NL-04-042, Reactor Vessel Material Surveillance Program: Preliminary Analysis Results for Capsule X

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Reactor Vessel Material Surveillance Program: Preliminary Analysis Results for Capsule X
ML041170399
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 04/19/2004
From: Dacimo F
Entergy Nuclear Northeast
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-04-042 IPEC-RPT-04-00005, Rev 0
Download: ML041170399 (39)
v • d • e


Text

Entergy Nuclear Northeast

'- Entergy Indian Point Energy Center 295 Broadway, Suite 1 RO. Box 249 Buchanan, NY 10511-0249 Fred Dacimo Ste Vice President Tel 914 734 6700 April 19, 2004 NL-04-042 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Indian Point Nuclear Generating Unit No. 3 Docket 50-286 Reactor Vessel Material Surveillance Program:

Preliminary Analysis Results for Capsule X

Dear Sir:

Pursuant to Appendix H to 10CFR50, Attachment I to this letter provides a summary technical report, "Summary Report IPEC-RPT-04-0005, Rev. 0, Preliminary Analysis of Capsule X -

Indian Point Unit 3 Reactor Vessel Surveillance Program."

The information in this report, which is pending final independent verification, validates the Indian Point 3 Technical Specifications pertaining to Reactor Coolant System heatup, cooldown and setpoints for the Low Temperature Overpressure Protection System. The report includes Charpy V-Notch testing analysis, Upper Shelf Energy (USE) analysis and credibility evaluation.

Tensile testing and dosimetry measurements are complete, and the results are being processed. It should be noted that neither tensile data nor dosimetry data affect the validation of the Technical Specifications. Tensile testing is part of the routine capsule test program, but the results are not applied to the embrittlement or USE calculations. Dosimetry measurements provide an independent check of the fluence models, but the official capsule fluence values, as provided in the report, are not affected by the measurement results. Tensile and dosimetry data will be included in the final report.

Entergy Nuclear Operations, Inc, will provide the final report to the NRC by July 30, 2004.

A 00%

NL-04-042 Page 2 of 2 There are no new commitments identified in this letter. If you have any questions, please contact Ms. Charlene Faison at 914-272-3378.

fred R. Dacimo 7071 'Site Vice President Indian Point Energy Center

Attachment:

Summary Report IPEC-RPT-04-00005 Rev 0, Preliminary Analysis of Capsule X, Indian Point 3 Reactor Vessel Surveillance Program cc:

Mr. Patrick D. Milano, Resident Inspector's Office Senior Project Manager Indian Point Unit 3 Project Directorate I, U.S. Nuclear Regulatory Commission Division of Reactor Projects I/Il P.O. Box 337 U.S. Nuclear Regulatory Commission Buchanan, NY 10511-0337 Mail Stop O-8-C2 Washington, DC 20555 Mr. Peter R. Smith, President NYSERDA Mr. Hubert J. Miller 17 Columbia Circle Regional Administrator, Region I Albany, NY 12203 U.S. Nuclear Regulatory Commission 475 Allendale Road Mr. Paul Eddy King of Prussia, PA 19406 New York State Dept. of Public Service 3 Empire Plaza Albany, NY 12223

ATTACHMENT I

SUMMARY

REPORT IPEC-RPT-04-00005 REV. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT 3 REACTOR VESSEL SURVEILLANCE PROGRAM ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

@ ENN QUALuTY RELATED ENN-DC147 Revision 2 Enfergy NUCLEAR ADMINISTRATIVE PROCEDURE MANAGEMENT S1 MANUAL INFOR uAn0NAL USE Page I11 of 13

. I.I ATTACHMENT 9.1 ENGINEERING REPORT COVER SHEET Engineering Report No. IPEC-RPT-04-00005 Rev. 0 Page I of 36

'Entergy ENTERGY NUCLEAR NORTHEAST Engineering Report Cover Sheet Summary Report IPEC-RPT-04-0005:

Preliminary Analysis of Capsule X Indian Point Unit 3 Reactor Vessel Surveillance Prozram Engineering Report Type:

New 3 Revision El Cancelled E] Superceded C]

Applicable Site(s)

II O P2 0 IP3 0 JAF El PNPS E Vw Quality-Relate>d: [I' N Prepared by: I 4&l;A.he Wsponsible Engmneer(rnKitgS att: 44Ž Verified/]

Reviewed by: 6&ACIA CA62Am . Date: yAlz/

Desiin Vedfler/Rie(P Name/Sign)

/J/f Approved by: 3zr Xp /k9S Date:

Supetvis&((Print Name/Sign)

Multiple Site Review Site Design Verifier/Reviewer (Print Name/Sign) Supervisor (Print Name/Sign) Date I: I"1

SUMMARY

REPORT IPEC-RPT-04-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM TABLE OF CONTENTS 1.0 Revision Summary 2 2.0 Purpose 3 3.0 Summary of Results 3 4.0 Evaluation 4 4.1 Charpy V-Notch Test Results 4 4.2 Credibility Evaluation 29 4.3 Tensile Data 35 4.4 Dosimetry Data 35 5.0 Conclusions 36 6.0 References 36 1.0 Revision Summary This is Rev 0 of the engineering report.

Page 2 of 36

SUMMARY

REPORT IPEC-RPT-04-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM 2.0 Purpose This report provides a summary of measurements associated with the testing program for Capsule X in the Indian Point Unit 3 (1P3) reactor vessel surveillance program. Capsule X was withdrawn from the IP3 reactor vessel at a service life of 15.5 Effective Full Power Years (EFPY) resulting in a predicted fluence of 0.874 x 1019 n/cm 2 .

The intent of the reactor vessel surveillance program is to validate the IP3 Technical Specifications associated with Reactor Coolant System (RCS) heatup, cooldown and setpoints for the Low Temperature Overpressure Protection System (LTOPS). In addition, the surveillance testing determines Upper Shelf Energy (USE) for limiting vessel materials at predicted End-of-Life (EOL) conditions.

