NL-04-092, Capsule X Material Surveillance Report

Capsule X Material Surveillance Report

ML042320088
Person / Time
Site:	Indian Point
Issue date:	07/29/2004
From:	Kansler M Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-04-092 WCAP-16251-NP Rev 0
Download: ML042320088 (205)

Text

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

"v Entergy 440 Hamilton Avenue White Plains, NY 10601 Tel 914 272 3200 Fax 914 272 3205 Michael R. Kansler President July 29, 2004 NL-04-092 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Indian Point Nuclear Generating Unit No. 3 Docket No. 50-286 Capsule X Material Surveillance Report

Reference:

1. Entergy letter to NRC, NL-04-042, regarding "Reactor Vessel Material Surveillance Program: Preliminary Analysis Results for Capsule X',

dated April 19, 2004.

Dear Sir:

Pursuant to Appendix H tol 0 CFR 50, Reference 1 provided a summary technical report, "Summary Report IPEC-RPT-04-00005 Rev. 0, Preliminary Analysis of Capsule X - Indian Point Unit 3 Reactor Vessel Surveillance Program". In Reference 1, Entergy Nuclear Operations, Inc. (ENO), also committed to provide a final report to the NRC by July 30, 2004. Attachment 1 to this letter transmits the final report WCAP-1 6251 -NP, entitled, "Analysis of Capsule X from Entergy's Indian Point 3 Reactor Vessel Radiation Surveillance Program".

The final report differs from the preliminary report (Reference 1) In that It includes dosimetry data and tensile test results. It also revises the Surveillance Program Weld Metal temperature shift data on Table 5.10 (shown as Table 10 in Reference 1). As the Indian Point 3 reactor vessel Is not weld limited, and as the revisions result in a better predicted-to-measured ratio, these revisions have no impact on the preliminary report's conclusions. Several minor typographical errors were corrected.

The results of the finalized report are consistent with those of the preliminary report. ENO concludes that no further actions are required to assure compliance with 10 CFR 50 Appendix H.

There are no new commitments made in this letter. If you have any questions, please contact Ms.

Charlene Faison at 914-272-3378.

r/

resident Entergy Nuclea perations, Inc.

Attachment:

I. WCAP-16251-NP, Analysis of Capsule X from Entergy's Indian Point Unit 3 Reactor Vessel Radiation Surveillance Program.

cc: Mr. Patrick D. Milano, Senior Project Manager Project Directorate I, Division of Licensing Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop O-8-C2 Washington, DC 20555 Mr. Samuel J. Collins Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident Inspector's Office Indian Point Unit 3 U.S. Nuclear Regulatory Commission P.O. Box 337 Buchanan, NY 10511-0337 Mr. Peter R. Smith President NYSERDA 17 Columbia Circle Albany, NY 12203 Mr. Paul Eddy New York State Dept. of Public Service 3 Empire Plaza Albany, NY 12223 2

ATTACHMENT 1 WCAP-1 6251-NP ANALYSIS OF CAPSULE X FROM ENTERGY'S INDIAN POINT UNIT 3 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM Entergy Nuclear Operations, Inc.

Indian Point Nuclear Generating Unit No. 3 Docket No. 50-286

Westinghouse Non-Proprietary Class 3 WCAP-16251-NP July 2004 Revision 0 Analysis of Capsule X from Entergy's Indian Point Unit 3 Reactor Vessel Radiation Surveillance Program S@ Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16251-NP, Revision 0 Analysis of Capsule X from Entergy's Indian Point Unit 3 Reactor Vessel Radiation Surveillance Program TJ. Laubham

1. Conermann S.L. Anderson July 2004 Approved: Q anAd ,

Gherurich Manager Reactor Component Design & Analysis Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355 402004 Westinghouse Electric Company LLC ADl Rights Reserved

Mi TABLE OF CONTENTS LIST OF TABLES ............ iv LIST OF FIGURES ............. vi PREFACE . .......... vii EXECUTIVE

SUMMARY

................. ix I

SUMMARY

OF RESULTS .1-2 INTRODUCTION .2-1 3 BACKGROUND .3-1 4 DESCRIPTION OF PROGRAM .......................................... 4-1 5 TESTING OF SPECIMENS FROM CAPSULE X .. 5-1 5.1 OVERVIEW .5-1 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS .5-3 5.3 TENSILE TEST RESULTS .5-5 5.4 WEDGE OPENING LOADING SPECIMEN TESTS .5-5 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY .. 6-1

6.1 INTRODUCTION

. &1 6.2 DISCRETE ORDINATES ANALYSIS .6-2 6.3 NEUTRON DOSIMETRY .6-5 6A CALCULATIONAL UNCERTAINTIES.........................................................................6-6 7 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE .7-8 REFERENCES.................I.....................I....................................................................................8-I APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS CREDIBILITY ............................ A-0 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS ............................. B-O APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD. C-0 APPENDIX D INDIAN POINT UNIT 3 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION ............. D-

iv LIST OF TABLES Table 4-1 Chemical Composition (wt %) of the Indian Point Unit 3 Reactor Vessel Surveillance Materials (Unirradiated) .4......................... -3 Table 4-2 Heat Treatment History of the Indian Point Unit 3 Reactor Vessel Surveillance Materials .... 4-4 Table 5-1 Charpy V-Notch Data for the Indian Point Unit 3 Lower Shell Plate R2803-3 Irradiated to a Fluence of 0.874 x 109 n/cm 2 (E > 1.0 MeV)

(Longitudinal Orientation)...............................................................................................5-6 Table 5-2 Charpy V-Notch Data for the Indian Point Unit 3 Lower Shell Plate R2803-3 Irradiated to a Fluence of 0.874 x 10'9 n/cm 2 (1 > 1.0 MeV)

(Transverse Orientation) .... 5-7 Table 5-3 Charpy V-notch Data for the Indian Point Unit 3 Surveillance Weld Material Irradiated to a Fluence of 0.874 x 10'9 n/cm2 (En1.0 MeV) ........................................... 5-8 Table 5-4 Charpy V-notch Data for the Indian Point Unit 3 Intermediate Shell Plate R2802-2 Irradiated to a Fluence of 0.874 x 10' 9 n/cm2 (E> 1.0 MeV)

(Longitudinal Orientation) .5-9 Table 5-5 Instrumented Charpy Impact Test Results for the Indian Point Unit 3 Lower Shell Plate R2803-3 Irradiated to a Fluence of 0.874 x 1019 n/cm 2 (E> 1.0 MeV)

(Longitudinal Orientation) .5-10 Table 5-6 Instrumented Charpy Impact Test Results for the Indian Point Unit 3 Lower Shell Plate R2803-3 Irradiated to a Fluence of 0.874 x 10' 9 n/cm2 (E> 1.0 MeV)

(Transverse Orientation) .5-11 Table 5-7 Instrumented Charpy Impact Test Results for the Indian Point Unit 3 Surveillance Weld Metal Irradiated to a Fluence of 0.874 x 1019 n/cm2 (E> 1.0 MeV) .5-12 Table 5-8 Instrumented Charpy Impact Test Results for the Indian Point Unit 3 Intermediate Shell Plate B2802-2 Irradiated to a Fluence of 0.874 x 10'9 n/cm2 (E> 1.0MeV)

(Longitudinal Orientation) .5-13 Table 5-9 Effect of Irradiation to 0.874 x 10'9 n/cm 2 (E> 1.0 MeV) on the Capsule "X" Notch Toughness Properties of the Indian Point Unit 3 Reactor Vessel Surveillance Materials .5-14 Table 5-10 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 .5-15

v LIST OF TABLES (Cont.)

Table 5-11 Tensile Properties of the Indian Point Unit 3 Capsule X Reactor Vessel Surveillance Materials Irradiated to 0.874 x 1019 n/cm2 (E> 1.0MeV) ......................... 5-16 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures At The Surveillance Capsule Center.................. 6-10 Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface ........................................ 6-14 Table 6-3 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV) Within The Reactor Vessel Wall ............ 6-18 Table 6-4 Relative Radial Distribution of Iron Atom Displacements (dpa) Within The Reactor Vessel Wall .61............. 618 Table 6-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Indian Point Unit 3.6-19 Table 6-6 Calculated Surveillance Capsule Lead Factors .6-19 Table 7-1 Recommended Surveillance Capsule Withdrawal Schedule .7-1

Ai LIST OF FIGURES Figure 4-1 Arrangement of Surveillance Capsules in the Indian Point Unit 3 Reactor Vessel ......... 4-5 Figure 4-2 Capsule X Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeters ...... I. . 4-6 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate R2803-3 (Longitudinal Orientation) . 5-17 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate R2803-3 (Longitudinal Orientation) ....................... 5-18 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate R2803-3 (Longitudinal Orientation) ...................................... 5-19 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate R2803-3 (Transverse Orientation) . 5-20 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate R2803-3 (Transverse Orientation) . 5-21 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Transverse Orientation) . 5-22 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for Indian Point Unit 3 Reactor Vessel Weld Metal . 5-23 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for Indian Point Unit 3 Reactor Vessel Weld Metal . 5-24 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for Indian Point Unit 3 Reactor Vessel Weld Metal ............................................................ 5-25 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2 (Longitudinal Orientation) ............................ 5-26 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2 (Longitudinal Orientation) . 5-27 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2 (Longitudinal Orientation) . 5-28 Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Indian Point Unit 3 Reactor Vessel Lower Shell Plate R2804-3 (Longitudinal Orientation) . 5-29

vii LIST OF FIGURES (Cont.)

Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Indian Point Unit 3 Reactor Vessel Lower Shell Plate R2803-3 (Transverse Orientation) ........................................ 5-30 Figure 5-15 Charpy Impact Specimen Fracture Surfaces for Indian Point Unit 3 Reactor Vessel Weld Metal .......................................................... 5-31 Figure 5-16 Charpy Impact Specimen Fracture Surfaces for Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2(Longitudinal Orientation) ....................................... 5-32 Figure 5-17 Tensile Properties for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Longitudinal Orientation) .......................................................... 5-33 Figure 5-18 Tensile Properties for Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2 (Longitudinal Orientation) .......................................................... 5-34 Figure 5-19 Fractured Tensile Specimens from Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Longitudinal Orientation) .......................................................... 5-35 Figure 5-20 Fractured Tensile Specimen from Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2 (Longitudinal Orientation) .......................................................... 5-36 Figure 5-21 Engineering Stress-Strain Curves for Lower Shell Plate B2803-3 Tensile Specimens A5 and A6 (Longitudinal Orientation) ........................................................ 5-37 Figure 5-22 Engineering Stress-Strain Curve for Intermediate Shell Plate B2802-2 Tensile Specimen NI (Longitudinal Orientation) .......................................................... 5-38 Figure 6-1 Indian Point Unit 3 rO Reactor Geometry at the Core Midplane .................................... 6-8 Figure 6-2 Indian Point Unit 3 rz Reactor Geometry .......................................................... 6-11

viii PREFACE This report has been technically reviewed and verified by:

Reviewer:

Sections 1 through 5, 7, 8, Appendices B, C and D C.M. Burton 49x -

Section 6 and Appendix A G K. Roberts

ix EXECUTIVE

SUMMARY

The purpose of this report is to document the results of the testing of surveillance Capsule X from Indian Point Unit 3. Capsule X was removed at 15.5 EFPY and post irradiation mechanical tests of the Charpy V-notch and tensile specimens were performed. 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/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 lol0 n/cm2 .

This evaluation lead to the following conclusions: 1) The measured 30 ft-lb shift in transition temperature values of the lower shell plate B2803-3 contained in capsule X (longitudinal & transverse) are greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift values are less than the two sigma allowance by Regulatory Guide 1.99, Revision 2. 2) The measured 30 ft-lb shift in transition temperature value of the weld metal contained in capsule X is less than the Regulatory Guide 1.99, Revision 2, prediction. 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, prediction. However, the shift value is less than the two sigma allowance by Regulatory Guide 1.99, Revision 2. 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 I OCFR50, Appendix G rn. 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 Appendix D.

Lastly, a brief summary of the Charpy V-notch testing can be found in Section 1. All Charpy V-notch data was plotted using a symmetric hyperbolic tangent curve fitting program.

1-1 1

SUMMARY

OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule X, the fourth capsule removed and tested from the Indian Point Unit 3 reactor pressure vessel, led to the following conclusions:

• The Charpy V-notch data presented in WCAP-8475131 , WCAP-9491 t 41 , WCAP-10300151 , and WCAP-1 1815(6] 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. Appendix C presents the CVGRAPH, Version 5.02, Charpy V-notch plots and the program input data.

• Capsule X received an average fast neutron fluence (E> 1.0 MeV) of 0.874 x 10i9 n/cm2 after 15.5 effective full power years (EFPY) of plant operation.

• 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 1910.6F 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 0 F for the longitudinal oriented specimens. See Table 5-9.

• 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 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 ft-lb transition temperature increase of 217.9 0 F for the longitudinal oriented specimens. See Table 5-9.

• Irradiation of the weld metal (heat number W5214) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 128.50 F and an irradiated 50 ft-lb transition temperature of 196.80 F. This results in a 30 ft-lb transition temperature increase of 193.2 0 F and a 50 ft-lb transition temperature increase of 242.80 F. See Table 5-9.

• 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 145.0 0 F. This results in a 30 ft-lb transition temperature increase of 152.6 0 F and a 50 ft-lb transition temperature increase of 166.50 F for the longitudinal oriented specimens. See Table 5-9.

• 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 5-9.

Summary of Results

1-2

• The average upper shelf energy of the lower shell plate B2803-3 (transverse orientation) resulted in an average energy decrease of 16 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 52 ft-lb for the longitudinal oriented specimens. See Table 5-9.

• 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 5-9.

• The average upper shelf energy of the intermnediate 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 5-9.

• A comparison, as presented in Table 5-10, of the Indian Point Unit 3 reactor vessel surveillance material test results with the Regulatory Guide 1.99, Revision 21'1 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) are greater than the Regulatory Guide 1.99, Revision 2, predictions. However, each shift value is less than the 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 less than the Regulatory Guide 1.99, Revision 2, prediction.

- 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, prediction. However, the shift value is less than the 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 life of the vessel (27.1 EFPY) as required by IOCFR5O, Appendix G PI.

Summary of Results

1-3

• The calculated end-of-license (27.1 EFPY) neutron fluence (E> 1.0 MeV) at the core midplane for the Indian Point Unit 3 reactor vessel using the Regulatory Guide 1.99, Revision 2 attenuation formula (i.e., Equation #3 in the guide) are as follows:

Calculated: Vessel inner radius* = 9.22 x IO"S n/cm2 Vessel 1/4 thickness = 5.50 x 10"8 n/cm2 Vessel 3/4 thickness = 1.95 x 1038 n/cm2

• Clad/base metal interface. (From Table 6-2)

Summary of Results

2-1 2 INTRODUCTION This report presents the results of the examination of Capsule X, the fourth capsule removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Indian Point Unit 3 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Indian Point Unit 3 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the pre-irradiation mechanical properties of the reactor vessel materials are presented in WCAP-8475, "Consolidated Edison Co. of New York Indian Point Unit No. 3 Reactor Vessel Radiation Surveillance Program"'3 . The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM El 85-62, "Recommended Practice for Surveillance Tests on Structural Materials for Nuclear Reactors." Capsule X was removed from the reactor after 15.5 EFPY of exposure and shipped to the Westinghouse Science and Technology Department Hot Cell Facility, where the post-irradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.

This report summarizes the testing of and the post-irradiation data obtained from surveillance capsule X removed from the Indian Point Unit 3 reactor vessel and discusses the analysis of the data.

Introduction

3-1 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The behtline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA302 Grade B Modified (base material of the Indian Point Unit 3 reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness during high-energy irradiation.

A method for ensuring the integrity of reactor pressure vessels has been presented in "Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section Xl of the ASME Boiler and Pressure Vessel Code 181. The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RTNDT).

RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDfT per ASTM E-208171) or the temperature 60 0 F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented perpendicular (transverse) to the major working direction of the plate. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (Kic curve) which appears in Appendix G to the ASME Code"'. The Klc curve is a lower bound of static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the Kk curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.

RTNT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effect%

of radiation on the reactor vessel material properties. The changes in mechanical properties of a given reactor pressure vessel steel, due to irradiation, can be monitored by a reactor vessel surveillance program, such as the Indian Point Unit 3 reactor vessel radiation surveillance program'31 , in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the initial RTNDT, along with a margin (M) to cover uncertainties, to adjust the RTNDT (ART) for radiation embrittlement. This ART (RTNDT initial + M + ARTNDT) is used to index the material to the Kic curve and, in turn, to set operating limits for the nuclear power plant that take into account the effects of irradiation on the reactor vessel materials.

Background

4-1 4 DESCRIPTION OF PROGRAM Eight surveillance capsules for monitoring the effects of neutron exposure on the Indian Point Unit 3 reactor pressure vessel core region (beltline) materials were inserted in the reactor vessel prior to initial plant start-up. The eight capsules were positioned in the reactor vessel between the thermal shield and the vessel wall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.

Capsule X was removed after 15.5 effective full power years (EFPY) of plant operation. This capsule contained Charpy V-notch, tensile, and Wedge Opening Loading (WOL) specimens as shown in Figure 4-2, which were made from intermediate shell plate B2802-2 (longitudinal). In addition this capsule contains Charpy V-notch and tensile specimens, also shown on Figure 4-2, made from lower shell plate B2803-3 (longitudinal & transverse) and submerged arc weld metal.

Test material obtained from the intermediate and lower shell plates (after thermal heat treatment and forming of the plate) were taken at least one plate thickness from the quenched edges of the plate. All test specimens were machined from the 1/4 thickness location of the plate, whereas the weld metal specimens were machined at various locations thru the weld thickness.

Charpy V-notch impact specimens from the intermediate shell plate B2802-2 and the lower shell plate B2803-3 were machined in the longitudinal orientation (longitudinal axis of the specimen parallel to the major working direction). In addition, Charpy V-notch impact specimens from the lower shell plate B2803-3 were also in the transverse orientation (longitudinal axis of the specimen perpendicular to the major working direction). The core region weld Charpy impact specimens were machined from the weldment such that the long dimension of each Charpy specimen was perpendicular to the weld direction. The notch of the weld metal Charpy specimens was machined such that the direction of crack propagation in the specimen was in the welding direction.

Tensile specimens from the intermediate shell plate B2802-2 and the lower shell plate B2803-3 were machined in both the longitudinal and transverse orientations. Tensile specimens from the weld metal were oriented with the long dimension of the specimen perpendicular to the weld direction. Capsule X only contained tensile specimens from the intermediate shell plate B2802-2 and lower shell plate B2803-3, both in the longitudinal orientation.

WOL test specimens from intermediate shell plate B2802-2 were machined in the longitudinal orientation so that the loading of the specimen would be in the longitudinal direction of the plate with the simulated crack propagation in the transverse direction. All specimens were fatigue pre-cracked according to ASTM E399-70T.

The chemical composition and heat treatment of the unirradiated surveillance materials are presented in Tables 4-1 and 4-2, respectively. The data in Table 4-1 and 4-2 was obtained from the surveillance capsule Z test report, WCAP- 11815.

Description of Program

4-2 Capsule X contained dosimeter wires of pure iron, copper, nickel, and aluminum-0. 15 weight percent cobalt (cadmium-shielded and unshielded).

The capsule contained thermal monitors made from two low-melting-point eutectic alloys and sealed in Pyrex tubes. These thermal monitors were used to define the maximum temperature attained by the test specimens during irradiation. The composition of the two eutectic alloys and their melting points are as follows:

2.5% Ag, 97.5% Pb Melting Point: 579 0 F (304'C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point: 590 0 F (310 0 C)

The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in Capsule X is shown in Figure 4-2.

Description of Program

4-3 Table 4-1 Cbemical Composition (wt%) of the Indian Point Unit 3 Reactor Vessel Surveillance Materials (Unirradiated)(_)

Intermediate Shell Plate Lower Shell Plate Element B2802-1 B2802-2 B2802-3 B2803-3 Weld Metal (b)

C 0.22 0.19 0.20 0.22 0.08 Mn 1.41 1.33 1.32 1.30 1.18 P 0.010 0.015 0.011 0.012 . 0.019 S 0.023 0.019 0.025 0.024 0.016 Si 0.28 0.21 0.26 0.28 0.17 Ni 0.50 0.53 OA9 0.52 1.02 (1.21)(c)

Cr 0.08 0.09 0.08 0.08 0.04 Mo 0.46 OA8 0.50 0.45 0.53 Cu 0.18 0.20 0.19 0.24 0.15 (0.166)(0)

Al 0.036 0.027 0.042 0.03 <0.01 V <0.01 <0.01 <0.01 <0.01 <0.01 Sn 0.014 0.017 0.014 <0.01 0.007 Cb <0.01 <0.01 <0.01 <0.01 <0.01 Zr <0.01 <0.01 <0.01 <0.01 <0.01 Ti <0.01 <0.01 <0.01 <0.01 <0.01 Notes:

(a) Data obtained from WCAP-1 181 5 and duplicated herein for completeness.

(b) Weld wire Heat Number W5214, Flux Type Linde 1092, and Flux Lot Number 3692. Surveillance weldment has the same heat and flux as the nozzle shell longitudinal weld seams 1-042A, B & C.

(c) Results of chemical analysis performed on irradiated Charpy V-notch Specimen W-1 5 from Capsule Y.

Description of Program

4-4 Table 4-2 Heat Treatment History of the Indian Point Unit 3 Reactor Vessel Surveillance Materials(a)

Material Temperature ( 0F) Time Coolant 1550 to 1650 4 hrs. Water-Quench Shell Plates 1225 4 hrs. Air-cooled 1125 to 1175 40 hrs. Furnace Cooled Weld Metal (Heat # W5214) 1125 to 1175 40 hrs. Furnace Cooled Notes:

(a) Data obtained from WCAP-1 1815 and duplicated herein for completeness.

Description of Program

4-5 REACTOR VESSEL r- THERMAL SHIELD lou- 0*

xw T

r A90*

Figure 4-1 Arrangement of Surveillance Capsules in the Indian Point Unit 3 Reactor Vessel Description of Program

4-6 LEGEND: A - LOWER SHELL PLATE B2803-3 (LONGITUDINAL)

AT - LOWER SHELL PLATE B2803-3 (TANGENTIAL)

W - WELD METAL (HEAT # W5214)

N - INTERMEDIATE SHELL PLATE B2802-2 (LONGITUDINAL)

Char 2 y _( Charpy & Tensile WOL WOL Charpy - WOL AT69 N8I Nl AT67 N6 DosimeterA) N6 N4 A40 W48 A39 I W47 A38 W46 _

TOP OF VESSEL.

• CENTER WOL Charpy Char _ WOL WOL Charpy _ Tensile AT66 I N5 AT65 I N4 AT64 I N3 AS N3 (DosimeterB) N2 Ni A37 I W45 I_ A36 IW44 I A35 W43 I A6 CENTER Dosimeter "A" Dosimeter "B" Charpy Charm Co 1: ... Co AT63 N2 AT62 I Ni Co(Cd)  ;.. *  :... : - Cu Co(Cd) Ni (Dosieter A) ...r .. i A34 I W42 I A33 W41 I .H - 579°F 590°F 0 B BOTTOM OF VESSEL Figure 4-2 Capsule X Diagram Showing The ILocation of Specimens, Thermal Monitors, and Dosimeters Description of Program

- -I N

5 TESTING OF SPECIMENS FROM CAPSULE X 5.1 OVERVIEW The post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens was performed in the Remote Metallographic Facility (RMF) at the Westinghouse Science and Technology Department. Testing was performed in accordance with IOCFR50, Appendices G and H'21 ,

ASTM Specification El 85-82r9', and Westinghouse Procedure RMF 8402110], Revision 2 as modified by Westinghouse RMF Procedures 81021111, Revision 1, and 81031121, Revision 1.

Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-8475"31 . No discrepancies were found.

Examination of the two low-melting point 5790 F (304'C) and 590'F (310 0 C) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 5790 F (3040 C).

The Charpy impact tests were performed per ASTM Specification E23-02a1131 and RMF Procedure 8103 on a Tinius-Olsen Model 74, 358] machine. The tup (striker) of the Charpy impact test machine is instrumented with a GRC 930-1 instrumentation system, feeding information into an IBM compatible computer. With this system, load-time and energy-time signals can be re'corded in addition to the standard measurement of Charpy energy (ED). From the load-time curve (Appendix B), the load of general yielding (PGy), the time to general yielding (tGy), the maximum load (PM), and the time to maximum load (tjl) can be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (PF), and the load at which fast fracture terminated is identified as the arrest load (PA).

