CAG to provide Guide 1.99, Revision 2.

plant-specific replacement energy cost estimates in present v from the proposed revision.

4

Background:

CAG reviewed the cost analysis for Revision 2 of Regulatory Guide 1.99 and I

offered to assist MEBR by providing more precise replacement energy cost In our initial review, we observed that the cost analysis was estimates.

dominated by the estimate of industry operating costs, i.e., replacement CAG's energy costs, and that the estimate was flawed in several resp in a recently published NUREG - prepared under contract for CAG by the Argonne National Laboratory.

Replacement Energy Costs:

This analysis was performed using the following assumptions which were listed in the October 30, 1984 memorandum of Lloyd Donnelly to Charles Serpan:

Replacement energy cost penalties are assumed to be constant in real 1.

This means terms over the remaining useful life of the reactor.

costs are not assumed to increase faster than the rate of general inflation.

For reactors already operational, replacement energy cost penalties 2.

are assumed to cormnence in 1985. Comercial operating life of a reactor is assumed to be 30 years. Thus, remaining useful life equals 30 minus number of years k operation prior to 1985.

For future reactors, replacement energy cost penalties will comence 3.

in estimated year of initial cognercial operation and continue for 30 years.

All costs will be expressed in 1984 constant dollars and discounted 4.

The final cost back to 1984 assuming a 10% real discount rate.

estimate will represent a 1984 present worth value in 1984 dollars.

The estimate of change in down time provided in the attachment to 5.

your memo of October 23, 1984 is assumed to be constant over the remaining useful life of each reactor.

M Van Kuiken, J.C., W.A. Buehring, and K.A. Guziel, Argonne National Laboratory, " Replacement Energy Costs for Nuclear Electricity 4012.

Generating Units in the United States," USNRC Report NUREG/CR-

MEBR furnished CAG with a list of the affected plants and the estimated impact of Revision 2 in terms of hours per year of plant-specific repla needed.

The costs were converted to hourly previously mentioned Agonne work.

One additional adjustment was required because CAG assumes the reactors would have been fully available at the time of the incremental figures.

outages, whereas the Argonne study is based on an average availability factor.

CAG calculated the yearly cost for each plant by multiplying the number of hours per year.the plant would not be producing due to Revision 2, and hence requiring replacement energy, times the hourly replacement energy cost rate.CAG EE5 r

Th] attached table contains the pertinent data. number of years remain L

The

' 27 plants were grouped by years of expected lifetime remaining.The present value of replacement energy costs were summed across each group.

I these annual costs were obtained using an annuity formula based on a I

percent discount rate.

This value is calculated to be worth value expressed in 1984 dollars.which is approximately sixty percent less t I

$70,742,600, cstimate given in the' original Value-Impact Analysis.

E-c.1l C

[,

=

b L.

s K.

B B'

'I CHANGES IN PLANT ENERGY REPLACEMENT COSTS DUE TO REGULATORY GUIDE 1.99, REVISION 2 Change in Plant Average Daily Change in Plant Down Time Due Energy Replacement Energy Replacement i

to Requirement Cost Costs Per Year (hours / year)

(thousands of (thousands of 1984 dollars) 1984 dollars) g i

Palisades 12.2 297 207.4 Sequoyah 2 10.8 306 189.0 Indian Point 2 9.5 580 314.4 Hatch 1 7.8 349 155.2 Hatch 2 7.8 362 161.5 Peach Bottom 2 7.7 521 228.7 Peach Bottom 3 7.7 513 225.6 Calvert Cliffs 1 7.1 440 178.2 Maine Yankee 5.5 646 203.0 Davis Besse 1 5.4 -

273 84.2 Fitzpatrick 5.4 574 177.1 4.8 306 84.0 Sequoyah 1 4.8 90 24.5 Watts Bar 1 4.8 310 85.0 Kewawnee Monticello 4.8 116 31.7 McGuire 2 4.2 527 126.4 Midland 2 4.2 334a 80.2 Bellefonte 1 4.2 329h 79.0 Bellefonte 2 4.2 329f 79.0 Washington Nuclear 1 4.2 380m 91.1 Millstone 1 4.2 511 122.6 4.2 343 82.3 Oyster Creek Browns Ferry 1 4.2

