ML061030088

WCAP-16526-NP, Rev. 0, Analysis of Capsule V from FPL Energy - Seabrook Unit 1 Reactor Vessel Radiation Surveillance Program.

ML061030088
Person / Time
Site:	Seabrook
Issue date:	03/31/2006
From:	Burgos B, Chapman D, Conermann J Westinghouse
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
WCAP-16526-NP, Rev 0
Download: ML061030088 (232)

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Westinghouse Non-Proprietary Class 3 WCAF'-16526-NP March 2006 RevisiDn 0 Analysis of Capsule V from FPL Energy -

Seabrook Unit 1 Reactor Vessel Radiation Surveillance Program

• Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16526-NP, Revision 0 Analysis of Capsule V from FPL Energy - Seabrook Unit 1 Reactor Vessel Radiation Surveillance Program B.N. Burgos D.M. Chapman J. Conermann March 2006 Approved:

• ElectronicallyApproved J.S. Carlson, Manager Primary Component Asset Management

• Electronically Approved Records are Authenticated in the Electronic Document Management System Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355 02006 Westinghouse Electric Company LLC All Rights Reserved

iii TABLE OF CONTENTS LIST OF TABLES .................. iv LIST OF FIGURES .................. vi PREFAC .................... ix EXECUTIVE

SUMMARY

................... x I

SUMMARY

OF RESULTS. 1-1 2 INTRODUCTION . 2-1 3 BACKGROUND :3-1 4 DESCRIPTION OF PROGRAM...................................................................................................4-1 5 TESTING OF SPECIMENS FROM CAPSULE V :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 COMPACT TENSION SPECIMEN TESTS 5 5-5 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY . .i- I

6.1 INTRODUCTION

.6- I 6.2 DISCRETE ORDINATES ANALYSIS .6-2 6.3 NEUTRON DOSIMETRY .6-4 6.4 CALCULATIONAL UNCERTAINTIES .6-5 7 SUJRVEILLANCE CAPSULE WITHDRAWAL SCHEDULE .'7-8 REFERENCES .81 APPENDI]X A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS CREDIBILITY .. A-0 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS .11-0 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD. C-0 APPENDIX D SEABROOK UNIT I SURVEILLANCE PROGRAM CREDIBILITY EVALUATION .D-0

iv LIST OF TABLES Table 4-1 Chemical Composition (wt %) of the Seabrook Unit I Reactor Vessel Surveillance Materials (Unirradiated) ........................ 4-3 Table 4-2 Heat Treatment History of the Seabrook Unit I Reactor Vessel and Surveillance Materials .... 4-4 Table 5-1 Charpy V-Notch Data for the Seabrook Unit I Lower Shell Plate R1808-3 Irradiated to a Fluence of 2.669 x 1019 n/cm 2 (E > 1.0 MeV)

(Longitudinal Orientation) ................ 5-6 Table 5-2 Charpy V-Notch Data for the Seabrook Unit 1 Lower Shell Plate RI 808-3 Irradiated to a Fluence of 2.669 x 1019 n/cm2 (E > 1.0 MeV)

(Transverse Orientation) ................. :5-7 Table 5-3 Charpy V-notch Data for the Seabrook Unit I Surveillance Weld Material Irradiated to a Fluence of 2.669 x 1019 n/cm2 (E> 1.0 MWV) ............................................. ;-8 Table 5-4 Charpy V-notch Data for the Seabrook Unit I Heat Affected Zone Material Irradiated to a Fluence of 2.669 x 1019 n/cm2 (E> 1.0 MeV) ........................................ '-9 Table 5-5 Instrumented Charpy Impact Test Results for the Seabrook Unit 1 Lower Shell Plate R1808-3 Irradiated to a Fluence of 2.669 x 10'9 n/cm 2 (E> 1.0 MeV)

(Longitudinal Orientation) ................ 5-10 Table 5-6 Instrumented Charpy Impact Test Results for the Seabrook Unit I Lower Shell Plate RI 808-3 Irradiated to a Fluence of 2.669 x 1019 n/cm 2 (E> 1.0 MeV)

(Transverse Orientation) ............... 5-11 Table 5-7 Instrumented Charpy Impact Test Results for the Seabrook Unit I Surveillance Weld Metal Irradiated to a Fluence of 2.669 x 10'9 n/cm 2 (E> 1.0 MeV) ..................... 5-12 Table 5-8 Instrumented Charpy Impact Test Results for the Seabrook Unit I Heat Affected Zone Material Irradiated to a Fluence of 2.669 x 10'9 n/cm 2 (E> I.OMeV) ........................... 5-13 Table 5-9 Effect of Irradiation to 2.669 x 10i9 n/cm2 (E> 1.0 MeV) on the Capsule V Toughness Properties of the Seabrook Unit I Reactor Vessel Surveillance Materials .................... 5-14 Table 5-10 Comparison of the Seabrook Unit I 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 Seabrook Unit I Capsule V Reactor Vessel Surveillance Materials Irradiated to 2.669 x 10i 9 n/cm 2 (E> 1.0MeV) .......................... 5-16 Table 6-1 Calculated Neutron Flux At The Surveillance Capsule Center ..................................... 6-7 Table 6-2 Calculated Neutron Fluence At The Surveillance Capsule Center ...... ......................... 6-8 Table 6-3 Calculated Iron Atom Displacement Rate At The Surveillance Capsule Center ........... 6-9 Table 6-4 Calculated Iron Atom Displacements At The Surveillance Capsule Center...............6-10 Table 6-5 Calculated Azimuthal Variation Of Neutron Flux At The Reactor Vessel Clad/Base Metal Interface . 6-11 Table 6-6 Calculated Azimuthal Variation Of Neutron Fluence At The Reactor Vessel Clad/Base Metal Interface . 6-12 Table 6-7 Calculated Azimuthal Variation Of Iron Atom Displacement Rates At The Reactor Vessel Clad/Base Metal Interface . 6-13 Table 6-8 Calculated Azimuthal Variation Of Iron Atom Displacements At The Reactor Vessel Clad/Base Metal Interface At The Reactor Vessel Clad/Base Metal Interface . 6-14 Table 6-9 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV) Within The Reactor Vessel Wall . 6-15 Table 6-10 Relative Radial Distribution Of Iron Atom Displacements (dpa) Within The Reactor Vessel Wall 6-16 Table 6-11 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Seabrook Station ........................................ 6- 17 Table 6-12 Calculated Surveillance Capsule Lead Factors ........................................ 6-17 Table 7-1 Recommended Surveillance Capsule Withdrawal Schedule ........................................ 7-1

vi

\\ LIST OF FIGURES Figure 4-1 Arrangement of Surveillance Capsules in the Seabrook Unit 1 Reactor Vessel ............... 4-5 Figure 4-2 Capsule V Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeters .4-6 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Seabrook Unit I Reactor Vessel Lower Shell Plate R1808-3 (Longitudinal Orientation) . 5-17 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Seabrook Unit I Reactor Vessel Lower Shell Plate RI 808-3 (Longitudinal Orientation) . 5-18 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Longitudinal Orientation) . 54-19 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Transverse Orientation) . 5--20 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Transverse Orientation) . 5-*21 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Seabrook Unit I Reactor Vessel Lower Shell Plate R1808-3 (Transverse Orientation) . 5-22 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for Seabrook Unit 1 Reactor Vessel Weld Metal . 5-23 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for Seabrook Unit 1 Reactor Vessel Weld Metal . 5-24 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for Seabrook Unit 1 Reactor Vessel Weld Metal . 5-25 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for Seabrook Unit I Reactor Vessel Heat Affected Zone Material . 5-26 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for Seabrook Unit 1 Reactor Vessel Intermediate Heat Affected Zone Material . 5-27 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for Seabrook Unit I Reactor Vessel Heat Affected Zone Material . 5-28 Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Longitudinal Orientation) . 5-29

vii LIST OF FIGURES (Cont.)

Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Seabrook Unit I Reactor Vessel Lower Shell Plate R1808-3 (Transverse Orientation) ......................................... 5-30 Figure 5-15 Charpy Impact Specimen Fracture Surfaces for Seabrook Unit I Reactor Vessel Weld Metal .......................................................... 5-31 Figure 5-16 Charpy Impact Specimen Fracture Surfaces for Seabrook Unit I Reactor Vessel Heat Affected Zone Material .......................................................... 5-32 Figure 5-17 Tensile Properties for Seabrook Unit 1 Reactor Vessel Lower Shell Plate RI 808-3 (Longitudinal Orientation) .......................................................... 5-33 Figure 5-18 Tensile Properties for Seabrook Unit 1 Reactor Vessel Lower Shell Plate RI 808-3 (Transverse Orientation) .......................................................... 5-34 Figure 5-19 Tensile Properties for Seabrook Unit I Reactor Vessel Weld Metal .............................. 5-35 Figure 5-20 Fractured Tensile Specimens from Seabrook Unit I Reactor Vessel Lower Shell Plate RI 808-3 (Longitudinal Orientation) .......................................................... 5-36 Figure 5-21 Fractured Tensile Specimens from Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Transverse Orientation) .......................................................... 5-37 Figure 5-22 Fractured Tensile Specimens from Seabrook Unit I Reactor Vessel Weld Metal .......... 5-38 Figure 5-23 Engineering Stress-Strain Curves for Lower Shell Plate R1808-3 Tensile Specimens KL-6 and KL-5 (Longitudinal Orientation) ................................................. 5-39 Figure 5-24 Engineering Stress-Strain Curves for Lower Shell Plate RI 808-3 Tensile Specimens KL-4 (Longitudinal Orientation) .......................................................... 5-40 Figure 5-25 Engineering Stress-Strain Curves for Lower Shell Plate R1808-3 Tensile Specimens KT-6 and KT-5 (Transverse Orientation) ..................................................... 5-41 Figure 5-26 Engineering Stress-Strain Curves for Lower Shell Plate R1808-3 Tensile Specimens KT-4 (Transverse Orientation) .......................................................... 5-42 Figure 5-27 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens KW-6 and KW-5 .......................................................... 5-43 Figure 5-28 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens KW4 ................ 5-44

viii LIST OF FIGURES (Cont.)

Figure 6-1. Seabrook Station rO Reactor Geometry with a Dual Capsule at the Core Midplane ...6--18 Figure 6-,2 Seabrook Station rO Reactor Geometry with a Single Capsule at the Core Midplane .6-49 Figure 6-3 Seabrook Station rO Reactor Geometry with a No Capsule at the Core Midplane .... 6--20 Figure 6-4 Seabrook Station rz Reactor Geometry with Neutron Pad ............................................

6-21

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

Reviewer:

Sections 1.through 5, 7, 8, Appendices B, C and D F.C. GifM

• ElectronicallyApproved Section 6 and Appendix A S. L. Anderson

• ElectronicallyApproved

• Electronically Approved Records are Authenticated in the Electronic Document Management System

x EXECUTIVE

SUMMARY

The purpose of this report is to document the results of the testing of surveillance Capsule V from FPL Energy - Seabrook Unit 1. Capsule V was removed at 12.39 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 V received a fluence of 2.669 x 1019 n/cm 2 after irradiation to 12.39 EFPY. The peak clad/base metal interface vessel fluence after 12.39 EFPY of plant operation was 7.05 x 1018 n/cm 2 .

This evaluation lead to the following conclusions: 1)All the surveillance materials tested resulted in measured 30 ft-lb shift in transition temperature values that are within the Regulatory Guide 1.99, Revision 2 [Ref. 1], predictions. 2). The measured percent decrease in upper shelf energy for all the surveillance materials contained in the Seabrook Unit 1 surveillance program are less than the Regulatory Guide 1.99, Revision 2 predictions. 3) 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 (32 EFPY) as required by IOCFR50, Appendix G [Ref. 2].

4) The Seabrook Unit I surveillance data from the lower shell plate R1808-3 and the surveillance weld metal were 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 V, the third capsule removed and tested from the Seabrook Unit I reactor pressure vessel, led to the following conclusions:

• The Charpy V-notch data presented in WCAP-101 0 [Ref. 3] and Duke Engineering Services Report DES-NFQA-98-01 [Ref. 4] were based on either hand-fit curves or computer fitted tanh Charpy curves. For consistency, 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 5.0.2, Charpy V-notch plots and the program input data.

• Capsule V received an average fast neutron fluence (E> 1.0 MeV) of 2.669 x 1019 n/cm 2 after 12.39 effective full power years (EFPY) of plant operation.

• Irradiation of the reactor vessel lower shell plate RI 808-3 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 32.30 F and an irradiated 50 ft-lb transition temperature of 67.10 F. This results in a 30 ft-lb transition temperature increase of 60.40 F and a 50 ft-lb transition temperature increase of 70.90 F for the longitudinal oriented specimens. See Table 5-9.

• In.adiation of the reactor vessel lower shell plate R1808-3 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 70.30 F and an irradiated 50 ft-lb transition temperature of 126.3°F. This results in a 30 ft-lb transition temperature increase of 61 .30 F and a 50 ft-lb transition temperature increase of 69.80 F for the transverse oriented specimens. See Table 5-9.

• In-adiation of the weld metal (heat number 4P6052) Charpy specimens resulted in an irradiated 3C ft-lb transition temperature of-32.90 F and an irradiated 50 ft-lb transition temperature of

-13.00 F. This results in a 30 ft-lb transition temperature increase of 41.70 F and a 50 ft-lb transition temperature increase of 36.90 F. See Table 5-9.

• The average upper shelf energy of the lower shell plate R1808-3 (longitudinal orientation) resulted in an average energy decrease of 13 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 112 ft-lb for the longitudinal oriented specimens. See Table 5-9.

• The average upper shelf energy of the lower shell plate R1808-3 (transverse orientation) resulteI in an average energy decrease of 10 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 69 ft-lb for the transverse oriented specimens. See Table 5-9.

• The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of 4 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 156 ft-lb for the weld metal specimens. See Table 5-9.

Summary ofResults

1-2

• A comparison, as presented in Table 5-10, of the Seabrook Unit 1 reactor vessel surveillance material program test results (from Capsules U, Y and V) with the Regulatory Guide 1.99, Revision 2 [Ref. 1] predictions led to the following conclusions:

- All six measured 30 ft-lb shifts in transition temperature values of the lower shell plate RI 808-3 (longitudinal & transverse) are within a + I a scatter band of the Regulatory Guide 1.99, Revision 2, predictions.

- All three measured 30 ft-lb shifts in transition temperature value of the weld metal are within a +/- Ia scatter band of the Regulatory Guide 1.99, Revision 2, predictions.

- The Seabrook Unit 1 surveillance data from the lower shell plate R1808-3 and the surveillance weld metal were found to be credible per the criteria in Regulatory Guide 1.99, Revision 2. All measured 30 ft-lb shifts in transition temperature values fall within the +/- IC scatter band of the predicted shift, as shown in Table 5-10. This complete credibility evaluation can be found in Appendix D

- The measured percent decrease in upper shelf energy for all the surveillance materials contained in the Seabrook Unit I surveillance program are less than 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 (32 EFPY) as required by IOCFR5O, Appendix G [Ref. 2].

• The calculated end-of-license (32 EFPY) neutron fluence (E> 1.0 MeV) at the core midplane for the Seabrook Unit I 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* = 1.72 x 109 n/cm2 Vessel 1/4 thickness = 1.02 x I0'9 n/cm2 Vessel 3/4 thickness = 3.64 x IO" n/cm 2

• Clad/base metal interface. (From Table 6-5)

Summary of Results

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

The surveillance program for the Seabrook Unit I 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-I01 10, "Public Service Company of New Hampshire Seabrook Station Unit No. I Reactor Vessel Radiation Surveillance Program." [Ref. 3]. The surveillance program was planned to cover the 40-year design lifi, of the reactor pressure vessel and was based on ASTM El 85-79, "Recommended Practice for Surveillance Tests on Structural Materials for Nuclear Reactors" [Ref. 5]. Capsule V was removed from the reactor after 12.39 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 V' removed from the Seabrook Unit I 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 beltline 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 lcw alloy, ferritic pressure vessel steels such as SA533 Grade B-i (base material of the Seabrook Unit I 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 XI of the ASME Boiler and Pressure Vessel Code [Ref. 6]. 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 (NDTT per ASTM E-208 [Ref. 7]) or the temperature 60'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 (KT, curve) which appears in Appendix G to the ASME Code [Ref. 6]. The Kic 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 K1, 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.

RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects 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 Seabrook Unit I reactor vessel radiation surveillance program [Ref. 3], 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 KI, 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 sunreillance capsules for monitoring the effects of neutron exposure on the Seabrook Unit I reactor pressure vNessel 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 V was removed after 12.39 effective full power years (EFPY) of plant operation. This capsule contained Charpy V-notch, tensile, and Compact Tension (CT) specimens as shown in Figure 4-2, which were mad- from lower shell plate R1808-3 (longitudinal and transverse) and the submerged arc weld metal.

Test material obtained from the lower shell plate (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 lower shell plate R1808-3 were machined in the longitudinal (longitudinal axis of the specimen parallel to the major working direction) and transverse (longitudinal axis of the specimen perpendicular to the major working direction) orientations. 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 lower shell plate R1808-3 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.

The chem:.cal 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 uprate analysis report, WC AP- 15745 [Ref. 8] and from the unirradiated surveillance program report, WCAP- IO 10

[Ref. 3].

Description of Program

4-2 Capsule V contained dosimeter wires of pure iron, copper, nickel, and aluminum-0.15 weight percent cobalt (cadmium-shielded and unshielded) and cadmium-shielded NP237 and U238 which will measure the integrated flux at specific neutron energy levels.

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: 579IF (3041C) 1.5% Ag, 1.0% Sn, 97.5% Pb Melting Point: 590'F (310C)

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

Description of Program

4-3 Table 4-1 Chemical Composition (wt%) of the Seabrook Unit 1 Reactor Vessel Surveillance Materials (Unirradiated) l Lower Shell Plate Element R1808-3(a) Weld Metal(b)

C 0.20 0.15 Mn 1.45 1.28 P 0.007 0.009 S 0.010 0.007 Si 0.24 0.14 Ni 0.59 (d) 0.049(c)

Cr 0.06 0.03 Mo 0.55 0.52 Cu 0 .0 7 (d) 0.047(c)

Al 0.028 0.003 V 0.003 0.003 Sn 0.011 0.004 Cb <0.01 <0.01 Zr <0.001 <0.001 Ti <0.01 <0.01 Notes:

(a) The source of the plate data is the chemistry presented in WCAP-IO 1 0 [Ref. 3], unless otherwise noted.

(b) Weld wire Heat Number 4P6052, Flux Type Linde 0091, and Flux Lot Number 0145. The weld material in the Seabrook Unit I surveillance program was made of the same wire and flux as all the reactor vessel beltline weld seams. The weld chemistry is the average of the CE and Westinghouse analyses from Reference 3, unless otherwise noted.

(c) Best Estimate Data from CE NPSD-1039,"Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," Revision 2 [Ref. 9].

(d) Average of original Surveillance test [Ref. 3] and the Lukens CMTR [Ref. 10].

Description of Program

4-4 Table 4-2 Heat Treatment History of the Seabrook Unit 1 Reactor Vessel and Surveillance Materials(a)

Material Temperature (OF) Time Coolant Austenitizing: 4 hrs. Water-Quench 1550 to 1650 Lower Shell Plates Tempered: 4 hrs. Air-cooled R1808-1 3 1225 Stress Relief: 16 hrs. Furnace Cooled 1125 to 1175 Austenitizing: 4 hrs. Water-Quench 1550 to 1650 Intermediate Shell Plate Tempered: 4 hrs.

R1806-1 3 1225 Stress Relief: 16.5 hrs Furnace Cooled 1125 to 1175 Lower Shell Plate Longitudinal Stress Relief: 16 Furnace Cooled Seam Welds 1125 to 1175 Intermediate Shell Longitudinal Stress Relief: 16.5 Furnace Cooled Seam Welds 1125 to 1175 Intermediate to Lower Shell Stress Relief: 12.5 hrs.Furnace Cooled Girth Seam Weld 1125 to 1175 1 Surveillance Program Test Material Surveillance Program Test Plate Austenitizing: 4 hrs. Water-Quench R1808-3 1550 to 1650 Tempered: 4 hrs. Air-cooled 1225 Stress Relief: 16.5 hrs Furnace Cooled 1125 to 1175 Weldment Stres Reli: 12.5 hrs. Furnace Cooled

__ __ __ __ _ __ __ _ _ _ _ _ _ _ _ _ _1 12 5 to _1175_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Notes:

(a) Data obtained from WCAP-101 10 [Ref. 3] and duplicated herein for completeness.

Description of Program

4-5 0o REACTOR VESSEL 180I PLA VIEW VESSEL WALL 11 CAPSULE c /- ASSEMBLY a1111111111I1 NEUmvoN PAI 3 CORE DAME!

FLEVATION VEW Figure 4-1. Arrangement of Surveillance Capsules in the Seabrook Unit 1 Reactor Vessel Description of Program

4-6 LEGEND: KL - LOWER SHELL PLATE Rl 808-3 (LONGITUDINAL)

KT - LOWVER SHELL PLATE R1808-3 (TANGENTIAL)

KW - WVELD METAL (HEAT # 4P6052)

KH- HEAT AFFECTED ZONE Tensile C Charpy Charpy Cor act KW27 _KH27 KW24 _KH24 KW26 KH26 KW23 [K1H23 KL8 K l KW4 I I KW25 I KH25 KW22 KH22 I TOP OF VESSEL CENTER Dosimeter A Charpy Dosimeter Tensile Chary Charpy Cha Chary IKW18 KH18 [ KT30 I KL30 KT27 KL27 KT24 KL24 KT2i KL21 KW17 KH17 472 KL5 KT29 I I KL29 KT26 KL26 KT23 KL23 KT20 KL20 IKW16 I KH16 KL I I KT28 l KL28 II KT25 l KL25 KT22 I L2 KT19 I IKL19I CENTER Dosimeter B Dosimeters A and B I---------

t Tensile - Spacer lKT6 I I , ----- _--

--- --_! I-----4 T5 llKT5 Co Wire- - Co-Cd Wire KT4 I Fe Wire - - Ni Wire BOTTOM OF VESSEL __

Cu Wire- ___--- ------ 4 7 579°F (Dosimeter A)

Dosimeter A 590°F (Dosimeter B)

Figure 4-2 Capsule V Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeters Description of Program

5-1 5 TESTING OF SPECIMENS FROM CAPSULE V 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

[Ref. 2], ASTM Specification E185-82 [Ref. 11], and Westinghouse Procedure RMF 8402, Revision 2

[Ref. 12] as detailed by Westinghouse RMF Procedures 8102, Revision 3 [Ref. 13], and 8103, Revision 2

[Ref.14].

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-101 0 [Ref.

3]. No discrepancies were found.

Examination of the two low-melting point 5790 F (3040 C) and 590'F (310C) cutectic 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 (304'C).

The Charpy impact tests were performed per ASTM Specification E23-02a [Ref. 15] and RMF Procedure 8103 [Ref. 14] on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy impact test machine i; instrumented with an Instron Impulse instrumentation system, feeding information into an IBM compatible computer. With this system, load-time and energy-time signals can be recorded 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 (tm) 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 (ay) was calculated from the three-point bend formula having the following expression:

aY = (PGY *L) /[B *(V- a)2

• C] li) 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 (4), 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 4 =

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

Testing of specimens from Capsule V

5-2 c-r = (PGY *L) /[B *(V_ a)2

• 1.21]=(3.305 *PGY*W)/[B *(Wa)2] (2)

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

cr = 33.3 *PGY (3) where cry is in units of psi and PGY 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 5, 6, and 7 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 [Ref. 15] and A370-97a [Ref. 16]. 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 E8-04 [Ref. 17] and E21-03 [Ref. 18], and Procedure RMF 8102 [Ref. 13]. 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 E83-93

[Ref. 19].

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 V

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 V, which received a fluence of 2.669 x 10'9 n/cm 2 (E> 1.0 MeV) in 12.39 EFPY of operation, are presented.

in Tables 5-1 through 5-12 and are compared with unirradiated results [Ref. 3] as shown in Figures 5-1 through 5-12.

The transition temperature increases and upper shelf energy decreases for the Capsule V materials are summarized in Table 5-9 and led to the following results:

Irradiation of the reactor vessel lower shell plate R1808-3 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 32.37F and an irradiated 50 ft-lb transition temperature of 67.17F. This results in a 30 ft-lb transition temperature increase of 60.40 F and a 50 ft-lb transition temperature increase of 70.90 F for the longitudinal oriented specimens. See Table 5-9.

Irradiation of the reactor vessel lower shell plate RI 808-3 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 70.30 F and an irradiated 50 ft-lb transition temperature of 126.30 F. This results in a 30 ft-lb transition temperature increase of 61.30 F and a 50 ft-lb transition temperature increase of 69.80 F for the transverse oriented specimens. See Table 5-9.

Irradiation of the weld metal (heat number 4P6052) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -32.9 0 F and an irradiated 50 ft-lb transition temperature of-13.00 F. This results in a 30 ft-lb transition temperature increase of 41.70 F and a 50 ft-lb transition temperature increase of 36.90 F. See Table 5-9.

Irradiation of the reactor vessel heat affected zone Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of-54.50 F and an irradiated 50 ft-lb transition temperature of-25.50 F. This results in a 30 ft-lb transition temperature increase of 139.2 0 F and a 50 ft-lb transition temperature increase of 114.7 0 F for the HAZ specimens. See Table 5-9.

The average upper shelf energy of the lower shell plate R1808-3 (longitudinal orientation) resulted in an average energy decrease of 13 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 112 ft-lb for the longitudinal oriented specimens. See Table 5-9.

The average upper shelf energy of the lower shell plate R1808-3 (transverse orientation) resulted in an average energy decrease of 10 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 69 ft-lb for the transverse oriented specimens. See Table 5-9.

The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease o:F4 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 156 ft-lb for the weld metal specimens. See Table 5-9.

Testing of Specimens from Capsule V

5-4 The average upper shelf energy of the heat affected zone material resulted in an average energy decrease of 0 ft-lb after irradiation. An irradiated average upper shelf energy of 130 ft-lb for the HAZ specimens was measured. See Table 5-9.

A comparison, as presented in Table 5-10, of the Seabrook Unit 1 reactor vessel surveillance material test results for Capsules U, Y and V with the Regulatory Guide 1.99, Revision 2 [Ref. I] predictions led to the following conclusions:

- All six measured 30 ft-lb shifts in transition temperature values of the lower shell plate R1808-3 (longitudinal & transverse) are within a +/- Ia scatter band of the Regulatory Guide 1.99, Revision 2, predictions.