The energy at maximum load (EM) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (Ep) is the difference between the total energy to fracture (ED) and the energy at maximum load (EM).

The yield stress (cy) was calculated from the three-point bend formula having the following expression:

0=(PGY*L)/[B*(W-a) 2 *C1J (1) where: L = distance between the specimen supports in the impact machine B = the width of the specimen measured parallel to the notch W = height of the specimen, measured perpendicularly to the notch a = notch depth The constant C is dependent on the notch flank angle [O. notch root radius (p) and the type of loading (i.e., pure bending or three-point bending). In three-point bending, for a Charpy specimen in which + =

450 and p = 0.010 inch, Equation I is valid with C = 1.21. Therefore, (forL=4W),

Testing of Specimens from Capsule X

5-2 an =(PGy*L) /I B *(W-a)2

• 1.21 I=(3.305 *PGy*W) /IB*(W- a )2 J (2)

For the Charpy specimen, B = 0.394 inch, W = 0.394 inch and a = 0.079 inch. Equation 2 then reduces to:

oi = 33-3

• PGY (3) where o, is in units of psi and Pay is in units of lbs. The flow stress was calculated from the average of the yield and maximum loads, also using the three-point bend formula.

The symbol A in columns 4, 5, and 6 of Tables 5-5 through 5-8 is the cross-section area under the notch of the Charpy specimens:

A=B *(W-a)=0.1241 sq.in. (4)

Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM Specification E23-02a[ 13 1 and A370-97a 114 1 . The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

Tensile tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specification E8-0l [151 and E21-92 (1998)1"6], and Procedure RMF 8102. All pull rods, grips, and pins were made of Inconel 718. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.

Extension measurements were made with a linear variable displacement transducer extensometer. The extensometer knife-edges were spring-loaded to the specimen and operated through specimen failure.

The extensometer gage length was 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-93117]

Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air. Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperatures. Chromel-Alumel thermocouples were positioned at the center and at each end of the gage section of a dummy specimen and in each tensile machine griper. In the test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower tensile machine griper and controller temperatures was developed over the range from room temperature to 550'F. During the actual testing, the grip temperatures were used to obtain desired specimen temperatures. Experiments have indicated that this method is accurate to +20 F.

The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.

Testing of Specimens from Capsule X

-5 . 5-3 5.2 CHARPY V-NOTCH IMPACT 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 10'9 n/cm 2 (E> 1.0 MeV) in 15.5 EFPY of operation, are presented in Tables 5-1 through 5-l l and are compared with unirradiated resultst 4 l as shown in Figures 5-1 through 5-12.

The transition temperature increases and upper shelf energy decreases for the Capsule X materials are summarized in Table 5-9 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 191.61F and an irradiated 50 ft-lb transition temperature of 223.80 F. This results in a 30 ft-lb transition temperature increase of 159.6 0 F and a 50 ft-lb transition temperature increase of 161.7 0 F for the longitudinal oriented specimens. See Table 5-9.

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 216.51F and an irradiated 50 ft-lb transition temperature of 327.4AF. This results in a 30 ft-lb transition temperature increase of 158.2 0 F and a 50 ft-lb transition temperature increase of 217.90 F for the longitudinal oriented specimens. See Table 5-9.

Irradiation of the weld metal (heat number W5214) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 128.50 F and an irradiated 50 ft-lb transition temperature of 196.80 F.

This results in a 30 ft-lb transition temperature increase of 193.21F and a 50 ft-lb transition temperature increase of 242.80 F. See Table 5-9.

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 98.1 'F and an irradiated 50 ft-lb transition temperature of 145.00 F. This results in a 30 ft-lb transition temperature increase of 152.6 0 F and a 50 ft-lb transition temperature increase of 166.5 0 F for the longitudinal oriented specimens. See Table 5-9.

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 5-9.

The average upper shelf energy of the lower shell plate B2803-3 (transverse orientation) resulted in an average energy decrease of 16 ftlb after irradiation. This results in an irradiated average upper shelf energy of 52 ft-lb for the longitudinal oriented specimens. See Table 5-9.

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 5-9.

Testing of Specimens from Capsule X

5-4 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 5-9.

A comparison, as presented in Table 5-10, of the Indian Point Unit 3 reactor vessel surveillance material test results with the Regulatory Guide 1.99, Revision 2 111 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) are greater than the Regulatory Guide 1.99, Revision 2, predictions. However, each shift value is less than the two sigma allowance by Regulatory Guide 1.99, Revision 2.

- The measured 30 fl-lb shift in transition temperature value of the weld metal contained in capsule X is less than the Regulatory Guide 1.99, Revision 2, predictions

- 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, prediction. However, the shift value is less than the 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 IOCFR5O, Appendix G 121.

The fracture appearance of each irradiated Charpy specimen from the various surveillance Capsule X materials is shown in Figures 5-13 through 5-16 and shows an increasingly ductile or tougher appearance with increasing test temperature.

The load-time records for individual instrumented Charpy specimen tests are shown in Appendix B.

The Charpy V-notch data presented in WCAP-8475 131 , WCAP-9491 141, WCAP-10300 151 , and WCAP-11815[61 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. This report also shows the composite plots that show the results from the previous capsule. Appendix C presents the CVGRAPH, Version 5.02, Charpy V-notch plots and the program input data.

Testing of Specimens from Capsule X

5-5 53 TENSILE TEST RESULTS The results of the tensile tests performed on the various materials contained in Capsule X irradiated to 0.874 x 1019 n/cm 2 (E> 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated results 131 as shown in Figures 5-17 and 5-18.

The results of the tensile tests performed on the lower Shell Plate B2803-3 (longitudinal orientation) indicated that irradiation to 0.874 x 10'9 n/cm2 (E> 1.0 MeV) caused approximately a 21 to 23 ksi increase in the 0.2 percent offset yield strength and approximately a 18 to 23 ksi increase in the ultimate tensile strength when compared to unirradiated datal 31 . See Figure 5-17.

The results of the tensile tests performed on the intermediate Shell Plate B2802-2 (longitudinal orientation) indicated that irradiation to 0.874 x 10'9 n/cm 2 (E> 1.0 MeV) caused approximately a 19 ksi increase in the 0.2 percent offset yield strength and approximately a 17 ksi increase in the ultimate tensile strength when compared to unirradiated data131. See Figure 5-18.

The fractured tensile specimens for the lower shell plate B2803-3 and intermediate shell plate B2802-2 material are shown in Figures 5-19 and 5-20. The engineering stress-strain curves for the tensile tests are shown in Figures 5-21 and 5-22.

5.4 WEDGE OPENING LOADING SPECIMEN TESTS Per the surveillance capsule testing contract, the Wedge Opening Loading Specimens were not tested and are being stored at the Westinghouse Science and Technology Center Hot Cell facility.

Testing of Specimens from Capsule X

5-6 Table 5-1 Charpy V-notch Data for the Indian Point Unit 3 Lower Shell Plate B2803-3 Irradiated to a Fluence of 0.874 x 1019 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 82 111 59 1.50 100 A35 350 177 78 106 57 1.45 100 A38 375 191 83 113 68 1.73 100 Testing of Specimens from Capsule X

5-7 Table 5-2 Charpy V-notch Data for the Indian Point Unit 3 Lower Shell Plate B2803-3 Irradiated to a Fluence of 0.874 x 10"ln/cm 2 (E> 1.0 MeV) (Transverse Orientation)

Sample Temperature Impact Energy Lateral Expansion Shear Number OF°C ftlbs - 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 Testing of Specimens from Capsule X

5-8 Table 5-3 Charpy V-notch Data for the Indian Point Unit 3 Surveillance Weld Metal Irradiated to a Fluence of 0.874 x 1019 n/cm2 (E> 1.0 MeV)

Sample Temperature Impact Energy Lateral Expansion Shear Number OFc 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 Testing of Specimens from Capsule X

5-9 Table 5-4 Charpy V-notch Data for the Indian Point Unit 3 Intermediate Shell Plate B2802-2 Irradiated to a Fluence of 0.874 x 10i9 n/cm2 (E> 1.0 MeV) (Longitudinal Orientation)

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

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 Testing of Specimens from Capsule X

5-10 Table 5-5 Instrumented Charpy Impact Test Results for the Indian Point Unit 3 Lower Shell Plate B2803-3 Irradiated to a Fluence of 0.874 x 1019 n/cm 2 (E>1.0 MeV) (Longitudinal Orientation)

Charpy Normalized Energies Yield Time to Time to Fast Test Energy (ft-lb/in 2 ) Load Yield Max. Max. Fract. Arrest Yield Flow Sample Temp. ED Charpy Max. Prop. PGY tcv Load tM Load PF Load PA Stress Stress No. (OF) (ft-lb) ED/A EM/A Ep/A (lb) (msec) PM (lb) (msec) (lb) (lb) ay (ksi) (ksi)

A37 100 7 56 19 37 2108 0.11 2304 0.13 2304 363 70 73 A34 150 21 169 68 101 3363 0.14 4187 0.22 4047 372 112 126 A36 175 22 177 68 110 3336 0.14 4173 0.22 4090 609 111 125 A33 200 27 218 65 152 3311 0.14 4061 0.22 3913 1082 110 123 A40 225 51 411 226 185 3331 0.14 4567 0.50 4529 2496 111 132 A39 280 82 661 236 425 3336 0.14 4671 0.51 n/a n/a III 133 A35 350 78 628 225 403 3176 0.14 4480 0.51 n/a n/a 106 127 A38 375 83 669 222 447 3165 0.14 4344 0.51 n/a n/a 105 125 Testing of Specimens from Capsule X

5-ll Table 5-6 Instrumented Charpy Impact Test Results for the Indian Point Unit 3 Lower Shell Plate B2803-3 Irradiated to a Fluence of 0.874 x 10" n/cm2 (E>1.0 MeV) (Transverse Orientation)

Charpy Normalized Energies Test Energy (ft-lb/in 2 ) _ Load Yield Max. Max. Fract. Arrest Yield Flow Sample Temp. ED Charpy Max. Prop. PcY t"y Load tM Load P 1 Load PA Stress Stress No. (OF) (ft-lb) ED/A Em/A E,/A (lb) (msec) PM (lb) (msec) (lb) (lb) cy (ksi) (i)

AT64 100 6 48 14 34 1416 0.09 1672 0.12 1659 455 47 51 AT69 175 20 161 68 93 3310 0.14 4149 0.22 4090 683 110 124 AT68 210 22 177 67 110 3407 0.15 4091 0.22 3927 987 113 125 AT67 225 33 266 66 200 3380 0.14 4113 0.21 4017 2273 113 125 AT66 250 44 355 162 193 3091 0.14 4136 0.41 3999 2397 103 120 AT65 325 47 379 143 236 3089 0.13 4027 0.37 3853 1829 103 118 AT62 375 54 435 162 274 2969 0.13 4063 0.41 n/a n/a 99 117 AT63 390 55 443 148 295 3028 0.13 4080 0.38 n/a n/a 101 118 Testing of Specimens from Capsule X

5-12 Table 5-7 Instrumented Charpy Impact Test Results 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)

Charpy Normalized Energies Yield Time to Time to Fast Test Energy (ft-lb/in 2) Load Yield Max. Max. Fract. Arrest Yield Flow Sample Temp. ED Charpy Max. Prop. PGY tGy Load tM Load PF Load Stress Stress No. (OF) (ft-lb) E0 /A EM/A E/A (lb) (msec) PM (lb) (msec) (lb) PA (lb) cry (ksi) (ksi)

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 111 124 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 111 127 W46 350 75 604 221 383 3285 0.14 4309 0.51 n/a n/a 109 126 Testing of Specimens from Capsule X

5-13 Table 5-8 Instrumented Charpy Impact Test Results for the Indian Point Unit 3 Intermediate Shell Plate B2802-2 Irradiated to a Fluence of 0.874 x 1019 n/cm2 (E>1.0 MeV) (Longitudinal Orientation)

Charpy Normalized EnergieS Yield Time to Time to Fast Test Energy Load Yield Max. Max. Fract. Arrest Yield Flow Sample Temp. ED Charpy Max. Prop. PGY tCY Load tM Load PF Load Stress cy Stress NO. (OF) (f-lb) EWDA EM/A EIYA (lb) (msec) PM (lb) (mseC) (tb) PA (lb) (ksl) (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 125 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 NI 250 104 838 311 527 3193 0.15 4446 0.68 n/a n/a 106 127 N8 300 105 846 312 534 3272 0.14 4494 0.67 - na n/a 109 129 N3 325 105 846 302 544 3068 0.14 4374 0.67 n/a nta 102 124 n.. ..----. . -

Testing of Specimens from Capsule X

5-14 Table 5-9 Effect of Irradiation to 0.874 x 10" n/cm (E>1.0 MeV) on the Capsule "X" Notch Toughness Properties of the Indian Point Unit 3 2

Reactor Vessel Surveillance Materials(c)

Average 30 (ft-lb)(a) Average 35 mil Lateral(b) Average 50 ft-lb(a) Average Energy Absorption(')

Material Transition Temperature (OF) Expansion Temperature (OF) Transition Temperature (OF) at Full Shear (ft-lb)

Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AE Lower Shell Piate 32.0 191.6 159.6 48.6 225.0 176.4 62.1 223.8 161.7 105 81 -24 B2803-3 (Long.)

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

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

Inter. Shell Plate -54.5 98.1 152.6 -38.6 147.3 185.9 -21.5 145.0 166.5 125 105 -20 B2802-2 (Long.)

a. "Average" is defined as the value read from the curve fit through the data points of the Charpy tests (see Figures 5-1, 5-4, 5-7 and 5-10).

b. "Average" is defined as the value read from the curve fit through the data points of the Charpy tests (see Figures 5-2, 5-5, 5-8 and 5-1 1).

Testing of Specimens from Capsule X

5-15 Table 5-10 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 Fluenceze Predicted Measured Predicted Measured (x 10" n/Cm2, ( 0F) (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 0.263 131.3 151.6 22 30 Program Y 0.692 185.0 172.0 27 43 Weld Metal Z 1.04 208.3 229.2 31 37 X 0.874 198.4 193.2 29 38 Intermediate Shell Plate B2802-2 X 0.874 146.2 152.6 30 16 (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 5.02 (See Appendix C)

(c) Values are based on the definition of upper shelf energy given in ASTM E185-82.

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

Testing of Specimens from Capsule X

5-16 Table 5-11 Tensile Properties of the Indian Point Unit 3 Capsule X Reactor Vessel Surveillance Materials Irradiated to 0.874 x 1019 n/cm2 (E > 1.0 MeV)

Material Sample Test 0.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Number Temp. Strength Strength Load Stress Strength Elongation Elongation in Area (OF) (ksi) (ksl) (kip) (ksi) (ksi) (%) (%)

Lower Shell Plate A5 250 81.7 101.0 3.68 147.8 74.9 11.3 21.4 49 B2803-3 (Long.) A6 550 75.4 100.0 3.90 149.7 79.5 10.5 18.6 47 Inter. Shell Plate N 225 73.3 90.2 2.98 165.7 60.6 11.3 22.4 63 B2802-2 (Long.) N 2 Testing of Specimens from Capsule X

5-17 LOWER SHELL PLATE B2803-3 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04AY2/2004 04:43 PM Data Set(s) Plotted Curve Plant Capsule Material Or. Heat #

1 Indian Point 3 UNIRR SA302B LT A-0512-2 2 Indian Point 3 T SA302B LT A-0512-2 3 Indian Point 3 z SA302B LT A-OS12-2 4 Indian Point 3 x SA302B LT A-0512-2 300 -

250 200-X150.

ISO _

6 100 50 0

-300 -2010 -100 0 100 200 300 400 500 600 Temperature In Deg F o Set I a Set 2

• Set 3 - a Set 4 Results Curve Fluence LSE USE d-USE T 030 d-T @30 T S0 d-T 050 2.2 105.0 .0 32.0 .0 62. 1 .0 2 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 81.0 -24.0 191.6 159.6 223. 8 161.7 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-18 LOWER SHELL PLATE B2803-3 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:57 PM Data Set(s) Plotted Curve Plant Capsule Material On. Heat #

1 Indian Point 3 UNIRR SA302B LT A-0512-2 2 Indian 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-0512-2 200 150 aa I

.19 e

I a 100 e

50 0 I-

-300 0 300 600 Temperature In Deg F 0 Set1 a Set2 0 Set3 A Set 4 Results Curve Fluence LSE USE d-USE T @35 d-T @35

.0 78.7 .0 48. 6 .0 2 .0 65.8 - 12.9 185.4 136. 8 3 .0 73.4 -5.2 218. 3 169.7 4 .0 65.1 - 13.5 225.0 176.4 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-19 LOWER SHELL PLATE B2803-3 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve PRinted on 04/02/2004 04:53 PM Data Set(s) Plotted Curve Plant Capsule Material Orl. Heat #

1 Indian Point 3 UNIRR SA302B LT A-0512-2 2 Indian 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-0512-2 125 _

100 _

75 50 _

o -

-300 -20 O -100 0 100 200 300 400 500 600

- Temperature In Deg F o Set 1 a Set2 0 Set3 A Set4 Results Curve Fluence LSE USE d-USE T @50 d-T 050 1  : .0 100.0 .0 68.8 - .0 2 .0 100.0 .0 204.7 135.9 3 .0 100.0 .0 205.3 136.5 4 .o 100.0 ' .0 206. 3 137.5 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-20)

LOWER SHELL PLATE B2803-3 (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:10 PM Data Set(s) Plotted Curve Plant Capsule Material On. Heat #

1 Indian Point 3 UNIRR SA302B TL A-0512-2 2 Indian Point 3 T SA302B TL A-0512-2 3 Indian Point 3 By SA302B TL A-0512-2 4 Indian Point 3 z SA302B A-0512-2 5 Indian Point 3 x SA302B TL A-0512-2 300 -

250-

,8 200-IL 21150 -

w 100 50 0

-300 -20( -100 0 100 200 300 400 500 600 Temperature in Deg F o Set1 a Set 2 0 Set 3 ^ Set 4 Set 5 Results Curve Fluence LSE USE d-USE T @30 d-T @30 T @50 d-T @50 2.2 68.0 .0 58. 3 .0 109.5 .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 4 2. 2 56. 0 - 12.0 216. 2 157. 9 266. 7 157.2 5 2. 2 52.0 -16.0 216. 5 158.2 327.4 217. 9 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-21 LOWER SHELL PLATE 2803-3 (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:36 PM Data Set(s) Plotted Curve Plant Capsule Material Oi. Heat #

2 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 TL A-0512-2 4 Indian Point 3 z SA302B TL A-0512-2 5 Indian Point 3 X SA302B IL A-0512-2 200 150

~2 EL 100 I 50 o _-

-300 0 300 600 Temperature In Deg F 0 Set 1 a Set 2

• Set 3 a Set 4 Set 5 Results Curve Fluence LSE USE d-USE T @35 d.T @35 I .0 63.1 .0 75.3 .0 2 .0 76.0 12.9 209.2 133.9 3 .0 53.3 -9.8 254.3 179.0 4 .0 53.9 -9.2 232.8 157. 5 5 .0 44.4 -18.7 269.3 194.0 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-22 LOWER SHELL PLATE B2803-3 (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02J2004 05:16 PM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat #

Indian Point 3 UNIRR SA302B TL A-0512-2 2 Indian Point 3 T SA302B TL A-0512-2 3 Indian Point 3 By SA302B TL A-0512-2 4 Indian Point 3 z SA302B TL A-0512-2 5 Indian Point 3 x SA302B TL A-0512-2 125-100 X 75 -

co S9 a

I2 50 _

a.

25-0-

-300 -20CI -100 0 100 200 300 400 500 600 Temperature In Deg F 0 SetI o Set 2 0 Set 3 A Set 4 Set 5 Results Curve Fluence LSE USE d-USE T @50 d-T @50

.0 100. 0 .0 92.2 .0 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 5 .0 100.0 .0 217. 1 124. 9 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-23 SURVEILLANCE WELD MATERIL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04102/2004 02:31 PM Data Set(s) Plotted Curve Plant Capsule Material On. Heat #

I Indian Point 3 UNIRR SAW NA W5214 2 Indian Point 3 T SAW NA W5214 3 Indian Point 3 SAW NA W5214 4 Indian Point 3 z SAW NA W5214 5 Indian Point 3 x SAW NA W5214 300 _

250 _

2200 -

150 _

w 100 0

-300 -201)0 -100 0 100 200 300 400 500 600 Temperature In Deg F 0 Set I a Set 2 0 Set 3 a Set 4 Set 5 Results Curve }luence LSE USE d-USE T030 d-T 030 T @50 d-T@50 1 2.2 120.0 .0 -64.7 .0 -46.0 .0 2 2.2 84. 0 -36.0 86.9 151.6 130. 6 176.6 3 2.2 69.0 -51.0 107. 3 172.0 164.2 210.2 4 2.2 76.0 -44.0 164.5 229:2 218.7 264. 7 5 2.2 74.0 -46.0 128.5 193.2 196. 8 242. 8 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for Indian Point Unit 3 Reactor Vessel Weld Metal Testing of Specimens from Capsule X

5-24 SURVEILLANCE WELD MATERIAL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:11 PM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat #

1 Indian Point 3 UNIRR SAW NA W5214 2 Indian Point 3 T SAW NA W5214 3 Indian Point 3 By SAW NA W5214 4 Indian Point 3 z SAW NA W5214 5 Indian Point 3 x SAW NA W5214 200 150 r-t aL 100 0

50 0 4-

-300 0 300 600 Temperature in Deg F 0 Set I a Set2 0 Set 3 b Set 4 Set 5 Results Curve Fluence LSE USE d-USE T @35 d-T @35 l .0 90. 8 .0 -59. 3 .0 2 .0 80. 3 - 10.5 113.3 172. 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 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for Indian Point Unit 3 Reactor Vessel Weld Metal Testing of Specimens from Capsule X

5-25 SURVEILLANCE WELD MATERIAL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 03:57 PM Data Set(s) Plotted Curve Plant Capsule Material Oi. 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 Indian Point 3 Z SAW NA W5214 5 Indian Point 3 X SAW NA W5214 125 -

100 _

75 _

gI g) a.5 50 -

25 -

0

-300 -201D -100 0 100 200 - 300 400 500 600 Temperature In Deg F 0 Set 1 D Set 2 0 Set 3 I A Set 4 Set 5 Results Curve Fluence LSE USE d-USE T @50 d.T @50

.0 100.0 .0 -47.8 .0 2 .0 100.0 .0 124.0 171.8 3 .0 100.0 .0 132.6 180.4 4 .0 100.0 .0 147. 5 195.3 5 .0 100.0 .0 144.5 192.3 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for Indian Point Unit 3 Reactor Vessel Weld Metal Testing of Specimens from Capsule X

5-26 INTERMEDIATE SHELL PLATE B2802-2 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:42 PM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat #

1 Indian Point 3 UNIRR SA302B LT A-0516-2 2 Indian Point 3 x SA302B LT A-0516-2 300 250 A 200 0

0 E 150 UJ z

B 100 50 0 -2Vis

-300 -2!00 -100 0 100 200 300 400 500 600 Temperature In Deg F 0 Seti a Set2 Results Curve Fluence LSE USE d-USE T @30 d-T @30 T @50 d-T @50 2.2 125.0 .0 - 54.5 .0 -21.5 .0 2 2.2 105. 0 -20.0 98. 1 152. 6 145.0 166.5 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2 (Longitudinal)

Testing of Specimens from Capsule X

5-27 INTERMEDIATE SHELL PLATE B2802-2 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:48 PM Data Set(s) Plotted Curve Plant Capsule Material 0. Heat #

1 Indian Point 3 UNIRR SA302B LT A-0516-2 2 Indian Point 3 X SA302B LT A-0516-2 200 -_

150 -_

C E2 EL 100 -

50-0

-300 0 300 600 Temperature In Deg F o SetI o Set 2 Results Curve nuence LSE USE d-USE T @35 d-T @35

.0 79.8 .0 -38.6 .0 2 .0 75.2 -4.6 147.3 185.9 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2 (Longitudinal)

Testing of Specimens from Capsule X

5-28 INTERMEDIATE SHELL PLATE B2802-2 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:45 PM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat #

I Indian Point 3 UlNIRR SA302B LT A-0516-2 2 Indian Point 3 x SA302B LT A-0516-2 125 100 A. 75 4-R

a. 50 2: -

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F 0 Seti1 Set2 Results Curve Fluence LSE USE d-USE T @50 d-T @50

.0 100.0 .0 22. 6 .0 2 .0 100.0 .0 153.2 130. 6 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2 (Longitudinal)

