ML20236M072
| ML20236M072 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 02/28/1986 |
| From: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| Shared Package | |
| ML20236M039 | List: |
| References | |
| TASK-ME-305-4, TASK-RE REGGD-01.099, REGGD-1.099, NUDOCS 8708100462 | |
| Download: ML20236M072 (20) | |
Text
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\q February 1986 3
f U.S. NUCLEAR REGULATORY COMMISSION 0FFICE OF NUCLEAR REGULATORY RESEARCH Division 1 3 .$* Task ME 305-4 DRAFT REGULATORY GUIDE AND VALUE/ IMPACT STATEMENT i k*****[
Contact:
P. N. Randall (301)443-7711 PROPOSED REVISION 2 TO REGULATORY GUIDE 1.99 RADIATION DAMAGE TO REACTOR VESSEL MATERIALS A. INTRODUCTION _ General Design Criterion 31, " Fracture Preve o ctor Coolant Pressure Boundary," of Appendix A, " General De ia for Nuclear Power Plants," to 10 CFR Part 50, " Domestic Licensin uction and Utilization Facilities," requires, in part, that the r lant pressure boundary be designed with sufficient margin to ensu h w en stressed under operating, maintenance, testing, and postulated acc kconditions,(1)theboundary behaves in a nonbrittle manner and robability of rapidly propagating i
- fracture.is minimized and that 1 reflects the uncertainties in deter- i mining the effects of irradiati on terial properties. Appendix G, " Fracture Toughness Requirements," a $ ppe H, " Reactor Vessel Meterial Surveillance i Program Requirements," whi im ement, in part, Criterion 31, necessitate the calculation of changes in fra ure toughness of reactor vessel materials caused by neutron radiation throughout the service life. This guide describes general A
procedures accep abl 4 o the NRC staff for calculating the effects of neutron ! radiation dam e he low-alloy steels currently used for light-water-cooled reactor ves s. The 1 e procedures given in Regulatory Position 1.1 of this draf t j guide a t the same as those given in the Pressurized Thermal Shock rule * )1
- I
*Section 50.61, " Fracture Toughness Requirements for Protection Against f Pressurized Thermal Shock Events," of 10 CFR Part 50 published July 23, 1985 (50 FR 299371 8708100462 870731 PDR ADOCK 05000483 p PDR This regulatory guide and the associated value/ impact statement are being issued in draf t form to involve the public in the l early stages of the development of a regulatory position in this area. They have not received complete staf f review and do j not represent an of ficial NRC staf f position. !
1 Public coMents are being solicited on both draf ts, the guide (including any implementation schedule) and the value/ impact l statement. Coments on the value/ impact statement should be accompanied by supporting data. Written coments may be j
) submitted to the Rules and Procedures Branch. DRR, ADri. U.S. Nuclear Regulatory Commission. Washington, DC 20$55. Coments may also be delivered to Room 4000, Nryland National Bank Buildin*j. 7735 Old r.corgetown Road, Bethesda, %ryland from 0:15 a.m. to 5:00 p.m. Copies of coments received may be examined at the PRC Public Document Room,1717 H 5treet NW. ,
Washington, DC. Coments will be most helpful if received by April 15, 1986. Requests for single copies of draf t guides (which may be reproduced) or for placement on an automatic distribution list for
$1ngle copies of future draf t guides in specific divisions should be opde in writing to the U.S. Nuclear Re3ulatory Comisdon, Washington, OC 20555. Attention: Director, Olvision of Technical Information and Document Control.
I
r for calculating RTPTS, the reference temperature that is to be compared to the screening criterion given in the rule. Issuance of this draft regulatory guide 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 specific plants for purposes not related to PTS, chiefly as the basis for calculating pressure-temperature limits as required by Appendix G to 10 CFR Part 50. Licensees may also consider and comment on the effect of the proposed change on the calculated PTS risk at their plant assuming that the Revision 2 correlation, if justified, would at some future time replace the RT PTS correlation in the PTS rule. After resolution of comments and general agreement is reached regarding the best way to calculate RTNDT, it will be appropriate to reevaluate the overall conservatism of the PTS rule and consider whether amendment of the PTS rule is desirable. Any information collection activities mentioned in this draft regulatory guide are contained as requirements in 10 CFR Part 50, which provides the regu-latory basis for this guide. The information collection requirements in 10 CFR Part 50 have been cleared under OMB Clearance No. 3150-0011.