- All three measured 30 ft-lb shifts in transition temperature value of the weld metal are within a 4t Ica scatter band of the Regulatory Guide 1.99, Revision 2, predictions.

- The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules U, Y and V contained in the Seabrook Unit I surveillance program arc less than the Regulatory Guide 1.99, Revision 2 predictions.

- The Seabrook Unit 1 surveillance data from the lower shell plate Rl 808-3 and the surveillance weld metal were found to be credible per the criteria in Regulatory Guide 1.99, Revision 2. All measured 30 ft-lb shifts in transition temperature values fall within the + la scatter band of the predicted shift, as shown in Table 5-10. This complete credibility evaluation can be found in Appendix D.

All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and arc predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the extended life of the vessel (32 EFPY) as required by 10CFR50, Appendix G [Ref. 2].

The fracture appearance of each irradiated Charpy specimen from the various surveillance Capsule V 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 (Note that some plots arc not available for the specimens identified in Tables 5-5 through 5-8 due to a computer malfunction).

The Charpy V-notch data presented in WCAP-I01 0 [Ref. 3] and Duke Engineering Services Report DES-NFQA-98-01 [Ref. 4] were based on various Charpy tanh curve fitting techniques, including hand fits and computer fits. 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 5.0.2, Charpy V-notch plots and the program input data.

Testing of Specimens from Capsule V

5-5 5.3 TENSILE TEST RESULTS The results of the tensile tests performed on the various materials contained in Capsule V irradiated to 2.669 x 1I)19 n/cm 2 (E> 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated results

[Ref. 3] as shown in Figures 5-17 through 5-19.

The results of the tensile tests performed on the Lower Shell Plate R1808-3 (longitudinal orientation) indicated that irradiation to 2.669 x 10' 9 n/cm 2 (E> 1.0 MeV) caused approximately a 5 to 7 ksi increase in the 0.2 percent offset yield strength and approximately a 3 to 6 ksi increase in the ultimate tensile strength when compared to unirradiated data [Ref. 3]. See Figure 5-17.

The results of the tensile tests performed on the Lower Shell Plate R1808-3 (transverse orientation) indicated 'that irradiation to 2.669 x 1019 n/cm2 (E> 1.0 MeV) caused approximately a 5 to 6 ksi increase in the 0.2 percent offset yield strength and approximately a 4 to 6 ksi increase in the ultimate tensile strength when compared to unirradiated data [Ref. 3]. See Figure 5-18.

The results of the tensile tests performed on the weld metal indicated that irradiation to 2.669 x 10'9 n/cm 2 (E> 1.0 MeV) caused approximately a 3 to 4 ksi increase in the 0.2 percent offset yield strength and approximately a 3 ksi increase in the ultimate tensile strength when compared to unirradiated data [Ref.

3]. See Figure 5-19.

The fractured tensile specimens for the lower shell plate R1808-3 (longitudinal and transverse orientations) and the weld metal are shown in Figures 5-20 through 5-22. The engineering stress-strain curves for the tensile tests are shown in Figures 5-23 through 5-28.

5.4 COMPACT TENSION SPECIMEN TESTS Per the surveillance capsule testing contract, the Compact Tension Specimens were not tested and are being stored at the Westinghouse Science and Technology Center Hot Cell facility.

Testing of Specimens from Capsule V

5-6 Table 5-1 Charpy V-notch Data for the Seabrook Unit 1 Lower Shell Plate R1808-3 Irradiated to a Fluence of 2.669 x 1019 n/cm 2 (E> 1.0 MeV) (Longitudinal Orientation)

Sample Temperature Impact Energy Lateral Expansion Shear Number OF 0C ft-lbs Joules mils mm  %

KL21 -50 -46 7 9 6 0.15 5 KLI9 0 -18 12 16 13 0.33 10 KL16 25 -4 30 41 26 0.66 15 KL20 50 10 48 65 35 0.89 20 KL25 50 10 40 54 29 0.74 20 KL29 75 24 41 56 33 0.84 20 KL26 75 24 65 88 40 1.02 30 KL24 100 38 67 j 91 46 1.17 60 KL30 125 52 81 110 62 1.57 70 KL27 150 66 98 133 73 1.85 95 KL17 175 79 102 138 74 1.88 100 KL22 225 107 111 151 72 1.83 100 KL28 250 121 123 167 80 2.03 100 KL23 300 149 122 ] 165 77 1.96 100 KLI8 350 177 114 155 76 1.93 100 Testing of Specimens from Capsule V

5-7 Table 5-2 Charpy V-notch Data for the Seabrook Unit 1 Lower Shell Plate R1808-3 Irradiated to a Fluence of 2.669 x 10"9 n/cm2 (E> 1.0 MeV) (Transverse Orientation)

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

KT28 -50 -46 3 4 0 0.00 5 KT17 0 -18 5 7 4 0.10 15 KT30 25 -4 17 23 12 0.30 15 KT19 50 10 22 30 19 0.48 20 KT16(a) 75 24 - - - - -

KT27 75 24 35 47 28 0.71 40 KT20 75 24 32 43 27 0.69 40 KT25 100 38 49 66 39 0.99 55 KT18 125 52 39 53 34 0.86 70 KT21 125 52 50 68 40 1.02 60 KT23 150 66 57 77 48 1.22 90 KT26 200 93 65 88 57 1.45 100 KT22 250 121 63 85 53 1.35 100 KT29 300 149 68 92 51 1.30 100 KT24 350 177 80 108 65 1.65 100

a. Invalid Test.

Testing of Specimens from Capsule V

5-8 Table 5-3 Charpy V-notch Data for the Seabrook Unit 1 Surveillance Weld Metal Irradiated to a Fluence of 2.669 x 10'9 n/cm 2 (E> 1.0 MeV)

Sample Temperature Impact Energy Lateral Expansion Shear Number OF ft-lbs Joules mils mml KW20 -90 -68 6 8 2 0.05 5 KW23 -50 -46 9 12 7 0.18 15 KW30 -35 -37 13 18 9 0.23 20 KW24 -25 -32 74 100 50 1.27 40 KW16 -25 -32 40 54 30 0.76 30 KW27 0 -18 70 95 49 1.24 60 KW29 25 -4 58 79 42 1.07 70 KW18 25 -4 108 146 73 1.85 80 KW28 50 10 124 168 80 2.03 90 KW22 75 24 167 226 93 2.36 100 KW25 75 24 170 231 86 2.18 98 KW26 100 38 142 193 92 2.34 98 KW17 200 93 150 203 85 2.16 100 KW19 250 121 150 203 89 2.26 100 KW21 300 149 155 210 80 2.03 100 Testing of Specimens from Capsule V

5-9 Table 54 Charpy V-notch Data for the Seabrook Unit 1 Heat Affected Zone Material Irradiated to a Fluence of 2.669 x 10'9 n/cm2 (E> 1.0 MeV)

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

KH20 -90 -68 36 49 28 0.71 10 KH24 -75 -59 23 31 15 0.38 20 KH16 -50 -46 28 38 25 0.64 25 KH23 -25 -32 36 49 27 0.69 45 KH22 0 -18 73 99 45 1.14 55 KH25 25 -4 89 121 60 1.52 85 KH30 25 -4 107 145 67 1.70 75 KH21 50 10 76 103 53 1.35 80 KH27 75 24 150 203 81 2.06 100 KH18 100 38 123 167 79 2.01 100 KH28 125 52 137 186 79 2.01 100 KH29 150 66 106 144 68 1.73 100 KH19 200 93 121 164 70 1.78 100 KH26 250 121 137 186 79 2.01 100 KH17 300 149 136 184 78 1.98 100 Testing of Specimens from Capsule V

5-10 Table 5-5 Instrumented Charpy Impact Test Results for the Seabrook Unit I Lower Shell Plate R1808-3 Irradiated to a Fluence of 2.669 x 109'n/cm 2 (E>1.0 MeV) (Longitudinal Orientation)

Normalized Energies Charpy (ft-lb/in2) Yield Time to Time to Fast Test Energy Load Yield Max. Max. Fract. Arrest Yield Flow Sample Temp. ED Charpy Max. Prop. P t GLoad tN1 Load PF Load PA Stress Stress No. (OF) (ft-lb) ED/A Es /A EIA (lb) (msec) Prq (lb) (msec) (lb) (lb) cy (ksi) (ksi)

KL21 -50 7 56 26 30 2578 0.13 2736 0.15 2736 0 86 88 KLI9 0 12 97 37 60 2739 0.14 3256 0.17 3233 269 91 100 KLI6 25 30 242 153 89 2530 0.13 4273 0.41 4268 61 84 113 KL20 50 48 387 292 95 2247 0.13 4474 0.67 4469 220 75 112 KL25 50 40 322 206 116 2434 0.14 4273 0.52 4262 219 81 112 KL29 75 41 330 174 156 2414 0.14 4303 0.45 4280 776 80 112 KL26 75 (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a)

KL24 100 67 540 295 245 2378 0.14 4444 0.67 4230 1237 79 114 KL30 125 81 653 289 364 2624 0.14 4185 0.67 3098 901 87 113 KL27 150 98 790 291 499 2617 0.13 4251 0.67 3017 1680 87 114 KLI7 175 102 822 279 543 2733 0.16 4091 0.67 n/a n/a 91 114 KL22 225 III 894 277 617 2360 0.14 4072 0.68 n/a n/a 79 107 KL28 250 (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a)

KL23 300 122 983 283 700 2391 0.15 4167 0.68 n/a n/a 80 109 KL18 350 114 919 282 637 2747 0.16 4092 0.68 n/a n/a 91 114

a. Computer malfunction, instrumented data not available.

Testing of Specimens from Capsule V

5-11 Table 5-6 Instrumented Charpy Impact Test Results for the Seabrook Unit 1 Lower Shell Plate R1808-3 Irradiated to a Fluence of 2.669 x 10'9 n/cm2 (E>1.0 MeV) (Transverse Orientation) _

-_. . - I I Normalized Energies Charpy (ft-lb/in2 ) Yield Time to Time to Fast Test Energy Load Yield Max. Max. Fract. Arrest Yield Flow Sample Temp. Charpy Max. Prop.

ED PGY tGY Load tN Load PF Load PA Stress Stress No. (OF) (ft-lb) ED/A Ep/A (lb) (msec) PM (lb) (msec) (lb) (lb) cy (ksi) (ksi)

KT28 -50 3 24 9 15 1132 0.09 1221 0.1 1221 0 38 39 KT17 0 5 40 9 31 1132 0.09 1246 0.11 1246 0 38 40 KT30 25 (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a)

KT19 50 22 177 98 80 2219 0.13 3948 0.32 3948 243 74 103 KT16 75 (b) (b) (b) (b) (b) (b) (b) (b) (b) (b) (b) (b)

KT27 75 35 282 135 147 2381 0.14 4290 0.38 4280 1307 79 III KT20 75 (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a)

KT25 100 49 395 213 182 2483 0.13 4178 0.53 4016 932 83 II KT18 125 39 314 125 189 2652 0.26 3662 0.42 3647 2218 88 105 KT21 125 (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a)

KT23 150 57 459 175 284 2490 0.14 3912 0.47 3410 2198 83 107 KT26 200 65 524 184 340 2304 0.13 3816 0.5 n/a n/a 77 102 KT22 250 63 508 187 321 2578 0.14 3835 0.49 n/a n/a 86 107 KT29 300 68 548 197 351 2508 0.13 3877 0.52 n/a n/a 84 106 KT24 350 80 645 197 447 2487 0.14 3971 0.51 n/a n/a 83 108

a. Computer malfunction, instrumented data not available.

b. Invalid Test Testing of Specimens from Capsule V

5-12 Table 5-7 Instrumented Charpy Impact Test Results for the Seabrook Unit 1 Surveillance Weld Metal Irradiated to a Fluence of 2.669 x 1019 n/cm 2 (E>I.O MeV)

Normalized Energies Charpy (ft-lb/in 2 ) Yield Time to Time to Fast Test EnergY Load Yield Max. Max. Fract. Arrest Yield Flow Sample Temp. En CharPY Max. PrOP. PGY tGY Load tN, Load PF Load Stress Stress No. (OF) (ft-lb) ED/A EM/A EP/A (lb) (msec) PM (lb) (msec) (lb) PA (lb) cry (ksi) (ksi)

KW20 -90 6 48 28 20 2089 0.11 2384 0.16 2384 383 70 74 KW23 -50 9 73 30 43 2278 0.13 2539 0.16 2537 453 76 80 KW30 -35 13 105 34 71 2552 0.13 3019 0.17 3006 651 85 93 KW24 -25 74 596 326 270 3158 0.14 4612 0.67 4278 800 105 129 KW16 -25 40 322 210 112 3235 0.14 4475 0.48 4470 643 108 128 KW27 0 70 564 231 333 2732 0.14 4535 0.52 4342 1465 91 121 KW29 25 (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a)

KWI8 25 108 870 317 553 2530 0.13 4430 0.7 3416 1604 84 116 KW28 50 124 999 306 693 2574 0.14 4411 0.68 3239 2283 86 116 KWV22 75 167 1346 302 1043 2777 0.16 4545 0.68 n/a n/a 92 122 KW25 75 (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a)

KW26 100 142 1144 302 842 2876 0.16 4324 0.68 1692 1344 96 120 KW1 7 200 (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a)

KWI9 250 150 1209 286 923 2111 0.12 4218 0.68 n/a n/a 70 105 KW21 300 (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a)

a. Computer malfunction, instrumented data not available.

Testing of Specimens from Capsule V

5-13 Instrumented Charpy Impact Test Results for the Seabrook Unit 1 Heat Affected Zone Material Irradiated to a Fluence of Table

2.669 x 1019n/cm 2 (E>1.0 MeV).' __

Normalized Energies

- - Charpy (ft-lb/inZ) Yield Time to Time to Fast Test Energy Load Yield Max. Max. Fract. Arrest Yield Flow Sample Temp. ED Charpy Max. Prop. PGY tGY Load tro Load PF Load Stress cy Stress No. (0 F) (ft-lb) ED/A EM/A EpIA (lb) (msec) PM (lb) (msec) (lb) PA (lb) (ksi) (ksi)

KH20 -90 36 290 141 149 2310 0.13 4915 0.37 4839 226 77 120 KH24 -75 23 185 80 105 2366 0.12 4065 0.26 3854 563 79 107 KHI6 -50 28 226 113 113 2380 0.12 4485 0.34 4475 262 79 114 KH23 -25 36 290 138 153 2366 0.12 4492 0.38 4414 750 79 114 KH22 0 73 588 309 279 2752 0.16 4568 0.67 4167 722 92 122 KH25 25 89 717 310 407 2774 0.15 4415 0.68 4181 2328 92 120 KH30 25 107 862 309 553 2180 0.12 4457 0.69 3331 965 73 III KH21 50 76 612 220 393 2391 0.14 4297 0.54 4204 2088 80 III KH27 75 (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a)

KHI8 100 123 991 294 697 2450 0.13 4321 0.68 n/a n/a 82 113 KH28 125 137 1104 288 816 2284 0.13 4268 0.68 n/a n/a 76 109 KH29 150 106 854 286 568 2367 0.14 4144 0.68 n/a n/a 79 108 KHI9 200 (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a) (a)

KH26 250 137 1104 272 832 2352 0.14 4299 0.68 n/a n/a 78 III KH17 300 136 1096 292 804 2218 0.13 4207 0.69 n/a n/a 74 107

a. Computer malfunction, instrumented data not available.

Testing of Specimens from Capsule V

5-14 Table 5-9 Effect of Irradiation to 2.669 x 1019 n/cm 2 (E>1.0 MeV) on the Capsule V Toughness Properties of the Seabrook Unit 1 Reactor Vessel Surveillance Materials Average 30 (ft-lb)(') Average 35 mil Lateral(b) Average 50 ft-Ib10) Average Energy Absorption(s)

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

Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AE Lower Shell Plate -28.1 32.3 60.4 -6.3 62.0 68.3 -3.8 67.1 70.9 125 112 13 RI 808-3 (Long.)

Lower Shell Plate 9.0 70.3 61.3 46.4 103.0 56.6 56.5 126.3 69.8 79 69 10 R1808-3 (Trans.) _

Weld Metal -74.6 -32.9 41.7 -47.1 -15.5 31.6 -49.9 -13.0 36.9 160 156 4 (Heat # 4P6052)

Heat Affected -193.7 -54.5 139.2 -120.2 -29.3 90.9 -140.2 -25.5 114.7 130 130 0 Zone

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 5-11).

Testing of Specimens from Capsule V

5-15 Table 5-10 Comparison of the Seabrook Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence(d) Predicted Measured Predicted Measureid (x loll n/cm2, (OF) (a) (0F) (b) (%) (a) (%)(Cl E > 1.0 MeV) l U 0.3142 30.0 39.7 14 5.6 Lower Shell Plate R1:308-3 Y 1.292 47.1 47.2 20 14.4 (Longitudinal)

V 2.669 55.5 60.4 24 10.4 U 0.3142 30.0 28.8 14 7.6 Lower Shell Plate R18308-3 Y 1.292 47.1 34.9 20 16.5 (Transverse)

V 2.669 55.5 61.3 24 12.7 U 0.3142 20.9 25.4 14 10.0 Surve illance Y 1.292 32.9 24.3 20 10.0 V 2.669 38.7 41.7 24 2.5 U 0.3142 94.4 --- 0 Heat Afficted Zone Y 1.292 147.7 --- 12.3 Material V 2.669 -- 139.2 - -- 0 Notes:

(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the Best Estimate weight percent values of copper and nickel of the surveillance material (plate material chemistry factor is 44 0F, weld chemistry factor is 30.7 0F).

(b) Calculated using measured Charpy data plotted using CVGRAPH, Version 5.0.2 (See Appendix C)

(c) Valu.-s 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 V

5-16 Table 5-11 Tensile Properties of the Seabrook Unit I Capsule V Reactor Vessel Surveillance Materials Irradiated to 2.669 x 109 n/cm 2 (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) (ksi) (kip) (ksi) (ksi) (%) (%) (%)

KL6 75 0F 78.9 97.8 3.18 190.7 64.7 11.3 25.1 66 Lower Shell Plate R1808-3 KL5 2000 F 73.8 91.7 2.95 166.9 60.1 10.2 23.4 64 (Longitudinal)

KL4 5500 F 68.2 91.7 3.30 145.6 67.2 9.8 19.7 54 KT6 75OF 76.9 96.8 3.55 139.5 72.3 11.6 24.1 48 Lower Shell Plate R1808-3 KT5 1500 F 74.4 92.7 3.30 166.6 67.2 10.1 20.9 60 (Transverse)

KT4 5500 F 68.2 91.7 3.65 132.2 74.4 10.2 17.6 44 KW6 75 0F 77.8 90.9 2.67 187.6 54.4 10.5 25.4 71 Program Weld Metal KW5 2000 F 74.2 85.6 2.53 189.9 51.4 9.6 23.7 73 KW4 5500 F 69.8 87.6 2.63 178.6 53.5 10.1 24.1 70 Testing of Specimens from Capsule V

547 LOWER SHELL PLATE R1808-3 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:01 PM Data Set(s) Plotted Curve Plant Capsule Material OrL Heat #

1 Seabrook I UMRRA SA533BI LT R1808-3 2 Seabrook I U SA533BI LT R1808-3 3 Seabrook I y SA533BI LT R1808-3 4 Seabrook I V SA533BI LT RI 808-3 300 250 -

. 200_

P150 -

Lu z

8 100 -

50 -

-300 .200 -100 0 100 200 300 400 500 600 Temperature in Deg F 0 Set 1 a Set 2 0 Set 3 A Set 4 Results Culve Fluence ISE USE d-USE T @.30 d-T 30 T (350 d-T N50 2.2 125.0 .0 -28.1 .0 -3.8 .0 2 2. 2 IIS.0 .7.0 11.6 39.7 34.0 37.8 3 2.2 107.0 -18.0 19. 1 47.2 45,7 49. 5 4 2.2 112.0 -13.0 32.3 60.4 67. 1 70.9 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Longitudinal Orientation)

Testing of Specimens from Capsule V

5-18 LOWER SHELL PLATE R1808-3 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:06 PM Data Set(s) Plotted Curve Plant Capsule Material On. Heat #

1 Seabrook I UNIRRA SA533BI LT R1808-3 2 Seabrook 1 U SA533BI LT R1808-3 3 Seabrook I y SA533B1 LT R1808-3 4 Seabrook I V SA533BI LT R1808-3 200 150 aL 100 e

it o -

-300 0 300 600 Temperature In Dog F o Set I a Set 2 o Set 3 A Set 4 Results Curve Fluence LSE USE d-USE T @35 d-T @35 1.0 80. 2 .0 -6.3 .0 2 1.0 83.2 3.0 24.6 30.9 3 1.0 77. 7 -2.5 41.9 48.2 4 1.0 78.7 -1.5 62.0 68.3 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Longitudinal Orientation)

Testing of Specimens from Capsule V

5-19 LOWER SHELL PLATE R1808-3 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:23 PM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat #

1 Seabrook 1 UNIRRA SA533B I LT Ri 808-3 2 Seabrook I U SA533BI LT R1808.3 3 Seabrook I y SA533BI LT R1 808-3 4 Seabrook 1 V SA533BI LT R1808-3 125 100 S-0 75 50 25 -

0 -

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Dog F 0 Set 1 I Set 2 0 Set 3 A Set 4 Results Curve Fluence LSE USE d-USE T @50 d-T 050

.0 100.0 .0 18.8 .0 2 .0 100.0 .0 48.0 29. 2 3 .0 100.0 .0 55.2 36.4 4 .0 100.0 .0 95.2 76.4 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Longitudinal Orientation)

Testing of Specimens from Capsule V

5-20 LOWER SHELL PLATE R1808-3 (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:28 PM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat1#

Seabrook I UNRIRA SA533BI TL R1808-3 2 Scabrook I U SA533BI TL R1808-3 3 Seabrook I Y SA533BI TL R1 808-3 4 Seabrook I V SA533Bt TL R1 808-3 300 250 -_

,R200-2 P 150 --

i; 100 50 -

00

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F 0 Set 1I = Set 2 0 Set 3 & Set 4 Results Curve Fluence LSE USE d-USE T g30 d-T @30 T@50 d-T ,50 2.2 79.0 .0 9. 0 .0 56.5 .0 2 2.2 73.0 -6.0 37.8 28.8 78. 1 21.6 3 2.2 66.0 -13.0 43.9 34.9 92. 1 35.6 4 2. 2 69. 0 -I0.0 70.3 61.3 126.3 69. 8 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Transverse Orientation)

Testing of Specimens from Capsule V

';-21 LOWER SHELL PLATE R1 808-3 (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:32 PM Data Set(s) Plotted clarve Plant Capsule Material Orl. Heat #

Seabrook I UNIRRA SA533BI TL R1808-3 2 Seabrook I U SA533B I TL R1808-3 3 Seabrook I Y SA533B I TL R1 808-3 4 Seabrook I V SA533BI TL R1808-3 200 150 Aq C

.2

£ 100 I

E 50 0-

-300 0 300 600 Temperature in Deg F 0 Set 1 D Set 2 o Set 3 A Set 4 Results Curve Fluence ISE USE d-USE T @35 d-T @35 1.0 62.3 .0 46.4 .0 2 1.0 66.0 3.7 65.4 19.0 3 1.0 57.4 -4.9 76. 1 29. 7 4 1.0 57.3 -5.0 103.0 56.6 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Transverse Orientation)

Testing of Specimens from Capsule V

5-22 LOWER SHELL PLATE R1808-3 (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/1212005 02:35 PM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat #

Seabrook I UNIRRA SA533BI TL R1808-3 2 Seabrook I U SA533BI TL R1808-3 3 Seabrook 1 Y SA533BI TL R1808-3 4 Seabrook I V SA533BI TL R1808-3 125 .

100 -

75 -

50 - -

25- I

-300 -2013 -100 0 100 200 300 400 500 600 Temperature in Dog F o Set I n Set 2 0 Set 3 & Set4 Results Curve Fluence LSE USE d-USE T50 d-T @50

.0 100. 0 .0 38.5 .0 2 .0 100.0 .0 71.8 33.3 3 .0 100. 0 .0 70.6 32. 1 4 .0 100.0 .0 92. 8 54. 3 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Seabrook Unit I Reactor Vessel Lower Shell Plate R1808-3 (Transverse Orientation)

Testing of Specimens from Capsule V

5-23 SURVEILLANCE PROGRAM WELD METAL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:40 PM Data Set(s) Plotted Curve Plant Capsule Material 0CI. Heat #

Seabrook I UNIRRA SAW NA 4P6052 I

1.. Seabrook I U SAW NA 4P6052

'i Scabrook I y SAW NA 4P6052 Seabrook I V SAW NA 4P6052 300 -

250 --

-S 200 -

92

150 -

Lu z

i; 100 r7 500

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F 0 Set I I Set 2 0 Sot 3 ^ Set 4 Results Curve Fluence LSE USE d-USE T @30 d.T @30 T @50 d-T V0

2. 2 160.0 .0 -74.6 .0 -49.9 .0 2 2.2 144. 0 -16.0 -49.2 25.4 -29.6 20. 3 3 2.2 144.0 -16.0 -50.3 24.3 -26.6 23.3 4 2. 2 156.0 -4.0 -32.9 41.7 -13.0 36. 9 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for Seabrook Unit 1 Reactor Vessel Weld Metal Testing of Specimens from Capsule V

5-24 SURVEILLANCE PROGRAM WELD METAL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:43 PM Data Sct(s) Plotted Curve Plant Capsule Material Orl. Heat #

1 Seabrook I UNIRRA SAW NA 4P6052 2 Seabrook I U SAW NA 4P6052 3 Scabrook I Y SAW NA 4P6052 4 Seabrook I V SAW NA 4P6052 200 150 M

E

.2 a 100 k &104 .......

