NRC Generic Letter 1988-11
UNITED STATES
LLNUCLEAR REGULATORY COMMISSION
- nWASHINGTON. D. C.20555 JUL 1 2 1988 TO ALL LICENSEES OF OPERATING REACTORS AND HOLDERS OF CONSTRUCTION PERMITS
SUBJECT: NRC POSITION ON RADIATION EMBRITTLEMENT OF REACTOR VESSEL MATERIALS
AND ITS IMPACT ON PLANT OPERATIONS (GENERIC LETTER 88-11 )
The purpose of this letter is to call your attention to the attached copy of Revision 2 to Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials," which became effective May 1988. It will be used by the NRC in re- viewing submittals regarding pressure-temperature (P-T) limits and for analyses other than pressurized thermal shock (PTS) that require an estimate of the embrittlement of reactor vessel beltline materials.
Licensees and permittees should use the methods described in Revision 2 to Regulatory Guide 1.99 to predict the effect of neutron radiation on reactor vessel materials as required by Paragraph V.A. of 10 CFR Part 50 Appendix G,
unless they can justify the use of different methods. The use of the Revi- sion 2 methodology may result in a modification of the pressure-temperature limits contained in Technical Specifications in order to continue to satisfy the requirements of Sec. V of 10 CFR Part 50, Appendix G. Within 180 days of the effective date of Revision 2, licensees should submit the results of their technical analysis and a proposed schedule for whatever actions they propose to take. In the event that such actions are necessary, their schedule is nego- tiable provided that all actions (hardware, procedures, and/or staff modifi- cations) are completed (fully implemented and operational) within 2 plant outages (approximately 3 years) after the effective date of Revision 2 to Regulatory Guide 1.99.
PWR licensees should note that the Low-Temperature-Overpressure Protection (LTOP) set points and enable temperatures, which are determined from the P-T
limits, may also have to be revised as d result of Revision 2. Since Revision 2, in general, results in a lowering of the Appendix G pressure curves and a shift to higher enable temperatures, the resulting narrowing of the operating window may restrict flexibility on heatup and cooldown operations.
Standard Review Plan 5.2.2, "Overpressure Protection," and the associated Branch Position RSB 5-2 is being changed to provide sone relief from this impact. Paragraph II.B, which requires protection "at low temperature," is being amended to define the required enable temperature for the LTOP system based on a fracture criterion. Automatic, or passive, protection of the upper end of the P-T limits will not be required but administratively controlled. At the lower end of the P-T limits, for example during startup, automatic protec- tion of the Appendix G P-T limits is still required for anticipated operational occurrences.
As plants age, it is expected that the operating window will continue to narrow and startup operations will become more difficult. Revision 2 accel- erates this narrowing of the operating window. Licensees are encouraged to 6
344 C-gecbQI>
(Rtx R l&0c0 oot
-2- JUL 121988 review system hardware and operating procedures to determine what changes could be made to reduce the likelihood of LTOP challenges. If changes can be implemented to demonstrate that the frequency of an LTOP event that would exceed Appendix G limits is expected to be much less than one per reactor life- time, then the staff would consider alternatives to Appendix G LTOP set points with appropriate justification of adequate safety from the standpoint of fracture prevention.
BWR licensees should note that the use of Revision 2 as the basis for P-T
limits for BWR pressure tests will require higher pressure test temperatures in many cases. The NRC does not accept the BWR Owners Group position that the margins given by following the procedures of Appendix G, 10 CFR Part 50 can safely be reduced.
With regard to the pressurized thermal shock issue in PWRs, the staff is presently considering an amendment to the PTS Rule, 10 CFR 50.61, that will replace the equations for RTP given in paragraph (b)(2) with the calculation procedure given in Section C.1 of Revision 2 to Reg. Guide 1.99, but will not change the screening criterion.
Based on calculations reported in the Regulatory Analysis, a number of reactor vessels will reach the screening criterion sooner, using Revision 2, and in a few cases that date will precede the end of license. To see if their plant falls in this category, licensees may wish to repeat the calculation of RT
values submitted to the NRC in response to the PTS Rule (January 23, 1986 EL
mittal) for the critical materials in the vessel beltline, using Section C.1 of Revision 2 to Regulatory Guide 1.99. The purpose of this suggestion is simply to provide early warning that further flux reduction should be con- sidered in some plants.
