ML20150C515

Rev 3 to North Anna Units 1 & 2 Reactor Vessel Fluence & Ref Temp PTS Evaluations

ML20150C515
Person / Time
Site:	North Anna
Issue date:	01/31/1988
From:	Heinecke C, Lippincott E, Weaver M WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20150C519	List: ML20150C515 ML20153A839
References
WCAP-11016, WCAP-11016-R03, WCAP-11016-R3, NUDOCS 8803210036
Download: ML20150C515 (80)

Text

WESTINGHOUSE CLASS 3 L

CUSTOMER DESIGNATED DISTRIBUTION WCAP-11016 Rev. 3 CUSTOMER DESIGNATED DISTRIBUTION NORTH ANNA UNITS 1 AND 2 REACTOR VESSEL FLUENCE AND RT EVALUATIONS PTS C. C. Heinecke E. P. Lippincott M. Weaver G. N. Wrights Work Performed for Virginia Power Company January 1988 Approved <;-

[ k % Af T. A. Meyett, Mdnager Structural Materials and Reliability Technology Approved by:

bv F. L. Lau, Manager Radiation and Systems Analysis Approved by:

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C. W. Hirst,, Manager Roactor Coolant System Components Licensing Work Performed Under Shop Order VHLJ-108 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 Generation Technology Systems Division P.O. Box 2728 oittsburgh, Pennsylvania 15230 2728 E803210036 $80315

auetten to POR ADOCK G5000333 h,

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TABLE OF CONTENTS

? ant TABLE OF CONTENTS i

LIST OF TABLES ii LIST OF FIGURES v

I.

INTRODUCTION 1

1.1 The Pressurized Thermal Shock Rule 1

I.2 The Calculation of RT 3

PTS II.

NEUTRON EXPOSURE EVALUATION 5

II.1 Method of Analysis 5

II.2 Fast Neutron Fluence Results 8

Rev.3 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 Sumary of Plant-Specific Material Properties 36 IV.

DETERMINATION OF RT VALUES FOR ALL BELTLINE 41 PTS REGION MATERIALS IV.1 Status of Reactor Vessel Integrity in Terms of RTPTS 41 versus Fluence Results IV.2 Discussion of Results 42 V.

CONCLUSIONS AND RECOMMENDATIONS 47 i Rev.3 VI.

REFERENCES 49 VII.

APPENDICES A.

Power Distribution A-1 B.

Weld Chemistry B-1 C.

RT Values of North Anna Units 1 and 2 Reactor Vessel C-1lRev.3 Nktline Region Matorials

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WESTINGHOUSE CLASS 3 LIST OF TABLES Page II.2-1 North Anna Unit 1 Fast Neutron (E>1.0 MeV) Exposure at 12 the Pressure Vessel Inner Radius - 0* Azimuthal Angle 11.2-2 North Anna Unit 1 Fast Neutron (E>1.0 MeV) Exposure at 13 the Pressure Vessel Inner Radius - 15' Azimuthal Angic 11.2-3 North Anna Unit 1 Fast Neutron (E>1.0 MeV) Exposure at 14 '

the Pressure Vessel Inner Radius - 30' Azimuthal Angle 11.2-4 North Anna Unit 1 Fast Neutron (E>1.0 MeV) Exposure at 15 '

the Pressure Vessel Inner Radius - 45' himuthal Angle 11.2-5 North Anna Unit 1 Fast Neutron (E>1.0 MeV) Exposure at 16 the 15* Surveillance Capsule Center 11.2-6 North Anna Unit 1 Fast Neutron (E>1.0 MeV) Exposure at 17 ;

the 25' Surveillance Capsule Center 11.2-7 North Anna Unit 1 Fast Neutron (E>1.0 MeV) Exposure at 18 the 35' Surveillance Capsule Center 11.2-8 North Anna Unit 1 Fast Neutron (E>1.0 MeV) Exposure at 19

• Rev.3 the 45' Surveillance Capsule Center 11.2-9 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exposure at 20 the Pressu'e Vessel Inner Radius O' Azimuthal Angle II.2-10 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exposure at 21 the Pressure Vessel Inner Radius 15' Azimuthal Angle II.2-11 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exposure at 22 the Pressure Vessel Inner Radius 30' Azimuthal Angle II.2-12 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exposure at 23 the Pressure Vessel Inner Radius 45' Azimuthal Angle 11.2-13 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exaosure at 24 the 15* Surveillance Capsule Center II.2-14 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exposure at 25 the 25' Surveillance Capsule Center II.2-16 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exposure at 26 the 35* Surveillanca Capsule Center UHM 0

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WESTINGHOUSE CLASS 3 LIST OF TABLES (Continued)

Page II.2-16 North Anna Unit 2 Fast Neutron (E>1.0 MeV) Exposure at 27 the 45' Surveillance Capsule Center III.3-1 North Anna Unit 1 Reactor Vessel Beltline Region 37 Material Properties III.3-2 North Anna Unit 2 Reactor Vessel Beltline Region 38 Material Properties IV.1-1 RT Values for North Anna Unit 1 43 Rev.3 PTS IV.1-2 RT Values for North Anna Unit 2 44 PTS A-1 North Anna Unit 1 Beginning-of-Cycle and End-of-Cycle A3 Fuel Assembly Burnups A-2 North Anna Unit 2 Beginning-of-Cycle and End-of-Cycle A-4 Fuel Assembly Burnups A-3 North Anna Unit 1 Core Power Distributions Used in the A-5 Fluence Analysis A-4 North Anna Unit 2 Core Power Distributions Used in the A-6 Fluence Analysis B.1-1 North Anna Unit 1 Intermediate to Lower Shell Circumferential B-2 Weld Chemistry From WOG Materials Data Base -

Wire Heat Number 25531 B.1-2 North Anna Unit 2 Intermediate to Lower Shell Circumferential B-3 Weld Chemistry From WOG Materials' Data Base - Wire Heat Number 716126 C.1-1 RT Values for North Anna Unit 1 Reaggor Vegsel Beltline C-2 RehIbn Materials at Fluence = 1.0 x 10 n/cm RT Values for North Anna Unit 1 Rea C-3 RehIbn Materials at Fluence = 5.0 x l']ggor Vegsel Beltline C.1-2 n/cm RehIbnMaterialsatFluence=1.0x10 Values for North Anna Unit 1 Reag,r Vegs C-4 C.1-3 RT r/cm C.1-4 RT Values for North Anna Unit 1 Reactor Vess'1 Beltline C-5 RehIbn Materials at Current (4.7 EFPY) Fluence Values C.1-5 RT Values for North Anna Unit 1 Reactor Vessel Beltline C-6 RehIbn Materials at License Expiration (30.7 EFPY)

Rev.3

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WESTINGHOUSE CLASS 3 c

LIST OF TABLES (Continued)

Page C.2-1 RT Values for North Anna Unit 2 Reaggor Vegsel Beltline C-7 RehIhnMaterialsatF.uence=1.0x10 n/cm RehIhnWaterialsat. Fluence =5.0x10 Values for North ;.nna Unit 2 Reaggor V C-8 C.2-2 RT n/cm C.2-3 RT Values for North Anna Unit 2 Reagger Vegsel Beltline C-9 RehI$nMaterialsatFluence=1.0x10 n/cm C.2-4 RT Values for North Anna Unit 2 Reactor Vessel Beltline C-10 Re5I$n Materiais at Current (3.s EFeY) Fluence values C.2-5 RT Values for North Anna Unit 2 Reactor Vessel Beltline C-11 RehI$nMaterialsatLicenseExpiration(31.4EFPY)

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WECINGHOUSE CLASS 3 LIST OF FIGURES Page II.1-1 North Anna Reactor Geometry 28 11.2-1 North Anna Unit 1 Maximum Fast Neutron (E>1.0 MeV) 29 Fluence at the Pressure Vessel Inner Radius as a Function of Full Power Operating Time II.2-2 North Anna Unit 2 Maximum Fast Neutron (E>1.0 MeV) 30 Fluence at the Pressure Vessel Inner Radius as a Function of Full Power Operating Time 3

II.2-3 North Anna Unit 1 Maximum Fast Neutron (E>1.0 MeV) 31 I Fluence at the Pressure Vessel Inner Radius as a i

Function of Azimuthal Angle II.2-4 North Anna Unit 2 Maximum Fast Neutron (E>1.0 MeV) 32 gey,3 Fluence at the Pressure Vessel Inner Radius as a Function of Azimuthal Angle II.2-5 North Anna Units 1 and 2 Relative Radial 33 Distribution of Fast Neutron (E>1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall II.2-6 North Anna Units 1 and 2 Relative Axial 34 Distribution of Fast Neutron (E>1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall III.1-1 Identification and Location of Beltline Region 39 Material for the North Anna Unit 1 Reactor Vessel III.1-2 Identification and Location of Beltline Region 40

~ Material for the North Anna Unit 2 Reactor Vessel IV.1-1 North Anna Unit 1 - RT Curves per PTS Rule Methodology 45 Rev.3 PTS IV.1-2 North Anna Unit 2 - RT Curves per PTS Rule Methodology 46 PTS A-1 North Anna Units 1 and 2 Core Description for Power A-7 l

Distribution Map e

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WESTINGHOUSE-CLASS 3 SECTION I INTRODUCTION The purpose of this report is to submit the reference temperature for

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pressurized thermal shock (RTPTS) values for the North Anna 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 II presents tha 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 RTPTS calculations from present through the projected 4

end-of-license fluence values.

