Text

Rev 1 to Customer Designated Distribution Surry Units 1 & 2 Reactor Vessel Fluence & Ref Temp PTS Evaluations
	ML20210B480
Person / Time
Site:	Surry
Issue date:	04/30/1987
From:	Hirst C, Lau F, Meyer T; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML18150A074	List: ML20210B480;
References
	TAC-59988, TAC-59989, WCAP-11015, WCAP-11015-R01, WCAP-11015-R1, NUDOCS 8705050334
	Download: ML20210B480 (90)
	v • d • e

{{#Wiki_filter:_ WESTINGHOUSE CLASS 3 WCAP-11015 Revision 1 CUSTOMER DESIGNATED DISTRIBUTION SURRY UNITS 1 AND 2 REACTOR VESSEL FLUENCE AND RT EVALUATIONS PTS C. C. Heinecke V. A. Perone M. Weaver G. N. Wrights Work Performed for Virginia Power Company April 1987 l APPROVED: 2 N b 11 APPROVED: E L. 6 [ Gr.o.u). I. A. Meyer, benager F. L. Lau, Manager Structural Materials Radiation and Systems and Reliability Technology Analysis APPROVED: M 4JLat# 3 / C.'W. Hirst, Manager 7/ et, Reactor Coolant System Compenents Licensing Although the information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its Licensees without the customer's approval. WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR ENERGY SYSTEMS P. O. BOX 355 'UENO N 8705050334 e79439 DR p ADOCK 05000280 PDR

l TABLE OF CONTENTS PAGE TABLE OF CONTENTS i LIST OF TABLES ii LIST OF FIGURES v I. INTRODUCTION 1 1.1 The Pressurized Thermal Shock Rule 1 1.2 The Calculation of RT 3 PTS II. NEUTRON EXPOSURE EVALUATION 5 11.1 Method of Analysis 5 11.2 Fast Neutron Fluence Results 8 III. MATERIAL PROPERTIES 35 III.1 Identification and Location of Beltline Region Materials 35 III.2 Definition and Source of Material Properties for All 35 Vessel Locations III.3 Summary of Plant-Specific Material Properties 36 IV. DETERMINATION OF RT VALUES FOR ALL BELTLINE 42 PTS REGION MATERIALS IV.1 Status of Reactor Vessel Integrity in Terms of RT 42 versus Fluence Results PTS IV.2 Discussion of Results 43 V. CONCLUSIONS AND RECOMMENDATIONS 48 VI. REFERENCES 50 VII. APPENDICES A. Power Distribution A-1 B. Weld Chemistry B-1 C. RT Values of Surry Units 1 and 2 Reactor Vessel C-1 EtlineRegionMaterials 2321s/0381s/04 tS8 710 j

1 LIST OF TABLES M 11.2-1 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 12 Pressure Vessel Inner Radius - 0* Azimuthal Angle 11.2-2 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 13 Pressure Vessel Inner Radius - 15' Azimuthal Angle 11.2-3 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 14 Pressure Vessel Inner Radius - 30' Azimuthal Angle 11.2-4 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 15 Pressure Vessel Inner Radius - 45' Azimuthal Angle II.2-5 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 16 15' Surveillance Capsule Center 11.2-6 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 17 25' Surveillance Capsule Center II.2-7 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 18 35' Surveillance Capsule Center 11.2-8 Surry Unit 1 Fast Neutron (E>1.0 MeV) Exposure at the 19 45* Surveillance Capsule Center 11.2-9 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 20 Pressure Vessel Inner Radius O' Azimuthal Angle 11.2-10 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 21 Pressure Vessel Inner Radius 15' Azimuthal Angle 11.2-11 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 22 Pressure Vessel Inner Radius 30' Azimuthal Angle II.2-12 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 23 Pressure Vessel Inner Radius 45' Azimuthal Angle !!.2-13 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 24 15' Surveillance Capsule Center 11.2-14 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 25 25' Surveillance Capsule Center 11.2-15 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 26 35' Surveillance Capsule Center 11.2-16 Surry Unit 2 Fast Neutron (E>1.0 MeV) Exposure at the 27 45' Surveillance Capsule Center 2321:44C34?10 jj

LIST OF TABLES (Continued) Page 111.3-1 Surry Unit 1 Reactor Vessel Beltline Region Material 38 Properties 111.3-2 Surry Unit 2 Reactor Vessel Beltline Region Material 39 Properties IV.1-1 RT Values for Surry Unit 1 44 PTS IV.1-2 RT Values for Surry Unit 2 45 PTS A-1 Surry Unit 1 Beginning-of-Cycle and End-of-Cycle Fuel A-3 Assembly Burnups A-2 Surry Unit 2 Beginning-of-Cycle and End-of-Cycle Fuel A-5 Assembly Burnups A-3 Surry Unit 1 Core Power Distributions Used in the Fluence A-7 Analysis A-4 Surry Unit 2 Core Power Distributions Used in the Fluence A-9 Analysis B.1-1 Surry Unit 1 Intermediate and Lower Shell Longitudinal B-2 Welds Chemistry From WOG Materials Data Base - Wire Heat Number 8T1554 i B.1-2 Surry Unit 1 Lower Shell Longitudinal Weld Chemistry From B-3 WOG Materials Data Base Wire Heat 299L44 B.1-3 Surry Unit 1 Beltline Circumferential Weld and Surry Unit 2 B-7 Intermediate Shell Longitudinal Weld Chemistry From WOG Materials Data Base - Wire Heat Number 72445 B.1-4 Surry Unit 2 Lower Shell Longitudinal Weld Chemistry From B-9 WOG Material Data Base - Wire Heat Number 8T1762 B.1-5 Surry Unit 2 Beltline Circumferential Weld Chemistry From B-10 WOG Materials Data Base - Wire Heat Number 0227 C.1-1 RTog ValuesforSurryUnit1ReactorfesselBeltline C-3 i 2 Rebidn Materials 9 Fluence = 1.0 x 10 n/cm RT}5nMaterials9 Fluence =5.0x10ValuesforSurryUnit1ReactorfesselBeltline C-4 C.1-2 g 2 Reh n/cm C.1-3 RT ValuesforSurryUnit1 Reactor 1gesselBeltline C-5 2 RehIbnMaterials9 Fluence =1.0x10 n/cm ,,,,.4,....,,,,,, s3,

LIST OF TABLES (Continued) Pace C.1-4 RT Values for Surry Unit 1 Reactor Vessel Beltline C-6 RehI5n Materials @ Current (7.4 EFPY) Fluence Values C.1-5 RT Values for Surry Unit 1 Reactor Vessel Beltline C-7 RehI5nMaterials@LicenseExpiration(28.8EFPY) C.2-1 RT ValuesforSurryUnit2 Reactor 1gesselBeltline C-8 RehI5n Materials @ Fluence = 1.0 x 10 n/cm2 C.2-2 RT Values for Surry Unit 2 Reactor gessel Beltline C-9 RehI5n Materiais @ Fluence = 5.0 x 101 2 n/cm g C.2-3 RT ValuesforSurryUnit2ReactorgesselBeltline C-10 J RehIq5n Materials @ Fluence = 1.0 x 10 n/cm 1 2 C.2-4 RT Values for Surry Unit 2 Reactor Vessel Beltline C-11 RehI5n Materials @ Current (7.6 EFPY) Fluence Values C.2-5 RT Values for Surry Unit 2 Reactor Vessel Beltline C-12 RehIbn Materials @ License Expiration (29.4 EFPY) sw.mw.ww so

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LIST OF FIGURES PAGE 11.1-1 Surry Reactor Geometry 28 11.2-1 Surry Unit 1 Maximum Fast Neutron (E>1.0 MeV) Fluence 29 at the Beltline Weld Locations as a Function of Full Power Operating Time !!.2-2 Surry Unit 2 Maximum Fast Neutron (E>1.0 MeV) Fluence 30 at the Beltline Weld Locations as a Function of Full Power Operating Time 11.2-3 Surry Unit 1 Maximum Fast Neutron (E>1.0 MeV) Fluence 31 at the Pressure Vessel Inner Radius as a function of Azimuthel Angle 11.2-4 Surry Unit 2 Maximum Fast Neutron (E>1.0 MeV) Fluence 32 at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle 11.2-5 Surry Units 1 and 2 Relative Radial Distribution 33 ( of Fast Neutron (E>1.0 MeV) Flux and Fluence Within [ the Pressure Vessel Wall ( 11.2-6 Surry Units 1 and 2 Relative Axial Distribution of 34 Fast Neutron (E>1.0 MeV) Flux and Fluence Within ths Pressure Vessel Well III.1-1 Identification and Location of Beltline Region Material 40 for the Surry Unit Reactor Vessel III.1-2 Identification and Location of Beltline Region Material 41 for the Surry Unit Reactor Vessel IV.1-1 Surry Unit 1 - RT Curves per PTS Rule Method (1) 46 DocketedBaseMatefkalandWOGDataBaseMeanWeldMaterial Properties IV.1-2 Surry Unit 2 - RT Curves per PTS Rule Method (1) 47 DocketedBaseMat$fkalandWOGDataBaseMeanWeldMaterial Properties A-1 Surry Units 1 and 2 Core Description for Power A-11 Distribution Map mwo.euno y

SECTION 1 INTRODUCTION The purpose of this report is to submit the reference temperature for pressurized thermal shock (RTPTS) values for the Surry Units 1 and 2 reactor vessels to address the Pressurized Thermal Shock (PTS) Rule. Section I discusses the Rule and provides the methodology for calculating RT PTS

Section 11 presents the results of the neutron exposure evaluation assessing the effects that past and present core management strategies have had on neutron fluence levels in the reactor vessel. Section III provides the reactor vessels beltline region material properties for both units. Section IV provides the RT calculations from present through the projected PTS end-of-license fluence values.

1.1 THE PRESSURIZED THERMAL SHOCK RULE The Pressurized Thermal Shock (PTS) Rule [1] was approved by the U.S. Nuclear Regulatory Commissioners on June 20, 1985, and appeared in the Federal Register on July 23, 1985. The date that the Rule was published in the Federal Register is the date that the Rule became a regulatory requirement. The Rule outlines regulations to address the potential for PTS events on pressurized water reactor (PWR) vessels in nuclear power plants that are operated with a license from the United States Nuclear Regulatory Commission (USNRC). PTS events have been shown from operating experience to be transients that result in a rapid and severe cooldown in the primary system coincident with a high or increasing primary system pressure. The PTS concern arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Such sn event may produce the propagation of flaws postulated to exist near the inner wall surface, thereby potentially affecting the integrity of the vessel. sm.+enne i

The Rule establishes the following requirements for all domestic, operating PWRs: Establishes the RT (measure of fracture resistance) Screening Criterion for the N$ctor vessel beltline region 270*F for plates, forgings, axial welds 300'F for circumferential weld materials 6 Months From Date of Rule: All plants must submit their present vaNhs at the expiration date of the operating license. values (per the pres RT The d$ this submittal must be received by the NRC for plants with operating licenses is January 23, 1986. 9 Months From Date of Rule: Plants projected to exceed the PTS Screening Criterion shall submit an analysis and a schedule for implementation of such flux reduction programs as are reasonably practicable to avoid reaching the Screening Criterion. The data for this submittal must be received by the NRC for plants with operating licenses by April 23, 1986. Requires plant-specific PTS Safety Analyses before a plant is within 3 years of reaching the Screening Criterion, including analyses of alternatives to minimize the PTS concern. Requires NRC approval for operation beyond the Screening Criterion. For applicants of operating licenses, values of the projected RTPTS are to be provided in the Final Safety Analysis Report. This requirement is added as part of 10CFP. Part 50.34. In the Rule, the NRC provides guidance regarding the calculation of the toughness state of the reactor vessel materials - the " reference temperature for nil ductility transition" (RTNDT). For purposes of the Rule, RTNDT is now defined as "the reference temperature for pressurized thermal shock" Each (RTPTS) and calculated as prescribed by 10 CFR 50.61(b) of the Rule. values from the time of USNRC licensed PWR must submit a projection of RTPTS the submittal to the license expiration date. This assessment must lie submitted within 6 months af ter the effective date of the Rule, on January 23, 1986, with updates whenever changes occur affecting projected values. The 2321s/040347.10 g

M calculation'must be made for each weld and plate, or forging, in the l'Eactor ~ values vessel beltline. The purpose of this report is to provide the RTPTS for Surry Units 1 and 2. I.2 THE CALCULATION OF RTPTS In the PTS Rule, the NRC Staff has selected a conservative and uniform method for determining plant-specific values of RTPTS at a given time. The prescribed equations in the PTS rule for calculating RTPTS are actually For the purpose of comparison with ore of several ways to calculate RTNDT. f r the reactor vessel must be the Screening Criterion, the value of RTPTS calculated for each weld and plate, or forging in the beltline region as given below. For each material, RT is the lower of the results given by PTS Equations 1 and 2. Equation 1: PTS = I + M + [-10.+ 470(Cu) + 350(Cu)(Ni)] f.270 0 RT Equation 2: RTPTS = I + M + 283 f where I = the initial reference transition temperature of the unirradiated material measured as defined in the ASME Code, NB-2331. If a measured value is not available, the following generic mean values must be used: 0*F for welds made with Linde 80 flux, and -56'F for welds made with Linde 0091, 1092 and 124 and ARCOS B-5 weld fluxes. l l l 2321s/08038710 3

