ML20132D328
| ML20132D328 | |
| Person / Time | |
|---|---|
| Issue date: | 04/18/1985 |
| From: | Minogue R NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| Shared Package | |
| ML20132C752 | List: |
| References | |
| RTR-REGGD-01.099, RTR-REGGD-1.099 NUDOCS 8509300140 | |
| Download: ML20132D328 (3) | |
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MEMORANDUM FOR: Victor Stello, Deputy Director for Regional Operations and Generic Requirements FROM:
Robert B. Minogue, Director Office of Nuclear Regulatory Research
SUBJECT:
CRGR REVIEW 0F PROPOSED REGULATORY GUIDE 1.99, REVISION 2. " RADIATION DAMAGE TO REACTOR VESSEL MATERIALS" The enclosed package is submitted in support of this request for CRGR review and approval to publish Revision 2 of Regulatory Guide 1.99 for public comment.
radiation damage trend Revision 2 (Enclosure 1) contains an update of the cIts provisions will affect any curves that have been in place since 1977.
plant whenever a calculation involving the fracture toughness of the Those calculations includei reactor vessel beltline is required.
pressure-temperature limits, analysis of transients such as low-temperature overpressurization, and the evaluation of flaws found in inservice inspection of the beltline.
Revision 2 incorporates the results of work begun for the NRC in 1981 at the time the pressurized thermal shock requirements were being developed, All aspects of as well as work done for EPRI in the same time period.
Revision 2 were developed in close cooperation between the Materials Engineering Branches in RES and NRR.
Background information about the implementation of Revision 2 for all plants and its impact on them is outlined in Enclosure 2 with references to e
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- MEETINGO79 i
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the other parts of the package required by the CRGR Charter. For further information on this review package contact P. N. Randall, Materials EngineeringBranch,RES,(X37711or28186).
Robert B. Minogue Office of Nuclear Regulatory Research
Enclosures:
- 1. Revision 2 of R.G. 1.99
- 2. Background Infonnation for CRGR review
- 3. Technical Basis
- 4. Changes to Standard Review Plan
- 5. Regulatory Analysis
- 6. PNL Value/ Impact Analysis
- 7. Cost Analysis Group's Analysis
- 8. Replacement Energy Cost Estimates by CAG e
1 l
l l
l 1
. . - 2 the other parts of the package required by the CRGR Charter. For further information on this review package contact P. N. Randall, Materials Engineering Branch, RES, (X37711 or 28186).
Robert B. Minogue Director Dffice of Nuclear Regulatory Research
Enclosures:
- 1. Revision 2 of R.G. 1.99
- 2. Background Information for CRGR review
- 3. Technical Basis
- 4. Changes to Standard Review Plan
- 5. Regulatory Analysis
- 6. PNL Value/ Impact Analysis .
- 7. Cost Analysis Group's Analysis AW .
- 8. Replacement Energy Cost A Estimates by CAG ME
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Enclosura Working Paper F 1
g May 21, 1985 Office of Nuclear Regulatory Research DRAFT REGULATORY GUIDE 1.99, REVISION 2 RADIATION DAMAGE TO REACTOR VESSEL MATERIALS j 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, l
" 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 Advisory Committee on Reactor Safeguards will be consulted concerning this guide.
B. DISCUSSION The principal examples of NRC requirements that necessitate calculation l of radiation damage are:
- 1. Paragraph V.A. of Appendix G requires: "The effects of neutron radiat, ion...are to be predicted from the results of pertinent radiation effect
- studies...." This guide provides such results in the form of calculative procedures that are acceptable to the NRC.
y 7 . , -
1 05/21/85 1bfp. 1 RG 1.99 REV 2
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- 2. Paragraph V.S. 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.
i
- 3. The definition of reactor vessel beltline given in Paragraph II.F.
of Appendix G requires identification of: "
... regions of the reactor vessel j that are predicted to experience sufficient neutron radiation damage to be I
considered in the selection of the most limiting material...." Paragraphs III.A.
l and IV. A.I. specify the additional test requirements for beltline materials j that supplement the requirements for reactor vessel materials generally.
i 4. Paragraph II.B. of Appendix H incorporates ASTM E185 by reference. s l 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
1.e., those expected to have the highest adjusted reference temperature or the l
lowest Charpy upper-shelf energy at end of life. Both measures of radiation l 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 l 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 l requires that a full curve of absorbed energy versus temperature be obtained i 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 j 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 tempera-ture; however, calculative procedures for the decrease in upper-shelf energy are unchanged, because the preparatory worn had not been completed in tint to 1
include them in Revision 2.
j i
i i 05/21/85 2 RG 1.99 REV 2 i
l The basis for equation (2) for ART NOT surface, given in Position C.1.a.12) ;i of this Guide, is contained in publications by G. L. Guthrie 2 and G. R. Odette.8 ,'
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 f factor, (3) the parameters in the chemistry factor should be the elements, l 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 101' n/cat (E>l MeV), steeper at low fluences and flatter at high fluences. Position C.I.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.s 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 atos (dpa)* in the analysis of the surveillance data base.5 2G. L. Guthrie, "Charpy Trend Curves Based on 177 PWR Data Points," from LWR Pressure Vessel Surveillance Dosimetry Improvement Program, Quarterly Prcgress Report April 1983 - June 1983, Hanford Engineering Development Laboratory, NUREG/CR-3391 Vol. 2, HEDL-TME 83-22.