The test results in this report are presented in accordance with the requirements of I0CFR50 Appendix H. The results are in the form of raw data and are in the process of formal review.

Upon completion of data review, Westinghouse will provide a formal report (WCAP-1625 1-NP)

(Reference 6. 1). Entergy will submit a copy of this report to the NRC no later than July 30, 2004.

The majority of the information provided in this report is taken directly from the preliminary version of Reference 6.1.

3.0 Summary of Results As noted below and in the sections that follow, the preliminary results of the testing confirm the validity of the Technical Specifications and identify EOL USE values greater than the required minimum of 50 ft-lbs for all limiting vessel plates and welds.

I) The measured 30 ft-lb shift in transition temperature values of the lower shell plate B2803-3 contained in capsule X (longitudinal & transverse) is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift values are less than two sigma allowance by Regulatory Guide 1.99, Revision 2 (Reference 6.2).

2) The measured 30 ft-lb shift in transition temperature value of the weld metal contained in capsule X is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift value is less than two sigma allowance by Regulatory Guide 1.99, Revision 2.
3) The measured 30 ft-lb shift in transition temperature value of the intermediate shell plate B2802-2 contained in capsule X (longitudinal) is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift value is less than two sigma allowance by Regulatory Guide 1.99, Revision 2.

Page 3 of 36

SUMMARY

REPORT IPEC-RPT-04-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM

4) The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules X contained in the Indian Point Unit 3 surveillance program are in good agreement with the Regulatory Guide 1.99, Revision 2 predictions.
5) All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the life of the vessel (27.1 EFPY) as required by 10CFR50, Appendix G.
6) The Indian Point Unit 3 surveillance data from the lower shell plate B2803-3 was found to be credible. This evaluation can be found in Section 4.2.
7) The calculated Chemistry Factor (CF), as defined in Reg Guide 1.99, Rev 2, is 167.9 degF, which is lower than the calculated CF of 170.9 degF currently in use at IP3 (References 6.3, 6.4). Therefore, the Technical Specifications controlling heatup, cooldown and LTOPS operation are slightly conservative for the present applicable service life of <20 Effective Full-Power Years.
8) A fluence evaluation utilizing the recently released neutron transport and dosimetry cross-section libraries was derived from the ENDF/B-VI database. Capsule X received a fluence of 0.874 x 1019 n/cm 2 after irradiation to 15.5 EFPY. The peak clad/base metal interface vessel fluence after 15.5 EFPY of plant operation was 5.86 x 10l1 n/cm 2 .

4.0 Evaluation 4.1 Charpy V-Notch Test Results The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule X, which received a fluence of 0.874 x 1O'9 n/cm2 (E> 1.0 MeV) in 15.5 EFPY of operation, are presented in Tables I through 10 and are compared with unirradiated results as shown in Figures 1 through 12.

The transition temperature increases and upper shelf energy decreases for the Capsule X materials are summarized in Table 9 and led to the following results:

Irradiation of the reactor vessel lower shell plate B2803-3 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation), resulted in an irradiated 30 ft-lb transition temperature of 191.6 0 F and an irradiated 50 ft-lb transition temperature of 223.80 F. This results in a 30 ft-lb transition temperature increase of 159.60 F and a 50 ft-lb transition temperature increase of 161.7°F for the longitudinal oriented specimens. See Table 9.

Irradiation of the reactor vessel lower shell plate B2803-3 Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of Page 4 of 36

SUMMARY

REPORT IPEC-RPT-04-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM 216.50 F and an irradiated 50 ft-lb transition temperature of 327.40 F. This results in a 30 ft-lb transition temperature increase of 158.2 0 F and a 50 fi-lb transition temperature increase of 217.91F for the longitudinal oriented specimens. See Table 9.

Irradiation of the weld metal (heat number W5214) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 128.5 0 F and an irradiated 50 ft-lb transition temperature of 196.80 F. This results in a 30 fl-lb transition temperature increase of 193.2 0 F and a 50 ft-lb transition temperature increase of 242.80 F. See Table 9.

Irradiation of the reactor vessel intermediate shell plate B2802-2 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation), resulted in an irradiated 30 ft-lb transition temperature of 98.10 F and an irradiated 50 ft-lb transition temperature of 145.00 F. This results in a 30 ft-lb transition temperature increase of 152.61F and a 50 ft-lb transition temperature increase of 166.5 0 F for the longitudinal oriented specimens. See Table 9.

The average upper shelf energy of the lower shell plate B2803-3 (longitudinal orientation) resulted in an average energy decrease of 24 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 81 ft-lb for the longitudinal oriented specimens. See Table 9.

The average upper shelf energy of the lower shell plate B2803-3 (transverse orientation) resulted in an average energy decrease of 24 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 52 ft-lb for the longitudinal oriented specimens.

See Table 9.

The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of 46 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 74 ft-lb for the weld metal specimens. See Table 9.

The average upper shelf energy of the intermediate shell plate B2802-2 (longitudinal orientation) resulted in an average energy decrease of 20 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 105 ft-lb for the longitudinal oriented specimens. See Table 9.

A comparison, as presented in Table 10, of the Indian Point Unit 3 reactor vessel surveillance material test results with the Regulatory Guide 1.99, Revision 2 predictions led to the following conclusions:

- The measured 30 ft-lb shift in transition temperature values of the lower shell plate B2803-3 contained in capsule X (longitudinal & transverse) is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, Page 5 of 36

SUMMARY

REPORT IPEC-RPT-04-00005 Rev'. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM the shift value is less than two sigma allowance by Regulatory Guide 1.99, Revision 2.

- The measured 30 ft-lb shift in transition temperature value of the weld metal contained in capsule X is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift value is less than two sigma allowance by Regulatory Guide 1.99, Revision 2.

- The measured 30 ft-lb shift in transition temperature values of the intermediate shell plate B2802-2 contained in capsule X (longitudinal) is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift value is less than two sigma allowance by Regulatory Guide 1.99, Revision 2.