M = the margin to be added to cover uncertainties in the values of initial RTNDT, c pper and nickel content, fluence, and calculation procedures. In Equation 1, M=48'F if a measured value of I was used, and M=59'F if the generic mean value of I was used. In Equation 2, M=0'F if a measured value of I was used, and M=34'F if the generic mean value of I was used. Cu and Ni = the best estimate weight percent of copper and nickel in the material. 1 f = the maximum neutron fluence, in units of 10 n/cm2 (E greater than or 19 equal to 1 MeV), at the clad-base-metal interface on the inside surface of the vessel at the location where the material in question receives the highest fluence for the period of service in question. 1 Note that the chemistry values given in equations 1 and 2 are best estimate values to be ' upper bound mean values. The margin, M, increases the RTPTS predictions. Thus, the mean material chemistry values are to be used when available so as not to compound conservatism. The basis for the Cu and Ni values used in the RTPTS calculations for Surry Units 1 and 2 are discussed in Section 111.2. i I b am,memno 4

SECTION II NEUTRON EXPOSURE EVALUATION This section presents the results of the application of Westinghouse derived adjoint importance functions to the calculation of the Surry Units 1 and 2 reactor vessel fluence for Virginia Power Company. The use of adjoint importance functions provides a cost effective tool to assess the effects that past and present core management strategies have had on neutron fluence levels in the reactor vessel. II.1 METHOD OF ANALYSIS A plan view of the Surry Units 1 and 2 reactor geometry at the core midplane is shown in Figure 11.1-1. Since the reactor exhibits 1/8th core symmetry only a 0*-45' sector is depicted. Eight irradiation capsules attached to the thermal shield are included in the design to constitute the reactor vessel surveillance program. The capsules are located at 45*, 55*, 65', 165*, 245', 285', 295*, and 305' relative to the reactor geometry flat at 0*. In performing the fast neutron exposure evaluations for the reactor geometry shown in Figure 11.1-1, two sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was utilized to provide baseline data derived from a design basis core power distribution against which cycle by cycle plant specific calculations can be compared. The second set of calculations consisted of a series of adjoint analyses relating the response of interest (neutron flux > 1.0 MeV) at several selected locations within the reactor geo.netry to the power distributions in the reactor core. These adjoint importance functions when combined with cycle specific core power distributions yield the plant specific exposure data for each operating fuel cycle. The forward transport calculation was carried out in R,0 geometry using the DOT discrete ordinates code [2] and the SAILOR cross-section library (3). The SAILOR library is a 47 group, ENDF/B-IV based data set produced specifically for light water reactor applications. Anisotropic scattering is treated with aP expansion of the cross-sections. An S angular quadrature was used. 3 6 zw,ioamn o 5

The design basis core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 3-loop plants. Inherent in the development of this design basis core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely that a single reactor would have a power distribution at the nominal +2e level for a large number of fuel cycles, the use of this design basis distribution is expected to yield somewhat conservative results. This is especially true in cases where low leakage fuel management has been employed. The design basis core power distribution data used in the analysis is provided in Appendix A of this report. The data listed in Appendix A represents cycle averaged relative assembly powers. cross-section The adjoint analyses were also carried out using the P3 approximation from the SAILOR library. Adjoint source locations were chosen at the center of each of the surveillance capsules as well as at positions along the inner radius of the pressure vessel. Again, these calculations were run in R,0 geometry to provide power distribution importance functions for the exposure parameter of interest (neutron flux > 1.0 MeV). Having the adjoint importance functions and appropriate core power distributions, the response of interest is calculated as: I (R,0,E) F (R,0,E) dE R dR de R =Ip /g /E p,g where: Response of interest (+ (E > 1.0 MeV)) at radius R and R = R,0 azimuthal angle 0. Adjoint importance function at radius R and azimuthal I I(R,0,E) = angle 0 for neutron energy group E. Full power fission density at radius R and azimuthal angle F(R,0,E) = e for neutron energy group E. m,$cnne 6

l The fission density distributions used reflect the burnup-dependent inventory of fissioning actinides, including U-235, U-238, Pu-239, and Pu-241. Core power distributions for use in the plant specific fluence evaluations for Surry Units 1 and 2 were derived from measured assembly and cycle burnups for each operating cycle to date of the two reactors. The specific power distribution data used in the analysis is provided in Appendix A of this report. The data listed in Appendix A represents cycle averaged relative assembly powers. Therefore, the adjoint results were in terms of fuel cycle averaged neutron flux which when multiplied by the fuel cycle length yields the incremental fast neutron fluence. The projection of reactor vessel fast neutron fluence into the future to the-expiration date of the operating license requires that a few key assumptions be made. Current neutron fluences, based on past core loadings, are defined as of September 30, 1985. The operating license for the Surry Units expires on June 25, 2008 (forty years after the construction permit was issued). This report includes fluence projections from September 30, 1985 to June 25,- 2008 using the cycle-averaged core power distribution of the current operating cycle (Cycle 8 for each Surry Unit) and an assumed future capacity factor of 80%. All fluence projections into the future reflect the low leakage fuel management strategy exemplified by the Cycle 8 core loadings. Finally, it has been assumed that the Surry cores will be uprated from 2441 MW to 2546 th MW at the beginning of Cycle 11. th The transport methodology, both forward and adjoint, using the SAILOR cross-section library has been benchmarked against the Oakridge National Laboratory (ORNL) Poolside Critical Assembly (PCA) facility as well as against the Westinghouse power reactor surveillance capsule data base (4). The benchmarking studies indicate that the use of SAILOR cross-sections and generic design basis power distributions produces flux levels that tend to be conservative by 7-22%. When plant specific power distributions are used with the adjoint importance functions, the benchmarking studies show that fluence predictions are within + 15% of measured values at surveillance capsule locations. 2321s/OdC387.10 7 i ~~ c. e

i 11.2 FAST NEUTRON FLUENCE RESULTS a Calculated fast neutron (E >1.0 MeV) exposure results for Surry Units 1 and 2 are presented in Tables 11.2-1 through II.2-16 and in Figures 11.2-1 through 11.2-6. Data is presented at several azimuthal locations on the inner radius of the pressure vessel as well as at the center of each surveillance capsule. In Tables 11.2-1 through 11.2-4 cycle-specifM maximum neutron flux and fluence levels at O',15*, 30', and 45' on the pressure vessel inner radius of Surry Unit 1 are listed for thf period of operation up to September 30, 1985, and projected to the expiration date of the operating license. Also presented are the design basis fluence levels predicted using the generic 3-loop core power distribution at the nominal + 2a level. Similar data for the center of surveillance capsules located at 15', 25', 35' and 45' are given in Tables 11.2-5 and 11.2-8, respectively. In addition to the calculated data given for the surveillance capsule locations, measured fluence data from previously withdrawn surveillance capsules are also presented for comparison with analytical results. In the case of Unit 1, capsules were removed from the 15* location at the end of cycle 1 and the 35' location at the end of cycle 4. Cycle-specific and design basis fast neutron flux and fluence data at the inner radius of the pressure vessel of Surry Unit 2 are given in Tables 11.2-9 through 11.2-12 for the period of operation up to September 30, 1985, and projected to the expiration date of the operating license. As in the case of Unit 1, data are presented for the O*,15*, 30', and 45* azimuthal angles. Evaluations of plant specific and design basis fluence levels at the four-surveillance capsule locations are given in Tables 11.2-13 and 11.2-16. For Unit 2, a surveillance capsule was removed from the 15' position following cycle 1. A dosimetry evaluation from this capsule withdrawal is listed in Table II.2-13. 2321:003e7.10 g

Several observations regarding the data presented in Tables II.2-1 thPough II.2-16 are worthy of note. These observations may be summarized as follows: 1. For both Surry units, calculated plant specific fast neutron (E > 1.0 MeV) fluence levels at the surveillance capsule center are in excellent agreement with measured data. The maximum difference between the plant specific calculations and the measurements is less than 8%. Differences of this magnitude are well within the uncertainty of the experimental results. 2. For both Surry units, the fast neutron (E > 1.0 MeV) flux incident on the pressure vessel during Cycle I was, on the average,14% less than predictions based on the design basis core power distributions. This result is consistent with the statement that the design basis power distributions produce flux levels that tend to be conservative by 7-22%. 3. The low leakage fuel management employed during cycle 8 of Surry Unit 1, which is used for projection into the future, has reduced the peak fast neutron flux (0* azimuthal position) on the pressure vessel by a factor of 1.47 relative to the design basis flux. (In subsequent discussions, factors of fast neutron flux reduction, defined as the ratio of the design basis flux to the cycle-specific flux, will be quoted.) The cycle 8 core loading produced flux reduction factors ranging from 1.45 to 1.61 at the other azimuthal locations. 4. In Surry Unit 2, the low leakage core loading used for projection into the future (cycle 8) yielded a flux reduction factor of 1.85 at the peak flux ) location and factors ranging from 1.35 to 1.59 at the remaining azimuths. 5. Comparing the flux reduction factors resulting from the cycle 8 low leakage core loadings in the Surry units, one observes differences that are attributable to the varying burnups of the fuel assemblies in peripheral locations (see burnup data in Appendix A). 2321s/040347 10 g

6. While the peak fast neutron flux location occurs at the 0* location in values both Surry units, the materials having the most limiting RTPTS (see Section IV) are not necessarily located in the peak flux. In Surry Unit 1, the limiting material is the circumferential weld which sees the peak flux at the 0* azimuthal position. On the other hand, the limiting material in Surry Unit 2 is a longitudinal weld which is situated at the 45* azimuth. Graphical presentations of the plant specific fast neutron fluence at key locations on the prossure vessel are shown in Figures 11.2-1 and 11.2-2 as a function of full power operating time for Surry Units 1 and 2, respectively. For both Units 1 and 2, pressure vessel data is presented for the O' location on the circumferential weld as well as for the 45* longitudinal welds (see Section 111.1). In regard to Figure 11.2-1 and 11.2-2, the solid portions of the fluence curves are based directly on the cycle-specific core loadings as of September 30, 1985. The dashed portions of these curves, however, involve a projection into the future. As mentioned in Section !!.1, the neutron flux average over cycle 8 of each Surry unit was used to project future fluence levels. It should be noted that implementation of a more severe low leakage pattern than that used in cycle 8 of each unit would act to reduce the projections of fluence at key locations. On the other hand, relaxation of the current low leakage patterns or a return to out-in fuel management would increase those projections. The RT assessment must be updated per 10CFR50.El(b)(1) PTS whenever, among other things, changes in core loadings significantly impact the fluence and RTPTS pr jections. In Figures 11.2-3 and 11.2-4, the azimuthal variation of maximum fast neutron (E > 1.0 MeV) fluence at the inner radius of the pressure vessel is presented as a function of azimuthal angle for Units 1 and 2, respectively. Data are presented for both current and projected expiration-of-operating-license conditions. In Figure II.2-5, the relative radial variation of fast neutron flux and fluence within the pressure vessel wall is presented. 10

Similar data showing the relative axial variation of fast neutron flux and fluence over the beltline region of the pressure vessel is shown in Figure 11.2-6. A three-dimersional description of the fast neutron exposure of the pressure vessel wall can be constructed using the data given in Figure 11.2-3 through II.2-6 along with the relation g(R, 0,Z) = e(0) F(R) G(2) Fast neutron fluence at location R, 0, Z within where: e (R,0,Z) = the pressure vessel wall e (0) Fast neutron fluence at azimuthal location 0 on = the pressure vessel inner radius from Figure 11.2-3 or 11.2-4 F (R) Relative fast neutron flux at depth R into the = pressure vessel from Figure 11.2-5 G (Z) Relative fast neutron flux at axial position Z from- = Figure 11.2-6 Analysis has shown that the radial and axial variations within the vessel wall are relatively insensitive to the implementation of low leakage fuel management schemes. Thus, the above relationship provides a vehicle for a reasonable evaluation of fluence gradients within the vessel wall. un,menno 11

~ - - TABLE 11.2-1 SURRY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 0* AZIMUTHAL ANGLE ") l J Beltline Region 2 Elapsed Cumulative Fluence (n/cm ) Irradiation Irradiation Avg. Flux Plant Desig b) Interval Time (EFPY) (n/cm -sec) Specific Basis 10 18 18 CY-1 1.1 5.03 X 10 1.70 x 10 2.02 x 10 10 18 18 CY-2 1.6 5.73 x 10 2.70 x 10 3.06 x 10 10 18 18 CY-2 2.3 5.22 x 10 3.87 x 10 4.39 x 10 10 18 18 CY-4 3.4 4.86 x 10 5.49 x 10 6.37 x 10 10 18 18 CY-5 4.6 4.40 x 10 7.10 x 10 8.54 x 10 10 19 19 CY-6 5.9 3.96 x 10 8.75 x 10 1.10 x 10 10 19 19 CY-7 6.8 5.91 x 10 1.05 x 10 .1.28 x 10 10 19 19 CY-8(9/30/85)(c) 7.4 4.05 x 10 1.13 x 10 1.39 x 10 10 19 19 9/30/85 - CY-10(d) 10.3 4.05 x 10 1.49 x 10 1.93 x 10 10 19 19 CY 5/25/2012(*) 28.8 4.22 x 10 3.96 x 10 5.55 x 10 Rev.1 (a) Applicable to the peak locations (0*, 90', 180*, 270') on the intermediate and lower shell plates and the intermediate to lower shell circumferential weld. 10 2 (b) Design basis fast neutron flux = 5.96 x 10 n/cm -sec at 2441 MWth i (c) 9/30/85 is the date at which the current neutron fluences are defined. (d) At the beginning of CY-11, the core thermal power will be uprated to 2546 1 MWth. Beyond 9/30/85 a 80% capacity factor is assumed. I (e) Exposure period from the onset of the uprating to the license expiration date. nu. omma 12