This request for information is covered by the Office of Management and Budget under Clearance Number 3150-0011, which expires December 31, 1989. Comments on burden and duplication may be directed to the Office of Management and Budget, Reports Management, Room 3208, New Executive Office Building, Washington, D.C. 20503.
Sincerely, Frank Jr.
Associate Director for Projects Office of Nuclear Reactor Regulation Enclosure:
Revision 2 to R.G. 1.99
Revision 2 U.S. NUCLEAR REGULATORY COMMISSION May 1988
) REGULATORY GUIDE
OFFICE OF NUCLEAR REGULATORY RESEARCH
REGULATORY GUIDE 1.99 ITask ME 305-4)
RADIATION EMBRITTLEMENT OF REACTOR VESSEL MATERIALS
A. INTRODUCTION tion requirements in 10 CFR Part SO have been cleared under OMB
Clearance No. 3150-0011.
General Design Criterion 31, "Fracture Prevention of Reactor Coolant Pressure Boundary," of Appendix A, "General Design B. DISCUSSION
Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," requires, in Sonme NRC requirements that necessitate calculation of radia- part, that the reactor coolant pressure boundary be designed with tion embrittlement are:
sufficient margin to ensure that, when stressed under operating, maintenance, testing, and postulated accident conditions, (1) the 1. Paragraph V.A of Appendix G requires the effects of neutron boundary behaves in a nonbrittle manner and (2) the probability radiation to be predicted from the results ofpertinent radiation effects of rapidly propagating fracture is minimized. General Design studies. This guide provides auch results in the form of calculative Criterion 31 also requires that the design reflect the uncertainties procedures that are acceptable to the NRC.
in determining the effects of irradiation on material properties.
Appendix G, "Fracture Toughness Requirements," and Appendix 2. Paragraph V.B of Appendix G describes the basis for setting H, "Reactor Vessel Material Surveillance Program Requirements," the upper limit for pressure as a function of temperature during which implement, in part, Criterion 31, necessitate the calculation heatup and cooldown for a given service period in terms of the of changes in fracture toughness of reactor vessel materials caused predicted value of the adjusted reference temperature at the end by neutron radiation throughout the service life. This guide describes of the service period.
general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlenent of the low-alloy steels 3. The definition of reactor vessel beltline given in Paragraph currently used for light-water-cooled reactor vessels. i.F of Appendix C requires identification of regions of the reactor vessel that are predicted to experience sufficient neutron radiation The calculative procedures given in Regulatory Position 1.1 of embrittlement to be considered in the selection of the most limiting this guide are ant the same as those given in tie Pressurized Thermal material. Paragraphs M.A and IV.A.1 specify the additional test Shock rule (§ 50.61, "Fracture Toughness Requirements for Pro- requirements for beltline materials that supplement th requirements for reactor vessel materials generally.
tection Against Pressurized Thermal Shock Events," of 10 CFR Part 50) for calculating RTpTS, the reference temperature that is to be compared to the screening criterion given in the rule. The 4. Paragraph KII of Appendix H incorporates ASTM E 185 information on which this Revision 2 is based may also affec the by reference. Paragraph 5.1 of ASTM E 185-82, "Standard Prac- basis for the PTS rule. The staff is presently considering whether tice for Conducting Sueillance Tests for Light-Water Cooled to propose a change to I 50.61. Nuclear Power Retor Vessels" (Re. 1), requires that the materials to be placed in surveillance be those that may limit operation of The Advisory Committee on Reactor Safeguards has been con- the reactor during its lifetime, i.e., those expected to have the highest sulted concerning this guide and has concurred in the regulatory adjusted reference temperature or the lowest Charpy upper-shelf position. energy at end of life. Both measures of radiation embrittlement must be considered. In Paragraph 7.6 of ASTM E 185-82, the require- Any infornation collection activities mertioned inthis regulatory ments for the number of capsules and the withdrawal schedule are based on the calculated amount of radiation emnbrittlement at end guide are contained as requirements in 10 CFR Part 50, which pro- vides the regulatory basis for this guide. The information collec- of life.