I.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 l

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 an event may produce the propagation of flaws postulated to exist near the inner wall surface, thereby potentially affecting the integrity I*

of the vessel.

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t!ESTfNGHOUSE CLASS 3 3

The Rule establishes the following requirements for all domestic, operating PWRs:

Establishes the RTPTS (measure of fracture resistance) Screening Criterion for the reactor vessel beltline region 270*F for plates, forgings, axial welds 300*F for circumferential weld materials 6 Months From Date of Rule:

All plants were required to submit their RT values at that time (per the prescribed methodology) and PTS values at the expiration date of the operating projected RTPTS license.

The date that this submittal was to have been received by the NRC for plants with operating licenses was January 23, 1986.

9 Honths From Date of Rule:

Plants projected to exceed the PTS Screening Criterion were required to 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 was to have been 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 epproval 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 10CFR Part 50.34.

l In the Rule, the NRC provides guidance regarding the calculation of the toughness state of the reactor vessel materials - the "reference temperature is for nil ductility transition" (RTNDT). For purposes of the Rule, RTNDT nn, man in 2

WESTINGHOUSE CLASS 3 now defined as "the reference temperature for pressurized thermal shock" (RTPTS) and calculated as prescribed by 10 CFR 50.61(b) of the Rule.

Each USNRC licensed PWR must submit a projection of RT values from the time of PTS the submittal to the license expiration date.

This assessment was to be submitted within 6 months after the effective date of the Rule, on January 23, 1986, with updates whenever changes occur affecting projected values. The calculation must be made for each weld and plate, or forging, in the reactor vessel beltline.,The purpose of this report is to provide the RTPTS values for North Anna 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 RT are actually PTS one of several ways to calculate RT For the purpose of comparison with NDT.

the Screening Criterion, the value of RT f r the reactor vessel must be PTS 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 l

l Equation 2:

PTS = I + M + 283 f.194 0

RT l

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WESTINGHOUSE CLASS 3 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

-l 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.

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M = the margin to be added to cover uncertainties in the values of initial RTNDT, copper and nickel content, fluence, and calculation procedures.

In Equati'.'n 1, M=48'F if a measured value of I was used, and M=59'F if the generi', mean value of I was used.

In Equation 2, M=0*F if a measured value of I wa, used, and M=34*F if the generic mean value of I was used.

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Cu and Ni = the best estimate weight percent of copper and nickel in the

material, 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.

Note, since the chemistry values given in equations 1 and 2 are best estimate mean values, and the margin, M, increases the RT values to be upper bound PTS predictions, 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 North Anna Units 1 and 2 are discussed in Section III.2.

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WESTINGHOUSE CLASS 3 SECTION-II NEUTRON EXPOSURE EVALUATION This section presents the results of the application of Westinghouse derived adjoint importance functions to the calculation of the North Anns 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.

11.1 NETH00 0F ANALYSIS A plan view of the North Anna 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 O'-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*, 293* and 305* relative to the reactor geometry flat at

-O'.

In performing the fast neutron exposure evaluations for the reactor geometry shown in Figure II.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 I

against which cycle by cycle plant specific calculat'ons can be compared.

The second set of calculations consisted of a series of adjoint analyses relating l

the response of interest (neutron flux > 1.0 NeV) at several selected locations within the reactor geometry 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.

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The forward transport calculation was carried out in R,0 geometry using the l

DOT discrete ordinates code (2) and the SAILOR cross-section library (3).

The l

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f-L WESTfNGHOUSE CLASS 3 SAILOR library is a 47 group, ENDF/B-IV based data set produced specifically l

for light water reactor applications. Anisotropic scattering is treated with aP expansion of the cross-sections. An S angular quadrature was used.

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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 2e 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

+2a 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 oesign 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.

The adjoint analyses were also carried out using the P3 cross-section approximation from the SAILOR library. Adjoint source locations were chosen at the center of each of the surveillance capsules as well as at positicns 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 exposuro 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:

R

1. # I (R,0,E) F (R,0,E) dE R dR de R,0 R 8 E where:

I Response of interest ($ (E > 1.0 MeV)) at radius R and R

=

R,0 azimuthal angle 8.

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WESTINGHOUSE CLASS 3 I (R,0,E) =

Adjoint importance function at radius R and azimuthal angle e for neutron energy group E.

F (R,0,E) =

Full power fission density at radius R and azimuthal angle 0 for neutron energy group E.

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 North Anna 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 North Anna Unit 1 expires on April 1, 2018 and the operating license for North Anna Unit 2 expires on August 21, 2020 (each forty years af ter the operating license was issued).

l This report includes fluence projections from September 30, 1985 to the Rev.3 j

respective license expiration dates using the cycle-averaged core power distribution of the current operating cycle (cycle 5 for Unit 1 and cycle 4 for Unit 2) and an assumed future capacity factor of 80%. All fluence projections into the future reflect the low leakage fuel management strategies exemplified by the core loadings of the current cycles.

Finally, it hat. been assumed that the core thermal power will be uprated from 2775 MW to 2893 th MW during cycle 6 of Unit 1 and during cycle 5 of Unit 2.

th The transport methodology, both forward and adjoint, using the SAILOR l

cross section library has been benchmarked against the Oakridge National nusmewe y

WESTINGHOUSE-CLASS 3 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.

11.2 FAST NEUTRON FLUENCE RESULTS Calculated fast neutron (E >1.0 MeV) exposure results for North Anna Units 1 and 2 are presented in Tables II.2-1 through II.2-16 and in Figures 11.2-1 through II.2-6.

Data is presented at several azimuthal locations on the inner-radius of the pressure vessel as well as a.t the center of each surveillance capsule.

In Tables 11.2-1 through II.2-4 cycle-specific maximum neutron flux and fluence levels at 0*,

15', 30*, and 45' on the ' pressure vessel inner radius of North Anna Unit 1 are listed for the 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 + 2e level. Similar data for the center of surveillance capsules located at 15', 25', 35', and 45' are given in Tables II.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, a capsule was removed from the 15' location at the end of cycle 1.

An error in the measured fluence has however precluded its use herein.

Cycle-specific and design basis fast neutron flux and fluence data at the inner radius of the pressure vessel of North Anna Unit 2 are given in Tables "2 " "

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WESTINGHOUSE CLASS 3 11.2-9 through II.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 0*,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 !!.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 also listed in Table II.2-13.

Several observations regarding the data presented in Tables II.2-1 through II.2-16 are worthy of note.

These observations may be summarized as follows:

1.

In North Anna Unit 2, calculated plant specific fast neutron (E > 1.0 MeV) fluence at the surveillance capsule center is in excellent agreement with measured data. The difference between the plant specific calculation and the measurement is less than 7%. Differences of this magnitude are well within the uncertainty of the experimental results.

2.

For both North Anna units, the fast neutron (E > 1.0 MeV) flux incident on the pressure vessel during Cycle 1 was, on the average,19% 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 5 of North Anna 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 1.90 relative to the design basis flux.

(In subsequent discussions, I

factors of fast neutron flux reduction, defined as the ratio of the design i

basis flux to the cycle-specific flux, will be quoted.)

The cycle 5 core j,

leading produced flux reduction factors ranging from 1.58 to 1,79 at the other azimuthal locations.

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WEST 1NGH0VSE CLASS 3 4.

In North Anna Unit 2, the low leakage core loading used for projection into the future (cycle 4) yielded a flux reduction factor of 1.58 at the peak flux location and factors ranging from 1.25 to 1.45 at the remaining azimuths.

5.

Comparing the flux reduction factors resulting from the low leakage core loadings in cycle 5 of North Anna Unit 1 and cycle 4 of North Anna Unit 2, one observes differences that are attributable to the varying burnups of l

the fuel assemblies in peripheral locations (see burnup data in Appendix

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A).

Graphical presentations of the plant specific fast neutron fluence at the peak location on the pressure vessel (O' azimuthal position) are shown in Figures 11.2-1 and 11.2-2 as a function of full power operating time for North Anna Units 1 and 2, respectively.

In regard to Figure II.2-1 and 11.2-2, the solid portions of the fluence curves are based directly on the cycle specific evaluations, as of September 30, 1985, presented in this report.

The dashed portions of these curves, however, involve a projection from September 30, 1985 to the respective license expiration dates.

As mentioned in Section II.1, the fluence projections are based on the cycle-averaged core power distribution of the current operating cycle (cycle 5 for Unit 1 and cycle 4 for Unit 2) and an assumed future capacity factor of 80%.

It should be noted that implement 6 tion of a more severe low leakage pattern than that of the current operating cycle 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.

In any event the RT assessment must be updated per PTS 10CFR50.61(b)(1) whenever, among other things, changes in core loadings significar.tly impact the fluence and RTPTS pr jections.

In Figures 11.2-3 and II.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-nnemsto go

WEST!NGH0VSE CLASS 3 license conditions.