TABLE 11.2-2 SURRY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 15'-AZIMUTHAL ANGLE Beltline Region 2 Elapsed Cumulative Fluence (n/cm ) Irradiation Irradiation Avg.2 Flux Plant Desigg) Interval Time (EFPY) (n/cm sec) Specific Basis 10 17 17 CY-1 1.1 2.40 X 10 8.12 x 10 9.21 x 10 10 18 18 CY-2 1.6 2.72 x 10 1.29 x 10 1.40 x 10 10 18 18 CY-3 2.3 2.49 x 10 1.84 x 10 2.00 x 10 10 18 18 CY-4 3.4 2.34 x 10 2.62 x 10 2.91 x 10 10 18 18 CY-5 4.6 2.06 x 10 3.38 x 10 3.90 x 10 10 18 18 CY-6 5.9 1.88 x 10 4.16 x 10 5.04 x 10 10 18 18 CY-7 6.8 2.50 x 10 4.92 x 10 5.87 x 10 CY-8(9/30/85)(b) 7.4 1.88 x 10 5.28 x 10 6.38 x 10 10 18 18 9/30/85 - CY-10(c) 10.3 1.88 x 10 6.96 x 10 8.82 x 10 10 18 10 10 19 19 CY 5/25/2012(d) 28.8 1.97 x 10 1.84 x 10 2.54 x 10 Rev.1 10 2 (a) Design basis fast neutron flux = 2.72 x 10 n/cm -sec at 24a1 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined. (c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed. (d) Exposure period from the onset of the uprating to the license expiration date. un.conna 13

TABLE 11.2-3 SURRY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 30' AZIMUTHAL ANGLE Beltline Region 2 Elapsed Cumulative Fluence'(n/cm ) Irradiation Irradiation Avg.2 Flux Plant Desigg) Interval Time (EFPY) (n/cm -sec) Specific Basis 10 17 17 CY-1 1.1 1.30 X 10 4.40 x 10 5.38 x 10 10 17 17 CY-2 1.6 1.54 x 10 7.09 x 10 8.16 x 10 10 18 18 CY-3 2.3 1.34 x 10 1.01 x 10 1.17 x 10 10 18 18 CY-4 3.4 1.30 x 10 1.44 x 10 1.70 x 10 10 18 18 CY-5 4.6 1.09 x 10 1.84 x 10 2.28 x 10 10 18 18 CY-6 5.9 1.02 x 10 2.27 x 10 2.95 x 10 9 18 18 CY-7 6.8 9.80 x 10 2.56 x 10 3.43 x 10 9 18 18 CY-8 (9/30/85)(b) 7.4 9.86 x 10 2.75 x 10 3.73 x 10 9 18 19 9/30/85-CY-10(c) 10.3 9.86 x 10 3.63 x 10 5.15 x 10 CY 5/25/2012(d) 28.8 1.03 x 10 9.63 x 10 1.48 x 1019 l'tev.1 10 18 10 2 (a) Design basis fast neutron flux = 1.59 x 10 n/cm -sec at 2441 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined. (c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80*/. capacity factor is assumed. (d) Exposure period from the onset of the uprating to the license expiration date. i 2321s/038 e s/041587.10 g

T TABLE II.2-4 SURRY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 45' AZIMUTHAL ANGLE (a) Beltline Region 2 Elapsed Cumulative Fluence (n/cm ) Irradiation Irradiation Avg.2 Flux Plant Desig[b} Interval Time (EFPY) (n/cm -sec) Specific Basis 9 17 17 CY-1 1.1 8.59 X 10 2.91 x 10 3.27 x 10 10 17 17 CY-2 1.6 1.05 x 10 4.75 x 10 4.96 x 10 9 17 17 CY-3 2.3 9.08 x 10 6.78 x.10 7.12 x 10 9 17 18 CY-4 3.4 8.71 x 10 9.68 x 10 1.03 x 10 9 18 18 CY-5 4.6 7.11 x 10 1.23 x 10 1.39 x 10 9 18 18 CY-6 5.9 6.86 x 10 1.51 x 10 1.79 x 10 9 18 18 CY-7 6.8 6.14 x 10 1.70 x 10 2.08 x.10 CY-8 (9/30/85)(c) 7.4 6.54 x 10 1.82 x 10 2.26 x 10 9 18 18' 9/30/85 - CY-10(d) 10.3 6.54 x 10 2.41 x 10 3.13 x 10 9 18 18 I 9 18 18 CY 5/25/2012 *) 28.8 6.82 x 10 6.39 x 10 9.01 x 10 Rev.1 (a) Applicable to the longitudinal welds at 45*, 135', 225*, 315* in the peak axial flux. 9 2 (b) Design basis fast neutron flux = 9.66 x 10 n/cm -sec at 2441 MWth (c) 9/30/85 is the date at which the current neutron fluences are defined. (d) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed. (e) Exposure period from the onset of the uprating to the license expiration date. am,menne 15

TABLE 11.2-5 SURRY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 15* SURVEILLANCE CAPSULE CENTER Beltline Region 2 Elapsed Cumulative Fluence (n/cm ) Avg.$ lux Desig,} Capsule F Plant Irradiation Irradiation (r/c sec) Specific Bas Data Interval Time (EFPY) 10 18 18 19 (*) CY-1 1.1 8.31 X 10 2.81 x 10 3.19 x 10 2.89 x 10 10 18 18 CY-2 1.6 9.42 x 10 4.46 x 10 4.84 x 10 10 18 18 CY-3 2.3 8.61 x 10 6.39 x 10 6.95 x 10 10 18 19 CY-4 3.4 8.11 x 10 9.08 x 10 1.01 x 10 10 19 19 CY-5 4.6 7.08 X 10 1.17 x 10 1.35 x 10 10 19 19 CY-6 5.9 6.47 x 10 1.44 x 10 1.75 x 10 10 19 19 CY-7 6.8 8.76 x 10 1.70 x 10 2.03 x 10 10 19 19 CY-8 (9/30/85)(b) 7.4 6.46 x 10 1.83 x 10 2.21 x 10 10 19 19 9/30/85 - CY-10(c) 10.3 6.46 x 10 2.40 x 10 3.05 x 10 10 19 19 CY 5/25/2012(d) 28.8 6.74 x 10 6.34 x 10 8.79 x 10 gey,1 10 2 (a) Design basis fast neutron flux = 9.43 x 10 n/cm -sec at 2441 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined. i l (c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed. (d) Exposure period from the onset of the uprating to the license expiration date. (e) Reflects adjustments made to the spectrum-averaged reaction cross sections reported in Reference 5. l l zw.monne 16

TABLE 11.2-6 SURRY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 25' SURVEILLANCE CAPSULE CENTER Beltline Region 2 Elapsed Cumulative Fluence (n/cm ) Irradiation Irradiation Avg.2 Flux Plant Desig a) Interval Time (EFPY) (n/cm -sec) Specific Basis 10 18 18 CY-1 1.1 5.26 x 10 1.78 x 10 2.02 x 10 10 18 18 CY-2 1.6 6.14 x 10 2.85 x 10 3.07 x 10 10 18 18 CY-3 2.3 5.40 x 10 4.06 x 10 4,41 x 10 10 18 18 CY-4 3.4 5.24 x 10 5.81 x 10 6.40 x 10 10 19 18 CY-5 4.6 4.48 x 10 7.44 x 10 8.58 x 10 10 18 19 CY-6 5.9 4.15 x 10 9.17 x 10 1.11 x 10 10 19 19 CY-7 6.8 4.17 x 10 1.04 x 10 1.29 x 10 10 19 19 CY-8(9/30/85)(b) 7.4 4.02 x 10 1.12 x 10 _ 1.40 x 10 9/30/85 - CY-10(C) 10.3 4.02.< l'010 19 19 1.48 x 10 1.94 x 10 10 19 19 CY 5/25/2012(d) 28.8 4.20 x 10 3.93 x 10 5.58 x 10 Rev.1 10 2 (a) Design basis fast neutron flux = 5.98 x 10 n/cm -sec at 2441 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined. (c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed. (d) Exposure period from the onset of the uprating to the license expiration date. nmem.ie 17 l

TABLE 11.2-7 SURRY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 35' SURVEILLANCE CAPSULE CENTER Beltline Region. 2 Elapsed Cumulative Fluence (n/cm ) Avg.klux Desig,) Capsule F Plant l Irradiation Irradiation (n/c sec) Specific Basi -Data Interval Time (EFPY) 10 18 16 CY-1 1.1 3.56 X 10 1.21 x 10 1.38 x 10 10 18 18 CY-2 1.6 4.28 x 10 1.95 x 10 2.10 x 10 10 18 18 CY-3 2.3 3.71 x 10 2.78 x 10 3.02 x 10 10 18 18 18Te) CY-4 3.4 3.58 x 10 3.97 x 10 4.38 x 10 4.31 x 10 10 18 18 CY-5 4.6 2.93 X 10 5.05 x 10 5.87 x 10 10 18 18 CY-6 5.9 2.78 x 10 6.21 x'10 7.58 x 10 10 18 18 CY-7 6.8 2.58 x 10 6.99 x 10 8.82 x 10 10 18 18 CY-8(9/30/85)(b) 7.4 2.68_x 10 7.50 x 10 9.59 x 10 10 18 19 9/30/85 - CY-10(c) 10.3 2.68 x 10 9.90 x 10 1.33 x 10 CY 5/25/2012(d) 28.8 2.80 x 10 2.62 x 10 3.82 x 1019 l Rev.1 10 19 10 2 (a) Design basis fast neutron flux = 4.09 x 10 n/cm -sec at 2441 MWth (b) 9/30/85 is the date at which the current neutron fluences are dafined. (c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed. (d) Exposure period from the onset of the uprating to the license expiration date. (a) Reflects adjustments made to the spectrum-averaged reaction cross sections reported in Reference 6. l mi.+mno 18

TABLE 11.2-8 SURRY UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 45' SURVEILLANCE j' CAPSULE CENTER Beltline Region - 2 Elapsed Cumulative Fluence (n/cm ) Irradiation Irradiation Avg.2 Flux Plant Desigg,) Interval Time (EFPY) (n/cm -sec) Specific Basis 10 17 18 CY-1 1.1 2.79 x 10 9.45 x 10 1.09 x 10 10 18 18 CY-2 1.6 3.43 x 10 1.55 x 10 1.66 x 10 10 18 18 CY-3 2.3 2.95 x 10 2.21 x 10 2.38 x 10 10 18 18 j CY-4 3.4 2.83 x 10 3.15 x 10 3.46 x 10 10 10 18 CY-5 4.6 2.29 x 10 3.98 x 10 4.64 x 10 l CY-6 5.9 2.21 x 10 4.91 x 10 5.98'x 1010 10 18 10 18 18 j CY-7 6.8 1.97 x 10 5.50 x 10 6.97 x 10 10 18 18 CY-8 (9/30/85)(b) 7.4 2.10 x 10 5.90 x 10 7.58 x 10 9/30/85 - CY-10(c) 10.3 2.10 x 10 7.78 x 10 10 18 - 1.05 x 1019 10 19 19 CY 5/25/2012(d) 28.8 2.19 x 10 2.06 x 10 3.01 x 10 ney,1 10 2 (a) Design basis fast neutron flux = 3.23 x 10 n/cm -sec at 2441 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined. (c) At the beginning of CY-11, the core thermal power will be uprated to 2546 j MWth. Beyond 9/30/85 a 80% capacity factor is assumed. (d) Exposure period from the onset of the uprating to the license expiration date. 1 i sw. mauve 19

TABLE 11.2-9 SURRY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 0' AZIMUTHAL ANGLE (a) Beltline Region p Elapsed Cumulative Fluence (n/cm ) Irradiation Irradiation Avg.2 Flux Plant Desig b) Interval Time (EFPY) (r./cm -sec) Specific Basis 10 18 19 CY-1 1.2 4.96 X 10 1.84 x 10 2.21 x 10 10 18 18 CY-2 1.9 5.16 x 10 3.02 x'10 3.57 x 10 10 18 18 CY-3 2.7 4.97 x 10 4.20 x 10 4.99 x 10 10 18 18 CY-4 3.8 5.21 x 10 6.01 x 10 7.06 x 10 10 18 18 CY-5 4.9 4.35 x 10 7.56 x 10 9.18 x 10 10 18 19 CY-6 6.2 4.16 x 10 9.25 x 10 1.16 x 10 10 19 19 CY-7 7.4 4.67 x 10 1.10 x 10 1.38 x 10 10 19 19 CY-8 (9/30/85)(c) 7.6 3.22 x 10 1.12 x 10 1.43 x 10 10 19 19 9/30/85 - CY-10(d) 10.7 3.22 x 10 1.44 x 10 2.01 x 10 CY 1/29/2013 ') 29.4 3.36 x 10 3.43 x 10 5.69 x 1019l Rev.1 I 10 19 (a) Applicable to the peak locations (0*, 90',180', 270') on the intermediate and lower shell plates and the intermediate to lower shell circumferential weld. 10 2 (b) Design basis fast neutron flux = 5.96 x 10 n/cm -sec at 2441 MWth (c) 9/30/85 is the date at which the current neutron fluences are defined. (d) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed. (e) Exposure period frora the onset of the uprating to the license expiration date. i I m,,meur so gn