USNRC REGULATORY GUIDES The guides are Issued In the following ten broad divisions:
Regulatory Guides are Issued to describe and make available to the 6. Products public methods acceptable to the NRC staff of Implementing 1. Power Reactors tech- 2. Research and Test Reactors 7. Transportation specific parts of the Commission's regulations to delineate 3. Fuels and Materials Facilities 5L Occupational Health niques used by the taf in evaluating specific problems or postu- 4. Environmental and Siting 9. Antitrust and Financial Review ted accident or to provide guidance to applicants. Regulatory and compliance with 5 Materials and Plant Protection 10. General Guides aeno sbtitutes fo regulations,r them is not required. Methods and solutlons different from those Mst out In the guides will be acceptable If they provide a basis for the findings requisite to the issuance or continuance of a permit or Copies of Issued guides may be purchased from the Government license by the Commission. Printing Offic at the current GPO price information on current This guide was Issued after consideration of comments received from Documents U.S Govrnment Printitng Ofitce s tOffic Boor the public, Comments and suggestions for Improvements In these 3702, Washington DC 20013-7082 telephone (202)27-2060
guides are encouraged at all times, and guidesreflectwill be revised, as (202)275-2171 appropriate, to accommodate comments and to new Informa- tlon or experience. Issued guides may also be purchased from the National Technical Written comments may be submitted to the Rules and Procedures Information Service on a standing order baSIs Detals on this Branch DRR AOM, U.S. Nuclear Regulatory Commission, service may be obtained by writing NTIS. 5285 Port Royal Road, Washington, DC 20555. Springfield. VA 22161.
- *
The two measures of radiation embrituement used in this guide Guthrie's derived formulas (Ref. 2) are 28'F for welds and 17'P
are obtained from the results of the Charpy V-notch impact test. for base metal despite extensive efforts to find a model that reduced Appendix G to 10 CFR Part 50 requires that a full curve of absorbed the fitting error. Thus the use of surveillance data from a given energy versus temperature be obtained through the ductile-to-brittle reactor (in place of the calculative procedures given in this guide)
transition temperature region. The adjustment of the reference requires considerable engineering judgment to evaluate the credibil- temperature, ARTNDT, is defined in Appendix G as the tempera- ity of the data and assign suitable margins. When surveillance data ture shift in the Charpy curve for the irradiated material relative from the reactor in question become available, the weight given to that for the unirradiated material measured at the 30-foot-pound to them relative to the information in this guide will depend on the energy level, and the data that formed the basis for this guide were credibility of the surveillance data as judged by the following
30-foot-pound shift values. The second measure of radiation criteria:
embrittlement is the decrease in the Charpy upper-shelf energy level, which is defined in ASTM E 185-82. This Revision 2 updates the 1. Materials in the capsules should be those judged most likely calculative procedures for the adjustment of reference temperature; to be controlling with regard to radiation embrittlement according however, calculative procedures for the decrease in upper-shelf to the recommendations of this guide.
energy are unchanged because the preparatory work had not been completed in time to include them in this revision. 2. Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough The basis for Equation 2 for ARTNDT (in Regulatory Position to permit the determination of the 30-foot-pound temperature and
1.1 of this guide) is contained in publications by G. L. Guthrie (Ref. the upper-shelf energy unambiguously.
2) and G. R. Odette et al. (Ref. 3). Both of these papers used surveillance data from commercial power reactors. The bases for 3. When there are two or more sets of surveillance data from their regression correlations were different in tat Odette made one reactor, the scatter of ARTNDT values about a best-fit line greater use of physical models of radiation embrittlement. Yet, the drawn as described in Regulatory Position 2.1 normally should be two papers contain similar recommendations: (1) separate correla- less than 28°F for welds and 17WP for base metal. Even if the fluence tion functions should be used for weld and base metal, (2) the func- range is large (two or more orders of magnitude), the scatter should tion should be the product of a chemistry factor and a fluence factor, not exceed twice those values. Even if the data fail this criterion
(3) the parameters in the chemistry factor should be the elements for use in shift calculations, they may be credible for determining copper and nickel, and (4) the fluence factor should provide a trend decrease in upper-shelf energy if the upper shelf can be clearly deter- curve slope of about 0.25 to 0.30 on log-log paper at 10"' n/cm1 mined, following the definition given in ASTM E 185-82 (Ref. 1).
(E > I MeV), steeper at low fluences and flatter at high fluences.