! 4 B. DISCUSSION Some NRC requirements that necessitate calculation of radiation damage I are:
- 1. Paragraph V.A of Appendix G requires the effects of neutron radia-tion to be predicted from the results of pertinent radiation effects studies.
This guide provides such results in the form of calculative procedures that are acceptable to the NRC.
- 2. Paragraph V.B of Appendix G describes the basis for setting the upper limit for pressure as a function of temperature during heatup and cooldown for a given service period in terms of the predicted value of the adjusted reference temperature at the end of the service period.
- 3. The definition of reactor vessel beltline given in Paragraph II.F of Appendix G requires identification of regions of the reactor vessel that are predicted to experience sufficient neutron radiation 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 requiremen,ts for reactor vessel materials generally. 2
) 4. Paragraph II.B of Appendix H incorporates ASTM E 185 by reference.
Paragraph 5.1 of ASTM E 185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels" (Ref.1), requires that the materials to be placed in surveillance be those that may limit operation of the reactor during its lifetime, i.e. , those expected to have the highest adjusted reference temperature or the lowest Charpy upper-shelf energy at end of life. Both measures of radiation damage must be considered. In Paragraph 7.6 of ASTM E185-82, the requirements for the number of capsules and the withdrawal schedule are based on~the calculated amount of radiation damage at end of life. 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 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 shif t in the Charpy curve for the irradiated material relative to that l for the unirradiated material measured at the 30-foot pound energy level, and L 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 l energy lev'el, which is defined in ASTM E 185-82. This proposed Revision 2 updates the calculative procedures for the adjustment of reference temperature; however, calculative procedures for the decrease in upper-shelf energy are unchanged because the preparatory work had not been completed in time to include them in this revision.
.The basis for Equation 2 for ART NDT surface (in Regulatory Position 1.1 l
of this guide) is contained in publications by G. L. Guthrie (Ref. 2) and G. R. Odette and P.M. Lambrozo (Ref 3). These authors used surveillance data from commercial power reactors as their data base,~but their analysis techniques l were different. They recommended that (1) separate correlation functions should be used for weld and base metal, (2) the function should be the product of a chemistry factor and a 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 1019 n/cm2 (E > 1 MeV), steeper at low fluences and flatter at high fluences. Regulatory Position 1.1 is a blend of the correlation functions presented by
)
these authors. Some test reactor data were used as a guide in establishing a 3
cutoff for the chemistry factor for low-copper materials. The data base for Regulatory Position 1.2 is that given by Spencer H. Bush (Ref. 4). The measure of fluence used in this guide is the number of neutrons per square centimeter having energies greater than 1 million electron volts (E > 1 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) (Ref. 5) in the analysis of the surveillance data base (Ref. 6). I However, the neutron energy spectrum does change significantly with loca-tion in the vessel wall; hence for calculating the attenuation of radiation damage through the vessel wall, it is necessary to use a damage function to determine ART NDT versus radial distance into the wall. The most widely accepted damage function at this time is dpa, and the attenuation formula (Ecuation 3) given in Regulatory Position 1.1 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. The original version and Revision 1 of this guide had a phosphorus term in the chemistry factor, but the studies on which this revi-sion was based found other elements such as phosphorus to be of secondary impor- - tance, i.e. , including them in the analysis did not produce a significantly better fit of the data. Scatter in the data base used for this guide is relatively significant, as evidenced by the fact that the standard deviations for Guthrie's derived formulas (Ref. 2) are 28 F for welds and 17 F for base metal despite extensive statistical analysis. Thus the use of surveillance data from a given reactor (in place of the calculative procedures given in this guide) requires consider-able engineering judgment to evaluate the credibility of the data and assign suitable margins. When surveillance data from the reactor in question become available, the weight given to them relative to the information in this guide
)
will depend on the credibility of the surveillance data as judged by the follow-ing criteria:
- 1. Materials in the capsules should be those judged most likely to be controlling with regard to radiation damage according to the recommendations of this guide.