- The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules X contained in the Indian Point Unit 3 surveillance program are in good agreement with the Regulatory Guide 1.99, Revision 2 predictions.

All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the extended life of the vessel (27.1 EFPY) as required by IOCFR50, Appendix G.

The fracture appearance of each irradiated Charpy specimen from the various surveillance Capsule X materials shows an increasingly ductile or tougher appearance with increasing test temperature.

The Charpy V-notch data presented in WCAP-8475, WCAP-9491, WCAP-1 0300, and WCAP-1 1815 (i.e., the surveillance program and the three other surveillance capsule reports) were based on hand-fit Charpy curves using engineering judgment. However, the results presented in this report are based on a re-plot of all applicable capsule data using CVGRAPH, Version 5.0.2, which is a hyperbolic tangent curve-fitting program.

Page 6 of 36

SUMMARY

REPORT IPEC-RPT-04-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X 6 INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM Table 1 Charpy V-notch Data for the Indian Point Unit 3 Lower Shell Plate B2803-3 Irradiated to a Fluence of 0.874 x 10'9 n/cm2 (E> 1.0 MeV) (Longitudinal Orientation)

Sample Temperature Impact Energy Lateral Expansion Shear Number OF l C ft-lbs . Joules mils mm  %

A37 100 38 7 9 2 0.05 10 A34 150 66 21 28 14 0.36 15 A36 175 79 22 30 15 0.38 20 A33 200 93 27 37 18 0.46 40 A40 225 107 51 69 36 0.91 70 A39 280 138 82 111i 59 1.50 100 A35 350 177 78 106 57 1.45 100 A38 375 191 83 113 68 1.73. 100 Page 7 of 36

SUMMARY

REPORT IPEC-RPT-04-0005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM Table 2 Charpy V-notch Data for the Indian Point Unit 3 Lower Shell Plate B2803-3 Irradiated to a Fluence of 0.874 x lot, n/cm 2 (E> 1.0 MeV) (Transverse Orientation)

Sample Temperature. Impact Energy Lateral Expansion Shear Number OF C ft-lbs Joules mils mm  %

AT64 100 38 6 8 0 0.00 15 AT69 175 79 20 27 14 0.36 25 AT68 210 99 22 30 14 0.36 30 AT67 225 107 33 45 .25 0.64 60 AT66 250 121 44 60 34 0.86 95 AT65 325 163 47 64 38 0.97 100 AT62 375 191 54 73 45 1.14 100 AT63 390 199 55 75 45 1.14 100 Page 8 of 36

SUMMARY

REPORT IPEC-RPT-04-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X 0 INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM Table 3 Charpy V-notch Data for the Indian Point Unit 3 Surveillance Weld Metal Irradiated to a Fluence of 0.874 x 1019 n/cm 2 (E> 1.0 MeV)

Sample Temperature Impact Energy Lateral Expansion Shear Number OFCC ft-lbs . Joules mils mm  %

W42 75 24 9 12 5 0.13 20 W41 125 52 49 66 36 0.91 50 W43 125 52 24 33 19 0.48 40 W48 150 66 35 47 26 0.66 45 W47 200 93 37 50 30 0.76 70 W44 250 121 67 91 52 1.32 95 W45 300 149 72 98 56 1.42 98 W46 350 177 75 102 57 1.45 100 Page 9 of 36

SUMMARY

REPORT IPEC-RPT-04-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM Table 4 Charpy V-notch Data for the Indian Point Unit 3 Intermediate Shell Plate B2802-2 Irradiated to a Fluence of 0.874 x 109 n/cm 2 (E> 1.0 MeV)

Sample Temperature Impaci Energy Lateral Expansion Shear Number OF l Ft-lbs Joules mils mm l  %

N2 25 -4 8 11 3 0.08 5 N6 75 24 24 33 14 0.36 15 N5 125 52 59 80 40 1.02 30 N7 150 66 40 54 .30 0.76 55 N4 200 93 58 79 44 1.12 65 NI 250 121 104 141 69 1.75 100 N8 300 149 105 142 71 1.80 .100 N3 325 163 105 142 68 1.73 100 Page 10 of 36

Table 5 Instrumented Charpy Impact Test Results for the Indian Point Unit 3 Lower Shell Plate B2803-3 Irradiated to a Fluence of 0.874 x 10s' n/cm2 (E>1.0 MeV) (Longitudinal Orientation)

Charpy Normalized Energies Yield Time to Time to Fast (ft/in) Load Yield MaX. MaX. Fract. Arrest Yield Flow Test Energy Charpy MAL Prop. PC', tcy Load tM Load PF Load PA Stress Stress Sample Temp. ED (ft-lb) EJ/A EM/A EPIA (Ob) (msec) PM (lb) (msec) (lb) (lb) ary (ksi) (ksl)

NO. (OF) 7 56 19 37 2108 0.11 2304 0.13 2304 363 70 73 A37 100 21 169 68 101 3363 0.14 4187 0.22 4047 372 112 126 A34 150 22 177 68 110 3336 0.14 4173 0.22 4090 609 III 125 A36 175 27 218 65 152 3311 0.14 4061 0.22 3913 1082 110 123 A33 200 411 226 185 3331 0.14 4567 0.50 4529 2496 111 132 A40 225 51 82 661 236 425 3336 0.14 4671 0.51 n/a n/a III 133 A39 280 78 628 225 403 3176 0.14 4480 0.51 n/a n/a 106 127 A35 350 375 83 669 222 447 3165 0.14 4344 0.51 n/a n/a 105 125 A38 Page 11 of 36

Table 6 Instrumented Cbarpy Impact Test Results for the Indian Point Unit 3 Lower Shell Plate B2803-3 Irradiated to a Fluence of 0.874 x 1019 n/cm1 (E>1.O MeV) (Transverse Orientation)