Regulatory Position 1.1 is a blend of the correlation functions 4. The irradiation temperature of the Charpy specimens in the presented by these authors. Somne test reactor data were used as capsule should match vessel wall temperature at the cladding/base a guide in establishing a cutoff for the chemistry factor for low- metal interface within +/-25FP.
copper materials. The data base for Regulatory Position 1.2 is that given by Spencer H. Bush (Ref. 4). S. The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the data base for The measure of fluence used in this guide is the number of that material.
neutrons per square centimeter having energies greater than I million electron volts (E > I MeV). The differences in energy spectra at To use the surveillance data from a specific plant instead of the surveillance capsule and die vessel inner surface locations do Regulatory Position 1, one must develop a relationship of ARTNDT
not appear to be great enough to warrant the use of a damage finc- to fluence for that plant. Because such data are limited in number ton such as displacements per atom (dpa) (Ref. 5) in the analysis and subject to scatter, Regulatory Position 2 describes a procedure of the surveillance data base (Ref. 6). in which the form of Equation 2 is to be used and the fluence fac- tor therein is retained, but the chemistry factor is determined by However, the neutron energy spectrum does change significantly the plant surveillance data. Of several possible ways to fit such data, with location in the vessel wall; hence for calculating the attenua- the method that minimizes the sums of the squares of the errors tion of radiation embrittlement through the vessel wall, it is was chosen somewhat arbitrarily. Its use is justified in part by the necessary to use a damage function to determine ARTNDT versus fact that "least squares" is a common method for curve fitting.
radial distance into the wail. Th most widely accepted damage fimu- Also, when there are only two data points, the least squares method ton at this time is dpa, and the attenuation formula (Equation 3) gives greater weight to the point with the higher ARTNDT; this given in Regulatory Position 1.1 is based on the attenuation of dpa seems reasonable for fitting surveillance data, because generally through the vessel wall. the higher data point will be the more recent and therefore will repre- sent more modem procedures.
Sensitivity to neutron radiation embrittlement may be affected by elements other than copper and nickel. The original version and C. REGULATORY POSITION
Revision I of this guide had a phosphorus term in the chemistry factor, but the studies on which this revision was based found other 1. SURVEILLANCE DATA NOT AVAILABLE
elements such as phosphorus to be of secondary importance, i.e.,
including them in the analysis did not produce a significantly bet- When credible surveillance data from the reactor in question ter fit of the data. are not available, calculation of neutron radiation embrittlement of the beldine of reactor vessels of light-water reactors should be based Scatter in the data base used for this guide is relatively signifi- on the procedures in Regulatory Positions 1.1 and 1.2 within the cant, as evidenced by the fact that the standard deviations for limitations in Regulatory Position 1.3.
1.99-2
1.1 Adjusted Reference Temperature Here, al is the standard deviation for the initial RTNDT. If a measured value of initial RTNDT for the material in question is The adjusted reference temperature (ART) for each material in available, oI is to be estimated from the precision of the test method.
the beltline is given by the following expression: If not, and generic mean values for that class of material are used, ol is the standard deviation obtained from the set of data used to ART = Initial RTNDT + ARTNDT + Margin (I) establish the mean.
Initial RTNDT is the reference temperature for the unirradiated The standard deviation for ARTNDT, oh. is 28 F for welds and material as defined in Paragraph NB-2331 of Section HI1of the 17'F for base metal, except that oa need not exceed 0.50 times ASME Boiler and Pressure Vessel Code (Ref. 7). If measured values the mean value of ARTNDT.
of initial RTNDT for the material in question are not available, generic mean values for that class* of material may be used if there 1.2 Charpy Upper-Shelf Energy are sufficient test results to establish a mean and standard devia- tion for the class. Charpy upper-shelf energy should be assumed to decrease as a function of fluence and copper content as indicated in Figure 2.
ARTNDT is the mean value of the adjustment in reference Linear interpolation is permitted.
temperature caused by irradiation and should be calculated as follows: 1.3 Limitations ARTNDT = (CF) f(0.