Testing of Specimens from Capsule X

5-29 A37, 100F A34,150W A36,175 F A33, 200$F A40, 225WF A39, 28OF A35, 350F A38, 375WF Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-30 AT64, 100 0 F AT69, 175 0 F AT68, 2100 F AT67, 225TF AT66, 250TF AT65, 32501F AT62,375TF AT63, 3900 F Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-31 W42,750 F W41,1250 F W43, 1250 F W48, 150'F W47, 2000 F W44, 250"F W45, 300"F W46, 3500 F Figure 5-15 Charpy Impact Specimen Fracture Surfaces for Indian Point Unit 3 Reactor Vessel Weld Metal Testing of Specimens from Capsule X

5-32 N2, 250 F N6, 75 0 F N5, 125 0 F N7, 150 0 F N4, 2000 F NI, 250 0F N8, 300 0F N3, 325 0F Figure 5-16 Charpy Impact Specimen Fracture Surfaces for Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-33 120 ULTIMATE YIELD STRENGTH 100 *A

_80 __

E60 -

40 0.2% YIELD STRENGTH 20 0-0 100 200 300 400 500 600 TEMPERATURE F)

Legend: Aand o are Unirradiated Aand

• are Irradiated to 8.74 x 10" nlcm 2 (E > 1.0 MeV) 80 REDUCTION INAREA 760 -

50 A

,40 130 -TOTAL ELONGATION 0 ~UNIFORM UNIFORM 0 100 200 300 400 500 600 TEMPERATURE ( F)

Figure 5-17 Tensile Properties for Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-34 100 - A ULTIMATE YIELD STRENGTH 80 -

0.60 -

= 40 0.2% YIELD STRENGTH 20 0 1 0 100 200 300 400 500 600 TEMPERATURE( F)

Legend: A and o are Unirradiated hand

• are Irradiated to 8.74 x I01s n/cm 2 (E > 1.0 MeV) 80 REDUCTION INAREA 70 - A

- 60 A

-0

>- 50

• a 40 0 320 - TOTAL ELONGATION

=30 Q 20 10 A ,

UNIFORM UNIFORM 0 100 200 300 400 500 600 TEMPERATURE ( F)

Figure 5-18 Tensile Properties for Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-35 Specimen A5 Tested at 2500F Specimen A6 Tested at 550"F Figure 5-19 Fractured Tensile Specimens from Indian Point Unit 3 Reactor Vessel Lower Shell Plate B2803-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-36 Specimen NI Tested at 225 0 F Figure 5-20 Fractured Tensile Specimen from Indian Point Unit 3 Reactor Vessel Intermediate Shell Plate B2802-2 (Longitudinal Orientation)

Testing of Specimens from Capsule X

I - 5-37 INDIAN POINT UNIT # 3

• X*CAPSULE 120 100 so 40 A-S 250 F 20 0

0 0.08 0.1 0.15 02 025 0.3 STRAIN, INMIN INDIAN POINT UNIT # 3

• X*CAPSULE 120 60 -

40 A-6 I $55 F 20 0

a 0.05 0.1 0.15 02 0.25 0.3 STRAIN, INAN Figure 5-21 Engineering Stress-Strain Curves for Indian Point Unit 3 Lower Shell Plate B2803-3 Tensile Specimens A5 and A6 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-38 INDIAN POINT UNIT #3 X'CAPSULE 120 100 80 I-I 40 N-1 225 IF 20 0

0 0.05 0.1 O.15 0.2 0.25 0.3 STRAIN, INAN Figure 5-22 Engineering Stress-Strain Curve for Indian Point Unit 3 Intermediate Shell Plate B2802-2 Tensile Specimen Ni (Longitudinal Orientation)

Testing of Specimens from Capsule X

6-1 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6.1 INTRODUCMION This section describes a discrete ordinates S, transport analysis performed for the Indian Point Unit 3 reactor to determine the neutron radiation environment within the reactor pressure vessel and surveillance capsules. In this analysis, fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements (dpa) were established on a plant and fuel cycle specific basis. An evaluation of the most recent dosimetry sensor set from Capsule X, withdrawn at the end of the twelfth plant operating cycle, is provided. In addition, to provide an up-to-date data base applicable to the Indian point Unit 3 reactor, the sensor sets from the previously withdrawn capsules (T, Y,and Z) were re-analyzed using the current dosimetry evaluation methodology. These dosimetry updates are presented in Appendix A of this report. Comparisons of the results from these dosimetry evaluations with the analytical predictions served to validate the plant specific neutron transport calculations. These validated calculations subsequently formed the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 54 Effective Full Power Years (EFPY).

The use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for the development of damage trend curves as well as for the implementation of trend curve data to assess the condition of the vessel. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves and improved accuracy in the evaluation of damage gradients through the reactor vessel wall.

Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, 'Analysis and Interpretation of Light-Water Reactor Surveillance Results," recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a database for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, "Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials."

All of the calculations and dosimetry evaluations described in this section and in Appendix A were based on the latest available nuclear cross-section data derived from ENDF/B-VI and made use of the latest available calculational tools. Furthermore, the neutron transport and dosimetry evaluation methodologies follow the guidance of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."P°l Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC approved methodology described in WCAP-14040-NP-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004?')

Radiation Analysis and Neutron Dosimetry

6-2 6.2 DISCRETE ORDINATES ANALYSIS A plan view of the Indian Point Unit 3 reactor geometry at the core midplane is shown in Figure 4-1.

Eight irradiation capsules attached to the thermal shield are included in the reactor design that constitutes the reactor vessel surveillance program. The capsules are located at azimuthal angles of 40, 1760, 1840, and 3560 (40 from the core cardinal axes) and 400, 140°, 2200, and 3200 (400 from the core cardinal axes) as shown in Figure 4-1. The stainless steel specimen containers are 1-inch square and are approximately 38 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 3 feet of the 12-foot high reactor core.

From a neutronic standpoint, the surveillance capsules and associated support structures are significant.

The presence of these materials has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus between the neutron pads and the reactor vessel. In order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model.

In performing the fast neutron exposure evaluations for the Indian Point Unit 3 reactor vessel and surveillance capsules, a series of fuel cycle specific forward transport calculations were carried out using the following three-dimensional flux synthesis technique:

(r,,~(r z) = 0(r, q(rr) )

,Oz) (, 2) 1=

where 0(r,O,z) is the synthesized three-dimensional neutron flux distribution, 0(rO) is the transport solution in r,0 geometry, ¢(rz) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and ¢(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the r,O two-dimensional calculation. This synthesis procedure was carried out for each operating cycle at Indian Point Unit 3.

For the Indian Point Unit 3 transport calculations, the r,0 model depicted in Figure 6-1 was utilized since the reactor geometry is octant symmetric. This r,6 model included the core, the reactor internals, the thermal shield - including explicit representations of the surveillance capsules at 40 and 400, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. This model formed the basis for the calculated results and enabled making comparisons to the surveillance capsule dosimetry evaluations. In developing the analytical model, nominal design dimensions were employed for the various structural components. Likewise, water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full power operating conditions. The coolant densities were treated on a fuel cycle specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, et cetera. The geometric mesh description of the r,0 reactor model consisted of 170 radial by 67 azimuthal intervals. Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a point-wise basis. The point-wise inner iteration flux convergence criterion utilized in the r,0 calculations was set at a value of 0.001.

Radiation Analysis and Neutron Dosimetry

. 6-3 The rz model used for the Indian Point Unit 3 calculations is shown in Figure 6-2 and extends radially from the centerline of the reactor core out to a location interior to the primary biological shield and over an axial span from an elevation one foot below the active fuel to one foot above the active fuel. As in the case of the rO model, nominal design dimensions and full power coolant densities were employed in the calculations. In this case, the homogenous core region was treated as an equivalent cylinder with a volume equal to that of the active core zone. The stainless steel former plates located between the core baffle and core barrel regions were also explicitly included in the model. The rz geometric mesh description of the reactor model consisted of 153 radial by 90 axial intervals. As in the case of the re calculations, mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a point-wise basis. The point-wise inner iteration flux convergence criterion utilized in the r,z calculations was also set at a value of 0.001.

The one-dimensional radial model used in the synthesis procedure consisted of the same 153 radial mesh intervals included in the rz models. Thus, radial synthesis factors could be determined on a mesh-wise basis throughout the entire geometry.

The core power distributions used in the plant specific transport analysis were provided by the Nuclear Fuels Division of Westinghouse. Specifically, the data utilized included cycle dependent fuel assembly initial enrichments, bum-ups, and axial power distributions. This power distribution information was provided on a fuel cycle specific basis for the first 13 reactor operating cycles at Indian Point Unit 3.

Each of these fuel cycle designs has been implemented at the plant. Also included in this fluence evaluation are the analyses for three preliminary future cycle designs (14, 15, and 16) that were created as a part of a power uprate study. This information was used to develop spatial and energy dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the.

neutron transport calculations provided data in terms of fuel cycle averaged neutron flux, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and bum-up history of individual fuel assemblies. From these assembly dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

All of the transport calculations supporting this analysis were carried out using the DORT discrete ordinates code Version 3.lI 21 and the BUGLE-96 cross-section library."3 The BUGLE-96 library provides a 67 group coupled neutron-gamma ray cross-section data set produced specifically for light water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P5 legendre expansion and angular discretization was modeled with an S16 order of angular quadrature.

Energy and space dependent core power distributions, as well as system operating temperatures, were treated on a fuel cycle specific basis.

The results of the cycle specific transport calculations were also used to provide projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 54 Effective Full Power Years (EFPY). These projections accounted for a power uprate from 3025.0 MWt to 3067.4 MWt occurring during fuel cycle 12 followed by an additional power uprate to 3216 MWt at the onset of cycle 14.

Radiation Analysis and Neutron Dosimetry

6-4 The projections beyond the end of cycle 16 assumed continued operation at the uprated core power level of 3216 MWt with a spatial core power distribution identical to the preliminary equilibrium cycle design intended for implementation in cycle 16.

Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-6. In Table 6-1, the calculated exposure rates and integrated exposures, expressed in terms of both neutron fluence (E > 1.0 MeV) and dpa, are given at the radial and azimuthal center of the octant symmetric surveillance capsule positions, i.e., for the 40 and 400 locations. These results, representative of the axial midplane of the active core, establish the calculated exposure of the surveillance capsules withdrawn to date as well as projected into the future. Similar information is provided in Table 6-2 for the reactor vessel inner radius at four azimuthal locations. The vessel data given in Table 6-2 were taken at the clad/base metal interface, and thus, represent maximum calculated exposure levels on the vessel.

Both calculated fluence (E > 1.0 MeV) and dpa data are provided in Table 6-1 and Table 6-2. These data tabulations include both plant and fuel cycle specific calculated neutron exposures at the end of the twelfth fuel cycle as well as future projections to 17.4, 19.3, 21.2, 23, 32, 34, 48, and 54 EFPY. The calculations for Cycle 13 account for an uprate from 3025.0 MWt to 3067.4 MWt. The projections beyond 17.4 EFPY are based on an additional power uprate to 3216 MWt.

Radial gradient information applicable to fast (E > 1.0 MeV) neutron fluence and dpa are given in Tables 6-3 and 6-4, respectively. The data, based on the cumulative integrated exposures from Cycles I through 16, are presented on a relative basis for each exposure parameter at several azimuthal locations.

Exposure distributions through the vessel wall may be obtained by multiplying the calculated exposure at the vessel inner radius by the gradient data listed in Tables 6-3 and 64.

The calculated fast neutron exposures for the four surveillance capsules withdrawn from the Indian Point Unit 3 reactor are provided in Table 6-5. These assigned neutron exposure levels are based on the plant and fuel cycle specific neutron transport calculations.

Updated lead factors for the Indian Point Unit 3 surveillance capsules are provided in Table 6-6. The capsule lead factor is defined as the ratio of the calculated fluence (E > 1.0 MeV) at the geometric center of the surveillance capsule to the corresponding maximum calculated fluence at the pressure vessel clad/base metal interface. In Table 6-6, the lead factors for capsules that have been withdrawn from the reactor (T,Y, Z and X) were based on the calculated fluence values for the irradiation period corresponding to the time of withdrawal for the individual capsules. For the capsules remaining in the reactor (S, U, V and W), the lead factor corresponds to the calculated fluence values at the end of Cycle 16, the last projected fuel cycle for Indian Point Unit 3.

Radiation Analysis and Neutron Dosimetry

6-5 6.3 NEUTRON DOSIMETRY The validity of the calculated neutron exposures previously reported in Section 6.2 is demonstrated by a direct comparison against the measured sensor reaction rates and via a least squares evaluation performed for each of the capsule dosimetry sets. However, since the neutron dosimetry measurement data merely serves to validate the calculated results, only the direct comparison of measured-to-calculated results for the most recent surveillance capsule removed from service is provided in this section of the report. For completeness, the assessment of all measured dosimetry removed to date, based on both direct and least squares evaluation comparisons, is documented in Appendix A.

The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsule X, that was withdrawn from Indian Point Unit 3 at the end of the twelfth fuel cycle, is summarized below.

Reaction Rates (rps/atom) M/C Reaction Measured Calculated Ratio 63Cu(n,a)' 0 Co 1.97E-17 1.81E-17 1.09 54Fe(n,p)mMn I.44E-15 1.60E-15 0.90 58 Ni(n,p) 5 Co 2.01 E- 15 2.14E- 15 0.94 Average: 0.98

% Standard Deviation: 10.2 The measured-to-calculated (M/C) reaction rate ratios for the Capsule X threshold reactions range from 0.90 to 1.09, and the average M/C ratio is 0.98 +/- 10.2% (Iac). This direct comparison falls well within the +/- 20% criterion specified in Regulatory Guide 1.190; furthermore, it is consistent with the full set of comparisons given in Appendix A for all measured dosimetry removed to date from the Indian Point Unit 3 reactor. These comparisons validate the current analytical results described in Section 6.2; therefore, the calculations are deemed applicable for Indian Point Unit 3.

Radiation Analysis and Neutron Dosimetry

6-6 6.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Indian Point Unit 3 surveillance capsule and reactor pressure vessel is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology was carried out in the following four stages as described in Reference 2:

1 - Comparison of calculations with benchmark measurements from the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL).

2 - Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H. B. Robinson power reactor benchmark experiment.

3 - An analytical sensitivity study addressing the uncertainty components resulting from important input parameters applicable to the plant specific transport calculations used in the neutron exposure assessments.

4 - Comparisons of the plant specific calculations with all available dosimetry results from the Indian Point Unit 3 surveillance program.

The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operational or geometric variables that impact power reactor calculations. The second phase of the qualification (H. B. Robinson comparisons) addressed uncertainties in these additional areas that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations. The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant specific input parameters. The overall calculational uncertainty applicable to the Indian Point Unit 3 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Indian Point Unit 3 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule and pressure vessel neutron exposures previously described in Section 6.2. As such, the validation of the Indian Point Unit 3 analytical model based on the measured plant dosimetry is completely described in Appendix A.

The following summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference 21.

Radiation Analysis and Neutron Dosimetry

6-7 Capsule Vessel IR PCA Comparisons 3% 3%

H. B. Robinson Comparisons 3% 3%

Analytical Sensitivity Studies 10% 11%

Additional Uncertainty for Factors not Explicitly Evaluated 5% 5%

Net Calculational Uncertainty 12% 13%

The net calculational uncertainty was determined by combining the individual components in quadrature.

Therefore, the resultant uncertainty was treated as random and no systematic bias was applied to the analytical results.

The plant specific measurement comparisons described in Appendix A support these uncertainty assessments for Indian Point Unit 3.

Radiation Analysis and Neutron Dosimetry

6-8 Figure 6-1 Indian Point Unit 3 r,6 Reactor Geometry 240 -

180 E

X 1 20 6 0-0 75 150 225 300 375 R Axis (cm)

Radiation Analysis and Neutron Dosimetry

6-9 Figure 6-2 Indian Point Unit 3 rz Reactor Geometry 225 -

175-125-75-25 25 r-a-25

-1

-1 0 75 . 150 225 300 375 R Axis (cm)

Radiation Analysis and Neutron Dosimetry

6-10 Table 6-1 Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cumulative Cumulative Neutron Flux Cycle Irradiation Irradiation I -n/

S Length Time Time Cycle IEFPSJ IEFPS] LEFPY] 40 400 1 4.30E+07 4.30E+07 1.4 1.97E+10 6.13E+10 2 2.89E+07 7.19E+07 2.3 2.21E+10 7.50E+10 3 3.02E+07 1.02E+08 3.2 2.57E+10 7.03E+10 4 3.59E+07 1.38E+08 4.4 2.30E+10 5.11E+10 5 3.60E+07 1.74E+08 5.5 2.06E+10 4.62E+10 6 3.77E+07 2.12E+08 6.7 1.85E+10 4.07E+10 7 3.41E+07 2.46E+08 7.8 1.71E+10 3.31E+10 8 3.57E+07 2.82E+08 8.9 2.03E+10 3.32E+10 9 4.88E+07 3.30E+08 10.5 1.80E+10 3.OOE+10 10 5.65E+07 3.87E+08 12.3 1.34E+10 2.92E+10 11 4.66E+07 4.33E+08 13.7 1.20E+10 2.60E+10 12 5.71 E+07 4.91E+08 15.5 1.25E+10 3.011E+10 13 5.98E+07 5.50E+08 17.4 1.29E+10 2.53E+10 14 5.83E+07 6.09E+08 19.3 1.48E+10 3.26E+10 15 5.92E+07 6.68E+08 21.2 1.44E+10 3.23E+10 16 5.92E+07 7.27E+08 23.0 1.50E+10 3.27E+10 Future 2.84E+08 1.0 IE+09 32.0 1.50E+10 3.27E+10 Future 6.31 E+07 1.07E+09 34.0 1.50E+10 3.27E+10 Future 4.42E+08 1.52E+09 48.0 1.50E+10 3.27E+10 Future 1.89E+08 1.71 E+09 54.0 1.50E+10 3.27E+10 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry

6-11 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cumulative Cumulative Neutron Fluence Cycle Irradiation Irradiation jncm2 Length Time Time Cycle IEFPS1 [EFPSI IEFPY1 40 400 1 4.30E+07 4.30E+07 1.4 8.47E+17 2.63E+18 2 2.89E+07 7.19E+07 2.3 1.49E+18 4.80E+18 3 3.02E+07 1.02E+08 3.2 2.26E+18 6.92E+18 4 3.59E+07 1.38E+08 4.4 3.09E+18 8.76E+18 5 3.60E+07 1.74E+08 S 5.5 3.83E+18 1.04E+19 6 3.77E+07 2.12E+08 6.7 4.53E+18 1.20E+19 7 3.41E+07 2.46E1+08 7.8 5.1 IE+18 1.31E+19 8 3.57E+07 2.82E1+08 8.9 5.83E+18 1.431E+19 9 4.88E+07 3.30E+08 10.5 6.71E+18 1.57E+g19 10 5.65E1+07 3.87E1+08 12.3 7.47E+18 1.74E+19 11 4.66E+07 4.33E1+08 13.7 8.03E+18 1.86E+19 12 5.71E+07 4.91E+08 15.5 8.74E+18 2.03E+19 13 5.98E+07 5.SOE5+08 17A 9.51E+18 2.18E+19 14 5.83E+07 6.09E+08 19.3 1.04E+19 2.37E+19 1S 5.92E+07 6.68E1+08 21.2 1.12E+19 2.56E+19 16 5.92E+07 7.27E1+08 23.0 1.21E+19 2.76E+19 Future 2.84E+08 I.011E+09 32.0 1.64E+19 3.69E+19 Future 6.31E+07 1.07E+09 34.0 1.73E+19 3.89E+19 Future 4.42E+08 1.52E+09 48.0 2.40E+19 5.34E+19 Future 1.89E+08 1.71E+09 54.0 2.68E+19 5.96E+19 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry

6-12 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cumulative Cumulative Iron Atom Cycle Irradiation Irradiation Displacement Rate Length Time Time Idpa/sl Cycle IEFPSI IEFPSI LEFPYI 40 400 1 4.30E+07 4.30E+07 1.4 3.1SE-11 1.03E-10 2 2.89E+07 7.19E+07 2.3 3.57E-11 1.27E-10 3 3.02E+07 1.02E+08 3.2 4.15E-11 I1.19E-10 4 3.59E+07 1.38E+08 4.4 3.71E-11 8.60E-I1 5 3.60E+07 1.74E+08 5.5 3.31E-11 7.76E-I 6 3.77E+07 2.12E+08 6.7 2.98E-11 6.84E-1 1 7 3.41E+07 2.46E+08 7.8 2.75E-1 I 5.54E-1I 8 3.57E+07 2.82E+08 8.9 3.27E-1 I 5.58E-l 1 9 4.88E+07 3.30E+08 10.5 2.90E-11 5.02E- 11 10 5.65E+07 3.87E+08 12.3 2.17E-1I 4.90E-I I I1 4.66E+07 4.33E4+08 13.7 1.93E-11 4.36E-1 1 12 5.71E+07 4.91E+08 15.5 2.00E-11 5.04E-I1 13 5.98E+07 5.50E+08 17.4 2.08E-11 4.23E-1 I 14 5.83E+07 6.09E+08 19.3 2.38E-11 5.47E-I I 15 5.92E+07 6.68E+08 21.2 2.32E-11 5.41E- II 16 5.92E+07 7.27E+08 23.0 2.42E-11 5.49E- II Future 2.84E+08 1.01 E+09 32.0 2.42E-1 1 5.49E-1 I Future 6.3 1E+07 1.07E+09 34.0 2.42E-1 1 5.49E-1 I Future 4.42E+08 1.52E+09 48.0 2.42E-1 I 5.49E-1 I Future 1.89E+08 1.71 E+09 54.0 2.42E-1 1 5A9E-1 I Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry

6-13 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cumulative Cumulative Iron Atom Cycle Irradiation Irradiation Displacements Length Time Time 'rI bj Cycle [EFPS] [EFPSI IEFTYI 40 400 1 4.30E+07 4.30E+07 A1. 1.37E-03 4.45E-03 2 2.89E+07 7.19E+07 2.3 2AOE-03 8.1 IE-03 3 3.02E+07 1.02E+08 3.2 3.65E-03 1.17E-02 4 3.59E+07 1.38E+08 4.4 4.98E43 1.48E-02 5 *3.60E+07 1.74E+08 5.5 6.17E-03 1.76E-02 6 3.77E+07 2.12E+08 6.7 7.30E-03 2.02E-02 7 3.41 E+07 2A6E+08 7.8 8.23E-03 2.20E-02 S 3.57E+07 2.82E+08 8.9 9.AOE-03 2.AOE-02 9 4.88E+07 3.30E+08 10.5 1.08E-02 2.65E-02 10 5.65E+07 3.87E+08 12.3 1.20E-02 2.93E-02 11 4.66E+07 4.33E+08 13.7 1.29E-02 3.13E-02 12 5.71E+07 4.91E+08 15.5 1.41E-02 3A2E-02 13 5.98E+07 5.50E+08 17A 1.53E-02 3.67E-02 14 5.83E+07 6.09E+08 19.3 1.67E-02 3.99E-02 15 5.92E+07 6.68E+08 21.2 1.81E-02 4.31E-02 16 5.92E+07 7.27E+08 23.0 1.95E-02 4.63E-02 Future 2.84E+08 1.O1E+09 32.0 2.64E-02 6.19E-02 Future 6.31 E+07 1.07E+09 34.0 2.79E-02 6.54E-02 Future 4.42E+08 1.52E+09 48.0 3.86E-02 8.96E-02 Future 1.89E+08 1.71E+09 54.0 4.32E-02 1.OOE-O1 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry

6-14 Table 6-2 Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Neutron Flux (E > 1.0 MeV)