Charpy Normalized Energies Yield Time to Time to Fast Test EnergY (ft-1b/in Load Yield MAS. Mal. Fract. Arrest Yield Flow Sample Temp. ED Charpy MaL Prop. PGY tGy Load tM LOad PF Load PA Stress Stress No. (F) (ft-lb) EJIA EM/A EV/A (lb) (msec) PM (lb) (msec) (Ob) (lb) ay (ksi) (ksi) 100 6 48 14 34 1416 0.09 1672 0.12 1659 455 47 51 AT64 20 161 68 93 3310 0.14 4149 0.22 4090 683 110 124 AT69 175 210 22 177 67 110 3407 0.15 . 4091 0.22 3927 987 113 125 AT68 33 266 66 200 3380 0.14 4113 0.21 4017 2273 113 125 AT67 225 44 355 162 193 3091 0.14 4136 0.41 3999 2397 103 120 AT66 250 47 379 143 236 3089 0.13 4027 0.37 3853 1829 103 118 AT65 325 54 435 162 274 2969 0.13 4063 0.41 n/a n/a 99 117 AT62 375 AT63 390 55 443 148 295 3028 0.13 4080 0.38 n/a n/a 101 118 Page 12 of 36

Table 7 Instrumented Charpy Impact Test Results for the Indian Point Unit 3 Surveillance Weld Metal Irradiated to a Fluence of 0.874 x 10" n/cm2 (E>1.0 MeV)

Charpy Normalized Energies Yied Time to Time to Fast Test Energy (ft-lbfin 2) Load Yield Mal. Max. Fract. Arrest Yield Flow Sample Temp. ED Charpy Max. Prop. PGc tGY Load tm Load Pr Load Stress Stress No. (VF) (ft-lb) EJ/A Em/A EV/A (lb) (msec) PM (lb) (msec) (lb) PA (lb) ay (ksi) (kil)

W42 75 9 73 36 36 3426 0.14 3696 0.16 3687 0 114 119 W41 125 49 395 226 169 3411 0.15 4363 0.52 4288 617 114 129 W43 125 24 193 68 126 3341 0.14 4109 0.22 4058 1313 III 124 W48 150 35 282 184 98 3416 0.14 4449 0.42 4417 1141 114 131 W47 200 37 298 150 148 3371 0.14 4260 0.37 4222 1713 112 127 W44 250 67 540 227 313 3486 0.14 4432 0.50 4251 2819 116 132 W45 300 72 580 218 362 3329 0.14 4303 0.50 3029 2501 II 127 W46 350 75 604 221 383 3285 0.14 4309 0.51 n/a n/a 109 126 Page 13 of 36

Table 8 Instrumented Charpy Impact Test Results for the Indian Point Unit 3 Intermediate Shell Plate B2802-2 Irradiated to a Fluence of 0.874 x 10w' n/em2 (E>1.0 MeV) (Longitudinal Orientation)

Charpy Normalized Energies Yield Time to Time to Fast EnerV (ft-lblin ) Load Yield Max Max. Fract. Arrest Yield Flow Test Temp. ED Charpy Max. Prop. PGY tC Load tH Load Pr Load Stress cy Stress Sample No. (OF) (ft-lb) ED/A EN/A E/A (lb) (msec) PM (b) (msec) (b) PA (lb) (ksi) (ksi)

N2 25 8 64 35 29 3649 0.15 3761 0.16 3761 0 121 123 N6 75 24 193 146 47 3388 0.14 4306 0.36 4303 0 113 128 N5 I ^5 59 475 327 148 3470 0.15 4594 0.68 4444 687 116 134 N7 150 40 322 185 138 3292 0.14 4338 0.44 4332 1570 110 127 N4 200 58 467 231 237 3239 0.14 4423 0.53 4369 2112 108 128 N1 250 104 838 311 527 3193 0.15 4446 0.68 n/a n/a 106 127 Ng 300 105 846 312 534 3272 0.14 4494 0.67 n/a n/a 109 129 N3 325 105 846 302 544 3068 0.14 4374 0.67 n/a n/a 102 124 Page 14 of 36

Effect of Irradiation to 0.874 x 101 n/cm2 (E>1.0 MeV) on the Capsule "X" Notch Toughness Properties of the Indian Point Unit 3 Table 9 Reactor Vessel Surveillance Materlaiste)

Average 30 (ft-lb)° Average 35 mil Lateral(b) Average 50 ft-lb(s) Average Energy Absorption("

Expanslon Temperature (IF) Transition Temperature (IF) at Full Shear (ft-lb)

Material Transition Temperature (IF)

Irradiated AT Unirradiated Irradiated AT Uniradiated Irradiated AE Uninfadiated Irradiated AT Unirradiated 225.0 176.4 62.1 223.8 161.7 105 81 -24 LowerShellPlate 32.0 191.6 159.6 48.6 B2803-3 (Long.)

158.2 75.3 269.3 194.0 109.5 327.4 217.9 68 52 .16 Lower Shell Plate 58.3 216.5 B2803-3 (Trans.)

-59.3 184.6 243.9 -46.0 196.8 242.8 120 74 -46 Weld Meul -64.7 128.5 193.2 (Heat # W5214)

_38.6 147.3 185.9 -21.5 145.0 166.5 125 105 -20 Inter. Shell Plate -54.5 98.1 152.6 B2802-2 (Long.)

1,4,7 and 10).

a. "Average" is defined as the value read from the curve fit through the data points of the Charpy tests (see Figures 2,S, 8 and I1).
b. "Average" is defined as the value read from the curve fit through the data points of the Charpy tests (see Figures Page 15 of 36

Table 10 Comparison of the Indian Point Unit 3 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence(d) Predicted Measured Predicted Measured (X 1019 n/Cm2 , (0 F) (") . (OF) (b) (%) () (%)(c)

E > 1.0 MeV)

Lower Shell Plate T 0.263 101.9 139.4 24 12 B2803-3 Z 1.04 161.6 167.8 33.5 22 (Longitudinal) X 0.874 153.9 .159.6 32 23 Lower Shell Plate T 0.263 101.9 105.9 24 16 B2803-3 Y 0.692 143.5 148.9 30 25 Z 1.04 161.6 157.9 33.5 18 (Transverse) X 0.874 153.9 158.2 32 24 Surveillance T 0.263 102.6 151.6 22 30 Program Y 0.692 144.4 172.0 26 43 Weld Metal Z 1.04 162.6 229.2 32 37 X 0.874 154.9 196.8 29 38 Intermediate Shell Plate X 0.874 146.2 152.6 30 16 B2802-2 (Longitudinal) _ ..