nG. R. Odette and P. M. Lombrozo, " Physically Based Regression Correlatioas of Embrittlement Data From Reactor Fressure Vessel Surveillance Programs,'
EPRI NP-3319, Final Report, January 1984, Prepared for Electric Power Research Institute.
sSpencer H. Bush, " Structural Materials for Nuclear Power Plants," 1974 ASTM Gillett Memorial Lecture, published in ASTM Journal of Testing and Euluation, November 1974, and its addendum, " Radiation Damage in Pressure Vessel Steels for Commercial Light-Water Reactors."
4 ASTM E 693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements Per Atom (dpa)."
- W.N. McElroy, Editor, " LWR Power Reactor Surveillance Physics - Dosimetry Data Base Compendium " NUREG/CR 3319. HEDL TME 84-2 March 1984.
05/21/85 3 RG 1.99 REV 2
l n
J 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 NDT versus radial distance into the wall. The most widely accepted damage 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 i than copper and nickel. Revisions 0 and 1 of this guide had a phosphorus ters !
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 j the data. i l 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 l
- f. ,
fomulas are 28'F for welds and 17'F for base metal, despite extensive statis- l tical analysis. Thus, the use of surveillance data from a given reactor (in l place of the calculative procedures given in this guide) requires considerable engineering judgment to evaluate the credibility of the data and assign suitable 1 I margins. When surveillance data from the reactor in question become available, the weight given to them relative to the information in this guide should j 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 l
controlling with regard to radiation damage according to the provisions of this guide.
- 2. Scatter in the plots of Charpy energy versus temperature for the 1rradiated and unirradiated conditions should be small enough to permit the l
determination of the 30 ft-lb temperature and the upper shelf energy unambig-l uously.
l i
, 3. When there are two or more surveillance data from one reactor, the f scatter of ART NDT values about a best fit line drawn as described in Posi-j tion C.2.a. 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) i 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 h
05/21/85 4 RG 1.99 REV 2
- ?
,s
l 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 NDT to fluence, it was deemed advisable (because of scatter) to determine 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 ARTMDT; which seems j reasonable for fitting surveillance data, because generally that datum will be 1 the more recent one and therefore will represent more modern procedures.
l C. REGULATORY POSITION )
l
- 1. SURVEILLANCE DATA NOT AVAILABLE l 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.
- a. The adjusted reference temperature (ART) for each material in the beltline is given by the following expression:
]
ART = Initial RTNDT + ARTNDT + Margin (1) l (1) " Initial RTNDT" is the reference temperature for the j unirradiated material as defined in Paragraph NB-2331 of Section III of the i ASME Boiler and Pressure Vessel Code. In cases where measured values of Initial RT NDT for the material in question are not available, generic mean values 05/21/85 5 RG 1.99 REV 2
for that class' of material say be used if there are sufficient test results to establish a mean and standard deviation for the class. Additional guidance for the estimation of initial RT NOT is given in W Sundard Review Plan, NUREG-0800, Section 5.3.2.
(2) " ARTET " is the mean value of the adjustment in reference
- temperature caused by irradiation and should be calculated as follows
I ART surface = [CF]f(0.28-0.10 log f) (2)
MOT l
) 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 nomally 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 1.0K 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>l MeV) divided by 1018 ,
The fluence factor, f0 .28 - 0.10 log f, is determined by calculation or from Figure 1.
To calculate ART NOT at any depth, (e.g. , at 1/4T or 3/4T), the following l attenuation formula should be used:
I f ARTNOT = [ARTNDT surface) e (3) l 4
j 'For the welds with which this guide is concerned, for estimating Initial
- RTNOT, class is generally determined by the welding flux; for base metal, by the ASTM Standard Specification.
l l
l l
05/21/85 6 RG 1.99 REV 2 l
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where "x" (in inches) is the depth into the vessel well 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) l If a measured value of Initial RTET for the material in question is l used, og may k taken as zero. If a pneric value of Initial RT 18 "d'
{ MDT ogshould be obtained from the same set of data (see paragraph C.1.a.(1)).
l The standard deviations for ART ET ' "'A" , are 28'F for welds and 17'F for base metal, except o gneed not exceed 0.50 times the mean value of ART ET surface.
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 l interpolation is permitted.
j c. Application of the foregoing procedures should be subject to
! the following limitations:
l (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
{ damage, and irradiation above 590*F may be considered to produce less damage.
i The correction factor used should be justified by reference to actual data.
l (3) Application of these procedures to fluence levels or to j 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.
