ML20137L791

Reactor Vessel Fluence & Rt Pressurized Thermal Shock (PTS) Evaluations

ML20137L791
Person / Time
Site:	Indian Point
Issue date:	01/31/1986
From:	Perone V, Rens T, Turner R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML093440566	List: ML20137L791
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR WCAP-11045, NUDOCS 8601280089
Download: ML20137L791 (51)

Text

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WCAP-11045 WESTINGHOUSE PROPRIETARY CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION INDIAN POINT UNIT 3 REACTOR VESSEL FLUENCE AND Ri p ;3 WAWAHOE V. A. Perone T. E. Rens R. L. Turner M. Weaver S. E. Yanichko Work Performed for New York Power Authority January 1906 APPROVED:

'/ M APPROVED: I w T. A. Meyer, Manager F. L. Lau, Manager Structural Materials Radiation and Systems and Reliability Technology Analysis APPROVED: /W C. W. Hirst, Manager Reactor Coclant System Components 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.

WESilHGHOUSE ELECTRIC CORPORATION NUCLEAR ENERGY SYSTEMS P. O. 00X 355 PliiSBURGil, PENNSYLVANI A 15230 4074e:1d/010006 0601200009Otgo 06 pon ADOCK Of PDH P

TABLE OF CONTENTS PAGE TABLE OF CONTENTS i LIST OF TABLES 11 LIST OF FIGURES v

1. INTRODUCTION 1 1.1 The Pressurized Thermal Shock Rule 1 1.2 The Calculation of RTpys 3 II. NEUIRON EXPOSURE EVALUATION 5 11.1 Method of Analysis 5 II.2 Fast Neutron Fluence Results 8 III. MATERIAL PROPERTIES 22 111.1 Identification and Location of Beltline Region Materials 22 III.2 Definition and Source of Material Properties for All 22 Vessel Locations III.3 Summary of Plant-Specific Material Properties 23 IV. DETERMINATION OF RTpis VALUES FOR ALL BELTLINE 26 REGION MATERIALS IV.1 Status of Reactor Vessel Integrity in Terms of RipTS 26 versus Fluence Results IV.2 Discussion of Results 27 V. CONCLUSIONS AND RECOMMENDATIONS 30 VI. REFERENCES 32 VII. APPENDICES A. Power Distribution A-1 B. Weld Chemistry B-1 C. RTpis Values of Indian Point Unit 3 Reactor Vessel C-1 Beltline Region Materials 1

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LIST OF 1 ABLES Page j 11.2-1 Indian Point Unit 3 Fast Neutron (E>1.0 MeV) Exposure at the 11 Reactor Vessel Inner Radius - 0* Azimuthal Angle II.2-2 Indian Point Unit 3 Fast Neutron (E>l.0 MeV) Exposure at the 12 Reactor Vessel Inner Radius - 15* Azimuthal Angle II.2-3 Indian Point Unit 3 Fast Neutron (E>l.0 MeV) Exposure at the 13 Reactor Vessel Inner Radius - 30* Azimuthal Angle 11.2-4 Indian Point Unit 3 Fast Neutron (E>l.0 MeV) Exposure at the 14 Reactor Vessel Inner Radius - 45' Azimuthal Angle 11.2-5 Indian Point Unit 3 Fast Neutron (E>1.0 MeV) Exposure at the 15 4* Surveillance Capsule Center 11.2-6 Indian Point Unit 3 Fast Neutron (E>l.0 MeV) Exposure at the- 16

, 40' Surveillance Capsule Center III.3-1 Indian Point Unit 3 Reactor Vessel Beltline Region Material 24 Properties IV.1-1 RTpisvalues for Indian Point Unit 3 28 A-1 Indian Point Unit 3 Core Power Distributions Used in A-3 the Fluence Analysis B .1 -1 Indian Point Unit 3 Intermediate to Lower Shell B -2 Circumferential Weld Chemistry From WOG Materials Data Base -

Wire Heat Number 13253 C .1 -1 RTpis Values for Indian Point Unit 3 Reactor Vessel Beltline C-2 Region Materials at Various Fluences C.1-2 RTPTS Values for Indian Point Unit 3 Reactor Vessel Beltline C -4 Region Materials 9 Current (4.3 EFPY) Fluence Values C.1-3 RTpis Values for Indian Point Unit 3 Reactor Vessel Beltline C-5 Region Materials @ License Expiration (20.0 EFPY) 4074e:ld/011086 11

LIST OF FIGURES PAGE II.1-1 Indian Point Unit 3 Reactor Geometry 17 11.2-1 Indian Point Unit 3 Maximum Fast Neutron (E>l.0 MeV) 18 Fluence at the Beltline Location as a Function of Full Power Operating Time II.2-2 Indian Point Unit 3 Maximum Fast Neutron (E>l.0 Met'; ! ! yfW. e 19 at the Reactor Vessel Inner Radius as a Function '

.of Azimuthal Angle 11.2-3 Indian Point Unit 3 Relative Radial Distribution 20' of Fast Neutron (E>l.0 MeV) Flux and Fluence Within the Reactor Vessel Wall II.2-4 Indian Point Unit 3 Relative Axial Distribution of 21 Fast Neutron (E>1.0 MeV) Flux and Fluence Within the Reactor Vessel Wall I I I .1 -1 Identification and Location of Beltline Region Material 25 for the Indian Point Unit 3 Reactor Vessel IV.1 -1 Indian Point Unit 3 - RTPTS Curves per PTS Rule Method [1] 29 Docketed Base Material and WOG Data Base Mean Weld Material Properties A-1 Indian Point Unit 3 Core Description for Power A-4 Distribution Map l

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SECTION I INTRODUCTION The purpose of this report is to submit the reference temperature for pressurized thermal shock (RT * "' " #" " " #'

PTS vessel to address the Pressurized Thennal Shock (PTS) Rule.Section I discusses the Rule and provides the methodology for calculating RT PTS'Section II 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 vessel beltline region material properties.Section IV provides the RT calculations from present through the projected end-of-license fluence PTS 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 PT3 events on pressurized water reactor (PWR) vessels in nuclear power plants that are operated with a license f rom 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 j 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 af fecting the integrity I

of the vessel.