Cycle Irradiation Irradiation In/ n-sl Length Time Time Cycle IEFPS] IEFPSJ (EFPYI 00 15° 300 450 1 4.30E+07 4.30E+07 1.4 6.01E+09 9.56E+09 1.20E+I0 1.79E+10 2 2.89E+07 7.19E+07 2.3 7.37E+09 1.17E+10 1.A9E+10 2.39E+10 3 3.02E+07 1.02E+08 3.2 7.71E+09 1.21E+10 1.41 E+10 1.99E+10 4 3.59E1+07 1.38E+08 4.4 6.97E+09 1.02E+10 1.09E+10 1.47E+10 5 3.60E+07 1.74E+08 5.5 6.17E+09 8.93E+09 9.49E+09 1.33E+10 6 3.77E+07 2.12E+08 6.7 5.56E+09 8.34E+09 8.97E1+09 1.18E+10 7 3.411E+07 2.46E+08 7.8 5.09E+09 8.78E+09 8.71 E+09 9.48E+09 8 3.57E+07 2.82E+08 8.9 6.02E+09 9.50E+09 8.08E+09 9.59E+09 9 4.88E+07 3.30E+08 10.5 5.47E+09 7.39E+09 7.22E+09 8.74E+09 10 5.65E+07 3.87E+08 12.3 4.27E+09 6.37E+09 7.05E+09 8.87E+09 11 4.66E+07 4.33E+08 13.7 3.68E+09 5.78E+09 6.78E+09 7.60E+09 12 5.71 E+07 4.91E+08 15.5 3.81E+09 5.67E+09 6.96E+09 8.85E+09 13 5.98E1+07 5.50E+08 17.4 3.96E+09 6.09E+09 6.66E+09 7.35E+09 14 5.83E+07 6.09E+08 19.3 4.55E+09 6.84E+09 7.62E+09 9.69E+09 15 5.92E+07 6.68E+08 21.2 4.45E+09 6.87E+09 7.78E+09 9.63E+09 16 5.92E+07 7.27E+08 23.0 4.65E+09 7.16E+09 7.96E+09 9.78E+09 Future 2.84E+08 1.011E+09 32.0 4.65E+09 7.16E+09 7.96E+09 9.78E+09 Future 6.31 E+07 1.07E+09 34.0 4.65E+09 7.16E+09 7.96E+09 9.78E+09 Future 4.42E+08 1.52E+09 48.0 4.65E+09 7.16E+09 7.96E+09 9.78E+09 Future 1.89E+08 1.71 E+09 54.0 4.65E+09 7.16E+09 7.96E+09 9.78E+09 Radiation Analysis and Neutron Dosimetry

6-15 Table 6-2 cont'd Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Neutron Fluence > 1.0 Mev)

Cycle Irradiation Irradiation _______

Length Time Time Cycle IEFPS - IEFPS] IEFPYJ 00 150 300 45° 1 4.30E+07 4.30E+07 1.4 2.58E+17 4.1 1E+17 5.15E+17 7.69E+17 2 2.89E+07 7.19E+07 2.3 4A7E+17 7.12E+17 8.97E+17 1.38E+18 3 - 3.02E+07 1.02E+08 3.2 -6.80E+17 1.08E+18 1.32E+18 1.98E1+18

.4 3.59E+07 1.38E+08 4.4 9.30E+i7 1.44E+18 1.71E+18 2.51E+18 5 3.60E+07 1.74E+08 5.5 1.15E1+18 1.76E+18 2.06E+18 2.99E+18 6 3.77E+07 2.12E+08 6.7 1.36E+18 2.08E1+18 2.39E+18 3.44E+18 7 3AIE+07 2.46E+08 7.8 1.54E+18 2.38E+18 2.69E+18 3.76E+18 8 3.57E+07 2.82E+08 8.9 1.75E+18 2.72E+18 ' 2.98E+18 4.10E+18 9 4.88E+07 3.30E+08 10.5 2.02E+18 3.08E+18 3.33E+18 4.53E+18

'10 5.65E+07 3.87E+08: 12.3 2.25E+18 3A2E+18 3.71E+18 5.OIE+18 II 4.66E+07 4.33E+08 13.7 2.42E1+1 8 3.69E+18 4.03E+18 5.36E1+18 12 5.71E+07 4.91E+08 15.5 2.64E+18 4.01E+18 4.42E+1' 5.86E+1 8 13 5.98E+07 5.50E+08 17.4 2.87E1+18 4.38E+18 4.82E+18 6.30E+18 14 5.83E+07 6.09E+08 19.3 3.13E+18 4.77E+18 5.26E+18 6.86E+18 15 5.92E+07 6.68E+08 21.2 3.39E+18 5.17E+18 5.71E+18 7.42E+18 16 5.92E+07 7.27E+08 23.0 3.66E+18 5.58E+18 6.17E+18 798E+18 Future 2.84E+08 Il.OIE+09 32.0 4.95E+18 7.57E+18 8.38E+18 1.07E+19 Future 6.311E+07 1.07E+09 34.0 5.24E+18 8.01E+18 8.87E+18 1.13E+19 Future ' 4.42E+08 1.52E+09 48.0 7.27E+18 1.1 IE+19 1.24E+19 1.56E+19 Future 1.89E+08 1.71E+09 54.0 '8.15E+18 1252E+19 1.39E+-19 1.74E+ 19 Radiation Analysis and Neutron Dosimetry

6-16 Table 6-2 cont'd Calculated Azimuthal Variation Of Fast Neutron Exposure Rates And Iron Atom Displacement Rates At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacement Rate Cycle Irradiation Irradiation Idl _as_

Length Time Time Cycle _EFPSI _ EFPSJ _ EFPY _ 00 150 300 450 I 4.30E+07 4.30E+07 1.4 9.74E-12 1.53E-11 1.93E-11 2.89E-11 2 2.89E+07 7.19E+07 2.3 1.19E-11 1.88E-11 2AOE-11 3.85E-1 1 3 3.02E+07 1.02E+08 3.2 1.25E-l1 1.94E-11 2.27E-11 3.22E-11 4 3.59E+07 1.38E+08 4.4 1.13E-11 1.63E-11 1.76E-11 2.38E-1 1 5 3.60E+07 1.74E+08 5.5 9.98E-12 1.43E-11 1.53E-11 2.15E-1 1 6 3.77E+07 2.12E+08 6.7 9.00E-12 1.33E-11 1.45E-1I 1.90E-1 1 7 3.41E+07 2.46E+08 7.8 8.26E-12 1.40E-11 1.40E-11 1.53E-1 I 8 3.57E+07 2.82E+08 8.9 9.76E-12 1.52E-11 1.30E-11 1.55E-1 1 9 4.88E+07 3.30E+08 10.5 8.84E-12 1.18E-11 1.16E-11 1.41E-11 10 5.65E+07 3.87E+08 12.3 6.92E-12 1.02E-11 1.14E-11 1.43E-1 I 11 4.66E+07 4.33E+08 13.7 5.96E-12 9.25E-12 1.09E-11 1.23E-1 1 12 5.71EE+07 4.91E+08 15.5 6.17E-12 9.OSE-12 1.12E-11 1.43E-1 I 13 5.98E+07 5.50E+08 17.4 6.41E-12 9.74E-12 1.07E-11 1.19E-11 14 5.83E+07 6.09E+08 19.3 7.37E-12 1.09E-11 1.23E-11 1.56E-l l 15 5.92E+07 6.68E+08 21.2 7.21E-12 1.lOE-11 1.25E-11 1.55E-1 1 16 5.92E+07 7.27E+08 23.0 7.53E-12 1.15E-11 1.28E-11 1.58E-1 I Future 2.84E+08 l.OIE+09 32.0 7.53E-12 1.15E-11 1.28E-11 1.58E-1 I Future 6.31E+07 1.07E+09 34.0 7.53E-12 1.15E-11 1.28E-11 1.58E-l l Future 4.42E+08 1.52E+09 48.0 7.53E-12 1.15E-11 1.28E-11 1.58E-1l Future 1.89E+08 I 1.71E+09 54.0 7.53E-12 1.15E-11 1.28E-11 1.58E-11 Radiation Analysis and Neutron Dosimetry

6-17 Table 6-2 cont'd Calculated Azimuthal Variation OfMaximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacements Cycle Irradiation Irradiation _ _ pal Length Time Time cle EFPS] IEFPS tEFJ Y) 00 150 300 450 4.30E+07 4.30E+07 1.4 4.19E-04 6.59E-04 8.31 E-04 1.24E-03 2 2.89E+07 7.19E+07 2.3 7.25E-04 1.14E-03 1.45E-03 2.24E-03 3 3.02E+07 1.02E+08 3.2 I.ioE-03 1.73E-03 2.13E-03 3.21E-03 4 3.59E+07 -1.38E+08 4.4 l.51 E-03 2.311E-03 2.76E-03 4.06E-03 5 3.60E+07 i.74E+08 5.5 1.87E-03 2.82E-03 3.31E-03 4.84E-03 6 3.77E+07 2.12E+08 6.7 2.21E-03 3.33E-03 3.86E-03 5.55E-03 7 3.41E+07 2.46E+08 7.8 2.49E-03 3.8lE-03 4.34E-03 6.07E-03 8 3.57E+07 2.82E+08 8.9 2.84E-03 4.35E-03 4.80E-03 6.63E-03 9 4.88E+07 3.30E+08 10.5 3.27E-03 4.92E-03 5.37E-03 7.32E-03 10 5.65E+07 3.87E+08 12.3 3.64E-03 5.47E-03 5.98E-03 8.09E-03 11 4.66E+07 4.33Ei+08 13.7 3.92E-03 S.90E-03 6.49E-03 8.66E-03 12 5.71E+07 4.91E+08 15.5 4.27E-03 6.42E-03 7.13E-03 9.48E-03 13 5.98E+07 5.50E+08 17.4 4.65E-03 7.00E-03 7.76E-03 1.02E-02 14 5.83E1+07 6.09E+08 19.3 5.08E-03 7.63E-03 8A77E-03 1. IE-02 15 5.92E1+07 6.68E+08 21.2 SA9E-03 8.27E-03 9.19E-03 1.20E-02 16 5.92E+07 7.27E+08 23.0 5.93E-03 8.93E-03 9.93E-03 1.29E-02 Future 2.84E+08 1.O}E+09 32.0 8.02E-03 121E-02 1.35E-02 1.73E-02 Future 6.31 E+07 1.07E+09 34.0 8.49E-03 1.28E-02 1.43E-02 1.83E-02 Future 4.42E+08 1.52E+09 48.0 1.188E-02 1.78E-02 1.99E-02 2.5 1E-02 Future 1.89E+08 1.711E+09 54.0 1.32E-02 2.OOE-02 2.23E-02 2.81E-02 Radiation Analysis and Neutron Dosimetry

6-18 Table 6-3 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV)

Within The Reactor Vessel Wall RADIUS AZIMUTHAL ANGLE l (cm) 00 150 300 450 220.35 1.000 1.000 1.000 1.000 225.87 0.544 0.546 0.550 0.540 231.39 0.262 0.262 0.266 0.256 236.90 0.121 0.121 0.124 0.116 242.42 0.055 0.054 0.056 0.050 Note: Base Metal Inner Radius = 220.35 cm Base Metal 1I4T = 225.87 cm Base Metal 1/2T = 231.39 cm Base Metal 3/4T = 236.90 cm Base Metal Outer Radius = 242.42 cm Table 6-4 Relative Radial Distribution Of Iron Atom Displacements (dpa)

Within The Reactor Vessel Wall RADIUS AZIMUTHAL ANGLE (cm) 00 150 300 450 220.35 1.000 1.000 1.000 1.000 225.87 0.641 0.638 0.649 0.639 231.39 0.394 0.390 0.403 0.388 236.90 0.238 0.235 0.246 0.228 242.42 0.135 0.132 0.140 0.119 Note: Base Metal Inner Radius = 220.35 cm Base Metal 1/4T = 225.87 cm Base Metal 1/2T = 231.39 cm Base Metal 3/4T = 236.90 cm Base Metal Outer Radius = 242.42 cm Radiation Analysis and Neutron Dosimetry

6-19 Table 6-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Indian Point Unit 3 Irradiation Time Fluence (E > 1.0 MeV) Iron Displacements Capsule IEFPY] In/cm 2l JdpaI T 1.4 2.63E+18 . 4A5E-03 Y 3.2 6.92E+18 1.17E-02 Z 5.5 1.04E+19 1.76E-02 X 15.5 8.74E+18 1.41E-02 Table 6-6 1 Calculated Surveillance Capsule Lead Factors Capsule ID And Location Status Lead Factor T (400) Withdrawn EOC I 3A3 Y (400) Withdrawn EOC 3 3.49 Z (40°) Withdrawn EOC 5 3.48 X (40) Withdrawn EOC 12 1.49 S(400) In Reactor 3.46 U (40) In Reactor 1.52 V (40 ) In Reactor 1.52 W(40) In Reactor 1.52 Note: Lead factors for capsules remaining in the reactor are based on cycle specific exposure calculations through the last analyzed fuel cycle, i.e., Cycle 16.

Radiation Analysis and Neutron Dosimetry

7-1 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following surveillance capsule removal schedule meets the requirements of ASTM El 85-82 and is recommended for future capsules to be removed from the Indian Point Unit 3 reactor vessel. This recommended removal schedule is applicable to 27.1 EFPY of operation.

Table 7-1 Recommended Surveillance Capsule Withdrawal Schedule Capsule Capsule Location Lead Factor () Withdwal EFPY (b Fluence (n/cm 2 ) (a)

T 400 3.43 1.4 2.63 x 10"' (c)

Y 400 3.49 3.2 6.92 x 10' (c)

Z 400 3.48 5.5 1.04 x 101 9 (c)

S 40° 3.46 (d) (d)

X 40 1.49 15.5 8.74 x 10" (c)

V 40 1.52 EOL (e, f) (el W 40 1.52 EOL (e, f) (e fl)

U 40 1.52 EOL (ef ) (el)

Notes:

(a) Updated in Capsule X dosimetry analysis.

(b) Effective Full Power Years (EFPY) from plant startup.

(c) Plant specific evaluation.

(d) Indian Point Unit 3 tried to remove Capsule S in May of 2001; however, the Capsule was not retrievable.

Therefore, the Capsule Removal Schedule was revised to exclude Capsule S and make use of a spare capsule in its place. Due to the presence of spare capsules, the RV surveillance program is not degraded by the elimination of Capsule S (e) If Indian Point Unit 3 is following a withdrawal schedule for the standard EOL (27.1 EFPY), then it is recommended to remove the 5" & standby capsules any time after 16.1 EFPY, but not to exceed 27.1 EFPY (EOL). This would satisfy the ASTM E 185-82 requirement to withdrawal @ EOL, not less than once or greater than twice the peak EOL vessel fluence. The projected fluence on the capsules will be between 9.22 x 10" n/cm2 (One times the EOL peak vessel) and 1.844 x 1019 n/cm2 (Two times the peak EOL vessel fluence),

depending on the exact withdrawal time. The standby capsules should also be withdrawn and placed in storage.

Alternative fluence measuring techniques must be applied.

(f) If Indian Point Unit 3 is following a withdrawal schedule for License Extension (45.3 EFPY), then it is recommended to remove the 5o and standby capsules any time after 28.2 EFPY, but not to exceed 45.3 EFPY (EOLE). This would satisfy the ASTM E 185-82 requirement to withdrawal @ EOL, not less than once or greater than twice the peak EOL vessel fluence. The projected fluence on the capsules will be between 1.48 x 1019 n/cm2 (One times the EOLE peak vessel) and 2.96 x 1019 n/cm2 (Two times the peak EOLE vessel fluence),

depending on the exact withdrawal time. The standby capsules should also be withdrawn and placed in storage.

Alternative fluence measuring techniques must be applied.

Surveillance Capsule Removal Schedule

8-1 8 REFERENCES

1. Regulatory Guide 1.99, Revision 2, RadiationEmbrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May, 1988.

2. Code of Federal Regulations, IOCFR50, Appendix 4 FractureToughness Requirements, and Appendix H, Reactor Vessel MaterialSurveillance ProgramrRequirements,U.S. Nuclear Regulatory Commission, Washington, D.C.

3. WCAP-8475, ConsolidatedEdison Co. of New York Indian Point Unit No. 3 Reactor Vessel Radiation Surveillance Program,S.E. Yanichko, et. al., dated January 1975.

4. WCAP-9491, Analysis of Capsule Tfrom the Indian Point Unit No. 3 Unit I Reactor Vessel Radiation Surveillance Program,J. A. Davidson, et. al., April 1979.

5. WCAP-10300, Analysis of Capsule Yfrom the PowerAuthority of the State of New York- Indian Point Unit No. 3 Unit I Reactor Vessel Radiation Surveillance Program, S.E. Yanichko, et. al., March 1983.

6. WCAP-I 1815, Analysis of Capsule Zfrom the New York Power Authority Indian Point Unit No. 3 Unit I Reactor Vessel Radiation Surveillance Program, S.EY Yanichko, et. al., March 1988.

7. ASTM E208, StandardTest Methodfor Conducting Drop-Weight Test to Determine Nil-Ductilitv Transition Temperatureof FerriticSteels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.

8.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G.FractureToughness Criteria for ProtectionAgainst Failure

9. ASTM E1 85-82, StandardPracticeforConducting Surveillance Testsfor Light-Water Cooled Nuclear PowerReactor Vessels.

10. Procedure RMF 8402, Surveillance Capsule Testing Program, Revision 2.

11. Procedure RMF 8102, Tensile Testing, Revision 1.

12. Procedure RMF 8103, Charpy Impact Testing, Revision 1.

13. ASTM E23-02a, StandardTest Methodfor Notched Bar Impact Testing of Metallic Materials, ASTM, 2002.

14. ASTM A370-97a, StandardTest Methods and DefinitionsforMechanical Testing of Steel Products, ASTM, 1997.

15. ASTM E8-01, Standard Test Methodsfor Tension Testing of Metallic Materials, ASTM, 2001.

References

8-2

16. ASTM E21-92 (1998), Standard Test Methods for Elevated Temperature Tension Tests ofMetallic Materials,ASTM, 1998.

17. ASTM E83-93, StandardPracticefor Verification and Classificationof Extensometers, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.

18. ASTM El 85-79, StandardPracticefor Conducting Surveillance Tests for Light-Water Cooled Nuclear PowerReactor Vessels

19. WCAP-14370, Use of the Hyperbolic Tangent Functionfor FittingTransition Temperature Toughness Data, T. R. Mager, et al, May 1995.

20. Regulatory Guide RG-1.190, Calculationaland Dosimetry Methodsfor Determining Pressure Vessel Neutron Fluence, U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

21. WCAP-1 4040-NP-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.

22. RSICC Computer Code Collection CCC-650, DOORS 3.1, One, Two- and Three-Dimensional Discrete OrdinatesNeutron/Photon TransportCode System, August 1996.

23. RSICC Data Library Collection DLC-1 85, BUGLE-96, Coupled47 Neutron, 20 Gamma-Ray Group Cross Section Library Derivedfrom ENDF/B-V for LWR Shielding and PressureVessel Dosimetry Applications, March 1996.

References

A-0 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS Appendix A

A-i A.1 Neutron Dosimetry Comparisons of measured dosimetry results to both the calculated and least squares adjusted values for all surveillance capsules withdrawn from service to date at Indian Point Unit 3 are described herein. The sensor sets from these capsules have been analyzed in accordance with the current dosimetry evaluation methodology described in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.lJ One of the main purposes for presenting this material is to demonstrate that the overall measurements agree with the calculated and least squares adjusted values to within i 20% as specified by Regulatory Guide 1.190, thus serving to validate the calculated neutron exposures previously reported in Section 6.2 of this report. This information may also be useful in the future, in particular, as least squares adjustment techniques become accepted in the regulatory environment.

A.1.1 Sensor Reaction Rate Determinations In this section, the results of the evaluations of the four neutron sensor sets withdrawn to date as part of the Indian Point Unit 3 Reactor Vessel Materials Surveillance Program are presented. The capsule designation, location within the reactor, and time of withdrawal of each of these dosimetry sets were as follows:

Azimuthal Withdrawal Irradiation Capsule ID Location Time Time [EFPYJ T 400 End of Cycle 1 1.4 Y 400 End of Cycle 3 3.2 Z 40° End of Cycle 5 5.5 X 40 EndofCycle 12 15.5 The azimuthal locations included in the above tabulation represent the first octant equivalent azimuthal angle of the geometric center of the respective surveillance capsules.

The passive neutron sensors included in the evaluations of Surveillance Capsules T, Y,Z, and X are summarized as follows:

Appendix A

A-2 Reaction Sensor Material Of Interest Capsule T Capsule Y Capsule Z Capsule X Copper 53Cu(na)fCo X X X X Iron 54 Fe(n,p) 4 Mn X X X X Nickel 53 Ni(np)5 Co X X X X Uranium-238 23 U(nm137 Cs X Neptunium-237 23'Np(n,f) 137Cs X Cobalt-Aluminum* 59 Co(ny)fCo X X X X

• The cobalt-aluminum measurements for this plant include both bare wire and cadmium-covered sensors.

Pertinent physical and nuclear characteristics of the passive neutron sensors are listed in Table A-1.

The use of passive monitors such as those listed above does not yield a direct measure of the energy dependent neutron flux at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:

• the measured specific activity of each monitor,

• the physical characteristics of each monitor,

• the operating history of the reactor,

• the energy response of each monitor, and

• the neutron energy spectrum at the monitor location.

The radiometric counting of the neutron sensors from Capsules T, Y, and Z was completed by the Westinghouse Analytical Services Laboratory located at the Waltz Mill site[A- A A-41. The radiometric counting of the sensors from Capsule X was carried out by Pace Analytical Services, Inc., also located at the Westinghouse Waltz Mill Site. In all cases, the radiometric counting followed established ASTM procedures. Following sample preparation and weighing, the specific activity of each sensor was Appendix A

A-3 determined by means of a high-resolution gamma spectrometer. For the copper, iron, nickel, and cobalt-aluminum sensors, these analyses were performed by direct counting of each of the individual samples. In the case of the uranium and neptunium fission sensors, the analyses were carried out by direct counting preceded by dissolution and chemical separation of cesium from the sensor material.

The irradiation history of the reactor over the irradiation periods experienced by Capsules T, Y,Z. and X was based on the monthly power generation of Indian Point Unit 3 from initial reactor criticality through the end of the dosimetry evaluation period. For the sensor sets utilized in the surveillance capsules, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. The irradiation history applicable to Capsules T, Y, Z. and X is given in'Table'A-2.

Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor,'reaction rates referenced to full-power operation were determined from the following equation:

R=

No F Y -PfP C, [1- e 1J] [e"ld]

where:

R = Reaction rate averaged over the irradiation period and referenced to operation at a core power level of Plf (rps/nucleus).

A = Measured specific activity (dps/gm).

No = Number of target element atoms per gram of sensor.

F = Weight fraction of the target isotope in the sensor material.

Y = Number of product atoms produced per reaction.

Pi = Average core power level during irradiation period j (MWt).

Pef = Maximum or reference power level of the reactor (MWt).

C, = Calculated ratio of ¢(E > 1.0 MeV) during irradiation period j to the time weighted average

$(E > 1.0 MeV) over the entire irradiation period.

x = Decay constant of the product isotope (1/sec).

tj = Length of irradiation period j (sec).

td = Decay time following irradiation period j (sec).

and the summation is carried out over the total number of monthly intervals comprising the irradiation period.

Appendix A

A4 In the equation describing the reaction rate calculation, the ratio [Pj]/[PAj] accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio Cj, which was calculated for each fuel cycle using the transport methodology discussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in flux level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, Cj is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those employing low leakage fuel management, the additional Cj term should be employed. The impact of changing flux levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from non-low leakage to low leakage fuel management or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another.

The fuel cycle specific neutron flux values along with the computed values for C3 are listed in Table A-3.

These flux values represent the cycle dependent results at the radial and azimuthal center of the respective capsules at the axial elevation of the active fuel midplane.

Since the construction of the surveillance capsules used in the Indian Point Unit 3 reactor design places individual sensors at several radial locations within the test specimen array, gradient corrections were applied to the measured reaction rates to index all of the sensor measurements to a common geometric location within the capsule. In the case of Indian Point Unit 3, the following radii apply to the locations of the various sensors:

Sensor Type Radius (cm)

Copper 211.18 Iron From Core Side Charpy 211.18 Nickel 211.18 Uranium 238 211.41 Neptunium 237 211.41 Iron From Vessel Side Charpy 212.18 Bare Cobalt-Aluminum 212.18 Cd Covered Cobalt-Aluminum 212.18 Appendix A

A-5 Gradient correction Factors used in indexing the measured results to the geometric center of the surveillance capsules (211.41 cm) were based on the transport calculations completed for Indian Point Unit 3 and were as follows:

40° Capsule 40 Capsule Sensor Type Radius (cm) Correction Correction Copper 211.18 0.955 0.956 Iron From Core Side Charpy 211.18 0.953 0.954 Nickel 211.18 0.953 0.955 Uranium 238 211.41 1.000 1.000 Neptunium 237 211.41 1.000 1.000 Iron From Vessel Side Charpy 212.18 1.156 1.149 Bare Cobalt-Aluminum 212.18 0.974 0.950 Cd Covered Cobalt-Aluminum 212.18 1.152 1.124 Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 231Umeasurements to account for the presence of 235U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.