- 2. Scatter in the plots of Charpy energy versus temperature for the I
irradiated and unitradiated conditions should be small enough to permit the 4
- _ _ _ _ _ _ . _ _ _ . _ _ _ _ __ ___J
i. J determination of the 30-foot pound temperature and the upper-shelf energy unambiguously.
- 3. When there are two or more sets of surveillance data from one reactor, the scatter of ARTNDT . values about a best-fit line drawn as described in Regula-tory Position 2.1 normally should be less than 28 F for welds and 17 F for base metal. Even if the fluence range is large (two or more orders of magnitude),
the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, i l l following the definition given in ASTM E 185-82 (Ref. 1).
- 4. The irradiation temperature of the Charpy specimens in the capsule ,
I should match vessel wall temperature at the cladding / base metal interface q within 125 F. l
- 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 { NDT the' slope, i.e., the fluence factor, from other than the plant data. Instead, Equation 2 (in Regulatory Position 1.1) 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 ARTNDl; this seems reasonable for fitting surveillance data, because generally the higher data point will 'be ) the more recent and therefore will represent more modern procedures. C. REGULATORY POSITION
- 1. SURVEILLANCE DATA NOT AVAILABLE 1 I
When credible surveillance data from the reactor in question are not avail-able, calculation of neutron radiation damage to the beltline of reactor vessels j
)
of light-water reactors should be based on the procedures in Regulatory Positions 1.1 and 1.2 within the limitations in Regulatory Position 1.3. 5 l
1.1 Adjusted Reference Temperature , The adjusted reference temperature (ART) for each material in the beltline is given by the following expression: ART = Initial RTNDT + ARTNDT + Margin (1) , Initial RT is the reference temperature for the unirradiated material-NDT as defined in. Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code-(Ref. 7). If measured values of Initial RT NDT f r the material in question are not available, generic'mean values for that class
- of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.
ART is the mean value of the adjustment in reference temperature caused NDT by irradiation and should be calculated as follows: N (2) ART NDT surface = M M I' CF ( F) is the chemistry factor, a function of copper and nickel content. CF is given in Table 1 for welds and in Table 2 for base metal (plates and forgings). Linear interpolation is permitted. In Tables 1 and 2 " weight percent copper" and " weight percent nickel" are the best-estimate values for the material, which will normally be the mean of the measured values for a plate or forging or for weld samples made with the weld wire heat number that matches the cri- i tical vessel weld. If such values are not available, the upper limiting values f given in the material specifications to which the vessel was built may be used. If not available, conservative estimates (mean plus one standard deviation) based on generic data may be used if justification is provided. If there is no information available, 0.35% copper and 1.0% nickel should be assumed. The 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/cm2 (E > 1 MeV) divided by 1028 l is generally determined, for the welds
*The class for estimating with which this guide is concerned, Initial RT"h the type of welding flux (Linde 80 or other); for base metal, by the ASTM Standard Specification.
6
The fluence f actor, f0 .28 - 0.100g f , is determined by calculation or r b . from Figure 1. To calculate ART NDT at any depth (e.g., at 1/4T or 3/4T), the following attenuation formula should be used: ARTNDT = [aRTNDT surface [ (3) where x (in inches) is.the depth into the vessel wall measured from the vescel inner (wetted) surface.
" Margin" is the quantity, F, that is to be added to obtain conservative, upper-bound values of adjusted refo ence temperature for the calculations required by Appendix G to 10 CFR Part 50.