['

288 68.9 Browns Ferry 2 4.2 288 68.9 Browns Ferry 3 4.2 288 68.9 Brunswick 1 4.2 341 81.9 Brunswick 2 4.2 e

341 81.9 4.2 130 31.1 Duane Arnold 4.2 106 25.2 Cooper La Salle 1 4.2 598 143.2 Susquehanna 1 4.2 545 130.6 Fermi 2 4.2 522 125.2 4.2 376 90.3 Perry 1 Perry 2 4.2 376f 90.3 Clinton 1 4.2 5271 126.4 Grand Gulf 2 4.2 553f 132.7 Hope Creek 1 4.2 564k 135.2 Limerick 1 4.2 555 133.1 Limerick 2 4.2 555f 133.1

CHANGES IN PLANT ENERGY REPLACEMENT COSTS DUE'TO REGULATORY GUIDE 1.99, REVISION 2 (continued)

Change in Plant Average Daily Change in Plant Down Time Due Energy Replacement Eneroy Replacement to Requirement Cost Cost: Per Year (hours / year)

(thousands of (thousands of 1984 dollars) 1984 dollars) f Nine Mile Point 2 4.2 683j 163.8 River Bend 1 4.2 396b 94.9 Shoreham 1 4.2 541c 129.8 Susquehanna 2 4.2 556 133.1 Cook 2 3.7 299 63.3 Cook 1 3.6 288 59.0 Arkansas 2 3.6 369 76.0 Trojan 3.6 324 66.6 San Onofre 2 3.6 758 155.9 San Onofre 3 3.6 758 155.9 Summer 1 3.6 399 82.1 St. Lucie 2 3.6 420 86.4 Midland 1 3.6 201d 41.4 Braidwood 1 3.6 383e 78.8 Braidwood 2 3.6 383f 78.8 Byron 1 3.6 580 119.2 Byron 2 3.6 580f 119.2 Callaway 1 3.6 349 71.6 Comanche Peak 1 3.6 711 146.2 Comanche Peak 2 3.6 684a 140.4 Marble Hill 1 3.6 444n 91.1 Marble Hill 2 3.6 444f 91.1 Millstone 3 3.6

['

856e 175.7 Wolf Creek 1 3.6 430 88.2 Shearon Harris 1 3.6 360e 73.8 Palo Verde 1 3.6 836 171.7 Palo Verde 2 3.6 8

822g 168.8 Palo Verde 3 3.6 822f 168.8 Seabrook 1 3.6 856a 175.7 Seabrook 2 3.6 856f 175.7 South Texas 1 3.6 773p 158.8 South Texas 2 3.6 773f 158.8 Waterford 3 3.6 496 101.9 Washington Nuclear 3 3.6 373m 76.7 Beaver Valley 2 3.6 254e 52.2 Catawba 1 3.6 487c 100.1 Catawba 2 3.6 487f 100.1 Vogtle 1 3.6 5531 113.8 Vogtle 2 3.6 553f 113.8

r CHANGES IN PLANT ENERGY REPLACEMENT COSTS DUE TO REGULATORY GUIDE 1.99, REVISION 2 (continued)

Change in Plant Average Daily Change in Plant Down Time Due Energy Replacement Energy Replacement to Requirement Cost Costs Per Year (hours / year)

(thousandsof (thousands of 1984 dollars) 1984 dollars) t 3.6 548 112.7 i

Pilgrim 1 Grand Gulf 1 3.0 553 94.8 3.0 598 102.3 La Salle 2 321 54.9 3.0 Washington Nuclear 2 3.0 331 56.7 Surry 2 North Anna 2 3.0 393 67.2 Three Mile Island 1 2.5 406 58.0 j