In Figure 11.2-5, the relative radial variation of fast neutron flux and fluence dchin the pressure vessel wall is presented.

e 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-dimensional description of the fast neutron exposure of the pressure vessel wall can be constructed using the data given in Figure II.2-3 through II.2-6 along with the relation o(R, 0,Z) = #(0) F(R) G(Z) where:

o(R,0,Z)

Fast neutron fluence at location R, 0, Z within

=

the pressure vessel wall e (0)

Fast neutron fluence at azimuthal location e on

=

the pressure vessel inner radius from Figure 11.2-3 or II.2-4 Relative fast neutron flux at depth R into the F (R)

=

pressure vessel from Figure II.2-5 Relative fast neutron flux at axial position Z from G(Z)

=

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.

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WESTINGHOUSE CLASS 3 TABLE II.2-1 NORTH ANNA UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE I8)

VESSEL INNER RADIUS - O' AZIMUTHAL ANGLE Beltline Region 2

Elapsed Cumulative Fluence (n/cm )

Irradiation Irradiation Avg.2 Flux Plant Desiggg)

Interval Time (EFPY)

(n/cm -sec)

Specific Basis 10 18 18 CY-1 1.1 5.30 x 10 1.89 x 10 2.42 x 10 10 18 18 CY-2 1.9 6.52 x 10 3.46 x 10 4.05 x 10 10 18 18 CY-3 2.8 4.56 x 10 4.82 x 10 6.07 x 10 10 18 18 CY-4 3.8 5.10 x 10 6.37 x 10 8.13 x 10 10 18 18 CY-5(9/30/85)(C) 4.7 3.56 x 10 7.35 x 10 9.99 x 10 10 18 1.13 x 10.9 1

9/30/85 - CY-6(d) 5.3 3.56 x 10 8.03 x 10 I

10 19 19 CY 4/1/2018 *)

30.7 3.71 x 10 3.77 x 10 6.79 x 10 gey (a) Applicable to the peak azimuthal locations (0*, 90',180', 270') on the core beltline.

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(b) Design basis fast neutron flux = 6.78 x 10 n/cm -see at 2775 MWth (c) 9/30/85 is the date at which the current neutron fluences are defined.

(d) During CY-6, the core thermal power will be uprated to 2893 kWth.

Beyond 9/30/85 an 80% capacity factor is assumed.

(e) Exposure period from the onset of the uprating to the license expiration date.

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b WESTINGHOUSE CLASS 3 TABLE !!.2-2 NORTH-ANNA 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 Desigga)

Interval Time (EFPY)

(n/cm -sec)

Specific Basis 10 17 18 CY-1 1.1 2.54 x 10 9.06 x 10 1.10 x 10 10 18 18 CY-2 1.9 3.10 x 10 1.65 x 10 1.85 x 10 10 18 18 CY-3 2.8 2.18 x 10 2.30 x 10 2.77 x 10 10 18 18 CY-4 3.8 2.34 x 10 3.01 x 10 3.71 x 10 18 CY-5(9/30/85)(b) 4.7 1.77 x 10 3.50 x 10 4.56 x 10 10 18 10 18 18 9/30/85 - CY-6(c) 5.3 1,77 x 10 3.84 x 10 5.15 x 10 19; gey 10 19 CY 4/1/2018(d) 30.7 1.84 x 10 1.86 x 10 3.09 x 10 10 2

(a) Design basis fast neutron flux = 3.09 x 10 n/cm -sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.

(c) During CY-6, the core thermal power will be uprated to 2893 MWth. Beyond 9/30/85 an 80% capacity factor is assumed.

(d) Exposure period from the onset of the uprating to the license expiration date.

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13

I

-l WESTINGHOUSE CLASS 3 TABLE 11.2-3 NORTH ANNA UNIT 1-FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 30' AZIMUTHAL ANGLE 1

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 1.39 x 10 4.97 x 10 6.45 x 10 10 17 10 CY-2 1.9 1.66 x 10 8.98 x 10 1.08 x 10 10 18 18 CY-3 2.8 1.19 x 10 1.25 x 10 1.62 x 10 10 18 18 CY-4 3.8 1.10 x 10 1.59 x 10 2.17 x 10 CY-5(9/30/85)(b) 4.7 1.01 x 10 1J87 x 10 2.67 x 10 10 18 18 9/30/85-CY-6(c) 5.3 1.01 x 10 2.06 x 10 3.01 x 10 10 18 18 CY 4/1/2018(d) 30.7 1~.05 x 10 1.06 x 10 1.81 x 10 Rev 10 19 19, 10 2

(a) Design basis fast neutron flux = 1,81 x 10 n/cm -sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.

(c) During CY-6, the core thermal power will be uprated to 2893 MWth.

Beyond 9/30/85 an 80% capacity factor is assumed.

l (d) Exposure period from the onset of the uprating to the license expiration

date, i

D e

p 14 L

WESTINGHOUSE CLASS 3 TABLE 11.2-4 NORTH ANNA UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 45' AZIMUTHAL ANGLE Beltline Region 2

Elapsed Cumulative Fluence (n/cm )

Irradiation Irradiation Avg.2 Flux Plant Desigh)

Interval Time (EFPY)

(n/cm -sec)

Specific Basis 9

17 17 CY-1 1.1 9.48 x 10 3.38 x 10 3.92 x 10 10 17 17 CY-2 1.9 1.08 x 10 5.98 x 10 6.57 x 10 9

17 17 CY-3 2.8 7.75 x 10 8.30 x 10 9.86 x 10 9

18 18 CY-4 3.8 6.96 x 10 1.04 x 10 1.32 x 10 9

18 18 CY-5(9/30/85)(b) 4.7 6.98 x 10 1.23 x 10 1.62 x 10 9

19 18 9/30/85 - CY-6(c) 5.3 6.98 x 10 1.37 x 10 1.83 x 10 9

18 19 CY 4/1/2018(d) 30.7 7.28 x 10 7.20 x 10 1.10 x 10

.Rev.

10 2

(a) Design basis fast neutron flux = 1.10 x 10 n/cm -sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.

(c) During CY-6, the core thermal power will be uprated to 2893 MWth.

Beyond 9/30/85 an 80% capacity factor is assumed.

(d) Exposure period from the onset of the uprating to the license expiration date.

~

l 2322s/010M410 15

WESTINGHOUSf CLASS 3 TABLE 11.2-5 NORTH ANNA UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 15' SURVEILLANCE CAPSULE CENTER Beltline Region 2

Elapsed Cumulative Fluence (n/cm )

Irradiation Irradiation Avg.2 Flux Plant Desigga) capsule Interval Time (EFPY)

(n/cm -sec)

Specific Basis Data 10 18 18 CY-1 1.1 8.79 X 10 3.13 x 10 3.82 x 10

(,)

11 19 19 CY-2 1.9 1.07 x 10 5.72 x 10 6.39 x 10 10 18 18 CY-3 2.8 7.47 x 10 7.95 x 10 9.59 x 10 10 19 19 CY-4 3.8 8.08 x 10 1.04 x 10 1.28 x 10 10 19 19 CY-5 (9/30/85)(b) 4.7 6.04 x 10 1.21 x 10 1.58 x 10 10 19 19 9/30/85 - CY-6(C) 5.3 6.04 x 10 1.32 x 10 1.78 x 10

-CY 4/1/2018(d) 30.7 6.30 x 10 6.37 x 10 1.07 x 1020 'Rev.3 10 19 11 2

(a) Design basis fast neutron flux = 1.07 x 10 n/cm -sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.

(c) During CY-6, the core thermal power will be uprated to 2893 MWth.

Beyond 9/30/85 an 80% capacity factor is assumed.

(d) Exposure period from the onset of the uprating to the license expiration date.

(e) An error in the reported data of Reference 5 has precluded its use in the comparison of measured and calculated surveillance capsule fluence.

4 b

WESTINGHOUSE CLASS 3 TABLE II.2-6 NORTH ANNA 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 Desigg")

Interval Time (EFPY)

(n/cm -sec)

Specific Basis 10 18 18 CY-1 1.1 5.60 x 10 2.00 x 10 2.43 x 10 10 18 18 CY-2 1.9 6.77 x 10 3.62 x 10 4.06 x 10 10 18 18 CY-3 2.8 4.90 x 10 5.09 x 10 6.10 x 10 10 18 18 CY-4 3.8 4.64 x 10 6.50 x 10 8.15 x 10 CY-5(9/30/85)(b) 4.7 4.04 x 10 7.61 x 10 1.00 x 10 10 18 19 9/30/85 - CY-6(C) 5.3 4.04 x 10 8.38 x 10 1.13 x 10 10 18 19 CY 4/1/2018(d) 30.7 4.21 x 10 4.21 x 10 6.81 x 10 10 19 19 10 2

(a) Design basis fast neutron flux = 6.80 x 10 n/cm -see at 2775 MWth

~

(b) 9/30/85 is the date at which the current neutron fluences are defined.

(c) During CY-6, the core thermal power will be uprated to 2893 MWth. Beyond 9/30/85 an 80% capacity factor is assumed.