Notes:

(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material.

(b) Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1 (c) Values are based on the definition of upper shelf energy given in ASTM El 85-82.

(d) The fluence values presented here are the calculated values, not the best estimate values.

Page 16 of 36

SUMMARY

REPORT IPEC-RPT-00400005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X 0 INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM I

LOWER SHELL PLATE B2803-3. (LONGITUDINAL ORIENTATION)

CVGRlAPH 5.02 Hyprbolic Tangent Cte Printed en 04102J2004 04:43 PM Dea Set(s) Plotted

  • Curve "lot CapJk Materiel ODL Heat Indian Point 3 SA302B LT A-9512-2 2 Indian Pont 3 SA3O2B LT A0512-2 3 IndiwiPoint 3 z SA302B LT A-0512-2 4 IndimnPoint 3 x . SA302B LT A-0512-2 n

.C W1-__

r1200--

_W- ----

__ __I A- I I-i t 0

100 a a

n - w- - --

-300 -200. -100 0 100 200 300 400 500 600 Temperatur In Deg F 0 Set I a Set 2 O Set 3 a Set4 Curn Fba LSE USE d6-UE T 03 d-T 030 T 050 d-T 050 2.2 105.0 .0 32.0 .0 62.1 .0 2J 2.2 92.0 -13.0 171.4 139.4 200.2 138.1 3 2.2 82.0 -23.0 199. 8 167.8 242.7 180.6 4 2.2 tl.0 .24.0 191.6 159.6 223.8 161.7 Figure 1 Charpy V-Notch Impact Energy vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Longitudinal Orientation)

Page 17 of 36

SUMMARY

REPORT IPEC-RPT-04-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM LOWER SHELL PLATE B2803-3 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.02 Hyperbolic Tanipt Cum Print on 04/2004 04:57 FM

  • Data Set(s) Ponctd Curve Plant Capsule Materal OrL Heatt Indim Point 3 UNIRR SA302B LT A-0512-2 2 Indin Point 3 T SA302B LT A-0512-2 3 Indian Point 3 z SA302B LT A-0512-2 4 Indian Point 3 X I SA302B LT A.0512-2' E

.9 a.

-300 0 300 600 Temperature In Deg F 0 Set I a Set 2 0 Set 3 A Set 4 Rewib Curve nMe ISE US: d-USE T S3S i-T O35

.0 7B.7 .0 41.6 .0 2 .0 65.5 -12.9 155.4 136. 5 3 .0 73.4 -5.2 21t.3 169.7 4 .0 65.1 -13.5 225.0 176.4 Figure 2 Charpy V-Notch Lateral Expansion vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Longitudinal Orientation)

Page 18 of 36

SUMMARY

REPORT IPEC-RPT-004*0005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM LOWER SHELL PLATE B2803-3 (LONGITUDINAL ORIENTATION)

CVGRAPH .0.2 Hypebolic Tenat Cuve Printed on 04Wn22004 04:53 PM Data Set(s) Plotted

  • Ciurve ant Capstik Mntl OrL Beat#

1 Indian Pont 3 UNMR SA302B LT A-OS12-2 2 Mnain Point 3 T SA302B LT A-0512-2 3 Indian Point 3 z SA302B LT A-0512-2 4 Indian Point 3 x , SA302B LT A-0512-2 1'w w7 . .

104-I 7 A! 5a 5 0 - . .

2 a

a - I

-_30 -200 -100 0 100 200 300 400 SOO 800 Temperature In Deg F o set t a Set 2 0 Set 3 A Set 4 FbNs" LsE U -USE T 0SO d-T S50

.0 I00.0 .0 6S.S .0 2 .0 100.0 .0 204.7 135.9 3 .o 100.0 .0 205.3 136.5 4 .o .100.0 .0 206.3 137.5 Figure 3 Charpy V-Notch Percent Shear vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Longitudinal Orientation)

Page 19 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM LOWER SHELL PLATE B2803-3 (TRANSVERSE ORIENTATION)

CVGRAPH 5.02 Hyperbolic Tangeat Cme Pzinted ma04/OZ7004 05:10 PM Dam Set(s) Plofzt Curl yPant Capsule Material OL. Beat J 1 JndiauPoint3 UNIRk SA302B . T A-0S12-2 2 Mndiu Pdnt3 T SA302B it A0512-2 3 rhdianPoint 3 Y SA3O2B 7t A-0512.2 4 IntdianPint3 Z SA302B it A-0512-2 5 IndianPoint3 X SA302B TL A-0512-2 4 200 U1.

P 150 a

'U t 100

-3 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F 0 Set I a Set 2 o Set 3 a Set 4 Set 5 Rensult curyt nlef LSE USE a-USE T 030 dQ 038 T O5e d-T S5I 2.2 68.0 .0 55.3 .0 W09.S .0 2 2.2 57.0 .11.0 164.2 105.9 256.4 146.9 3 2.2 51.0 -17.0 207.2 148.9 362.1 252.6 2.2 56.0 -i2.0 216.2 157.9 266.7 157.2 5 2.2 52.0 -36.0 216.5 158.2 327.4 217.9 Figure 4 Charpy V-Notch Impact Energy vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Transverse Orientation)

Page 20 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULEX X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM LOWER SHELL PLATE 2803-3 (TRANSVERSE ORIENTATION)

CVGRAPH 5.02 Hypcuoc Tant Cwe Purnted on 04004 O5.34 PM Data Set(s) Plotted Curre Plant Capsule Materl OrL Heat #

