ML040330964

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Surveillance Capsule Test Report
ML040330964
Person / Time
Site: San Onofre Southern California Edison icon.png
Issue date: 01/27/2004
From: Scherer A
Southern California Edison Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
BAW-2454
Download: ML040330964 (133)


Text

F1I SOUTHERN CALIFORNIA AEInSI EDISON An EDISON INTERNATIONiA0 Company A. Edward Scherer Manager of Nuclear Regulatory Affairs January 27, 2004 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

Subject:

Docket No. 50-362 Surveillance Capsule Test Report San Onofre Nuclear Generating Station, Unit 3

Reference:

Letter from F. R. Nandy (SCE) to the Document Control Desk (NRC) dated May 3,1991,

Subject:

Docket No. 50-362, Surveillance Capsule Test Report, San Onofre Nuclear Generating Station Unit 3

Dear Sir or Madam:

This letter provides as an enclosure Framatome ANP, Inc. report "Analysis of the 263 Degree Capsule, Southern California Edison Company, San Onofre Unit 3 Nuclear Generating Station - Reactor Vessel Material Surveillance Program," Framatome ANP Document No. BAW-2454 dated January 2004. This report provides the test results and analysis for the second surveillance capsule which was removed from the San Onofre Unit 3 reactor vessel on January 27, 2003, as required by 10 CFR 50, Appendix H. No immediate revision of the reactor coolant system pressure-temperature (P-T) limits in existing Technical Specification (TS) 3.4.3 is necessary. While the predicted 20 Effective Full Power Year (EFPY) Adjusted Reference Temperature (ART) has increased by approximately 21 degrees, the current Unit 3 TS P-T curves are anticipated to be valid for at least the next 4 years with consideration of current (code case N-640) methodology which was recently incorporated into the ASME code. SCE will update the Unit 3 P-T curves as necessary within 2 years, to ensure continued compliance.

P.O. Box 128 7 4c San Clemente, CA 92674-0128 949-368-7501 Fax 949-368-7575

Document Control Desk January 27, 2004 The San Onofre Unit 3 reactor vessel surveillance capsule test report for the first surveillance capsule was submitted by the reference.

If you have any questions or need additional information regarding this matter, please contact me or Mr. Jack Rainsberry at (949) 368-7420.

Sincerely, Enclosure cc:

B. S. Mallett, Regional Administrator, NRC Region IV B. M. Pham, NRC Project Manager, San Onofre Units 2, and 3 C. C. Osterholtz, NRC Senior Resident Inspector, San Onofre Units 2 & 3

iny-I BAW-2454 January 2004 Analysis of the 2630 Capsule Southern California Edison Company San Onofre Unit 3 Nuclear Generating Station

-- Reactor Vessel Material Surveillance Program --

t FRAMATOME ANP

BAW-2454 January 2004 Analysis of the 2630 Capsule Southern California Edison Company San Onofre Unit 3 Nuclear Generating Station

-- Reactor Vessel Material Surveillance Program --

by J. B. Hall J. W. Newman, Jr.

Framatome ANP Document No. 77-2454-00 (See Section 9 for document signatures.)

Prepared for Southern California Edison Company Prepared by Framatome ANP, Inc.

3315 Old Forest Road P. 0. Box 10935 Lynchburg, Virginia 24506-0935 A

FRAMATOME ANP

BAW-2454 Executive Summary This report describes the results of the examination of the second capsule (the 2630 capsule) of the Southern California Edison Company San Onofre Unit 3 Nuclear Generating Station as part of their reactor vessel surveillance program (RVSP). The objective of the program is to monitor the effects of neutron-irradiation on the mechanical properties of the reactor vessel materials by testing and evaluation of tensile and Charpy V-notch impact specimens. The San Onofre Unit 3 RVSP was designed and furnished by Combustion Engineering, Inc., based on ASTM Standard E 185-73.

The 2630 capsule was removed from the San Onofre Unit 3 reactor vessel at the end of cycle eleven for testing and evaluation. The capsule received a maximum fast fluence of 2.471 x 10'9 n/cm2 (E > 1.0 MeV). Based on cycle 11 flux (including power uprated conditions) extrapolated to 32 EFPY, the projected peak fast fluence of the San Onofre Unit 3 reactor vessel beltline region clad/vessel interface is 4.028 x 1O'9 n/cm2 (E.> 1.0 MeV). Fluence values were calculated using Framatome ANP's NRC approved fluence methodology, which is fully Regulatory Guide 1.190 compliant.