Corrections were also made to the 2mU and W3Np sensor reaction rates to account for gamma ray induced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the Indian Point Unit 3 fission sensor reaction rates are summarized as follows:

Correction Capsule T Capsule Y Capsule Z Capsule X 2 5U Impurity/Pu Build-in N/A 0.858 N/A N/A N/A 0.958 N/A N/A Net 238U Correction N/A 0.822 N/A N/A Np(yf) N/A 0.985 N/A N/A These factors were applied in a multiplicative fashion to the decay corrected uranium and neptunium fission sensor reaction rates. Note that the 23'U and 237Np sensors were included only in Capsule Y.

Appendix A

A-6 Results of the sensor reaction rate determinations for Capsules T, Y,Z. and X are given in Table A4. In Table A-4, the measured specific activities, decay corrected saturated specific activities, and computed reaction rates for each sensor indexed to the radial center of the capsule are listed. The fission sensor reaction rates are listed both with and without the applied corrections for 235U impurities, plutonium build-in, and gamma ray induced fission effects.

A.1.2 Least Squares Evaluation of Sensor Sets Least squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations resulting in a Best Estimate neutron energy spectrum with associated uncertainties. Best Estimates for key exposure parameters such as 4(E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least squares methods, as applied to surveillance capsule dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrum within their respective uncertainties. For example, R, +/- 8r, =F, (aig+/-i e8Q

)(Og +/-i a) g relates a set of measured reaction rates, R&to a single neutron spectrum, f through the multigroup dosimeter reaction cross-section, cri, each with an uncertainty 8. The primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the Indian Point Unit 3 application, the FERRET code[AAl was employed to combine the results of the plant specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters (+(E > 1.0 MeV) and dpa) along with associated uncertainties for the four in-vessel capsules withdrawn to date.

The application of the least squares Iethodology requires the following input:

1 - The calculated neutron energy spectrum and associated uncertainties at the measurement location.

2 - The measured reaction rates and associated uncertainty for each sensor contained in the multiple foil set.

3 - The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set.

For the Indian Point Unit 3 application, the calculated neutron spectrum was obtained from the results of plant specific neutron transport calculations described in Section 6.2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section A.1.1.

The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section libraryl(A'. The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard El 018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E 706 (IIB)."

Appendix A

A-7 The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum were input to the least squares procedure in the form of variances and covariances.

The assignment of the input uncertainties followed the guidance provided in ASTM Standard E 944, "Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance."

The following provides a summary of the uncertainties associated with the least squares evaluation of the Indian point Unit 3 surveillance capsule sensor sets.

Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, irradiation history corrections, and corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM National Consensus Standards for reaction rate determinations for each sensor type.

After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least squares evaluation:

Reaction Uncertainty

' 3Cu(n,a)f0 Co 5%

54Fe(n,p) 54Mn 5%

'Ni(n,p) Co 5%

23'U(nf)'3 7Cs 100/s 2 "Np(nf) 37 Cs  ; 10%

59Coony)6"Co 5%

These uncertainties are given at the or level.

Dosimetry Cross-Section Uncertainties The reaction rate cross-sections used in the least squares evaluations were taken from the SNLRML library. This data library provides reaction cross-sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross-sections and uncertainties are provided in a fine multigroup structure for use in least squares adjustment applications. These cross-sections were compiled from the most recent cross-section evaluations and they have been tested with respect to their accuracy and consistency for least squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as in the fluence and energy characterization of 14 MeV neutron sources.

Appendix A

A-8 For sensors included in the Indian Point Unit 3 surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.

Reaction Uncertainty 63Cu(n,a)OCo 4.084.16%

'Fe(n,p) 4 Mn 3.05-3.11%

5 Ni(n,p)5 Co 4.49-4.56%

23'U(n,f)137 Cs 0.54-0.64%

237Np(n,f) 137 Cs 10.32-10.97%

59 Co(n,y)60Co 0.79-3.59%

These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in LWR irradiations.

Calculated Neutron Spectrum The neutron spectra input to the least squares adjustment procedure were obtained directly from the results of plant specific transport calculations for each surveillance capsule irradiation period and location. The spectrum for each capsule was input in an absolute sense (rather than as simply a relative spectral shape). Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements.

While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks and the dosimetry cross-section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:

M , =R.2 +R *Rg *Pgg*

where Rl specifies an overall fractional normalization uncertainty and the fractional uncertainties RI and R8. specify additional random group-wise uncertainties that are correlated with a correlation matrix given by:

Pe = [1-_ ]8g + 0 e'"

where (g-g') 2 H- 2y 2 Appendix A

A-9 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range y (0 specifies the strength of the latter term). The value of 6 is .0 when g = g', and is 0.0 otherwise. '

The set of parameters defining the input covariance matrix for the Indian Point Unit,3 calculated spectra was as follows:

Flux Normalization Uncertainty (R.) 15%

Flux Group Uncertainties (R,, Rol)

-(E > 0.0055 MeV) 15%

(0.-68 eV < E < 0.0055 MeV) 29%

(E < 0.68 eV) 52%

Short Range Correlation (0)

(E > 0.0055 MeV) ' 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Flux Group Correlation Range (y)

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 Appendix A

A-10 A.1.3 Comparisons of Measurements and Calculations Results of the least squares evaluations of the dosimetry from the Indian Point Unit 3 surveillance capsules withdrawn to date are provided in Tables A-5 and A-6. In Table A-5, measured, calculated, and best-estimate values for sensor reaction rates are given for each capsule. Also provided in this tabulation are ratios of the measured reaction rates to both the calculated and least squares adjusted reaction rates.

These ratios of M/C and M/BE illustrate the consistency of the fit of the calculated neutron energy spectra to the measured reaction rates both before and after adjustment. In Table A-6, comparison of the calculated and best estimate values of neutron flux (E > 1.0 MeV) and iron atom displacement rate are tabulated along with the BE/C ratios observed for each of the capsules.

The data comparisons provided in Tables A-5 and A-6 show that the adjustments to the calculated spectra are relatively small and well within the assigned uncertainties for the calculated spectra, measured sensor reaction rates, and dosimetry reaction cross-sections. Further, these results indicate that the use of the least squares evaluation results in a reduction in the uncertainties associated with the exposure of the surveillance capsules. From Section 6.4 of this report, it may be noted that the uncertainty associated with the unadjusted calculation of neutron fluence (E > 1.0 MeV) and iron atom displacements at the surveillance capsule locations is specified as 12% at the I a level. From Table A-6, it is noted that the corresponding uncertainties associated with the least squares adjusted exposure parameters have been reduced to 6% - 7% for neutron flux (E > 1.0 MeV) and 7% - 9/h 0 for iron atom displacement rate. Again, the uncertainties from the least squares evaluation are at the 1a level.

Further comparisons of the measurement results with calculations are given in Tables A-7 and A-8.

These comparisons are given on two levels. In Table A-7, calculations of individual threshold sensor reaction rates are compared directly with the corresponding measurements. These threshold reaction rate comparisons provide a good evaluation of the accuracy of the fast neutron portion of the calculated energy spectra. In Table A-8, calculations of fast neutron exposure rates in terms of+(E > 1.0 MeV) and dpa/s are compared with the best estimate results obtained from the least squares evaluation of the capsule dosimetry results. These two levels of comparison yield consistent and similar results with all measurement-to-calculation comparisons falling well within the 20% limits specified as the acceptance criteria in Regulatory Guide 1.190.

In the case of the direct comparison of measured and calculated sensor reaction rates, the M/C comparisons for fast neutron reactions range from 0.90 to 1.23 for the 14 samples included in the data set. The overall average M/C ratio for the entire set of Indian Point Unit 3 data is 1.06 with an associated standard deviation of10.1%.

In the comparisons of best estimate and calculated fast neutron exposure parameters, the corresponding BE/C comparisons for the capsule data sets range from 0.91 to 1.18 for neutron flux (E > 1.0 MeV) and from 0.91 to 1.15 for iron atom displacement rate. The overall average BE/C ratios for neutron flux (E > 1.0 MeV) and iron atom displacement rate are 1.02 with a standard deviation of 11.9% and 1.01 with a standard deviation of 10.9%, respectively.

Based on these comparisons, it is concluded that the calculated fast neutron exposures provided in Section 6.2 of this report are validated for use in the assessment of the condition of the materials comprising the beltline region of the Indian Point Unit 3 reactor pressure vessel.

Appendix A

A-1l Table A-I Nuclear Parameters Used In The Evaluation OfNeutron Sensors Target 90% Response Fission Monitor Reaction of Atom Range Product Yield Material Interest Fraction (MeY) Half-life (%)

Copper 63 Cu (nc) 0.6917 4.9- 12.1 5.271 y Iron Fe (np) 0.0585 2.1-8.9 312.1 d Nickel 5 Ni (np) 0.6808 1.5 - 8.8 70.82 d Uranium-238 - 2U (nf) 1.0000 1.3 - 7.0 30.07 y 6.02 Neptunium-237 237Np (nf) 1.0000 0.4-4.4 30.07 y 6.17 Cobalt-Aluminum "Co (ny) 0.0015 non-threshold 5.271 y Note: The 90% response range is defined such that, in the neutron spectrum characteristic of the Indian Point Unit 3 surveillance capsules, approximately 90% of the sensor response is due to neutrons in the energy range specified with approximately 5% of the total response due to neutrons with energies below the lower limit and 5% of the total response due to neutrons with energies above the upper limit.

Appendix A

A-12 Table A-2 Monthly Thermnal Generation During The First Twelve Fuel Cycles Of The Indian Point Unit 3 Reactor Thermal Thermal 1Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 1976 4 7394 1979 4 1803659 1983 4 0 1976 5 570059 1979 5 2236503 1983 5 0 1976 6 1074795 1979 6 2056010 1983 6 0 1976 7 1109470 1979 7 1601186 1983 7 0 1976 8 1375290 1979 8 1618305 1983 8 0 1976 9 605492 1979 9 564515 1983 9 0 1976 10 1613060 1979 10 0 1983 10 0 1976 11 1859918 1979 11 0 1983 11 0 1976 12 1687081 1979 12 0 1983 12 0 1977 1 1158756 1980 1 0 1984 1 0 1977 2 1805507 1980 2 298849 1984 2 0 1977 3 1964160 1980 3 991653 1984 3 0 1977 4 1479229 1980 4 695552 1984 4 0 1977 5 2018472 1980 5 1824582 1984 5 0 1977 6 1935000 1980 6 1996621 1984 6 282244 1977 7 1941232 1980 7 1088451 1984 7 0 1977 8 1996896 1980 8 1999791 1984 S 0 1977 9 1897557 1980 9 1905318 1984 9 0 1977 10 371890 1980 10 0 1984 10 0 1977 11 0 1980 11 0 1984 11 0 1977 12 833079 1980 12 725986 1984 12 1978 I 1951512 1981 1 2074821 1985 1 1978 2 1161885 1981 2 0 1985 2 1542818 1978 3 1955510 1981 3 0 1985 3 2058008 1978 4 1616402 1981 4 1088240 1985 4 2144331' 1978 5 1885029 1981 5 1394908 1985 5 2041 716 1978 6 440030 1981 6 1393640 1985 6 2105400 1978 7 0 1981 7 1458736 1985 7 0 1978 8 283997 1981 8 1441828 1985 8 0 1978 9 1823475 1981 9 196767 1985 9 0 1978 10 2135478 1981 10 0 1985 10 788038 1978 11 2119504 1981 11 386680 1985 11 2046223!

1978 12 1603787 1981 12 1960672 1985 12 2002904 1979 1 2114526 1982 I 2055809 1986 I 2254965 1979 2 2021349 1982 2 1563844 1986 2 1981394 1979 3 1647683 1982 3 1440377 1986 3 I 2163376 Appendix A

A-13 Table A-2 cont'd Monthly Thermal Generation During The First Twelve Fuel Cycles Of The Indian Point Unit 3 Reactor

- Thermal Thermal . Thermal

. Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 1986 4' 1801432 1989 4 0 1992 4 1243176

.1986 5 261216 1989 5 0 1992 5 0 1986 6 1179256 1989 6 186875 1992 6 0 1986 7 181463' 1989 7 2168026 1992 7 0' 1986 8 0 1989 8 2202457 1992 8 1455404 1986 9 1469788 1989 9 2136821 1992 9 658303 1986 10 2177363 1989 10 1883145 1992 10 1138362 1986' 11 2010045 1989 11 1883145 1992 11 1740734 1986 12 .2247782 1989 12 2251639 1992 12 2249459 1987 1 2217919 '1990 1 2228855 1993 1 -1827512 1987 2 1680057 1990 2 2033379 1993 2 1857911 1987 3 1930676 '1990 3 135483 1993 3 0 1987 4 1809631 1990 4 1547737 1993 4 0 1987 5 57335 1990 5 2241934 1993 5 0 1987 6 0 :1990 6 2058672 1993 6 0 1987 7 0 1990 7 2175820 1993 7 0 1987 8 0 1990 8 2034186 1993 8 0 1987 9 1400877 1990 9 926462 1993 9 0 1987 10 2160202 1990 10 0 1993 10 0 1987 11 2143211 1990 11 0 1993 11 0 1987 12 2035559 -1990 12 189055 ' 1993 12 0 1988 1 ' 2224853 1991 1 2072170 1994 1 ' 0 1988 2 1979688 '1991 2 2060225 1994 2 0 1988 3 2194305 1991 3 '1440264 1994 3 0

-1988 4 1979091 1991 4 1425214 1994 4 0 1988 5 761416 1991 5 1247238 1994 5 0 1988 6 1759009 1991 6 2197470 1994 6 0 1988 7 2155305 1991 7 2256537 1994 7 0 1988 8 2084174 1991 8 2168594 1994 8 0 1988 9 .2092028 1991 9 2186839 1994 9 0 1988 10 831920 1991 10 1325804 1994 10 0 1988 11 521805 1991 11 1910528 1994 11 0 1988 12 2203860 1991 12 2278127 1994 12 0 1989 1 2192602 1992 1 2275140 1995 I 0 1989 2 180873 1992 2 2133566 1995 2 0 1989 3 0 1992' 3 2133566 1995 3 0 Appendix A

A-14 Table A-2 cont'd Monthly Thermal Generation During The First Twelve Fuel Cycles Of The Indian Point Unit 3 Reactor Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Mouth (MWt-hr) 1995 4 0 1998 4 2177403 2001 4 1959848 1995 5 0 1998 5 2224644 2001 5 416282 1995 6 149 1998 6 2030124 2001 6- 2156571 1995 7 1360205 1998 7 2119531 2001 7 2245664 1995 8 2219597 1998 8 1295166 2001 8 2243771 1995 9 992785 1998 9 2025197 2001 9 2216498 1995 10 0 1998 10 2251102 2001 10 2216498 1995 11 0 1998 11 1314785 2001 11 2216498 1995 12 0 1998 12 1904078 2001 12 2264345 1996 1 0 1999 1 2244323 2002 I 2267346 1996 2 0 1999 2 2030095 2002 2 2047713 1996 3 0 1999 3 1851999 2002 3 2265719 1996 4 1610268 1999 4 2172953 2002 4 2190676 1996 5 1970640 1999 5 2225331 2002 5 2269544 1996 6 2157246

'O 1999 6 2170027 2002 6 2192665 1996 7 2095611 1999 7 2198814 2002 7 2247948 1996 8 2174775 1999 8 2011341 2002 8i 2213933 1996 9 2120695 1999 9 613391 2002 9 2181936 1996 10 2002741 1999 10 594041 2002 2265746 1996 11 1955649 1999 11 2082383 2002 11 1711030 1996 12 2099613 1999 12 2268422 2002 12 2175283 1997 1 1210149 2000 1 2266690 2003 1 2177101 1997 2 631387 2000 2 2118635 2003 2 2067096 1997 3 2242233 2000 3 2265943 2003 3 2056477 1997 4 2166533 2000 4 2188810 2003 4 121323 1997 5 959250 2000 5 2264330 1997 6 0 2000 6 1877829 1997 7 0 2000 7 2256059 1997 8 0 2000 8 2250086 1997 9 686135 2000 9 2179194 1997 10 2087877 2000 10 2115678 1997 11 2177701 2000 11 2190661.

1997 12 1273755 2000 12 2056591 1998 1 2079247 2001 I 2268374 1998 2 2031050 2001 2 2050102 1998 3 2234917 2001 3 2269861 Appendix A

A-15 Table A-3 Calculated Cj Factors at the Surveillance Capsule Center Core Midplane Elevation Fuel (E> 1.0 MeV) In/cm2-s]

Cycle Capsule T Capsule Y Capsule Z Capsule X I 6.13E+10 6.13E+10 6.13E+10 1.97E+10 2 7.50E+10 7.50E+10 2.21E+10 3 7.03E+10 7.03E+10 2.57E+10 4 5.11E+10 2.30E+10 5 4.62E+10 2.06E+10 6 1.85E+10 7 1.71E+10 8 2.03E+10 9 1.80E+10 10 1.34E+10 11 1.20E+10 12 1.25E+10 Average 6.13E+10 6.78E+10 5.99E+10 1.78E+10 Appendix A

A-16 Fuel CJ Cycle Capsule T Capsule Y Capsule Z Capsule X 1 1.00 0.90 1.02 1.11 2 1.11 1.25 1.24 3 1.04 1.17 1.44 4 0.85 1.29 5 0.77 1.16 6 1.04 7 0.96 8 1.14 9 1.01 10 0.75 11 0.67 12 0.70 Average 1.00 1.00 1.00 1.00 Appendix A

A-17 Table A-4 Measured Sensor Activities And Reaction Rates Surveillance Capsule T Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate Reaction Location (dps/g) (dpslg) (rpslatom) 63Cu (n,a) "OCo Top 4.87E+04 3.25E+05 3.10E+05 4.74E-17 Bottom 4.59E+04 - 3.06E+05 2.93E+05 4.46E-17 Average 4.60E-17 5Fe (np) 'Min W-40 1.37E+06 3.50E+06 3.34E+06 5.29E-15 W-37 1.25E+06 3.19E+06 3.04E+06 4.83E-15 B-9 1.34E+06 3.42E+06 3.26E+06 5.17E-15 A-32 1.17E+06 2.99E+06 3.46E+06 5.48E-15 AT-58 1.09E+06 2.79E+06 3.22E+06 5.1OE- 15 AT-54 1.18E+06 3.02E+06 3.49E+06 5.53E-15 Average 5.23E-15

'Ni (n,p) 58 Co l Middle 8.IIE+06 l 5.00E+07 1 4.77E+07 1 6.82E-15 Averake 6.82E-15 59 Co (ny) 60Co Top 7.63E+06 5.09E+07 4.96E+07 3.24E-12 Middle 6.84E+06 4.57E+07 4.45E+07 2.90E-12 Bottom 7.68E+06 5.13E+07 4.99E+07 3.26E-12 Average _ 3.13E-12 59 Co (ny) 60Co (Cd) Top 3.08E+06 2.06E+07 2.37E+07 1.55E-12 Middle 3.03E+06 2.02E+07 2.33E+07 1.52E-12 Bottom 3.02E+06 2.02E+07 2.32E+07 1.52E-12 Average 1.53E-12 Notes:

1) Measured specific activities are indexed to a counting date of November 7, 1978.

2) The location of the iron sensors corresponds to individual Charpy specimens from which the iron samples were extracted.

Appendix A

A-18 Table A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule Y Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate Reaction Location (dpslg) (dps/g) (dps/g) (rpslatom) 63Cu (na) 60Co Top 8.58E+04 3.09E+05 2.95E+05 4.50E-17 Bottom 8.93E+04 3.21E+05 3.07E+05 4.68E-17 Average 4.59E-17 54Fe (n,p) 54Mn W-16 1.05E+06 3.13E+06 2.98E+06 4.73E-15 W-12 1.04E+06 3.10E+06 2.96E+06 4.68E-15 W-9 1.06E+06 3.16E+06 3.01E+06 4.77E-15 AT-37 8.98E+05 2.68E+06 3.10E+06 4.91E-15 AT-33 8.70E+05 2.59E+06 3.OOE+06 4.75E-15 AT-30 8.14E+05 2.43E+06 2.81E+06 4.45E-15 Average 4.72E-15 5 Ni (n,p) 5 Co Middle 4.05E+06 4.64E+07 4.42E+07 6.33E-15 Average 6.33E-15 2U (n,f) ' 37Cs (Cd) Middle 3.32E+05 4.71E+06 4.71E+06 3.10E-14 23'U (nf) 137Cs (Cd) Including 235U, 239Pu, and yfission corrections: 2.54E-14 237Np (n,f) 137Cs (Cd) Middle 2.34E+06 3.32E+07 3.32E+07 2.12E-13 237Np (n f) 137CS (Cd) Including yfission correction: 2.09E-13 59 Co (ny) 60Co Top 1.52E+07 5.47E+07 5.33E+07 3.48E-12 Middle 1.42E+07 5.1lE+07 4.98E+07 3.25E-12 Bottom 1.52E+07 5.47E+07 5.33E+07 3.48E-12 Average 3.40E-12 59 Co (n,y) WCo (Cd) Top 5.56E+06 2.OOE+07 2.3 1E+07 1.50E-12 Middle 5.36E+06 1.93E+07 2.22E+07 1.45E-12 Bottom 5.79E+06 2.08E+07 2.40E+07 1.57E-12 Average 1.51E-12 Notes:

1) Measured specific activities are indexed to a counting date of October 20, 1982.

2) The average 238U (nf) reaction rate of 2.54E-14 includes a correction factor of 0.858 to account for 23U impurities

& plutonium build-in & an additional factor of 0.958 to account for photo-fission effects in the sensor.

3) The average 237Np (n,f) reaction rate of 2.09E-13 includes a correction factor of 0.985 to account for photo-fission effects in the sensor.

4) The location of the iron sensors corresponds to individual Charpy specimens from which the iron samples were extracted.

Appendix A

A-19 Table A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule Z Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate Reaction Location (dps/g) (dps/g) (dps/g) (rps/atom) 63Cu (na)60Co Top l.O1E+05 2.93E+05 2.80E+05 4.27E-17 Bottom 9.92E+04 2.88E+05 2.75E+05 4.20E-17 Average _ _  : 4123E-17 54 Fe (n,p) 54Mn W-64 9.92E+05 2.85E+06 2.71E+06 4.30-15 W-61 9.93E+05 2.85E+06 2.72E+06 4.311E-15 TI 9.66E+05 2.77E+06 2.64E+06 4.19E-15 AT-82 7.94E+05 2.28E+06 2.64E+06 4.18E-15 AT-78 8.25E+05 2.37E+06 2.74E+06 4.34E-15 A-56 7.65E+05 2.20E+06 2.54E+06 4.02E-15 Average 4.22E-15 55 Ni (np) 53Co Middle 4.03E+06 3.83E+07 3.65E+07 J 5.22E-15

_ Average _ Jl 5.22E-15 59 Co (ny) 6 Co Top 1.47E+07 4.27E+07 4.16E+07 2.71E-12 Middle 1.36E+07 3.95E+07 3.85E+07 2.511E-12 Bottom 1.49E+07 4.33E+07 4.21E+07 2.75E-12 Average _ 2.66E-12 59 Co'(n y) 60Co (Cd) Top 5.90E+06 1.71E+07 1.97E+07 1.29E-12 Middle 5.82E+06 1.69E+07 1.95E+07 1.27E-12 Bottom 5.92E+06 1.72E+07 1.98E+07 1.29E-12 Average - 1.28E-12 Notes:

I) Measured specific activities are indexed to a counting date of November 1, 1987.

2) The location of the iron sensors corresponds to individual Charpy specimens from which the iron samples were extracted.

Appendix A

A-20 Table A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule X Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate Reaction Location (dps/g) (dps/g) (dpslg) (rps/atom) 63CU (n,a) 60Co Top 6.94E+04 1.38E+05 1.32E+05 2.01E-17 Bottom 6.65E+04 1.32E+05 1.26E+05 1.92E-17 Average 1.97E-17

'Fe (np) mMn AT-67 2.78E+05 8.08E+05 9.28E+05 1.47E-15 AT-64 2.56E+05 7.44E+05 8.55E+05 1.36E-15 AT-69 2.83E+05 8.22E+05 9.45E+05 1.50E-1 5 Average 1.44E-15 5 8Ni (np) Co Middle 5.3 1E+05 1.47E+07 l 1.40E+07 2.01E-15 59 Co (n,) 60CO Top 4.09E+06 J 8.12E+06 7.71E+06 5.03E-13 Middle 4.42E+06 8.77E+06 8.33E+06 5.44E-13 Bottom 4.68E+06 9.29E+06 8.82E+06 5.76E-13 Average 5.41E-13 5 9Co (n y) 6OCo (Cd) Top 1.67E+06 3.31E+06 3.72E+06 2.43E-13 Middle 1.71E+06 3.39E+06 3.81E+06 2.49E-13 Bottom 1.75E+06 3.47E+06 3.90E+06 2.55E-13 Average _ 2.49E-13

1) Measured specific activities are indexed to a counting date of January 23,2004.