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The Rule establishes the following requirements for all domestic, operating

, PWRs:

Establishes the RTPTS (measure of f racture resistance) Screening Criterion for the reactor vessel beltline region '

) 270"F for plates, forgings, axial welds i 300*F for circumferential weld materials 1

, 6 Months From Date of Rule: All plants must submit their present Ripy3 values (per the prescr! bed methodology) and projected Ripts

! values at the expiration date of the operating license. The date that this submittal must be received by the NRC for plants with' operating

, licenses is January 23, 1986.

I 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 c 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.

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, RT NDT 5 now defined as "the reference temperature for pressurized thermal shock" (Rip 73) and calculated as prescribed by 10 CFR 50.61(b) of the Rule. Each j USNRC licensed PWR must submit a projection of RT py3 values from the time of

the submittal to the license _ expiration date. This' assessment must be submitted within 6 months after the effective date of the Rule,~on January 23, 9

1986, with updates whenever changes in core loadings, surveillance measurements, or other information indicate a significant change in projected 1

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 <

RT PTS vahes for Indan Mnt Unh 3.

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I.2 THE CALCULATION OF RT PTS j In the PTS Rule, the NRC Staff has selected a conservative and uniform method for determining plant-specific values of RT at a given time.

PTS i

l The prescribed equations in the PTS rule for calculating RT PTS are actually one of several ways to calculate RT NDT. For the purpose of comparison with the Screening Criterion, the value of RT PTS for the reactor vessel must be

calculated for each weld and plate, or forging in the beltline region as given below. For each material, RT PTS is the lower of the results given by Equations 1 and 2.

Equation 1:

RT PTS = 1 + M + [-10 + 470(Cu) + 350(Cu)(Ni)] f

• Equation 2:

0 RT PTS = I + M + 283 f .194 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.

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.

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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 10I9n/cm2 (E greater than or 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, causes the RT PTS values to be upper bound.

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 RT PTS calculations for Indian Point Unit 3 are discussed in Section 111.2.

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SECTION II NEUTRON EXPOSURE EVALUATION This section presents the results of the application of Westinghouse derived ,

l adjoint importance functions to the calculation of the Indian Point Unit 3 reactor vessel fluence for New York Power Authority. 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 l in the reactor vessel.

11.1 METHOD OF ANALYSIS

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A plan view of the Indian Point Unit 3 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' i surveillance program. Four capsules are located synnetrically at 4* f rom the

> core cardinal axes while four more are positioned 40* from the core cardinal i.

axes. The relative locations of the two sets of capsules are shown in Figure

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i II.1-1.

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 f rom a design basis core power distribution against which cycle by cycle plant specific calculations can be compared. The i second set of calculations consisted of a series of adjoint analyses relating i

the response of interest (fast neutron flux (E > 1.0 MeV)) at several selected l

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.

The forward transport calculation was carried out in R,8 geometry using the j DOT- discrete ordinates code [2] 'and the SAILOR cross-section library [3]. The l SAILOR library is a 47 group. ENOF/B-IV based data set produced specifically for light water reactor applications. Anisotropic scattering is treated with-q aP 3expansi n f the cross-sections. An S angular quadrature was used.

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l The design basis core power distribution utilized in the. forward analysis was derived f rom statistical studies of long-term operation of Westinghouse 4-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 l

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 P cross-secdon 3

approximation f rom 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 reactor vessel. Again, these calculations were run in R,e geometry to provide power distribution importance functions for .

the exposure parameter of interest (fast neutron flux (E > 1.0 MeV)). Having the adjoint importance function; and appropriate core power distributions, the response of interest is calculated as:

R R,0 " R I sI IE(R,0,E) F (R,0,E) dE R dR de where:

R g ,,

=

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

, azimuthal angle e.

I (R,0,E) = Adjoint importance function at radius R and azimuthal angle e for neutron energy group E.

F (R,e,E) =

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

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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 Indian Point Unit 3 plant specific fluence evaluations were taken from the nuclear design reports for cycles 1 through 5 [4, 5, 6, 7, 8]. 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 the core loadings in cycles 1 through 4, are defined as of June 8, 1985. The operating license for Indian Point Unit 3 expires on August 13, 2009 (forty years after the construction permit was issued). This report includes fluence projections from June 8, 1985 to August 13, 2009 using the cycle-averaged core power distribution of the carrent operating cycle (Cycle 5) and an assumed future capacity f actor of 65%. All fluence projections into the future reflect the low leakage fuel management strategy exemplified by the Cycle 5 core loading.

1 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 [9). The benchmarking studies indicate th&t 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 i 15% of measured values at si:rveillance capsule locations.

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11.2 FAST NEUTRON FLUENCE RESULTS 4 Calculated fast neutron (E >1.0 MeV) exposure results for Indian Point Unit 3 are presented in Tables 11.2-1 through 11.2-6 and in Figures 11.2-1 through 11.2-4. Data is presented at several azimuthal locations on the inner radius of the reactor vessel as well as at the center of each surveillance capsule.

f In Tables 11.2-1 through 11.2-4 cycle-specific maximum neutron flux and i fluence levels at 0*,15*, 30*, and 45* on the reactor vessel inner radius of 2

Indian Point Unit 3 are -listed for the period of operation up to June 8,1985, and projected to the expiration date of the operating license. Also presented are the design basis fluence levels predicted using the generic 4-loop core power distribution' at the nominal + 2a level. Similar data for the center

, of surveillance capsules located at 4* and 40* are given in Tables 11.2-5 and 11.2-6, respectively.

In addition to the calculated data given for the surveillance capsule locations, measured fluence data from previously withdrawn surveillance

}

1 capsules are also presented for comparison with analytical results. Capsules l 1

were removed from the 40' location at the end of cycle 1 and at the end of j cycle 3.