~A I

50 LX ^ , ,

0

-300 0 300 Boo Temperature in Deg F 0 Set I a Set 2 0 Set 3 A Set 4 Results Curve Fluwece LSE USE d-USE T @35 d-T @35 1.0 89.4 .0 -47.1 .0 2 1.0 88.9 .5 -32.3 14.8 3 1.0 83.4 -6. 0 -22. 8 24. 3 4 1.0 88.3 -1. 1 -15.5 31.6 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for Seabrook Unit 1 Reactor Vessel Weld Metal Testing of Specimens from Capsule V

5-25 SURVEILLANCE PROGRAM WELD METAL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:47 P.Mf Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat #

Seabrook 1 UNIRRA SAW NA 4P6052 I Seabrook I U SAW NA 4P6052 Seabrook 1 y SAW NA 4P6052 Seabrook I V SAW NA 4P6052 125! -

100 -

75 -

2 El r.

, 50 _

25 -

0

-300 -2010 -100 0 100 200 300 400 500 600 Temperature in Dog F o Set I a Set 2 o Set 3 - Set 4 Results Curve Fluence LSE USE d-USE T ,50 d-T @50 1 .0 100.0 .0 -41.9 .0 2 .0 100.0 .0 -25.6 16.3 3 .0 100.0 .0 -22.6 19.3 4 .0 100.0 .0 -6.8 35. 1 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for Seabrook Unit 1 Reactor Vessel Weld Metal Testing of Specimens from Capsule V

5-26 HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:52 PM Data Set(s) Plotted Curve Plant Capsule Material Or. Heat #

Seabrook I UNIRIA SA533BI NA RI 808-3 2 Seabrook I U SA533BI NA R1808-3 3 Scabrook I y SA533BI NA R1808-3 4 Seabrook I V SA533BI NA R1 808-3 300 -

250 -

w 200-12 P 150 -_

uJ 100 -,.

50

-300 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F 0 Set II a Set 2 0 Set 3 A Set 4 Results Curve Fluence LSE USE d-USE T @30 d-T @30 T @50 d-T @50 2.2 130.0 .0 -193.7 .0 .140.2 .0 2 2. 2 134.0 4.0 - 99.3 94.4 -59. 1 81.1 3 2. 2 114.0 -16.0 -46.0 147.7 -24.2 116.0 4 2.2 130.0 .0 -54.5 139.2 -25.5 114.7 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for Seabrook Unit 1 Reactor Vessel Heat Affected Zone Material Testing of Specimens from Capsule V

5-27 HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:55 PM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat #

Seabrook I UNIRRA SA533BI NA RI 808-3 2 Seabrook I U SA533BI NA R1808-3 3 Seabrook I y SA533B I NA R1 808-3 4 Seabrook I V SA533BI NA R1 808-3 200 150 a

a 100 0

50

-300 0 300 600 Temperature In Deg F 0 Set I m Set 2 0 Set 3 A Set 4 Results Curve Fluence LSE USE d-USE T @35 d-T @35 1.0 81.0 .0 -120.2 .0 2 1.0 82. 8 1.8 .57.2 63. 0 3 1.0 72. 8 -8. 1 -18.9 101. 3 4 1.0 77. 6 -3.3 -29.3 90.9 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for Seabrook Unit 1 Reactor Vessel Heat Affected Zone Material Testing of Specimens from Capsule V

5-28 HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/1212005 02:58 PM Data Set(s) Plotted Curve Plant Capsule Material On. Heat #

1 Seabrook I UNIRRA SA533BI NA R1808-3 2 Seabrook I U SA533B1 NA R1808-3 3 Scabrook 1 Y SA533BI NA R1808-3 4 Seabrook I V SA533B1 NA R1808-3 125 100 L. 75 0

x t.

50 25 -

o

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Dog F 0 Set I o Set 2 0 Set 3 A Set 4 Results Curve Fluence LSE USE d-USE T @50 d-T @50

.0 100.0 .0 -102.1 .0 2 .0 100.0 .0 -57.8 44.3

.3 .0 100. 0 .0 -34.2 67. 9 4 .0 100.0 .0 -15.8 86. 3 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for Seabrook Unit 1 Reactor Vessel Heat Affected Zone Material Testing of Specimens from Capsule V

I 5-29 K1 'I -A0 °F 1(1 10 (OF W1(16. ;5OF O0 F KM 20r KL25, 500 F

.1 Ta 7C0L1 T-T 9i 7°OT TI r %A I flfnO 7 T'-T 'an I 'fO T- TT 'I 7 1C KL17, 175 0F KL22, 225-F KL28, 250 0F KL23, 300°F KL18, 350°F Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Longitudinal Orientation)

Testing of Specimens from Capsule V

5-30 WTf3 -,5noF KT1 7 0F KT30 2.SOF KT1 9 50nF KT1 A 75OF WT9n 7!0jF WT95 1O)(iFn WT1 R 1 2;OP WT21 1 25FP KT23, 1500 F KT26, 2000 F KT22, 2500 F KT29, 3000 F KT24, 3500 F Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Transverse Orientation)

Testing of Specimens from Capsule V

U t

5-31 TW27 OOF R7W29 25?F KW18, 25 0F KW2gg 0 O

nF KW22W 75°F KW25, 75 0F KW26, 100 0F KW17, 200 0F KW19, 250 0F KW21, 300 0F Figure 5-15 Charpy Impact Specimen Fracture Surfaces for Seabrook Unit I Reactor Vessel Weld Metal Testing of Specimens from Capsule V

5-32 I(H2f -900 F KH244 -75°F KH16 -50 0 F KH23, -25"F KH22. 0nF WTH2.r 2oF KH3n 2;5°F KH1. n0oF KH2T7 75 0F KH14, 100OF FI KH28, 125 0F KH29, 150 0F KH19, 200°F KH26, 250°F KH17, 300 0F Figure 5-16 Charpy Impact Specimen Fracture Surfaces for Seabrook Unit 1 Reactor Vessel Heat Affected Zone Material Testing of Specimens from Capsule V

I L

5-33 120 - ULTIMATE TENSILE STRENGTH 100-

.a 80 -

cu'60 X60 0.2% YIELD STRENGTH to 40 20 0 100 200 300 400 500 600 TEMPERATURE( F)

Legend: A and o are Unirradiated Aand

• are Irradiated to 2.669 x 1019 n/cm2 (E > 1.0 MeV) 80 -

REDUCTION INAREA 70 60 50 a-40 30 TOTAL ELONGATION 20 10 UNIFORM UNIFORM 0 I l

--- -I I 0 100 200 300 400 500 600 TEMPERATURE ( F)

Figure 5-17 Tensile Properties for Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Longitudinal Orientation)

Testing of Specimens from Capsule V

5-34 120 -

1oo - ULTIMATE YIELD STRENGTH 60 0.2% YIELD STRENGTH te 40 20 0

0 100 200 300 400 500 600 TEMPERATURE( F)

Legend: A and o are Unirradiated Aand are Irradiated to 2.669 x 1019 n/cm 2 (E > 1.0 MeV) 70 REDUCTION IN AREA 60-

-- 50 -1

> 40

~30 30TOTAL ELONGATION a 20 10-UNIFORM UNIFORM 0 100 200 300 400 500 600 TEMPERATURE ( F)

Figure 5-18 Tensile Properties for Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Transverse Orientation)

Testing of Specimens from Capsule V C C°Z

5-35 100 ULTIMATE TENSILE STRENGTH 80 0 ST 40- 60 ~60 0.2% YiIELD STRENGTH U)

Ix40 1 U) 20 0

100 200 300 400 500 600 TEMPERATURE( F)

Legend: A and o are Unirradiated Aand

• are Irradiated to 2.669 x 10'8 n/cm 2 (E > 1.0 MeV) 80 70

_ 60 REDUCTION IN AREA

50 I--

i 40 I--

Q 30 TOTAL ELONGATION a 20 A

10 AA UNI FORM ELONGATION 0

0 100 200 300 400 500 600 TEMPERATURE ( F)

Figure 5-19 Tensile Properties for Seabrook Unit 1 Reactor Vessel Weld Metal r r)

Testing of Specimens from Capsule V

5-36 Specimen KL-6 Tested at 750 F Specimen KL-5 Tested at 200'F Specimen KL-4 Tested at 550'F Figure 5-20 Fractured Tensile Specimens from Seabrook Unit 1 Reactor Vessel Lower Shell Plate R1808-3 (Longitudinal Orientation)

Testing of Specimens from Capsule V

5-37 Specimen KT-6 Tested at 75 0 F Specimen KT-5 Tested at 150'F Specimen KT-4 Tested at 5500 F Figure 5-21 Fractured Tensile Specimen from Seabrook Unit I Reactor Vessel Lower Shell Plate R1808-3 (Transverse Orientation)

Testing of Specimens from Capsule V

5-38 Specimen KW-6 Tested at 75 0 F Specimen KW-5 Tested at 200OF Specimen KW-4 Tested at 5500 F Figure 5-22 Fractured Tensile Specimen from Seabrook Unit 1 Reactor Vessel Weld Metal Testing of Specimens from Capsule V

_-39 SEABROOK UNIT I CAPSULE V 100 80 I- 60 CD al T_

of 40 KL6 75-F 20 0

C 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN SEABROOK UNIT I CAPSULE V 100 1 80 0e 6 60-w) t-F 40-KL5 200'F 20 0

0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN. INIIN Figure 5-23 Engineering Stress-Strain Curves for Seabrook Unit 1 Lower Shell Plate R1808-3 Tensile Specimens KL-6 and KL-5 (Longitudinal Orientation)

Testing of Secimens from Capsule V

5-40 SEABROOK UNIT I CAPSULE V 100 -

80 -

U, 60 -

I-ow 0,

40-KL4 550'F 20 0

0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN1N Figure 5-24 Engineering Stress-Strain Curve for Seabrook Unit 1 Lower Shell Plate R1808-3 Tensile Specimen KL-4 (Longitudinal Orientation)

Testing of Specimens from Capsule V

5_41 SEABROOK UNIT 1 CAPSULE V 100 82

,6 6)

U) 4')

KT6 75'F 201 (I

0 0.05 0.1 0.15 02 0.25 0.3 STRAIN, INAN SEABROOK UNIT I CAPSULE V 100 80 ve 60 V) w 40 KT5 150'F 20 0

0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, ININ Figure 5-25 Engineering Stress-Strain Curves for Seabrook Unit 1 Lower Shell Plate R1808-3 Tensile Specimens KT-6 and KT-5 (Transverse Orientation)

Testing of Specimens from Capsule V

5-42 SEABROOK UNIT I CAPSULE V 100 80 Y-co 60 or Cd, C) 40 KT4 550'F 20 0

0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN Figure 5-26 Engineering Stress-Strain Curve for Seabrook Unit I Lover Shell Plate R1808-3 Tensile Specimen KT4 (Transverse Orientation)

Testing of Specimens from Capsule V

5-43 SEABROOK UNIT I CAPSULE V Ico 80 o5 a,, 63 w

43 KW6 75TF 2) 0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN SEABROOK UNIT I CAPSULE V 100 V5 6 6C w

f-4C 20 0

0.05 0.1 0.15 0.2 0.25 0.3 STRAIN. INIIN Figure 5-27 Engineering Stress-Strain Curves for Seabrook Unit 1 Weld Metal Tensile Specimens KW-6 and KW-5 Testing of Specimens from Capsule V

5-44 SEABROOK UNIT 1 CAPSULE V 100 s0 V5

<, 60 40 in 40 KW4 550'F 20 0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, INnN Figure 5-28 Engineering Stress-Strain Curve for Seabrook Unit 1 Weld Metal Tensile Specimen KV-4 (Longitudinal Orientation)

Testing of Specimens from Capsule V

5-1 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6.1 INTRODUCTION

This section describes a discrete ordinates Sn transport analysis performed for the Seabrook Station 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 V,withdrawn at the end of the tenth plant operating cycle, is provided. In addition, to provide an up-to-date data base applicable to the Seabrook Station reactor, sensor sets from previously withdrawn capsules (U and Y) were re-analyzed using the current do imetry 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 60 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 yea!s, 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 perAtom." 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" [Ref. 20]. Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC approved methodology described in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004 [Ref. 21].

Radiation Analysis and Neutron Dosimetry

6-2 6.2 DISCRETE ORDINATES ANALYSIS A plan view of the Seabrook Station reactor geometry at the core mid-plane is shown in Figure 4-1.

Figures 6-1 through 6-3 represent all of the unique octants of the full-core shown in Figure 4-1. Six irradiation capsules attached to the reactor vessel are included in the reactor design that constitutes the reactor vessel surveillance program. The capsules are located at azimuthal angles of 58.5°,610, 121.50, 238.50, 2410, and 301.50 as shown in Figure 4-1. These full core position correspond to the following octant symmetric locations represented in Figures 6-1 and 6-2: 290 from the core cardinal axis (for the 610 and 2410 dual surveillance capsule holder locations) and 31.50 from the core cardinal axis (for the 121.50 and 301.50 single surveillance capsule holder locations, and for the 58.50 and 238.50 dual surveillance capsule holder locations). The stainless steel specimen containers are 1.25-inch by 1.0-inch and are approximately 56.0 inches in height. The containers are positioned axially such that the test specimens are centered on the core mid-plane, thus spanning the central 5 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 core barrel 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 Seabrook Station 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:

0(r, 0, z) = b(r, 0) *0(r,z) where 4(rO,z) is the synthesized three-dimensional neutron flux distribution, 4(r,0) is the transport solution in rO 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,0 two-dimensional calculation. This synthesis procedure was carried out for each operating cycle at Seabrook Station.

For the Seabrook Station transport calculations, the octant r,0 models shown in Figures 6-1 through 6-3 were utilized since, with the exception of the neutron pads, the reactor is octant symmetric. These octant r,0 models use reflective boundary conditions and include the core, the reactor internals, the neutron pads

- including explicit representations of octants not containing surveillance capsules and octants with surveillance capsules at 290 and 31.50, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. These models formed the basis for the calculated results and enabled making comparisons to the surveillance capsule dosimetry evaluations.

In developing these analytical models, nominal design dimensions were employed for the various structural components. Moderator water temperatures, and hence coolant densities, in the reactor core and bypass 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 Radiation Analysis and Neutron Dosimetry

.5-3 grids, guide tubes, et cetera. The geometric mesh description of the rO reactor models consisted of 183 radial by 99 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.

The rz model used for the Seabrook Station calculations is shown in Figure 6-4 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 below the lower core plate to above the upper core plate. As in the case of the r,0 models, 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 these reactor models consisted of 153 radial by 188 axial intervals. As in the case of the r1O 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 0.00 1.

The one-dimensional radial models 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.

Data used in the transport analyses were taken from the appropriate Seabrook Station cycle-specific fuel information provided by FPL Energy and the Nuclear Fuels unit of Westinghouse. As-operated conditions at a power level of 3411 MWt are reflected through Cycle 10. The extracted data represented cycle dependent fuel assembly enrichments, burn-ups, and radial and axial power distributions. 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 the 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 3.1 [Ref. 22] and the BUGLE-96 cross-section library [Ref. 23]. 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.

Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-12. In Tables 6-1 through 6-4, 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 three azimuthally symmetric surveillance capsule positions (290 dual capsule, 31.50 dual capsule, and 31.5° Radiation Analysis and Neutron Dosimetry

6-4 single capsule). These results, representative of the axial mid-plane 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 Tables 6-5 through 6-8 for the reactor vessel inner radius at five azimuthal locations. The vessel data given in Tables 6-5 through 6-8 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 Tables 6-1 through Table 6-8. These data tabulations include both plant and fuel cycle specific calculated neutron exposures at the end of the tenth operating fuel cycle and future projections at multiple times through 60 EFPY, including 32 EFPY.

The projections were based on the assumption that the core power distributions and associated plant operating characteristics from Cycle 11 were representative of future plant operation. The future projections are also based on the current reactor power level of 3659 MWt and core average temperature of 593 'F.

Radial gradient information applicable to fast (E > 1.0 MeV) neutron fluence and dpa are given in Tables 6-9 and 6-10, respectively. The data, based on the cumulative integrated exposures from Cycles 1 through 10, 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 6-9 and 6-10.

The calculated fast neutron exposures for the three surveillance capsules withdrawn from the Seabrook Station reactor are provided in Table 6-11. These assigned neutron exposure levels are based on the plant and fuel cycle specific neutron transport calculations performed for the Seabrook Station reactor.

Updated lead factors for the Seabrook Station surveillance capsules are provided in Table 6-12. 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 6-12, the lead factors for capsules that have been withdrawn from the reactor (U, Y,and V) 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 (W, X, and Z), the lead factor corresponds to the calculated fluence values at the end of Cycle 10, the latest completed operating fuel cycle for Seabrook Station.

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, are documented in Appendix A.

Radiation Analysis and Neutron Dosimetry

65-5 The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsule V,that was withdrawn from Seabrook Station at the end of the tenth fuel cycle, is summarized below.

Reaction Rates (rps/atom) M/C Reaction Measured Calculated Ratio 63 Cu(na)60Co 4.12E-17 3.79E-17 1.09 54 Fe(n,p)54 Mn 4.31E-15 4.12E-15 1.05 58Ni(n,p)58Co 6.06E-15 5.76E-15 1.05 238U(n,f)137Cs(Cd) 2.36E-14 2.19E-14 1.08 237 Np(n,t1)37 Cs(Cd) 2.14E-13 2.12E-13 1.01 Average: 1.05

% Standard Deviation: 3.0 The measured-to-calculated (M/C) reaction rate ratios for the Capsule V threshold reactions range from 1.01 to 1.09, and the average M/C ratio is 1.05 +/- 3.0% (lay). 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 Seabrook Station reactor. These comparisons validate the current analytical results described in Section 6.2; therefore, the calculations are deemed applicable for Seabrook Station.

6.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Seabrook Station surveillance capsule and reactor pressure vessel is based on the recommended approach provided in Regulatory Guide 1.19D. In particular, the qualification of the methodologywas carried out in the following four stages:

I - 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.

Radiation Analysis and Neutron Dosimetry

6-6 4 - Comparisons of the plant specific calculations with all available dosimetry results from the Seabrook Station 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 Seabrook Station analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Seabrook Station 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 Seabrook Station 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.

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%/0 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 Seabrook Station.

Radiation Analysis and Neutron Dosimetry

6-7 Table 6-1 Calculated Neutron Flux At The Surveillance Capsule Center Cumulative Cumulative Neutron Flux (E > 1.0 MeV)

Cycle Irradiation Irradiation In/cm2 -sI Length Time Time Dual Dual Single Cycle IEFPSI IEFPSI [EFPY] 290 31.50 31.50 1 2.88E+07 2.88E+07 0.91 1.OOE+11 1.09E+11 1.08E+I 2 2.75E+07 5.63E+07 1.78 8.03E+10 8.44E+10 8.35E+10 3 3.81 E+07 9.44E+07 2.99 7.97E+10 8.77E+10 8.69E+10 4 3.81E+07 1.33E+08 4.20 5.61E+10 5.98E+10 6.38E+10 5 4.33E+07 1.76E+08 5.57 6.13E+10 6.62E+10 6.55E+10 6 4.69E+07 2.23E+08 7.06 6.1OE+10 6.56E+10 6.49E+10 7 4.44E+07 2.67E+08 8.46 6.79E+10 7.52E+ 10 7.45E+10 8 3.73E+07 3.04E+08 9.65 6.08E+10 6.64E+10 6.58E+10 9 4.26E+07 3.47E+08 11.00 6.51E+10 7.14E+10 7.08E+10 10 4.41E+07 3.91E+08 12.39 6.48E+10 7.07E+10 7.OOE+10 11 4.41E+07 4.35E+08 13.79 6.1lE+10 6.63E+10 6.56E+10 Future 2.59E+08 6.94E+08 22.00 6.11E+10 6.63E+10 6.56E+ 10 Future 1.89E+08 8.84E+08 28.00 6.1lE+10 6.63E+10 6.56E+10 Future 2.21 E+08 1.011E+09 32.00 6.11E+10 6.63E+I 0 6.56E+10 Future 2.52E+08 1.14E+09 36.00 6.11E+10 6.63E+10 6.56E+10 Future 1.89E+08 1.33E+09 42.00 6.111E+10 6.63E+10 6.56E+10 Future 1.89E+08 1.511E+09 48.00 6.11E+10 6.63E+10 6.56E+10 Future 1.89E+08 1.70E+09 54.00 6.11E+10 6.63E+10 6.56E+10 Future 1.89E+08 1.89E+09 60.00 6.11E+10 6.63E+10 6.56E+10 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry

6-8 Table 6-2 Calculated Neutron Fluence At The Surveillance Capsule Center Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation In/cm 2 i Length Time Time Dual Dual Single Cycle IEFPSI [EFPSJ IEFPY] 290 31.50 31.50 1 2.88E+07 2.88E+07 0.91 2.88E+18 3.14E+18 3.1lE+18 2 2.75E+07 5.63E+07 1.78 5.09E+18 5.46E+18 5.41E+18 3 3.81E+07 9.44E+07 2.99 8.13E+18 8.80E+18 8.72E+18 4 3.8 1 E+07 1.33E+08 4.20 1.03E+19 1.1E+19 1.12E+19 5 4.33E+07 1.76E+08 5.57 1.29E+19 1.40E+19 1.40E+19 6 4.69E+07 2.23E+08 7.06 1.58E+19 1.70E+19 1.70E+19 7 4.44E+07 2.67E+08 8.46 1.88E+19 2.04E+19 2.03E+19 8 3.73E+07 3.04E+08 9.65 2.1lE+19 2.28E+19 2.28E+19 9 4.26E+07 3.47E+08 11.00 2.38E+19 2.59E+19 2.58E+19 10 4.411E+07 3.91E+08 12.39 2.67E+19 2.90E+19 2.89E+19 11 4.41E+07 4.35E+08 13.79 2.94E+19 3.19E+19 3.18E+19 Future 2.59E+08 6.94E+08 22.00 4.52E+19 4.91E+19 4.88E+19 Future 1.89E+08 8.84E+08 28.00 5.68E+19 6.16E+19 6.12E+19 Future 2.21 E+08 1.01 E+09 32.00 6.45E+19 7.00E+19 6.95E+19 Future 2.52E+08 1.14E+09 36.00 7.22E+19 7.84E+ 19 7.78E+19 Future 1.89E+08 1.33E+09 42.00 8.38E+19 9.09E+19 9.02E+19 Future 1.89E+08 1.51E+09 48.00 9.54E+19 1.03E+20 1.03E+20 Future 1.89E+08 1.70E+09 54.00 1.07E+20 1.16E+20 1.15E+20 Future 1.89E+08 1.89E+09 60.00 1.19E+20 1.29E+20 1.28E+20 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry

65-9 Table 6-3 Calculated Iron Atom Displacement Rate At The Surveillance Capsule Center Cumulative Cumulative Iron Atom Displacement Rate Cycle Irradiation Irradiation Idpa/sI Length Time Time Dual Dual Single Cycle JEFPSJ IEFPSJ IEFPYI 290 31.50 31.50 1 2.88E+07 2.88E+07 0.91 1.96E-10 2.14E-10 2.12E-10 2 2.75E+07 5.63E+07 1.78 1.57E-10 1.65E-10 1.63E-10 3 3.81 E+07 9.44E+07 2.99 1.55E-10 1.711E-10 1.69E-10 4 3.81 E+07 1.33E+08 4.20 1.09E-10 1.16E-10 1.23E-10 5 4.33E+07 1.76E+08 5.57 1.19E-10 1.28E-10 1.27E-10 6 4.69E+07 2.23E+08 7.06 1.18E-10 1.27E-10 1.26E-10 7 4.44E+07 2.67E+08 8.46 1.32E-10 1.46E-10 1.44E-10 8 3.73E+07 3.04E+08 9.65 1.18E-10 1.29E- 10 1.27E-10 9 4.26E+07 3.47E+08 11.00 1.27E-10 1.39E-10 1.37E-10 10 4.41E+07 3.91E+08 12.39 1.26E-10 1.37E-10 1.36E-10 11 4.411E+07 4.35E+08 13.79 1.19E-10 1.28E-10 1.27E-10 Future 2.59E+08 6.94E+08 22.00 1.19E-10 1.28E-10 1.27E-10 Future 1.89E+08 8.84E+08 28.00 1.19E-10 1.28E-10 1.27E-10 Future 2.21 E+08 1.01 E+09 32.00 1.19E-10 1.28E-10 1.27E-10 Future 2.52E+08 1.14E+09 36.00 1.19E-10 1.28E-10 1.27E-10 Future I.89E+08 1.33E+09 42.00 1.19E- 10 1.28E-10 1.27E-10 Future 1.89E+08 1.51E+09 48.00 1.19E- 10 1.28E-10 1.27E-10 Future 1.89E+08 1.70E+09 54.00 1.19E-10 1.28E-10 1.27E-10 Future I.89E+08 1.89E+09 60.00 1.19E- 10 1.28E-10 1.27E-10 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry

6-10 Table 64 Calculated Iron Atom Displacements At The Surveillance Capsule Center Cumulative Cumulative Iron Atom Displacements Cycle Irradiation Irradiation Idpal Length Time Time Dual Dual Single Cycle IEFPSJ JEFPSJ [EFPYI 290 31.50 31.50 1 2.88E+07 2.88E+07 0.91 5.65E-03 6.16E-03 6.09E-03 2 2.75E+07 5.63E+07 1.78 9.97E-03 1.07E-02 1.06E-02 3 3.8 1E+07 9.44E+07 2.99 1.59E-02 1.72E-02 1.70E-02 4 3.81E+07 1.33E+08 4.20 2.OOE-02 2.16E-02 2.17E-02 5 4.33E+07 1.76E+08 5.57 2.52E-02 2.72E-02 2.72E-02 6 4.69E+07 2.23E+08 7.06 3.07E-02 3.31 E-02 3.31E-02 7 4.44E+07 2.67E+08 8.46 3.66E-02 3.96E-02 3.95E-02 8 3.73E+07 3.04E+08 9.65 4.1OE-02 4.44E-02 4.43E-02 9 4.26E+07 3.47E+08 11.00 4.64E-02 5.03E-02 5.011E-02 10 4.41E+07 3.91 E+08 12.39 5.19E-02 5.64E-02 5.61E-02 11 4.41E+07 4.35E+08 13.79 5.71 E-02 6.20E-02 6.17E-02 Future 2.59E+08 6.94E+08 22.00 8.78E-02 9.53E-02 9.46E-02 Future 1.89E+08 8.84E+08 28.00 1.1 OE-0 I 1.20E-01 1.19E-0O Future 2.21 E+08 1.0 IE+09 32.00 1.25E-01 1.36E-01 1.35E-0O Future 2.52E+08 1.14E+09 36.00 1.40E-01 1.52E-01 1.51E-01 Future 1.89E+08 1.33E+09 42.00 1.63E-01 1.76E-01 1.75E-01 Future 1.89E+08 1.51E+09 48.00 1.85E-01 2.01 E-0 1 1.99E-01 Future 1.89E+08 1.70E+09 54.00 2.07E-01 2.25E-01 2.23E-0l Future 1.89E+08 1.89E+09 60.00 2.30E-01 2.49E-01 2.47E-01 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry

6-11 Table 6-5 Calculated Azimuthal Variation Of Neutron Flux At The Reactor Vessel Clad/ Base Metal Interface Cumulative Cumulative Neutron Flux (E > 1.0 MeV)