28
- 0.10 log f) (2) Application of the foregoing procedures should be subject to the following limitations:
CF (1F) is the chemistry factor, a function of copper and nickel content. CF is given in Table I for welds and in Table 2 for base 1. The procedures apply to those grades of SA-302, 336, 533, metal (plates and forgings). Linear interpolation is permitted. In and 508 steels having minimum specified yield strengths of 50,000
Tables I and 2 "weight-percent copper" and "weight-percent psi and under and to their welds and heat-affected zones.
nickel" are the best-estimate values for the material, which will normally be the mean of the measured values for a plate or forging 2. The procedures are valid for a nominal irradiation temperature
0
or for weld samples made with the weld wire heat number that of 550'F. Irradiation below 525 F should be considered to pro- matches the critical vessel weld. If such values are not available, duce greater embrittlement, and irradiation above 590°F may be the upper limiting values given in the material specifications to which considered to produce less embrittlement. The correction factor used the vessel was built may be used. If not available, conservative should be justified by reference to actual data.
estimates (mean plus one standard deviation) based on generic data may be used if justification is provided. If there is no information 3. Application of these procedures to fluence levels or to cop- available, 0.35% copper and 1.0% nickel should be assumed. per or nickel content beyond the ranges given in Figure 1 and Tables
2 The neutron fluence at any depth in the vessel wall, f (10'"n/cm , I and 2 or to materials having chemical compositions beyond the E > I MeV), is determined as follows: range found in the data bases used for this guide should be justified by submittal of data.
f= fsurf (e -0.24x) (3)
2. SURVEILLANCE DATA AVAILABLE
2 where fsurf (10" n/cm , E > I MeV) is the calculated value of the neutron fluence at the inner wetted surface of the vessel at the When two or more credible surveillance data sets (as defined location of the postulated defect, and x (in inches) is the depth into in the Discussion) become available from the reactor in question, the vessel wall measured from the vessel inner (wetted) surface. they may be used to determine the adjusted reference temperature Alternatively, if dpa calculations are made as part of the fluence and the Charpy upper-shelf energy of the beldine materials as analysis, the ratio of dpa at the depth in question to dpa at the inner described in Regulatory Positions 2.1 and 2.2, respectively.
surface may be substituted for the exponential attenuation factor in Equation 3. 2.1 Adjusted Reference Temperature The fluence factor, f0 .
28
- 0.10 log f, is determined by calcula- The adjusted reference temperature should be obtained as tion or from Figure 1. follows. First, if there is clear evidence that the copper or nickel content of the surveillance weld differs from that of the vessel weld,
"Margin" is the quantity, OF, that is to be added to obtain con- i.e., differs from the average for the weld wire heat number servative, upper-bound values of adjusted reference temperature associated with the vessel weld and the surveillance weld, the for the calculations required by Appendix G to 10 CFR Part 50. measured values of ARTNDT should be adjusted by multiplying them by the ratio of the chemistry factor for the vessel weld to that for the surveillance weld. Second, the surveillance data should be fitted using Equation 2 to obtain the relationship of ARTNDT to Margin =2 V/ of + off (4) fluence. To do so, calculate the chemistry factor, CF, for the best fit by multiplying each adjusted ARTNDT by its corresponding fluence factor, summing the products, and dividing by the sum of wibe class for esimating itial RT DT is generally determined, for the welds the squares of the fluence factors. The resulting value of CF when with which this Wideisconcerned, by di type of welding flux (Linde 30 or other);
for base mewl, by the ASTM Standard Specification. entered in Equation 2 will give the relationship of ARTNDT to
1.99-3
TABLE I
0
CHEMISTRY FACTOR FOR WELDS. F
Wt-% 0 0.20 0.40 0.60 0.80 1.00 1.20
0 20 20 20 20 20 20 20
0.01 20 20 20 20 20 20 20
0.02 21 26 27 27 27 27 27
0.03 22 35 41 41 41 41 41
0.04 24 43 54 54 54 54 54
0.05 26 49 67 68 68 68 68
0.06 29 52 77 82 82 82 82
0.07 32 55 85 95 95 95 95
0.08 36 58 90 106 108 108 108
0.09 40 61 94 115 122 122 122
0.10 44 65 97 122 133 135 135
0.11 49 68 101 130 144 148 148
0.12 52 72 103 135 153 161 161
0.13 58 76 106 139 162 172 176
0.14 61 79 109 142 168 182 188
0.15 66 84 112 146 175 191 200
0.16 70 88 115 149 178 199 211
0.17 75 92 119 151 184 207 221
0.18 79 95 122 154 187 214 230
0.19 83 100 126 157 191 220 238
0.20 88 104 129 160 194 223 245
0.21 92 108 133 164 197 229 252
0.22 97 112 137 167 200 232 257
0.23 101 117 140 169 203 236 263
0.24 105 121 144 173 206 239 268
0.25 110 126 148 176 209 243 272
0.26 113 130 151 180 212 246 276
0.27 119 134 155 184 216 249 280
0.28 122 138 160 187 218 251 284
0.29 128 142 164 191 222 254 287
0.30 131 146 167 194 225 257 290
0.31 136 151 172 198 228 260 293
0.32 140 155 175 202 231 263 296
0.33 144 160 180 205 234 266 299
0.34 149 164 184 209 238 269 302
0.35 153 168 187 212 241 272 305
0.36 158 172 191 216 245 275 308
0.37 162 177 196 220 248 278 311
0.38 166 182 200 223 250 281 314
0.39 171 185 203 227 254 285 317
0.40 175 189 207 231 257 288 320
fluence that fits the plant surveillance data in such a way as to to obtain mean values of shift, ARTNDT. In calculating the margin, minimize the sum of the squares of the errors. the value of O, may be reduced from the values given in the last paragraph of Regulatory Position 1.1 by an amount to be decided To calculate the margin in this case, use Equation 4; the values on a case-by-case basis, depending on where the measured values given there for oA may be cut in half. fall relative to the mean calculated for the surveillance materials.