2) The location of the iron sensors corresponds to individual charpy specimens from which the iron samples were extracted.

Appendix A

A-21 Table A-5 Comparison of Measured, Calculated, and Best Estimate Reaction Rates At The Surveillance Capsule Center Capsule T

_ _l Reaction Rate IrpsIatoml Best Reaction Measured Calculated Estimate M/C MIBE 3

- Cu(n,a)OOCo 4.60E-17 3.80E-17 4.58E-17 1.21 1.00 54Fe(n,p) 54Mn 5.23E-15 4.25E-15 5.12E-15 1.23 1.02

-"Ni(np) 5 sCo 6.82E-15 5.87E-15 6.95E-15 1.16 0.98 23U(nfI) 7Cs (Cd) 237Np(nt)137Cs (Cd) - --

59Co(ny)fCo 3.13E-12 2.48E-12 3.11E-12 1.26 1.01 S"Co(n,y)WCo (Cd) 1.53E-12 1.29E-12 I.53E-12 1.19 1.00 Cazpsule Y

__ _ __ __ __ _Reacto Rate I /at__ ___

Best Reaction Measured Calculated Estimate M/C MIBE 63 Cu(n,af)Wo 4.59E-1 7 4.16E-17 4.46E-1 7 1.10 1.03 "4Fe(n,p)54Mn 4.72E-15 4.69E-15 4.SOE-15 1.01 0.98

• 58Ni(npf)58Co 6.33E-15 6.47E-1 5 6.56E-1 5 0.98 0.96 B'U(n,t)' Cs (Cd) 2.54E-14 2.34E-14 2.41E-14 1.09 1.05.

3

• N~~) 3 Cs (Cd) 2.09E-13 1.85E-13 2.00E-13 1.13 1.05 S9CoQnT)ECCo 3.40E-12 2.75E-12 3.37E-12 1.24 1.01 "Co(nT)'Co (Cd) 1.51E-12 I1.44E-12 I1.52E-12 1.05: '0.9 Appendix A

A-22 Table A-5 Comparison of Measured, Calculated, and Best Estimate Reaction Rates At The Surveillance Capsule Center Capsule Z Y

Redrtifl Rate Irnit/ntnmI Best Reaction Measured Calculated Estimate MYC MIBE 63Cu(na)fCo 4.23E-17 3.76E-17 4.06E-17 1.13 1.04 54 Fe(n,p)54Mn 4.22E-15 4.18E-15 4.16E-15 1.01 1.01 58 Ni(n) 53Co 5.22E-15 5.76E-15 5.54E-15 0.91 0.94 23 IU(n f)t)1377CsCs(Cd) 237Npn (Cd) 59 Co(n,y) 6 "Co 2.66E-12 2.41E-12 2.65E-12 1.10 1.00 59Co(n,7y)oCo (Cd) 1.28E-12 1.26E-12 1.28E-12 1.02 1.00 Capsule X Reaction Rate rpslatoml Best Reaction Measured Calculated Estimate MJC MIBE 63 Cu(na)6Co 1.97E-17 1.81E-17 1.88E-17 1.09 1.05 54Fe(n,p)54 Mn 1.44E-15 1.60E-15 1.50E-15 0.90 0.96 5 Ni(n,p)5 Co 2.01E-15 2.14E-15 2.03E-15 0.94 0.99 23

'U(n f01 7 Cs (Cd) 237Np(nf 317 Cs (Cd) 59Co(n,7)6Co 5.41E-13 5.49E-13 5.40E-13 0.99 1.00 59Co(n,y)fCo (Cd) 2.49E-13 2.79E-13 2.50E-13 0.89 1.00 Appendix A

A-23 Table A-6 Comparison of Calculated and Best Estimate Exposure Rates

- At The Surveillance Capsule Center

• (E> 1.0 MeV) InIcm2 _sl Best Uncertainty Capsule ID Calculated Estimate (1C) BE/C T 6.13E+10 7.21E+10 7% 1.18 Y 6.78E+10 7.03E+10 6% 1.04 Z 5.99E+10 5.63E+10 7% 0.94 X 1.78E+10 1.62E+10 7% 0.91 Note: Calculated results are based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period.

Iron Atom Displacment Rate Walk-Best Uncertainty Capsule ID Calculated Estimate icr) BEIC T 1.03E-10 1.19E-10 9o/e 1.15 Y u.14E-10 1.19E-10 7% 1.03 Z L.OIE-10 9.46E-I1 90/ 0.94 X 2.87E-1 2.60E-11 8% 0.91 Note: Calculated results are based on the synthesized transport calculations taken at the core

,midplane following the completion of each respective capsules irradiation period.

Appendix A

A-24 Table A-7 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions WC Ratio Reaction Capsule T Capsule Y Capsule Z Capsule X 63 Cu(n,a)OCo 1.21 1.10 1.13 1.09 5'4Fe(n,p)mMn 1.23 1.01 1.01 0.90 55 Ni(n,p) 5 Co 1.16 0.98 0.91 0.94 238u(np)1 37 cs (Cd) 1.09 237 Np(nf)137 Cs (Cd) 1.13 Average 1.20 1.06 1.01 0.98

% Standard Deviation 2.9 6.1 10.8 10.2 Note: The overall average MWC ratio for the set of 14 sensor measurements is 1.06 with an associated standard deviation of 10.1%.

Table A-8 Comparison of Best Estimate/Calculated (BE/C) Exposure Rate Ratios BEIC Ratio Capsule ID (E > 1.0 MeY) dpals T 1.18 1.15 Y 1.04 1.03 Z 0.94 0.94 X 0.91 0.91 Average 1.02 1.01

% Standard Deviation 11.9 10.9 Appendix A

A-25 Appendix A References A-i. Regulatory Guide RG-1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 1995.

A-2. WCAP-9491, 'Analysis of Capsule T from the Indian Point Unit No. 3 Unit 1 Reactor Vessel Radiation Surveillance Program," 1.A. Davidson, et. al., April 1979.

A-3 WCAP-I 0300, "Analysis of Capsule Y from the Power Authority of the State of New York Indian Point Unit No. 3 Unit I Reactor Vessel Radiation Surveillance Program," S.E. Yanichko, et. al., March 1983.

A-4 WCAP-11815, "Analysis of Capsule Z from the New York Power Authority Indian PointUnit No. 3 Unit I Reactor Vessel Radiation Surveillance Program," S.E. Yanichko, et. al., March 1988.

A-5 A. Schmittroth, FERRETDataAnalysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

A-6. RSIC Data Library Collection DLC-1 78, "SNLRML Recommended Dosimetry Cross-Section Compendium", July 1994.

Appendix A

B-o APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS

• Specimen prefix "A" denotes Lower Shell Plate B2803-3, Longitudinal Orientation

• Specimen prefix "Ar' denotes Lower Shell Plate B2803-3. Transverse Orientation

• Specimen prefix "W" denotes Weld Material

• Specimen prefix "N" denotes Intermediate Shell Plate B2802-2, Long. Orientation

• Load (1) is in units oflbs

• Time (1) is in units of milli seconds Appendix B

B-1 5M.00.

4000.00 3000.00 2000.00-1000.00 I 0.00 0.00 1.00 2D. 3.00 4.00 5.00 Sim TimeA1(0)

A37, 100°F

-J 0.00 1.00 2.00 3.00 4.00 5.00 6.O0 Time41 (ms)

A34, 1500 F Appendix B

B-2 5000.00 4000.00 2000.00.{

looom.

0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-i (ms)

A36, 1750 F 5M0000 4000.00 300 0 .W 2000.00

• 1000.00 X 0.00 .

0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (b)

A33, 2000 F Appendix B

B-3 5000.00 4000.00 830000Ot 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

A40, 2250 F 5000.00 4000.00 j3000.00 200.00 I 000.00 0.00 0.00 1.00 2.00 ~ 3.00 4 00 5.00 600(

tine-I (ms)

A39, 280TF Appendix B

B-4 5000.00 4000.00 3000.OO 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 S.0 6.00 Tine-1 (ins)

A35, 3500 F 5000.00 4000.00 3000.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-I (ms)

A38, 3750 F Appendix B

B-5 50o.00 4000.00 3000.00 2000.00 1000.00 nnnlI .

• 0.00 100 2.00 3.00 4.00 5.00 6.00 Tk4-l (MS)

AT64, 100°F 5000X 4000O0 I

300000-2000.00-1000.0 nrin 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

AT69, 175TF Appendix B

B-6 500.00 4000.00 X 3000.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tine-1 (ms)

AT68, 210 0 F 5000.00 4000.00 X 3000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.0 5.00 6.00 Time-1 (ms)

AT67, 225-F Appendix B

B-7 5000.00 4000.00 20.00.0 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 TOne- (Is)

AT66, 250TF 5000.00 4000.00 3.000.0 2000.00 1000.00 0.00 0.00 100 2.00 3.00 4.00 500 D60 time-1 (Ms)

AT65, 325TF Appendix B

B-8 Tibl-1 (ms)

AT62, 3750 F 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-I (rns)

AT63, 390 0 F Appendix B

B-9 S.0000 4000.00 I 3000.0 2W00000 0.00 1.00 2.00 3.00 6.00 6.00 Tine-I (ms)

W42, 75TF S60000-4000 00

• 3000.00 2000.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 r.me- (Ms)

W41, 125-F Appendix B

B-10 5D0.00.

4000.00 3000.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 400 5.00 6.00 Time-I (ms)

W43, 125-F 5000.00 4000.00 3000.00 2000.00 100w.00 I 0.00 , -.

0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-I (ms)

W48, 1500 F Appendix B

B-11t 4000.00 3000.00 2000.00-1000.0o 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Trne-1 (Ms)

W47, 200rF 5000.00 3000.l0 2000.00 1000.00 -

0.00 1 00 2.00 3.o 4.00 5.00 6.00 Tne-4 (ms)

W44, 250OF Appendix B

B- 12 5M0000 4000.00

• 3000.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

W45, 3000 F 5000.00 4000.00

• 3000.00 2000.00 1000.00 0.00 0.00 100O 2.00 3.00 4.00 5.00 6.00 Time-I (ms)

W46, 3500 F Appendix B

B-13 4000.00 3000.00 2000.00 1000.00.

0.00.

0.00 1.00 2.00 3.00 4.00 51D0 6.00 Tkie-I (me)

N2, 25 0 F 4000.100 2000.001 1000.00~

0.00 0.00 1.00 2.00 3.00 4.00 5.00 600O Time- (ims)

N6, 750F Appendix B

B-14 j 3000.1 .00 J 200.

2000.1D.0 1000.1Do.

0.1DO 0.00 1.00 2.00 3.00 4.00 5.00 6.00 5ne-, (m2)

N5, 125°F 5000.100 00 P00 43000. DO 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

N7, 1500 F Appendix B

B-15 5000.00 4000.00

- 3000.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-I (ms)

N4, 2000 F 0sooo.0 400.00 300000J\

2000.00 1000.00I , -,,> I ,

0.00 0.00 1DO 2.00 3.00 4.00 5.00 6.00 rffne-I (ns)

Nl, 250DF Appendix B

B-16 5000.

40001 30001 20001 30-1000.(

01 XI.

0.00 1.00 2.00 3.00 4.00 5.00 TNm-1 (ms)

NS, 300°F 5000.1DO-4000.1DO, a

DO 2000.1DO 1000.1DO 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-I (ms)

N3, 325PF Appendix B

c-0 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD Appendix C

C-l Contained in Table C-1 are the upper shelf energy values used as input for the generation of the Charpy V-notch plots using CVGRAPH, Version 5.0.2. The definition for Upper Shelf Energy (USE) is given in ASTM El 85-82, Section 4.18, and reads as follows:

"upper sheifenergy level - the average energy value for all Charpy specimens (normally three) whose test temperature is above the upper end of the transition region. For specimens tested in sets of three at each test temperature, the set having the highest average may be regarded as defining the upper shelf energy."

If there are specimens tested in set of three at each temperature Westinghouse reports the set having the highest average energy as the USE (usually unirradiated material). If the specimens were not tested in sets of three at each temperature Westinghouse reports the average of all 100% shear Charpy data as the USE. Hence, the USE values reported in Table C-I and used to generate the Charpy V-notch curves were determined utilizing this methodology.

The lower shelf energy values were fixed at 2.2 ft-lb for all cases.

Table C-1 Upper Shelf Energy Values Fixed In CVGRAPH [ft-lbI Material Unirradiated Capsule T Capsule Y Capsule Z Capsule X Lower Shell Plate lOS ft-lbs 92 ft-lbs 82 ft-lbs 81 ft-lbs B2803-3 (Long.)

Lower Shell Plate 68 fl-lbs 57 ft-lbs 51 ft-lbs 56 ft-lbs 52 ft-lbs B2803-3 (Trans.)

Weld Metal 120 ft-lbs 84 ft-lbs 69 ft-lbs 76 ft-lbs 74 ft-lbs (Heat # W5214)

Inter. Shell Plate 125 ft-lbs 105 ft-lbs B2802-2 (Long.)

Appendix C

UNIRRADIATED (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:39 PM Page I Coefficients of Curve 1 A = 53.6 B = 51A C = 70.68 TO = 67.02 D = O.OOE+00 Equation is A + B * [Tanh((T-To)Y(C+DT))]

Upper Shelf Energy-105.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=32.0 Deg F Temp@50 fh-lbs=62.1 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: UNIRR Pluence: n/cmA2 300 250

, 200 10 0

U-8 100 50 o 1=

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-20.00 12.50 10.28 2.22

-20. 00 13.00 10.28 2.72

10. 00 18.00 19.28 - 1.28

40. 00 28.00 34. 86 -6. 86

40. 00 38. 00 34. 86 3. 14

40. 00 37.00 34. 86 2. 14

75. 00 59.00 59.38 -. 38

75. 00 66.00 59. 38 6. 62

75. 00 53.50 59.38 -5. 88 C-2

UNIRRADIATED (LONGITUDINAL ORIENTATION)

Page 2 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: UNIRR Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 160. 00 96.00 98.09 -2. 09 160. 00 106.00 98.09 7.91 160. 00 92. 00 98. 09 -6. 09 210. 00 104. 50 103. 23 1.27 210.00 105.50 103. 23 2.27 210.00 105.00 103.23 1.77 Correlation Coefficient = .993 C-3

UNIRRADIATED (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/0212004 04:56 PM Page 1 Coefficients of Curve 1 A = 39.33 B = 39.33 C = 67.91 TO = 56.09 D = O.OOE+00 Equation is A + B * [Tanh((T-To)I(C+DT))]

Upper Shelf L.E.=78.7 Lower Shelf L.E.=.0(Fixed)

Temp.@LE. 35 mils=48.6 Deg F Plant: Indian Point 3 Material: SA302B Heat: A70512-2 Orientation: LT Capsule: UNIRR Fluence: n/cmA2 200 150 0

F 100 so 50 6

50

-300 0 300 600 Temperature In Deg F Charpy V-Notch Data Temperature Input LE. Computed LE. Differential

- 20. 00 8.00 7. 56 .44

- 20. 00 10. 00 7.56 2. 44 10.00 16.00 16. 10 - 10 40.00 24.00 30. 18 -6. 18.

40. 00 32. 00 30. 18 1.82

40. 00 31.00 30. 18 .82 75.00 50. 00 50.01 - .01 75.00 55.00 50.01 4.99 75.00 48.00 50. 01 -2. 01 C-4

UNIRRADIATED (LONGITUDINAL ORIENTATION)

Page 2 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: UNIRR Fluence: n/cmn2 Charpy V-Notch Data Temperature Input LE. Computed LE. Differential 160.00 76. 00 75. 14 . 86 160.00 75. 00 75. 14 -. 14 160.00 72.00 75. 14 -3. 14 210.00 80.00 77. 82 2. 18 210.00 80.00 77. 82 2. 18 210.00 75. 00 77. 82 -2. 82 Correlation Coefficient = .995 C-5

UNIRRADIATED (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:52 PM Page 1 Coefficients of Curve 1 A =50. B =50. C = 74.77 TO = 68.72 D= O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 68.8 Plant: Indian Point 3 Material: SA302B Heat: A-O512-2 Orientation: LT Capsule: UNIRR Fluence: n/cmn2 125 100 0

a) 75 Co 0

0 50 25 o 4--

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 20. 00 13.00 8.52 4.48

- 20. 00 13.00 8.52 4.48

10. 00 17.00 17.21 -. 21

40. 00 34.00 31.68 2. 32

40. 00 38.00 31.68 6. 32 40.00 37.00 31. 68 5. 32

75. 00 41.00 - 54. 19 - 13. 19 75.00 53.00 54. 19 .1. 19 75.00 46.00 54. 19 -8. 19 C-6

UNIRRADIATED (LONGITUDINAL ORIENTATION)

Page 2 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: UNRR Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 160. 00 1 00. 00 91.99 8.01 1 60. 00 100. 00 91.99 8.01 160. 00 1 00. 00 91.99 8.01 210.00 100.00 97.77 2. 23 210. 00 100. 00 97. 77 2. 23 210. 00 100. 00 97. 77 2. 23 Correlation Coefficient = .986 C-7

CAPSULE T (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:39 PM Page 1 Coefficients of Curve 2 A = 47.1 B = 44.9 C = 61.77 TO = 196.14 D = O.OOE+00 Equation is A + B * [Tanh((T-To)Y(C+DT))]

Upper Shelf Energy=92.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=171.4 Deg F Temp@50 ft-lbs=200.2 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: T Fluence: n/cmA2

. 1uu 250

, 200 0

U-

100 a---

I---

I I

I, 50 a -

0

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

70. 00 12.00 3.69 8.31 1 35. 00 19.50 13. 10 6.40 175.00 33.00 32. 31 .69 200.00 46.00 49. 90 3. 90 210.00 50. 00 57. 01 - 7. 01 250.00 92.00 78. 64 - 13.36 300. 00 88.00 88. 99 - . 99 400. 00 96. 00 91. 88 4. 12 Correlation Coefficient = .980 C-8

CAPSULE T (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:56 PM Page 1 Coefficients of Curve 2 A = 32.9 B = 32.9 C = 69.91 TO = 180.88 D = 0.00E+O0 Equation is A + B * [Tanh((T-To)y(C+DT))]

Upper Shelf LE.=65.8 Lower Shelf LE.=.0(Fixed)

Temp.@L.E. 35 mils=1 85.4 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: T Fluence: n/cm1n2 200 150

.9 c

C c

2 a 100 0

0 I I

-300 0 300 600 Temperature In Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

70. 00 9. 00 2.65 6. 35 135. 00 17.00 13.95 3.05 175. 00 28. 00 30. 13 -2. 13 200.00 4 l. 00 41.67 - . 67 210. 00 39.00 45. 86 -6. 86 250. 00 7 1. 00 57. 79 13.21 300. 00 63. 00 63. 68 -. 68 400. 00 60. 00 65. 67 -5.67 Correlation Coefficient = .959 C-9

CAPSULE T (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:52 PM Page 1 Coefficients of Curve 2 A = 50. B = 50. C = 48.79 TO = 204.64 D = O.OOE+00 Equation is A + B

• ITanh((T-To)Y(C+DT))]

Temperature at 50% Shear = 204.7 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: T Fluence: n/cmA2 125 100 L..

w 75 W

.C/

V) cI a- 50 25 0 4-

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

70. 00 10.00 . 40 9.60 135.00 20. 00 5.44 14. 56 175.00 25.00 22. 88 2. 12 200. 00 40. 00 45.26 -5. 26 210.00 50.00 55.47 -5.47 250.00 99. 00 86. 52 12.48 300. 00 100. 00 98. 03 1.97 400.00 100. 00 99.97 .03 Correlation Coefficient = .982 C-10

CAPSULE Z (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:39 PM Page 1 Coefficients of Curve 3 A = 42.1 B = 39.9 C = 83.48 TO = 225.87 D = O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=82.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temnp@30 ft-lbs=199.8 Deg F Temp@50 ft-lbs--242.7 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: Z Fluence: nlcmA2 300 250 T3 200 00 150 C

z 8 100t 50 0 1-

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 150.00 18. 00 13. 35 4. 65 200. 00 29.00 30. 12 - 1. 12 200. 00 31.00 30. 12 . 88 225.00 39.00 41.68 -2. 68 250. 00 52.00 53. 32 - 1. 32 325.00 81. 00 75.21 5.79 400. 00 82.00 80. 79 1.21 Correlation Coefficient = .992 C-11

- UNIRRADIATED (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:39 PM Page 1 Coefficients of Curve 1 A =53.6 B = 51.4C = 70.68 TO = 67.02 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=105.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Ternp@30 ft-lbs=32.0 Deg F Temp@50 ft-lbs=62.1 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: UNIRR Fluence: n/cmA2 300 250 n

Q 200 0

IL150 8 100 50 0 4=

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-20. 00 12.50 10.28 2.22

- 20. 00 13. 00 10.28 2.72 10.00 18.00 19.28 - 1.28

40. 00 28.00 34. 86 -6.86

40. 00 38.00 34. 86 3. 14

40. 00 37.00 34. 86 2.14 75.00 59.00 59. 38 - .38 75.00 66.00 59. 38 6.62 75.00 53.50 59. 38 -5.88 C-12

UNIRRADlATED (LONGITUDINAL ORIENTATION)

Page 2 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: UNIRR Fluence: n/cm 2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 160.00 96. 00 98.09 -2.09 160.00 106.00 98. 09 7.91 160.00 92. 00 98.09 - 6. 09 210.00 104. 50 103. 23 1.27 210. 00 105. 50 103.23 2.27 210.00 105.00 103.23 1.77 Correlation Coefficient = .993 C-1 3

UNIRRADIATED (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:56 PM Page I

- Coefficients of Curve 1 A = 39.33 B = 39.33 C = 67.91 TO = 56.09 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=78.7 Lower Shelf L.E.=.O(Fixed)

Temp.@L.E. 35 mils-48.6 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: UNIRR Fluence: n/cmA2 200 I

150 E

0 C

a. 100 50 0

I.

i

-300 0 300 600 Temperature In Deg F Charpy V-Notch Data Temperature Input L.E. Computed LE. Differential

- 20. 00 8.00 7. 56 .44

- 20. 00 10.00 7. 56 2.44 10.00 16.00 16. 10 -. 10 40.00 24.00 30. 18 - 6. 18

40. 00 32.00 30. 18 I. 82 40.00 3 1. 00 30. 18 .82 75.00 50. 00 50. 01 - .01 75.00 55.00 50.01 4.99 75.00 48.00 50.01 -2. 01 C-14

UNIRRADIATED (LONGITUDINAL ORIENTATION)

Page 2 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: UNIRR Fluence: ncMnA2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 160.00 76.00 75. 14 . 86 160. 00 75.00 75. 14 -. 14 160.00 72.00 75. 14 -3. 14 210.00 80. 00 77. 82 2. 18 210.00 80. 00 77. 82 2. 18 210.00 75. 00 77. 82 -2.82 Correlation Coefficient = .995 C-15

UNIRRADIATED (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:52 PM Page 1 Coefficients of Curve i A =50. B =50. C = 74.77 TO = 68.72 D=O.OOE+00 Equation is A + B * [Tanh((T-To)y(C+DT))J Temperature at 50% Shear = 68.8 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT. Capsule: UNIRR Fluence: n/cmA2 125 100 I 1--_---

co.