Several observations regarding the data presented in Tables 11.2-1 through 11.2-6 are worthy of note. These observations may be summarized as follows:

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1. Calculated plant specific fast neutron (E > 1.0 MeV) fluence levels at the I surveillance capsule center are in good agreement with measured data. The

, average difference between the plant specific calculations and the i measurements is less than 125. Differences of this magnitude are within i the uncertainty of the experimental results.

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2. The peak fast neutron (E > 1.0 MeV) flux incident on the reactor vessel i (45' azimuthal position), averaged over the fuel cycles where fresh fuel l was loaded adjacent to the 45' position (cycles 1-3)', was 12 percent less than predictions based on the design basis core power distribution.

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3. During cycles 4 and 5, when low leakage fuel management was used on the core periphery at the 45* position, the average peak fast neutron (E > 1.0 Mev) flux on the reactor vessel was reduced by 28 percent relative to that existing prior to the implementation of low leakage.

Graphical presentations of the plant specific fast neutron fluence at key I locations on the reactor vessel are shown in Figure II.2-1. Reactor vessel data is presented for the 45* location on the circumferential weld and shell plates as well as for all of the longitudinal welds (see Section III.1).

In regard to Figure II.2-1, the solid portions of the fluence curves are based directly on the cycle-specific core loadings of the first four fuel cycles.

The dashed portions of these curves, however, involve a projection into the future. As mentioned in Section 11.1, the neutron flux average over cycle 5 was used to project future fluence levels.

It should be noted that implementation of a more severt low leakage pattern than that used in cycle 5 would reduce the projections of fluence at key locations. On the other hand, relaxation of the current low leakage pattern or a return to out-in fuel management would increase those projections. The RT PTS assessment must be updated per 10CFR50.61(b)(1) whenever, among other things, changes in core loadings significantly impact the fluence and RT PTS projections.

In Figure 11.2-2, the azimuthal variation of maximum fast neutron (E > 1.0 MeV) fluence at the inner radius of the reactor vessel is presented as a function of azimuthal angle. Data are presented for both current and projected expiration-of-operating-license conditions. In Figure 11.2-3, the relative radial variation of fast neutron flux and fluence within the reactor vessel wall is presented. Similar data showing the relative axial variation of fast neutron flux and fluence over the beltline region of the reactor vessel is shown in Figure 11.2-4. A three-dimensional description of the fast neutron exposure of the reactor vessel wall can be constructed using the data given in Figure II.2-2 through II.2-4 along with the relation

+(R, 0,Z) = $(e) F(R) G(Z) 4074e:1d/0ll686 9

where: + (R,0,Z) = Fast neutron fluence at location ~R, e, Z within the reactor vessel wall

+ (e) -

Fast neutron fluence at azimuthal location e on the reactor vessel inner radius f rom Figure 11.2-2

, F (R) = Relative f ast neutron flux at depth R into the reactor vessel f rom Figure 11.2-3 G (Z) - Relative fast neutron flux at axial position Z from Figure 11.2-4 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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TABLE 11.2-1 INDIAN POINT UNIT 3 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE REACTOR VESSEL INNER RADIUS - 0* AIIMUTHAL ANGLEI *},

l Beltline Region Elapsed. Cumulative Fluence (n/cm2)

Irradiation Irradiation Avg.pFlux Plant Desig&b}

Interval Time (EFPY) (n/cm -sec) SDecific Basis CY-1 1.2 6.34 X 10 9 2.43 x 10" 3.10 x 10 U CY-2 2.1 7.90 x 10 9 4.67 x 10 U 5.39 x 10 U CY-3 3.2 8.39 x 10 9 7.47 x 10 lI '8.08 x 10 U CY-4 IC} 4.3 7.70 x 10 9 1.02 x 10 18 1.09 x 10 18 18 18 CY-5(d) 5.5 7.20 x 10 9 1.29 x 10 1.39 x 10 18 CY-6+8/13/2009I ') 20.0 7.20 x 10 9 4.59 x 10 5.10 x 10 18 (a) Applicable to longitudinal weld 2-042C in the intermediate shell.

(b) Design basis fast neutron flux = 8.08 x 109 n/cm2-sec at 3025 MWth (c) Current neutron fluences are defined as of the end of CY-4 (6/8/85).

(d) Fuel cycle projection. Beyond the end of CY-4 a 65% capacity factor is assumed.

(e) The average fast neutron flux derived from CY-5 was used to make fluence projections to the license expiration date (8/13/2009).

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TABLE 11.2-2 INDIAN POINT UNIT 3 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE REACTOR VESSEL INNER RADIUS - 15' AZIMUTHAL ANGLE (*}

Beltline Region Elapsed Cumulative Fluence (n/cm2 )

Irradiation Irradiation Avg.pFlux Plant Interval ~ Time (EFPY) (n/cm -sec) Desig[b}

SDecific Basis 10 CY-1. 1.2 1.00 X' 10 3.84 x 10 I 4.92 x 10 I CY-2 2.1 1.25 x 10 10 7.38 x 10 I 8.54 x 10 II CY-3 3.2 1.32 x 10 10 1.18 x 10 18 1.28 x 10 18 CY-4 IC) 4.3 1.11 x 10 10 1.57 x 10 18 1.73 x 10 18 CY-5(d) 5.5 1.02 x 10 10 1.95 x 10 18 2.21 x 10 18 CY-648/13/2009I ') 20.0 1.02 x 10 10 6.63 x 10 18 8.08 x 10 18 (a) Applicable to longitudinal welds 3-042A and 3-042B in the lower shell.

(b) Design basis fast ~ neutron flux.= 1.28 x 1010 n/cm2-sec at 3025 MWth (c) Current neutron fluences are defined as of the end of CY-4 (6/8/85).

i (d) Fuel cycle projection. Beyond the end of CY-4 a 65% capacity factor is assumed.

(e) The average fast neutron flux derived from CY-5 was used to make fluence projections to the license expiration date (8/13/2009).

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-- - . . ~ . . . , _ - - - .