If this procedure gives a higher value of adjusted reference temperature than that given by using the procedures of Regulatory 2.2 Charpy Upper-Shelf Energy Position 1.1, the surveillance data should be used. If this procedure gives a lower value, either may be used. The decrease in upper-shelf energy may be obtained by plot- ting the reduced plant surveillance data on Figure 2 of this guide For plants having surveillance data that are credible in all respects and fitting the data with a line drawn parallel to the existing lines except that the material does not represent the critical material in as the upper bound of all the data. This line should be used in the vessel, the calculative procedures in this guide should be used preference to the existing graph.
1.99-4
2. Holders of licenses and permits should use the methods Technical Specifications in order to continue to satisfy the described in this guide to predict the effect of neutron radiation on requirements of Section V of Appendix G, 10 CFR Part 50.
reactor vessel materials as required by Paragraph V.A of Appen- dix G to 10 CFR Part 50, unless they can justify the use of dif- 3. The recommendations of Regulatory Position 3 are essen- ferent methods. The use of the Revision 2 methodology may result tially unchanged from those used to evaluate construction permit in a modification of the pressure-temperature limits contained in applications docketed on or after June 1, 1977.
1.99-6
TABLE 2 CHEMISTRY FACTOR FOR BASE METAL. VF
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
0.37 162 177 196 220 248 278 308
0.38 166 182 200 223 250 281 313
0.39 171 185 203 227 254 285 317
0.40 175 189 207 231 257 288 320
3. REQUIREMENT FOR NEW PLANTS D. IMPLEMENTATION
For beltline materials in the reactor vessel for a new plant, the The purpose of this section is to provide information to applicants content of residual elements such as copper, phosphorus, sulfur, and licensees regarding the NRC staff's plans for using this and vanadium should be controlled to low levels.* The copper con- regulatory guide. Except in those cases in which an applicant pro- tent should be such that the calculated adjusted reference temperature poses an acceptable alternative method for complying with specified at the 1/4T position in the vessel wall at end of life is less than portions of the Commission's regulations, the methods described
200°F. In selecting the optimum amount of nickel to be used, its in this guide will be used as follows:
deleterious effect on radiation embrittlement should be balanced against its beneficial metallurgical effects and its tendency to lower 1. The methods described in Regulatory Positions I and 2 of the initial RTNDT- this guide will be used by the NRC staff in evaluating all predic-
- For mmre infonnation, see the Appendix t ASMI Stadrd Specification A 533 tions of radiation embrittlement needed to implement Appendices (Ref. 8). G and H to 10 CFR Part 50.