(
i 2. SURVEILLANCE DATA AVAILABLE i
i When two or more credible surveillance data as defined in the Discussion,
$ Section 8, 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.
j 05/21/85 7 RG 1.99 REV 2 i
t____.__. ___ . _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ . _ _ __ _ _
- a. The adjusted reference temperature should be obtained by first t fitting the surveillance data using Equation 2, paragraph C.1.a.(2), to obtain the relationship of ART NDT surface to fluence. To do so, calculate the cMm-1stry factor, "CF," for the best fit as follows. Multiply each measured ART by its corresponding fluence factor, sum the products and divide by the MDT sum of the squares of the fluence factors. The resulting value of CF when entered in Equation 2 will give the relationship of ARTNDT ""'I*C' ** II"'"C' 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 C.1.a.(3), except the values given there for og may be cut in half.
If this procedure gives a higher value of adjusted reference temperature than that given by using the procedures of paragraph C.1.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 be controlled to low levels. The copper content should be such that the calcula,ter, 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 utilizing this regulatory guide. i Except in t',ose cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, th po.litions described in this guide will be used by the NRC staff as follows:
05/21/85 8 RG 1.99 REV 2 l J
c
- 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 General Design Criterion 31 or as called for in Appendices G and H to l 10 CFR Part 50 submitted on or after (60 days after publication); however, if l 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 fom, 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.I. 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 confom 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 for whom the allowable operating period has been extended, when judged by the criteria of Revision 2, should submit the appropriate revision j to their Technical Specifications no later than 90 days prior to the expiration l of their current operating period.
- 3. The recommendations of regulatory position C.3 are unchanged from those used to evaluate construction pemit applications docketed on or after June 1, 1977.
i 05/22/85 9 RG 1.99 REV 2 4
_ _ _ - . _ . , . _ - . . . - , _ ~ _, _ _ _ _ -
f TABLE I ,
CHEMISTRY FACTOR FOR WELDS, 'F Copper, Nickel, Wt. %
Wt. X 0 0.20 0.40 0.60 0.80 1.00 1.20 0 20 20 20 20 20 to 20 0.01 20 20 20 20 20 20 20 l 0.02 '
21 26 27 27 27 27 27 0.03 22 35 41 41 41 41 41 0.04 24 43 54 54 64 54 54 0.05 26 49 67 68 68 68 68 0.06 29 52 77 82 82 82 82 0.07 32 55 85 95 95 95 95 0.08 36 58 90 106 108 108 108 0.09 40 61 94 115 122 122 122 0.10 44 65 97 122 133 135 135 0.11 49 68 101 130 144 148 148 0.12 52 72 103 135 153 161 161 0.13 58 76 106 139 162 172 176 0.14 61 79 109 142 168 182 188 0.15 66 84 112 146 175 191 200 0.16 70 88 115 149 178 199 211 0.17 75 92 119 151 184 207 221 0.18 79 95 122 154 187 214 230 0.19 83 100 126 157 191 220 238 0.20 88 104 129 160 194 223 245 .
0.21 92 108 133 164 197 229 252 0.22 97 112 137 167 200 232 257 0.23 101 117 140 169 203 236 263 0.24 105 121 144 173 206 239 268 ,
0.25 110 126 148 176 209 243 272 1 0.26 113 130 151 180 212 246 276 0.27 119 134 155 184 216 249 280 0.28 122 138 160 187 218 251 284 0.29 128 142 164 191 222 254 287 0.30 131 146 167 194 225 257 290 0.31 136 151 172 198 228 260 293 0.32 140 155 175 202 231 263 296 0.33 144 160 180 205 234 266 299 0.34 149 164 184 209 238 269 302 0.35 153 168 187 212 241 272 305 0.36 158 172 191 216 245 275 308 0.37 162 177 196 220 248 278 311 0.38 166 182 200 223 250 281 314 0.39 171 185 203 227 254 285 317 0.40 175 189 207 231 257 288 320 05/21/85 10 RG 1.99 REV 2
- _ - = - _ _ _ _ _ _ _ _ - - . - -
TABLE 11 CHEMISTRY FACTOR FOR BASE METAL, *F Copper, Nickel, Wt. X j Wt. E O 0.20 0.40 0.60 0.80 1.00 1.20 0 20 20 20 20 20 to 20 0.01 20 20 20 20 20 20 20 0.02 20 20 20 20 20 20 to 0.03 20 20 20 to to 20 20 0.04 22 26 26 26 26 26 26 0.05 25 31 31 31 31 31 31 O.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 i 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 1 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 1
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 i
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 05/21/85 11 RG 1.99 REV 2
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I il T- 1 i!R iid .T =,
h - -
h
~1 .
t : ib _
" .g.lpp_
4 :
r_ :-
- l'::1 J ?>-
c:
r r-
- . -~
i . .- l
. c -
- e 2 L -
1:
m J 4 :
- lt3 JE 4
l' i 1-
' }I
.}':L 4 - -- - - -
I' --- 4 '
kn r -
!h[$.
- 1}p
/ F .hl-- +
l - --.
~
_r c:t.
@m!4 ij c i
- ~
R oj I;.
-~.- - _ l ii 8
10 18 4 4 l 4 6 8 2 j 17 4 6 8 10 18 2 2 X 10 FLUENCE,n/cm2 (E > 1MeV) 1 I
Figure 2 Predicted Decrease in Shelf Enerw as a Function of Copper Content and Fluence.
1 1
1