TABLE II.2-3 INDIAN POINT UNIT 3 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE REACTOR VESSEL INNER RADIUS - 30' AZIMUTHAL ANGLE (*

l Beltline Region Elapsed Cumulative Fluence (n/cm2 )

Irradiation Irradiation Avg.2 Flux Plant Desig Interval Time (EFPY) (n/cm -sec) Specific 8 asis [b) 10 CY-1 1.2 1.26 X 10 4.85 x 10 lI 6.14 x 10 II CY-2 2.1 1.60 x 10 10 9.40 x 10 II 1.09 x 10 18 CY-3 3.2 1.55 x 10 10 1.45 x 10 18- 1.60 x 10 I8 CY-4(c) 4.3 1.21 x 10 10 1.88 x 10 18 2.17 x 10 18 CY-5(d) 5.5 1.08 x 10 10 2.29 x 10 18 2.76 x 10 18 CY-648/13/2009I ') 20.0 1.08 x 10 10 7.26 x 10 18 1.01 x 10 I9 t

(a) Applicable to longitudinal welds 2-042A and 2-0428 in the intermediate shell.

(b) Design basis fast neutron flux = 1.60 x 1010 n/cm2-sec at 3025 MWth 1

(c) Current neutron fluences are defined as of the end of CY-4 (6/8/85).

(d) Fuel cycle projection. Beyond the end of CY-4 a 65% capacity factor is assumed.

(e) The average fast neutron flux derived from CY-5 was used to make fluence projectionstothelicenseexpirationdate(8/13f2009).

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TABLE 11.2-4 INDIAN POINT UNIT 3 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE REACTOR VESSEL INNER RADIUS - 45' AZIMUTHAL ANGLE I ")

Beltline Region Elapsed Cumulative Fluence (n/cm2 )

Irradiation Irradiation Avg.2 Flux . Plant Interval Desig?b}

Time (EFPY) (n/cm -sec) Specific Basis 1.86 X 10 10 7.13 x 10I I I

CY-1 1.2 9.41 x 10 II CY-2 2.1 2.56 x 10 10 1.44 x 10 18 1.64 x 10 18 CY-3 3.2 2.17 x 10 10 2.16 x 10 18 2.45 x 10 18 1.61 x 10 10 2.73 x 10 18 3.32 x 10 18 l

CY-4(c) 4.3 CY-5(d) 5.5 1.50 x 10 10 3.29 x 10 18 4.23 x 10 18 CY-64t;/13/2009I 'I 20.0 1.50 x 10 10 1.02 x 10 I9 1.55 x 10 I9 (a) Applicable to longitudinal weld 3-042C in the lower shell, the intermediate {

to lower shell circumferential weld 9-042, and all shell plates.

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(b) Design basis fast neutron flux = 2.45 x 1010 n/cm2-sec at 3025 MWth (c) Current neutron fluences are defined as of the end of CY-4 (6/8/85).

(d) Fuel cycle projection. Beyond the end of CY-4 a 65% capacity factor is-assumed.

(e) The average fast neutron flux derived from CY-5 was used to make fluence projections to the licer expiration date (8/13/2009).

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TA8LE 11.2-5 INDIAN POINT UNIT 3 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 4* 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 lI I CY-1 1.2 2.01 x 10 7.72 x 10 9.91 x 10 10 18 18 i CY-2 2.1 2.51 x 10 1.48 x 10 1.72 x 10 10 18 18 CY-3 3.2 2.67 x 10 2.37 x 10 2.58 x 10 CY-4(b) 4.3 2.44 x 10 10 3.24 x 10 18 3.50 x 10 18 CY-5(c) 5.5 2.27 x 10 10 4.08 x 10 18 4.46 x 10 18 CY-6+8/13/2009(d) 20.0 2.27 x 10 10 1.45 x 10 I9 1.63 x 10 I9 1

(a) Design basis fast neutron flux = 2.58 x 1010 n/cm2-sec at 3025 MWth (b) Current neutron fluences are defined as of the end of CY-4 (6/8/85).

(c) Fuel cycle projection. Beyond the end of CY-4 a 65% capacity factor is assumed.

! (d) The average fast neutron flux derived from CY-5 was used to make fluence projections to the license expiration date (8/13/2009).

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TABLE II.2-6 INDIAN POINT UNIT 3 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 40* SURVEILLANCE CAPSULE CENTER Beltline Region Elapsed Cumulative Fluence (n/cm2 )

4 Irradiation Irradiation Avg.2 Flux Plant Desigg,) Capsule Interval Time (EFPY) (n/cm -sec) Specific Basis Data CY-1 1.2 6.04 X 10 10 2.32 x.10 18 3.07 x 10 18 2.79 x 10 18(e)

CY-2 2.1 8.31 x 10 10 4.68 x 10 18 5.34 x 10 18 CY-3 3.2 7.11 x 10 10 7.04 x 10 18 8.01 x 10 18 7.51 x 10 18(f)

CY-4(b) 4.3 5.28 x 10 10 8.91 x 10 18 1.08 x 10 I9 ,

4.89 X 10 10 1.07 x 10 I9 1.38 x 10 I9 CY-5(c) 5.5 20.0 4.89 x 10 10 I9 5.05 x 10 I9 CY-648/13/2009(d) 3.32 x 10 (a) Design basis fast neutron flux = 8.00 x 1010 n/cm2-sec at 3025 MWth (b) Current neutron fluences are defined as of the end of CY-4 (6/8/85).

(c) Fuel cycle projection. Beyond the end of CY-4 a 65% capacity factor is assumed.

(d) The average' fast neutron flux derived from CY-5 was used to make fluence projections to the license expiration date (8/13/2009).

, (e) Peflects adjustments made to the spectrum-averaged reaction cross sections reported in Reference 10.

(f) Reflects adjustments made to the spectrum-averaged reaction cross sections reported in Reference 11.