(d) Exposure period from the onset of the uprating to the license expiration date.

l l

nn'* *

• 17 l

WESTINGHOUSE CLASS 3 TABLE 11.2-7 NORTH ANNA UNIT 1 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 1.1 3.86 x 10 1.38 x 10 1.66 x 10 10 18 18 CY-2 1.9 4.54 x 10 2.47 x 10 2.78 x 10 10 18 18 CY-3 2.8 3.22 x 10 3.43 x 10 4.17 x 10 10 18 18 CY-4 3.8 2.90 x 10 4.31 x 10 5.58 x 10 10 18 18 CY-5(9/30/85)(b) 4.7 2.80 x 10 5.08 x 10 6.86 x 10 10 18 18 9/30/85 - CY-6(C) 5.3 2.80 x 10 5.61 x 10 7.75 x 10 10 19 l9 CY 4/1/2018(d) 30.7 2.92 x 10 2.50 x 10 4.66 x 10 Rev..

10 2

(a) Design basis fast neutron flux = 4.65 x 10 n/cm -sec at 2775 MWth

~

(b) 9/30/85 is the date at which the current neutron fluences are defined.

(c) During CY-6, the core thermal power will be uprated to 2893 MWth.

Beyond 9/30/85 an 80% capacity facter is assumed.

(d) Exposure period from the onset of the uprating to the license expiration date.

G i m. m m eio 18

WESTINGHOUSE CLASS 3 TABLE 11.2-8 NORTH ANNA UNIT 1 FAST NEUTRON (E > 0 MeV) EXPOSURE AT THE 45' SURVEILLANCE 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 18 18 CY-1 1.1 3.08 x 10 1.10 x 10 1.31 x 10 10 18 18 CY-2 1.9 3.51 x 10 1.94 x 10 2.19 x 10 10 18 18 CY-3 2.8 2.49 x 10 2.69 x 10 3.29 x 10 10 18 18 CY-4 3.8 2.24 x 10 3.3) x 10 4.40 x 10 10 18 18 CY-5(9/30/85)(b) 4.7 2.24 x 10 3.98 x 10 5.41 x 10 10 18 18 9/30/85-CY-6(C) 5.3 2.24 x 10 4.41 x 10 S.11 x 10 10 19 19 CY 4/1/2018(d) 30.7 2.34 x 10 2.32 x 10 3.68 x 10 Rev.

10 2

(a) Design basis fast neutron flux = 3.67 x 10 n/cm -see at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.

(c) During CY-6, the core thermal power will be uprated to 2893 HWth. Beyond 9/30/85 an 80% capacity factor is assumed.

(d) Exposure period from the onset of the uprating to the license expiration date.

4 I

nuvo,om,o 19

WESTINGHOUSE CLASS 3 TABLE 11.2-9 1

NORTH ANNA UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 0' AZIMUTHAL ANGLE (a)

Beltline Region 2

Elapsed Cumulative fluence (n/cm )

Irradiation Irradiation Avg.2 Flux Plant Desigfb)

Interval Time (EFPY)

(n/cm -sec)

Specific Basis 10 18 18

)

CY-1 1.0 5.30 X 10 1.73 x 10 2.22 x 10 10 18 18 CY-2 1.6 6.54 x 10 2.97 x 10 3.50 x 10 10 18 18 CY-3 2.7 3.87 x 10 4.26 x 10 5.75 x 10 10 18 18 CY-4(9/30/85)(C) 3.5 4.30 x 10 5.31 x 10 7.41 x 10 10 18 18 9/30/85 - CY-5(d) 4.4 4.30 x 10 6.56 x 10 9.38 x 10 10 19 19 CY 8/21/2020(*)

31.4 4.48 x 10 4.47 x 10 6.96 x 10 (a) Applicable to the peak azimuthal locations (0*, 90*,180*, 270') on the cor; beltline.

10 2

(b) Design basis fast neutron flux = 6.78 x 10 n/cm -see at 2775 MWth (c) 9/30/85 is the date at which the current neutron fluences are defined.

(d) During CY-5, the core thermal power will be uprated to 2893 MWth.

Beyond 9/30/85 an 80% capacity factor is assumed.

4 (e) Exposure period from the onset of the uprating to the license expiration date.

2222' " "

20 l

WESTINGHOUSE CLASS 3 TABLE II.2-10 NORTH ANNA UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER P.ADIUS - 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 18 CY-1 1.0 2.54 X 10 8.29 x 10 1.01 x 10 10 18 18 CY-2 1.6 3.11 x 10 1.42 x 10 1.60 x 10 10 18 18 CY-3 2.7 2.12 x 10 2.12 x 10 2.62 x 10 10 18 18 CY-4 (9/30/85)(b) 3.5 2.15 x 10 2.65 x 10 3.38 x 10 10 18 4,'28 x 1018 9/30/85 - CY-5(C) 4.4 2.15 x 10 3.27 x 10 10 19 19 CY 8/21/2020(d) 31.4 2.24 x 10 2.24 x 10 3.17 x 10 10 2

(a) Design basis fast neutron flux = 3.09 x 10 n/cm -sec at 2775 MWth (b) 9/30/85 is the date at which the curr it neutron fluences are defined.

(c) During CY-5, the core thermal power will be uprated to 2893

.th.

Beyond 9/30/85 an 80% capacity factor is assumed.

(d)

Exposure period from the onset of the uprating to the license expiration date.

l l

4 i

8"'*"""

21

WESTINGHOUSE CLASS 3_

TABLE 11.2-11 NORTH ANNA UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 30' AZIMUTHAL ANGLE Beldine Region 2

Elapsed Cumulative Fluence (n/cm )

Irradiation Irradiation Avg.2 Flux P1 ant Desigg,).

Interval Time (EFPY)

(n/cm -sec)

Specific Basis 10 17 17 CY-1 1.0 1.39 X 10 4.54 x 10 5.92 x 10 10 17 17 CY-2 1.6 1.68 x 10 7.72 x 10 9.35 x 10 10 18 18 CY-3 2.7 1.41 x 10 1.24 x 10 1.54 x 10 10 18 18 CY-4(9/30/85)(b) 3.5 1.25 x 10 1.54 x 10 1.98 x 10 10 18 18

~

9/30/85 - CY-5(C) 4.4 1.25 x 10 1.91 x 10 2.50 x 10 10 19 19 C'l-5-8/21/2020(d) 31.4 1.30 x 10 1.3 x 10 1.85 x 10 10 2

(a) Design basis fast neutron flux = 1.81 x 10 n/cm -sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.

(c) During CY-5, the core thermal power will be uprated to 2893 MWth. Beyond 9/30/85 an 80% capacity factor is assumed.

(d) Exposure period from the onset of the uprating to the license expiration date.

e 2322s 4 10444 10 22

WESTINGHOUSE CLASS 3 f-TABLE 11.2-12 NORTH ANNA UNIT 2 FAST NEUTRON (E > 0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 45' 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 9

17 17 CY-1 1.0 9.37 X 10 3.07 x 10 3.60 x 10 10 17 17 CY-2 1.6 1.11 x 10 5.17 x 10 5.68 x 10 9

17 17 CY-3 2.7 9.73 x 10 8.39 x 10 9.33 x 10 CY-4(9/30/85)(b) 3.5 8.80 x 10 1.06 x 10 1.20 x 10 9

18 18 9/30/85-CY-5(C) 4.4 8.80 x 10 1.31 x 10 1.52 x 10 9

18 18 9

18 19 CY-5-8/21/2020(d) 31.4 9.17 x 10 9.13 x 10 1.12 x 10 10 2

(a) Design basis fast neutron flux = 1.10 x 10 n/cm -sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.

{

(c) During CY-5, the core thermal power will be uprated to 2893 MWth. Beyond 9/30/85 an 80% capacity factor is assumed.

(d) Exposure period from the onset of the uprating to the license expiration date.

i' l

88 " ""*"

23

~

WESTINGHOUSE CLASS 3 TABLE 11.2-13 NORTH ANNA UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 15' SURVEILLANCE CAPSULE CENTER Beltline Region 2

Elapsed Cumulative Fluence (n/cm )

Irradiation Irradiation Avg.2 Flux Plant Desigg)

Capsule Interva' Time (EFPY)

(n/cm -sec)

Specific Basis Data 10 18 18 18 CY-1 1.0 8.78 X 10 2.87 x 10 3.50 x 10 2.70 x 10 11 18 18 CY-2 1.6 1.08 x 10 4.92 x 10 5.53 x 10 10 18 18 CY-3 2.7 7.22 x 10 7.31 x 10 9.07 x 10 10 18 19 CY-4(9/30/85)(b) 3.5 7.38 x 10 9.13 x 10 1.17 x 10 10 19 19 9/30/85 - CY-5(c) 4.4 7.38 x 10 1.13 x 10 1.48 x 10 10 19 20 CY 8/21/2020(d) 31.4 7.69 x 10 7.68 x 10 1.1 x 10 11 2

(a) Design basis fast neutron flux = 1.07 x 10 n/cm -sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.

(c) During CY-5, the core thermal power will be uprated to 2893 MWth. Beyond 9/30/85 an 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 and dosimeter location reported in Reference 6.