Cycle Irradiation Irradiation In/cm 2-sl Length Time Time Cycle [EFPS] [EFPS] IEFPYI 00 150 300 450 I 2.88E+07 2.88E+07 0.91 1.41 E+10 2.15E+10 2.53E+10 2.76E+10 2 2.75E+07 5.63E+07 1.78 1.33E+10 2.02E+10 2.06E+10 2.1OE+10 3 3.8 1E+07 9.44E+07 2.99 1.19E+10 1.68E+10 2.011E+10 2.211E+10 4 3.81E+07 1.33E+08 4.20 1.09E+10 1.52E+10 1.54E+10 1.51lE+10 5 4.33E+07 1.76E+08 5.57 8.43E+09 1.32E+10 1.58E+10 1.60E+10 6 4.69E+07 2.23E1+08 7.06 9.88E+09 1.42E+10 1.60E+10 1.56E+10 7 4.44E+07 2.67E+08 8.46 8.69E+09 1.311E+10 1.73E+10 1.75E+10 8 3.73E+07 3.04E+08 9.65 9.03E+09 1.31E+ 10 1.55E+10 1.62E+10 9 4.26E+07 3.47E+08 11.00 8.97E+09 1.36E+10 1.66E+10 1.76E+10 10 4.41 E+07 3.91E+08 12.39 9.27E+09 1.36E+10 1.67E+10 1.75E+10 11 4.411E+07 4.35E+08 13.79 9.40E+09 1.40E+10 1.64E+10 1.65E+10 Future 2.59E+08 6.94E+08 22.00 9.40E+09 1.40E+10 1.64E+10 1.65E+10 Future 1.89E+08 8.84E+08 28.00 9.40E+09 1.40E+10 1.64E+10 1.65E+10 Future 2.21 E+08 1.0 IE+09 32.00 9.40E+09 1.40E+ 10 1.64E+10 1.65E+10 Future 2.52E+08 1.14E+09 36.00 9.40E+09 1.40E+10 1.64E+10 1.65E+10 Future 1.89E+08 1.33E+09 42.00 9.40E+09 1.40E+ 10 1.64E+10 1.65E+10 Future 1.89E+08 1.51 E+09 48.00 9.40E+09 1.40E+10 1.64E+10 1.65E+10 Future 1.89E+08 1.70E+09 54.00 9.40E+09 1.40E+10 1.64E+10 1.65E+10 Future 1.89E+08 1.89E+09 60.00 9.40E+09 1.40E+10 1.64E+10 1.65E+10 Neutron Flux (E > 1.0 MeV) per EFPY used in future projections 2.97E+17 4.42E+ 17 5.18E+17 5.211E+17 In/cm2- EFPYI Radiation Analysis and Neutron Dosimetry

6-12 Table 6-6 Calculated Azimuthal Variation Of Neutron Fluence At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation in/cm2 1 Length Time Time Cycle IEFPSJ IEFPSJ IEFPY] 00 150 300 450 1 2.88E+07 2.88E+07 0.91 4.06E+17 6.18E+17 7.27E+17 7.94E+17 2 2.7513+07 5.63E+07 1.78 7.60E1+17 1.16E+18 1.28E+18 1.35E+18 3 3.81 E+07 9.4413+07 2.99 1.2213+18 1.80E+18 2.04E1+18 2.20E+18 4 3.81E+07 1.33E+08 4.20 1.63E+18 2.38E+18 2.63E+18 2.77E+18 5 4.33E+07 1.76E+08 5.57 1.99E+18 2.94E+18 3.31E+18 3.46E+18 6 4.69E+07 2.23E+08 7.06 2.44E1+18 3.59E+18 4.03E+18 4.17E+18 7 4.4413+07 2.67E+08 8.46 2.82E+18 4.16E+ 18 4.80E+18 4.94E+18 8 3.73E+07 3.04E+08 9.65 3.16E+18 4.65E+18 5.38E+18 5.55E+18 9 4.26E+07 3.47E+08 11.00 3.54E+18 5.23E+18 6.0813+18 6.29E+18 10 4.4113+07 3.911E+08 12.39 3.95E+18 5.82E+18 6.811E+18 7.05E+18 11 4.41 E+07 4.35E+08 13.79 4.3513+18 6.42E+18 7.511E+18 7.76E+18 Future 2.5913+08 6.94E+08 22.00 6.74E1+18 9.99E+18 1.17E+19 1.20E+19 Future 1.89E+08 8.8413+08 28.00 8.52E+18 1.26E+19 1.48E+19 1.511E+19 Future 2.21 E+08 1.01 E+09 32.00 9.7113+18 1.44E+ 19 1.6913+19 1.72E+19 Future 2.52E+08 1.14E+09 36.00 1.09E+19 1.62E+19 1.8913+19 1.92E+19 Future 1.8913+08 1.3313+09 42.00 1.2713+19 1.8813+19 2.2013+19 2.2413+19 Future 1.89E+08 1.511E+09 48.00 1.45E+19 2.15E1+19 2.511E+19 2.55E+19 Future 1.89E+08 1.7013+09 54.00 1.62E+19 2.411E+19 2.82E+19 2.86E+19 Future 1.89E+08 1.89E+09 60.00 1.80E+19 2.68E+19 3.13E+19 3.1713+19 Radiation Analysis and Neutron Dosimetry

6-13 Table 6-7 Calculated Azimuthal Variation Of Iron Atom Displacement Rates At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacement Rate Cycle Irradiation Irradiation Idpa/sl Length Time Time Cycle IEFPSI IEFPSI IEFPYi 00 150 300 450 1 2.88E+07 2.88E+07 0.91 2.19E-11 3.29E-11 3.89E-11 4.36E- l1 2 2.75E+07 5.63E+07 1.78 2.06E-11 3.10E-11 3.17E-11 3.32E-l1l 3 3.81E+07 9.44E+07 2.99 1.85E-11 2.58E-11 3.11IE-11 3.49E-1I 4 3.81E+07 1.33E+08 4.20 1.69E-11 2.34E-11 2.38E-11 2.39E-I I 5 4.33E+07 1.76E+08 5.57 1.31E-1I 2.03E-11 2.44E-11 2.53E-1 6 4.69E+07 2.23E+08 7.06 1.53E-11 2.18E-11 2.47E- 11 2.46E- 11 7 4.44E+07 2.67E+08 8.46 1.35E-11 2.011E-11 2.67E-11 2.77E- 11 8 3.73E+07 3.04E+08 9.65 1.40E-11 2.011E-11 2.40E-11 2.56E-11 9 4.26E+07 3.47E+08 11.00 1.40E- 11 2.1OE-11 2.56E- 11 2.78E-11 10 4.41E+07 3.91E+08 12.39 1.44E-11 2.09E-11 2.58E-11 2.76E-I I 11 4.41 E+07 4.35E+08 13.79 1.46E-11 2.16E- 11 2.53E-11 2.60E-1 I Future 2.59E+08 6.94E+08 22.00 1.46E-11 2.16E-11 2.53E-11 2.60E- II Future 1.89E+08 8.84E+08 28.00 1.46E-11 2.16E-11 2.53E-11 2.60E-11 Future 2.2 1E+08 1.OIE+09 32.00 1.46E-11 2.16E-11 2.53E-11 2.60E-I Il Future 2.52E+08 1.14E+09 36.00 1.46E-11 2.16E-11 2.53E-11 2.60E- 11 Future 1.89E+08 1.33E+09 42.00 1.46E-11 2.16E-11 2.53E-11 2.60E-1I I Future 1.89E+08 1.51 E+09 48.00 1.46E- 11 2.16E-11 2.53E- 11 2.60E- 11 Future 1.89E+08 1.70E+09 54.00 1.46E-11 2.16E- 11 2.53E-11 2.60E- 11 Future 1.89E+08 1.89E+09 60.00 1.46E-11 2.16E-11 2.53E-11 2.60E-I I Radiation Analysis and Neutron Dosimetry

6-14 Table 6-8 Calculated Azimuthal Variation Of Iron Atom Displacements At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacements Cycle Irradiation Irradiation Idpal Length Time Time Cycle [EFPSl IEFPSI IEFPYl 00 150 300 450 1 2.88E+07 2.88E+07 0.91 6.30E-04 9.49E-04 1.12E-03 1.25E-03 2 2.75E+07 5.63E+07 1.78 1.18E-03 1.78E-03 1.97E-03 2.14E-03 3 3.81 E+07 9.44E+07 2.99 1.89E-03 2.76E-03 3.15E-03 3.47E-03 4 3.81 E+07 1.33E+08 4.20 2.53E-03 3.65E-03 4.06E-03 4.38E-03 5 4.33E+07 1.76E+08 5.57 3.09E-03 4.52E-03 5.l l E-03 5.47E-03 6 4.69E+07 2.23E+08 7.06 3.79E-03 5.5 1E-03 6.22E-03 6.58E-03 7 4.44E+07 2.67E+08 8.46 4.39E-03 6.40E-03 7.4 1E-03 7.81 E-03 8 3.73E+07 3.04E+08 9.65 4.9 lE-03 7.16E-03 8.30E-03 8.77E-03 9 4.26E+07 3.47E+08 11.00 5.50E-03 8.04E-03 9.38E-03 9.93E-03 10 4.41E+07 3.91 E+08 12.39 6.13E-03 8.96E-03 1.05E-02 1.12E-02 11 4.41E+07 4.35E+08 13.79 6.75E-03 9.87E-03 1.16E-02 1.23E-02 Future 2.59E+08 6.94E+08 22.00 1.05E-02 1.54E-02 1.80E-02 1.89E-02 Future 1.89E+08 8.84E+08 28.00 1.32E-02 1.94E-02 2.28E-02 2.38E-02 Future 2.21 E+08 1.011E+09 32.00 1.51 E-02 2.22E-02 2.60E-02 2.71 E-02 Future 2.52E+08 1.14E+09 36.00 1.69E-02 2.49E-02 2.92E-02 3.04E-02 Future 1.89E+08 1.33E+09 42.00 1.97E-02 2.90E-02 3.40E-02 3.53E-02 Future 1.89E+08 1.51E+09 48.00 2.25E-02 3.30E-02 3.87E-02 4.02E-02 Future 1.89E+08 1.70E+09 54.00 2.52E-02 3.71 E-02 4.35E-02 4.52E-02 Future 1.89E+08 1.89E+09 60.00 2.80E-02 4.12E-02 4.83E-02 5.0IE-02 Radiation Analysis and Neutron Dosimetry

6-15 Table 6-9 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV)

Within The Reactor Vessel Wall RADIUS AZIMUTHAL ANGLE (cm) 00 150 300 450 220.11 1.000 1.000 1.000 1.000 225.59 0.573 0.568 0.563 0.559 231.06 0.283 0.278 0.274 0.270 236.54 0.135 0.131 0.128 0.126 242.01 0.065 0.060 0.058 0.057 Note: Base Metal Inner Radius = 220.11 cm Base Metal I/4T = 225.59 cm Base Metal 1/2T = 231.06 cm Base Metal 3/4T = 236.54 cm Base Metal Outer Radius = 242.01 cm Radiation Analysis and Neutron Dosimetry

6-16 Table 6-10 Relative Radial Distribution Of Iron Atom Displacements (dpa)

Within The Reactor Vessel Wall RADIUS AZIMUTHAL ANGLE (cm) 00 150 300 450 220.11 1.000 1.000 1.000 1.000 225.59 0.645 0.639 0.639 0.647 231.06 0.393 0.384 0.385 0.395 236.54 0.239 0.229 0.230 0.237 242.01 0.143 0.129 0.130 0.132 Note: Base Metal Inner Radius = 220.11 cm Base Metal 1/4T = 225.59 cm Base Metal 1/2T = 231.06 cm Base Metal 3/4T = 236.54 cm Base Metal Outer Radius = 242.01 cm Radiation Analysis and Neutron Dosimetry

6-17 Table 6-11 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Seabrook Station Irradiation Time Fluence (E > 1.0 MeV) Iron Displacements Capsule [EFPY] [n/cm 2 ]

[dpa]

U 0.91 3.142E+18 6.156E-03 Y 5.57 1.292E+19 2.518E-02 V 12.39 2.669E+19 5.190E-02 Table 6-12 Calculated Surveillance Capsule Lead Factors Capsule ID And Location Status Lead Factor U (31.50 Dual) Withdrawn EOC 1 3.96 Y (29.0° Dual) Withdrawn EOC 5 3.74 V (29.0° Dual) Withdrawn EOC 10 3.78 X (31.50 Dual) In Reactor 4.11 W (31.50 Single) In Reactor 4.10 Z (31.50 Single) In Reactor 4.10 Note: Lead factors for the capsules remaining in the reactor are based on cycle specific exposure calculations through the most recently completed operating fuel cycle, i.e., Cycle 10.

Radiation Analysis and Neutron Dosimetry

6-18 Figure 6-1 Seabrook Station rO Reactor Geometry with a Dual Capsule at the Core Midplane 240 180 E

x A:

c 120 60 0

0 75 150 225 300 R Axis (cm)

Note: The x-axis is Qo reference angle. Capsule centers are at 29.00 and 31.5o from the reference angle. The neutron pad segment extends 22.50 from the 45o minor axis.

Radiation Analysis and Neutron Dosimetry

6-49 Figure 6-2 Seabrook Station rO Reactor Geometry with a Single Capsule at the Core Midplane 240 180

• 1.E He 120 60 0

0 75 150 225 300 R Axis (cm)

Note: The major x-axis is Oo reference angle. The capsule center is at 31.50 from the reference angle. The neutron pad segment extends 20.00 from the 45o minor axis.

Radiation Analysis and Neutron Dosimetry

6-20 Figure 6-3 Seabrook Station r,0 Reactor Geometry with No Capsule at the Core Midplane 240 180 E

I-.

x 120 60 0

0 75 150 225 300 R AxIs (cm)

Note: The major x-axis is Oo reference angle. The neutron pad segment extends 12.50 from the 45o minor axis.

Radiation Analysis and Neutron Dosimetry

6-21 Figure 64 Seabrook Station rz Reactor Geometry with Neutron Pad 300 200 I

E

-4 x

P4

-100

-200

-300

-400 0 75 150 225 300 Radiation Analysis and Neutron Dosimetry

7-1 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following surveillance capsule removal schedule meets the requirements of ASTM E185-82 [Ref. 11]

and is recommended for future capsules to be removed from the Seabrook Unit I reactor vessel. This recommended removal schedule is applicable to 32 EFPY of operation.

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

U 31.5° Dual 3.96 0.91 3.142 x lo0 (c)

Y 29.0° Dual 3.74 5.57 1.292 x 1019 (c)

V 29.0° Dual 3.78 12.39 2.669 x 1019 (c) 3.48 x i019 (d)

X 31.5° Dual 4.11 15 or 22 (d) 4.91 x 10 9(d)

W 31.50 Single 4.10 Standby (e) (e)

Z 31.50 Single 4.10 Standby (e) (e)

Notes:

(a) Updated in Capsule V dosimetry analysis.

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

(c) Actual plant evaluation calculated fluence.

(d) Capsule X will reach about 15 EFPY and a fluence of approximately 3.48 x 10'9 n/cm 2 (E>l MeV) (t the end. of operating cycle 12 which will support 2 times EOL fluence for a 32 EFPY (40 year) license. However, if a 48 EFPY license renewal is planned, Capsule X may be removed at - 22 EFP'Y (cycle 17) while not exceeding 2 times the 60 year EOL fluence of 5.10 x 10'9 n/cm 2 . It is recomrmended that Capsule X be withdrawn and placed in storage.

(e) Based on the Capsule X withdrawal, it is recommended that the standby capsules be withdrawn within one cycle of the removal of Capsule X. This would allow for meaningful metallurgical data for approximating 60 years of plant operation. It is recommended that these capsules be withdrawn and placed in storage.

Surveillance Capsule Removal Schedule

8-1 8 REFERENCES

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

2. Code of Federal Regulations, 10CFR50, Appendix G FractureToughness Requirements, and Appendix H, Reactor Vessel MaterialSurveillance ProgramRequirements, U.S. Nuclear Regulatory Commission, Washington, D.C.

3. WCAP- 10110, Public Service Company of New HampshireSeabrook Station Unit No. 1 Reactor Vessel Radiation SurveillanceProgram,L.R. Singer, et. al., dated March 1983.

4. DES-NFQA-98-01, Duke Engineering Services Report, Analysis of SeabrookStation Unit I Reactor Vessel Surveillance Capsules Uand K, E.C. Biemiller and GM. Sloan, May 1998.

5. ASTMV[ E1 85-79, StandardPracticeforConductingSurveillance Testsfor Light- Water Cooled Nuclear PowerReactor Vessel, American Society for Testing and Materials.

6.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix Q FractureToughness Criteria for PnrtectionAgainst Failure

7. ASTM E208, StandardTest Methodfor ConductingDrop- Weight Test to Determine Nil-Ductility TransitionTemperature ofFerriticSteels, American Society for Testing and Materials.

8. WCAIP-15745, Seabrook Unit 1 Heatup and Cooldown Limit Curvesfor Normal Operation,T.J.

Laubham, December 2001.

9. CE Report NPSD-1039, Revision 2, Best Estimate Copper and Nickel Vltes in CE Fabricated Reactor Vessel Welds, CEOG Task 902, C-E Owners Group, June 1997.

10. Material Certification Report, Lukens Steel Company, Job No. 733142-005, Plate Heat No. D1136-2, Plate Code No. R-1803-3, J.M. Arnold, July 10, 1975.

11. ASTM E 185-82, StandardPracticefor Conducting Surveillance Testsfor Light- Water Cooled NuclearPowerReactor Vessels, American Society for Testing and Materials.

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

13. Procedure RMF 8102, Tensile Testing, Revision 3.

14. Procedure RMF 8103, Charpy Impact Testing, Revision 2.

15. ASTM E23-02a, StandardTest Methodfor Notched Bar Impact Testing of Metallic Materials, American Society for Testing and Materials.

References

8-2

16. ASTM A370-97a, StandardTest Methods andDefinitionsfor MechanicalTesting ofSteel Products, American Society for Testing and Materials.

17. ASTM E8-04, Standard Test Methodsfor Tension Testing of Metallic Materials, American Society for Testing and Materials.

18. ASTM E21-03 a, StandardTest Methodsfor Elevated Temperature Tension Tests of Metallic Materials,American Society for Testing and Materials.

19. ASTM E83-93, StandardPracticefor Verification and ClassificationofExtensometers,American Society for Testing and Materials.

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

21. WCAP-14040-NP-A, Revision 4, Methodology Used to Develop Cold OverpressureMitigating System Setpoints andRCS Heatup and Cooldown Limit Curves, May 2004.

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

23. RSICC Data Library Collection DLC-185, BUGLE-96, Coupled47Neutron,20 Gamma-Ray Group Cross Section LibraryDerivedfrom ENDFIB-VIfor LWR Shielding andPressure Vessel Dosimetry Applications, March 1996.

References

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

A-1 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 Seabrook Station 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."E[A i 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 +/- 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.l.I Sensor Reaction Rate Determinations In this section, the results of the evaluations of the three neutron sensor sets withdrawn to date as part of the Seabrook Station 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 IEFPYJ U 31.50 End of Cycle 1 0.91 Y 29.00 (dual) End of Cycle 5 5.57 V 29.00 (dual) End of Cycle 10 12.39 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 U, Y,and V are summarized as follows:

Appendix A

A-2 Reaction Sensor Material Of Interest Capsule U Capsule Y Capsule V Copper 6 3 Cu(n,a)60Co X X X 54 Iron Fe(np)5 4 Mn X X X Nickel 58Ni(n,p) 58Co X X X Uranium-238 23 8U(n,f)' 37Cs X X ** X 237 137 Neptunium-237 Np(n,f) Cs X X X 59 Cobalt-Aluminum* Co(ny)60Co X X ** X

• The cobalt-aluminum measurements include both bare and cadmium-covered sensors.

• • Measurements not included in the evaluations as they fell outside the Westinghouse experience base in excess of 3-sigma, indicating a specious measurement result.

Since all of the dosimetry monitors were accommodated within the dosimeter block centered at the radial, azimuthal, and axial center of the material test specimen array, gradient corrections were not required for these reaction rates. 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 do 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 tfie 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 arc!

well known. In particular, the following variables are of interest:

• the :mneasured 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.

Results fro:n the radiometric counting of the neutron sensors from U and W are documented in Reference A-2. The radiometric counting of the sensors from Capsule V was carried out by Pace Analytical Services, Inc., 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 determined by means of a high-resolution gamma spectrometer. For Appendix A

A-3 the copper, iron, nickel and cobalt-aluminum sensors, 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 U, Y,and V was based on the reported monthly power generation of Seabrook Station for Cycles I through 10. 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 U, Y, and V 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:

RA No F Y E-jCjf [I - etJ] [e-Ad]

Pref where:

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

A = Measured specific activity (dps/gm).

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

F = Atom fraction of the target isotope in the target element.

Y = Number of product atoms produced per reaction.

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

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

Cj = Calculated ratio of 4(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).

Appendix A

A-4 and the summation is carried out over the total number of monthly intervals comprising the irradiation period.

In the equation describing the reaction rate calculation, the ratio [Pj]/[Pref] accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fiel 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 Cj 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.

Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 238Uand 27 Np measurements to account for the presence of 23513 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 238U 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 Seabrook Station fission sensor reaction rates are summarized as follows:

Correction Capsule U Capsule Y Capsule V 2 "5U Impurity/Pu Build-in 0.872 0.834 0.787 2 38 U(y,f) 0.966 0.967 0.967 Net 238U Correction 0.842 0.806

• 0.761 237 Np(y,f) 0.990 0.990 0.990

• Uranium measurement results for Capsule Y were not used to support Reference A-2.

These factors were applied in a multiplicative fashion to the decay corrected uranium fission sensor reaction rates.

Appendix A

A-S Results of the sensor reaction rate determinations for Capsules U, Y,and V are given in Tables A4a, 4b, and 4c, respectively. 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 in these tables. The fission sensor reaction rates are listed both with and without the applied corrections for 238U impurities, plutonium build-in, and gamma ray induced fission effects.

Examination of the results in these tables revealed that uranium fission monitor reaction rates for Capsule Y were inconsistent with measurement data obtained from comparable reactors. Similar observations were made for the Capsule Y cadmium covered cobalt-aluminum reaction rates, although it is recognized that non-threshold reactions are involved in these sensors. As a result of these observations, these measurements were not utilized in the least squares adjustment calculations for the capsules.

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, i R, E(0ig ++/- aig )(Og ++/- 5) relates a set of measured reaction rates, Ri, to a single neutron spectrum, f through the multigroup dosimeter reaction cross-section, cOig, each with an uncertainty 5. 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 least squares evaluation of the Seabrook Station surveillance capsule dosimetry, the FERRET code[A-3] 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 (4(E > 1.0 MeV) and dpa) along with associated uncertainties for the three in-vessel capsules withdrawn to date.

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

I - 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.

Appendix A

Ak-6 For the Seabrook Station 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 library.(A4] The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard E 1018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E 706 (IIB)".

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. Tnie 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 Seabrook Station 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 6 3 CU(na)6 OCo 5%

54 Fe(n,p) 54 Mn 5%

"Ni(n,p)"Co 5%

2 38 37 U(n,f) Cs 10%

237Np(n,f)137 Cs 10%

59 Co(n,y)60 Co 5%

These uncertainties are given at the lc level.

Appendix A

A-7 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 multi-group 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.

For sensors included in the Seabrook Station surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.

Reaction Uncertainty

+

63 Cu(nsx)6 0 Co 4.084.16%

54Fe(np)5 4Mn 3.05-3.11%

5 8 Ni(np)5 8 Co 4.494.56%

23 8 U(n,f)' 37 Cs 0.54-0.64%

23 7Np(n,f) 37Cs 10.32-10.97%

59 Co(n,y) 6 0 Co 0.79-3.59%

These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in Light Water Reactor (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.

Appendix A

A-8 While the uncertainties associated with the reaction rates were obtained from the measurement procedu: es 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:

Mgg. =R 2 +Rg *Rg *Pgg where Rn specifies an overall fractional normalization uncertainty and the fractional uncertainties R. and Rg. specify additional random group-wise uncertainties that are correlated with a correlation matrix given by:

Pgg. =[1-07 OeH where H = -g') 2 2y 2 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 5 is 1.0 when g = g', and is 0.0 otherwise.

The set of parameters defining the input covariance matrix for the Seabrook Station calculated spectra was as follows:

Flux Normalization Uncertainty (R,) 15%

Flux Group Uncertainties (R., Rg')

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 29%

(E < 0.68 eV) 52%

Short Range Correlation (0)

Appendix A

A-9 (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 A.1 .3 Comparisons of Measurements and Calculations Results of the least squares evaluations of the dosimetry from the Seabrook Station 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.

A distinction should be made between the Best Estimate/Calculation, or [BE]/[C], ratios and the Measurement/Calculation, or [M]/[C], ratios. In this case, Best Estimate values refer to the combination of calculation and measurement via a least squares adjustment procedure to arrive at the best estimate of the neutron flux (E > 1.0 MeV) with an associated uncertainty. The least squares procedure provides a weighting of calculated and measured input based on the energy response and uncertainty associated with each input parameter. The [BE]/[C] ratios, therefore, represent a comparison of the results of the least squares adjustment with the analytical prediction of the neutron flux (E > 1.0 MeV). The [M]/[C] ratios, on the other hand, provide a direct comparison of actual calculated and measured individual foil reaction rates. Using the [M]/[C] data, a direct comparison of calculated and measured neutron flux (E > 1.0 MeV) can not be made without a suitable weighting of the individual foil results.

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 McV) and iron atom displacements at the surveillance capsule locations is specified as 12% at the Isa level. From Table A-6, it is noted that the corresponding uncertainties associated with the least squares adjusted exposure parameters have been Appendix A

A-10 reduced to 7% for neutron flux (E > 1.0 MeV) and 8% for iron atom displacement rate. Again, the uncertainties from the least squares evaluation arc at the Is 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 4(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 0.86-1.21 for the 14 samples included in the data set.

The overal]l average M/C ratio for the entire set of Seabrook Station data is 1.04 with an associated standard deviation of 9.8%.

In the comparisons of best estimate and calculated fast neutron exposure parameters, the corresponding BE/C comparisons for the capsule data sets range from 0.94-1.04 for neutron flux (E > 1.0 MeV) and from 0.95-1.04 for iron atom displacement rate. The overall average BE/C ratios for neutron flux (E > 1.0 MeV) and iron atom displacement rate are 1.01 with a standard deviation of 5.4% and 1.00 with a standard deviation of 4.4%, 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 Seabrook Station reactor pressure vessel.