TABLE 11.2-10 SURRY UNIT 2 B ST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 15' AZIMUTHAL ANGLE Beltline Region 2 Elapsed Cumulative Fluence (n/cm ) Irradiatien Irradiation Avg.2 Flux Plant Desigg) Interval Time (EFPY) (n/cm -sec) Specific Basis 10 17 10 CY-1 1.2 2.37 X 10 8.81 x 10 1.01 x 10 10 10 18 CY-2 1.9 2.52 x 10 1.45 x 10 1.63 x 10 10 18 18 CY-3 2.7 2.49 x 10 2.05 x 10 2.28 x 10 10 18 18 CY-4 3.8 2.50 x 10 2.92 x 10 3.22 x 10 10 18 18 CY-5 4.9 2.06 x 10 3.65 x 10 4.19 x 10 10 18 18 CY-6 6.2 2.00 x 10 4.46 x 10 5.30 x 10 10 18 18 CY-7 7.4 2.10 x 10 5.25 x 10 6.32 x 10 CY-8(9/30/85)(b) 7.6 1.72 x 10 5.37 x 10 18 10 18 - 6.51 x 10 9/30/85 - CY-10(c) 10.7 1.72 x 10 7.06 x 10 9.17 x 10 10 10 18 l9lRev.1 CY 1/29/2013(d) 29.4 1.80 x 10 1.77 x 10 2.60 x 10 10 19 10 2 (a) Design basis fast neutron flux = 2.72 x 10 n/cm -sec at 2441 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined. (c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed. (d) Exposure period from the onset of the uprating to the license expiration date. am.Senno 21

TABLE 11.2-11 l l SURRY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 30' AZIMUTHAL ANGLE ^ Beltline Region 2 Cumulative Fluence (n/cm ) Elapsed Desigg Irradiation Irradiation Avg.2 Flux Plant Basis,) t -Interval Time (EFPY) (n/cm -sec) Specific 10 17 17 CY-1 1.2 1.30 X 10 4.81 x 10 5.91 x 10 10 17' 17 CY 1.9 1.41 x 10 8.01 x 10 9.53 x 10 10 - 18 1.33 x 10 8 1 CY 2.7 1.45 x 10 1.15 x 10 10 18 18 CY 3.8 1.38 x 10 1.63 x 10 1.88 x 10 10 18 18 CY-5 4.a 1.13 x 10 2.03 x 10 2.45 x 10 10 18 18 CY-6 6.2 1.14 x 10 2.49 x 10 3.10 x 10 10 10 18 I CY-7 7.4 1.07 x 10 2.89'x 10 3.69 x 10 10 18 18 CY-8 (9/30/85)(b) 7.6 1.00 x 10 2.96 x 10 3.80 x 10 10 18 18 9/30/85 - CY-10(c) 10.7 1.00 x 10 3.95 x 10 5.36 x 10 10 19 19 CY 1/29/2013(d) 29.4 1.05-x 10 - 1.02 x 10 1.52 x 10 Rev.1 l 10 2 (a) Design basis fast neutron flux = 1.59 x 10 n/cm -sec at 2441 MWth (b) 9/30/85 is the data at which the current neutron fluences are defined. (c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed. (d) Exposure period from the onset of the uprating to the license expiration date. 22 --.-,---,-,=w n- ~ p r -,,4

TABLE 11.2-12 SURRY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 45' AZIMUTHAL ANGLE (a) Beltline Region 2 Elapsed Cumulative Fluence (n/cm ) Irradiation Irradiation Avg.2 Flux Plant Desig[b) Interval Time (EFPY) (n/cm -sec) Specific Basis 9 17 17 CY 1.2 8.61 X 10 3.20 x 10 3.59 x 10 9 17 17 CY-2 1.9 9.74 x 10 5.41 x 10 5.79 x 10 9 17 17 CY-3 2.7 9.86 x 10 7.76 x 10 8.09 x 10 9 18 18 CY-4 3.8 9.20 x 10 1.10 x 10 1.14 x 10 9 18 18 CY-5 4.9 7.95 x 10 1.38 x 10 1.49 x 10 9 18 18 CY-6 6.2 7.70 x 10 1.69 x 10 1.88 x 10 9 19 18 CY-7 7.4 7.66 x 10 1.98 x 10 2.24 x 10 CY-8 (9/30/85)(c) 7.6 7.14 x 10 2.03 x 10 2.31 x 10 9 18 18 9/30/85-CY-10(d) 10.7 7.14 x 10 2.73 x 10 3.26 x 10 9 18 18 CY 1/29/2013(*) 29.4 7.44 x 10 7.14 x 10 9.22 x 1018lRev,l 9 18 (a) Applicable to the longitudinal welds at 45*, 135*, 225*,'315' in the peak axial flux. 9 2 (b) Design basis fast neutron flux = 9.66 x 10 n/cm -sec at 2441 MWth (c) 9/30/85 is the date at which the current neutron fluences are defined. (d) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed, i (e) Exposure period from the onset of the uprating to the license expiration date. l 23 l

TABLE 11.2-13 SURRY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 15' SURVEILLANCE CAPSULE CENTER Beltline Region 2 Elapsed Cumulative Fluence (n/cm ) F Plant -Capsule Avg.$ lux Desig$,) Irradiation Irradiation (n/c sec) Specific Basi Data Interval Time (EFPY) 10 18 18 18 (*) CY-1 1.2 8.21 X 10 3.05 x 10 3.50 x 10 3.01 x 10 10 18 18 CY-2 1.9 8.73 x 10 5.04 x 10 5.65 x 10 10 18 18 CY-3 2.7 8.63 x 10 7.09 x 10 7.89 x 10 10 19 19 ~ CY-4 3.8 8.67 x 10 1.01 x 10 1.12 x 10 10 19 19 CY-5 4.9 7.06 X 10 1.26 x 10 1.45 x 10 10 19 19 CY-6 6.2 6.87 x 10 1.54 x 10 1.84 x 10 10 19 19 CY-7 7.4 7.25 x 10 1.81 x 10 2.19 x 10 10 19 19 CY-8 (9/30/85)(b) 7.6 5.92 x 10 1.86 x 10 2.26 x 10 10 19 19 9/30/85 - CY-10(c) 10.7 5.92 x 10 2.43 x 10 3.18 x 10 10 19 19 CY 1/29/2013(d) 29.4 6.17 x 10 6.09 x 10 9.01 x 10 p,ey,1 10 2 (a) Design basis fast neutron flux = 9.43 x 10 n/cm -sec at 2441 MWth ] (b) 9/30/85 is the date at which the current neutron fluences are defined. (c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed. (d) Exposure period from the onset of the uprating to the license expiration date. .(e) Reflects adjustments made to the spectrum-averaged reaction cross sections reported in Reference 7. i j rmwo" 24

TABLE 11.2-14 SURRY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 25* SURVEILLANCE CAPSULE CENTER Beltlineilegion 2 Elapsed Cumulative Fluence (n/cm ) Irradiation Irradiation Avg.2 Flux Plant Desigg} Interval Time (EFPY) (n/cm -sec) Specific Basis 10 18 18 CY-1 1.2 5.24 X 10 1.95 x 10 2.22 x 10 10 18 18 CY-2 1.9 5.62 x 10 3.23 x 10 3.58 x 10 10 18 18 i CY-3 2.7 5.77 x 10 4.60 x 10 5.01 x 10 10 18. 18 CY-4 3.8 5.57 x 10 6.54 x 10 7.09 x 10 10 18 18 CY-5 4.9 4.53 x 10 8.14 x 10 9.21 x 10 10 19 19 CY-6 6.2 4.59 x 10 1.00 x 10 1.16 x 10 10 19 19 CY-7 7.4 4.29 x 10 1.16 x 10 1.39 x 10 10 19 19 CY-8 (9/30/85)(b) 7.6 3.98 x 10 1.19 x 10 1.43 x 10 9/30/85 - CY-10(c) 10.7 3.98 x 10 1.58 x 10 2.02 x 10 10 19 19 29.4 4.15 x 10 4.04 x 10 5.71 x 1019lpev.1 Id) 10 19 CY 1/29/2013 10 2 (a) Design basis fast neutron flux = 5.98 x 10 n/cm -sec at 2441 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined. (c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed. (d) Exposure period from the onset of the uprating to the license expiration date. i i 2321s/04C387 IC 25

TABLE 11.2-15 SURRY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 35' SURVEILLANCE CAPSULE CENTER Beltline Region 2 Elapsed Cumulative Fluence (n/cm ) Irradiation Irradiation Avg.2 Flux Plant Desigga) Interval Time (EFPY) (n/cm -sec) Specific basis 10 18 18 CY-1 1.2 3.55 X 10 1.32 x 10 1.52 x 10 10 18 18 CY-2 1.9 3.92 x 10 2.21 x 10 2.45 x 10 10 18 18 CY-3 2.7 4.03 x 10 3.17 x 10 3.42 x 10 10 18 18 CY-4 3.8 3.80 x 10 4.49 x 10 4.85 x 10 10 18 18 CY-5 4.9 3.14 X 10 5.61 x 10 6.30 x 10 10 18 18 CY-6 6.2 3.12 x 10 6.88 x 10 7.96 x 10 10 18 19 CY-7 7.4 3.01 x 10 8.01 x 10 9.50 x 10 10 18 18 CY-8 (9/30/85)(b) 7.6 2.81 x 10 8.20 x 10 9.79 x 10 10 19 19 9/30/85 - CY-10(c) 10.7 2.81 x 10 1.10 x 10 1.38 x 10 Id) 10 19 l9 29.4 2.93 x 10 2.83 x 10 3.91 x 10 Rev.1 CY 1/29/2013 10 2 l (a) Design basis fsst neutron flux = 4.09 x 10 n/cm -sec at 2441 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined. (c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed. (d) Exposure period from the onset of the uprating to the license expiration date. 2321s @ 0387 10 g

i TABLE 11.2-16 SURRY UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 45* SURVEILLANCE CAPSULE CENTER 8eltline Region 2 Elapsed Cumulative Fluence (n/cm ) Irradiation Irradiation Avg.2 Flux Plant Desigg) Interval Time (EFPY) (n/cm -sec) Specific Basis 10 18 18 CY-1 1.2 2.80 X 10 1.04 x 10 1.20 x 10 10 18 10 CY-2 1.9 3.17 x 10 1.76 x 10 1.94 x 10 10 18 10 CY-3 2.7 3.21 x 10 2.52 x 10 2.70 x 10 10 16 18 CY-4 3.8 2.99 x 10 3.56 x 10 3.83 x 10 10 18 18 CY-5 4.9 2.58 x 10 4.48 x 10 4.97 x 10 10 18 18 CY-6 6.2 2.49 x 10 5.49 x 10 6.29 x 10 10 18 18 CY-7 7.4 2.48 x 10 6.42 x 10 7.50 x 10 CY-8 (9/30/85)(b) 7.6 2.30 x 10 6.58 x 10 7.73 x 10 10 18 18 9/30/85 - CY-10(c) 10.7 2.30 x 10 8.84 x 10 1.09 x 10 10 18 19 CY 1/29/2013(d) 29.4 2.40 x 10 2.31 x 10 3.08 x 10 Rev.1 10 19 l9 10 2 (a) Design basis fast neutron flux = 3.23 x 10 n/cm -sec at 2441 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined. (c) At the beginning of CY-11, the core thermal power will be uprated to 2546 MWth. Beyond 9/30/85 a 80% capacity factor is assumed. (d) Exposure period from the onset of the uprating to the license expiration date. 2321s/043387 10 g7

16003 19 l l i l O' (MAJOR AXIS) 15' (CAPSULES V T & X,V)

ff 25' (CAPSULES S,Z,X & Y,W,U)

^ / / / / 35' (CAPSULES Y,W 6 Z,T) / Q 45' (CAPSULE U a S) I / AxxxxN, Q / PRESSURE VESSEL f vmmx i / t / l l W\\ THERMAL SHIELD / / / / / / / I / I / / ,I //g CORE BARREL y // /// BAFFLE -ll/ / / / REACTOR CORE 1// /// + UNIT I SURV. CAPSULE I.D. & UNIT 2 SURV. CAPSULE I.D. Figure 11.1-1. Surry Reactor Geometry 28

16003 20 1020 7 5 3 / / / CIRCUMFERENTIAL WELD, O' f / / / / 10 19 s c 7 W S 5 / ' w s' O / d / LONGITUDINAL WELDS, 45' p 3 / zo / x / y / W f lois ACTUAL 7

PROJECTED 5

3 LICENSE EXPIRATION 9/30/85 I I "I I I I 10 37 O 10 20 30 40 50 60 70 Rev.1 OPERATING TIME (EFPY) Figure 11.2-1. Surry Unit 1 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Beltline Weld Locations as a Function of Full Power Operating Time 29

16003-21 1020 7 5 3 /' / / CIRCUMFERENTIAL WELD, O' f / / / / / n 10 39 ~ c 7 ~ y 'a h 5 ,/ / LONGITUDINAL WELDS, 45' f / 7 Z / Ox / / D i z win d 10 18 ACTUAL 7

PROJECTED I

5 3 LICENSE EXPIRATION 9/30/85 I i 4 I i l 10 17 O IO 20 ~ 30 40 50 60 70 Rev.1 OPERATING TIME (EFPY) Figure 11.2-2. Surry Unit 2 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Beltline Weld Locations as a Function of Full Power Operating Time 30

16003 22 1020 7 ACTUAL 5

PROJECTED f

f 3 'ss C g \\ w \\ W N w s a 's% J b-io19 %'N z %s o s s tr 7 s g s~~- E 5 LICENSE EXPIRATION U1< Rev.1 b_ 3 9/30/85 10 38 i l l i O 10 20 30 40 50 60 70 AZIMUTHAL ANGLE (DEG.) Figure 11.2-3. Surry Unit 1 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle 31

16003 23 1020 7 ACTUAL 5

PROJECTED N

eu 3 ~~~'s N N s d \\ g g 'ss 5 '%s s~s% aJ 10 19 %'s z s's o N 7 s' a Rev.1 LLIz 5 LICENSE EXPIRATION b <L 3 9/30/85 10 18 I I I I I I O 10 20 30 40 50 60 70 AZIMUTHAL ANGLE (DEG.) Figure 11.2-4. Surry Unit 2 Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle 32