1.99-5
REFERENCES
1. American Society for Testing and Materials, "Standard Prac- 5. American Society for Testing and Materials, "Standard Prac- tice for Conducting Surveillance Tests for Light-Water Cooled tice for Characterizing Neutron Exposures in Ferritic Steels in Nuclear Power Reactor Vessels," ASTM E 185-82, July 1982.* Terms of Displacements per Atom (DPA)," ASTM E 693-79, August 1979.*
2. G. L. Guthrie, "Charpy Trend Curves Based on 177 PWR Dam 6. W. N. McElroy, "LWR Pressure Vessel Surveillance Dosimetry Points," in"LWR Pressure Vessel Surveillance Dosimetry Im- Improvement Program: LWR Power Reactor Surveillance provement Program," NUREG/CR-3391, Vol. 2, prepared by Physics-Dosimetry Data Base Comnpendium," NUREG/
Hanford Engineering Development Laboratory, HEDL-TME CR-3319, prepared by Hanford Engineering Development
83-22, April 1984.A* Laboratory, HEDL-TME 85-3, August 1985.**
3. G. R. Odete et al., "Physically Based Regression Correlations 7. American Society of Mechanical Engineers, Section HI,
of Embrittlement Data from Reactor Pressure Vessel "Nuclear Power Plant Components," of ASME Boiler and Surveillance Programs," Electric Power Research Institute, Pressure Vessel Code. New York (updated frequently). tt NP-3319, January 1984.t
8. American Society for Testing and Materials, "Standard Specification for Pressure Vessel Plates, Alloy Steel, Quenched
4. S. H. Bush, "Structural Materials for Nuclear Power Plants," nd Tempered, Manganese-Molybdenum nd Manganese- in Journalof Testing and Ewluadon, American Society for Molybdenum-Nickel," ASTM A 533/A 533M-82, September Testing nd Materials, November 1974. 1982.*
- Copis may be obtaied from the American Society for Testing and Materials, 1916 Rae Ste, Phhidelphia, PA 19103.
- Copies may be obtained from the Superintendent of Documents, U.S. Government Pdintn Office, Past Office Box 37082. Wasingo, DC 20013-7082.
ICopies may be obtained from the Electric Power Research Institute, 3412 Hilview Avenue, Palo Alto, CA 94304.
ttCopics mny be obtained from the American Society of Mchanical Engines, 345 E. 47th Sateet, New York, NY 1O017.
1.99-7
2-0 . . . 111 I II I I I I 111IIIIIIIIIIIIIIIIIIIII I -iII.?.1 r III~IIIIIII~ II~ I
III~lIIIII~tII
- - l U.
10" 6 III INta t Fluence. n/cm2 (E > I May)
FIGURE I Fluence Factor for Use in Equation 2, the Expression for hRTNDT
60
50
40
- -1.
X,
.: -I
, ." I
rI.it
1+t]
% COPPER
i M211-t.
t
- + --1-j- I-, I
- - -t:t-17
- t I. -L-t I
-4-1-4- -- 1-44--
t-TiTT
!--'
71-1
, t1ATi
. I -if .z:
'It414;-44
-T
tiL
4trdt'l w-tlt H- +!H!
,
I
t;
itt bi BASE I S
30 II 41
,
.1 METAL WELDS !:>
! . -L I I i I ; I i iI i im I
0.35 0.30- mil1'41- ~-1iiWM4 Ell
- i Ii I l 1.lulull: _
-?
I
-i-I-
I
i _1a_; i, .,
HWi r _r
-+ . I,
- t 0.30 0.25- -i---i -tl+/-L-
'if-t I,&
-tiH- Ift1 , II I. I,
C
a)
4- - !
0.25 0.20- Ki
117- .
-+ 04-1-4.Li B- ii;R
1 11
44
-i+/-40I
iF
I Ir, LEE.F65$
r-1.
,I 1+++Ht RI HREE
'0.20 0.15- ;-, -H- I=
a-
02 t 0.15 0.10- 7m1II Tl 111111 T 1
. 11I1I Llll cn II 10.10 0.05- -
H., I. I 1 -I.t...
H! 1'.I I I I I I I1111111 I . .I.. I I I I 1 1111.11 I I ...
.111I.1I I 1111 . .1 I i I I I I II I I I I I! i I', T
w )
'a l U)
- I
0 tit:
TM t t.
4 T
4 :4 r;4 1-4 T-,r F-UWq 7TF;4:1'T-J. 74:-:i 4-4.-F-
+
44-1
2 X 1017 4 6 8 1018 2 4 6 8 1019 2 4 6
2 FLUENCE, n/cm (E> lMeV)
FIGURE 2 Predicted Decrease in Shelf Energy as a Function of Copper Content and Fluence
REGULATORY ANALYSIS
A copy of the regulatory analysis prepared for this Regulatory a fee at the Commission's Public Document Room at 1717 H Street Guide 1.99, Revision 2, is available for inspection and copying for NW., Washington, DC, under Regulatory Guide 1.99, Revision 2.
1.99-10