Appendix A

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

63 Copper Cu(n,a) 0.6917 5.0- 11.9 5.271 y 54 Iron Fe (n,p) 0.0585 2.1 - 8.5 312.1 d Nickel 58Ni (n,p) 0.6808 1.5 - 8.3 70.82 d Uranium-238 238U (n,f) 1.0000 1.3 - 6.9 30.07 y 6.02 Neptunium-237 ... Np (n,f) 1.0000 0.3 - 3.8 30.07 y 6.17 Cobalt-Aluminum 59Co (ny) 0.0015 non-threshold 5.271 y Note: The 90% response range is defined such that, in the neutron spectrum characteristic of the Seabrook Station 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 Thermal Generation During The First Eleven Fuel Cycles Of The Seabrook Station Reactor (Reactor Power of 3411 MWt for Cycles 1 through 10)

Thermal Thermal Thermal Generation Generation Generation Month Year (MWt-hr) Month Year (MWt-hr) Month Year (MWt-}Ir) 3 1990 33071 3 1993 2537402 3 1996 2537453 4 1990 127021 4 1993 2449287 4 1996 2452203 5 1990 36338 5 1993 2278654 5 1996 25376Ci2 6 1990 566463 6 1993 2455637 6 1996 24557LQ 7 1990 1565543 7 1993 2256834 7 1996 2537879 8 1990 2041581 8 1993 2537509 8 1996 25376E,8 9 1990 2316184 9 1993 1737708 9 1996 2456614 10 1990 2229260 10 1993 1930953 10 1996 2528667 11 1990 1150226 11 1993 2060928 11 1996 245522:9 12 1990 2534639 12 1993 2502022 12 1996 2530869 1 1991 2535476 1 1994 1997765 1 1997 25387'3 2 1991 1669920 2 1994 907870 2 1997 22930E;1 3 1991 2355827 3 1994 2537552 3 1997 2538966 4 1991 1936043 4 1994 580745 4 1997 2452033 5 1991 2535734 5 1994 0 5 1997 708891 6 1991 2012648 6 1994 0 6 1997 71987 7 1991 1642082 7 1994 6114 7 1997 2508503 8 1991 0 8 1994 2381880 8 1997 2537484 9 1991 0 9 1994 2455597 9 1997 2365268 10 1991 882592 10 1994 2541007 10 1997 2538688 11 1991 2454605 11 1994 2458300 11 1997 2455554 12 1991 2444608 12 1994 2537531 12 1997 377665 1 1992 2534000 1 1995 2535750 1 1998 1157471 2 1992 2373536 2 1995 2293874 2 1998 2291800 3 1992 2536703 3 1995 2537703 3 1998 2536574 4 1992 2451882 4 1995 2452508 4 1998 2453798 5 1992 2537326 5 1995 2536759 5 1998 2538922 6 1992 2451366 6 1995 1470050 6 1998 823532 7 1992 2537068 7 1995 2211297 7 1998 1659121 8 1992 2524004 8 1995 2530752 8 1998 2538922 9 1992 356668 9 1995 2455018 9 1998 2457267 10 1992 0 10 1995 2540835 10 1998 2545859 11 1992 913663 11 1995 192874 11 1998 1401145 12 1992 2237482 12 1995 1498045 12 1998 231109.2 1 1993 1870107 1 1996 2219912 1 1999 252348?

2 1993 2291849 2 1996 1729081 2 1999 2286553 Appendix A

A-13 Table A-2 cont'd Monthly Thermal Generation During The First Eleven Fuel Cycles Of The Seabrook Station Reactor (Reactor Power of 3411 MWt for Cycles I through 10)

Thermal Thermal Thermal Generation Generation Generation Month Year (MWt-hr) Month Year (MWt-hr) Month Year (MWt-hr) 3 1999 2098281 3 2002 2537261 3 2005 2320022 4 1999 0 4 2002 2452671 4 2005 6 5 1999 1136402 5 2002 221101 End of Cycle 10 6 1999 2454228 6 2002 2300450 7 1999 2537590 7 2002 2537434 8 1999 2537590 8 2002 2537360 9 1999 2455221 9 2002 2455438 10 1999 2541350 10 2002 2540688 11 1999 2456066 11 2002 2455452 12 1999 2536872 12 2002 2537330 1 2000 2247080 1 2003 2537341 2 2000 2367593 2 2003 2291839 3 2000 2535922 3 2003 2537349 4 2000 2450384 4 2003 2452116 5 2000 2535500 5 2003 2537347 6 2000 2188492 6 2003 2455417 7 2000 2537189 7 2003 2537360 8 2000 2540146 8 2003 2537086 9 2000 2458833 9 2003 2424603 10 2000 1506515 10 2003 428942 11 2000 0 11 2003 2136981 12 2000 0 12 2003 2537377 I 2001 5464 1 2004 2536808 2 2001 2290454 2 2004 2373586 3 2001 1375432 3 2004 2537323 4 2001 2452144 4 2004 2452078 5 2001 2470138 5 2004 2537313 6 2001 2453937 6 2004 2455393 7 2001 2534771 7 2004 2535478 8 2001 2464421 8 2004 2537354 9 2001 2456046 9 2004 2455433 10 2001 2133594 10 2004 2540662 11 2001 2367786 11 2004 2455355 12 2001 2536839 12 2004 2537232 1 2002 2537472 1 2005 2537167 2 2002 2291719 2 2005 2291629 Appendix A

A-14 Table A-3 Calculated Cj Factors at the Surveillance Capsule Center Core Midplane Elevation Fuel *(E > 1.0 MeV) [n/cm2 -sI

_Cycle EFPY Capsule U Capsule Y Capsule V 1 0.91 1.09E+11 1.OOE+11 1.00E+11 2 0.87 8.03E+10 8.03E+10 3 1.21 7.97E+10 7.97E+10 4 1.21 5.61E+10 5.61E+10 5 1.37 6.13E+10 6.13E+10 6 1.49 6.10E+10 7 1.41 6.79E+10 8 1.18 6.08E+10 9 1.35 6.51E+10 10 1.40 6.48E+10 Average (EFPY-weighted) 1.09E+11 7.35E+ 10 6.83E+ 10 EFPY Capsule U Capsule Y Capsule V 1 0.91 1.00 1.36 1.47 2 0.87 1.09 1.18 3 1.21 1.08 1.17 4 1.21 0.76 0.82 5 1.37 0.83 0.90 6 1.49 0.89 7 1.41 1.00 8 1.18 0.89 9 1.35 0.95 10 1.40 0.95 Average (EFPY-wcighted) 1.00 1.00 1.00 Appendix A.

A-15 Table A-4a Measured Sensor Activities And Reaction Rates Surveillance Capsule U Corrected Measured Saturated Reaction Reaction Average Reaction Location Activity Activity Rate Rate Reaction (dps/g) (dps/g) (rps/atom) (rps/atoe) Rate 63Cu (na) 6 0Co (Cd) Top 5.15E+04 4.61E+05 7.03E-17 Middle 4.67E+04 4.18E+05 6.37E-17 Bottom 4.60E+04 4.11E+05 6.26E-17 6.55E-17 6.55E-17 14 Fe (n,p) 14Mn Top 2.13E+06 4.3 1E+06 6.84E-15 Middle 1.92E+06 3.88E+06 6.1SE-15 Bottom 1.87E+06 3.78E+06 5.99E-15 6.32E-15 6.32E-15 58 Ni (n p) 5 8Co (Cd) Top 5.39E+07 6.46E+07 9.24E-15 Middle 4.97E+07 5.96E+07 8.53E-15 Bottom 4.88E+07 5.85E+07 8.37E-15 8.71E-15 8.71E-15 238U (nf) 137Cs (Cd) Middle 1.25E+05 6.03E+06 3.96E-14 3.96E-14 3.34E-14 237Np (n,f) 137Cs (Cd)

Middle 9.95E+05 4.79E+07 3.06E-13 3.06E-13 3.02E-13 59 Co (n,y) 60Co Top 1.05E+07 9.43E+07 6.15E-12 Middle l.lOE+07 9.84E+07 6.42E-12 Bottom 1.18E+07 1.06E+08 6.90E-12 6.49E-12 6.49E-12

' 9Co (n,y) 60Co (Cd) Top 5.65E+06 5.05E+07 3.30E-12 Middle 5.46E+06 4.88E+07 3.19E-12 Bottom 5.93E+06 5.30E+07 3.46E-12 3.31E-12 3.31E-12 Notes: 1) Measured specific activities are indexed to a counting date of July 25, 1991.

2) The corrected average 23 8U (n,f) and 238 Np (n,f) reaction rates account for the plutonium build-in and photo-fission effects described in Section A. 1.1.

Appendix A

A-16 Table A-4b Measured Sensor Activities And Reaction Rates Surveillance Capsule Y Aveage Correctedc Measured Saturated Reaction Reaction ARerage Reaction Location Activity Activity Rate Rate Reaction (dps/g) (dpslg) (rpslatom) (rps/atom) (rps/atoweL 63Cu (n, a) 60Co (Cd) Top 2.38E+05 5.07E+05 5.30E-17 Middle 2.04E+05 4.34E+05 4.54E-17 Bottom 2.10E+05 4.49E+05 4.69E-17 4.84E-17 4.84E-17

' 4Fe (n,p) 54Mn Top 4.3 1E+07 5.65E+07 5.06E-15 Middle 3.87E+07 5.07E+07 4.54E-15 Bottom 3.81E+07 4.99E+07 4.47E-15 4.69E-15 4.69E-I';

"Ni (n,p) "Co (Cd) Top 6.57E+07 8.02E+07 7.71E-15 Middle 5.89E+07 7.18E+07 6.91E-15 Bottom 5.79E+07 7.06E+07 6.80E-15 7.14E-15 7.14E-15 23sU (ni) 137Cs (Cd) Middle 3.91E+05 3.31E+06 2.17E-14 2.17E-14 1.75E-14 237 Np (n,f) '7Cs (Cd) Middle 3.68E+06 3.11E+07 1.99E-13 1.99E-13 1.97E-13 Co (n,) 60 Co Top 1.77E+10 3.78E+10 3.70E-12 Middle 1.811E+10 3.87E+10 3.78E- 12 Bottom 1.73E+10 3.68E+10 3.60E-12 3.69E-12 3.69E-12

' 9Co (n,() 60Co (Cd) Top 8.27E+08 1.76E+09 1.73E-13 Middle 8.45E+08 1.80E+09 1.76E-13 Bottom 9.07E+08 1.94E+09 1.89E-13 1.79E-13 1.79E-13 Notes: 1) Measured specific activities are indexed to a counting date of May 10, 1997.

2) The corrected average 238 U(n,f) and 238Np (n,f) reaction rates account for the plutonium build-in and photo-fission effects described in Section A. 1.1.

3) The uranium fission monitor and cadmium covered cobalt-aluminum reaction rates are inconsistent with measurement data obtained from comparable reactors. These measurements were not utilized in the least squares adjustment calculations.

Appendix A

A-17 Table A-4c Measured Sensor Activities And Reaction Rates Surveillance Capsule V Average Corrected Measured Saturated Reaction Reaction vreactie Reaction Location Activity Activity Rate Rate Reaction (dps/g) (dps/g) (rps/atom) (rps/atom) Rate (rps/tom)(rpslatom) 6 3Cu (na) 60CO (Cd) Top 1.98E+05 2.94E+05 4.49E-17 Middle 1.74E+05 2.59E+05 3.95E-17 Bottom 1.73E+05 2.57E+05 3.92E-17 4.12E-17 4.12E-17 14Fe (n,p) 14Mn Top 1.79E+06 2.93E+06 4.64E-15 Middle 1.61E+06 2.64E+06 4.18E-15 Bottom 1.59E+06 2.60E+06 4.13E-15 4.32E-15 4.32E-15 5 8Ni (np) 58 Co (Cd) Top 7.40E+06 4.5 1E+07 6.46E-15 Middle 6.77E+06 4.13E+07 5.91E-15 Bottom 6.65E+06 4.06E+07 5.81E-15 6.06E1-15 6.06E-15 2 38 U(n,f) 37Cs (Cd) Middle 1.13E+06 4.73E+06 3.11E-14 3.11E-14 2.37E-14 23 7Np (n,f) '37Cs (Cd) Middle 8.08E+06 3.38E+07 2.16E1-13 2.16E-13 2.14E-13 59 Co (n,y) 60 Co Top 3.41E+07 5.07E+07 3.31E-12 Middle 3.43E+07 5.1 OE+07 3.33E-12 Bottom 3.42E+07 5.0813+07 3.32E-12 3.32E-12 3.32E-12 5

"Co (n,y) 60Co (Cd) Top 1.77E+07 2.63E+07 1.72E-12 Middle 1.85E+07 2.75E+07 1.79E-12 Bottom 1.87E+07 2.78E+07 1.81E-12 1.77E-12 1.77E-12 Notes: 1) Measured specific activities are indexed to a counting date of September 24, 2005.

2) The corrected average 23 8U (n,f) and 23gNp (n,f) reaction rates account for the plutonium build-in and photo-fission effects described in Section A. 1.1.

Appendix A

A-18 Table A-5 Comparison of Measured, Calculated, and Best Estimate Reaction Rates At The Surveillance Capsule Center Reaction Rate irps/atoml Best Reaction Measured Calculated I Estimate M/C M/BE Capsule U 3

Cu(n,a)60 Co 6.55E-17 5.57E-17 6.27E- 17 1.18 1.04 5 4Fe(n,p) 5 4Mn 6.32E-15 6.35E-15 6.43E-15 1.00 0.98 58 Ni(np)58 Co 8.71E-15 8.93E-15 8.91E-15 0.98 0.98 238 tj (nf) 137Cs (Cd) 3.33E-14 3.46E-14 3.34E-14 0.96 1.00 237 Np (n,f) 37Cs (Cd) 3.02E-13 3.43E-13 3.12E-13 0.88 0.97 5"Co (ny) 60CO 6.49E-12 4.93E-12 6.35E-12 1.32 1.02 59CoD (n y) 60 Co (Cd) 3.31E-12 3.43E-12 3.36E-12 0.97 0.99 Capsule Y 3 Cu(na)6 0 Co 1 4.84E-17 4.01E-17 4.74E-17 1.21 1.02

-4 Fe(n,p) 5 4 Mn 4.69E- 15 4.40E- 15 4.85E-15 1.07 0.97

8Ni(n p)58Co 7.14E- 15 6.16E- 15 6.90E- 15 1.16 1.03 237 Np (nf) 37Cs (Cd) 1.97E-13 2.29E-13 2.15E-13 0.86 0.92 "Co (n,y) 60 CO 5

3.69E-12 3.20E-12 3.68E-12 1.15 1.00 Capsule V 6 3 Cu(na) 6 0 Co 4.12E-17 3.79E- 17 4.08E-17 1.09 1.01 5"Fe(n,p)"Mn 4.31E-15 4.12E-15 4.35E-15 1.05 0.99

8Ni(np) 5 8Co 6.06E- 15 5.76E- 15 6.08E-15 1.05 1.00 238U'(n, f) 37Cs (Cd) 2.36E-14 2.19E-14 2.29E-14 1.08 1.03 237Np

(n,f) 37Cs (Cd) 2.14E-13 2.12E-13 2.17E-13 1.01 0.99 5

S'Co (ny) 60Co 3.32E-12 2.95E-12 3.26E-12 1.13 1.02 59 Co (n,y) 60 Co (Cd) 1.77E-12 2.07E-12 1.81E-12 0.86 0.98 Appendix A

A-19 Table A-6 Comparison of Calculated and Best Estimate Exposure Rates At The Surveillance Capsule Center

• (E > 1.0 MeV) [n/cm2 -sJ Best Uncertainty Capsule ID Calculated Estimate (iy) BE/C U 1.1OE+11 1.04E+11 6% 0.94 Y 7.42E+10 7.65E+10 7% 1.03 V 6.89E+10 7.18E+10 6% 1.04 Iron Atom Displacement Rate [dpa/s]

Best Uncertainty Capsule ID Calculated Estimate (1a) BE/C U 2.13E-10 2.03E-10 8% 0.95 Y 1.43E-10 1.44E-I0 8% 1.01 V 1.32E-10 1.37E-10 7% 1.04 Note: Calculated results arc based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period.

Appendix A

A-20 Table A-7 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions M/C Ratio Reaction Capsule U Capsule Y Capsule V 63 Cu(na)6 0Co (Cd) 1.18 1.21 1.09 14Fe(np)4 Mn 1.00 1.07 1.05 58 Ni(np)58 Co (Cd) 0.98 1.16 1.05 23 8U (nf) 137Cs (Cd) 0.96 1.08 23 7Np (nf) 137Cs (Cd) 0.88 0.86 1.01 Average 1.00 1.07 1.05

% Standard Deviation 10.9 14.30 2.89 Note: The overall average M/C ratio for the set of 14 sensor measurements is 1.04 with an associated standard deviation of 9.8%.

Appendix A

A-21 Table A-8 Comparison of Best Estimate/Calculated (BE/C) Exposure Rate Ratios BE/C Ratio Capsule ID *(E > 1.0 MeV) dpa/s U 0.94 0.95 Y 1.03 1.01 V 1.04 1.04 Average 1.01 1.00

% Standard Deviation 5.4 4.4 Appendix A

A-22 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 2001.

A-2 Westinghouse Calculation Note CN-REA-03-38, "Seabrook Unit I (NAH) RPV Fluence Evaluation for 7.4% Uprate Program," September 2003.

A-3 A. Schmittroth, FERRETDataAnalysisCore, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

A-4 R.SIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section Compendium," July 1994.

Appendix A

13-0 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS INSTRUMENTED CHARPY IMPACT TEST CURVES

• Specimen prefix "KL" denotes Lower Plate, Longitudinal Orientation

• Specimen prefix "KT" denotes Lower Plate, Transverse Orientation

• Specimen prefix "KW" denotes Weld Material

• Specimen prefix "KH" denotes Heat-Affected Zone material Appendix B

I B-1 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)

L21, -50 0 F 5000.00-4000.00-3000.00 -

2000.00 -

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

KLI9, 0F 5000.00 4000.00

, 3000.00

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

KL16, 25 0 F Appendix B GON

I B-2 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tke-1 (ms)

KL20, 50°F 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tmne-I (ms)

KL25, 500 F 40D.00 3000.00-2000.00 1000.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Trme-1 (Ms)

KL29, 750 F I_

ca5 I Apppendix B

B-3 5w0.00-4000.00-

'7 300.00-2000.00:

1oW.OO 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

KL24, 100 0 F 5000.00 4000.00 a

X 3000.00

-j 2000.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tim-L1(ms)

KL30, 125°F 5000.00 4000.00 A

' 3000.00 2000.00 1000.00

[,,,I i 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

KL27, 150 0 F Appendix B CoC

11-4 5000.00 4000.00 3000.00]

2000.00 1000.00 nnn .

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

KL17, 1750 F e Tvm4 (ms)

KL22, 225 0 F 3.00 6.00 Trne-I (ms)

KL23, 300 0F Appendix B

B-5 0

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

KL18, 3500 F 5000.00 4000.00 3 3000.00 2000.00 1000.00 0.00 -I 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

KT28, -50 0 F 5000.00 4000.00

' 3000.00 20

-J 2000.00-1000.00 0.00-0.00 1.00 2.0 0 3.00 4.00 5.00 6.00 Time-I (ms)

KT17, 0F Appendix B C07

U v

B-6 5000.00 4000.00 3000.00 20W.00 1000.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

KT19, 500 F 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

KT27, 750 F 500.001-4000.001 X 3000.00-2000.00' 1000.00' 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tine-1 (ns)

KT25, 100 0 F o Co Appendix B

B-7 500.0O.

4000.00 .

a 3000.00 2000.00 1000.00 0.00 g . _ I , I I t o.l 00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

KT18, 1250 F 5000.00 4000.00

, 3000.00

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

KT23, 1500 F 5000.00 4000.00 -

i 3000.00 .

o

-J 2 0 00

.00 .

1000.00 'J

.00 3.00 6.00 Time-I (Ms)

KT26, 20 0 °F Appendix B c-0

13-8 S0OO 4000fl0 i 300000 2000.00 1000.00 0.00 1.00 2 .00 3.00 4.00 SW0 6.00 Tine-I (ms)

KT22, 250 0 F 5000.00 4000.00 i 3000.00 2000.00 1000.00 nm, 0oWo 1.00 2.00 3.00 4.00 5.00 6.00 Tine-1 (Ms)

KT29, 3000 F 5000.00 4000.00 00 2000.00 OOO.

1 1000.001 nm, , , ,a 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Trne-I (ms)

KT24, 3500 F Appendix B

B-9 4a0 2000.00

-I M.00, 1000,00-000 t___ -

1O0 2.00 300 4.00 SDO 6.00 Tyro-i (ns)

KW20, -90 0 F 5000.00-4000.00-8 3000.00 2000.00 1000.00 Time-1 (ms)

KW23, -50 0 F 5000.00 4000.00 n

X 3000.00 2000.00 1000.00 1

wII ,

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

W30, -35 0 F CAd Appendix B

I B-10 5000.00 4000.00 3000.00 2000.00 1000.00 n o - t IIl - I 0.00 1 .0 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

KW24, -25 0 F 5000.0*

4000.00

- 3000.00 2000.00 1000.00.

al .w 11UI I 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

KW16, -25 0F 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

KW27, 0F Appendix B c UI

I B-lI 500.00-4000.00-a 3000.00 2000.0 1000.00

[n [I, . i 0.00 1 .00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

KW18, 25-F 5000.001 4000.00-xs 3000.00-am' 0

-J 2000.00-1000.00-0.00 0.1 Time-1 (ms)

KW28, 50 0 F 500.00 -

4000.00

3000.00*

aD 0

-J 2000.00 1000.00 00 3.00 6.00 Time-I (ms)

KW22, 75-F C B Appendix B

W B-12 5M.000 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 Time-1 (ins)

KW26, 1000 F 4000.00 3000.00 2000.00}

1000.00l , , , , , _

0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 rime-I (Ms)

KW19, 250 0 F 40.000 3000.00 1000.00 0.00 o00 1s0 2.0o 3.0 4.o0 5.00 6.00 rowe. (me)

KH20, -90 0 F Appendix B C(-3

B-13 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-1 (ms)

KH24, -750 F 5000.00 4000.00 -

300.00 2000.00 -

1000.00 I 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

KH16, -50 0F 5000.00 400000 3000.00 2000.00 -

10W.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-i (ms)

KH23, -25 0F Appendix B 6AAT

I-B-14 5000.00 4000.00 3000.00

-j 2000.00 1000.00 0.00 0.t 30 1.00 2.00 3.00 4.00 5.00 6.00 Tine-1 (ms)

KH22, 0F 5000.00 4000.00 a

i 3000 00

-j 2000.00 1000.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ims)

KH25, 25-F 00 3.00 6.00 Tine- (ms)

KH30, 250 F Appendix B

(-r;

I B-15 x 3000.00-0

-J 2000.00-1000.00W 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

KH21, 50°F 50.000-4000.00-A 7 3000.00 a

0

-J Time-1 (ms)

KH18, 100 0F 5000.00 4000.00 A

013000.00 05

-j 2000.00 1000.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (rns)

KH28, 125 0F Appendix B cClC

B-16 50.00 400.0 3000 M

200.0 1000.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tme-i (ms)

KH29, 1500 F 5000.0 4000.00 a 3000.00 2000.00 1000.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tm-e-1 (ms)

KH26, 250 0 F 5000.00 4000.00 i 3000.X 2000.00 1000.00 "M . l @ l l- l ll 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tkme-1 (Ms)

K1117, 300 0 F Appendix B

C-o 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 E185-82, Section 4.18, and reads as follows:

"uppershelf energy 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."

Westinghouse typically reports the average of all Charpy data 2 95% shear as the USE. In some instances, there may be data deemed 'out of family' and are removed from the determination of the USE based on engineering judgement. The USE values reported in Table C-l 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 Material Unirradiated Capsule U Capsule Y Capsule V Lower Shell Plate 125 ft-lbs 118 ft-lbs 107 ft-lbs 112 ft-lbs RI 808-3 (Long.)

Lower Shell Plate 79 ft-lbs 73 ft-lbs 66 ft-lbs 69 fl-lbs RI 808-3 (Trans.)