Margin =24oj+oj (4) If a measured value of Initial RT f r the material in question is NDT used, o may be taken as zero. If a generic value of Initial RT is used, f NDT o g should be obtained from the same set of data (see Regulatory Position 1.1). j The standard deviations for ART NDT' a, are 28 F for welds and 17 F for base metal, except o need g not exceed 0.50 times the mean value of ART NDT surface. 1.2 Charpy Upper-Shelf Energy Charpy upper-shelf energy should be assumed to decrease as a function of fluence and copper content as indicated in Figure 2. Linear interpolation is permitted. 1.3 Limitations Application af the foregoing procedures should be subject to the following limitations:
- 1. The procedures apply to those grades of SA-302, 336, 533, and 508 steels having minimum specified yield strengths of 50,000 psi and under and to their welds and heat-affected zones.
- 2. The procedures are valid for a nominal irradiation tempera-ture of 550 F. Irradiation below 525 F should be considered to produce greater 7
J
l' 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.
- 3. Application of these procedures to fluence levels or to copper or nickel' content beyond the ranges given in Figure 1 and Tables 1 and 2 or to '
materials having chemical compositions beyond the range found in the data bases
'used for this guide should be justified by submittal of data.
L2. SURVEILLANCE DATA AVAILABLE j , l When two or more credible surveillance data sets (as defined in the , Discussion) become available from the reactor in question, they may be used to determine the adjusted reference temperature and the Charpy upper-shelf energy of the beltline materials as described in Regulatory Positions 2.1 and 2.2' , t respectively. L 2.1 Adjusted Reference Te perature ; The adjusted reference temperature should be obtained by first fitting the surveillance data using Equation 2 to obtain the relationship of ART NDT surface to fluence. To do so, calculate the chemistry factor, CF, for the best fit by multiplying each measured ART NDT by its corresponding fluence factor, summing the products, and dividing by the sum of the squares of the fluence factors. The resulting value of CF.when entered in Equation 2 will give the relationship of ART surface to fluence that fits the plant surveillance data in such a-NDT way as to minimize the sum of the squares of the errors. To calculate the margin in this case, use Equation 4; the values given there for o may be cut in half. 3 If this procedure gives a higher value of adjusted reference temperature than that given by using the procedures of Regulatory Position 1.1, the surveillance data should be used. If this procedure gives a lower value, either may be used. l 1 2.2 Charpy Upper-Shelf Energy The decrease in upper-shelf energy may be obtained by plotting the reduced plant sur.eillance data on Figure 2 of this guide and fitting the data with a lin? drawn parallel to the existing lines as the upper bound of all the data, m This line should be used in preference to the existing graph. 8 ___ _ ._______________.______________i
'3. REQUIREMENT FOR NEW PLANTS i
For beltline materials in the reactor vessel for a new plant, the content of residual elements such as' copper, phosphorus, sulfur, and vanadium should be controlled to low levels." The copper content should be such that the calculated adjusted reference temperature at the 1/4T position in the vessel wall at end of life is less than '200 F. D. IMPLEMENTATION The purpose of this section is to provide information to applicants and licensees regarding the NRC staff's plans for using this regulatory guide. This proposed revision has been released to encourage public participation in its development. Except in those cases in which an applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the methods to be described in the final guide reflecting public comments will be used by the NRC staff as follows:
?
i'~ l}' l '. The methods described in Regulatory Positions 1 and 2 of this guide will be used in evaluating all predictions of radiation damage needed to imple-ment Appendices G and H to 10 CFR Part 50 for applications submitted 60 days after publication lof the final guide; however, if an applicant wishes to use the recommendations of Regulatory Positions 1 and 2 in developing submittals before that date, the pertinent portions of the submittal will he evaluated on the basis of this guide.
- 2. The owners of all operating reactors and all applicants for an opera-ting license should review the basis for the pressure-temperature limits in their Technical Specifications for consistency with Regulatory Position 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 operat-ing procedures, as appropriate, to conform with the criteria of the final version of Revision 2 of this guide and submit the appropriate revision to their Tech-nical Specifications within 3 years of the date of publication of Revision 2 of this guide in final form. Those for whom the allowable operating period has . .).
*For more information, see the Appendix to ASTM Standard Specification A 533 (Ref. 8).