The results of the tension tests indicated that the San Onofre Unit 3 surveillance materials exhibited normal behavior for neutron-irradiation. The Charpy impact data for the San Onofre Unit 3 surveillance materials exhibited a higher increase in ductile-to-brittle transition temperature than predicted by Regulatory Guide 1.99, Revision 2. Therefore the surveillance data was used to establish the end of license adjusted reference temperature prediction for this beltline limiting material. The surveillance materials exhibited a characteristic decrease in upper-shelf energy as a result of neutron-irradiation. None of the San Onofre Unit 3 beltline materials are predicted to fall below the 50 ft-lb limit required in 10 CFR 50 Appendix G before 32 EFPY.

A ii FRAMATOME ANP

BAW-2454 Table of Contents Page

1. Introduction.....

1-1~~~~~.

1.

Introduction......................................................................................................................

1-1

2.

Background.......................

2-1

3.

Surveillance Program Description

.3-1

4.

Tests of Unirradiated Material

.4-1

5.

Post-Irradiation Testing

.5-1 5.1.

Capsule Disassembly and Inventory

.5-1 5.2.

Thermal Monitors

.5-1 5.3.

Tension Test Results 5-1 5.4.

Charpy V-Notch Impact Test Results

.5-2 5.5.

Hardness Testing

.5-3 5.6.

Chemical Analysis 5-3

6.

Neutron Fluence............................................................................................................... 6-1 6.1.

Introduction..........................................................................................................

6-1 6.2.

Fluence Results 6-2 6.3.

Dosimetry Activity.

6-3 6.4.

Capsule Lead Factors

.6-3

7.

Discussion of Capsule Results 7-1 7.1.

Chemical Composition Data

.7-1 7.2.,

Unirradiated Material Property Data 7-1 7.3.

Irradiated Property Data

.7-1 7.3.1. Tensile Properties 7-2 7.3.2. Impact Properties 7-2 7.3.3. Hardness................................................................................................... 7-3 7.4.

Surveillance Capsule Withdrawal Schedule.

7-3 7.5 Predicted Adjusted Reference Temftperatures

................ 7-3 7.6 Predicted Decrease in Charpy Upper Shelf Energy

........................................ 7-5

8.

Summary of Results

.8-1

9.

Certification 9-1

10.

References 10-1 A

iii FRAMATOME ANP

BAW-2454 Table of Contents (cont.)

Appendices Page A.

Unirradiated and Irradiated Tensile Data for the San Onofre Unit 3 RVSP Materials...........................................................

A-I B.

Unirradiated and Irradiated Charpy V-Notch Impact Surveillance Data for the San Onofre Unit 3 RVSP Materials Using Hyperbolic Tangent Curve-Fitting Method...............

B-1 C.

Charpy V-Notch Shift Comparison: Original Hand Curve Fit vs. Hyperbolic Tangent Curve Fit.C-l D.

Fluence Analysis Methodology.D-1 E.

Reactor Vessel Surveillance Program Background Data and Information.E-l List of Tables Table Page 3-1.

Test Specimens Contained in the 2630 San Onofre Unit 3 Capsule.3-3 3-2.

Chemical Composition of the 2630 San Onofre Unit 3 Capsule Surveillance Materials.3-4 5-1.

Tensile Properties of the 2630 San Onofre Unit 3 Capsule Reactor Vessel Surveillance Materials, Irradiated to 2.471 x IO'9 n/cm2 (E>1.0 MeV).5-4 5-2.

Charpy V-Notch Impact Results for the 2630 San Onofre Unit 3 Capsule Base Metal Plate C-6802-1, Irradiated to 2.471 x 1019 n/cm2 (E> 1.0 MeV) Transverse Orientation.5-5 5-3.

Charpy V-Notch Impact Results for the 263° San Onofre Unit 3 Capsule Weld Metal, C-6802-2/C-6802-3 (Wire Heat 90069), Irradiated to 2.471 x 1019 n/cm2 (E>1.0 MeV).5-6 5-4.

Charpy V-Notch Impact Results for the 2630 San Onofre Unit 3 Capsule Heat-Affected-Zone Material, Irradiated to 2.471 x 1019 n/cm2 (E>1.0 MeV).

5-7 5-5.

Charpy V-Notch Impact Results for the 263° San Onofre Unit 3 Capsule Standard Reference Material, HSST Plate 01 Irradiated to 2.471 x I0'9 n/cm (E>1.0 MeV).

5-8 5-6.

Hyperbolic Tangent Curve Fit Coefficients for the 2630 San Onofre Unit 3 Capsule Surveillance Materials.5-9 A

iv FRAMATOME ANP

BAW-2454 List of Tables (cont.)

Table Page 5-7.

Rockwell Hardness B Measurements for the 2630 San Onofre Unit 3 Capsule Base Metal Plate C-6802-1, Irradiated to 2.471 x 10 n/cm2 (E>1.0 MeV)........................... 5-10 5-8.

Rockwell Hardness B Measurements forthe 2630 San Onofre Unit 3 Capsule Weld Metal, C-6802-2/C-6802-3 (Wire Heat 90069), Irradiated to 2.471 x 10'9 n/cm2 (E> 1.0MeV).......................