16o03-24 lO 7 5 3 199.39 W z i.O 204.90 0 2 d 7 N CLAD N 5 IR _J u. 3 I/M b 8 215.13 Z I 50.i 220.24 w E 7 3/4T 45* 5 O' 3 REACTOR VESSEL OR I I I, I I i O.01 195 199 203 207 211 215 219 223 RADIUS (cm) Figure 11.2-5. Surry Units 1 and 2 Relative Radial Distribution of Fast Neutron (E > 1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall 33

16003-25 I.O 7 5 3 w O b D O.I X 7 D d 5 Z 0 3 F w Z W> [ O.OI _- _J 7 W 5 3 CORE MIDPLANE O.001 -300 -200 -100 O 100 200 300 DISTANCE FROM CORE' MIDPLANE (cm) Figure 11.2-6. Surry Units 1 and 2 Relative Axial Variation of Fast Neutron (E > 1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall 34

SECTION III. MATERIAL PROPERTIES 4 For'.th'e RTPTS calculation, the best estimate copper and nickel chemical ~ composition of the reactor vessel beltline material is necessary. The material properties for the Surry Units 1 and 2 beltline region will be i presented in this section. III.1 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS The beltline region is defined by the Rule [1] to be "the region of-the 4 reactor vessel'(shell material including welds, heat affected zones, and t plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron irradiation damage to be considered in the l selection of the most limiting material with regard to radiation damage." Figures III.1-1 and 111.1-2 identify and indicate the location of all beltline { region materials for the Surry Units 1 and 2 reactor vessels. III.2 DEFINITION AND SOURCE OF MATERIAL. PROPERTIES FOR ALL VESSEL LOCATIONS Material property values for the shell plates, which have been docketed with the NRC in Reference 8, were derived from vessel fabrication test certificate results. The property data for the welds have also been docketed with the NRC t in Reference 8, however, the weld properties cannot be used in the same direct j manner as the properties for the plates. Fast neutron irradiation-induced changes in the tension, fracture, and impact properties of reactor vessel materials are largely dependent on chemical composition, particularly in the copper concentration. The' variability'in irradiation-induced property changes, which exists in general, is compounded - by the variability of copper concentr tion within the weldments. To address the variation in chemistry, Babcock & Wilcox (B&W) performed a reactor vessel beltline weld chemistry study of eight B&W vessels, including 2321s/04C34t10 35 . c., J.

e i Surry Units 1 an'd 2, and reported the results in BAW-1799 [9] for the Westinghouse Owners Group (WOG). The scope of the work included collecting existing sources of chemistry data, performing extensive chemical analysis on the available archive reactor vessel weldments, and developing predictive methods with the aid of statistical analyses to establish the chemistry of the i reactor vessel beltline weldments-in question. In addition to the B&W report BAW-1799, the WOG Reactor Vessel Beltline Region r I Weld Metal Data Base was used. The WOG data base, which was developed in 1984 j. and is continually being updated, contains information from weld qualification-l records, surveillance capsule reports, the B&W report BAW-1799, and the Materials Properties Council (MPC) and the NRC Mender MATSURV data bases. j. ,For each of the welds in the Surry Units 1 and 2 beltline region, a material data search was performed using the WOG data base. Searches were performed for materials having the identical weld wire heat number as those in the Surry vessels, but any combination of wire and flux was allowed. For all of the data found for a particular wire, the copper, nickel, phosphorous and silicon values were averaged and the standard deviations were calculated. Although phosphorous and silicon are not needed for the PTS Rule, they are provided for the sake of completeness. The information obtained from the data base searches is found in Appendix B. l 111.3

SUMMARY

OF PLANT-SPECIFIC MATERIAL PROPERTIES A summary of the pertinent chemical and mechanical properties of the beltline region plate and weld materials of the Surry Units 1 and 2 reactor vessels are respectively given in Tables III.3-1 and 111.3-2 along with the references for t this information. Although phosphorus is no longer used in'the calculation of RT with respect to the PTS rule [1], it is given for reference since it NDT is currently used in the Regulatory Guide 1.99, Revision 1 trend curve [10]. The initial RT value of 0*F, which is shown for all of the Surry_ Units 1 NDT and 2 reactor vessel beltline weldments made with Linde 80 flux, is the i 2321s/040387.10 36 i

generic mean value defined in the PTS rule [1] for welds made with Linde 80 flux. Grau Lo flux was used in manufacturing the circumferential weld in Surry Unit 2. The initial RT value of 0*F was estimated per the NRC NDT Standard Review Plan [11]. The data in Tables 111.3-1 and 111.3-2 are used to evaluate the RT values PTS for the Surry Unit 1 and 2 reactor vessels.. 4 mi.menno 37 i i

TABLE III.3-1 SURRY UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Cu Ni. P I' (Wt.%) (Wt.%) (Wt.%) (*F) Source Intermediate Shell Plate C4326-1: 0.11 0.55 0.008 10(a) - Ref. [8] Intermediate Shell Plate C4326-2: 0.11 0.55 0.008 0(a) Ref. [8] fa) p,f,(g) Lower Shell Plate C4415-1: 0.11 0.50 0.014 20 Lower Shell Plate C4415-2: 0.11 0.50 0.014 0(a) Ref. [8] Longitudinal Welds - Intermed. & Lower Shells L1, L3,' L4 Heat No. 8T1554, Linde 80 l Flux 8579: 0.18 0.63 0.014 0(b) WOG Material Data Base, BAW 1799 (9) Longitudinal Weld - Lower Shell L2 Heat No. 299L44, Linde 80, Flux 8596: i 0.35 0.67 0.014 0(b) WOG Material Data Base, BAW 1799~[9] Rev.1 Circumferential Weld - Intermed. to Lower Shell WOS, Heat No. 72445, Linde 80 Flux 8579/8632: 0.21 0.58 0.015 0(b) WOG Material Data Base BAW-1799 [9] Notes: Position MTEB hl (alue for this plate is estimated according to Branch The initial RT v (a) 11] The initial RT values for the welds are the generic mean value i (b) definedbytheEISrule[1]forweldswithLinde80 flux. g 222" * * "' 38 4

I TABLE 111.3-2 SURRY UNIT 2' REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Cu Ni P I (Wt.%) (Wt.%) (Wt.%) (*F) Source Intermediate Shell Plate C4339-1: 0.11 0.54 0.012 30(*) Ref.~[8] Intermediate Shell Plate C4208-2: 0.15 0.55 0.008 -30(a) Ref. [8] I Lower Shell Plate C4331-2: 0.12 0.60 0.009 10(*) Ref. [8]' Lower Shell Plate C4339-2: 0.11 0.54 0.012 10(a)' Ref. [8] l Longitudinal Welds - Intermed. Shell L4 and L3, Heat No. 72445, Flux 8597: 0.21 0.58 0.016 0(b) WOG Material Data Base Longitudinal Welds - Lower shell L1 and L2, Heat No. 8T1762, Flux 8597/8632: ID) 0.29 0.55 0.013 O WOG I Material i Data Base, BAW 1799 [9] Circumferential Weld - Intermed. to Lower Shell WO5, Heat No. 0227, Grau to LW320: 0.19 0.56 0.017 0(a) WOG l Material Data Base Notes: (a) The initial RT value for these plates and weld are estimated accordingtoBkdchPositionMTEB5-2(11] (b) The initial RT values for the welds are the generic mean value defined by the hIS rule [1] for welds with Linde 80 flux. N i i i 1 I' ' 2321s/0dG347.10 3g -n_mn..-, w -.-,.s m-~-

s-r-,

y e-g e w,-- n---- ,- - ~-- - --vp,w--,- m ,-,w--- , -.- -,,---- p

FIGURE III.1-1 3DENTIFICAT10N AND LOCATION OF BELTLINE RER10N MATERIAL FOR THE SURRY UNIT 1 REACTOR VESSEL CIRCUMFERENTIAL SEAMS VERTICAL $EAMS, 270' \\ L4 ' eU C4326-1 - WO6 t 9.0" ' I 45 CORE l" j l O' 180' CORE w C4326-2 L3 goo 144" 3f-CL _J L_ 19.7" 14 -WO5 270* i C4415-1 L2 15' WE O' 180* 48.3" C4415-2 L1 90' i l 40 i -- ---,---..,-- - -.,,... ~. - -, _ -.. - - - - - - - -... _ -,. _. - _ - - - - - -.

FIGURE III.1-2 3DENTIFICAT10N AND LOCATION OF BELTLINE REGION MATERIAL FOR THE SURRY UNIT 2 REACTOR VESSEL CIRCUMFERENTIAL SEAMS VERTICAL SEAMS. 270' I C4339-1 4 - WO6 9.0" A ? 45 CORE C4 O' 180' j g CORE 33 C4208-2 L3 144" 90* 3 CL u 19.7" mr-(( -WO5 I 270* C-4331-2 L2 l ? I f 15' WE O' I f 180*- 48.3" C4339-2 L----- L1 90* 41 .--------w------,-,-nw-,,v,-w w-,,,-----w ,m-~~ ---,,----,----,--w---,m--w--,-r,

SECTION IV DETERMINATION OF RT VALUES FOR ALL BELTLINE REGION MATERIALS PTS Using the methodology prescribed in Section I.2, the results of the fast neutron exposure provided in Section II, and the material properties discussed in Section III, the RT values for Surry Units 1 and 2 can now be PTS determined. IV.1 STATUS OF REACTOR VESSEL INTEGRITY IN TERMS OF RT VERSUS FLUENCE PTS RESULTS 1 values were generated for Using the prescribed PTS Rule methodology, RTPTS all beltline region materials of the Surry Units 1 and 2 reactor vessels as a function of several fluence values and pertinent vessel lifetimes. The tabulated results from the total evaluation are presented in Appendix C for j all beltline region materials for both units. j Figures IV.1-1 and IV.1-2 present the RT values for the limiting PTS longitudinal weld, circumferential weld and shell plate of the Surry Units 1 and 2 vessels in terms of RTPTS versus fluence

curves.

The curves in these figures can be used: o to provide guidelines to evaluate fuel reload options in relation to the NRC RT Screening Criterion for PTS (i.e., RT values can be PTS PTS readily projected for any options under consideration, provided fluence is known),and o to show the current (7.4 EFPY for Surry 1 and 7.6 EFPY for Surry 2), and end-of-license (28.8 EFPY for Surry 1 and 29.4 EFPY for Surry 2) RTPTS values using actual and projected fluence. Rev.1

The EFPY can be determined using Figure 11.2-1 for Unit 1 and Figure 11.2-2 for Unit 2.

mi,m:xno 42 l n.

values for all Table IV.1-1 and IV.1-2 provide a summary of the RTPTS beltline region materials for the lifetime of interest. IV.2 DISCUSSION OF RESULTS As shown in Figures IV.1-1 and IV.1-2, the welds are the governing locations for both reactor vessels relative to PTS. All the RT values remain below PTS the NRC' screening values for PTS using the projected fluence values through value at license expiration i the license expiration. The most limiting RTPTS is 269'F for the lower shell longitudinal weld L2 of Unit 1 and 225'F for the Rev.1 longitudin 1 welds in the lower shell of Unit 2. l i r i 4 a 1 mi acune 43 e ---,,----n.- ..a ,-.,,,,,.g..

iS1 TABLE IV.1-1 RT VALUES FOR SURRY UNIT 1 g73 RT Values (*F) PTS Present End-of-License Screening Location Vessel Material (7.4 EFPY) (28.8 EFPY) Criteria 1 Intermediate shell plate 123 149 270 C4326-1 2 Intermediate shell plate 113 139 270 C4326-2 3 Lower shell plate C4415-1 131 156 270 4 Lower shell plate C4415-2 111 136 270 5 Intermediate and lower shell 131 160 270 longitudinal welds L1, L3, L4 6 Intermediate to lower shell 195 249 300 circumferential weld WO5 3 7 Lower shell longitudinal 208 269 270 Rev.1 weld L2 i mi.4m.miur to 44 t

TABLE IV.1-2 RT VALUES FOR SURRY UNIT 2 FTS RT Values ~('F) PTS Present End-of-License Screening Location Vessel Material (7.4 EFPY) (29.4 EFPY) Criteria r 1 Intermediate shell plate C4339-1 142 165 -270 i i 2 Intermediate shell plate C4208-2 110 143 270 3 Lower shell plate C4331-2 132 .158 270 4 Lower shell plate C4339-2 122 145 270 1 5 Intermediate shell longitudinal 144 179 270' 4 welds L4,L3 f 6 Lower shell longitudinal weld 177 225 270 L1,L2 l 7 Intermediate to lower shell 179 222 300 i V

circumferential weld WO5 4

i i s mi.e.our io 45 i

FIGURE IV.1-1 SURRY UNIT 1 - RTPTS CURVES PER PTS RULE METHOD [1] DOCKETED BASE MATERIAL AND WOG DATA BASE MEAN WELD MATERIAL PROPERTIES 340 e Present (7.4 EFPY) E License Expiration (28.8 EFPY) 320 NRC SCREENING VALUE CIRCUMFERENTIAL WELDS NRC SCREENING VALUE ~ FOR PLATES AND LONGITUDINAL WELDS 260 { 240 LONGITUDINAL WELD