(Heat #4P6052) 160 ft-lbs 144 ft-lbs 144 ft-lbs 156 ft-lbs Heat Affected Zone 130 ft-lbs 134 ft-lbs 113 ft-lbs 130 ft-lbs Material Appendix C

(-2 UNIRRADIATED LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 01:25 PM Page I Coefficients of Curve I A 63.6 B = 61.A C - 62.35 T0 10.17 D = O.OOE+00 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-28.1 Deg F Ternp50 ft-lbs-3.8 Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: LT Capsule: UNIRRA Fluence: n/cmr2 300 250

,- 200

150

100 0 50 n

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

-60.00 10.00 13.90 -3.90

- 60. 00 12. 00 13. 90 - I. 90

-30. 00 18. 00 2S. 74 .10.74

-30. 00 26. 00 28. 74 -2. 74

- 30. 00 38.00 28. 74 9.26

.o 0 49. 00 53. 68 -4.68

. 00 50. 00 53. 68 -3. 68

.o 0 70. 00 53.68 16.32

40. 00 85.00 90. 93 -5.93

- I Appendix C

C-3 UNIRRADIATED LOWER SHELL PLATE RI 808-3 (LONGITUDINAL)

Page 2 Plant: Seabrook I Material: SA533BI Heat: RI808-3 Orientation: LT Capsule: LNIRRA Fluence: n/cmA2 Charpy V-Notch Data Ternperature Input CVN Computed CVN Differential

40. 00 97. 00 90. 93 6.07

75. 00 90. 00 111.36 -21. 36 75.00 124. 00 111.36 12.64 1 1 0. 0 0 1 1 7. 0 0 120.20 - 3.20 1 1 0. 0 0 126. 00 120. 20 5.80 200. 00 123. 00 124. 72 -1. 72 200. 00 127. 00 124. 72 2.28 200. 00 127. 00 124. 72 2.28 300.00 124. 00 124.99 -. 99 Correlation Coefficient - .981 Appendix C

C-4 CAPSULE U LOWER SHELL PLATE RI 808-3 (LONGITUDINAL)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 01:25 PM Page I Coefficients of Curve 2 A = 60.1 B = 57.9 C - 55.93 TO = 43.81 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=l 18.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Ternp@30 ft-lbs=l 1.6 Deg F Ternp@50 fl-lbs--34.0 Deg F Plant: Seabrook I Material: SA533BI Heat: R180S-3 Orientation: LT Capsule: U Fluence: nCM^A2 300 250 a, 200 0

P 150 1-1 z

Z; 100 D_____A _____ ___._____ ____

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

- 40. 00 I1. 00 7. 71 3. 29

. 00 28.00 22.20 5. 80

20. 00 34.00 36.84 - 2. 84

40. 00 62.00 56. 17 5. 83

40. 00 52.50 56. 17 -3. 67 7 0. 00 102.50 85.39 17. 11

70. 00 60.00 85. 39 -25. 39

90. 00 88.00 99.37 - I 1.37 100.00 1 7.50 104.31 13. 19

-=1 I Appendix C

C-5 CAPSULE U LOWER SHELL PLATE RI 808-3 (LONGITUDINAL)

Page 2 Plant: Seabrook I Material: SAS33BI Heat: RI 808-3 Orientation: LT Capsule: U Fluence: n/cm'2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 125.00 122. 00 111. 9 8 10. 02 12 5. 00 1 20. 00 111.98 8. 02 150. 00 1 17.00 1 15.46 1.54 225.00 I 12. 5 0 117.82 -5. 32 325.00 I 1 2. 0 0 I 1 8. 0 0 -6. 00 550.00 121.50 I 1 8. 0 0 3.5 0 Correlation Coefficient - .962 Appendix C

C-6 CAPSULE Y LOWER SHELL PLATE RI 808-3 (LONGITUDINAL)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 01:25 PM Page I Cocfficicnts of Curve 3 A = 54.6 B -52.4 C = 63.25 TO = 51.25 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper ShelfEnergy=I07.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Tcmp@30 ft-lbs=l 9.1 Deg F Temp@50 ft-lbs=45.7 Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: LT Capsule: Y Fluence: n/cMA2 300 250 aJ200 0;150

-X_ - -1 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 Diffcrential

-40. 00 8. 50 7.74 .76

.00 18. 50 19.50 -1. 00 25.00 42. 00 34. 02 7. 98

50. 00 55.00 53. 56 1.44

68. 00 73. 50 68. 16 5. 34

71. 00 50. 00 70.45 - 20. 45 100. 00 92. 00 88.52 3. 4 8 125.00 103. 00 97.72 5. 28 150.00 112. 50 102.58 9.92

_ I Appendix C

C-7 CAPSULE Y LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

Page 2 Plant: Seabrook 1 Material: SA533BI Hcat: RI 808-3 Orientation: LT Capsule: Y Fluence: nrcmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 175. 00 107.00 104.95 2.05 200.00 111.50 106.06 5. 4 4 200. 00 110.50 106.06 4.44 250.00 103.50 106.80 -3. 30 300. 00 104.50 106.96 -2. 46 350.00 102.00 106.99 -4. 99 Correlation Coefficient - .979 Appendix C

(-8 CAPSULE V LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 01:25 PM Page I Coefficients of Curve 4 A - 57.1 B = 54.9 C = 84.65 TO = 78.06 D = O.OOE+00 Equation is A + B I [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=1 12.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Ternp@30 fl-lbs=32.3 Deg F Temp@50 h-lbs=67.1 Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: LT Capsule: V Fluence: n/cmA2 300 250

, 200 Za 0

P 150 z

8 100 50

4.4 0

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

-50. 00 7.00 7.28 -. 28

.00 12.00 17.19 -5. 19

25. 00 30.00 26.58 3.42 50.00 48.00 39.54 8.46

50. 00 40.00 39.54 .46

75. 00 41.00 55.12 -14.12

75. 00 65. 00 55. 12 9. 88 I 0. 00 67.00 71.02 -4. 02 125.00 8l.00 84.77 -3.77 Appendix C

C-9 CAPSULE V LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

Page 2 Plant: Seabrook 1 Material: SA533B1 Heat: RI 808-3 Orientation: LT Capsule: V Fluencc: n/cMA2 Charpy V-Notch Data Temperaturc Input CVN Computed CVN Differential 150.00 98.00 95.04 2.96 175. 00 102.00 101.91 . 09 225. 00 Ill. 00 108. 69 2. 31 250. 00 123. 00 110.14 12. 86 300. 00 122. 00 111.42 10.58 350. 00 114.00 111.82 2. 18 Correlation Coefficient =.986 Appendix C

c-i UNIRRADIATED LOWER SHELL PLATE RI 808-3 (LONGITUDINAL)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:04 PM Page 1 Coefficients of Curve 1 A = 40.6 B = 39.6 C = 70.98 TO = 3.76 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=80.2 Lower Shelf L.E.=1.O(Fixed)

Tetnp.@L.E. 35 mils-=6.3 Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: LT Capsule: UNIRRA Fluence: n/cm\2 200 150

.I I I . 1 5&100 5i 50 0

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

- 60. 00 18. 00 12.27 5.73

- 60. 00 14. 00 12.27 1 .7 3

- 30. 00 18.00 23. 07 -5. 07

-30. 00 I8. 00 23. 07 -5. 07

-30. 00 25.00 2 3. 07 I .9 3

.00 3 5. 00 38.50 -3. 50

.00 3 5. 00 38.50 -3. 50

.00 48.00 38.50 9. 50 40.00 58.00 59. 23 - 1.23

-I__

Appendix C

C-lI UNIRRADIATED LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

Page 2 Plant: Scabrook I Material: SA533B1 Heat: RI 808-3 Orientation: LT Capsule: UN'IRRA Flucncc: n/cmA2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

40. 00 61. 00 59. 23 1. 77 75.00 63. 00 70. 82 -7. 82 75.00 79. 00 70. 82 8. 18 110.00 72. 00 76. 42 -4.42 110.00 80. 00 76. 42 3. 58 200. 00 79. 00 79. 89 - . 89 200.00 80. 00 79. 89 .11 200. 00 84. 00 79. 89 4. 11 300. 00 77.00 80. 18 -3.18 Correlation Coericient - .983 Appendix C

C-i2 CAPSULE U LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/1212005 02:04 PM Page 1 Coefficients of Curve 2 A =42.09 B = 41.09 C = 63.34 TO 35.58 D e O.OOE+00 Equation is A + B * [Tanh((T-ToY(C+DT))]

Upper Shelf LE.=83.2 Lower Shelf LE.=1.0(Fixed)

Ternp.@L.E. 35 niuls=24.6 Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: LT Capsule: U Fluence: nlcmA2 200 150 E

.2 a 100 5l 1,13 1-- a B~_______

50 T 13 1 0

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

-40. 00 9. 00 7. 92 1. 08

. 00 25.00 21.16 3.84

20. 00 31.00 32. 18 - 1. 18

40. 00 47. 00 44. 95 2. 05

40. 00 42. 00 44.95 -2. 95

70. 00 71. 00 62.45 8.55

70. 00 49. 00 62.45 - 13. 45

90. 00 68. 00 70.68 -2. 68 1 00. 00 77. 00 73. 67 3.33

- I Appendix C

C-13 CAPSULE U LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

Page 2 Plant: Seabrook I Material: SA533BI Heat: RI 808-3 Orientation: LT Capsule: U Fluencc: n/eMA2 Charpy V-Notch Data Temperature Input LE. Computed LE. Differential 125.00 85.00 78.57 6.43 125.00 83.00 78.57 4.43 150. 00 84.00 81.01 2. 99 225.00 76. 00 82. 97 - 6. 97 325.00 85.00 83. 16 1. 84 550. 00 78. 00 83. 17 - 5. 17 Correlation Coefficient =.974 Appendix C

C-i4 CAPSULE Y LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

CVGRAPH 5.0.2 Hyperbolic Tangent Curvc Printed on 12/12/2005 02:04 PM Page I Coefficients of Curve 3 A - 39.35 B = 38.35 C = 63.27 TO = 49.07 D = 0.OOE+00 Equation is A + B * [Tanh((T-Toy(C+DT))]

Upper Shelf L.E.=77.7 Lower Shelf LE.=1.0(Fixed)

Temp.@L.E. 35 rnils=41.9 Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: LT Capsule: Y Fluence: n/cnA2 200 150 .

'i a 100 50

.................. ,., ,-s I. , .

A

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

-40. 00 2.0 0 5.33 -3. 33

.00 13. 00 14. 42 -1. 42

25. 00 30. 00 25.43 4. 57

50. 00 41.00 39. 92 1.08

68. 00 55. 00 50. 50 4. 50

71. 00 42. 00 52. 14 -10. 14 100.00 65.00 64. 93 .07 125. 00 75.00 71.33 3.67 150. 00 76.00 74. 68 1. 32

-___ _ I Appendix C

c-I15 CAPSULE Y LOWER SHELL PLATE RI 808-3 (LONGITUDINAL)

Page 2 Plant: Seabrook 1 Material: SA533B1 Heat: RI 808-3 Orientation: LT Capsule: Y Fluence: nlcm'2 Charpy V-Notch Data Tcrmperature Input L.E. Computed L.E. Differential 175. 00 75.00 76.30 - 1.30 200. 00 77.00 77.07 - . 07 200. 00 8I. 00 77.07 3. 93 250. 00 75.00 77.58 -2. 58 300. 00 77.00 77.68 - . 68 350. 00 76.00 77.70 - 1.70 Correlation Coefficient = .989 m m Appendix C

C-16 CAPSULE V LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12112/2005 02:04 PM Page I Cocfficients of Curve 4 A = 39.85 B - 38.85 C = 90.62 TO - 73.36 D - O.OOE+00 Equation is A + B * [Tanh((T-ToY(C+DT))]

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

Temp.@L.E. 35 minls=62.0 Deg F Plant: Seabrook I Material: SA533B I Heat: RI 808-3 Orientation: LT Capsule: V Fluence: n/cMA2 200 150 + t E

8 100 E

A~

50 ,,a,,

. ..-- /-TI i

An-

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

-50. 00 6.00 5. 7 9 .21

. 00 13. 00 1 3. 8 5 - . 85

25. 00 26.00 20. 88 5. 12

50. 00 3 5. 00 3 0. 05 4. 95

50. 00 29. 00 30.05 -1.05 7 5. 00 33. 00 40.55 -7. 5 5

75. 00 40.00 40. 55 - .55 1 00. 00 46. 00 50. 95 -4. 95 125.00 62.00 59. 87 2.13

-I__

Appendix C

C-17 CAPSULE V LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

Page 2 Plant: Scabrook I Material: SA533BI Heat: RI 808-3 Orientation: LT Capsule: V Fluence: n/cMA2 Charpy V-Notch Data Temperature Input L.E. Computed LE. Differential 150.00 73.00 66.61 6.39 175.00 74.00 71.25 2.75 225. 00 72.00 76.06 -4.06 250. 00 80.00 77.16 2.84 300.00 77. 00 78. 18 -1.18 350.00 76.00 78.53 -2.53 Correlation Cocfficient - .988 Appendix C

C-l8 UNIRRADIATED LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:22 PM Page 1 Coefficients of Curve I A=50. B=50.C=70.68 TO=18.8 D=0.O0E+00 Equation is A + B * [Tanh((T-To)I(C+DT))]

Temperature at 50% Shear = 18.8 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: LT Capsule: UNIRRA Fluence: n/cm^2 125 100 75 L.

C4 0

50 25 o I-

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

- 60.00 18.00 9.71 8.29

. 60. 00 18.00 9.7 1 8.29

-30. 00 18. 00 20. 09 -2. 09

- 3 0. 00 25. 00 20.09 4.9 1

- 30.00 18. 00 20. 09 -2. 09

.00 34. 00 37.01 -3. 01

. 00 30.00 37. 01 -7. 01

.00 40. 00 3 7. 0 1 2. 9 9

40. 00 56.00 64.56 -8. 56 Appendix C

c-I9 UNIRRADIATED LOWER SHELL PLATE RI 808-3 (LONGITUDINAL)

Page 2 Plant: Seabrook I Material: SA533BI Heat: RI808-3 Orientation: LT Capsule: UNIRRA Fluencc: n/crnA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

40. 00 64. 00 64. 5 6 -. 56

75. 00 79. 00 83. 07 -4. 07 7 5. 00 100. 00 83. 07 16. 9 3 110. 00 91.00 92.96 -1. 96 1 10.00 100. 00 92. 96 7.04 200.00 100. 00 99.41 .59 200. 00 100. 00 99. 41 .59 200.00 100. 00 9 9. 4 1 . 59 300. 00 100. 00 99. 97 . 03 Correlation Coefficient = .985 Appendix C

C-20 CAPSULE U LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:22 PIM Page I Coefficients of Curve 2 A=50. B=50.C = 51.03 T0O47.91 D=O.OOE+00 Equation is A + B * [Tanh((T-ToY(C+DT))]

Temperature at 50% Shear - 48.0 Plant: Seabrook I Material: SA533B1 Heat: R1808-3 Orientation: LT Capsule: U Fluence: n/cmA2 125 100 75 2

t co a

50 25 0 +--

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

.40. 00 .00 3. 09 -3. 09

.00 25. 00 13.26 I 1. 74

20. 00 2 5. 00 25.09 -. 09

40. 00 50.00 42.31 7. 6 9 4 0. 00 30.00 4 2. 3 1 - 1 2. 3 1

70. 00 85.00 70.39 14.61

70. 00 50.00 70.39 - 20.39

90. 00 80. 00 83. 88 -3. 88 1 00. 00 100.00 88.51 I I. 4 9

-I__

Appendix C

C-21 CAPSULE U LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

Page 2 Plant: Seabrook I Material: SA533BI Heat: RI 808-3 Orientation: LT Capsule: U Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 125.00 1 00. 00 95.35 4.65 125.00 100.00 95.35 4.65 150.00 100.00 98.20 1.80 225.00 100. 00 99.90 .10 325. 00 1 00. 00 100.00 . 00 550. 00 100. 00 100.00 . 00 Correlation Coefficient = .967 Appendix C

C-22 CAPSULE Y LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:22 PM Page I Coefficients of Curve 3 A = 50. B = 50. C - 57.07 TO -55.13 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 55.2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: LT Capsule: Y Fluence: n/cmA2 125 100 91 L-co 75 S - I S

0. 50 II 25 0

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

-40. 00 .0 0 3.44 -3.44

. 00 10. 00 12.65 -2. 65 25.00 30.00 25. 81 4. 19

50. 00 50. 00 45.52 4. 48

68. 00 70. 00 61.09 8.91 71.00 45. 00 63.56 - 18. 56 1 00. 00 85. 00 82. 81 2. 19 125.00 95.00 92.05 2. 95 150.00 100.00 96. 53 3.47

-_ I _n Appendix C

C-23 CAPSULE Y LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

Page 2 Plant: Seabrook I Material: SA533BI Heat: R1 808-3 Orientation: LT Capsule: Y Fluence: nlcmA2 Charpy V-Notch Data Ternperature Input Percent Shear Computed Percent Shear Differential 175.00 100. 00 98.52 1.48 200. 00 100. 00 9 9. 3 8 .62 200.00 100.00 99. 3 8 .62 250. 00 100.00 99. 89 .11 300. 00 100. 00 99. 98 .0 2 350. 00 1 0 0. 00 100.00 . 00 Correlation Coefficicnt -. 986

- -- I Appendix C

C-24 CAPSULE V LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/1212005 02:22 PIM Page 1 Coefficients of Curve 4 A=50. B-50.C=54.84 T0=95.14 D=O.OOE+00 Equation is A + B * [Tanh(CT-To)/(C+DT))]

Temperature at 50%/o Shear = 95.2 Plant: Seabrook I Material: SA533B I Heat: RI 808-3 Orientation: LT Capsule: V Fluence: ntcmA2 125 100 ----- -A-----

A,' 1

' I

/ I

_TI7 75 U) 4 a.

50 A

25 A.

A -.

0

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

- 50.00 5.00 .50 4.50

.00 10. 00 3.02 6. 98 25.00 15. 00 7. 19 7.81

50. 00 20. 00 16. 16 3.84 50.00 20. 00 16. 16 3. 84 75.00 20. 00 32. 42 - 12.42 75.00 30. 00 32.42 - 2. 42 I 0. 00 60. 00 54.42 5. 58 125.00 70. 0 0 74. 82 -4.82 Appendix C

C-25 CAPSULE V LOWER SHELL PLATE R1808-3 (LONGITUDINAL)

Page 2 Plant: Seabrook 1 Material: SA533BI Heat: R1808-3 Orientation: LT Capsule: V Fluencc: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 150. 00 95. 00 8 8. 09 6. 91 175.00 100.00 94. 85 5. 15 225.00 100.00 99. 13 .87 250.00 100. 00 99. 65 . 35 300. 00 1 00. 00 99. 94 . 06 350. 00 100. 00 99. 99 .01 Correlation Coefficient - .991 f

Appendix C

C-26 UNIRRADIATED LOWER SHELL PLATE R1808-3 (TRANSVERSE)

CVGRAPH 5.02 HIyperbolic Tangent Curve Printed on 12/12/2005 02:26 PM Page I Coefficients of Curve I A =40.6 B = 38.4 C =89.14 TO = 34.22 D = O.OOE+00 Equation is A + B * [Tanh((T-Toy(C+DT))]

Upper Shelf Energy=79.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 fl-lbs=9.0 Deg F Temp@50 fl-lbs=56.5 Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: UNIRRA Fluencc: n/cmA2 300 250 5, 200 12 P0 150

100 50 0 , 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 CVN Computed CVN Differendal

- 75. 00 12. 00 8. 30 3.7 0

-50. 00 12. 00 12.28 - .28

- 5 0. 0 0 17.00 12.28 4.72

. 00 27. 00 2 6. 54 .46

. 00 2S. 00 26. 54 1.46

.00 34.00 26.54 7.4 6

40. 00 35.00 43.09 - 8. 09

40. 00 3 6. 00 43. 09 -7. 09

40. 00 40.00 43. 09 -3. 09

- I Appendix C

C-27 UNIRRADIATED LOWER SHELL PLATE R1808-3 (TRANSVERSE)

Page 2 Plant: Seabrook 1 Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: UNIRRA Fluence: n/cmA2 Charpy V-Notch Data Ternperature Input CVN Computed CVN Differential 75.00 46.00 57.04 -I 1. 04 75.00 58.00 57.04 .96 75.00 68.00 57.04 10.96 125. 00 74. 00 7 0. 14 3. 8 6 125.00 79.00 70.14 8.86 200.00 80.00 77. 18 2.82 200.00 81.00 77. 18 3.82 200.00 82.00 77.18 4.82 300. 00 77.00 78. 80 - 1.80 Correlation Coefficient = .974 Appendix C

C-28 CAPSULE U LOWER SHELL PLATE RI 808-3 (TRANSVERSE)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:26 PM Page I Coefficients of Curve 2 A = 37.6 B = 35.4 C = 69.07 TO = 52.77 D = O.OOE+00 Equation is A + B * [Tanh((T-Toy(C+DT))]

Upper Shelf Energy=73.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=37.8 Deg F Ternp@50 fl-lbs=78.1 Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: U Fluence: n/cmnA2 300 250

, 200

150 Lu 8 100 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 Differcntial

-40. 00 12. 00 6. 72 5.28

.00 15.00 14.82 .18

20. 00 19.00 21. 96 -2. 96 40.00 32.50 31.13 1 .37

70. 00 45. 00 46. 25 -1.25 100. 00 59. 50 58. 63 .87 120. 00 58. 00 64. 16 -6.16 125.00 71.50 65. 22 6.28 150. 00 71. 00 69. 00 2.00 I

Appendix C

C-29 CAPSULE U LOWER SHELL PLATE RI 808-3 (TRANSVERSE)

Page 2 Plant: Scabrook I Material: SA533BI Heat: RI 808-3 Orientation: TL Capsule: U Fluence: r/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 1 75.00 73. 50 71. 00 2.50 200.00 69. 50 72.02 - 2.52 200. 00 67.00 72. 02 - 5.02 225. 00 81.50 72. 52 8.98 325. 00 79. 50 72. 97 6. 53 550. 00 70. 00 73. 00 -3. 00 Correlation Coefficient =.983 Appendix C

C-30 CAPSULE Y LOWER SHELL PLATE R1808-3 (TRANSVERSE)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:26 PM Page I Coefficients of Curve 3 A = 34.1 B = 31.9 C = 71.23 TO = 53.09 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy 66.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=43.9 Deg F Ternp@50 ft-lbs=92.1 Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: Y Fluence: n/cme2 300 250 1-,200 P 150 a