75 0~

50 0

25 0 i 1 / ___

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

-20.00 13. 00 8. 52 4. 48

- 20. 00 13.00 8.52 4.48 10.00 17.00 17.21 - .21

40. 00 34.00 31.68 2.32

40. 00 38.00 31. 68 6. 32

40. 00 37.00 31. 68 5.32

75. 00 41. 00 54. 19 - 13. 19

75. 00 53.00 54. 19 - 1. 19 75.00 46.00 54. 19 - 8. 19 C-16

UNIRRADIATED (LONGITUDINAL ORIENTATION)

Page 2 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: UMNRR Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 160. 00 100.00 91. 99 8.01 1 60. 00 100. 00 91.99 8. 01 1 60. 00 100.00 91.99 8. 01 210.00 100. 00 97. 77 2. 23 210. 00 100. 00 97.77 2.23 210.00 100.00 97. 77 2.23 Correlation Coefficient = .986 C-17

CAPSULE T (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:39 PM Page 1 Coefficients of Curve 2 A = 47.1 B = 44.9 C = 61.77 TO = 196.14 D = 0.OOE+00 i Equation is A + B * [Tanh((T-Toy(C+DT))I Upper Shelf Energy=92.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=171.4 Deg F Temp@50 ft-lbs=200.2 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Onientation: LT Capsule: T Fluence: nlcmA2 300 250 co

, 200 LI P150 w

z 100 50 0 =------4-----t---

-300 -200 -100 0 100 i 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 70.00 12. 00 3. 69 8. 31 135. 00 19. 50 13. 10 6.40 175.00 33.00 32.31 . 69 200. 00 46. 00 49. 90 3. 90 210. 00 50. 00 57. 01 -7. 01 250.00 92.00 78. 64 13. 36 300. 00 88.00 88.99 . 99 400. 00 96. 00 91. 88 4. 12 Correlation Coefficient = .980 C-l8

CAPSULE T (LONGITUDINAL ORIENTATION)

CYGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04102/2004 04:56 PM Page 1 Coefficients of Curve 2 A = 32.9 B = 32.9 C = 69.91 TO = 180.88 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=65.8 Lower Shelf L E.=.O(Fixed)

Temp. @LE. 35 mils=185.4 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-O512-2 Orientation: LT Capsule: T Fluence: n/cmA2 200 150 0

a 100 50 50 0

-300 0 300 600 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

70. 00 9. 00 2. 65 6. 35 135. 00 17. 00 13.95 3.05 175. 00 28. 00 30. 13 -2. 13 200.00 4 1. 00 41.67 -. 67 210. 00 39. 00 45. 86 -6. 86 250. 00 7 1. 00 57.79 13.21 300.00 63. 00 63. 68 -. 68 400. 00 60. 00 65. 67 -5. 67 Correlation Coefficient = .959 C-19

CAPSULE T (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:52 PM Page 1 Coefficients of Curve 2 A=5O. B = 50. C = 48.79 TO = 204.64 D = O.OOE+O0 Equation is A + B * [Tanh((T-Toy(C+DT))]

Temperature at 50% Shear = 204.7 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: T Fluence: n/cmA2 125 100

.1 75 0

60 -

25-

. -300 -.200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 70.00 10. 00 . 40 9.60 135.00 20. 00 5. 44 14.56 175.00 25.00 22. 88 2. 12 200.00 40. 00 I 4 5. 26 - 5. 26 210.00 50. 00 55. 47 - 5. 47 250. 00 99. 00 86. 52 12. 48 300. 00 100.00 98. 03 1. 97 400.00 100.00 99.97 .03 Correlation Coefficient = .982 C-20

CAPSULE Z (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/0212004 04:39 PM Page 1 Coefficients of Curve 3 A = 42.1 B = 39.9 C = 83.48 TO = 225.87 D =0.OOE+00 Equation is A + B * [Tanh((T-To)I(C+DT))]

Upper Shelf Energy=82.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=199.8 Deg F Temp@50 ft-lbs=242.7 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: Z Fluence: n/cmr2 300 250

, 200 0

0 IL E 150 z

> 100 50 0 4-

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 150. 00 18.00 13. 35 4. 65 200. 00 29.00 30. 12 -1. 12 200. 00 31.00 30. 12 . 88 225. 00 39.00 41.68 -2. 68 250. 00 52.00 53. 32 - 1. 32 325. 00 81. 00 75.21 5. 79 400. 00 82. 00 80. 79 1.21 Correlation Coefficient = .992 C-21

CAPSULE Z (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04102/2004 04:56 PM Page. 1 Coefficients of Curve 3 A = 36.71 B = 36.71 C = 98.7 TO = 222.82 D = O.OOE+00 Equation is A + B * [Tanh((T-To)y(C+DT))]

Upper Shelf L.E.=73.4 Lower Shelf L.E.=.O(Fixed)

Temp.@L.E. 35 mils=218.3 Deg P Plant: Indian Point 3 Material: SA302B Heat: A-O512-2 Orientation: LT Capsule: Z Fluence: n/cm^2

. 200 150 0

E C

0 E 100  !: f .

50 Q

...4 0-

-300 0 300 600 Temperature In Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 150.00 18.00 13. 66 4. 34 200. 00 26. 00 28. 37 -2. 37 200. 00 27.00 28. 37 - 1. 37 225. 00, 36.50 37. 52 -1. 02 250. 00 47. 5 0 46.57 .93 325.00 68.50 65. 19 3.31 400. 00 69.00 71. 44 -2.44 Correlation Coefficient = .991 C-22

CAPSULE Z (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:52 PM Page 1 Coefficients of Curve 3 A = 50. B = 50. C = 78.21 TO = 205.29 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 205.3 Plant: Indian Point 3 Material: SA302B Heat: A-O0512-2 Orientation: LT Capsule: Z Fluence: n/cmA2 125 100 Do 75 U

0, 50 0.

25 0 4-

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 150. 00 20. 00 19.56 .44 200. 00 45.00 46. 62 - 1. 62 200. 00 50. 00 46. 62 3. 38 225. 00 60. 00 62. 34 -2.34 250. 00 75.00 75. 83 - . 83 325. 00 100. 00 95.53 4.47 400. 00 100. 00 99. 32 .68 Correlation Coefficient = .997 c-23

CAPSULE X (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:39 PM Page 1 Coefficients of Curve 4 A = 41.6 -B = 39.4 C = 62.02 TO = 21035 D = O.OOE+00 Equation is A + B

• ITanh((T-To)/(C+DT))1 Upper Shelf Energy=-l8.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=191.6 Deg F Temp@50 ft-lbs=223.8 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-O512-2 Orientation: LT Capsule: X Fluence: nlcmA2 300 250

, 200 06 I

150 i; 100 50 o0 0

O ---- ----- t 1------- l ' I

-300 -200 -100 0 100 200 300 400 500 600

- Temperature In Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 100.00 7.00 4.38 2.62 150. 00 21.00 12. 05 8. 95 175. 00 22. 00 21.30 . 70 200. 00 27. 00 35.09 - 8.09 225.00 5 1. 00 50.74 .26 280.00 82. 00 73. 46 8. 54 350.00 78.00 80. 14 - 2. 14 375. 00 83.00 80. 61 2. 39 Correlation Coefficient = .984 C-24

CAPSULE X (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:56 PM Page 1 Coefficients of Curve 4 A = 32.56 B = 32.56 C = 71.68 TO = 219.55 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf LE.=65.1 Lower Shelf L.E.=.O(Fixed)

Temp. @L.E. 35 mils=225.0 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: X Fluence: nfcrnA2 200 150 a

a 100 50 0

-300 0 300 600 Temperature In Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 100. 00 2.00 2.24 - .24 150.00 14.00 8. 18 5.82 175. 00 15.00 14.58 .42 200. 00 18.00 23. 89 -5. 89 225. 00 36. 00 35. 03 .97 280. 00 59.00 54.94 4.06 350. 00 57.00 63.45 - 6.45 375.00 68. 00 64.27 3.73 Correlation Coefficient = .984 C-25

CAPSULE X (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:52 PM Page 1 Coefficients of Curve 4 A = 50. B = 50. C = 49.18 TO= 2063 D = O.OOE+O0 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 206.3 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: LT Capsule: X Fluence: n/cmA2 125 100 co 75 a-Qa, 50 25 o 4-

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 100.00 10. 00 1.31 8.69 150. 00 15.00 9.20 5.80 175. 00 20.00 21. 88 - 1. 88 200. 00 40. 00 43.63 -3.63 225.00 70. 00 68. 15 1.85 280. 00 100.00 95. 25 4. 75 350.00 100.00 99.71 .29 375.00 100. 00 99.90 .10 Correlation Coefficient = .995 C-26

UNIRRADIATED (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:07 PM Page I Coefficients of Curve i A = 35.1 B = 32.9 C = 79.37 TO = 70.69 D = O.OOE+00 Equation is A + B * [Tanh((T-To)Y(C+DT))]

Upper Shelf Energy=68.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temnp@30 ft-lbs=58.3 Deg F Temp@50 ft-lbs=109.5 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: TL Capsule: UNIRR Fluence: n/cmA2

,Q #

. 4UU

-I 250 i I i a 200 0

12 P 150 w

z B 100 50 A I S

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-20. 00 9.00 8.28 . 72

- 20. 00 7.00 8.28 - 1.28

- 20. 00 II. 00 8.28 2.72

40. 00 29. 50 22.98 6.52

40. 00 24.00 22.98 1.02 40.00 17.50 22.98 -5.48

75. 00 34. 00 36. 89 -2.89 75.00 4 1. 00 36. 89 4. 11 75.00 33.50 36. 89 -3.39 C-27

UNIRRADIATED (TRANSVERSE ORIENTATION)

Page 2 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: TL Capsule: UNIRR Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 125.00 54. 00 54. 65 - . 65 125.00 59.00 54. 65 4. 35 125.00 46.00 54. 65 -8.65 160.00 65.00 61.73 3.27 160. 00 66. 00 61.73 4.27 160.00 ~-59. 50 61.73 -2.23 210.00 62. 00 66. 09 -4.09 210.00 70. 00 66. 09 3.91 210.00 65.00 66. 09 -1. 09 210.00 70. 50 66. 09 4.41 210.00 68. 00 66. 09 1.91 210.00 70. 00 66. 09 3.91 Correlation Coefficient = .984 C-28

UNIRRADIATED (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 0410212004 05:34 PM Page 1 Coefficients of Curve I A = 31.57 B = 31.57 C = 78.27 TO = 66.73 D = O.OOE+00 Equation is A + B * [Tanh((T-To)J(C+DT))]

Upper Shelf L.E.=63.1 Lower Shelf L.E.=.O(Fixed)

Temp. @L.E. 35 mils=75.3 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: TL Capsule: UNIRR Fluence: n/cmA2 200 150 E

C 0

E 100 Lb 50 0 4-

-300 0 300 600 Temperature In Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

- 20. 00 6.00 6.21 -. 21

-20. 00 4.00 6.21 -2. 21

- 20. 00 8.00 6.21 1. 79

40. 00 27.00 21. 19 5.81

40. 00 20.00 21. 19 - 1. 19

40. 00 18.00 21. 19 - 3. 19 75.00 33.00 34. 89 - 1. 89 75.00 37.00 34.89 2. 11

75. 00 34.00 34. 89 -. 89 c-29

UNIRRADIATED (TRANSVERSE ORIENTATION)

Page 2 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: TL Capsule: UNIRR Fluence: n/cmA2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 125.00 50. 00 51.51 - 1. 51 125.00 54.00 51.51 2. 49 125.00 47. 00 51.51 -4.51 160.00 63.00 57. 80 5.20 160. 00 60.00 57. 80 2. 20 160. 00 57.00 57. 80 -. 80 210.00 62. 00 61. 55 .45 210.00 60.00 61. 55 -1.55 210.00 57.00 61. 55 -4.55 210.00 64. 00 61.55 2.45 210.00 60. 00 61.55 - 1. 55 210.00 63.00 61.55 1.45 Correlation Coefficient = .991 C-30

UNIRRADIATED (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:14 PM Page 1 Coefficients of Curve I A =50. B =50. C = 83.96 TO = 92.15 D =0.OOE+00 Equation is A + B

• fTanh((T-To)/(C+DT))]

Temperature at 50% Shear = 92.2 Plant: Indian Point 3 Material: SA302B Heat: A-O512-2 Orientation: TL Capsule: UNIRR Pluence: n/cmA2 125 100 -

I..

C, 75 CD 50 0-IL 00 0

0 25 0

0

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

20. 00 5.00 6.47 -1. 47

20. 00 5.00 6.47 -1. 47

20. 00 9. 00 6.47 2.53 40.00 32.00 22.41 9.59

40. 00 33.00 22. 41 10.59

40. 00 21. 00 22.41 -1.41

75. 00 41. 00 39. 93 1. 07 75.00 47.00 39.93 7.07 75.00 42.00 39. 93 2.07 C-31

UNIRRADIATED (TRANSVERSE ORIENTATION)

Page 2 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: TL Capsule: UNIRR Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 125.00 47.00 68.62 -21.62 125.00 51.00 68. 62 - 17. 62 125.00 41.00 68. 62 - 27.62 160.00 100.00 83.43 16. 57 160.00 100. 00 83.43 16.57 160.00 100. 00 83.43 16. 57 210.00 100.00 94.31 5. 69 210.00 100. 00 94.31 5.69 210.00 1 00. 00 94.31 5. 69 210. 00 100.00 94. 31 5. 69 210.00 100.00 94.31 5. 69 210. 00 100. 00

• 94.31 5. 69 Correlation Coefficient = .950

• C-32

CAPSULE T (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04102/2004 05:07 PM Page 1 Coefficients of Curve 2 A = 29.6 B 27.4 C = 97A3 TO = 162.76 D = O.OOE+00 Equation is A + B * [Tanh((T-To)Y(C+DT))]

Upper Shelf Energy=57.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=164.2 Deg F Temp@50 ft-lbs=256.4 Deg F Plant: lndian Point 3 Material: SA302B- Heat: A-0512-2 Orientation: TL Capsule: T Fluence: n/cmA2 300 250

, 200 0a 0.

150 C

w z

i 100I 50 0 *o=

-300 -200 -100 0 100 ¢ 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

70. 00 13. 50 9. 30  ! 4. 20 120. 00 22.50 18. 29 4. 21 150.00 22.00 26. 03 -4. 03 175.00 30.00 33. 02 - 3.02 210.00 36.50 41.93 -5. 43 225.00 47.50 45.06 2.44 250.00 56. 00 49. 16 6. 84 300.00 58.00 S3. 91 4.09 Correlation Coefficient = .961 C-33

CAPSULE T (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:34 PM Page 1 Coefficients of Curve 2 A = 38.02 B = 38.02 C = 150.8 TO = 221.15 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=76.0 Lower Shelf L.E.=.O(Fixed)

Temp.@L.E. 35 mils=209.2 Deg F Plant: Indian Point 3 Material: SA302B Heat A-0512-2 Orientation: TL Capsule: T Fluence: nlcm^2 200 150 A

E a 100 a

50 00 -- -- 0 3---------0

-300 0 300 600 Temperature In Deg F Charpy V-Notch Data Temperature Input L.E. Computed LE. Differential

70. 00 10. 00 9.03 .97 120. 00 17.00 15. 76 1.24 150.00 19.00 21.31 -2.31 175.00 28.00 26. 74 1.26 210.00 32.00 35. 22 - 3.22 225.00 39. 00 38.99 .01 250. 00 49.00 45. 21 3.79 300. 00 55.00 56. 27 - 1.27 Correlation Coefficient = .990 C-34

I CAPSULE T (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:14 PM Page 1 Coefficients of Curve 2 A = 50. B = 50. C = 70.63 TO = 200.34 D -O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 200.4 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: TL Capsule: T- Fluence: n/cmA2 125 100

.1 75

0. 50 25 0 4-

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

70. 00 10. 00 2.43 7.57 120.00 25.00 9.32 15.68 150.00 20. 00 19.38 .62 17 5. 00 30.00 32.79 .- 2.79 210.00 45.00 56.79 - 11. 79 225.00 60. 00 66. 78 -6. 78 250. 00 100.00 80. 32 19.68 300.00 100. 00 94. 39 5.61 Correlation Coefficient = .952 C-35

CAPSULE Y (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:07 PM Page 1 Coefficients of Curve 3 A = 26.6 B = 24.4 C = 86.38 TO = 195.04 D = O.OOE+OO Equation is A + B * [Tanh((T-To)l(C+DT))]

Upper Shelf Energy--51.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs--207.2 Deg F Temp@50 ft-lbs=362.1 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: TL Capsule: Y Fluence: nlcmA2 300 250

_ 200 08 0

IL 150 i;100 50 0-

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 150. 00 16.00 14.92 1.08 200. 00 23.00 28. 00 -5.00 225.00 39.50 34. 74 4.76 225.00 36.00 34. 74 1.26 300. 00 42. 50 47.05 -4. 55 325.00 48.00 48.71 - .71 400. 00 57. 00 50. 58 6.42 450. 00 56. 50 50. 87 5.63 Correlation Coefficient = .958 C-36

CAPSULE Y (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:34 PM Page 1 Coefficients of Curve 3 A = 26.67 B = 26.67 C = 104.32 TO = 220.53 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))l Upper Shelf LE.=53.3 Lower Shelf L.E.=.O(Fixed)

Temp.@LE. 35 mils=254.3 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: TL Capsule: Y Fluence: n/cmA2 200 150 E

C 0

E 100 C

50 0 -

-300 0 300 600

- Temperature In Deg F Charpy V-Notch Data Temperature I Input L.E. Computed LE. Differential 150.00 10.00 10.96 - . 96 200. 00 15.50 21. 49 - 5. 99 225. 00 31. 50 27. 82 3. 68 225.00 33.50 27. 82 5. 68 300. 00 36. 50 43. 80 -7. 30 325.00 50.50 47.00 3.50 400.00 49. 50 . 51. 69 - 2. 19 450. 00 55. 50 52. 70 2. 80 Correlation Coefficient = .957 C-37

CAPSULE Y (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Pnnted on 04/02/2004 05:14 PM Page 1 Coefficients of Curve 3 A = 50. B ='50. C = 79.94 TO = 221.88 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)Y(C+DT))]

Temperature at 50% Shear = 221.9 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: TL Capsule: Y Fluence: n/cmA2 125 100 t.

Ca 0 75 0

0 DL

0. 50 25 0 1 , - I

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 150.00 3 1. 00 14. 21 16. 79 200. 00 36. 00 36. 65 -. 65 225.00 47.00 51. 95 -4. 95 225. 00 42. 00 51. 95 -9. 95 300.00 100.00 87. 59 12.41 325. 00 100. 00 92.96 7.04 400. 00 1 00. 00 98. 85 1. 15 450.00 1 00. 00 99. 67 .33 Correlation Coefficient = .963 C-38

CAPSULE Z (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:07 PM Page 1 Coefficients of Curve 4 A = 29.1 B = 26.9 C = S031 TO = 214A2 D = O.OOE+00 Equation is A + B * [Tanh((T-To)y(C+DT))]

Upper Shelf Energy=56.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=216.2 Deg F Temp@45 ft-lbs=266.7 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: TL Capsule: Z Fluence:- nIcmn2 300 250 a°, 200 46.,

0 0) b 150 C) w z

6 100l 50 0

-300 -200 -100 0 100 - 200 300 400 500 600 Temperature in Deg F

Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 175.00 10.00 11.49 - 1.49 200. 00 23.00 21. 59 I1.41 225.00 40. 00 34. 67 5.33 225. 00 3 1. 00 34. 67 -3. 67 250. 00 40. 00 45. 48 -5. 48 275.00 56.00 51. 56 4. 44 350.00 60. 00 55. 76 4. 24 425.00 52.00 55. 99 - 3. 99 Correlation Coefficient = .968

-C-39

CAPSULE Z (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:34 PM Page 1 Coefficients of Curve 4 A = 26.96 B = 26.96 C = 59.95 TO = 214.3 D = O.OOE+0O Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=53.9 Lower Shelf L.E.=.O(Fixed)

Temp.@L.E. 35 nils=232.8 Deg P Plant Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: TL Capsule: Z Fluence: n/cmn2 200 150

.9 E

C 0

a 100 6

C) o 50 0

-300 0 300 600 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 175.00 10.50 11. 44 - .94 200. 00 21.50 20. 64 .86 225. 00 36. 00 31.72 4. 28 225. 00 28.50 31.72 -3. 22 250. 00 39.00 41. 35 - 2.35 275.00 48.50 47. 63 .87 350. 00 56.00 53. 34 2. 66 425.00 51. 50 53. 86 -2.36 Correlation Coefficient = .986 C-40

CAPSULE Z (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:14 PM Page I Coefficients of Curve 4 A =50. B =50. C = 42.51 TO = 203.29 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 203.3 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: TL Capsule: Z Fluence: n/cmA2 125 100 a.-

Ca C) 75 co C)

C) 50 C-25 0

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 175.00 20. 00 20.90 - . 90 200. 00 .50. 00 46. 13 3. 87 225.00 90. 00 73. 52 16.48 225.00 - 50.00 73.52 - 23.52 250. 00 95.00 90. 00 5.00 275.00 I100.00 I 96. 69 3.31 350.00 100. 00 99. 90 .10 425.00 100.00 -100.00 .00 Correlation Coefficient = .933 C-41

CAPSULE X (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:07 PM Page 1 Coefficients of Curve 5 A = 27.1 B = 24.9 C = 75.44 TO = 207.61 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))J Upper Shelf Energy=52.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=216.5 Deg F Temp@50 ft-lbs=327.4 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: ll Capsule: X Fluence: n/cmA2 250

,. 200 a

0 150 0

z

> 100 50 I~

t~~CAi< -

I- __ I- -- I - - -

0

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 100. 00 6.00 4.92 1. 08 175. 00 20. 00 16. 96 3.04 210. 00 22.00 27. 89 -5. 89 225. 00 33.00 32. 74 .26 250. 00 44.00 39. 78 4.22 325. 00 47. 00 49. 88 - 2. 88 375. 00 54. 00 51.42 2.58 390. 00 55. 00 51.61 3.39 Correlation Coefficient = .981 c-42

CAPSULE X (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/0212004 05:34 PM Page 1 Coefficients of Curve 5 A = 22.2 B _ 22.2 C = 75.96 TO = 219.26 D _ O.OOE+00 Equation is A + B

• ITanh((T-To)/(C4DT1'))]

-Upper Shelf LE.=44.4 Lower Shelf L.E.=.O(Fixed)

Temp.@L.E. 35 mils=269.3 Mg F Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: TL Capsule: X Fluence: nIcmA2 200 150 W

E C

0 8 100 A

50

. . 7 0

300 0 300 600 Temperature In Deg F Charpy V-Notch Data Temperatu re Input LE. Computed L.E. Differential 100.0 0 .00 1.84 -1.84 175.0 0 14.00 10.55 3.45 210.0 )0 14.00 19.50 -5.50 225.0 )0 25.00 23.87 1.13 250. 0)0 34.00 30.72 3.28 325. C10 38.00 41.81 -3.81 375. C10 45.00 43.67 1.33 390. C10 45.00 43.91 1.09 Correlation Coefficient = .980 C-43

CAPSULE X (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:14 PM Page 1 Coefficients of Curve 5 A = 50. B = 50. C = 37.97 TO = 217.1 D = 0OOE+O0 Equation is A + B * [Tanh((T-ToY(C+DT))]

Temperature at 50% Shear = 217.1 Plant: Indian Point 3 Material: SA302B Heat: A-0512-2 Orientation: TL Capsule: X Fluence: n/cmA2 125 100 I

0 Z.

75 I

CO) j lu 4..1 I 0 0 CL 50 I I

--- V 25 I f 7 I, I

aI i i i I i i i

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 100. 00 15. 00 .21 14.79 175. 00 25. 00 9. 82 15. 18 210.00 30. 00 40.76 - 10.76 225. 00 60.00 60. 26 - .26 250. 00 95. 00 84. 98 10.02 325.00 100.00 99.66 .34 375. 00 100. 00 99. 98 .02 390. 00 1 00. 00 99.99 .01 Correlation Coefficient = .978 C-44

UNIRRADIATED (WELD)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 02:26 PM Page 1 Coefficients of Curve 1 A = 61.1 B = 58.9 C = 47.03 TO = -37.11 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=120.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-64.7 Deg F Temp@50 ft-lbs=-46.0 Deg F Plant- Indian Point 3 Material: SAW Heat: W5214 Orientation: NA ' Capsule: UNIRR Fluenoe: n/cmA2 300 250 n 200 8

0

£1.- ;150 150

> 100 50 0 I I1 1

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 150. 00 5.00 3. 16 1. 84 150. 00 2.00 3. 16 - 1. 16 150. 00 4.50 3. 16 1. 34 100. 00 29. 00 9. 80 19. 20 100. 00 18. 00 9. 80 8. 20 100. 00 25.50 9. 80 15.70

- 50. 00 35.00 45.35 - 10. 35

-50. 00 33. 00 45. 35 - 12. 35

-50. 00 32.50 45. 35 - 12. 85 C-45

UNIRRADIATED (WELD)

Page 2 Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: UNIRR Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-35. 00 78.00 63. 74 14.26

-35. 00 69. 50 63. 74 5.76

-35. 00 54. 50 63. 74 -9. 24

-20. 00 87.00 81. 63 5.37

- 20. 00 82.00 81. 63 .37

- 20. 00 89. 00 81.63 7. 37 10.00 100.00 106. 00 - 6.00 10.00 105. 00 106. 00 - 1. 00 10.00 113.50 106. 00 7. 50