9
r-been extended, when judged by the criteria of the final version 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.
- 3. The recommendations of Regulatory Position 3 are unchanged from those used to evaluate construction permit applications docketed on or after June 1, 1977.
s_ 1 ( l 1 -- l 10 l l .. . . _ - - - _ _ _ J
TABLE 1 CHEMISTRY FACTOR FOR WELDS, F
)
Copper, Nickel, Wt-% Wt-% 0. 0.20 0.40 0.60 0.80 1.00 1.20 0 20 20 20 20 20 20 20
'0.01 20 20 20 20 20 20 20 0.02 21 26. 27 27 27 27 27 0.03 22 35 41 41 41 41 41 0.04 24 43 54 54 54 54 54 0.05 26 49 67 68 68 68 68 0.06. 29 52 77 82 82 -82 82 0.07 32 55 85 95 95- 95 95-0.08 36 58 .90 106 108 108 108 1 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 j
-- 0.16 70' 88 115 149 178 199 211
(
-) - .O.17- 75 .92 119 151 184 207 221 0.18 79 95 122 154 187 214 230 0.19 83 100 126 157 '191 220 238 0.20 88 104 129 160 194 223 245
~0.21' 92 108 133 164 197 229 252 0.22 97 112 137 167 200 232 257 0.23 101 117 140 169 203 236 263 0.24 105 121 144 173 206 259 268 0.25 110 126 148 176 209 243 272 '
0.26' 113 130 151 180 212 246 276 O.27 119 134 155 '184 '216 249 280 l 0.28 122 138 160 '187 218 251 284 1 0.29 128 142 164 191 222 254 287 l h- 0.30 131 146 167 194 225 257 290
- 0.~ 31 136 151 172 198 228 260 293 0.32 140 155 175 202 231 263 296 l L 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 1 11 I:. L _ _ ----- - _ _ _ _ _ J
J.. L .. e i TABLE 2 CHEMISTRY FACTOR FOR BASE METAL, *F Copper, Nickel, Wt-% l Wt-% 'O 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 L 0.03 20 20 20 20 20 20 20 j' O.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 l 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 O.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 1 4 0.17 69 88 110 127 132 135 135 l 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 d 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- 1 0.28 119 138 160 187 216 233 239 0.29 124 142 164 191 221 241 248 I 0.30 129 146 167 194 225 249 257 j 0.31 134 151 172 198 228 255 266 1 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 0.37 162 177 196 220 248 278 308 0.38 166 182 200 223 250 281 313 ( '- 1 0.39 171 185 203 227 254 285 317 0.40 '175 189 207 231 257 288- 320 12 i
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l- . F l REFERENCES l' l- . 1. American Society for Testing and Materials, " Standard Practice for Conduct-ing Surveillance Tests for Light-Water Cooled Nuclear Power Reactor' Vessels," E 185-82, July 1982.*
- 2. G. L. Guthrie, "Charpy Trend Curves Based on 177 PWR Data Points," in " LWR Pressure Vessel Surveillance Dosimetry Improvement Program," NUREG/CR-3391, Vol. 2, prepared by Hanford Engineering Development Laboratory, HEDL-TME 83-22, April 1984.**
- 3. G. R. Odette et al., " Physically Based Regression Correlations of Embrittle-ment Data from Reactor Pressure Vessel Surveillance Programs," Electric Power Research Institute, NP-3319, January 1984.t
- 4. S. H. Bush, " Structural Materials for Nuclear Power Plants," in " Journal of Testing and Evaluation," American Society for Testing and Materials, November 1974.*
- 5. American Society for Testing and Materials, " Standard Practice for Charac-terizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (DPA)," E 693-79, August 1979.*
- 6. W. N. McElroy, " LWR Pressure Vessel Surveillance Dosimetry Improvement Program: LWR Power Reactor Surveillance Physics--Dosimetry' Data Base
- Compendium," NUREG/CR-3319, prepared by Hanford Engineering Development Laboratory, HEDL-TME 85-3, August 1985.**
- 7. American Society of Mechanical Engineers, Section III,~" Nuclear Power 1 Plant Components," of_ASME Boiler and Pressure Vessel Code, New York (updated frequently).tt
- 8. American Society for Testing and Materials, " Standard Specification for Pressure Vessel Plates, Alloy Steel, Quenched and Tempered, Manganese-Molybdenum and Manganese-Molybdenum-Nickel," A 533/A 533M-82, September 1982.*
I
]