Indian Point 3 1.8 649 66.6 1.8 537 55.1 1.8 476 49.0 Zion 2 Crystal River 3 387 39.8 North Anna 1 1.8 27.9 Beaver Valley 1 1.8 271 McGuire 1 1.8 527 54.2 Diablo Canyon 2 1.8 850 87.3 Salem 2 1.7 585 56.8 Watts Bar 2 1.2 310f 21.2 Diablo Canyon 1 1.2 852 58.3 1.2 351 24.0 Farley 1 Salem 1 1.2 570 39.0 Arkansas ANO 1 1.2 360 24.6 0

N/A N/A Robinson 2 0

N/A N/A Oconee 2 0

N/A N/A Fort Calhoun AS N/A N/A O

Rancho Seco 0

N/A N/A Oconee 1 0

N/A N/A 0

N/A N/A Ginna Oconee 3 0

N/A N/A Farley 2 0

N/A N/A St. Lucie 1 0

N/A N/A Big Rock Point 0

N/A N/A Dresden 2 0

N/A N/A Dresden 3 Nine Mile Point 1 0

N/A N/A 0

N/A N/A Quad Cities 1 0

N/A N/A Quad Cities 2 O

N/A N/A Vennont Yankee 7.2 Prairie Island 2

- 1.2 106

- 1.2 537

- 36.7 Zion 1 Calvert Cliffs 2

- 1.2 440

- 30.1

6-COSTS CHANGES IN PLANT ENERGY REPLAC ON 2 E

(continued)

Changt in Plant Average DailyEnergy Replacement Energ Chan9e in Plant Costs Per Year Down Time Due Cost (thoustnds of to Requirement (thousandsof 1984 dollars)

(hours / year) 1984 dollars)

g

- 32.9 321 19.8 1.8 115 84.7 3.0 424 84.7 Surry 1 Point 8each 1 3.5 424

- 23.8 Turkey Point 3 3.5 115 21.6 Turkey Point 4 3.6 106 2.2 Paint Beach 2 3.6 10 39.8

~

Prairie Island 1 3.6 131

-183.4 5.3 669

-268.9 l

Lacrosse 4.8 436

-282.2 Yankee Rowe Millstone 2

-10.8 295 Haddam Neck

-16.8 l

San Onofre 1

=

NOTES FOR COLUMN 2 y

~v Based only on the Summer 1986 estimate, d Sumner 1985 cost estimates.

1985/86, Spring 1986, an a.

Average of Winter 6 estimates.

Average of Sumner 1985 through Spring 198for Midland 2 b.

Based only on the Sumner 1986 estimate c.

j; d.

to size.

Average of Spring 1986 and Sumner 1986.

they are the same size.

Used same estimate as for Unit 1 because e.

i 6

Average of Fall 1985 through Sumner 198,

Bellefonte is 1235 net f.

t MWe.

Sequoyah 1 and 2 are $306 each 9 1,148 n Clinton is 950 net MWe, g.

b $329.

h.

t MWe, LaSalle 1 and 2 are $598 each 9 1,078 ne be $527.

therefore its cost is estimated to Unit 2 is 1,080 not MWe, 1.

MWe.

Nine Mile Point 1 is $386 9 610 net be $683.

therefore its cost is estimated to J.

a r

7 e

Salem 1 is $570 9 1,079 net MWe. Hope Creek 1 is 1,067 net MWe, k.

therefore its cost is estimated to be $564.

Hatch 1 is $349 9 764 net MWe. Vogtle 1 and 2 are 1,210 net MWe each, 1.

therefore their cost is estimated to be $553 each.

C.

WNP 2 is $331 9 1,103 net MWe. WNP 1 is 1,266 net MWe, therefore its cost is estimated to be $380. WNP 3 is 1,242 net MWe, therefore its cost is estimated to be $373.

Er Marble Hill's costs are based on the ECAR average, n.

net MWe. South Texas 1 and 2 are both Comanche Peak 1 is $711 9 1,150 i

1,250 net MWe, therefore their costs are estimated to be $773 each.

L p.

U.S. NRC, Office of Nuclear Regulatory Research; Van Kuiken, J.C.,

SOURCES:

i et al. " Replacement Energy Costs for Nuclear Electricity Generating Units in the United States," USNRC Report NUREG/CR-4012, r

October,1984.

9 I

t, 5

F l

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