4074e:1d/011686

16267 1 O' 4*(CAPSULES U,V.W,X) l REACTOR VESSEL 1

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Figure 11.1-1. Indian Point Unit 3 Reactor Geometry 17

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Figure 11.2-1. Indian Point Unit 3 Maximum Fast Neutron (E>1 MeV) Fluence at the Beltline Location as a Function of Full Power Operating Time 18 I

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l I l  !  ! I 1017 o lo 20 30 40 so 60 70 AZIMUTHAL ANGLE (DEGREE)

Figure 11.2-2. Indian Point Unit 3 Maximum Fast Neutron (E>1.0 MeV) Fluence at the Reactor VesselInner Radius as a Function of Azimuthal Angle 19

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Figure 11.2-3. Indian Point Unit 3 Relative Radial Distribution of Fast Neutron (E>1.0 MeV) Flux and Fluence Within the Reactor Vessel 20

16267-5 lo.o _

8 -

6 -

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Figure l1.2-4. Indian Point Unit 3 Relative Axial Variation of Fast Neutron (E>1.0 MeV) Flux and Fluence Within the Reactor Vessel Wall 21

SECTION III MATERIAL PROPERTIES For the RT PTS calculation, the best estimate copper and nickel chemical composition of the reactor vessel beltline material is necessary. The material properties for the Indian Point Unit 3 beltline region will be presented in this section.

111.1 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."

Figure III.1-1 identifies the location of all beltline region materials for the reactor vessel.

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 12, were derived from vessel fabrication test certificates. The property data for the welds were derived from weld qualification test records and have also been reported in Reference 12. The tests were performed by the reactor vessel vendor, Combustion Engineering (CE), at the time of fabrication.

Fast neutron irradiation-induced changes in the tensile, f racture, 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 concentration within the weldments.

4074e:ld/010886 22

r For each weld in the Indian Point Unit 3 beltline region, a material data search was performed using the WOG Reactor Vessel Beltline Region Weld Materials Data Base. The WOG data base, which was developed in 1984 and is continually being updated, contains information from weld qualification records (including those for Indian Point Unit 3), surveillance capsule reports, the B&W report BAW-1799, and the Materials Properties Council (MFC) and the NRC Mender MAISURV data bases.

Searches were performed for materials having the identical weld wire heat number as those in the Indian Point Unit 3 vessel, 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 f rom the data base searches is found in Appendix B.

III.3

SUMMARY

OF PLANT-SPECIFIC MATERIAL PROPER 1IES A summary of the pertinent chemical.and mechanical properties of the beltline region plate and weld materials of the Indian Point Unit 3 reactor vessel are given in Table 111.3-1. 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 currently used in the Regulatory Guide 1.99, Revision 1 trend curve [13].

All of the initial RT NDT values (I) are given in Table 111.3-1. The longitudinal weld initial RI s e generic mean value as defined by the NDT Pls rule [1].

The data in Table 111.3-1 is used to evaluate the RT p73 values for the Indian Point Unit 3 reactor vessel.

4074e:ld/011086 23

TABLE III .3-1 INDIAN POINT UNIT 3 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES Cu Ni P 1(a)

(Wt.%) (Wt.%) (Wt.%) CF1 Intermediate Shell Plate B2802-1: .20 .50 .010 5 Intermediate Shell Plate B2802-2: .22 .53 .015 -4 Inttrmediate Shell Plate B2802-3: .20 .49 .011 17 Lower Shell Plate B2803-1: .19 .47 .012 49 Lower Shell Plate B2803-2: .22 .52 .011 -5 Lower Shell Plate B2803-3: .24 .52 .012 74 Longitudinal Welds 042 A,B,C and 3-042 A,B,C, Wire Heat 348009, Flux Linde 1092 WOG Data Base .19 1.00 .012 -56 Circum. Weld 9-042 - Intermed. to Lower Shell, Wire Heat 13253, Flux Linde 1092 Weld Qualification Value .27 .74 .023 -70 WOG Data Base Mean .25 .72 .02 ---

(a) All Plate and Circumferential Weld RTNDT (I) values are actual values.

The remaining value for the Longitudinal welds is a generic mean value as defined by the PTS rule [1].

4074e:1d/010886 24

16267-6 CIRCUMFERENTIAL SEAMS VERTICAL SEAMS 2-0428 90*

~

B2802-2 g C 8-042 21" f =

30 L 2-042C H J CORE J

O'- -- -

A 180*

CORE Y I m '

30' N

< 82802-1 B2802-3 144" 8

$ 2-042A Q_ .-_- __ _ z 270*

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u

/ g CORE 3-042A O*

~< ,.

180*

I B2803-1 3-042C 270*

Figure 111.1-1. Identification and Location of Beltline Region Material for the indian Point Unit 3 Reactor Vessel 25

SECTION IV DETERMINATION OF RT VALUES FOR ALL BELTLINE REGION MATERIALS PTS Using the methodology prescribed in Section 1.2, the results of the fast neutron exposure provided in Section II, and the material properties discussed

,in Section III, the RT PTS values for Ind an Point Un h 3 can now be

'~

determined. ,

IV.1 STATUS OF REACTOR VESSEL INTEGRITY IN . TERMS OF RT PTS SUS M ENM RESULTS i

Using the prescribed PTS Rule methodology, RT values were generated for PTS all beltline region materials of the Indian Point Unit 3 reactor vessel 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.

Figure IV.1-1 presents the RT values for the limiting _ longitudinal weld, PTS circumferential weld and limiting basemetal of the Indian Point Unit 3 vessel in terms of RT PTS 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 PTS Screening Criterion for PTS (i.e., RT values can be PTS readily projected for any options under consideration, provided fluence is known), and o to show the current (4.29 EFPY), and end-of-license (20.0 EFPY) RI PTS values using actual and projected fluener.

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

4074e:1d/011686 26

_ . _ . - . - . ,- . ~-

i i

t

lable IV.1-1 provides a sununary of the RT pj3 values for all beltline region-

j. materials for the lifetime of. interest.

1 IV.2 DISCUSSION OF RESULTS j

i As shown in Figure IV.1-1 and Table IV.1-1, . lower shell plate B2803-3 is the limiting location relative to PTS. At license expiration, plate 82803-3 is

seen to have an RT PTS value of 269'F. All of the RT values, including PTS that for plate B2803-3,. are below the NRC screening criteria through license

expiration.