" Copies may be obtained from the American Society for Testing and Materials, f
1916 Race Street, Philadelphia, PA 19103
** Copies may be obtained from the Superintendent of Documents, U.S.
Government Printing Office, Post Office Box 37082, Washington, DC l 20013-7082. i tCopies may be obtained from the Electric Power Research Institute, 3412 i Hillview Avenue, Palo Alto, CA 94304. ! ttCopies may be obtained from the American Society of Mechanical Engineers, 1 345 E. 47th Street, New York, NY 10017. 15 l i
]
+
ORAFT REGULATORY ANALYSIS
- 1. STATEMENT OF THE PROBLEM.
Knowledge of fracture toughness as a function of temperature of reactor vessel beltline materials is needed for decisions dealing with the preven-tion of. fracture of the reactor vessel. Appendix G to 10 CFR Part 50 requires the effects of neutron radiation to be predicted from the results of pertinent l radiation effects studies. Many plants now have surveillance data of their own, but such data are not yet available for the new plants. The NRC staff must rely on calculated values based on the chemical composition of the vessel materials and the neutron fluence for licensing oecisions. The calculative procedures r eed to be expanded and upgraded.
- 2. OBJECTIVE i
The objective of proposed Revision 2 to Regulatory Guide 1.99 is to upgrade the calculative procedures in Revision 1 by basing them on the mort pertinent radiation data and the best available understanding of radiation damage in reactor -
)!
vessel materials in accordance with Appendix G to 10 CFR Part 50. Revision 2 i to Regulatory Guide 1.99 would provide a basis acceptable to the NRC staff to account for the effects of neutron radiation on the fracture toughness of reactor vessel matorials. 'This basis is needed in the calculation of pressure-temperature limits, in the analysis of transients that threaten vessel integrity, and in the analysis of beltline flaws.
- 3. ALTERNATIVES
.The following alternatives were considered with respect to revising )
Regulatory Guide 1.99: i 3.1 Alternative 1 Retain the guidance in Revision 1 of Regulatory Guide 1.99. t 16
_ _ , _ _ = _ _ _ _ _ _ - . . _ - . ._ _ .-- . _ _ - 1 3.2 Alternative 2 J Withdraw Regulatory Guide 1.99 3.3 Alternative 3-Revise Regulatory Guide 1.99 to update the guidance'and calculative procedures contained in Revision 1.
- 4. CONSEQUENCES, COSTS, AND BENEFITS OF EACH ALTERNATIVE 4.1 ' Alternative 1 Consequences: If Revision 1 is retained as the basis for. pressure-temperature limits, about half of the plants will be operating with limits that provide a reduced margin of safety against vessel fracture, that'is, the operator will be misled by erroneous pressure-temperature limits as to the potential !
severity of system transients and will have less time to take corrective action
~~ }. --' to-avoid hazardous pressure-temperature conditions.