6 23223roiosas to 24

WESTINGHOUSE CLASS 3 TABLE II.2-14 NORTH ARNA UNIT 2 FAST NEUTRON (E > 0 MeV) EXPOSURE AT THE 25' SURVEILLANCE CAPSULE CENTER Beltlino Region 2

Elapsed Cumulativa Fluence (n/cm )

Irradiation Irradiation Avg.2 Flux Plant Desigfa)

Interval Time (EFPY)

(n/cm -sec)

Specific Basis 10 18 18 CY-1 1.0 5.58 X 10 1.83 x 10 2.22 x 10 10 18 19 CY-2 1.6 6.78 x 10 3.11 x 10 3.51 x 10 10 18 18 CY-3 2.7 5.54 x 10 4.95 7. 10 5.77 x 10 10 18 18 CY-4 (9/30/85)(b) 3.5 4.93 x 10 6.16 x 10 J.44 x 10 9/30/85-CY-5(c) 4.4 4.93 x 10 7.59 x 10 9.41 x 10 10 18 18 10 19 19 CY 8/21/2020(d) 31.4 5.14 x 10 5.13 x 10 6.98 x 10 10 2

(a) Design basis fast neutron flux = 6.80 x 10 n/cm -sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.

t' (c) During CY-5, the coro thermal power will be uprated to 2893 MWth. Beyond 9/30/85 an 80% capacity factor is assumed.

l l

(d) Exposure period from the onset of the uprating to the license expiration i

date.

l l

e

WESTINGHOUSE CLASS 3 TABLE II.2-15 NORTH ANNA 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 Desigf a'/

Interval Time (EFPY)

(n/cm -sec)

Specific Basis 10 18 18 CY-1 1.0 3.83 X 10 1.25 x 10 1.52 x 10 10 18 18 CY-2 1.6 4.59 x 10 2.12 x 10 2.40 x 10 10 18 18 CY-3 2.7 3.92 x 10 3.42 x 10 3.94 x 10 10 18 18 CY-4 (9/30/85)(b) 3.5 3.50 x 10 4.28 x 10 5.09 x 10 9/30/M - CY-5(c)"

4.4 3.50 x 10 5.30 x 10 6.43 x 10 10 18 18-CY 8/21/2020(d)

/1.4 3.65 x 10 3.64 x 10 4 78 x 10 10 19 19 10 2

(a) Design basis fast neutron flux = 4.65 x 10 n/cm -sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are dafined.

(c) During CY-5, the core thermal power will be uprated to 2893 HWth, i3eyond 9/30/85 an 80% capacity factor is assumed.

(d)

Exposure period from the onset of the uprating to the license expiration date.

L 2'"'"""

26 l

WESTINGHOUSE CLASS 3 TABLE II.2-16 NORTH ANNA UNIT 2 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 45' SURVEILLANCE 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 1.0 3.04 X 10 9.96 x 10 1.20 x 10 10 18 18 CY-2 1.6 3.60 x 10 1.68 x 10 1.90 x 10 10 18 18 CY-3 2.7 3.15 x 10 2.73 x 10 3.11 x 10 CY-4(9/30/85)(b) 3.5 2.84 x 10 3.43 x 10 4.02 x 10 10 18 18 9/30/85 - CY-5(c) 4.4 2.84 x 10 4.25 x 10 5.08 x 10 10 18 18 W 5 - 8/21/2020(d) 31.4 2.97 x 10 2.95 x 10 3.77 x 10 10 19 19 10 2

l-(a) Design basis fast neutron flux = 3.67 x 10 n/cm -sec at 2775 MWth (b) 9/30/85 is the date at which the current neutron fluences are defined.

(c) During CY-5, the core thermal power will be uprated to 2893 MWth. Beyond 9/30/85 an 80% capacity factor is assumed.

(d) Exposure period from the onset of the uprating to the license expiration date.

O e

l 27

16003 1 O'

(MAJOR AXIS)

~

15' (CAPSULES V,X)

I 25' (CAPSULES Y,W,U) p

/

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35' (CAPSULES Z,T)

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g 45' (CAPSULE S) l

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xxxxxy j j

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Figure 11.1-1. North Anna Reactor Geometry 28 1

16o03 2 1020 7

5 0

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LICENSE 9/30/85 EXPIRATION I

I I

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IO 20 30 40 50 60 70 Rev.3 OPERATING TIME (EFPY) l.

Figure 11.2-1. North Anna Unit 1 Maximum Fast Neutron (E > 1.0 MeV)

Fluence at the Pressure VesselInner Radius as a Function of Full Power Operating Time 29 l

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16003 3 1020 7

5 0*

n N

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Figure 11.2-2. North Anna Unit 2 Maximum Fast Neutron (E > 1.0 MeV) t Fluence at the Pressure Vessel Inner Radius as a Function of Full Power Operating Time 30

l 16003 4 io20

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l 7

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PROJECTED 5

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10 20 30 40 50 60 70 AZIMUTHAL ANGLE (DEG.)

Figure l1.2 3. North Anna Unit 1 Maximum Fast Neutron (E > 1.0 MeV)

Fluence at the Pressure VesselInner Radius as a Function of Azimuthal Angle 31

16003 5 1020

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ACTUAL

---

PROJECTED 5

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LICENSE EXPIRATION 7

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3 9/30/85 10 18 I'

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Figure ll.2-4. North Anna Unit 2 Maximum Fast Neutron (E > 1.0 MeV)

Fluence at the Pressure VesselInner Radius as a Function of Azimuthal Angle 32

16003 6 10 7

s 3

199.39 dz I.0 204.66 3

t 7

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CLAD g

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3 REACTOR VESSEL OR l

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0.01 195 199 203 207 211 215 219 223 RADIUS (em)

Figure l1.2-5. North Anna Units 1 and 2 Relative Radial Distribution of Fast Neutron (E > 1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall 33 l

l

16003 7 1.0 7

5 3

W OZW DJ O.1 I

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i00 200 300 OISTANCE FROM CORE MIOPLANE (cm)

Figure l1.2-6. North Anna Units 1 and 2 Relative Axial Variation of Fast Neutron (E > 1.0 MeV) Flux and Fluence Within the Pressure Vessel Wall 34

WESTINGHOUSE CLASS 3 SECTION III MATERIAL PROPERTIES For the RT calculation, the best estimate copper and nickel chemical PTS composition of the reactor vessel beltline material is necessary.

The material properties for the North Anna Units 1 and 2 beltline region will be 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 reactor vessel (shell material including welds, heat affected zones, and 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 selection of the most limiting material with regard to radiation damage."

Figures III.1-1 and III.1-2 identify and indicate the location of all beltline region materials for the North Anna Units 1 and 2 reactor vessels.

III.2 DEFINITION AND SOURCE OF MATERIAL PROPERTIES FOR ALL VESSEL LOCATIONS Material property values for the forgings, which have been docketed with the NRC in Reference 7, were derived from vessel fabrication test certificate results. The property data for the welds have also been docketed with the NRC in Reference 7, however, the weld properties cannot be used in the same direct manner as the properties for the forgings.

1 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 l'

by the variability of copper concentration within the weldments.

2322s/0106H 10 35

WEST!NGHOUSE CLASS 3 For each weld in the North Anna Units 1 and 2 beltline region, a material data search was performed using the WOG Reactor Vessel Beltline Region Weld Metal Data Base. The WOG data base, which was developed in 1984 and is continually being updated, contains information from weld qualification records, surveillance capsule reports, the B&W report BAW-1799(8], and the Materials Properties Council (MPC) and the NRC Mender MATSURV data bases.

Searches on the WOG Data Base were performed for materials having the identical weld wire heat number as those in the North Anna vessels, with any flux 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.

III.3

SUMMARY

OF PLANT-SPECIFIC MATERIAL PROPERTIES A summary of the pertinent chemical and mechanical properties of the beltline region shell and weld materials of the North Anna Units 1 and 2 reactor vessels are respectively given in Tables III.3-1 and III.3-2 along with the references for this information. Although phosphorus is no longer used in the calculation of RT with respect to the PTS rule (1), it is given for NDT reference since it is used in the Regulatory Guide 1.99, Revision 1 trend l

curve (9).

The initial RTNDT shown for all North Anna Units 1 ano 2 shells and weldments are the actus.1 values.

i The data in Tables III.3-1 and 111.3-2 are used to evaluate the RTPTS values for the North Anna Unit 1 and 2 reactor vessels.

nn,meus n 3g

WESTINGHOUSE CL'4SS 3 TABLE III.3-1 NORTH ANNA UNIT 1 REAC. TOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Cu Ni P

I (Wt.%)

(Wt.%) (Wt.%)

('F)

Source Intermediate Shell 04 Her.t 0.12 0.82 0.010 17 Ref. [7]

990311/298244 Circumferential Weld - Intermed, to Lower Shell WO4, Heat 25531, Smit 89, Flux 1211:

0.086 0.11 0.02 19 WOG Material Data Base Lower Shell 03 Heat 0.15 0.80 0.009 38 Ref. (7) 990400/292332 i

I swuctosss so 37 l

7-.