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! 4074e:ld/0ll686 27 i

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1 TABLE IV.1-1 RT VALUES FOR INDIAN POINT UNIl 3 PTS i

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RT PTS Values (*F)

Present End-of-License Screening Location Vessel Material (4.29 EFPY) (20.0 EFPY) Criteria 1 Intermediate she'll plate B2802-1 137 173 270 3

2 Intermediate shell plate B2802-2 139 179 270 3 Intermediate shell plate B2802-3 148 184 270 4 Lower shell plate B2803-1 175 208 270 5 Lcwer shell plate B2803-2 137 177 270 6 Lower shell plate B2803-3 225 269 270 7 Limiting Longitudinal Weld 3-042C 104 147 270 8 Intermediate to lower shell 114 166 300 circumferential weld 9-042 Longitudinal weld 2-042C 80 119 270 Longitudinal weld 3-042A,8 90 131 270 l

Longitudinal weld 2-042A,8 94 134 270 i

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! 4074e:1d/011686 28 l

16267-7 400 I l

l LIMITING BASEMETAL 350 -

NRC RTPTS SCREENING VALUE (3OO*F) -

CIRCUMFERENTIAL WELD 300 ________________________________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

NRC RTPTS SCREENING VALUE (270*F) -

PLf]E AND pONGp,W{ nap _WE{DS, 250 - (269)

- (225) b 1 m w

200 1

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'1010 019 iO20 NEUTRON FLUENCE (n/cm2)

! LEGEPOt & = CURRENT LIFE. (4.29 EFPY) Ato e = END OF LICENSE (20.O EFPY)

RTPTS VALUES USING PLANT SPECIFIC Af0 PROJECTED PLANT SPECIFIC 4

FLUENCE VALUES (LIMITING FLUENCES USED FOR ALL LOCATIONS)

Figure IV.1-1. Indian Point Unit 3 RTPTS Curves Per PTS Rule Method [1] Docketed Base Material and WOG Data Base Mean Material Properties l

1 s

I 29 1

. - - - - - ~ . , - --

SECTION V CONCLUSIONS AND RECOMMENDATIONS Calculations have been completed in order to submit RT values for the PTS l Indian Point Unit 3 reactor vessel in meeting the requirements of the NRC Rule l

l for Pressurized Thermal Shock [1]. This work entailed a neutron exposure evaluation and a reactor vessel material study in order to determine the i RT PTS values.

j 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 Indian Point Unit' 3 reactor vessel. Explicit calculations were performed for the first five fuel cycles. Projection of the fast neutron exposure beyond June 8, 1985 was based on continued implementation of low leakage fuel management similar to that employed during cycle 5.

i Plant specific evaluations have demonstrated that during fuel cycles using '

out-in-in fuel management, the maximum fast neutron (E > 1.0 MeV) flux incident on the reactor vessel was, on the average,12 percent less than predictions based on the design basis core power distributions. With regard to the low leakage fuel management strategy in place at Indian Point Unit 3, the plant specific evaluations have shown that the average fast neutron (E > 1.0 MeV) flux at the 45* azimuthal position (peak location) was reduced

by 28 percent relative to that prior to the implementation of low leakage.

l i It should be noted that significant deviations from the low leakage scheme t

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 relative power on the periphery) would reduce the projection. On'the other hand, a relaxation of the loading pattern toward higher relative power on the core periphery would increase the projections beyond those reported. As each future fuel cycle evolves, the loading patterns should be evaluated to j determine their potential impact on projections made in this report.

l 4074e:1d/011686 30

J j lhe fast neutron-fluence values from the plant specific calculations have been i compared directly with measured fluence levels derived from neutron dosimetry i contained in surveillance capsules withdrawn from Indian Point Unit 3. The 4

ratio of calculated to measured fluence values ranges from 0.83 to 0.94 for the two capsule data points. This reasonably good agreement between calculation and measurement supports the use of this analytical approach to

! perform a plant specific evaluation for the Indian Point Unit 3 reactor.

l Material property values for the Indian Point Unit 3 reactor vessel beltline

region components were determined. The pertinent chemical and mechanical

! properties for the shell plates remain the same as those that were originally i reported-in the vessel fabrication test certificates. The weld material j properties are obtained from the WOG Material Data Base.

4 Using the prescribed PTS Rule methodology, RT values were g,enerated for PTS all beltline region materials of the Indian Point Unit 3 reactor vessel:as a j function of several fluence values and pertinent vessel lifetimes. All of the

{

RT PTS values remain below the NRC screening values for PTS using the projected fluence exposure through the expiration date of'the operating ,

j license. The most limiting value at end-of-license is 269'F for the lower

shell plate B2803-3. A review of the material properties of shell plate

! B2803-3 indicates that a high initial value of RT is the primary factor NDT causing the high RTPTS **1"

i The results in this report are provided to enable New York Power Authority to j comply with the initial 6 months submittal requirements of the USNRC PTS rule.  ;

}

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l 4074e:ld/0ll686 31

SECTION VI REFERENCES

1. Nuclear Regulatory Commission,10CFR Part 50, " Analysis of Potential Pressurized Therwel 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 Data Library Collection DLC-76." Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P , Cross-Section' Library for Light Water 3

Reactors.

4. WCAP-8360, " Core Physics Characteristics of the Indian Point Nuclear Plant Unit III, Cycle 1," P. J. Sipush, et al. 0ctober 1974.

5. WCAP-9244, "The Nuclear Design and Core Management of the Indian Point Plant Unit No. 3, Cycle 2," D. M. Lucof f, et al., January 1978.

6. WCAP-9599, "The Nuclear Design and Core Management of the Indian Point Plant Unit No. 3, Cycle 3," J. A. Penkrot, et al., September 1979.

7. WCAP-10051, "The Nuclear Design and Core Management of the Indian Point Plant Unit No. 3, Cycle 4," M. A. Kotun and M. F. Muenks, March 1982.

8. WCAP-10839, "The Nuclear Design and Core Management of the Indian Point Plant Unit No. 3, Cycle 5," M. A. Kotun and R. H. Pitulski, June 1985.

9. Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology -

4 to be published.

4074e:ld/010886 32 l

10._WCAP-9491, " Analysis of Capsule T from the Indian Point Unit No. 3 Reactor Vessel Radiation Surveillance Program," J. A. Davidson, S. L. Anderson and W. T. Kaiser, April 1979.

1

11. WCAP-10300, " Analysis of Capsule Y f rom the Power Authority of the State of New York, Indian Point Unit 3 Reactor Vessel Radiation Surveillance Program," S. E. Yanichko, S. L. Anderson and W. T. Kaiser, March 1983.

12. Letter f rom W. J. Cahill, Jr., of Consolidated Edison to the Director of i Nuclear Reactor Regulation, Mr. R. W. Reid, Docket 50-286, March 8,1978.

i

' 13. " 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.

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4 q 4074e:ld/011086 33 i

. . . - . _- . _ . _~ .-. . .

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l APPENDIX A CORE POWER DISTRIBUTIONS .

Core power distributions used in the plant specific fast neutron exposure analysis of the Indian Point Unit 3 reactor vessel were derived from the i following fuel cycle nuclear design reports: ,

) Fuel Cycle Nuclear Desian Report 1

i i 1 WCAP-8360 2 WCAP-9244 3 WCAP-9599 4 WCAP-10051 5 WCAP-10839 di I A schematic diagram of the core configuration applicable to Indian Point l Unit 3 is shown in Figure A-1. Cycle averaged relative assembly powers for each fuel cycle are listed in Table A-1 along with the design basis core power ,

distribution.

j

! On Figure A-1 and in Table A-1 an identification number is assigned to each fuel assembly location. Three regions consisting of subsets of fuel l assemblies are defined. In performing the adjoint evaluatio'ns, the relative power in the fuel assemblies comprising Region 3 has-been adjusted to account for known biases in the prediction of power in the peripheral fuel assemblies f while the relative power in the fuel assemblies comprising Region 2 has been l maintained at the cycle average value. Due to the extreme self-shielding of  :

the reactor core neutrons born in the fuel assemblies comprising Region 1 do

< not contribute significantly to the neutron exposure of either the surveillance capsules or the reactor vessel. Therefore, core power l distribution data for fuel assemblies in Region 1 are not listed in Table A-1. '

l l In each of the adjoint evaluations, within assembly spatial gradients have j been superimposed on the average assembly power levels. For the peripheral i

f 4074e:ld/010886 A-1 i

J l assemblies (Region 3), these spatial gradients also include adjustments to

} account for analytical deficiencies that tend to occur near the boundaries of j the core region.

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l 4074e:ld/010886 A-2

TABLE A-1 INDIAN POINT UNIT'3 CORE POWER DISTRIBUTIONS USED IN THE FLUENCE ANALYSIS ,

l Design . Plant Specific Cycle Averaged Relative Assembly Power Basis Fuel Cycle i Fuel Relative Assembiv Power 1 2 3 4 5 1 1.06 0.76 1.00 1.07 0.98 0.81 2 1.09 0.82 1.01 1.06 0.96 0.94 3 1.01 0.71 0.94 0.99 0.92 0.83 4 0.81 0.61 0.76 0.79 0.49 0.43 i 5 1.15 0.87 1.05 1.02 0.93 0.74 6 0.75 0.52 0.76 0.64 0.40 0.37

! 7 1.02 1.00 0.96 1.11 1.10 1.09 j 8 1.10 1.00 1.24 1.19 1.14 1.22 9 1.00 0.98 0.94 0.98 1.16 1.04 10 1.05 0.97 1.16 1.22 1 ~.10 1.06 11 1.07 1.07 0.98 0.96 1.12 1.08 12 1.00 0.94 1.04 1.00 1.08 1.14

! 13 1.05 0.97 1.14 0.74 0.79 0.82 1

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t j Note: . Refer to Figure A-1 for fuel assembly location.

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4074e:1d/010886 A-3

16267 8 i

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7 8 9 10 5 6 45*

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14 15 16 11 12 13

17 18 19 20 21 I

22 23 24 25 i

i 26 27 28 1 l 29 30 REGION ASSEMBLIES i

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31 2 7-13  !

3 l-6 I Figure A-1. Indian Point Unit 3 Core Description for Power Distribution Map i

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APPENDIX B WELD CHEMISTRY l

Table B.1-1 provides the weld data output f rom the WOG Material Data Base.

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 i are used in the RT analysis.

PTS i

Weld Chemistry Data Source and Plant:

AEP -

Donald C. Cook Unit 1 Cu -

Weight % of Copper INT -

Indian Point Unit 3 KEP -

Mihama Unit 1 Ni -

Weight % of Nickel P -

Weight % of Phosphorous PGE -

Diablo Canyon Unit 1 PNJ -

Salem Unit 2

) PSE -

Salem Unit 1 SC -

Surveillance Capsule Si -

Weight % of Silicon WQ -

Weld Qualification i

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i __ _ _ _ . _ . _ . . , _ . _ . _ _ - _

_ ~ . . -

9 TABLE B-1 INDIAN POINT UNIT 3 INTERMEDIATE TO LOWER SifELL WELD CHEMIG TRY FROM WOG MA TERI ALS DATA BASE.

WIRE HEAT NUMBER 13253 SELECT AEPORT l

13 u!K ulRE FLO: FLUI WELCCHEM Eu Ni P Si PLANI KSCRIPil0N HEAT TVPE TYFE LOT LATA SOURCE 0290 13253 5-4 MOS LINDE 1092 3774 PNJ,5C 0.230 0.710 0.017 0.290 F6E N0lILE 10 INTER SHELL co 0:94 13253 9-4 MDI LINDE 10 2 3791 AEP,5C 0.270 0.740

• 0.023 0.190 AEP SURVEILLIEE WELD l

h GI94 13253 INT INTER TO LOWER SHELL 0294 13253 FEP N0ZILE TO INTER SNELL 0294 13253 PSE INTER TO LOWER SHELL sen 0.250000 0.725000 0.020000 0.235000 stt.dev. 0.020204 0.021213 0.004243 0.077702

.3.4...... .... ...... .................s.......................................................................................