Cos t's: There would, of course, be no change in the present operating costs. Benefits: For those plants that would have to raise their pressure-temperature limits if Revision 2 were adopted, the resulting additional con-straints on operation would be avoided or at least postponed. 4.2 Alternative 2-Consequences: The NRC staff yeviews several pressure-temperature limits L per year-(plus occasional transients and flaw indications), and there would be no published basis for its reviews. There is presently nothing equivalent to !. Regulatory Guide 1.99 in the ASME Code. Because.each utility should use the most up-to-date information.as the basis for their pressure-temperature limits, the impact on most plants would be similar to that resulting from the use of Revision 2. I-17
Costs: There would be considerable added costs to the utilities for pre-paration, and_to the NRC for review, of case-by-case submittals justifying the values of. adjusted reference temperature that would be needed as the basis for pressure-temperature limits. Benefits: No benefits beyond those provided by adoption of Revision 2 would result from having no guidance on radiation damage estimates. (Even with Revision 2, utilities would be permitted to submit their own bases for pressure-temperature limits if they supplied full justification.) 4.3 Alternative 3 Consequences: Adoption of Revision 2 would upgrade the calculative pro-cedures in Revision 1 of Regulatory Guide 1.99 by basing them on the most per-l- tinent radiation data and the best available understanding of radiation damage l in reactor vessel materials. Consequently, the margin of safety would be restored to levels consistent with NRC regulations if Revision 2 becomes the basis for pressure-temperature limits at each plant. Costs: The cost of adopting Revision 2 would derive mainly from the. cost j of purchased power during delays in startup caused by the more restrictive - operating limits. Pacific Northwest Laboratories estimated that eight plants are expected to have to raise RT NDT 50 to 100 F. These plants would have an estimated delay of 2 hours per heatup, with six heatup/cooldown cycles per year for a cost of $150,000 per year per plant for the next 25 years. Seventy-seven plants, including those undergoing licensing, are expected to have to raise RT NDT 20 to 50 F per year; they are assumed to incur half that amount. The best-estimate present value of these industry operating costs was given by PNL as $101 million and the net cost to industry was $63 million. The bottom-line, l best-estimate cost per person-rem saved reported by PNL is from $3,500 to $5,600. l NRC's Cost Analysis Group estimates the net cost to industry to be much lower. Based on their review and other analyses done by the NRC staff, the figure was reduced to approximately $900 to $1,400 per person-rem saved. Benefits: The procedures for calculating radiation damage to reactor ! i vessel materials would be upgraded. They would be based on the most pertinent i radiation data and the best available understanding of radiation damage processes. l 1 1 I 1 18
) Thus, the margin of safety against vessel failure would be restored to levels believed to exist when Revision 1 was issued.
4.4 Conclusions Revision 2 to Regulatory Guide 1.99 should be issued to provide updated guidance on calculating radiation damage to reactor vessel materials.
- 5. IMPACTS ON OTHER REQUIREMENTS The calculative procedures given in Regulatory Position 1.1 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 the proposed Revision 2 to Regulatory Guide 1.99 for public comment in no way affects the recently promul-gated PTS rule. Licensees and the technical commelity are requested to comment on the technical merits of this proposal, including its effect on their plants for purposes not related to PTS, chiefly as the basis for calculating pressure-temperature limits as required by Appendix G. Licensees may also consider and comment on the proposed change's effect on the calculated PTS risk at their plant, assuming that, if justified, the proposed correlation would at some future time replace the RT PTS correlation in the PTS rule. Following resolution of comments and after general agreement is reached regarding the best way to cal-culate RTNDT, it will be appropriate to reevaluate the overall conservatism of the PTS rule.
- 6. CONSTRAINTS There have been no constraints identified that would affect the implementa-tion of Revision 2 to Regulatory Guide 3.99.
- 7. DECISION RATIONALE Based on the analyses of the safety issues, system impacts, and costs, it is recommended that Revision 2 of Regulatory Guide 1.99 be issued for public
*Section 50.61, " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," of 10 CFR Part 50, published July 23, 1985 (50 FR 29937).
19
. _ _ _ _ _ _ _ _ _ _ _ D
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comment. The analyses hue shown that the guide is needed because it provides part of the basis for ensuring safe operation of reactors during startup and shutdown and for the evaluation of transients and flaws found in service. Periodic updating of the guide is consistent with the requirements of Appendix G to 10 CFR Part 50. !
- 8. IMPLEMENTATION i
All operating plants should review their pressure-temperature limits within a 3 year period after issuance of the final guide and revise them if necessary. Within that period, plants would be allowed to continue to follow the present schedule for updating their pressure-temperature limits. 1 1 1 i l l l l l 20 1
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