5-10 5-9.

Rockwell Hardness B Values for the 263° San Onofre Unit 3 Capsule Heat-Affected-Zone Material, Irradiated to 2.471 x 10 19 n/cm2 (E > 1.0 MeV)

(Converted from Knoop)......................

5-10 5-10. Rockwell Hardness B Measurements for the 2630 San Onofre Unit 3 Capsule Standard Reference Material, HSST Plate 01, Irradiated to 2.471 x 10'9 n/cm2 (E > 1.0 MeV).............:

........ 5-11 5-11.

Chemical Composition of the 2630 San Onofre Unit 3 Capsule Base Metal Plate C-6802-1 and Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069).5-11 6-1.

San Onofre Unit 3 Fast Fluence Dosimetry.6-5 6-2.

3D Coordinates for San Onofre Unit 3 Points of Interest.........................................

6-5 6-3.

3D Synthesized Fluxes.6-6 6-4.

Cumulative Fluence Estimates at the Wetted Surface.6-7 6-5.

Cumulative Fluence Estimates at the Vessel/Clad Interface.......................................... ;;.6-7 6-6.

C/M Ratios..........................................

6-7 7-1.

Chemical Composition Data for San Onofre Unit 3 Reactor Vessel Surveillance Base Metal Plate C-6802-1.7-7 7-2.

Chemical Composition Data for San Onofre Unit 3 Reactor Vessel Surveillance Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069)....................................................... 7-7 7-3.

Summary of San Onofre Unit 3 Reactor Vessel Surveillance Capsules Tensile Test Results..........................................................

7-8 7-4.

Measured vs. Predicted 30 ft-lb Transition Temperature Changes for 2630 San Onofre Unit 3 Capsule Surveillance Materials - 2.471 x 10'9 n/cm2.................................... 7-9 7-5.

Measured vs. Predicted Upper-Shelf Energy Decreases for the 2630 San Onofre Unit 3 Capsule Surveillance Materials-2.471 x 10 n/cm2............................................. 7-10 V

FRAMATOME ANP

BAW-2454 List of Tables (cont.)

Table Page 7-6.

Summary of San Onofre Unit 3 Reactor Vessel Surveillance Capsules Charpy Impact Test Results (Based on Tanh Reevaluation)...................................

7-11 7-7.

Hardness Data for San Onofre Unit 3 Reactor Vessel Surveillance Materials................ 7-12 7-8.

Surveillance Capsule Withdrawal Schedule...........................................................

7-12 7-9.

Credibility Assessment for San Onofre Unit 3 Reactor Vessel Beltline Limiting Plate C-6802-1..........................................................

7-13 7-10. Predicted Adjusted Reference Temperature for the Uprated Peak Fluence in San Onofre Unit 3 Reactor Vessel Beltline Limiting Plate C-6802-1................................ 7-14 7-11. Predicted Charpy Upper Shelf Energy at 32 EFPY for the San Onofre Unit 3 Reactor Vessel Beltline Region Materials at the 1/4T Location with Power Uprate..........................................................

7-15 A-1.

Tensile Properties of Unirradiated Base Metal Plate C-6802-1, Transverse Orientation..........................................................

A-2 A-2.

Tensile Properties of Unirradiated Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069)...........................................

A-2 A-3.

Tensile Properties of Unirradiated Heat Affected Zone Metal....................................... A-3 A-4.

Tensile Properties of Base Metal Plate C-6802-1, Irradiated to 8 x 1018 n/cm2 (E> 1.0 MeV) Transverse Orientation.A-3 A-5.

Tensile Properties of Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069)

Irradiated to 8 x 1018 n/cm2 (E>1.0 MeV).A-4 A-6.

Tensile Properties of Heat Affected Zone Metal Irradiated to 8 x 1018 n/cm 2 (E>1.0 MeV).A-4 B-1.

Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Base Metal Plate C-6802-1, Transverse Orientation.B-2 B-2.

San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Base Metal Plate C-6802-1, Irradiated to 8 x 1018 n/cm2 (E>1.0 MeV) Transverse Orientation.

B-3 B-3.

Hyperbolic Tangent Curve Fit Coefficients for San Onofre Unit 3, Base Metal Plate C-6802-1, Transverse Orientation.

B-4 B-4.

Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Base Metal Plate C-6802-1, Longitudinal Orientation.B-5 B-5.

San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Base Metal Plate C-6802-1, Irradiated to 8 x 1018 n/cm2 (E>1.0 MeV) Longitudinal Orientation.

B-6 A

vi FRAMATOME ANP

BAW-2454 List of Tables (cont.)

Table Pace B-6.