LIMITING g

o ---CIRCUMFERENTIAL WELD cn 220 F Q.p 200 180 160 140 lINITING PLATE 120 Rev.1 gg i 18 19 20 10 10 10 FLUENCE N/CMXX2 J

l 'l l i FIGURE IV.1-2 SURRY UNIT 2 - RTPTS CORVES PER PTS RULE METHOD [1] DOCKETED BASE MATERIAL AND WOG DATA BASE MEAN WELD MATERIAL PROPERTIES 340 _ e Present (7.6 EFPY) e License Expiration (29.4 EFPV) 320 - NRC SCREENING VALUE FOR CIRCUMFERENTIAL WELDS E ~ NRC SCREENING VALUE FOR PLATES AND LONGITUDINAL WELDS y 260 e f LIMITING LONGITUDINAL WELD CIRCLMFERENTIAL lELD w Cl_r 220 CC 2W 180 160 N LIMITING BASE PLATE 140 120 Rev.1 I@ 19 1020 18 10 10 FLUENCE.N/CMXX2

i i SECTION'V CONCLUSIONS AND RECOMMENDATIONS 1 i Calculations have been completed in order to submit RTPTS values for the Surry Units 1 and 2 reactor vessels in meeting the requirements of the NRC Rule for Pressurized Thermal Shock [1]. This work entailed a neutron exposure l evaluation and a reactor vessel material study in order to determine the RT values. PTS s Detailed fast neutron exoosure evaluations using plant specific cycle by cycle core power distributions and state-of-the-art neutron transport methodology have been completed for the Surry Units 1 and 2 pressure vessels. Explicit ) j calculations were performed for the first eight operating cycles of both units l as of September 30, 1985. For both units, projection of the fast neutron exposure beyond September 30, 1985 was based on continued implementation of low leakage fuel management similar to that employed during cycle 8 of each unit. I In regard to the low leakage fuel management already in place at the Surry 1 Units, the plant specific evaluations have demonstrated that for the cycle 8 case the peak fast neutron flux at the 0* azimuthal position has been reduced 4 by a factor of 1.47 in Unit 1 and a factor of 1.85 in Unit 2 relative to the flux based on the design basis core power distribution. 1 I This location represents the maximum fast neutron flux incident on the reactor j pressure vessel. At other locations on the vessel, as well as at the surveillance capsules, the impact of low leakage will differ from the data presented above. i It should be noted that significant deviations from the low leakage scheme ) i already in place will affect the exposure projections beyond the current operating cycle. A move toward a more severe form of low leakage (lower i relative power on the periphery) would tend to reduce the projection. On the t other hand, a relaxation of the loading pattern toward higher relative power umacune 48

i on the core periphery would increase the projections beyond those reported. As each future fuel cycle evolves, the loading patterns should be analyzed to determine their potential impact on vessel and capsule exposure. ^ I The fast neutron fluence values from the plant specific calculations have been compared directly with measured fluence levels derived from neutron dosimetry contained in surveillance capsules withdrawn from each of the Sorry Units. For Unit 1, the ratio of calculated to measured fluence values ranges from-t l' O.92 to 0.97 for the t/c capsule data points. The corresponding ratio for Unit 2 is 1.01 for the capsule removed from that reactor. This excellent agreement between calculation and measurement supports the use of this analytical approach to perform a plant specific evaluations for the Surry i reactors. d i l Material property values for the Surry Units 1 and 2 reactor vessel beltline l region components were determined. The pertinent chemical and mechanical f properties for the shell plates remain the same as those that have been docketed with the NRC in Reference 8. The weld material properties are obtained from the WOG Material Data Base, 4 i l Using the prescribed PTS Rule methodology, RTPTS values were generated for all beltline region materials of the Surry Units 1 and 2 reactor vessels as a ) function of several fluence values and pertinent vessel lifetimes. For both values remain below the NRC screening values reactor vessels, all the RTPTS i for PTS using the projected fluence exposure through the expiration date of I the operating license. The most limiting values at end-of-license (28.8 EFPY for Surry Unit 1 and 29.4 EFPY for Surry Unit 2) are 269'F and 225'F for the j j longitudinal weld in the lower shell for Unit 1 and the longitudinal welds in f the lower shell of Unit 2, respectively. i { The results in this report are provided to enable Virginia Power Company to I comply with the initial 6 months submittal requirements of the USNRC PTS Rule. i i m i. = m i. 49 i

SECTION VI REFERENCES 1. Nuclear Regulatory Commission, 10CFR Part 50, " Analysis of Potential Pressurized Thermal Shock Events," Federal Register, Vol. 50, No. 141, I July 23, 1985. 2. Soltesz, R. G., Disney, R. K., Jedruch, J. and Ziegler, S. L., " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation Vol. 5 - Two Dimensional, Discrete Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5, August 1970. 3. " SAILOR RSIC Data Library Collection DLC-76." Coupled, Self-Shielded, 47 l Neutron, 20 Gamma-Ray, P, Cross-Section Library for Light Water 3 Reactors. 1 4. Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology to be published. 5. "Surry Unit No. 1 Pressure Vessel Irradiation Capsule Program: Examination and Analysis of Capsule T," J. S. Perrin, et al., June 24, 1975. 6. "Suny Unit No. 1 Pressure Vessel Irradiation Capsule Program: Examination and Analysis of Capsule W," J. S. Perrin, et al., March 30, 1979. i L-7. "Surry Unit No. 2 Pressure Vessel Irradiation Capsule Program: i Examination and Analysis of Capsule X," J. S. Perrin, et al., September 2, 1975. I 8. Letter f om C. M. Stallings of Virginia Power to E. G. Case of the NRC, Serial No. 081, February 15, 1978. j 50 mi.Scune l l I

9. B&W Owners Group Report, BAW-1799, "B&W 177-FA Reactor Vessel Beltline Weld Chemistry Study", July 1983.

10. " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," Regulatory Guide 1.99 - Revision 1, U.S. Nuclear Regulatory Commission, Washington, April 1977.

i i

11. NUREG-0800 - U.S. NRC Standard Review Plan, Branch Technical Position 5-2, I

Revision 1. July 1981.

12. Letter from K. L. Basehore of Virginia Power to D. R. Beynon, Jr. of Westinghouse Electric Corporation transmitting measured fuel assembly and cycle burnups for the Surry and North Anna Units, dated October 7, 1985.

i -l 1 2321:440M7to 51

APPENDIX A POWER DISTRIBUTIONS i Core power distributions used in the plant specific fast neutron exposure analysis of the Surry pressure vessels were derived from the measured fuel assembly and cycle burnup data supplied by Virginia Power (12). The beginning of-cycle (B0C) and end-of-cycle (EOC) fuel assembly burnups, based 3 on incore flux maps, were provided for selected peripheral fuel assembly locations for each of the first 7 cycles of operation. In addition, estimated data was provided for the current cycle of operation (Cycle 8). Table A-1 { shows the Surry Unit 1 fuel assembly and cycle burnups for Cycles 1 through 8. Similar data for Surry Unit 2 are shown in Table A-2. (The fuel assembly locations in the Surry cores are numbered according to Figure A-1). I Cycle-averaged relative assembly powers for each cycle were computed using the l following relation i Relative Assembly Power = EOC Assembly Burnup - BOC Assembly Burnup l Cycle Burnup i f and are shown in Tables A-3 and A-4 for Surry Units 1 and 2, respectively. l The cycle-averaged relative assembly powers representing the design basis core power distribution are also shown in Tables A-3 and A-4. i j Due to the extreme self-shielding of the reactor core, neutrons born in fuel } assemblies inboard of those for which burnup data were requested do not contribute significantly to the fast neutron exposure either at the surveillance capsules or at the pressure vessel. Therefore, power l distribution data for these interior assemblies are not given in Tables A-3 and A-4. In each of the adjoint evaluations, within assembly spatial gradients have been superimposed on the average assembly power levels. For peripheral l j am..enne A-1 l

assembly locations 1, 2, 3, 4, 5 and their symmetric partners, these spatial gradients also include adjustments to account for analytical deficiencies that tend to occur near the boundaries of the core region. 1 1 I 1 ? i t 4 i I k 1 h i l i t mm. enne A-2 j i w-c.-- p v.-.-%.

---i--

--me-

e--

w-+ --r m%, y .... - ~, - + + w,. --a

TABLE A-1 SURRY UNIT 1 BEGINNING-0F-CYCLE AND END-0F-CYCLE FUEL ASSEMBLY BURNUPS Fuel Assembly Burnup (MWO/MTU) Fuel Cycle (a) 1(13547) 2(6915) 3(8944) 4(13107)- Assembly BOC EOC BOC EOC BOC E0C BOC EOC 1 0 10615 0 6365 6400 13380-0 9825 2 0 8425 0 5075 5075 10763 0 7905 3 0 12275 0 7305 6450 14473 0 11825 4 0 8623 0 5438 6050 11730 0 8328 5 0 9285 0 6070 5450 11668 0 8928 6 0 12390 14550 20175 8650 18655 8300-20720 7 0 13900 0 6453 7300 16923 0 13320 8 0 13753 14475 21513 13725 21370 14475 28715 9 0 14180 14250 20000 8425 18570 0 14088 10 0 13328 0 7415 7400 17093 0 13373 11 0 14468 16275 23005 7625 16175 11725 25593 12 0 11555 0 6883 8625 17920' 13425 25895 4 (a) The number in parenthesis beside the cycle number is the fuel cycle length in MWD /MTU. l m i. m ontic A-3 3

. TABLE A-1 (Continued) SURRY UNIT 1 BEGINNING-0F-CYCLE AND END-OF-CYCLE FUEL ASSEMBLY BURNUPS. 4 -Fuel Assembly Burnup (MWD /MTU) Fuel Cycle (a) 1(13547) 2(6915) 3(8944) 4(13107) 1 Assembly BOC EOC BOC E0C B0C E0C BOC E0C 1 0 10730 0 10965 0 11125 8794 18233 2 14088 20363 16552 23188 0 8883 26041 31676 3 0 13685 0 14238 20034 28135 13824 .24780 4 15570 22105 25593 32388 25995 30730 28496 34121 5 15128 22115 17374 25063 25701 30363 22530 28747 6 7965 22515 10958 26340 25981 37125 11123 27319 7 0 15335 0 16045 0 13693 0 16178 8 8330 24743 15336 33000 19146 32393 30220 44443 9 0 16553 0 18945 0 14253 0 17569 10 0 15450 0 16473 0 12925 0 15824-11 0 17623 0 19590 0 15205 0 18023 12 7853 21533 10725 25888 28995 37585 15271 28592 (a) The number in parenthesis beside the cycle number is the fuel cycle length in MWD /MTV. mi.conno A-4

i TABLE A-2 SURRY UNIT 2 BEGINNING-0F-CYCLE AND END-0F-CYCLE FUEL ASSEMBLY BURNUPS Fuel Assembly Burnup (MWD /MTU) Fuel Cycle (a) 5(13971)- 6(16006) 7(14802) 8(13000) Assembly BOC EOC BOC EOC BOC EOC BOC E0C 1 0 11390 0 .7095 7095 13775 0 10955 2 0 9145 0 5938 0 ~6140 0 8943 3 0 13423 0 8678. 6215 14755 0 12663 4 0 9445 0 6215 0 6853 0 9385' 5 0 10178 0, 7095 0 7425 0 9995 6 0 13515 16950 24860 13515 20985 17450 29975 7 0 15183 0 10365 5938 15980 0 14290 8 0 15058 14775 24468 10878 20975 14750 28998 9 0 15593 17350 25908 9145 19815 0 15420 i 10 0 14650 0 10040 7095 17413 0 14118-11 0 15845 15850 25125 10178 20620 0 15698 4 12 0 12608 0 10090 13505 23145 17425 29873 1 i (a) The number in parenthesis beside the cycle number is the fuel cycle length in MWD /MTU. sm. Sonne A-5

TABLE A-2 (Continued) SURRY UNIT 2 BEGINNING-OF-CYCLE AND END-OF-CYCLE FUEL ASSEMBLY BURNUPS Fuel Assembly Burnup (MWD /MTU) Fuel Cycle (a) 5(13971) 6(16006) 7(14802) 8(13000) Assembly BOC EOC B0C E0C BOC EOC B0C E0C 1 0 10250 0 11210 0 11405 32167 37372 2 16268 22393 17672 24465 16811 24225 26980 31662 3 0 13125 0 14890 14893 25975 18204 27394 4 22130 28423 16405 24128 25586 31918 25974 31173 5 9388 17700 16812 25588 18846 26980 24223 30606 6 8880 23265 15202 30325 15639 30560 11406 25115 7 0 14575 0 15648 0 15320 0 13554 8 14118 29685 16876 34278 19868 36033 15323 31390 9 0 16813 0 18845 0 17;88 0 15761 10 0 15045 0 16810 0 15973 0 13606 11 0 16408 0 19508 0 18208 0 16497 12 15403 28278 10264 25745 19149 32600 15975 28199 l (a) The number in parenthesis beside the cycle number is the fuel cycle length in MWD /MTU. 1 mi. enne A-6

TABLE A-3 SURRY UNIT 1 CORE POWER DISTRIBUTIONS USED IN THE FLUENCE ANALYSIS Plant Specific Cycle Averaced Relative Assembly Power Design Fuel Cycle Basis Assembly Relative Power 5 6 7 8 1 1.00 0.78 0.92 0.78 0.75 2 0.83 0.62 0.73 0.64 0.60 3 1.21 0.91 1.06 0.90 0.90 4 0.86 0.64 0.79 0.64 0.64 5 0.92 0.69 0.88 0.70 0.68 6 0.98 0.91 0.81 1.12 0.95 7 1.10 1.03 0.93 1.08 1.02 8 1.00 1.02 1.02 0.85 1.09 9 1.05 1.05 0.83 1.13 1.07 10 1.08 0.98 1.07 1.08 1.02 11 1.06 1.07 0.97 0.96 1.06 12 0.95 0.85 1.00 1.04 0.95 m woonsi,o p.7