I Indimn Point 3 UNIR SA302B it A-0512-2 2 Indi nPoint3 T SA302B 1L A-0512-2 3 Indian Point 3 Y SA302B . 'L A-0512-2 4 IndianPoint3 Z , . SA302B IL A.0512-2 5 Indian Point 3

  • X SA302B 1t A-0512-2

. 4. . 4 A

.2 a

I

-300 0 300 600 Tempemature In Deg F o Set 1 a Set 2 0 Set 9 A Set 4 Set 5 R mu '

cum nac ISE USE d-USE T @35 d-T 035

.0 63.1 .0 75.3 0 2 .0 76.0 12.9 209.2 133.9 3 .0 53.3 -9.3 254.3 179.0 4 .0

  • 53.9 -9.2 232.8 157.5 5 .0 44.4 -13.7 269.3 194.0 Figure 5 Charpy V-Notch Lateral Expansion vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Transverse Orientation)

Page 21 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM LOWER SHELL PLATE B2803-3 (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hypcbolic Tangcnt Cum Phinted on 04102r2004 05:16 PM Data Set(s) Plotted Cume Plant Capsule Materal Oi. Rest I Indian Point 3 UNIRR SA302B TL A-0512-2 2 Indian Point 3 T SA302B TL A-0512-2 3 Indian Point 3 Y SA302B *L A-0512-2 4 Indian Point 3 Z SA302B 1L A-0512-2 5 Indian Point3 X SA302B TL A-0512-2

. U

-300 -200 -100 0 100 200 300 400 500 600 Temperntum In Deg F 0 Set I a Set 2 o Set 3 a Set 4 Set 5 RIt -

cumn Fbe 15K US1 d.aUSIE TO50 d-T 050

.0. 100.0 .0 92.2 .0 i 2 .0 100.0 .0 200.4 108.2 3 .0 100.0 .0 221.9 129.7 4 .0 100.0 . .0 203.3 111.1

.0 100.0 .0 217.2 124.9 S

Figure 6 Charpy V-Notch Percent Shear vs. Temperature for Indian Point Unit 3

.Reactor Vessel Lower Shell Plate B2803-3 (Transverse Orientation)

Page 22 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM SURVEILLANCE WELD MATERIAL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printod on 04102/2004 02:31 PM Data Set(s) Plotted Curve Plant Capsule Material Orl Heat #

1 Indian Point 3 UNIRR SAW NA W5214 2 Indian Point 3 T SAW NA W5214 3 Indian Point 3. y SAW NA W5214 4

5 Indian Point 3 Indian Point 3 z SAW NA WS214 WS214 SAW NA

.,juv I 250 - - __-

-T 200 II I I

w ISO - -

P I

I1nn II p-i_ o._

50 61

.300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F a Set 1 a Set 2 0 Set 3 A Set 4 Set 5 Resunt Curve Flue.ce ISE USE d-USE T Q30 d.T 030 T @50 d.T QSD 2.2 120.0 .0 -64.7 .0 -46.0 * .0 2 2.2 34.0 -36.0 36.9 151. 6 130. 6 176.6 3 2. 2 69.0 -51. 0 107. 3 172.0 164.2 210.2 4 2.2 76.0 -44. 0 164.5 229.2 213.7 264.7 S 2.2 74. 0 -46.0 123.5 193.2 196. 242.

Figure 7 Charpy V-Notch Impact Energy vs. Temperature for Indian Point Unit 3 Reactor Vessel Weld Metal Page 23 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM SURVEILLANCE WELD MATERIAL CVGRAPH 5.0.2 Hyperbolic Tangent Cwue Printed on 04/02/2004 04:11 PM Data Set(s) Plotted Curve Plant Capsule Material ' . Hleat #

1 Indian Point 3 UNIRR SAW NA WS214 2 Indian Point 3 *SAW NA W5214 3 Indian Point 3 y SAW NA W5214 4 Indian Point 3 z SAW NA W5214 S Indian Point 3 X SAW NA W5214 A

C

.2 a

I di

-300 0 300 600 Temperature In Deg F 0 Set I a Set 2 0 Set 3 a Set 4 Set 5 eumts Curve Pluencg ISE USE d-USE T Q35 d-T Q35 1 .0 90.3 .0 -59.3 .0 2 .0 FO. 3 -310. 5 113.3 172.6 3 .0 65.6 -25.2 145.0 204. 3 4 .0 73. 5 - 17.3 187. 8 247. 1 5 .0 62.1 -28. 7 184.6 243.9 Figure 8 Charpy V-Notch Lateral Expansion vs. Temperature for Indian Point Unit 3 Reactor Vessel Weld Metal Page 24 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM SURVEILLANCE-WELD MATERIAL CVGRAPH 5.0.2 Hlypboac Tangent Curve Puinted on 0412004 03:57 PM Data Set(s) Plotted Curve Pant Capsule Material Orl Hleat#

Indian Point 3. UNIUk SAW NA W5214 2 Indian Point 3 T y

SAW NA W5214 3 Indian Point 3 SAW NA W5214 4 Indian Point 3 SAW NA W5214 5 Indian Point 3 xz SAW NA W5214 o - . -- I _..a. I -. I-II

-300 -200 -100 0 100 200 300 400 soO 600 Temperature In Dog F C0 Set I D Set 2

  • Set 3 A Set 4 sot 5 Cw" n~ac tsR USe d4USE T 050 .- T @30

.0 100.0 .o -47.3 .0 2 .0 100.0 .0 124.0 171. s 3 .0 100.0 .0 132.6 150.4 4 .0 100.0 .o. 147.5 195.3 5 .0 100.0 .o 1a:5 192.3 Figure 9 Charpy V-Notch Percent Shear vs. Temperature for Indian Point Unit 3 Reactor Vessel Weld Metal Page 25 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM INTERMEDIATE SHELL PLATE B2802-2 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hypnbolic Tangent Cume Printed on 0402U1004 05:42 PM Data Set(s) Plotted Curve Plant Capsule Mattrbl OrL Heat#

2 Idian Point 3 UNxR *SA302B LT A.0516-2 2 Inhan Point 3 X SA302B LT A0516-2 300

%?!!.n ,

200 IL 6 100 _1_ _ __

50 6K

.a

^

n I

-r--

l _ ,- .