uj 8 100 50

~~~~~.~.-_.-.-._ -j.,f-. ~

8'""-'.---'-

0

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

-40. 00 8.50 6.56 I. 94

.00 12.00 13. 93 -1. 93 25.00 2 6. 0 0 2 2. 1 3 3.8 7

50. 00 34.50 32.72 1.78

68. 00 34. 00 40. 68 .6. 68 71.00 44.00 41.96 2.04 100.00 48.50 52.52 -4. 02 125.00 60. 50 58. 52 1.98 150.00 66.50 62.06 4. 4 4

- I Appendix C

C-31 CAPSULE Y LOWER SHELL PLATE R1808-3 (TRANSVERSE)

Page 2 Plant: Seabrook I Material: SA533BI Heat: RI 808-3 Orientation: TL Capsule: Y Fluence: nlcmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 175.00 67.50 63. 98 3.52 200. 00 70. 00 64.98 5.02 200.00 68.50 64. 98 3. 52 250. 00 60.00 65. 75 -5. 75 300.00 64. 00 65. 94 -1.94 350. 00 67. 00 65. 98 1. 02 Correlation Coefficient = .94 Appendix C

C-32 CAPSULE V LOWER SHELL PLATE R1808-3 (TRANSVERSE)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:26 PM Page 1 Coefficients of Curve 4 A = 35.6 B = 33.4 C = 88.74 TO = 85.32 D = O.OOE+00 Equation is A + B * [Tanh((T-ToY(C+DT))]

Upper Shelf Energy=69.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Ternp@30 ft-lbs=70.3 Deg F Temp@50 ft-lbs=126.3 Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: V Flucnce: n/cmA2 300 250

, 200 I 150

100 Ai 50 0

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

- 50. 00 3.00 5.22 -2. 22

. 00 5.00 10.72 - 5. 72

25. 00 17.00 15.85 1.15 50.00 22.00 22.97 - . 97 75.00 35.00 31.73 3.27

75. 00 32.00 31.73 . 27 100.00 49. 00 41. 08 7.92 125.00 39.00 49.61 - 10. 61 125.00 50.00 49.61 .39 I

Appendix C

C-33 CAPSULE V LOWER SHELL PLATE R1808-3 (TRANSVERSE)

Page 2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Oricntation: TL Capsule: V Fluence: n/cmA2 Charpy V-Notch Data Ternperature Input CVN Computed CVN Differential 1 50. 00 57. 00 56. 39 .61 200.00 65.00 64. 3 2 . 68 250. 00 63. 00 67.41 -4. 41 300. 00 68. 00 68. 48 - . 48 350.00 80. 00 68.83 11.17 Correlation Coefficient = .976 Appendix C

C-34 UNIRRADIATED LOWER SHELL PLATE R1808-3 (TRANSVERSE)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:30 PM Page I Coefficients of Curve I A - 31.65 B = 30.65 C = 96.79 TO = 35.77 D = O.OOE+O0 Equation is A 4 B * [Tanh((T-Toy(C+DT))]

Upper Shelf L.E.=62.3 Lower Shelf L.E.=l.O(Fixed)

Ternp.@L.E. 35 nils=46.4 Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: UNIRRA Fluence: n/crn^2 200 150 IA E

.2 a 100 5o 0

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

-75. 00 7. 00 6. 64 . 36

-50. 00 7. 00 9.9 0 -2. 90

- 50. 00 11. 00 9.90 1.10

.00 23. 00 2 0. 8 1 2. 19

. 00 20. 00 2 0. 8 1 - .81

. 00 24.00 20. 8 1 3. 19

40. 00 30. 00 3 2.99 - 2. 99 40.00 30. 00 32. 99 -2. 99

40. 00 34. 00 32.99 1. 01 I

Appendix C

C-35 UNIRRADIATED LOWER SHELL PLATE RI 808-3 (TRANSVERSE)

Page 2 Plant: Seabrook I Material: SA533BI Heat: RI 808-3 Orientation: TL Capsule: UNIRRA Fluence: n/cmA2 Charpy V-Notch Data Temperature Input LE. Computed L.E. Differential 75.00 36.00 43.43 -7. 43 75.00 44.00 43.43 .57 75.00 51.00 43.43 7.57 125.00 55.00 53.92 1.08 125.00 56.00 53. 92 2.08 200.00 62.00 60.31 1.69 200.00 57.00 60.31 -3.31 200.00 59.00 60. 31 -1.31 300.00 63.00 62.04 .96 Corfelation Coefficient = .986 Appendix C

C.36 CAPSULE U LOWER SHELL PLATE R1808-3 (TRANSVERSE)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:30 PM Page I Cocfficients of Curve 2 A = 33.49 B = 32.49 C = 86.66 TO = 61.26 D =O.OOE+O0 Equation is A + B * [Tanh((T-Toy(C+DT))]

Upper Shelf L.E.=66.0 Lower Shelf LE.=1.0(Fixed)

Tcmp.eL.E. 35 nuls=65A Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: U Fluence: m'crnA2 200 150 IE

.2 a 100 i,.

E 50 O

-300

_ 9--'6'/_

0

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

- 40. 00 9. 00 6. 72 2.28

.00 14.00 13. 71 .29

20. 00 16.00 19. 09 -3. 09

40. 00 28.00 25. 67 2.33

70. 00 36. 00 36.75 -. 75 100. 00 45. 00 47.11 - 2. 11 120. 00 51.00 52. 65 -1. 65 125. 00 59. 00 53. 84 5.1 6 150.00 59.00 58.55 .45 Appendix C

C-37 CAPSULE U LOWER SHELL PLATE R1808-3 (TRANSVERSE)

Page 2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: U Fluence: nlcmA2 Charpy V-Notch Data Tempeature Input L.E. Computed L.E. Differential 175.00 62. 00 61.58 . 42 200.00 61. 00 63. 43 -2. 43 200. 00 62. 00 63.43 - 1. 43 225.00 65. 00 64. 52 .48 325.00 71. 00 65. 82 5. 18 550. 00 62.00 65.97 -3.97 Correlation Coefficient - .991 Appendix C

C-38 CAPSULE Y LOWER SHELL PLATE R1808-3 (TRANSVERSE)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:30 PM I Page 1 Coefficients of Curve 3 A = 29.22 B = 28.22 C = 63.86 TO = 62.81 D = O.OOE+00 Equation is A + B * [Tanh((T-ToY(C+DT))]

Upper Shelf L.E.-57.4 Lower Shelf L.E.l .0(Fixed)

Ternp.@L.E. 35 rnils-76.1 Deg F Plant: Seabrook I Material: SA533B I Heat: RI 808-3 Orientation: TL Capsule: Y Fluence: n/cn^A2 200 150 i

.2 a 100 S

50 0

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

-40. 00 1.00 3. 17 -2. 17

.00 6. 00 7. 9 2 - 1. 92

25. 00 1 7. 00 14.22 2. 78 50.00 24. 00 23.63 . 37

68. 00 28.00 31.51 -3. 51

71. 00 3 6. 00 32.82 3.18 100.00 43. 00 44. 02 - 1.02 125.00 48.00 50.39 -2. 39 150.00 58.00 53.98 4. 02

- U Appendix C

C-39 I

CAPSULE Y LOWER SHELL PLATE R1808-3 (TRANSVERSE)

Page 2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: Y Fluence: n/crnA2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 175.00 56.00 55.81 .19 200.00 58.00 56. 68 1.32 200. 00 56.00 56. 68 - . 68 250. 00 53. 00 57. 28 - 4.28 300.00 60. 00 57.40 2. 60 350. 00 57. 00 57. 43 -. 43 Correlation Coefficient - .992 Appendix C

C-40 CAPSULE V LOWER SHELL PLATE R1808-3 (TRANSVERSE)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:30 PM Page 1 Coefficients of Curve 4 A 29.15 B =28.15 C = 85.43 TO = 84.88 D= 0.00E+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf LE.=57.3 Lower Shelf L.E.=I.0(Fixed)

Temp.TL.E. 35 mils=103.0 Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: V Fluence: n/cmA2 200 150 E

0C 8 100 E

4-

/ ..

50  : . d

, ^-'--s- 1'--'----'------

0

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

- 50. 00 .00 3.30 -3.30

.00 4.00 7.79 -3.79

25. 00 12.00 12. 12 - . 12

50. 00 1 9. 00 1 8.25 .75 75.00 2S. 00 25.91 2.09

75. 00 27. 00 25.91 1. 09 1 00. 00 39.00 34. 08 4. 92 125.00 34. 00 41.47 - 7. 47 125. 00 40. 00 41.47 - 1. 47 Appendix C

C-4 CAPSULE V LOWER SHELL PLATE RI 808-3 (TRANSVERSE)

Page 2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: V Fluence: n/cmA2 Charpy V-Notch Data Temperanure Input L.E. Computed L.E. Differential 150. 00 48. 00 47.23 .77 200. 00 5 7. 00 5 3. 73 3. 27 250. 00 53. 00 56. 14 -3. 14 300. 00 51. 00 56.93 - 5. 93 350.00 65. 00 57. 18 7. 82 Correlation Coefficient - .978 Appendix C

C-42 UNIRRADIATED SHELL PLATE R1808-3 (TRANSVERSE)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:33 PM Page 1 Coefficients of Curve I A = 50. B - 50. C = 8038 TO = 38.49 D =O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 500/0 Shear = 38.5 Plant: Scabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: UNIRRA Fluence: n/cmA2 125 100 Y' 75 aE 0

A! 50 25 0 _.

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

- 75. 00 9. 00 5.60 3.40

- 50. 00 1S. 00 9. 96 8.04

- 50. 00 14. 00 9. 96 4.04

. 00 29. 00 27. 73 1.27

.00 29. 00 27.73 1.27

. 00 25.00 27. 73 - 2. 73 40.00 52. 00 50. 94 1.06 40.00 43. 00 50. 94 -7. 94

40. 00 52.00 50. 94 1.06 Appendix C

C43 UNIRRADIATED SHELL PLATE R1808-3 (TRANSVERSE)

Page 2 Plant: Seabrook I Material: SA533B1 Heat: RI 808-3 Orientation: TL Capsule: UNIRRA Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

75. 00 54. 00 7 1. 2 7 -17. 27

75. 00 59. 00 7 1. 2 7 - 12.27

75. 00 92. 00 7 1. 2 7 20. 73 125. 00 100. 00 89. 59 1 0. 41 125. 00 1 00. 00 89.59 10.41 200. 00 100. 00 98.23 1. 77 200. 00 100. 00 98.23 1. 77 200. 00 100. 00 98.23 1. 77 300. 00 100. 00 99. 8 5 .15 Correlation Coefficient =.970 Appendix C

C-44 CAPSULE U LOWER SHELL PLATE R1808-3 (TRANSVERSE)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:33 PM Page I Coefficients of Curve 2 A= 50. B = 50. C = 69.69 T0 = 71.76 D = O.OOE+00 Equation is A + B * [Tanh((T-ToY(C+DT))]

Temperature at 50% Shear = 71.8 Plant: Seabrook I Material: SA533B I feat: R1 808-3 Orientation: TL Capsule: U Fluence: n/cMnA2 125 100 75

>1 I II 0

a. 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

-40.00 . 00 3. 89 -3. 89

.00 20. 00 11.31 8.69 20.00 20.00 18.46 1.54

40. 00 30. 00 2 8. 67 1.33 70.00 45. 00 4S. 74 -3. 74 100.00 65.00 69. 22 -4. 22 120.00 60. 00 79. 97 - 19. 97 125. 00 100. 00 82. 17 17. 83 150. 00 100. 00 90. 43 9.57 Appendix C

C-45 CAPSULE U LOWER SHELL PLATE R1808-3 (TRANSVERSE)

Page 2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: U Fluence: n/cMnA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 1 75. 00 100. 00 95. 09 4. 91 200. 00 100. 00 97. 54 2. 46 200.00 100. 00 97. 54 2. 46 225. 00 100. 00 98.78 1. 22 325. 00 1 00. 00 99. 93 . 07 550.00 1 00. 00 100.00 . 00 Correlation Coefficient - .975 Appendix C

C-46 CAPSULE Y LOWER SHELL PLATE R1808-3 (TRANSVERSE)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:34 PM Page 1 Coefficients of Curve 3 A=50. B=50.C-62.9 TO-70.6 D=O.OOE+00 Equation is A + B * [Tanh((T-ToY(C+DT))]

Temperature at 50h/o Shear - 70.6 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: Y Fluence: n/cmA2 125 100 I 75 a.

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 Diffctential

-40. 00 . 00 2. 8 8 -2. 88

.00 5.00 9. 58 -4. 58 25.00 30.00 1 9. 0 0 1 1. 0 0

50. 00 40. 00 34. 19 5. 8t

68. 00 45. 00 47.94 -2. 94 71.00 40. 00 50.32 - IO. 3 2 100. 00 70. 00 71. 81 - I .8 1 125. 00 90. 00 84. 94 5. 06 150. 00 1 00. 00 9 2. 5 9 7.41

-m ______________________

Appendix C

C-47 CAPSULE Y LOWER SHELL PLATE R1808-3 (TRANSVERSE)

Page 2 Plant: Seabrook I Material: SA533B1 Heat: R1808-3 Orientation: TL Capsule: Y Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 175. 00 95. 00 96.51 - 1. 51 200. 00 100.00 98.39 I .61 200. 00 100.00 98.39 1 .61 250.00 100.00 99. 67 .33 300. 00 100. 00 99. 93 .07 350.00 1 00. 00 99.99 .01 Correlation Coefficient - .990 Appendix C

C-48 CAPSULE V LOWER SHELL PLATE R1808-3 (TRANSVERSE)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:34 PM Page I Coefficients of Curve 4 A = 50. B - 50. C = 76.03 TO = 92.75 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear= 92.8 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsule: V Fluence: n/cMA2 125 100 S.) 75 a-50 25 0 I , -,4

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

- 50. 00 5.00 2. 2 9 2.71

. 00 1 5. 00 8. 02 6. 9 8

25. 00 1 5. 00 14. 40 .60 5 0. 00 20. 00 24. 5 1 -4. 51

75. 00 40.00 38.53 1.47 75.00 40. 00 38. 53 1.47 1 00. 0 0 55. 00 54.75 .25 125.00 70. 00 70. 02 -. 02 125.00 60. 00 70. 02 - 10. 02

- ___________________________________I Appendix C

C-49 CAPSULE V LOWER SHELL PLATE R1808-3 (TRANSVERSE)

Page 2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: TL Capsulc: V Fluence: r/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Conputed Percent Shear Differential 150.00 90. 00 81. 84 8. 16 200. 00 100.00 94. 3 8 5. 62 250. 00 100.00 98.43 1. 57 300. 00 100. 00 99. 5 7 . 43 350. 00 100.00 99. 89 .11 Correlation Coefficient -. 992 Appendix C

C.50 UNIRRADIATED WELD CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:38 PM Page I Coefficients of Curve I A = 81.1 B = 78.9 C = 69.9 TO = -20.77 D O.OOE+00 Equation is A + B * [Tanh((T-ToY(C+DT))]

Upper Shelf Energy160.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Tcmp@30 fl-Ths=-74.6 Deg F Ternp@50 ft-lbs-49.9 Deg F Plant: Seabrook 1 Material: SAW Ileat 4P6052 Orientation: NA Capsule: UNIRRA Fluence: n/cmA2 300 250

], 200 0

U-P 150 w

100 50 0-

.300 -200 -100 0 100 200 300 400 500 600 Temperature in Dog F Charpy V-Notch Data Tempcrature Input CVN Computed CVN Difterential

-I 00. 00 8.00 17. 02 -9. 02

-I 00. 00 26. 00 17. 02 8. 98

-75. 00 12. 00 29.79 - 17. 79

-75. 00 43. 00 29.79 13.21

-75. 00 76. 00 29.79 46.21

- 60. 00 14. 00 40. 95 - 26.95

- 60. 00 15. 00 40. 95 -25. 95

- 50. 00 76. 00 4 9. 9 1 2 6. 09

-50. 00 84. 00 4 9. 9 1 34.09

- I Appendix C

C-51 UNIRRADIATED WELD Page 2 Plant: Seabrook 1 Material: SAW Heat: 4P6052 Orientation: NA Capsule: UNIRRA Fluence: n/cmA2 Charpy V-Notch Data Teiperature Input CVN Computed CVN Differential

-40.00 12. 00 59.93 -47.93

-40. 00 13.00 59.93 -46.93

- 25.00 69. 00 76. 33 -7. 33

-25.00 90.00 76.33 13.67

- 25.00 115.00 76.33 38. 67

. 00 114.00 103. 88 10. 12

. 00 115.00 103. 88 11.12

40. 00 120.00 136.42 - 16. 42

40. 00 128.00 136.42 - 8. 42 120.00 154.00 157.24 -3.24 120.00 154. 00 157.24 -3.24 120. 00 175.00 157.24 17.76 250. 00 157.00 159. 93 -2. 93 Correlation Coefficient - .894

-I Appendix C

C-52 CAPSULE U WELD CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:38 PM Page I Coefficients of Curve 2 A = 73.1 B = 70.9 C - 53.28 TO = -11.68 D O.OQE+OQ Equation is A + B * [Tanh((T-ToY(C+DT))]

Upper Shelf Energy=144.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 fl-lbs--49.2 Deg F Temp@50 ft-lbs-29.6 Deg F Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: U Fluence: nJcKA2 300 iI 250 A 200 0

U-P150 z

50 so 0 - -- -- - - --- _ _ I , ,I ,. I I

-3( )0 -200 -100 0 100 200 300 400 500 600 Temperature In Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

- 80. 00 7.00 12.33 - 5. 3 3

-50. 00 35.00 29.40 5. 60

-40. 00 48.00 3 8. 61 9. 3 9

- 20. 00 30.50 62.12 -31. 62

-20. 00 69. 00 62. 12 6. 88

.00 101.50 88.40 13.10

.00 93. 50 88.40 5. 10

40. 00 123. 00 126. 18 -3. 18

70. 00 133. 00 137.69 -4. 69

_ I Appendix C

C-53 CAPSULE U WELD Page 2 Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: U Fluence: n/cMA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 1 00. 00 138.00 141.89 -3.89 150. 00 143. 00 143. 67 - . 67 225.00 142. 50 143. 98 - 1.48 325. 00 157.00 144.00 13. 00 550. 00 153.00 144. 00 9. 0 0 Correlation Coefficient -. 975 Appendix C

C-54 CAPSULE Y WELD CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12112/2005 02:38 PM Page I Coefficients of Curve 3 A-73.1 B -70.9 C - 64.44 TO = 4.86 D = O.OOE+00 Equation is A + B * [Tanh((T-ToY(C+DT))]

Upper Shelf Energy=-144.0(Fixed) Lower Shelf Energy=2.2(Fixcd)

Temp@30 fl-lbs-50.3 Deg F Temp@50 ft-lbs-26.6 Deg F Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: Y Fluence: n/cmA2 300 250 52 200 P150 8 100 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

- 80. 00 4. 50 14. 75 10.25

-40. 00 27. 50 37.86 -10.36

-25.00 44. 00 51. 6 3 -7. 63

. 00 104.50 78.43 26. 07 25.00 102. 00 103. 78 - 1. 78 50.00 132.50 122. 14 10.3 6 68.00 132. 50 130. 61 1. 89 72.00 1 04. 00 132.05 -28. 05 100. 00 I1 9. 0 0 138.73 - 1 9. 73

_ I Appendix C

C-55 CAPSULE Y WELD Page 2 Plant: Seabrook 1 Material: SAW Heat: 4P6052 Orientation: NA Capsule: Y Fluence: n/cmA2 Charpy V-Notch Data Ten raturc Input CVN Computed CVN Diffcrential 125.00 126. 00 141.52 - 15.52 175.00 149. 00 143.47 5.53 200. 00 140. 00 143. 75 -3.75 200. 00 147. 50 143.75 3.75 250. 00 135.50 143.95 - S.45 300. 00 146.50 143. 99 2.51 Correlation Coefficient - .958 Appendix C

C-56 CAPSULE V WELD CVGRAPH 5.0.2 Hypcrbolic Tangent Curve Printed on 12/12/2005 02:38 PM Page I Coefficients of Curve 4 A - 79.1 B - 76.9 C = 55.82 TO 9.19 D = 0.0OE+OO Equation is A + B * [Tanh((T-Toy(C+DT))J Upper Shelf Energy=156.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 fl-lbs=-32.9 Deg F Temp@50 ft-lbs 13.0 Deg F Plant: Seabrook 1 Material: SAW Heat: 4P6052 Orientation: NA Capsule: V Fluence: n/crm2 300 250 0

A 200 100 A 150 -a---

I /

Lu I A 1/4, 100 50 a .1..

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

-90. 00 6. 00 6.4 8 .. 4 8

- 50. 00 9. 00 18. 67 -9. 67

-35. 00 13. 00 28.40 -15. 40

-25. 00 74.00 37. 12 36.88

- 25. 00 40. 00 37. 12 2. 88

.0 0 70. 00 66.56 3. 44

25. 00 58.00 100. 32 -42. 32

25. 00 108.00 100.32 7. 68

50. 00 124. 00 127.07 -3. 07 Appendix C

C-57 CAPSULE V WELD Page 2 Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: V Fluence: n/cmA2 Charpy V-Notch Data Tctperature Input CVN Computed CVN Differential

75. 00 1 67. 00 142.71 24. 29

75. 00 170. 00 142.71 27.29 100. 00 142.00 150. 28 - 8. 28 200. 00 150.00 155. 84 - 5. 84 250.00 1 50. 00 1 5 5. 97 - 5. 97 300. 00 1 55. 00 156. 00 - 1.00 Correlation Coefficicnt = .950 Appendix C

C-58 UNIRRADIATED WELD CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:42 PM Page I Cocfficients of Curve I A = 45.2 B = 44.2 C - 60.81 TO = -32.83 D = O.OOE+00 Equation is A + B * (Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=89A Lower Shelf L.E.=1.O(Fixed)

Temp.@L.E. 35 rnils=47.1 Deg F Plant: Seabrook I Material: SAW Heat 4P6052 Orientation: NA Capsule: UNIRRA Fluence: n/cmA2 200 150 j

C

.2 2

a 100 S

50 o -

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

- 00. 00 3.00 9. 75 -6. 75

- 100.00 14. 00 9. 75 4. 2 5

-75. 00 6. 00 18. 67 - 12. 67

-75. 00 28. 00 18. 67 9. 3 3

-75. 00 52. 00 1 8. 67 33.33

- 60. 00 7. 00 26. 67 -19. 67

- 60. 00 10. 00 2 6. 67 - 16.67

-50. 00 53. 00 3 3. 05 19.95

-50. 00 58. 00 33.05 24. 95

- U Appendix C

C-59 UNIRRADIATED WELD Page 2 Plant: Seabrook I Matcrial: SAW Heat: 4P6052 Orientation: NA Capsule: LTNIRRA Fluence: n/cm12 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

.40. 00 7. 00 40. 02 - 33. 02

-40.00 11. 00 40. 02 -29. 02

- 25. 00 45. 00 50.86 - 5. 8 6

- 25. 00 60. 00 50. 86 9. 14

-25. 00 74. 00 50. 86 23. 14

.00 75. 00 66. 99 8.01

. 00 72. 00 66. 99 5. 01

40. 00 72. 00 82.02 - 10.02

40. 00 77. 00 82. 02 - 5. 02 120. 00 90. 00 88. 83 1.17 120. 00 91.00 88. 83 2. 17 120.00 90. 00 88.83 1. 17 250.00 90. 00 89.40 .60 Correlation Coefficient = .857 Appendix C

C-60 CAPSULE U WELD CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:42 PM Page 1 Coefficients of Curve 2 A 44.94 B = 43.94 C - 46.23 TO -- 21.71 D = 0.OOE+00 Equation is A + B * [Tanh((T-Toy(C+DT))]

Upper Shelf L.E.=88.9 Lower Shelf LE.1=.O(Fixed)

Ternp.@L-E. 35 mils=-32.3 Deg F Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: U Fluence: nfcmn2 200 150 C

a I

a 100

.- r:

0 50 q __ _X r

. /

. cl,;

0

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

-80. 00 4. 00 7. 5 3 -3. 53

-50. 00 2 6. 00 20. 97 5.03

-40. 00 3 6. 00 28.41 7.5 9

- 20. 00 25.00 46.56 - 21.56

-20. 00 51.00 46.56 4.44

. 00 73. 00 64. 17 8.83

. 00 65. 00 64. 17 .83

40. 00 83. 00 83. 18 -. 18

70. 00 90. 00 87. 24 2.7 6 i I Appendix C

C-61 CAPSULE U WELD Page 2 Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: U Fluence: n/cmA2 Charpy V-Notch Data Terpcrature Input L.E. Computed L.E. Differential 100.00 91. 00 88.42 2.58 1 5 0. 00 89.00 88.82 . IS 225.00 87.00 88.87 -1. 87 325.00 88. 00 88. 87 - . 87 550.00 85. 00 88. 87 -3. 87 Correlation Coefficient -. 970 Appendix C

C-62 CAPSULE Y WELD CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:42 PM Page I Coefficients of Curve 3 A = 42.21 B =41.21 C = 32.23 TO = -17.14 D = O.OOE+00 Equation is A + B * [Tanh((T-Toy(C+DT))]

Upper Shelf L.E.=83.4 Lower Shelf L.E.=l.O(Fixed)

Temp.@L.E. 35 mils-22.8 Deg F Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: Y Fluence: nlcmA2 200 150 E

S C

0 a 100 C

50 0 4---

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

- 80. 00 .00 2. 63 -2.63

-40. 00 17.00 17. 07 -. 07

- 25.00 29. 00 32.36 -3.36

. 00 71.00 62. 28 8.72 25.00 67. 00 77.81 -10. 81

50. 00 77. 00 82. 17 -5. 17

68. 00 83. 00 83.01 - . 01

72. 00 72.00 83. 10 -11. 10 100. 00 8 1. 00 83.37 -2.37 U

Appendix C

C-63 CAPSULE Y WELD Page 2 Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: Y Fluence: n/cMA2 Charpy V-Notch Data Temperature Input L.E. Computed LE. Differential 125.00 84.00 83. 41 .59 17 5. 00 91. 00 83. 43 7. 5 7 200. 00 87. 00 83. 43 3.57 200. 00 89.00 83.43 5. 57 250.00 8 6. 00 83.43 2. 57 300.00 8 7. 00 83.43 3. 5 7 Correlation Coefficient -. 979 Appendix C

C-64 CAPSULE V WELD CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12112/2005 02:42 PM Page 1 Coefficients of Curve 4 A - 44.63 B = 43.63 C = 543 TO = -339 D = 0.OOE+00 Equation is A + B * [Tanh(T-Toy(C+DT))]

Upper Shelf L.E.=88.3 Lower Shelf L.E.=1.0(Fixed)

Tenp.@L.E. 35 rnils-15.5 Deg F Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: V Fluence: n/ctn^2 200 a

150 C

'E 1005 50

... ......-- A2,.

a

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

-90. 00 2. 00 4.45 -2.45

-50. 00 7. 00 14.29 -7. 29

-35.00 9. 00 21. 76 - 12. 76

-25. 00 50.00 28. 13 21. 87

-25. 00 30. 00 28. 13 1.87

.00 49. 00 47.36 1.64

25. 00 42. 00 65.58 -23. 58

25. 00 73.00 65.58 7.42

50. 00 80. 00 77.56 2.44 I

-, I Appendix C

C-65 CAPSULE V WELD Page 2 Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: V Fluence: n/cmA2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

75. 00 93. 00 83.66 9.34

75. 00 8 6. 00 83.66 2.34 1 00. 00 9 2. 00 8 6. 3 8 5.62 200. 00 8 5. 00 88.22 -3. 22 250.00 8 9. 00 88. 26 .74 300.00 80. 00 8 8. 27 -8. 27 Correlation Coefficient = .949 Appendix C

C-66 UNIRRADIATED WELD CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/1212005 02:45 PM Page I Coefficients of Curve I A = 50. B-5.C -82.2 TO = -41.94 D = O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50O% Shear = -41.9 Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: UNIRRA Fluence: n/cenI2 125 100 75 fa 0

0 1! 50 25 O 4--

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

- 100.00 1 8.00 19.58 -1. 58

- 100. 00 25. 00 19. 5 8 5.42

-75. 00 2 3. 00 30. 91 - 7.91

-75. 00 34. 00 30.91 3.09

-75. 00 48. 00 30.91 17.09

- 60. 00 29. 00 39. 19 - 10. 19

- 60. 00 29. 00 39. 19 - 10. 19

- 50. 00 52. 00 45.11 6. 89

-50. 00 54. 00 45.11 8.89

__ N Appendix C

C-67 UNIRRADIATED WELD Page 2 Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: UNIRRA Fluencc: n/cmA2 Charpy V-Notch Data Tenerature Input Percent Shear Computed Percent Shear Differential

-40.00 29. 00 51.18 -22.18

-40.00 37. 00 51.18 - 14. 18

-25.00 55. 00 60. 16 -5. 16

- 25.00 64.00 60. 16 3.84

- 25. 00 84. 00 60. 16 23. 84

. 00 82. 00 73. 50 S. 50

. 00 82. 00 73. 50 8.50

40. 00 75. 00 88. 01 - 13.01

40. 00 82.00 8 8. 01 -6. 01 120.00 1 00. 00 98. 09 1.91 120. 00 100.00 98. 09 1. 91 120. 00 100. 00 98. 09 1 . 91 250. 00 1 00. 00 99. 92 .08 Correlation Coefficient = .929 Appendix C

C-68 CAPSULE U WELD CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:45 PM Page I Coefficients of Curve 2 A =50. B = 50. C = 41.7 TO = -25.65 D - O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = -25.6 Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: U Fluence: m'cMA 2 125 100

_ _ [

I I-0 75 J!

co II V II 9 50 it cto 25 0

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

- 80.00 I 0 . 00 6. 87 3.13

- 50. 00 30. 00 23.72 6.28

-40.00 40.00 33. 44 6. 56

- 20. 00 30.00 56. 74 - 26.74

- 20. 00 55. 00 56. 74 - 1.74

. 00 1 00. 00 77.39 22. 61

.00 8 0. 00 77. 39 2.61

40. 00 90. 00 95. 89 -5. 89

70. 00 100.00 98. 99 1. 01

-111110 N __ _

Appendix C

C-69 CAPSULE U WELD Page 2 Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: U Fluencc: nfcmA2 Charpy V-Notch Data Tenmerature Input Percent Shear Computed Percent Shear Differential 1 00. 00 100. 00 99. 76 .24 150. 00 100. 00 99. 98 . 02 225. 00 100. 00 100. 00 . 00 325. 00 100.00 100.00 . 00 550. 00 100. 00 100. 00 . 00 Correlation Coefficient -. 952 Appendix C

C-70 CAPSULE Y WELD CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:45 PM Page 1 Coefficients ofCurve 3 A - 50. B - 50. C - 50.67 TO - -22.64 D = O.00E+00 Equation is A + B * [Tanh((T-ToY(C+DT))]

Temperature at 50% Shear = -22.6 Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: Y Fluence: n/crm2 125 100 a-75 en I

so

• I 25 I