TABLE A-3 (Continued) SURRY UNIT 1 CORE POWER DISTRIBUTIONS USED IN THE FLUENCE ANALYSIS Plant Specific Cycle Averaced Relative Assembly Power l Design Fuel Cycle l Basis Assembly Relative Power 5 6 7 8 1 1.00 0.75 0.66 0.93 0.65 2 0.83 0.44 0.40 0.74 0.39 3 1.21 0.95 0.86 0.68 0.76 4 0.86 0.45 0.41 0.40 0.39 5 0.92 0.49 0.47 0.39 0.43 6 0.98 1.01 0.93 0.93 1.12 7 1.10 1.07 0.97 1.14 1.12 8 1.00 1.14 1.07 1.10 0.98 9 1.05 1.15 1.15 1.19 1.21 10 1.08 1.07 1.00 1.08 1.09 11 1.06 1.22 1.19 1.27 1.24 12 0.95 0.95 0.92 0.72 0.92 mwo<cw sa p.g

TABLE A-4 SURRY UNIT 2 CORE POWER DISTRIBUTIONS USED IN THE FLUENCE ANALYSIS Plant Specific Cycle Averaced Relative Assembly Power Design Fuel Cycle Basis Assembly Relative Power 5 6 7 8 1 1.00 0.77 0.78 0.71 0.80 2 0.83 0.61 0.66 0.65 0.65 3 1.21 0.90 0.96 0.91 0.93 4 0.86 0.64 0.69 0.73 0.69 5 0.92 0.68 0.78 0.79 0.73 6 0.98 0.91 0.87 0.79 0.92 7 1.10 1.02 1.14 1.07 1.05 = 8 1.00 1.01 1.07 1.07 1.04 9 1.05 1.05 0.95 1.13 1.13 10 1.08 0.99 1.11 1.10 1.03 11 1.06 1.07 1.02 1.11 1.15 12 0.95 0.85 1.11 1.02 0.91 mm.conne A-9

TABLE A-4 (Continued) SURRY UNIT 2 CORE POWER DISTRIBUTIONS USED IN THE FLUENCE ANALYSIS Plant Specific Cycle Averaged Relative Assembly Power Design Fuel Cycle Basis Assembly Relative Power 5 6 7 8 l 1 1.00 0.73 0.70 0.77 0.40 2 0.83 0.44 0.42 0.50 0.36 3 1.21 0.94 0.93 0.75 0.71 4 0.85 0.45 0.48 0.43 0.40 5 0.92 0.60 0.55 0.55 0.49 l 6 0.98 1.03 0.95 1.01 1.05 l 7 1.10 1.04 0.98 1.03 1.04 8 1.00 1.11 1.09 1.09 1.24 9 1.05 1.20 1.18 1.20 1.21 10 1.08 1.08 1.05 1.08 1.05 11 1.06 1.17 1.22 1.23 1.27 12 0.95 0.92 0.97 0.91 0.94 sm.**'" A-10

16o03 26 O' (MAJOR AXIS) l BAFFLE // CORE BARREL \\\\\\\\ l 2 \\ 45 x\\\\\\K p 6 7 3 4 \\ s\\\\ 8 9 10 5 / II I2 I Figure A 1. Surry Units 1 & 2 Core Description for Power Distribution Map A-11

APPENDIX B WELD CHEMISTRY Tables B.1-1 through B.1-4 provide the weld data output from the WOG Material Data Base. Given are the searches of all available data for the wire heat in the Surry Units 1 and 2 reactor vessels beltline region. The pertinent material chemical compositions are given, along with the wire / flux identification. The mean chemistry values and the population standard deviation are then calculated. The mean values of copper and nickel are used in the RTPTS analysis. Weld Chemistry Data Source and Plant: AN1 Arkansas Nuclear 1 BAW-1799 Babcock & Wilcox Report Number B&W Babcock & Wilcox CE Combustion Engineering CR3 Crystal River 3 Cu Weight % of Copper CWE Zion 1 ESA Emission Spectrographic Analysis MATSURV NRC Mender MATSURV Data Base MPC Materials Preperties Council Data Base Ni Weight % of Nickel OC1 Oconee 1 P Weight % of Phosphorous SC Surveillance Capsule Si Weight % of Silicon TMI Three Mile Island 1 VPA Surry 1 VIR Surry 2 WEP Point Beach 1 WO Weld Qualification sm. e4:in to B-1

TABLE B.1-1 Surry Unit 1 Intermediate and Lower Shell Longitudinal Wel(r Chemistry From WOG Materials Data Base - W* 7 Heat Number 8T1554

.=
=*

= ID WIRE u!RE Flus Ft tJu WE L DO(M Ces Mt P Se PLANT DESCRIPTION e4E AT TvFE T 74 LOT DATA SI AJRCE 4340 873354 ese-MD-est LiasDE On 9754 BAW-8 799.Wrt o.tG6 0.390 0.084 0.450 CR3 NOTTLE 70 INTER SHFtt es341 ST 8*sS4 ete-MD-es t LINDE Bo BM4 W-8 799.WD CH3 NONLE TD INff R SHFLL

s!42 8T 8554 age-to-est LINDE es e579 Wee-8 799.We n ten o.4So 0.015

0. 44e s TMs LawrR SHELL LONIe o

WPA INTER GHFLL LONP. WA LOWER SHFLL Lupa3

- 749 ST tm Pee-Mrl-es!

LINrt 9* 8479 feet-8 799.Wr* O.19e *

o. 6* en 6.034

0. *Ve eh7"A RT 3*134 ree-eO-pe t LINDt EM H5$4 Wee-I F99.ta?

0.39e so. 6 t o 0.086 es. 4 9ts rJ 07*4 8T 8534 ree-en-es t LINDE 80 H F*,4 FeAte-t ?99.WLR a.16o n.A?O O.086 0.450 CR3 NO22LE TO INTFR SHrlt e FJ es*are e*. 8 5 720*

o. 5 76non 0. 08 30(=) 0.4*wano se d.4 pew.

0.0T56?6 n n78972 o.tooIeem) O.019494 I 1 1

Report BAW-1799 reconsnends a snean value of copper equal to 0.18 wt1 with a standard deviation of 0.07 wt%.
- BAW-1799 recorrends that a mean value of nickel equal to 0.63 wt%.

TABLE B.1-2 Surry Unit 1 Lower Shell Longitudinal Weld Chemistry From WOG Materials Data Base-Wire Heat 299L44 ID tet#E aeInE Fa tes F t UE 4E E D[3( M Esa No F' Sa PLANT IESCRIPTION e4 AT t vF1 1791 10 Data StE 53CF ~~--- ---------- OD 2'*L ** 'We sq0-ast L ge@E 90 9630 Pts.ea7

o. 2gt.

o. 720 0.p07 p,ygO Kt NOllLF TO $NTER LL K2 TM8 INTER TO L[mE R 9E LL VME 95tVF IL L ANCF WF L D O277 299 Lee 79e_paeg-ge g L gsq gen gw.gt Pts.edO o

9me es. Plo to. t e 19 to. e*sO Kt Nullt E TO INTER 9 ELL DC2 IME INTER TO ((nER SNFLL TMt SL83W ILL ANCE WEL D 02fr5 29'ates see eq0-ast L lostf 94 R596 VPA.SC n.2Se o.6m*

e.ott 0.370 TMs LDhdE R 94LL LONG WPA StsWVEILLANCE WFLD 0344 299 Lee e9e-940 est LineDE 90 BASS Pnks - 3 799.E SA O. N o O. h'.O o.035 0.460 Oct NOllLE TO INTER 9 ELL DC2 IMI INTER TO LOWF R 94 L L TMI Sts3VF ILL ANCE tdFL D 03e7 299 Lee see-eq0-as s (3edet fan sieSe mees-0799.te;A

o. 39..

630 0.083 0.490 Kg adullt E TO INTER 9 ELL 002 IMS INTER TO LDMER 9 ELL TMt Sts7VFILLANCE tdELD 0349 299 tee poe_,40 est L io@F RO 9630 Dame 3799.FSA

o. r '.e*

e*.640 0.Ote 0.470 El NOllL E TO INTER 9 ELL K2 IMS INTER TO LOhER DELL g TMt SURVIILLANCE edEL D e 03e9 299 Lee see-ec set t INEE f>

H6Se Pabs - t F99.t SA O.76 *

o. 69f t 0.033 p.460 003 Nullt E TO INTER 9 ELL (48 UC2 TMt IssTER TO LOMER 9 ELL TMt StfrVEILL ANCE tsELD 0330 299 Lee Pee-ec-asI Liter Rrt pgSe Dame-g799.FSA O.730 O.70n O.O32 O.420 DC3 NDE EL E 70 INTER 9 ELL DC2 TMt INTER TO LOMFR 9 ELL TMS SEJRVEILLANCE tdFLD 0333 299 Lee steeg L gnery gen ynge,o pane.gygg.ESA O.?AO o.690 0.032 0.440 DCB psOlltE TO INTER 9 ELL DC2 TMs INTER TO t.tDdER DELL TMt SURVE IL LANCE tEL D OTS2 299 Lee pos-*c -ast LleCF f>e 9650 enas-t ?99.ESA O.34 *

o. 69e

e.oge v.470 008 esulltf TO INith 9 ELL DC2 IMt INTER T3 LOWER 9 ELL TMt StstVEILLANCE tdFLD 01rS3 29'at es pos -eO-an g LIMDF Eb o

$3630 fanas - 3 799,t SA

o. ems e 730 0.o33 o.S$d 003 NullL L ID INTER 9 ELL DC2 IMt INTER TO LOWER 9 ELL iMI SURVEILLANCE edtLD 03Se 2gaites pos-ec-agt L IN[T 90 96SD Dabs-IF99.ESA

o. Tese

0. 70 0 0.034 0.4 70 OCR 9esilLE TO INTER DELL OC2 TMt INTER TO LONFR 9 ELL TMI SURVEILLANCE edFLD et35 294 Lee fee-+o-es t LgneE 90 9650 DAns-IF99.ESA e.370

p. FisO O.014 0.470 000 NOllLE TO INTER SBELL DC2 TMt INTER TO LOMER DELL VMS StsvVrILLANEE ndELD

336 294 Lee seg L3edDF 193 9650 Dame-g 799,ESA O.370 e.70p O.o83 0.eHO 003 NOllt E TO INTER SELL OC2 TMs INTER TO LthdER DELL VMS 95tVF ILLANCE IdFLD Rev.1 0137 799 Lee pes-en-ast L3e@f 90 Ib630 stAae-I F99.FSA 6.3a0 0.700 0.085 0.550 000 edrilit t TO tastrR 9 ELL DC2

Ea d d d d d d d d d d d - d d k* I"s ISI IS* IS" ESI aiiwe$*s.$S*$*I 1 1 1 i.ns $* I I I** $* oss.5s.ss.ws."s .f ws.fs.S ws fs .ss.w n es-es-ee si.iai.1i1iIiiiivsi slo asi es ee-es. sloai a i 1 1 aloi i i i sl slo si slo si. slo si e]s. slo sto s s s s es-es-ee-eso ee-es-ed-es-se .; E{E g t. $.1:$. 0 1,$,E{EsO, U"E"i:i~,s,*,W"{~Es"{~E"~E"~ ~U "U ~ ,~-{t,, O, t{~g{ e,,e, 1 ~,,- s b > ~, ,, i ~, 1 s,- EE5lEE59EE59EE59EE50EE5lEE50EE50EE50EE50EE50EEEEEE5EEE5EEEEEEE 3 R 3 3 I i R b b e m 4 A e h m e 4 4 4 4 4 h A O 5 5 5 O C O 5 5 C O O i i i i i i i a i i i s i i i 9 f, E 9 9 9 9 k l a 4 h 4 x x x x d d i d d d d d d d i i d d d r n b b f d f f f f 5 5 5 5 5 5 5 3 5 5 5 5 5 m w 4 I I I I I i i I i i i i i i i m m S 8 A 3 4 3 8 R A 3 s A R 8 8 8 8 E 8 8 8 8 i i t S i 1 8 E 8 8 8 1 i i E 8 I I I I I I I I I I I I I I I 3 s s a a i i i i i i i i i i i i i i i 3 k k k k f. k k k k k k k k k k r

a. a. a.

a a. a a. a a. a

a. a.

s a. B-4

1 t u J d d gd y g dJ da d da dJ ydg gNg d e.ns oss oss ss.el i N y liIfeNgd teI lea 5 e8 Xe3 Xe3 teIf XaI Xe g Re BSI R tel ws oss.fs oss iag!IgiIg!1g111agiagiag!ivs.ws.se.wsoss.ns.ns giag!Ig0!Ig!agiagiag! -ee-ee-es-ese!O d O 0 0 0

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a. s. a.

a i E E E E E E E e e B-5

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l11.1l.1.ii:i i i fi!liifi 1 t 1 1 1 1 1. 1. 1. 1. 1. 11:.1. i illi;i;f;isi;l;i,1,1;ig1;l;ig:1 l 'a 'n 'n ', 1; f; i; I 'n 'n 's a t -l '. ill!!!!!!!!!ililill!!!!!!!!!!!!!!!! E il tillilli!!illi!!i!!illillilli!!i!!i!!i!!illi!!illi E l a 1 !!! I 'l 5 'l !I i! !' I }p sieses e e e e e i e a e e e e e e e T E' iiiiii i i i i iiiiiii! i i "y 4i4444 4 e i 4 4 4 4 4 4 4 4 4 e i i i ))R ! !!Ii I i ! !I I I I 'l I ,_ u g 444444 4 i s 4 4 4 4 4 4 4 4 4 4 4 4 4 59 Hi!881 A A i I ? A i E a gh 444444 4 4 4 e e a 4 4 4 4 i i s e i i ?k I. 5. k k k k.k !. l

5..!.