60. 00 115. 00 118.14 - 3. 14

60. 00 1 19. 00 118. 14 .86

60. 00 121.50 118.14 3. 36 160. 00 124. 00 119.97 4.03 160. 00 125. 00 119. 97 5.03 160.00 112.00 119.97 - 7. 97 Correlation Coefficient = .981 C46

UNIRRADIATED (WELD)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/0212004 04:10 PM Page 1 Coefficients of Curve I A = 45A1 B = 45.41 C = 58.14 TO = -45.8 D = O.OOE+00 Equation is A + B * [Tanh((T-ToY(C+DT))]

Upper Shelf L.E.=90.8 Lower Shelf L.E.=.O(Fixed)

- -Temp.@L.E.35 nfls=-59.3 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: UNIRR Fluence: n/cmA2 200 150 2

E 50 C,

-0 so10 it

-300 0 300 600 Temperature In Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 150. 00 2. 00 2.45 -. 45 150. 00 2.00 2.45 - .45 150.00 4. 00 2. 45 1.55 100.00 22.00 12. 19 9.81 100. 00 16.00 12. 19 3. 81 100. 00 23. 00 12. 19 10.81

- 50. 00 34. 00 42. 14 - 8. 14

-50. 00 30.00 42. 14 - 12. 14

-50. 00 30. 00 42. 14 - 12. 14 C47

UNIRRADIATED (WELD)

Page 2 Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: UNIRR Fluence: n/cmA2 Charpy V-Notch Data Temperature Input LE. Computed L.E. Differential

-35. 00 66.00 53.75 12. 25

-35. 00 56.00 53.75 2.25

-35. 00 47.00 53.75 -6. 75

- 20. 00 69.00 64. 34 4.66

- 20. 00 63.00 64. 34 -1. 34

- 20.00 74.00 64. 34 9.66 10.00 78.00 79. 20 - 1.20 10.00 8 1. 00 79. 20 1.80 10.00 85.00 79. 20 5.80

60. 00 89. 00 88. 49 .51 60.00 84. 00 88. 49 -4. 49 60.00 90. 00 88. 49 1.51 160. 00 88.00 90. 74 -2. 74 160. 00 89.00 90. 74 - 1. 74 160.00 90. 00 90.74 - .74 Correlation Coefficient = .979 C48

UNIRRADIATED (WELD)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 03:54 PM Page 1 Coefficients of Curve I A =50. B =50. C = 60.8 TO = -47.89 D= O.OOE+00 Equation is A + B * [Tanh((T-ToY(C+DT))j Temperature at 50% Shear = 47.8 Plant: Indian Point 3 Material: SAW - Heat: W5214 Orientation: NA Capsule: UNIRR Fluence: n/cmA2 125 100 I-co 75 C)

(0 L) 2E 50 25 0 4-

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 150. 00 5.00 3. 36 1.64 150. 00 5.00 3.36 1.64 150. 00 9.00 3.36 5.64 100. 00 20. 00 15.26 4.74 100. 00 18.00 15.26 2.74 100.00 23.00 15.26 7.74

- 50.00 40.00 48.26 -8. 26

-50. 00 47. 00 48.26 - 1.26

-50.00 40. 00 48.26 -8.26 C-49

UNIRRADIATED (WELD)

Page 2 Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: UNIRR Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 35. 00 64. 00 60. 44 3.56

-35. 00 67.00 60. 44 6. 56

-35. 00 40. 00 60. 44 20.44

- 20. 00 77.00 71.45 5. 55

- 20. 00 77.00 71.45 5.55

- 20.00 81. 00 71.45 9. 55

10. 00 8 1. 00 87. 04 -6. 04 10.00 82.00 87.04 -5. 04 10.00 100.00 87. 04 12. 96

60. 00 100.00 97.20 2. 80

60. 00 100.00 97. 20 2. 80

60. 00 100. 00 97.20 2. 80 1 60. 00 100.00 99. 89 .11 1 60. 00 100.00 99. 89 .11 160. 00 100.00 99. 89 .11 Correlation Coefficient = .980 C

CAPSULE T (WELD)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 02:26 PM Page 1 Coefficients of Curve 2 A = 43.1 B = 40.9 C = 87.09 TO = 115.75 D = O.OOE+00 Equation is A + B * [Tanh((T-To)Y(C+DT))J Upper Shelf Energy=84.0(Fixed) Lower Shelf Energy-2.2(Fixed)

Temp@30 ft-lbs=86.9 Deg F Ternp@50 ft-lbs=130.6 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: T Fluence:. n/cmA2 300 250

, a 200 06 0

UO.

@150 50 0

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

, 00 13. 00 7. 56 5. 44

70. 00 17.50 23. 39 -S. 89 110.00 48. 00 40. 40 7. 60 150.00 55. 50 58. 40 -2. 90

.150.00 53.00 58. 40 -5. 40 165.00 66.00 64.04 1.96 210.00 78.00 75. 58 2. 42 300. 00 90.50 82. 83 7.67 Correlation Coefficient = .979 C-51

CAPSULE T (WELD)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:10 PM Page 1 Coefficients of Curve 2 A = 40.17 B = 40.17 C = 112.89 TO = 127.9 D = O.OOE+00 Equation is A + B * [Tanh((r-ToY(C+DT))I Upper Shelf LE.-80.3 Lower Shelf L.E.=.O(Fixed)

Temp.@L.E. 35 mils=113.3 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: T Fluence: n/cmA2 200 150 S

.2 C

a 100 J

50 0

-300 0 300 600 Temperature In Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

. 00 9.00 7.55 1.45

70. 00 17. 00 21.20 -4. 20 110. 00 45. 00 33. 86 11.14 150. 00 46.00 47. 94 - 1. 94 150.00 35. 00 47.94 - 12. 94 165.00 55.00 52. 92 2. 08 210.00 73.00 65. 14 7. 86 300. 00 74. 00 7 6. 7 1 -2. 71 Correlation Coefficient = .949 C-52

CAPSULE T (WELD)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 03:54 PM Page 1 Coefficients of Curve 2 A = 50. B = 50. C = 103.39 TO = 123.94 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT)))

Temperature at 50% Shear = 124.0 Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: T Fluence: n/cmA2 125 100 S-so) 75-co r:

-C 0~

25o _. _.

30

-300 -200 -1 00 0 100  ; 200 -300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

. 00 15. 00 8. 34 6. 66

70. 00 20. 00 26. 05 -6. 05 110.00 55. 00 43. 30 11. 70 150. 00 60.00 62. 34 -2. 34 150.00 55. 00 - 62. 3 4 -7. 34 1 65. 00 60. 00 68. 88 - 8.88 210.00 98. 00 84.09 13.91 300.00 100.00 96. 79 3. 21 Correlation Coefficient - .958 C-53

CAPSULE Y (WELD)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 02:26 PM Page 1 Coefficients of Curve 3 A = 35.6 B = 33.4 C = 90.16 TO = 122.54 D = O.OOE+00 Equation is A + B * [Tanh((T-To)Y(C+DT))I Upper Shelf Energy=69.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=107.3 Deg F Temp@50 ft-lbs=164.2 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: Y Fluence: n/cmA2 300-250

, 200 0

0 IL 2e 150 0

w Ul z

> 100 50 0

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

25. 00 20. 00 9.08 10. 92 7 2. 00 19.50 18. 62 .88 12 5. 00 31. 00 36.51 -5.51 1 25. 00 29.50 36.51 -7.01 1 50. 00 49.00 45. 47 3.53 200. 00 67. 50 58. 84 8.66 300. 00 69.50 67.72 1.78 400. 00 68. 50 68. 86 -. 36 Correlation Coefficient = .960 c-54

CAPSULE Y (WEL D)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04:10 PM Page 1 Coefficients of Curve 3 A = 32.81 B = 32.81 C = 94.65 TO = 138.68 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=65.6. Lower Shelf L.E.=.O(Fixed)

Temp. @LE.35 mils=145.0 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: Y Fluence: n/cMA2 200

.150 (n

0 E10 100 50 O. ...4

.0

-300 0 300 600 Temperature In Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 25.00 12. 50 5.45 7.05 72.00 13.50 12. 89 .61 125.00 25. 00 28. 10 -3. 10 125. 00 22.00 28. 10 -6. 10 150. 00 41.00 36.72 4.28 200. 00 55.50 51.52 3.98 300.00 60. 00 63. 52 -3.52 400. 00 66. 50 65. 36 1. 14 Correlation Coefficient = .979 C-55

CAPSULE Y (WELD)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 03:54 PM Page 1 Coefficients of Curve 3 A = 50. B = 50. C = 22.27 TO = 132.59 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 132.6 Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: Y Fluence: nlcmA2 125 100 2-0 75 CO IV 0 50 25 0

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

25. 00 27. 00 .01 26. 99

72. 00 19. 00 .43 18.57 125.00 38.00 33.59 4.41 125.00 28.00 33.59 -5. 59 150. 00 84.00 82. 69 1.31 200. 00 98.00 99. 77 - 1. 77 300. 00 100. 00 100.00 .00 400. 00 100.00 100. 00 .00 Correlation Coefficient = .978 C-56

CAPSULE Z (WELD)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04102/2004 02:26 PM Page 1 Coefficients of Curve 4 A =39.1 B = 36.9 C= 97.52 TO = 188.96 D= O.OOE+00 Equation is A + B * [Tanh((T-To)I(C+DT))]

Upper Shelf Energy=76.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=164.5 Deg F Temp@50 ft-lbs=218.7 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: Z Fluence: n/cmA2 300 250 a 200 10.

0 U-P 150 C0) w z

100 50 0

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 100.00 10.00 12.45 -2. 45 150. 00 2 1. 00 25. 10 -4. 10 150.00 44. 00 25. 10 18.90 175. 00 26. 00 33. 85 -7. 85 200. 00 33.00 43. 26 - 10.26 225. 00 57. 00 52. 15 4. 85 300.00 7 5. 00 69. 14 5.86 400. 00 77.00 75.04 1.96 Correlation Coefficient = .929 C-57

CAPSULE Z (WELD)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 04: 10 PM Page 1 Coefficients of Curve 4 A = 36.77 B = 36.77 C = 129.86 TO = 194.01 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))j Upper Shelf L.E.=73.5 Lower Shelf L.E.=.O(Fixed)

Temnp.@L.E. 35 mils=187.8 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: Z Fluence: n/cmA2 200 150 E

.2 4 1 0 o 50 o 4-

-300 0 300 600 Temperature In Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 100.00 1 1. 00 14. 00 -3. 00 150.00 20. 00 24. 77 -4. 77 150. 00 40.00 24.77 15. 23 175. 00 28. 00 31.43 - 3.43 200. 00 29. 50 38.47 - 8.97 225. 00 49. 00 45. 39 3.61 300. 00 64. 50 61.52 2.98 400.00 69.00 70. 59 - 1. 59 Correlation Coefficient = .935 C-58

CAPSULE Z (WELD)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 03:54 PM Page 1 Coefficients of Curve 4 A _ 50. B = 50. C = 103.22 TO = 147.45 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT)))

Temperature at 50% Shear = 147.5 Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: Z Fluence: n/cmA2 125 100 I.-

LU 0 75 to C.

50 25 0 4-

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 100.00 30. 00 28.51 1. 49 150. 00 55.00 51.23 3.77 150.00 50. 00 51.23 - 1.23 175.00 65.00 63.04 1.96 200.00 55.00 73.46 - 18.46 225.00 95.00 81.80 13.20 300. 00 I100.00 95.05 4. 95 400. 00 100. 00 99.26 .74 Correlation Coefficient = .942 C-59

CAPSULE X (WELD)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/0212004 02:36 PM Page I Coefficients of Curve 5 A = 38.1 B = 35.9 C = 118.98 TO = 155.76 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=74.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=128.5 Deg F Temp@50 ft-lbs=196.8 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: X Fluence: n/cm^2 300 250 -

-,p 200 0

0 LL E 150 0

z

>) 1 00 50 o 4=-=

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

75. 00 9.00 16. 89 -7.89 125.00 49.00 29. 02 19.98 125. 00 24.00 29. 02 -5.02 150.00 35.00 36. 36 - 1. 36 200. 00 37.00 50. 87 - 13. 87 250. 00 67. 00 61.78 5.22 300. 00 72.00 68. 16 3.84 350. 00 75.00 71. 36 3.64 Correlation Coefficient = .906 C-60

CAPSULE X (WELD)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Pnnted on 04/0212004 04:11 PM Page 1 Coefficients of Curve S5 A = 31.07 B = 31.07 C = 134.21 TO = 167.48 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=62.1 Lower Shelf L.E.=.O(Fixed)

Temp.@L.E. 35 mils=184.6 Deg F Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: X Fluence: n/cmA2

. 200 150 to E

.2 o a 100 C

Lb 50 o~ .- I l,_ - Ii . . I

-300 0 300 600 Temperature In Deg F Charpy V-Notch Data Temperature Input LE. Computed L.E. Differential 75.00 5. 00 .12. 51 -7. 51 125. 00 36. 00 21. 55 14.45 125. 00 19.00 21. 55 -2. 55 150.00 26. 00 - 27. 05 - 1. 05 200. 00 30. 00 38. 45 --8. 45 250.00 -52. 00 4 8. 08 3. 92 300. 00 56.00 54. 57 1.43 350.00 57. 00 58. 30 - 1. 30 Correlation Coefficient = .924 C-61

CAPSULE X (WELD)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 03:55 PM Page 1 Coefficients of Curve 5 A = 50. B = 50. C = 102.97 TO = 144.46 D = O.OOE+0O Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 144.5 Plant: Indian Point 3 Material: SAW Heat: W5214 Orientation: NA Capsule: X Fluence: n/cmn2 125 -

100 -

I-75 U) i~

0 50 -

a-25-0

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 75.00 20. 00 20. 60 -. 60 125.00 50.00 40. 66 9. 34 125.00 40. 00 40.66 - . 66 150.00 45.00 52. 69 -7. 69 200. 00 70. 00 74. 63 -4. 63 250. 00 95. 00 88. 59 6.41 300.00 98.00 95. 35 2. 65 350. 00 100. 00 98. 19 1.81 Correlation Coefficient = .984 c-62

UNIRRADIATED (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:41 PM Page 1 Coefficients of Curve 1 A = 63.6 B = 61A C = 84.72 TO = -2A6 D = O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=125.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-54.5 Deg F Temp@50 ft-lbs=-21.5 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-0516-2 Orientation: LT Capsule: UNIRR Fluence: n/cmr2 300 250

,. 200 0

0 LL

150

C w

z

> 100 50 0

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

- 100.00 5. 00 13.36 -8. 36

- 100. 00 6.00 13.36 -7.36

-I 00. 00 5. 00 13. 36 -8.36

- 40. 00 50. 50 38.04 12. 46

.40.00 43.00 38.04 4. 96

-40.00 44. 50 38.04 6. 46 10.00 79.00 72. 57 6.43

10. 00 75.00 72. 57 2.43

• 10. 00 61.00 72.57 -11. 57 C-63

UNIRRADIATED (LONGITUDINAL ORIENTATION)

Page 2 Plant: Indian Point 3 Material: SA302B Heat: A-0516-2 Orientation: LT Capsule: UNIRR Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

40. 00 100.00 92. 03 7.97

40. 00 92.00 92. 03 - . 03

40. 00 65. 50 92. 03 -26.53 110. 00 123.00 116.93 6.07 1 10. 00 121. 00 116. 93 4.07 1 10.00 129.00 116.93 12.07 210.00 133.50 124. 19 9. 31 210.00 115.00 124. 19 -9. 19 210.00 125.00 124. 19 .81 Correlation Coefficient = .974 C-64

UNIRRADIATED (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:47 PM Page 1 Coefficients of Curve 1 A = 39.9 B = 39.9 C = 69.47 TO = -30.12 D = O.OOE+00 Equation is A + B * [Tanh((T-To)l(C+DT))J Upper Shelf L.E.=79.8 Lower Shelf L.E.=.O(Fixed)

Temp. @L.E. 35 nils=-38.6 Deg F Plant: Indian Point 3 Material: SA302B Heat: A-O516-2 Orientation: LT Capsule: UNIRR Fluence: nlcm^2 200 150

.2 2E a 100 C) 50 0 I-

-300 0 300 600 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

- 100. 00 2.00 9. 41 - 7. 4 1

- 100. 00 2.00 9.41 -7. 41

- 100. 00 5. 00 9.41 -4. 41

-40. 00 43.00 34.26 8.74

-40. 00 40.00 34.26 5. 74

-40. 00 38. 00 34. 26 3.74 10.00 61. 00 60. 68 .32 10.00 60. 00 60. 68 - . 68

10. 00 5 1. 00 60. 68 -9. 68 C-65

UNIRRADIATED (LONGITUDINAL ORIENTATION)

Page 2 Plant: Indian Point 3 Material: SA302B Heat: A-0516-2 Orientation: LT Capsule: UNJRR Fluence: n]cmA2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

40. 00 72.00 70. 44 1.56

40. 00 74.00 70. 44 3. 56

40. 00 56. 00 70.44 - 14. 44 110.00 88.00 78.41 9. 59 110.00 83.00 78.41 4. 59 I 10. 00 84.00 78. 41 5.59 210.00 75. 00 79.72 -4.72 210.00 82. 00 79.72 2.28 210.00 73.00 79. 72 -6. 72 Correlation Coefficient = .972 C-66

UNIRRADIATED (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:44 PM Page 1 Coefficients of Curve I A =50. B= 50. C = 7231 TO = 22-57 D= O.OOEiOO Equation is A + B * [Tanh((T-To)l(C+DT))]

Temperature at 50% Shear = 22.6 Plant: Indian Point 3 Material: SA302B Heat: A-0516-2 Orientation: LT Capsule: UNERR Fluence: n/cmA2 125 100 s

75 co

0. 50 25 0 I-

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 100. 00 . 00 3.26 -3. 26

-100. 00 . 00 3.26 - 3.26

- 100. 00 . 00 3.26 -3. 26

- 40. 00 20. 00 15.05 4.95

- 40. 00 18.00 15.05 2.95

-40. 00 20. 00 15.05 4.95 10.00 40. 00 41. 40 -1. 40

10. 00 45.00 41.40 3.60 10.00 45.00 41.40 3. 60 C-67

UNIRRADIATED (LONGITUDINAL ORIENTATION)

Page 2 Plant: Indian Point 3 Material: SA302B Heat: A-0516-2 Orientation: LT Capsule: UNIRR Fluence: nIcmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

40. 00 59.00 61. 82 -2. 82

40. 00 59. 00 61. 82 -2. 82

40. 00 48.00 61. 82 - 13. 82 110. 00 100.00 91. 82 8. 18 110. 00 100.00 91. 82 8. 18 1 0. 00 100.00 91. 82 8. 18 210. 00 100. 00 99.44 .56 210.00 100.00 99. 44 . 56 210. 00 100. 00 99. 44 .56 Correlation Coefficient = .990 C-68

CAPSULE X (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:41 PM Page 1 Coefficients of Curve 2 A = 53.6 B = 51.4 C = 109.99 TO = 152.64 D = O.OOE+00 Equation is A + B * (Tanh((T-To)/(C+iDT))]

Upper Shelf Energy=105.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=98.1 Deg F Temp@50 ft-lbs=145.0 Deg F Plant: Indian Point 3 Material: SA302B f Heat: A-0516-2 Orientation: LT Capsule: X Fluence: n/cmA2 300 250 J 200 0

U-P 150 w

t100 50 0

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input CVN . Computed CVN Differential 25.00 8.00 11. 39 -3. 39

.75. 00 24.00 22. 34 1.66 125.00 59.00

• 40. 95 18.05 150.00 40. 00  ; 52. 37 - 12. 37 200. 00 S8. 00

• 74. 46 - 16.46 2 5 0. 00 104.00 90. 04 13..96 300. 00 105. 00 98. 40 6.60 325.00 105.00 100.71 4.29 Correlation Coefficient = .951 C-B9

CAPSULE X (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:47 PM Page I Coefficients of Curve 2 A = 37.62 B = 37.62 C = 118.27 TO = 155.47 D = O.OOE+00 Equation is A + B

• lTanh((T-To)/(C+DT))]

Upper Shelf L.E.=75.2 Lower Shelf L.E.=.O(Fixed)

Temp. @L.E. 35 mils=147.3 D)eg F Plant: Indian Point 3 Material: SA302B Heat: A-0516-2 Orientation: LT Capsule: X Fluence: n/cmA2 200 150 M

E C

a. 100 50 0

-300 0 300 600 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

25. 00 3.00 7.46 -4. 46 75.00 14.00 15.36 - 1.36 125. 00 40.00 28. 14 11. 86 150. 00 30. 00 35. 88 -5. 88 200. 00 44.00 51. 15 -7. 15 250. 00 69.00 62. 58 6.42 300. 00 7 1. 00 69.23 1.77 325. 00 68. 00 71. 19 - 3. 19 Correlation Coefficient = .968 C-70

CAPSULE X (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 04/02/2004 05:44 PM Page 1 Coefficients of Curve 2 A = 50. B = 50. C = 83.85 TO = 153.16 D = O.OOE+00 Equation is A + B * [Tanh((T-Toy(C+DT))]

Temperature at 50% Shear = 153.2 Plant: Indian Point 3 Material: SA302B Heat: A-0516-2 Orientation: LT Capsule: X Fluence: n/cmA2 125 100 75 V)

C) 0~ 50 25 0

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

25. 00 5.00 4. 49 .5 1

75. 00 15. 00 13.42 1. 58 125. 00 30.00 33.81 -3. 81 150.00 55.00 48. 12 6. 88 200. 00 65.00 75. 35 - 10. 35 250. 00 100.00 9.0. 97 9.03 300.00 100.00 97.08 2.92 325. 00 100.00 98.37 1.63 Correlation Coefficient = .988 C-71

D-O APPENDIX D INDIAN POINT UNIT 3 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION Appendix D

D-l INTRODUCTION:

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 have 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 afull margin term). 3) The surveillance weld heat is not the same heat as the beltline welds (intermediate/lowershell longitudinal wield & intermediateto lower shell girth weld), thus should not be used for vessel material predictions (see discussion under Criterionl);'

Appendix D

D-2 EVALUATION:

Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The beitline 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 Lindel092).

Per WCAP-8475, the Indian Point Unit 3 surveillance program was based on ASTM E 185-62. 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 betline 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 beitline 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 befltine region. Therefore it was chosen as the surveillance weld material.

Hence, Criterion I is met for the Indian Point Unit 3 reactor vessel.

Appendix D

D-3 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 28TF for welds and 17*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 280 F for welds and less than 170F 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.

Appendix D

D-4 TABLE D-1 Calculation of Chemistry Factors using Indian Point Unit 3 Surveillance Capsule Data Material Capsule Capsule P) FF°) W lRTNT FF*&RTNWDT FF2 Lower Shell T 0.263 0.637 139.4 88.798 0.406 Plate B2803-3 Z 1.04 1.01 167.8 169.478 1.02 (Longtudina X 0.874 0.962 159.6 153.535 0.925 Lower Sheil T 0.263 0.637 105.9 67A58 OA06 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 CFB2W3.3 - ZFF A RTmT) + S( FF2 ) (924.499) + (5.507) = 167.9 0 F Notes:

(a) f= fluence. Calculated fluence from Section 6 of this report [x 1019 nlcm 2 , E > 1.0 MeVI.

(b) FF = fluence factor - tO-.2 o r.

(c) ARTNDT values are the measured 30 ft-lb shift values taken from Figures 5-1 and 5-4, herein [°F].

Appendix D

D-5 The scatter of ARTmT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table D-2.

Table D-2:

Indian Point Unit 3 Surveillance Capsule Data Scatter about the Best-Fit Line for Surveillance Forging Materials.

Material Cpu CF FF Measured Predicted Scatter < 170 F (OF) ARTmT ARTN=(2) ARTry (IF) (Base Metals)

Lower Shell Plate T 167.9 0.637 139.4 107.0 32.4 No B2803-3 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 T 167.9 0.637 105.9 107.0 -1.1 Yes Lower Shell Plate B2803-3 Y 167.9 0.897 148.9 150.6 -1.7 Yes (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 FF) Per Equation 2 of Reg. Guide 1.99 Rev. 2 Position 1.1.

Conclusion for Criterion 3:

Table D-2 indicates that only 1 of 7 data points falls outside the +/- la of 17*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.

Appendix D

D-6 Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 257F.

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 25'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.

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, ORNLJTM-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 criteria is met.

CONCLUSION:

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.

AppendixDD