=

WESTINGHOUSE CLASS 3

]

3 TABLE 111.3-2 NORTH ANNA UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Cu Ni P

I (Wt.%)(Wt.%)

(Wt.%)

(*F).

Source Intermediate Shell 04 Heat 0.09 0.83 0.011 75 Ref. (7) 990496/292424 Lower Shell 03 Heat 0.13 0.83 0.013 56 Ref. [7]

990533/297355 Circumferential Weld - Intermed to Lower Shell WO4, Heat 716126, Grau Lo 26 0.065 0.044 0.015

-48 WOG Material Data Base ner.mo u io 38 I

FIGURE 111.1-1 l-Identification and Location of North Anna Unit No.1 Reactor Vessel Beltline Material H

d Weld Seam WO5 no*

6 coat l

C Forginq 04 l

~ ~ wo-o-

w l

144.0" g

B 5

1 C oR E (17

,o.

d Weld Seam WO4 no-1 Coat Foraina 03

- 30 o.

$o h

ir t

8 l

49.3"

,o.

I r

g i

39

\\

FIGURE III.1-2 1

Identification and Location of North Anna Unit No. 2 Reactor Vessel Beltline Material

/

170*

Weld Seam V05 gs -

Forcinq 04 o.

150' t

144,0,,

d 3

5 v

1 CORE (g,.

,3 vo' veTd seam V04

\\

\\

Fordi"9 p

o' 160'

\\

1 s

9 o.r i

st' r

g 1

l l

l 40 1

WESTINGHOUSE CLASS 3 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 North Anna Units 1 and 2 can now be PTS determined.

IV.1 STATUS OF REACTOR VESSEL INTEGRITY IN TERMS OF RT VERSUS FLUENCE PTS RESULTS values were generated for Using the prescribed PTS Rule methodology, RTPTS all beltline region materials of the North Anna 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 all beltline region materials for both units.

Figures IV.1-1 and IV.1-2 present the RT values for the limiting shell PTS and circumferential weld of the North Anna Units 1 and 2 vessels in terms of RT versus fluence

• curves.

The curves in these figures can be used:

PTS o

to provide guidelines to evaluate fuel reload options in relation to the l

NRC RT Screening Criterion for PTS (i.e., RTPTS values can be PTS I

readily projected for any options under consideration, provided fluence is known),and o

to show the current (4.7 EFPY for North Anna 1 and 3.5 EFPY for North Anna

2) and end-of-license (30.7 EFPY for North Anna 1 and 31.4 EFPY for North Anna 2) RT values using actual and projected fluence.

! ev.3 R

PTS l

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

1 mwo,csu n 41 l

WESTINGHOUSE CLASS 3 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 lower shells are the governing locations for both reactor vessels relative to PTS. All the RT values PTS remain below the NRC screening values for PTS using the projected fluence value at license values through license expiration.

The most limiting RTPTS expiration is 233'F for the lower shell of Unit 1 and 237*F for the lower Rev.3 shell of Unit 2.

e 2322stottesa 10 l

WESTINGHOUSE CLASS 3 TABLE IV.1-1 RT VALUES FOR NORTH ANNA UNIT 1 PTS RT Values ('F)

PTS Present End-of-License Screening Location Vessel Material (4.7 EFPY)

(30.7 EFPY)

Criteria-1 Intermediate shell 04 139 181 270 2

Intermediate to lower shell 98 118 300 circumferential weld WO4 l

I 3

Lower shell 03 180 233 270 lRev.3 l'

l*

l rus.ioia. io 43

{

WESTINGHOUSE CLASS 3 TABLE IV.1.

RTPTS. VALUES FOR NORTH ANNA UNIT 2 RT Values (*F)

PTS Present End-of-License Screening location Vessel Material (3.5 EFPY)

(31.4 EFPY)

Criteria 1

1,1termediate shell 04 172 210 270 2

Lower shell 03 179 237

-270 3

Intermediate to icwer shell 20 36 300 circumferential weld WO4

= -

9 i

l f

un.io,one se 44

FIGURE IV.1-1 NORTH ANNA UNIT 1 - RT CURVES PER PTS RULE MElH00 [1] DOCKETED PTS BASE MATERIAL AND WOG DATA BASE MEAN WELD MATERIAL PROPERTIES 320 NRC SCREENING VALUE FOR CIRCUMFERENTIAL WELDS 280 NRC SCREENING VALUE FOR FORGINGS 260 240 220 200 Limiting Forging 180 (Lower Shell) m p

160 y

Q-140 i--

fr 120 100 W

circumferential veld 80 60 40 O

Present (4.7 EFPY) 20

~

m Lloanes E9 ration (30.7 EFPY)

?

0 Rev.3 1018 1019 390 FLUENCE. N/CMXX2 l

\\

l i

{!l y

S E

(

D I

2 T

0 R

i 1

E P

e O

e R

P i

L A

e I

]R 1 E

[T e

)

A M

A DM O

i HD H

TL S

EE M W R

)

n YP EN W

F LA 0

E UE RM R

2 W

4 S E G

e N

C 1

T S P A 3

I

(

B G

/

R S -

M

) n EA D

S Y o PT L

A E -

N Pi Ft SD W

G O

VG L -

8 R r

E a N

E 1

I T

5. i RO A

T 0 U I

p I

UW I

1 C

T -

M E

(3 x a

E D

N I

SN E -

L a

te TA R

P E

S ns F -

G -

a E

en TL RA M

N -

D C

se g

I L

4 ec U -

s I

D G

E ri R

C W

E R

R PL 0

U O

I 2T L

L A

C F

9 TM A

F em R

R I

I NE O -

O T

US F

F -

N E

A AB E -

F R

N U

U -

E ND L

L F

AE A - A -

M T

V V

U C

HE TK G -

G.

R RC N

N I

C I

I O0 N0 N

N E

E -

E -

E 2

R R

C -

C 1

S S

V C

C 8

I R

R -

N N

1 E

0 R

U 1

G 0

0 0 0

0 0

0 0

0 0

0 0 0 0 0 0

8 6 4

2 0

8 4

2 0

8 8 4 2

I F

3 2 2 2 2

2 1

1 1

1 n [o S"

~

~

1l

WESTINGHOUSE CLASS 3 SECTION V CONCLUSIONS AND RECOMMENDATIONS values for the Calculations have been completed in order to submit RTPTS North Anna 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 evaluation and a reactor vessel material study in order to determine the RT values.

PTS Detailed fast neutron exposure evaluations using plant specific cycle by cycle core power distributions and state-of-the-art neutron transport methodology have been completed for the North Anna Units 1 and 2 pressure vessels.

Explicit calculations were performed for the operating cycles of both units as of September 30, 1985. For both units, projection of the fast neutron exposure beyond the current operating cyclo was based on continued implementation of low leakage fuel management similar to that employed during cycle 5 for Unit 1 and cycle 4 for Unit 2.

In regard to the low leakage fuel management already in place at the North Anna Units, the cycle-specific evaluations have demonstrated that for the low leakage case (cycle 5 in Unit 1 and cycle 4 in Unit 2) the peak fast neutron flux at the 0* azimuthal position has been reduced by a facte of 1.90 in Unit 1 and a factor of 1.58 in Unit 2 relative to the flux based on the design basis core power distribution.

This location represents the maximum fast neutron flux incident on the reactor 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 l

presented above.

(,

It should be noted that significant deviations from the low leakage schemo l

already in place will affect the exposure projections beyond the current operating cycle.

A move toward a more severe form of low leakage (lower relative power on the periphery) would tend to reduce the projection. On the l

other hand, a relaxation of the loading pattern toward higher relative power l

47

WESTINGHOUSE CLASS 3 on the core periphery would increase the projections beyond those reported.

As each future fuel cycle evolve. the loading patterns should be analyzed to determine their potential impac5 an vessel and capsule exposure.

The fast neutron fluence values from the plant specific calculations have been compared directly with measured fluence lev'els derived from neutron dosimetry contained in the surveillance capsules withdrawn from each of the North Anna Units. For Unit 1, an error in the reported measured fluence precluded its use. For Unit 2, the ratio of calcul'ated to measured fluence 4s 0.94 for the 15' surveillance capsule withdrawn following cycle 1.

This excellent agreereent between calculation and measurement supports the use of this analytical approach to perform a plant specific evaluations for the North Anna reactors.

Material property values for the North' Anna Units 1 and 2 reactor vessel beltline region components were determined. The pertinent chemical and mechanical properties for the shells remain the same as those that have been docketed with the NRC in Reference 7.

The weld material pr perties are determined using the WOG Meterials Data Base.

Using the prescribed PTS Rule methodology, RT values were generated for PTS all beltline region materials of the North Anna Units 1 and 2 reactor vessels as a function of several fluence values and pertinent vessel lifetimes. For both reactor vessels, all the RT values remain below the NRC screening PTS values for PTS using the projected fluence exposure through license expiration.

The most limiting values at end-of-license (30.7 EFPY for North Anna 1 and 31.4 EFPY for North Anna 2) are 233'F and 237*F for the lower Rev.3 shells of Unit 1 and Unit 2, respectively.

e em,aus to 48 l

WESTINGHOUSE CLASS 3 SECTION VI REFERENCES 1.