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8

!s ,i,il. .n ,sct i t i i i nit i i i i i i i

u. y

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TABLE B.1-4 Surry Unit 2 Lower Shell Longitudinal Weld Chemistry From WOG Material Data Base - Wire Heat Number ST1762 3D asIRE esgset Flus i t US WFLDOfse Cae see P 5e PLAssT tlE EftIPT f De et at 7,FT 1 es 5 t us geAt a CJEElf Wes t L isdEE mm m97 pas.ed)

u. f M*

e.S 10 0.087 0.330 CWE 18fffR M LL LGd3 VIR LOWEIR 94LL Lpal pose est L ineDE me 8% 7 pas.an3 es..wan o.630 a.coe 0.330 CR3 latfER 94.LL LDE3 om eT37 2 Ladt INTER 94LL Ladi e*214 ST S FR CadE LOWN singtL LonG Yng getter 94LL L(ING VIR LOWR 94tL LonG . bs Ent s 7Q ree ag-est L geet p. su' At ges.edo 0.39* en.6tp "p.*f7 0.eTo 003 LDWR 94LL LONG adQifLE TO IssTER 9 ELL edEP e neM

e. e*.s ev. m.e

a. 3h ass 3 laster 94LL LOdB smm pO4sg a tourg goe pes *.=

ses.ndo metI LOKR 94LL LONG =e.'789 9117C LR3 SteTLR 94LL LGsG sW".*> *- thAss 8 F99.tdC

*. I b '
- .4**

.037 0.440 UL3 LDE R 94 LL LUDO e4 P sollt E ftp lesTE R a;pq LL

- - *4 et t FM two-M) est LipdDE. *

.**.y ses t FQ pgs +(3 est L gpurg te n 3694 goes 3fvy,e#7 es. 7t so. &wt n 434 es,4 *a Lk3 LOWE R DE LL LLP8G C3 - -.._- -~~~.. u.t.7.3 ***e.*4e641 p.nt?t&7 a.4*23331 ..ews.3 ..e,s..e n.e.o. m _ ames.

Since these values are limited to the weld metal qualification test reports, report BA'J.1799 reconnends that a mean value of ccpper equal to 0.29 wt1 with a standard deviation equal to 0.07 wt% be applied because a bias exists in the retest of similar type welds.
- BAW.1799 recomunends that a mean value of nickel equal to 0.55 wt%

be applied.

og., 5 e.!.esl n I E ll 3 t e as EN 3 l 68 d4 i Et 3 II I

11 M

-= da .e e= ~t }5 m m e, 1s yg 3 y G. 9 iI L .E S f5 l6 33 S a "e 8 89

I d-[

SS E se [g Ed n ll Ii ig ii ll u 10

APPEN0!X C RT VALUES OF SURRY UNITS 1 AND 2 PTS REACTOR VESSEL BELTLINE REGION MATERI ALS C.1 SURRY UNIT 1 values, as a function of both Tables C.1-1 through C.1-5 provide the RTPTS constant fluence and constant EFPY (assuming the projected fluences values), for all beltline region materials of the Surry Unit I reactor vessel. The RTPTS values are calculated in accordance with the PTS rule, which is Reference (1) in the main body of this report. The vessel location numbers in the following tables correspond to the vessel materials identified below and in Table 111.3-1 of the main report. Locatier. Vossol Paterial 1 Intermediate shell plate C4326-1 2 Intermediate shell plate C4326 2 3 Lower shell plate C4415 1 4 Lower shell plate C4415 2 5 Intermediate and lower shells longitudinal welds L1, L3, L4 6 Intermediate to lower shell circumferential weld WOS 7 Lower shell longitudinal weld L2 lpey,g C.2 SURRY UNIT 2 Tables C.2-1 through C.2 $ provide the RTPTS values, as a function of both constant fluence and constant EFPY (assuming the projected fluence values), for all beltline region materials of the Surry Unit 2 reactor vessel. The Rip 73 values are calculated in accorda'nce with the PTS rule, which is Reference (1) in the main body of this report. The vessel location nutbers sm,wune C-1

in the following tables correspond to the vessel materials identified below and in Table !!!.3-2 of the main report. Location Vessel Material 1 Intermediate shell plate C4339-1 2 Intermediate shell plate C4208-2 3 Lower shell plate C4331-2 4 Lower shell plate C4339-2 5 Intermediate shell longitudinal welds L4,L3 6 Lower shell longitudinal welds L2,L1 7 Intermediate to lower shell circumferential weld WO5 m s w. une.,s c.g

l TABLE C.1-1 f! PTS VALUES FOR SURRY UNIT 1 REACTOR VESSEL 10 2 BELTLINE REGION MATERI ALS 0 FLUENCE = 1.0 x 10 n/cm LOC PLANT CU NI P 1 VALUE TYPE FLUENCE RTPTS 1 VPA .11 .55 .008 10. Actual B.M. .10E+19 92. 2 VPA .11 .55 .008 0. Actual 5.M. .10E+19 82. 3 VPA .11 .50 .014 20. Actual B.M. .10E+19 101, 4 VPA .11 .50 .014 0. Actual B.M. .10E+19 81. 5 VPA .18 .63 .014 0. Generic L.W. .10E+19 120. 6 VPA .21 .58 .016 0. Generic C.W. .10E+19 129. 7 VPA .35 .67 .014 0. Generic L.W. .10E+19 186. Rev.1 NOTES: B.M. = Base Material (Plate) L.W. = Longitudinal Wald C.W. = Circumferential Wold Reference temperatures are in 'F e m..m..... i. C-3

TABLE C.1-2 RT VALUES FOR SURRY UNIT 1 REACTOR VESSEL PTS 10 2 BELTLINE REGION MATERI ALS 0 FLUENCE = 5.0 x 10 n/cm LOC PLANT CU NI P I VALUE TYPE FLUENCE RTPTS 1 VPA .11 .55 .008 10. Actual B.M. .50E+19 110. 2 VPA .11 .55 .008 0. Actual B.M. .50E+19 100. 3 VPA .11 .50 .014 20. Actual B.M. .50E+19 119. 4 VPA .11 .50 .014 0. Actual B.M. .50E+19 99. 5 VPA .18 .63 .014 0. Generic L.W. .50E+19 154. 6 VPA .21 .58 .016 0. Generic C.W. .50E+19 168. 7 VPA .35 .67 .014 0. Generic L.W. .50E+19 255. Rev.1 i I am.nei. ww e C-4

f TABLE C.1-3 MPTS VALUES FOR SURRY UNIT 1 REACTOR VESSEL 0 2 BELTLINE REGION MATERIALS 0 FLUENCE = 1.0 x 10 n/cm LOC PLANT CU NI P 1 VALUE TYPE FLUENCE RTPTS 1 VPA .11 .55 .008 10. Actual B.M. .10E+20 121. 2 VPA .11 .55 .008 0. Actual B.M. .10E+20 111. 3 VPA .11 .50 .014 20. Actual B.M. .10E+20 129. 4 VPA .11 .50 .014 0. Actual B.M. .10E+20 109. 5 VPA .18 .63 .014 0. Generic L.W. .10E+20 173. 6 VPA .21 .58 .016 0. Generic C.W. .10E+20 190. 7 VPA .35 .67 .014 0. Generic L.W. .10E+20 296.lpey,3 me ewsww te c.5

l TABLE C.1-4 RT VALUES FOR SURRY UNIT 1 REACTOR VESSEL PTS BELTLINE REGION MATERIALS 0 CURRENT (7.4 EFPY) FLUENCE VALUES LOC PLANT CU NI P I VALUE TYPE FLUENCE RTPTS 1 VPA .11 .55 .008 10. Actual B.M. .11E+20 123. 2 VPA .11 .55 .008 0. Actual B.M. .11E+20 113. 3 VPA .11 .50 .014 20. Actual B.M. .11E+20 131. 4 VPA .11 .50 .014 0. Actual B.M. .11E+20 111. 5 VPA .18 .63 .014 0. Generic L.W. .18E+19 131. 6 VPA .21 .58 .016 0. Generic C.W. .11E+20 195. 7 VPA .35 .67 .014 0. Generic L.W. .18E+19 208.lpey,1 l m i. e i. wi m is C-6

TABLE C.1-5 RT VALUES FOR SURRY UNIT 1 REACTOR VESSEL PTS BELTLINE REGION MATERIALS @ LICENSE EXPIRATION (28.8 EFPY) LOC PLANT CU NI P I VALUE TYPE FLUENCE RTPTS 1 VPA .11 .55 .008 10. Actual B.M. .40E+20 149. 2 VPA .11 .55 .008 0. Actual B.M. .40E+20 139. 3 VPA .11 .50 .014 20. Actual B.M. .40E+20 156. 4 VPA .11 .50 .014 0. Actual B.M. .40E+20 136. 5 VPA .18 .63 .014 0. Generic L.W. .64E+19 160. 6 VPA .21 .58 .016 0. Generic C.W. .40E+20 249, 7 VPA .35 .67 .018 0. Generic L.W. .64E+19 269. Rev.1 sw.iow.w w,o c.y

TABLE C.2-1 RT VALUES FOR SURRY UNIT 2 REACTOR VESSEL BELTLIN[ ION MATERIALS 0 FLUENCE = 1.0 x 10 n/cm 18 2 LOC PLANT CU NI P I VALUE TYPE FLUENCE RTPTS 1 VIR .11 .54 .012 30. Actual B.M. .10E+19 112. 2 VIR .15 .55 .008 -30. Actual B.M. .10E+19 66. 3 VIR .12 .60 .009 10. Actual B.M. .10E+19 96. 4 VIR .11 .54 .012 10. Actual B.M. .10E+19 92. 5 VIR .21 .58 .016 0. Generic L.W. .10E+19 130. 6 VIR .29 .55 .013 0. Generic L.W. .10E+19 157. 7 VIR .19 .56 .017 0. Generic C.W. .10E+19 122. NOTES: B.M. = Base Material (Plate) L.W. = Longitudinal Weld C.W. = Circumferential Weld Reference temperatures are in *F mivem.Smno C-8

TABLE C.2-2 RT VALUES FOR SURRY UNIT 2 REACTOR VESSEL PTS 10 2 BELTLINE REGION MATERIALS 9 FLUENCE = 5.0 x 10 n/cm LOC PLANT CU NI P I VALUE TYPE FLUENCE RTPTS 1 VIR .11 .54 .012 30. Actual B.M. .50E+19 130. 2 VIR .15 .55 .008 -30. Actual B.M. .50E+19 92. 3 VIR .12 .60 .009 10. Actual B.M. .50E+19 117, 4 VIR .11 .54 .012 10. Actual B.M. .50E+19 110. 5 VIR .21 .58 .016 0. Generic L.W. .50E+19 168. 6 VIR .29 .55 .013 0. Generic L.W. .50E+19 210, 7 VIR .19 .56 .017 0. Generic C.W. .50E+19 156. m i. e m.4 m en e C-9

TABLE C.2-3 EPTS VALUES FOR SURRY UNIT 1 REACTOR VESSEL 0 2 BELTLINE REGION MATERIALS 9 FLUENCE = 1.0 x 10 n/cm LOC PLANT CU NI P I VALUE TYPE FLUENCE RTPTS 1 VIR .11 .54 .012 30. Actual B.M. .10E+20 140. 2 VIR .15 .55 .008 -30. Actual B.M. .10E+20 107. 3 VIR .12 .60 .009 10. Actual B.M. .10E+20 130. 4 VIR .11 .54 .012 10. Actual B.M. .10E+20 120. 5 VIR .21 .58 .016 0. Generic L.W. .10E+20 190. 6 VIR .29 .55 .013 0. Generic L.W. .10E+20 241. 7 VIR .19 .56 .017 0. Generic C.W. .10E+20 176. am.mmi.wuno C-10

TABLE C.2-4 VALUES FOR SURRY UNIT 2 REACTOR VESSEL RJPTS BELTLINE REGION MATERIALS 0 CURRENT (7.6 EFP!) FLUENCE VALUES LOC PLANT CU NI P I VALUI TYPE FLUENCE RTPTS 1 VIR .11 .54 .012 30. Actual B.M. .11E+20 142. 2 VIR .15 .55 .008 -30. Actual B.M. .11E+20 110. 3 VIR .12 .60 .009 10. Actual B.M. .11E+20 132. 4 VIR .11 .54 .012 10. Actual B.M. .11E+20 122. 5 VIR .21 .58 .016 0. Generic L.W. .20E+20 144. 6 VIR .29 .55 .013 0. Generic L.W. .20E+19 177. 7 VIR .19 .56 .017 0. Generic C.W. .11E+20 179. am.,eni.wser.io C-11

( TABLE C.2-5 RT VALUES FOR SURRY UNIT 2 REACTOR VESSEL PTS BELTLINE REGION MATERIALS @ LICENSE EXPIRATION (29.4 EFPY) LOC PLANT CU NI P I VALUE TYPE FLUENCE RTPTS 1 VIR .11 .54 .012 30. Actual B.M. .34E+20 165. s 2 VIR .15 .55 .008 -30. Actual B.M. .34E+20 143. 3 VIR .12 .60 .009 10. Actual B.M. .34E+20 158. 4 VIR .11 .54 .012 10. Actual B.M. .34E+20 145. 5 VIR .21 .58 .016 0. Generic L.W. .71E+19 179. 6 VIR .29 .55 .013 0. Generic L.W. .71E+19 225. 7 VIR .19 .56 .017 0. Generic C.W. .34E+20 222. Rev.1 j l am.mm.io mno C-12 ---}}

v t e Letter Sequence Other
TAC:59988, (Approved, Closed) TAC:59989, (Approved, Closed)
Initiation Acceptance... Administration Questions Answers Results Approval Other: ML18149A459, ML20140D065, ML20207B523, ML20210B480, ML20237C058
MONTHYEAR1986-12-12 [Table View]

ML20210B480

Text

SUMMARY

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