a.

/

0

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

- 80.00 .00 9.42 -9.42

.40. 00 30. 00 33.51 -3. 51

- 25. 00 50. 00 47.68 2. 32

. 00 85.00 70. 97 14.03 25.00 75. 00 86. 77 - II.77

50. 00 100.00 94.62 5.38

68. 00 90. 00 97.28 -7. 28 72.00 75. 00 97. 67 - 22. 67 100.00 90. 00 99. 22 -9. 22

- I Appendix C

C-71 CAPSULE Y WELD Page 2 Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: Y Fluence: n/cm^2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 1 25. 00 9 0. 00 9 9. 7 1 -9. 71 175. 00 1 00. 00 99.96 .04 200.00 100. 00 99.98 .02 200. 00 1 00. 00 99. 98 .02 250.00 100. 00 100.00 .00 300.00 100.00 100. 00 . 00 Correlation Coefficient -. 958 Appendix C

C-72 CAPSULE V WELD CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/1212005 02:45 PM Page I Coefficients of Curve 4 A=50. B=50.C=51.03 TO=-6.88 D-O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = -6.8 Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: V Fluence: n/cMA2 125 100 I I _ I _

I_ _ I_ I I- k . - iI 75

hA a.

I ;I I I 50 7 11 I I

Ai A i I 25 I I4N jA 0 I I I

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

-90. 00 5. 00 3.71 1.29

-50. 00 15. 00 15.58 - . 58

-35. 00 20. 00 24.94 -4.94

-25.00 40. 00 32.96 7.04

-25. 00 30. 00 32. 96 -2. 96

. 00 60. 00 56.70 3.30 25.00 70. 00 77. 72 - 7. 72

25. 00 80. 00 77.72 2.28 50.00 90.00 90.28 -. 28

- I Appendix C

C-73 CAPSULE V WELD Page 2 Plant: Seabrook I Material: SAW Heat: 4P6052 Orientation: NA Capsule: V Fluence: n/cm\2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

75. 00 100.00 96.12 3.8 8 75.00 98. 00 96. 12 1. 88 100. 00 98. 00 9 8. 5 1 - .51 200. 00 100.00 99. 9 7 .03 250. 00 100. 00 100.00 .0 0 300. 00 100.00 100. 00 . 00 Correlation Cofricient =.995 Appendix C

C-74 UNIRRADIATED HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/1212005 02:50 PM Page I Coefficicnts of Curve I A = 66.1 B= 63.9 C = 139.95 TO = -104.19 D = O.OOE+OO Equation is A + B * [Tanh((T-Toy(C+DT))]

Upper Shelf Energy=130.0(Fixed) Lower Shelf Energy2.2(Fixed)

Temp@30 fl-lbs=-193.7 Deg F Temp@50 fl-lbs=-140.2 Deg F Plant: Seabrook I Material: SA533BI Heat: RI808-3 Orientation: NA Capsule: UNIRRA Flucnce: nt/cm2 300 250

], 200

150 a

z

100 50 0

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

-200.00 15.00 2B. I I -13.11

- 200. 00 19.00 2 8. I 1 -9. 11

-175.00 26.00 36.27 - l0. 27

-175.00 31.00 36.27 - 5. 2 7

- 175.00 56.00 36. 27 19.73

-150.00 47. 00 45. 90 1. 10

- 150. 00 73.00 45.90 27. 10

-1 50.00 80.00 45. 90 34. 10

- 100.00 38.00 68.01 -30.01

__I Appendix C

C-75 UNIRRADIATED HEAT AFFECTED ZONE Page 2 Plant: Seabrook 1 Material: SA533BI Heat: RI 808-3 Orientation: NA Capsule: UNIRRA Fluence: n/cm12 Charpy V-Notch Data Temperature Input CVN Cornputed CVN Differential

-100.00 47.00 68.01 -21.01

-100.00 86.00 68.01 17.99

-60.00 54.00 85.63 -31.63

-60.00 55.00 85.63 -30.63

-60.00 108.00 85.63 22.37

.00 108.00 106.48 1.52

.00 121.00 106.48 14.52 40.00 125.00 115.56 9.44 40.00 142.00 115.56 26.44 110.00 108.00 124.28 -1 6.28 110. 00 132.00 124.28 7.72 110.00 132.00 124.28 7.72 250.00 142.00 129.20 12.80 Correlation Coefficient -. 886 Appendix C

CG76 CAPSULE U HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:50 PM Page I Coeflicients of Curve 2 A -68.1 B = 65.9 C = 106.37 TO = -29.18 D - O.OOE+O0 Equation is A + B * [Tanh((T-Toy(C+DT))]

Upper Shelf Energy=134.0(Fixed) Lower Shelf Energr=2.2(Fixed)

Temp@30 ft-lbs-99.3 Deg F Temp@50 ft-lbs--59.1 Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: U Fluence: n/cmnA2 300 250

-, 200 w;150 z

6 100 50 0 i i  ! , I

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

- 80. 00 32. 00 38. 81 -6. 81

-40. 00 89. 00 61.42 27.58

- 40. 00 79.00 61.42 17.5 8

- 20. 00 62. 50 73. 77 - 11. 27

.00 52. 00 85. 74 -33. 74

20. 00 71. 00 96.57 - 25.57

40. 00 121.50 105.79 15.71

70. 00 130.00 116.32 13.68 1 00. 00 147. 50 123.33 24. 17

-. I Appendix C

C-77 CAPSULE U HEAT AFFECTED ZONE Page 2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: U Fluence: nfcmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 150. 00 142.50 129.61 12. 89 175.00 111.00 131.22 -20.22 200. 00 138.00 132.25 5. 75 225.00 118.50 132.90 - 14.40 325.00 161.50 133.83 27. 67 550. 00 132.00 134.00 -2. 00 Corrclation Coefficient - .860 1

Appendix C

C-78 CAPSULE Y HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:50 PM Page 1 Coefficients of Curve 3 A = 58.1 B = 55.9 C = 53.47 TO - -16.47 D = O.OOE+00 Equation is A + B * (Tanh((r-Toy(C+DT))]

Upper Shelf Energy=l 14.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs--46.0 Deg F Temp@50 ft-lbs--24.2 Deg F Plant: Seabrook I Material: SA533BI hleat: R1808-3 Orientation: NA Capsule: Y Fluence: n/cnA2 300 250 s 200 1

LL P 150 z

z; 100 I ___0__

50 0

.300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Tenierat= Input CVN ConPuted CVN Diffcrential

- 80.00 15.50 11.70 3. 80

.40.00 32.50 34.97 -2.47

- 25.00 46.50 49. 26 - 2.76

.00 71.00 74.79 -3.79 25.00 116.00 94.44 21.56

50. 00 100.00 105.41 -5.41 68.00 95.50 109.45 -13.95

72. 00 102.00 110.06 - S. 06 I 0. 00 102.00 112.58 -10.58

- I Appendix C

C-79 CAPSULE Y HEAT AFFECTED ZONE Page 2 Plant: Seabrook I Material: SA533BI Heat: RI 808-3 Orientation: NA Capsule: Y Fluence: n/cm"2 Charpy V-Notch Data Ternperature Input CVN Computed CVN Differential 125. 00 107.00 113. 44 -6. 44 175.00 Ill. 5 0 113. 9 1 -2. 41 200. 00 148.00 I 13. 97 34.03 200. 00 115. 50 113. 97 1. 53 250.00 108.00 113. 99 - 5. 9 9 300. 00 120.50 114. 00 6.5 0 Correlation Coefficient = .938 Appendix C

C-80 CAPSULE V HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:50 PM Page 1 Coefficients of Curve 4 A = 66.1 B = 63.9 C = 75.78 TO -- 6.05 D -0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy- I30.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Tcmp@30 fl-lbs-54.5 Deg F Temp@50 ft-lbs--25.5 Deg F Plant: Scabrook I Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: V Fluence: n/crnA2 300 250 __ I __ I __

a, 200 0

U-P 150 A A A

e; 100 A

I 50 a -.---.-----.- - 4- - I.- -

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

- 90. 00 36. 00 14. 77 21. 23

- 75. 00 23. 00 20. 02 2. 98

-50. 00 28. 00 32. 70 -4. 70

-25. 00 36. 00 50.44 - 14.44

.00 73. 00 71.19 1.81

25. 00 8 9. 00 90. 9 1 - 1. 91

25. 00 107.00 90. 9 1 .16. 09 50.00 76. 00 106. 29 - 30. 29

75. 00 150. 00 116.53 33.47 I

Appendix C

C-81 CAPSULE V HEAT AFFECTED ZONE Page 2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: V Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 100. 00 123.00 122.66 .34 125.00 137.00 126.10 10. 90 150.00 106.00 127.95 -21. 95 200. 00 121.00 129.45 -8.45 250.00 137.00 129.85 7. 15 300.00 136.00 129.96 6. 04 Correlation Coefficient = .930 Appendix C

C-82 UNIRRADIATED HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:53 PM Page I Coefficients of Curve I A = 40.98 B = 39.98 C - 148.95 TO = -97.82 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=8 1.0 Lower Shelf L.E.1=.0(Fixed)

Ternp.@L.E. 35 inils=-120.2 Deg F Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: UNIRRA Fluence: n/cne2 200 150 E

a 100 e 50 o

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

- 200. 00 10. 00 17.18 -7. 18

- 200. 00 I 0. 00 17.18 -7. 18

-175. 00 17. 00 21. 94 -4. 94

-175. 00 17. 00 21.94 - 4. 94

-175. 00 32. 00 21.94 10. 06

- 150.00 30.00 27. 52 2. 4 8

- 150.00 41. 00 27.52 13.48

- 150. 00 50. 00 27. 52 22. 48

- 100. 00 21.00 40.40 -19.40

- _________________________________U Appendix C

C-83 UNIRRADIATED HEAT AFFECTED ZONE Page 2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: UNIRRA Fluence: n/cmA2 Charpy V-Notch Data Temperature Input LE. Computed LE. Differential

- 100. 00 26.00 40. 40 -14.40

-1 00. 00 54. 00 40.40 13. 60

- 60. 00 33. 00 50. 92 -17.92

- 60. 00 39. 00 50. 92 - 11. 92

- 60. 00 60.00 50. 92 9. 08

. 00 66. 00 64. 02 1.98

. 00 74.00 64. 02 9.98

40. 00 76. 00 70. 10 5. 90

40. 00 83. 00 70. 10 12. 90 I 10. 00 64.00 76. 34 -12. 34 110. 00 78.00 76.34 1.66 110 . 00 78.00 76. 34 1.66 250. 00 76. 00 80.22 -4. 22 Correlation Coefficient =.888 Appendix C

C-84 CAPSULE U HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:54 PM Page I Cocfficicnts of Curve 2 A 41.89 B - 40.89 C = 107.38 TO - -38.99 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=82.8 Lower Shelf LE.= .O(Fixed)

Tomp.@LE. 35 mils-57.2 Deg F Plant: Scabrook I Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: U Fluence: n/cmA2 200 150 a 100 I! D, 0 ___

5.

50

.~~~~ ~ ____--a 0

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

- 80. 00 2 1. 00 26. 99 -5. 99

-40. 00 5 6. 00 41.51 14.49

-40. 00 5 2. 00 41. 51 1 0. 4 9

- 20. 00 42. 00 49. 05 -7. 05

.00 40. 00 56. 12 - 16. 12

20. 00 51. 00 62. 34 -1. 34

40. 00 74. 00 67. 51 6. 4 9

70. 00 82.00 73. 29 8. 7 1 100.00 85.00 77. 07 7. 93 I

Appendix C

C-85 CAPSULE U HEAT AFFECTED ZONE Page 2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: U Fluence: n/cmr2 Charpy V-Notch Data Tenpemture Input LE. Computed LE. Differential 150. 00 85.00 80.44 4.56 175. 00 80. 00 81.30 - 1.30 200.00 79. 00 81. 85 - 2.85 225. 00 78. 00 82. 19 -4. 19 325.00 80.00 82.70 -2. 70 550.00 82.00 82.79 -. 79 Correlation Coefficient = .907 Appendix C

cG86 CAPSULE Y HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:54 PM Page I Cocfficients of Curve 3 A - 36.91 B = 35.91 C = 56.33 TO -- 15.98 D - O.OOE+00 Equation is A + B * [Tanh(CI-Toy(C+DT))]

Upper Shelf L.E.72.8 Lower Shelf L.E.=l.0(Fixed)

Temp.@L.E. 35 mils'18.9 Deg F Plant: Scabrook I Material: SA533B I Hcat RI808-3 Orientation: NA Capsule: Y Fluence: n/cmA2 200 150 E

.2 EL 100 50 0

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

- 80. 00 7. 00 7.71 -. 71

-40. 00 22. 00 22. 46 -. 46

-25. 00 30. 00 31.21 - 1.21

.0 0 46. 00 46. 83 - . 83

25. 00 68. 00 59. 23 8.77

50. 00 66. 00 66.52 -. 52 68.00 59. 00 69.35 - 10.35

72. 00 67.00 69. 79 -2. 79 100.00 68. 00 71.67 -3. 67 m I Appendix C

C-87 CAPSULE Y HEAT AFFECTED ZONE Pagc 2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: Y Fluence: n/ern2 Charpy V-Notch Data Temperature Input LE. Computed LE. Differential 125.00 69.00 72.34 -3. 34 175.00 76.00 72.74 3. 26 200. 00 79.00 72. 78 6.22 200. 00 75.00 72.78 2. 22 250.00 70. 00 72. 81 -2. 81 300.00 78. 00 72. 82 5. 18 Correlation Coefficient - .977 Appendix C

C-88 CAPSULE V HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:54 PM Page 1 Coefficients of Curve 4 A - 39.31 B = 38.31 C = 86.25 T0 - -19.63 D - O.OOE+00 Equation is A + B * [Tanh((T-Toy(C+DT))]

Upper Shelf L.E.=77.6 Lower Shelf L.E.=l.O(Fixed)

Tenp.@L.E. 35 nils-29.3 Deg F Plant: Seabrook I Material: SA533BI Heat: RI 808-3 Orientation: NA Capsule: V Fluence: n/cmA2 200 150 M

E

.2 E 100 S

• A A _____________________.

50 AA

/J:

0

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

.90. 00 28.00 13. 53 14.47

.75. 00 15. 00 17. 62 -2. 62

-50. 00 25. 00 26.35 - 1.35

-25.00 27. 00 36.93 - 9.93

. 00 45.00 47. 88 -2. 88 25.00 60. 00 57.53 2.47 25.00 67. 00 57.53 9.47

50. 00 53.00 64.90 - 11. 90

75. 00 81. 00 69. 94 11. 06

- I Appendix C

C-89 CAPSULE V HEAT AFFECTED ZONE Page 2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: V Fluence: n/cmA2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 1 00. 00 79. 00 73. 12 5. 88 125. 00 79.00 75.03 3. 97 150.00 6 8. 0 0 7 6. 1 5 - 8. 15 200. 00 7 0. 00 77. 1 5 -7. 15 250. 00 79.00 77.47 1. 53 300. 00 78.00 77.57 . 43 Correlation Coefficient - .943 Appendix C

C-90 UNIRRADIATED HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:56 PM Page I Coefficients of Curve I A = 50. B - 50. C = 127.83 TO = -102.2 D = O.OOE+00 Equation is A + B * [Tanh((T-Toy(C+DT))]

Temperature at 50% Shear = -102.1 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: UNIRRA Flucncc: n/cmA2 125 100 g- 75 m.

50 25

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

-200. 00 13.00 17. 80 -4. 80

-200. 00 1 3. 00 17. 80 -4. 80

-175. 00 23. 00 24. 25 -1. 25

-175. 00 1 8. 0 0 24. 25 -6. 25

-1 75. 00 30.00 24. 25 5.75

.150. 00 3 7. 00 32. 13 4. 87

-150. 00 40. 00 32. 13 7. 87

-150. 00 54. 00 32. 13 21.87 100. 00 29. 00 50. 86 -21. 86

- __________________________U Appendix C

C-91 UNIRRADIATED HEAT AFFECTED ZONE Page 2 Plant: Seabrook I Material: SA533BI Heat: RI 808-3 Orientation: NA Capsule: UNIRRA Fluence: n/cmA2 Charpy V-Notch Data Tenperature Input Percent Shear Computed Percent Shear Differential

-100.00 3 8. 00 50. 86 -12. 86

-I 0 0. 0 0 79.00 50. 86 28. 14

- 60. 00 5 5. 00 65.93 -10. 93

- 60. 00 37. 00 65. 93 -28. 93

- 6 0. 00 67.00 65. 93 1 . 07

. 00 92. 00 83. 19 8. 81

. 00 92. 00 83. 19 8. 81

40. 00 1 00. 00 90. 25 9.75

40. 00 100.00 90. 25 9. 75 I I 0. 00 100.00 96. 5 1 3. 4 9 I I 0. 00 100. 00 9 6. 5 1 3. 4 9 110. 0 0 100. 00 9 6. 5 1 3. 4 9 250.00 1 00. 00 99. 60 . 40 Correlation Coefficient = .925 Appendix C

C-92 CAPSULE U HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:57 PM Page I Coefficients of Curve 2 A = 50. B = 50. C = 113.01 TO = -57.88 D = O.OOE+00 Equation is A + B * [Tanh((T-To)(C+DT))]

Temperature at 50%1o Shear = -57.8 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: U Fluence: n/cmA2 125 100 75 V) 0.

50 25 0 O-- I I

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

- 80. 0 0 45.00 40. 34 4.66

- 40. 00 60. 00 57. 84 2. 16

-40. 00 60. 00 57. 84 2. 16

-20. 00 70. 00 66. 16 3.84

.00 50. 00 73. 58 -23. 58

20. 00 70. 00 79. 87 -9. 87

40. 00 1 00. 00 84. 97 1 5. 03

70. 00 1 0 0. 00 90.58 9. 4 2 1 00. 00 100. 00 94. 24 5. 7 6

- I Appendix C

C-93 CAPSULE U HEAT AFFECTED ZONE Page 2 Plant: Seabrook 1 Material: SA533BI Heat: RI808-3 Orientation: NA Capsule: U Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 150.00 1 00. 00 97.54 2.46 175. 00 100.00 9S. 40 1.60 200. 00 100. 00 98.97 1 . 03 225. 00 100. 00 9 9. 33 .67 325.00 100. 00 99. 89 .11 550. 00 100. 00 100.00 . 00 Correlation Coefficient - .917 Appendix C

C-94 CAPSULE Y HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:57 PML4 Page I Coefficients of Curve 3 A = 50. B = 50. C = 71.89 TO = -34.3 D = O.OOE+O0 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50%h Shear = -34.2 Plant: Seabrook 1 Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: Y Fluence: n/cmr2 125 100 I-co 75 a:

50 25 0 4-

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

- 80. 00 15.00 21.90 -6. 90

-40. 00 5 0. 00 46.04 3. 9 6

-25. 00 60. 00 56.43 3. 5 7

. 00 70.00 72.20 -2. 20

25. 00 90.00 83.89 6.11

50. 00 80.00 9 1. 2 6 - 11. 26 68.00 90. 00 94.51 -4. 51

72. 00 95. 00 95.06 -. 06 100. 00 1 00. 00 97. 67 2. 3 3

___M Appendix C

C-95 CAPSULE Y HEAT AFFECTED ZONE Page 2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: Y Fluencc: nfcmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 125.00 100. 00 98. 82 1.18 1 7 5. 00 100.00 99. 70 .30 200. 00 100. 00 99.85 .15 200. 00 100. 00 99. 8 5 .15 250. 00 1 00. 00 99. 96 . 04 300.00 1 00. 00 99.99 .01 Correlation Coefficient - .985 Appendix C

C-96 CAPSULE V HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 12/12/2005 02:57 PM Page 1 Coefficients of Curve 4 A = 50. B = 50. C - 68.59 TO = -15.88 D = O.OOE+00 Equation is A + B * [Tanh((T-Toy(C+DT))]

Temperature at 50% Shear= -15.8 Plant: Seabrook 1 Material: SA533BI Ileat: RI 808-3 Orientation: NA Capsule: V Fluence: n/cmAn2 125 100 75 0

2 50 a-25 o 4-

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

• 90. 00 10.00 10.33 - .33

- 75.00 20. 00 15. 14 4. 86

- 50. 00 25. 00 26. 99 -1.99

- 25.00 45.00 43.39 1.61

.0 0 55.00 61.37 -6.37

25. 00 85. 00 76.71 8.29

25. 00 75. 00 76. 71 -1.71

50. 00 80. 00 87. 23 -7. 23

75. 00 100. 00 93. 40 6.60 Appendix C

C-97 CAPSULE V HEAT AFFECTED ZONE Page 2 Plant: Seabrook I Material: SA533BI Heat: R1808-3 Orientation: NA Capsule: V Fluence: n/cMA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 100. 00 1 00. 00 96.70 3. 3 0 125.00 100. 00 9S. 38 1.62 150. 00 100. 00 9 9. 2 1 . 79 200.00 100.00 99. 82 . I8 250.00 100.00 99. 9 6 . 04 300.00 100. 00 99.99 . 01 Correlation Coefficient - .992 Appendix C

1)-0 APPENDIX D SEABROOK UNIT 1 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION Appendix D

D-1 Introduction Regulatory Guide 1.99, Revision 2 [Ref. D-1] 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 three surveillance capsules removed from the Seabrook Unit I 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 Seabrook Unit I reactor vessel surveillance data and determine if the Seabrook Unit I surveillance data are credible.

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

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements," as follows [Ref. D-2]:

...the reactor vessel (shell materialincluding welds, heal affected zones, and plates orforgings) that directly surroundsthe effective height of the active core and adjacentregions of the reactorvessel that arepredictedto experience sufficient neutron radiationdamage to be consideredin the selection of the most limiting material with regardto radiationdamage.

The Seabrook Unit I reactor vessel consists of the following beltline region materials:

Intermediate shell plates R1806-1,2,3 Lower shell plates R1808-1,2,3 Intermediate shell longitudinal (axial) weld seams 10 1-124 A,B,C Lower shell longitudinal (axial) weld seams 101-142 A,B,C Intermediate to lower shell circumferential (girth) weld seam 101-171 Appendix D

E1-2 From WCAP-I1 100 [Ref. D-3], selection of the surveillance material was based on an evaluation of initial toughness (characterized by the reference temperature, RTNDT and C, upper shelf energies), and the predicted effect of chemical composition (residual copper and phosphorus) and neutron fluence on the toughness (RTNDT shift) during reactor operation. Lower shell plate numbered R1808-3 (Heat DI 136-2) was selected as the surveillance base metal since it had the highest adjusted EOL RTNDT of the six beltline region plates. Weld Heat 4P6052 was selected because it is the same heat used in fabrication of all of the axial and circumferential welds.

Based on this discussion, Criterion I is met for the Seabrook Unit I reactor vessel.

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

Based on engineering judgment, the scatter in the data presented in these plots is small enough to permit the determination of the 30 ft-lb temperature, and the USE of the Seabrook Unit I surveillance materials unambiguously. Hence, the Seabrook Unit I 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 28 0 F 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 [Ref. D-4].

The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized io determine a best-fit line for these data and to determine if the scatter of ARTNDT values about this line is less than 23°F for welds and less than 17'F for the plate.

The Seabrook Unit I lower shell plate R1808-3 and surveillance weld will be evaluated for credibility.

The surveillance weld is made from weld wire Heat 4P6052 and also located in the Millstone 3 vessel surveillance program. Since there are now three data points available that are specific to the Seabrook surveillance program, only that data will be evaluated to determine credibility.

Table D-1 contains the calculation of chemistry factors for the Seabrook Unit 1 reactor vessel beltline materials contained in the surveillance program. These chemistry factors are calculated per Regulatory Guide 1.99, Revision 2, Position 2.1.

Appendix D

D-3 Table D-1 Calculation of Chemistry Factors using Seabrook Unit I Surveillance Capsule Data Material Capsule Capsule jPa) FF(b) ARTNDT(C) FF*ARTNDT FF2 Intermediate Shell Plate U 0.3142 0.682 39.7 27.08 0.465 R1808-3 Y 1.292 1.071 47.2 50.55 1.147 (Longitudinal) V 2.669 1.262 60.4 76.22 1.593 Intermediate Shell Plate U 0.3142 0.682 28.8 19.64 0.465 R1808-3 Y 1.292 1.071 34.9 37.38 1.147 (Transverse) V 2.669 1.262 61.3 77.36 1.593 SUM: 288.23 6.410 CF = X(FF

• RTNDT) + X( FF2) = (288.23) + (6.410) = 45.0 0F U 0.3142 0.682 25.4 17.32 0.465 Weld Metal 4P6052 Y 1.292 1.071 24.3 26.03 1.147 V 2.669 1.262 41.7 52.63 1.593 SUM: 95.98 3.205 CF = X(FF

• RTNDT) . Z( FF2) = (95.98) . (3.205) = 30.00 F Notes:

(a) f= Calculated fluence from the Seabrook Unit I capsule V dosimetry analysis results, (x 1019 n/cm 2, E > 1.0 MeV).

(b) FF = fluence factor==0.2 8 . I log 0 (c) ARTNDT values are the measured 30 ft-lb. shift values for Seabrook Unit I taken from Appendix C.

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

Appendix D

['-4 Table D-2 Seabrook Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line for Surveillance Forging Materials Material Capsule CF FF Measured Predicted Scatter (Base Metals)

(Slopebest fi) ARTNDT ARTNDT ARTNDT <28( F (Wteld)

U 45.00 F 0.682 39.7 0 F 30.7 0F 9.0F Yes Lower Shell Plate R11i08-3 Y 45.0F 1.071 47.2 0F 48.2 0F - 1.00 F Yes (Longitudlinal)

V 45.0F 1.262 60.4 0 F 56.8 0F 3.6 0F Yes U 45.0F 0.682 28.8 0F 30.70 F -1.90 F Yes Lower Shell Plate RI1808-3 Y 45.0F 1.071 34.9 0F 48.2 0 F - 13.3OF Yes (Transverse)

V 45.0 0 F 1.262 61.3 0F 56.80 F 4.50F Yes U 30.00 F 0.682 25.40F(a) 20.50 F 4.90F Yes Surveillance Weld Material Y 30.00 F 1.071 24.30F(a) 32.1 0 F -7.8 0F Yes (Heat # 4T'6052) 1 V 30.00 F 1.262 41.70F(a) 37.9 0F 3.8 0F Yes Note:

(a) The measured surveillance *veld metal ARTNDT values have not been adjusted by a ratio factor since variability of thc non coated copper wire without nickel addition is a function of measurement variability and not clear evidence that the surveillance weld differs from the 0.047% copper and 0.049% nickel best estimate.

Table D-3 indicates that no data points fall outside the +/- lc of 17°F scatter band for the lower shell plate R1808-3 surveillance data. Therefore, the plate surveillance data are deemed credible. No weld data points fall outside the +/- 1c of 28°F scatter band for the surveillance weld data; therefore, the weld data are also deemed credible per the third criterion. For future evaluation of ART and RTPTS values, reduced margin terms for both the plate and weld materials is permitted per Regulatory Guide 1.99, Revision 2.

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 +/- 25°F.

The capsule specimens are located in the reactor between the fuel and the vessel wall opposite the center of the core. The test capsules are in baskets attached to the vessel wall. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens are subjected to 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.

Appendix D

D-5 The Seabrook Unit 1 vessel surveillance program does not have correlation monitor material as a part of that program. Therefore, Seabrook takes exception to meeting the requirements of this criterion.

Conclusion Based on the preceding responses to all five criteria of Regulatory Guide 1.99, Revision 2, the Seabrook Unit I surveillance plate and weld data are deemed credible.

References D-1. Regulatory Guide 1.99, Revision 2, RadiationEmbrittlement ofReactor Vessel Materials, U.S. Nuclear Regulatory Commission, May, 1988.

D-2. Code of Federal Regulations, IOCFR50, Appendix Q FractureToughness Requirements, U.S.

Nuclear Regulatory Commission, Washington, D.C.

D-3. WCAP-I0 100, PublicService Company of New HampshireSeabrook Station Unit No. I Reactor Vessel Radiation SurveillanceProgram,L.R. Singer, ct. al., dated March 1983 D4. ASTM E1 85-82, StandardPracticeforCondutctingSirveillance Tests forLight- Water Cooled NuclearPower Reactor Vessels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.

Appendix D