Nuclear Regulatory Commission,10CFR Part 50, "Analysis of Potential Pressurized Thermal Shock Events," Federal Register, Vol. 50, No. 141, 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 C.sta Library Collection DLC-76." Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P, Cross-Section Library for Light Water 3

Reactors.

4.

Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology -

to be published.

5.

BAW-1E39. "Analysis of Capsulo V from the Virginia Electric and Power i

Company North Anna Unit No. 1 Reactor Vessel Materials Surveillance Program," A.

t.. Lowe, Jr., et al., May 1981.

l l

6.

BAW-1794, "Analysis of Capsule V from the Virginia Electric and Power i

Company North Anna Unit No. 2 Reactor Vessel Materials Surveillance Program," A. L. Lowe, Jr., et al., October 1983.

7.

Letter from C. M. Stallings of Virginia Power to H. R. Denton of the NRC, Serial No. 5018, December 11, 1978.

l 8.

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

l l*

49 l

t d

WESTfNGHOUSE CLASS 3 9.

"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.

10. Letter from K. L. Basehore of Virginia Power Company 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.

un.ioio.u io 50

WESTINGHOUSE CLASS 3 APPENDIX A POWER DISTRIBUTIONS Core power distributions used in the plant specific fast neutron exposure analysis of the North Anna pressure vessels were derived from the measured fuel assembly and cycle burnup data supplied by Virginia Power (10].

The beginning-of-cycle (BOC) and end-of-cycle (E0C) fuel assembly burnups, based on incore flux maps, were provided for selected peripheral fuel assembly locations for each of the previous cycles of operation.

In addition, estimated data was provided for the current cycle of operation (cycle 5 of Unit 1 and cycle 4 of Unit 2). Table A-1 shows the North Anna Unit 1 fuel assembly and cycle burnups for Cycles 1 through 5.

Similar dats for North Anna Unit 2 are shown in Table A-2.

(The. fuel assembly locations in the North Ann'a cores are numbered according to Figure A-1.)

Cycle-averaged relative assembly powers for each cycle were computed using the following relation Relative Assembly Power = EOC Assembly Burnuo - BOC Assembly Burnuo Cycle Burnup and are shown in Tables A-3 and A-4 for North Anna Units 1 and 2, respectively. The cycle-averaged relative assembly powers representing the design basis core power distribution are also shown in Tables A-3 and A-4.

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 distribution data for these interior assemblies are not given in Tables A-3 and A-4.

A-1

WESTlNGHOUSE CLASS 3 In each of the adjoint evaluations, within assembly spatial gradients have been superimposed on the average assembly power levels. For peripheral assembly locations 1, 2, 3, 4, 5 and their symmetric partners, these spatial gradients also incit adjustments to account for analytical deficiencies that tend to occur near the boundaries of the core region.

f 9

i 1

e en A-2 r

WESTINGHOUSE CLASS 3 TABLE A-1 NORTH ANNA UNIT 1 BEGINNING-0F-CYCLE AND END-0F-CYCLE FUEL ASSEMBLY BURNIJPS Fuel Assembly Burnup (WWD/MTU)

Fuel Cycle *)

l 1(15892) 2(10711) 3(13335) 4(13478) 5(13000)

Assembly BOC EOC BOC EOC BOC EOC BOC EOC BOC EOC 1

0 11500 0

9570 0

9040 0

9920 9922 16202 2

0 9095 0

7845 26466 31468 16299 22625 30715 34608 3

0 13550 0

10895 0

12105 0

12120 16867 25462 4

0 9435 0

7953 22432-28195 27414 32475 31976 36424 5

0 10458 0

8165 24721 30658 28455 33675 26933 32090 6

0 15070 17095 27140 7992 21585 11819 25375 17518.29523 7

0 15428 0

12530 0

14045 0

14180 0

12705 8

0 16810 19120 28850 12484 27413 14047 29058 13050 28565 9

0 16465 18795 28688 0

15898 0

15810 0

14790 10 0

15560 0

12483 0

14448 0

13005 0

13435 11 0

17045 15428 26468 0

16295 0

15853 0

15637 12 0

14200 18558 27125 7979 20795 9004 20950 12139 23738 l

l (a) The number in parentheses beside the cycle number is the fuel cycle length in MWD /MTV.

t l

n n.c o.u."

A-3

WESTINGHOUSE CLASS 3 TABLE A-2

-NORTH ANNA UNIT 2 BEGINNING-0F-CYCLE AND END-0F-CYCLE FUEL ASSEMBLY BURNUPS Fuel Assembly Burnup (MWD /MTU)

Fuel Cycle (a) 1(14494) 2(8436) 3(14717) 4(16000)

Assembly BOC E0C BOC E0C B0C EOC BOC E0C 1

0 10500 0

7510 6060 13685 13424 21998 2

0 8288 0

6145 23520 28638 17561' 24292 3

0 12325 0

8715 0

13700 18422 29624-4 0

8530 0

6253 9377 17560 18900 26484 5

0 9448 0

6645 6644 16075 18259 27092 6

0 13680 17310 25095 18485 30175 13684 29054 7

0 14020 0

9760 0

14875 0

16715 8

0 15383 17472 25438 21863 35763 16037 33669 9

0 14895 17130 25330 0

17445 0

18759 10 0

14035 0

93'/8 8717 23665 0

17269 11 0

15370 16775 24860 0

18420 0

19628 12 0-12820 8204 16878 6204 21983 13878 28673 (a) The number in parentheses beside the cycle number is the fuel cycle length in MWD /MTU.

e h

O WESTINGHOUSE CLASS 3 TABLE A-3 NORTH ANNA UNIT 1 CORE POWER DISTRIBUTIONS USED IN THE FLUENCE ANALYSIS Plant Specific Cycle Averaged Relative Assembly Power-Design Fuel Cycle Basis Assembly Relative Power 1

2 3

4 5-1 1.00 0.72 0.89 0.68 0.74 0.48 2

0.83 0.57 0.73 0.38 0.47 0.30 3

1.21 0.85 1.02 0.91 0.90 0.66 4

0.86 0.59 0.74 0.43 0.38 0.34-5 0.92 0.66 0.76 0.45 0.39 0.40 6

0.98 0.95 0.94 1.02 1.01 0.92 7

1.10 0.97 1.17 1.05 1.05 0.98 8

1.00 1.06 0.91 1.12 1.11 1.19 9

1.05 1.04 0.92 1.19 1.17 1.14 i.

10 1.08 0.98 1.17 1.08 0.97 1.03 11 1.06 1.07 1.03 1.22 1.18 1.20 12 0.95 0.89 0.80 0.96 0.89 0.89 t

l

~

l l

l' un. mens io A-5 l

WESTINGHOUSE CLASS 3

'1 TABLE A-4 NORTH ANNA UNIT 2 CORE POWER DISTRIBUTIONS USED IN THE FLUENCE ANALYSIS Plant Specific Cycle Averaged Relative Assembly Power-Design Fuel Cycle Basis Assembly Relative Power 1

2 3

4 1

1.00 0.72 0.89 0.52 0.54 2

0.83 0.57 0.73 0.35 0.42 3

1.21 0.85 1.03 0.93 0.70 4

0.86 0.59 0.74 0.56 0.47 5

0.92 0.65 0.79 0.64 0.55 6

0.98 0.94 0.92 0.79 0.96 7

1.10 0.97 1.16 1.01 1.05 8

1.00 1.06 0.94 0.94 1.10 9

1.05 1.03 0.97 1.19 1.17 l

10 1.08 0.97 1.11 1.02 1.08 11 1.06 1.06 0.96 1.25 1.23 12 0.95 0.89 1.03 1.07 0.93 l

l l

a nin.,ema in A-6

16003 8 O'

(MAJOR AXIS) g BAFFLE l[//

CORE BARREL 7

NN\\K I

2

\\

45-s N N N K' 6

7 3

4

\\

\\\\

s 8

9 10 5

/

l II I2 l

l Figure A 1. North Anna Units 1 & 2 Core Description for Power Distribution Map A-7

WESTINGHOUSE CLASS 3 APPENDIX B WELD CHEMISTRY Tables B.1-1 and B.1-2 provide the weld data output from the WDG Material Data Base. Given are the searches of all available data for the wire heat in the North Anna 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 RT analysis.

PTS Weld Chemistry Data Source and Plant:

Weight % of Copper Cu NRC Mender MATSURV Data Base NATSURV Materials Properties Council Data Base MPC Weight % of Nickel Ni Rotterdam RDM P

Weight % of Ph.)sphorous Surveillance Capsule SC Si Weight % of Silicon Ringhals 2 SSP North Anna 2 VGB North Anna 1

.VRA Weld Qualification WQ e

nnsmo 9_l

TABLE B.1-1 North Anna Unit 1 Intermediate to Lower Shell Circumferential Weld Chemistry From WOG Materials Data Base - Wire Heat I

Number 25531 ID WIRE WIRE FLUI FLUM WE'L DCHEM Cu N6 P

St FLANY DESCRIPTION HE AT T vF E TYPE LOT DATA EXM M E 0.260 SSP N0lILE TO INTER fi> ELL 0646 2553 SMIT 40 SMI T 809 1281 RDM.Wo WA SURVEILLANCE WI~LD 0728 25531 BMIT 40 SMIT 99 1281 VRA.SC O.006 0.810 0.020 0.350 SSP NO22LE 80 INTER SHELL VRA tiURVEILLANCE WELD rD;,

0.006000 O.110000 0.020u00 0.30".hM O.0000*4 0.000000 0.00sXMM O.063640

.een std.dev.