.1, ^

U-_ .200 -100 aI 100 200 30 400 500 600 Temperature In Deg F 0 Set I c Set2 Cum nlae 15E USE d.un S@030 6T7030 TOSO d-T OS0 2.2 . 125.0 .0 -54.5 .0 -21.5 .0 2 2.2 105.0 -20.0 91.1 152.6 145.0 166.5 Figure 10 Charpy V-Notch Impact Energy vs. Temperature for Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2 (Longitudinal)

Page 26 of 36

SUMMARY

REPORT IPEC.RPT-004-OU005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM INTERMEDIATE SHELL PLATE B2802-2 (LONGITUDINAL ORIENTATION) r-Iv. A u C n 7. Uvnsnlir T"...A -

hn flANV"P7( nt-A MA Data Set(s) PlIoted Plant Capsule Mate - OrTl Heat Indi nPoint 3 UNIRR SA302B LT A-0516-2 2 IndisnPoint3 X SA302B LT AM0516-2 200

.3 IC

.2 a

IL 01 - __ __ __ __

-300 0 300 600.

Temperature In Dog F 0 Sit o Set 2 Remuts Iluence L USE d-USE T 035 d-T @35

.0 79.1 .0 -31. 6 .0 2 .0 75.2 -4.6 147.3 115.9 Figure 11 Charpy V-Notch Latteral Expansion vs. Temperature for Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2 (Longitudinal)

Page 27 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM INTERMEDIATE SHELL PLATE B2802-2 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.02 Hyperbolic Tangent Curve Printed on 04102fl004 05:45 PM r%-. af.l nt 1%61 rj.

Curve Plant capsule Material 0n. Heat #

2* Indian Point 3 CN4 SA3O2B * " LT A40516-2 2 Indian Point 3 x SA302B *LT A-0516-2 75 - - L 50- a

  • it - -

i ny

-300 -200 K

-100 0 100 200 300 400 500 600 Temperature In Deg F 0 Set I a Set 2 Fhwm Ls USE d-USE TQ 5b d-T 05

.0- 100.0 .0 22.6 .0 2 .0 100.0 .0 153.2 130.6 Figure 12 Charpy V-Notch Percent Shear vs. Temperature for Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2 (Longitudinal)

Page 28 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM 4.2 Credibility Evaluation Regulatory Guide 1.99, Revision 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

To date there has been four surveillance capsules removed from the Indian Point Unit 3 reactor vessel. To use these surveillance data sets, they must be shown to be credible. In accordance with the discussion of Regulatory Guide 1.99, Revision 2, there are five requirements that must be met for the surveillance data to be judged credible.

The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Indian Point Unit 3 reactor vessel surveillance data and determine if the Indian Point Unit 3 surveillance data is credible.

It should be noted that only surveillance plate B2803-3 will be evaluated for credibility for the following reasons: 1) The surveillance plates B2802-1, 2, and 3 do not contain sufficient irradiated data sets to be used in vessel material predictions, 2) The limiting surveillance plate B2803-3 has a significantly larger initial RTNDT, where the remaining surveillance materials could not become limiting even with non-credible surveillance data (i.e. using axill margin term). 3) The surveillance weld heat is not the same heat as the beltline welds (intermediate/lowershell longitudinalweld

& intermediateto lower shell girth *weld), thus should not be used for vessel material predictions (see discussion under Criterion 1).

Page 29 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements", as follows:

"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 radiation damage to be considered in the selection of the most limiting material with regard to radiation damage."

The Indian Point Unit 3 reactor vessel consists of the following beltline region materials:

  • Intermediate Shell Plates B2802-1, 2, 3
  • Lower Shell Plates R2803-1, 2, 3
  • Intermediate & Lower Shell Longitudinal Weld Seams (Heat # 34B009, Flux Type Linde 1092),
  • Intermediate to Lower Shell Circumferential Weld Seam (Heat # 13253, Flux Type Linde 1092).

Per WCAP-8475, the Indian Point Unit 3 surveillance program was based on ASTM El 85-62. When the surveillance program material was selected it was believed that copper and phosphorus were elements most important to embrittlement of the reactor vessel steels. Lower shell plate B2803-3 had the highest copper weight percents, the highest initial RTNDT and the lowest USE of all plate materials in the beltline region. Thus, it was selected as one of the beltline plate materials included in the surveillance capsules. Since Indian Point Unit 3 had eight surveillance capsules, there was sufficient room for additional plate materials, thus, specimens from each of the intermediate shell plates were also included, but not to the extent as lower shell plate B2803-3.

The weld material in the Indian Point Unit 3 surveillance program was made of the weld wire heat W5214, flux type linde 1092. This is the same heat as that from the nozzle shell longitudinal welds, but the same flux type as those welds within the beltline region. In addition, predictions made at the time the capsule program was developed indicated that weld heat W5214, linde 1092 would produce similar predictions as those weld heats within the beltline region and thus deemed the surveillance weld heat W5214 representative of the beltline region.

Therefore it was chosen as the surveillance weld material.

Page 30 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM Hence, Criterion 1 is met for the Indian Point Unit 3 reactor vessel.

Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper shelf energy unambiguously.

Based on engineering judgment, the scatter in the data presented in these plots is small enough to permit the determination of the 30 ft-lb temperature and the upper shelf energy of the Indian Point Unit 3 surveillance materials unambiguously. Hence, the Indian Point Unit 3 surveillance program meets this criterion.

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of ARTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 280 F for welds and 17 0 F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82.

The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these ARTNDT values about this line is less than 281F for welds and less than 17'F for the plate. Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2.