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APPENDIX C RT VALUES OF INDIAN POINT UNIT 3 PTS REACTOR VESSEL BELTLINE REGION MATERIALS Tables C.1-1 through C.1-3 provide the RT PTS values, as a function of both constant fluence and constant EFPY (assuming the projected fluences values),

for all beltline region materials of the Indian Point Unit 3 reactor vessel.

The RT PTS values are calculated in accordance with the PTS rule, which is l 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 l

1 Intermediate shell plate B2802-1 2 Intermediate shell plate B2802-2 3 Intermediate shell plate B2802-3 i 4 Lower shell plate B2803-1

. 5 Lower shell plate B2803-2 6 Lower shell plate B2803-3 7 Limiting Longitudinal Weld 3-042C 8 Intermediate to lower shell Circumferential j Weld 9-042 i

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! 4074e:ld/010886 C-1

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

TABLE C-1-1 RTPTS VALUES FOR THE INDIAN POINT UNIT 3 REACTOR VESSEL BELTLINE REGION MATERIALS AT VARIOUS FLUENCES

!!D ! PLANT! CU ! NI  ! P !  !  ! VALUE  ! TYPE ! RTPTS VALUE AT FLUENCE

.10E+19 .50E+19 .10E+20 .20E+20 t INT 0.200 0.500 0.010 5 ACTUAL B.M. 117 152 172 196 2 INT 0.220 0.530 0.015 -4 ACTUAL B.M. 116 155 178 206 3 INT 0.200 0.490 0.011 17 ACTUAL B.M. 129 163 183 208 4 INT 0.190 0.470 0.012 49 ACTUAL B.M. 156 189 208 230 l 5 INT 0.220 0.520 0.011 -5 ACTUAL B.M. 115 154 176 204 i r3 6 INT 0.240 0.520 0.012 74 ACTUAL B.M. 201 243 268 299 g, 7 INT 0.187 1.000 0.012 -56 GENERIC L.W. 80 122 146 176 8 INT 0.250 0.725 0.020 -54 ACTUAL C.W. 86 136 165 200 1

1 Notes: ID = Location of vessel saterial (see page C-1)

! = Initial value of RTNDT, actual or GENERIC Value=" ACTUAL" or '6ENERIC* denotes type of initial RTNDT value 8.M. = Base Metal L.W. = Longitudinal Weld C.W. =Circumferential Weld Reference temperatures are in deg F

TABLE C.1-1 (CONT)

RTPTS VALUES FOR THE INDIAN POINT UNIT 3 REACTOR VESSEL BELTLINE REGION MATERIALS AT VARIOUS FLUENCES

! VALUE ' TYPE ! RTPTS VALUE AT FLUENCE

!ID ! PLANT! CU '

NI  ! P  !  !

.70E+20 40E+20 .60E+20 ACTUAL B.M. 226 246 254 1 INT 0.200 0.500 0.010 5 ACTUAL B.M. 239 262 271 2 INT 0.220 0.530 0.015 -4 B.M. 237 257 265 ACTUAL 3 INT 0.200 0.490 0.011 17 B.M. 258 276 284 49 ACTUAL 4 INT 0.190 0.470 0.012 B.M. 237 259 269 5 INT 0.220 0.520 0.011 -5 ACTUAL ACTUAL B.M. 335 360 370 6 INT 0.240 0.520 0.012 74 L.W. 211 236 245 S' 7 INT 0.187 1.000 0.012 -56 GENERIC 243 271 283

-54 ACTUAL C.W.

B INT 0.250 0.725 0.020

t TABLE C.1-2 RTPTS VALUES FOR THE INDIAN POINT UNIT 3 REACTOR VESSEL l BELTLINE REGION MATERIALS AT CURRENT LIFE 14.29 EFPV)

FLUENCE VALUES l

s

!!D ! PLANT! CU  ! NI  ! P  !  !  ! VALUE  ! TYPE ! FLUENCE ! RTPTS !

1 INT 0.200 0.500 0.010 5 ACTUAL B.M. 0.27E+19 137

, 2 INT 0.220 0.530 0.015 -4 ACTUAL B.M. 0.27E+19 139

) 3 INT 0.200 0.490 0.011 17 ACTUAL B.M. 0.27E+19 148 4 INT 0.190 0.470 0.012 49 ACTUAL B.M. 0.27E+19 175 '

l 5 INT 0.220 0.520 0.011 -5 ACTUAL B.M. 0.27E+19 137 f

& INT 0.240 0.520 0.012 74 ACTUAL B.M. 0.27E+19 225 c, 7 INT 0.187 1.000 0.012 -56 GENERIC L.N. 0.27E+19 104 j, 8 INT 0.250 0.725 0.020 -54 ACTUAL C.N. 0.27E+19 114 1

1 l 4

1 i

i

TABLE C.1-3 RTPTS VALUES FOR THE INDIAN POINT UNIT 3 REACTOR VESSEL BELTLINE REGION MATERIALS e END-OF-LICENSE (20.0 EFPY) FLUENCE VALUES

!ID ! PLANT! CU  ! NI  ! P  !  !  ! VALUE  ! TYPE ! FLUENCE ! RTPTS !

1 INT O.200, 0.500 0.010 5 ACTUAL B.M. O.10E+2O 173 2 INT O.220 0.530 0.015 -4 ACTUAL B.M. O.10E+20 179 3 INT O.200 0.490 0.011 17 ACTUAL B.M. O.10E+20 184 4 INT O.190 0.470 0.012 49 ACTUAL B.M. O.10E+20 208 5 INT O.220 0.520 0.011 -5 ACTUAL B.M. O.10E+20 177 7

m 6 INT O.240 0.520 0.012 74 ACTUAL B.M. O.10E+20 269 7 INT O.187 1.000 ,0.012 -56 GENERIC L.W. O.10E+20 147 8 INT O.250 0.725 0.020 -54 ACTUAL C.W. O.10E+20 166

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