6 8

e 9

TABLE B.1-2 North Anna Unit 2 Intermediate to Lower Shell Circumferential Weld Chemistry From WOG Materials Data Base - Wire Heat Number 716126 SELECT REM)RT ID WIRE WERE

'FLUM FL UM WELDCHEN Cas NA P

St PLANT DESCRIPTION HEAT T YF E TYPE BOT DAfA BOUNCE 0630 786176 S 3MO LW320 26 RDrt,WQ O.068 0.030 0.082 0.237 0631 786126 93F10 L W 320 26 RDN.WO O.062 0.030 0.083 0.227 0632 786526 S 3MO L W320 26 RDN,WO O 079 0.040 0.089 0.227 e633 716826 63NO LW320 26 RDN WO O.064 0.040 0.018 0.225 0634 716826 S 3MO LW320 26 RDet,WQ O.060 0.000 0.086 0.898 U149 786526 53P90 LW320 26 Wie.Sc O.Oue 0.084 0.087 0.250 m

mean 0.069000 0.050667 0.015667 0.225667 etd.dow.

0.018662 0.024782 0.002582 0.020704

WESTINGHOUSE CLASS 3 APPENDIX C RT VALUES OF NORTH ANNA UNITS 1 AND 2 PTS REACTOR VESSEL BELTLINE REGION MATERIALS

-C.1 NORTH ANNA 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 Nortn Anna Unit i reactor vessel.

The RT values are calculated in accordance with the PTS rule, which is PTS 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 III.3-1 of the main report.

Location Vessel Material 1

Intermediate shell 04 2

Intermediate to lower shell circumferential weld WO4 3

Lower shell 03 1*

C.2 NORTH ANNA UNIT 2 l

values, as a function of both

. Tables C.2-1 through C.2-5 provide the RTPTS constant fluence and constant EFPY (assuming the projected fluence values),

for all beltline region materials of the North Anna Unit 2 reactor vessel.

The RT values are calculated in accordance with the proposed PTS rule, PTS which is Reference (1) in the main body of this report.

The vessel location numbers in the following tables corrcspond to the vessel materials identified below and in Table III.3-2 of the main report.

location Vessel Material 1

Intermediate shell 04 2

Lower shell 03 3

Intermediate to lower shell circumferential weld WO4 nn.m eme C-1

TABLE C.1-1 18 2

VALUES FOR NORTH ANNA UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIALS 9 FLUENCE n/cm RTp73 RT 1 Tves irtucues- = ------------- py3 e

I watus LOC netauri cu i nr i

= - - - - - - - - - -

1 i vna i.32

.ean.osos tv. i acTuat i s.m.

i.soe+ses vos.:

2 vna

.oes. tis.omoi to. I acTuat c.w. i. sot +ses os.:

a

-.------------------------ ---------------- ---=

g 3

vaa

.isi.soi.coe se. i actuat i s.m. i.ioe+tei sat.:

- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - - - - - - - - - - - - -

Notes:

B.M. = Base Material (Shell)

C.W. = Circumferential Weld Reference Temperature are in 'F g

TABLE C.1-2 18 2

RT VALUES FOR NORTH ANNA UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIALS 9 FLUENCE = 5.0 x 10 g

PTS ma ee I watur s tres artwesca RT LOC netassvi cu

................................---.....--....-......--- py3 1

vna I.121.e28.olo

17. I actual I e.m.

I.soE+1el 132.1

..-_......=_

2 vna i.oes. :.omo:

se. e actuat c.w.

i. soc +ses er.:

3

vna

.sse. son.ooes ae. e actuat i s.m.

i. soc +ses svi.:

e (a)

TABLE C.1-3 D

2 RT VALUES FOR NORTH ANNA UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIALS 9 FLUENCE =1.0 x 10 n/cm PTS

'l RT LOC tasset cu e est i e I

watus t Yves set w esca PTS l

1 I VRA B.128.e29.0101

17. I ACTUAL I e. 88.

l. TOE *208 144.I f

_.______._...______..==-

2

vea :.Oei. ins.020

... e actual i c.w.

i.iOr+20 sos.:

- =.......... _ _ _ _ -....... _. _

3 8 v** ' '*' 808 00*'

• • • 0^'

l t

O k

P e

- ~,

l l

TABLE C.1-4 RT VAMS MR MRTH ANNA UNIT 1 REACMR ESE BMINE MN 0 NNT H.7 M ME VAN p73

(

L(jC.setasers cu a ses e c: I e watur a swee snussectiRT

...._.........................................._........pys.

1 I Vea I.121.82.0:01 17.

aC7ual : 5.M.

1.74E+191 139.1 2

vna :.oes.ts!.o20 is.

actuat : c.w. i.74e+,,e too.:

3

vna

.ns. son.ooe se. i actuat i s.M.

s.74t+ise iso.:

e (JI

~ :

TABLE C.1-5 VALUES FOR NORTH ANNA UNIT 1 REACTOR VESSEL BELTLINE MATERIAL 9 LICENSE EXPIRATION (30.7 RTPTS l

RT

________._______.______..__.____.______________________P_T_S_

I PI

[

I VaLUE

TVPE IFLUENCE LOC _IPLaNTI CU l NI 1

vna

.t2:.eas.osos ti. i acunat i s.N.

.38E+20 181:

i i

2 : vea

.os. ::.oaos so.

AcTuat c.w.

.38E+20 118 t

3 : vna :.ns. son.cose as.

actuat e.N.

.38E+20 233.'

,t i Rev.3 l

l n

i em l

l O

e s

e p

e TABLE C.2-1 RT VALUES FOR NORTH ANNA UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS 9 FLUENCE = 1.0 x 1018,jc,2 PTS I

0b LOC IPLANil Cu 9 NI I P I I

I WAl

=-.-

_' '***' t*4't 1

9 vos

.oss' 838 0888 78- '

"C'"

2 vos

.i::.ezi. ossa

. i aCrust I a.m I. tot +1el 152.1

-=-----

3 8 WGs 8 878 05' * ~**-,,croat

.w.

i., t+ ion ts.:

,,eem.-O-----

O e

N O

A TABLE C.2-2 RT VALUES FOR NORTH Ai4NA UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS 9 FLUENCE = 5.0 x 1013,7 PTS ten erumsscas RT I

e vatus

.........................................................f.TS enasere cu a m ef LOC 1

Va5 :.oS

.83:.o89:

75. I aCTuas. I S.98.

e. Soc +fSe 17f.:

............_ -=

_ =--........._ =---=..

=-......

2

vas :.sas.sas.oisi ss. i actuas.

s.m. a. soc + e #7s.:

_===...==__--

3'

vas

.or.oss.ois

-4a. i actuat i c.w.

s. sos.ss zo.:

-.___ -=

n

..............=

3 9

0 0

TABLE C.2-3 I9 2

RT VALUES FOR NORTH ANNA UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS 9 FLUENCE = 1.0 x 10 n/cm PTS RTa LOC irtauri cu we ei I

vatut Tves artuenct:= -............... I.T.S l

I : VGs t.oSt. Sat.ott

75. I acTuat t 5.m. i. toe +2ol let.:

....= --=____......................_._- - - - - -=.........._..

2

vos

.33:.sas.osas ss. i acivat i s.m. s. toe +2o: isa.:

......................................=---

3 : vos :.or.csi.ois 4s. i actuat a c.w. s.toseaos 24.

_-......_...__.__ -- =_

+

O eW i

~

m Wd w

W wDd w

e.

m w

szw Eg 9

sidisi n%

m

=:

W

e.. 2 l

I.

5:

X: N X :

=w

., =.

I

.i E

E. : M.."l

.. u i w

w=

I w

id W. -. 8. :.

": 8.

z 5

R.

-s W

u.

u.

l r

m.

s E

4:

4

.a 3

w i

m e

~;

l d

lW w

5. :. 5.. 5..

5 w

=

......-.S..

l r

u w

l i :

! 9 l:

=

l N

B; I

- !. - l. - !. - f i

b I.

i

=D g.

=D z

m N

M

~

.a M

aO A

gwD a<>

4 A&L HM C-10

~

w TABLE C.2-5 VALUES FOR NORTH ANNA UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIALS 9 LICENSE EXPIRATIO RTPTS

_.................T..S

TYPEIFLUENete P

I P I I t vaLUE LOC.IPLANr t CU t NI 1

vsa :.os.ssa.oss:

7s. e actuat a s.m.

45E+20 '210.8

_=

2 vos :.ts:.ess.oisi ss. i actuat i s.m.

4SE+20 237.8

.=__

3 : vos :. ors.ost.ois

-4s. i actuat c.w.

.45E+20 36.8 n.

H~

4 T

r 0

4 L

l I

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