Page 31 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM TABLE 11 Calculation of Chemistry Factors using Indian Point Unit 3 Surveillance Capsule Data Material Capsule Capsule FF(b) ARTNDT(C) FF*ARTND FF2

01) T Lower Shell T 0.263 0.637 139.4 88.798 0.406 Plate B2803-3 Z 1.04 1.01 167.8 169A78 1.02 (Longitudinaol) X 0.874 0.962 159.6 153.535 0.925 Lower Shell T 0.263 0.637 105.9 67.458 0.406 Plate B2803-3 Y 0.692 0.897 148.9 133.563 0.805 (Transverse) z 1.04 1.01 157.9 159.479 1.02 X 0.874 0.962 158.2 152.188 0.925 SUM: 924.499 5.507 CF 2&03.3 = X(FF z ARTKDT) . X( FF2) = (924.499) + (5.507) = 167.91F Notes:

(a) f = fluence. Calculated fluence = [x lol 9 n/cm2 , E > 1.0 MeV].

(b) FF = fluence factor =f° zs *O.Ilog .

(c) ARTNDT values are the measured 30 fl-lb shift values taken from Figures I and 4, herein [IF].

The scatter of ARTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table 12.

Page 32 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM TABLE 12 Indian Point Unit 3 Surveillance Capsule Data Scatter about the Best-Fit Line for Surveillance Forging Materials.

CF Measured Predicted Scatter < 170 F

(°F) ARTNDT ARTNDTr ARTNDT (OF, (Base Metals)

Lower Shell T 167.9 0.637 139.4 107.0 32.4 No B280P 3 Z 167.9 1.01 167.8 169.6 -1.8 Yes (Longitudinal) X 167.9 0.962 159.6 161.5 -1.9 Yes Lower Shell T 167.9 0.637 105.9 107.0 -1.1 Yes Plate Y 167.9 0.897 148.9 150.6 -1.7 Yes B2803-3 ____ ___

(Transverse) Z 167.9 1.01 157.9 169.6 -11.7 Yes X 167.9 0.962 158.2 161.5 -3.3 Yes NOTES:

(a) Predicted ARTNDT (CF

Table 12 indicates that only I of 7 data points falls outside the +1-lc of 170 F scatter band for the lower shell plate B2803-3 surveillance data. One out of 7 data points is still considered credible. Therefore the lower shell plate B2803-3 surveillance data is deemed "credible" per the third criterion.

Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the claddingfbase metal interface within +/- 25 0F.

The capsule specimens are located in the reactor between the thermal shield and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the Thermal Shield. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 250 F. Hence, this criterion is met.

Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.

Page 33 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM The Indian Point Unit 3 surveillance program does contain correlation monitor material, but not in Capsule X. Past capsule results for the correlation monitor material is contained in NUREG/CR-6413, ORNL/TM-13133, which shows a plot of residual vs. fast fluence. The data shown in this report indicates that the CMM tested to date falls within acceptable limits. Hence, this criterion is met.

Therefore, based on the preceding responses to all five criteria of Regulatory Guide 1.99, Revision 2, the Indian Point Unit 3 surveillance plate (B2803-3) is credible.

Page 34 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM 4.3 Tensile Testing All tensile testing of the applicable Capsule X materials has been performed, but the testing results have not been compiled or reviewed at this time. However, tensile results do not affect the validation of the Technical Specifications nor the EOL USE calculations. Therefore, it is acceptable to conclude that the tensile test results, when completed, will not result in any required change to plant design basis documents.

A complete analysis of tensile testing data will be included with the final report.

4.4 Dosimetry Dosimetry data has been gathered and is being reduced at this time. The dosimetry results do not directly affect the Technical Specifications validation, since the associated evaluations (such as those involving the determination of Chemistry Factors or Fluence Factors) are based on calculated fluence.

A fluence evaluation utilizing the recently released neutron transport and dosimetry cross-section libraries was derived from the ENDF/B-VI data-base.

Capsule X received a fluence of 0.874 x 1019 n/cm2 after irradiation to 15.5 EFPY. The peak clad/base metal interface vessel fluence after 15.5 EFPY of plant operation was 5.86 x 101 n/cm2 .

Fluence is determined to ensure consistency between measured and predicted values. Per Reference 6.6, this is required to be within 20%. Based on experience gained from the results of the three previous surveillance capsules, actual agreement is expected to be well within this criterion. Dosimetry measurements are not a required parameter for reporting under 10CFR50 Appendix H.

The predictions of fluence factors through license expiration are consistent with those that have previously been transmitted to the NRC (Reference 6.5) as part of the Appendix K 1.4% power uprate project. The calculated fluences were evaluated at that time to include the most up-to-date methodology, and the same considerations are unchanged for this application.

A detailed evaluation of measured versus predicted fluence will be included in the final report.

Page 35 of 36

SUMMARY

REPORT IPEC-RPT-004-00005 Rev. 0 PRELIMINARY ANALYSIS OF CAPSULE X INDIAN POINT UNIT 3 REACTOR VESSEL SURVEILLANCE PROGRAM 5.0 Conclusions The preliminary results of the Capsule X data indicate that the IP3 Technical Specifications and EOL USE parameters for the reactor vessel remain valid and do not require revision. These conclusions are pending final review and verification of the data by Westinghouse and will be confirmed upon issuance of the final WCAP report.

6.0 References 6.1 WCAP-16251 -NP Rev 0 (Preliminary), "Analysis of Capsule X from Entergy's Indian Point 3 Reactor Vessel Surveillance Program,"

Westinghouse Electric Co, April 2004 (preliminary) 6.2 Regulatory Guide 1.99 Rev 2, "Radiation Damage to Reactor Vessel Materials," USNRC, February 1986 6.3 WCAP-16037 Rev 1, "Final Report on Pressure-Temperature Limits for Indian Point Unit 3 NPP," Westinghouse Electric Co, May 2003 6.4 Indian Point Unit 3 Technical Specifications Amendment 220, December 2003 6.5 Entergy Report IP3-RPT-MULT-03614, Rev 0, "Indian Point Nuclear Generating Unit Number 3; 1.4% Measurement Uncertainty Recapture Power Uprate Application Report," May 2002 6.6 Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," USNRC, Office of Nuclear Regulatory Research, March 2001 Page 36 of 36

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