Hyperbolic Tangent Curve Fit Coefficients'for San Onofre Unit 3, Base Metal Plate C-6802-1, Longitudinal Orientation....................................

B-7 B-7.

Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069)..

B-8 B-8.

San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069) Irradiated to 8 x 1018 n/cm2 (E>.0 MeV).

B-9 B-9.

Hyperbolic Tangent Curve Fit Coefficients for San Onofre Unit 3 Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069)...........................

B-10 B-10. Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Heat-Affected-Zone Material..................

B-I 1 B-l.

San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Heat-Affected-Zone Material, Irradiated to 8 x 1618 n/cm2 (E>1.OMeV)..

B-12 B-12. Hyperbolic Tangent Curve Fit Coefficients for San Onofre Unit 3 Heat-Affected-Zone Material.

B-13 B-13. Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Standard Reference Material...............

'B-14

~

~

~~~~~ ~ ~~

~

~

~~.

B14 B-14. Hyperbolic Tangent.Curve Fit Coefficients for San Onofre Unit 3 Standard Reference Material.B-.............

' B-15 C-1.

Comparison of Curve Fit Transition Temperature Shifts for San Onofre Unit 3 Surveillance Material, Base Metal Plate C-6802-1, Transverse Orientation.C-2 C-2.

Comparison of Curve Fit Transition Temperature Shifts for San Onofre Unit 3 Surveillance Material, Base Metal Plate C-6802-1, Longitudinal Orientation.C-3 C-3.

Comparison of Curve Fit Transition Temperature Shifts for San Onofre Unit 3 Surveillance Material, Weld Metal, C-6802-2/C-6802-3 (Wire Heat 90069).C-4 C-4.

Comparison of Curve Fit Transition Temperature Shifts for San Onofre Unit 3 Surveillance Material, Heat-Affected-Zone Material....................................

.;.;.C-5 D-1.

Bias Correction Factors.......................

D-9 D-2.

Calculational Fluence Uncertainties D9...........................

D-9 E-1.

Description of the San Onofre Unit 3 Reactor Vessel Beltline Region Material.. ;.E-3 vi.

A vii FRAMATO ME AN P

BAW-2454 List of Figures Fisure Page 3-1.

Reactor Vessel Cross Section and Elevation Views Showing the Locations of RVSP Capsules in San Onofre Unit 3 Reactor Vessel................................. 3-5 3-2.

Surveillance Capsule Assembly Showing Locations of Specimens and Monitors for the 2630 San Onofre Unit 3 Capsule.............................................................

3-6 5-1.

Photographs of Thermal Monitors Removed from the San Onofre Unit 3 2630 Capsule Top Compartment.............................................................

5-12 5-2.

Photographs of Thermal Monitors Removed from the San Onofre Unit 3 263° Capsule Middle Compartment.............................................................

5-13 5-3.

Photographs of Thermal Monitors Removed from the San Onofre Unit 3 2630 Capsule Bottom Compartment.............................................................

5-14 5-4.

Tension Test Stress-Strain Curve for Base Metal Plate C-6802-1, Transverse Orientation, Specimen No. 2J7, Tested at 70IF...............................

5-15 5-5.

Tension Test Stress-Strain Curve for Base Metal Plate C-6802-1, Transverse Orientation, Specimen No. 2L6, Tested at 2500F................................

5-16 5-6.

Tension Test Stress-Strain Curve for Base Metal Plate C-6802-1, Transverse Orientation, Specimen No. 2KA, Tested at 550'F...................................... 5-17 5-7.

Tension Test Stress-Strain Curve for Weld Metal, C-6802-2/C-6802-3 (Wire Heat 90069), Specimen No. 3JL, Tested at 70'F............................

5-18 5-8.

Tension Test Stress-Strain Curve for Weld Metal, C-6802-2/C-6802-3 (Wire Heat 90069), Specimen No. 3KE, Tested at 250'F..............................

5-19 5-9.

Tension Test Stress-Strain Curve for Weld Metal, C-6802-2/C-6802-3 (Wire Heat 90069), Specimen No. 3KC, Tested at 550'F...........................................................

5-20 5-10.

Tension Test Stress-Strain Curve for Heat-Affected-Zone Material, Specimen No. 4KE, Tested at 70'F...........................................................

5-21 5-11.

Tension Test Stress-Strain Curve Heat-Affected-Zone Material, Specimen No. 4KA, Tested at 250'F...........................................................

5-22 5-12.

Tension Test Stress-Strain Curve for Heat-Affected-Zone Material, Specimen No. 4J4, Tested at 550'F...........................................................

5-23 5-13.

Photographs of Tested Tension Test Specimens and Corresponding Fracture Surfaces - Base Metal Plate C-6802-1, Transverse Orientation.................................... 5-24 5-14.

Photographs of Tested Tension Test Specimens and Corresponding Fracture Surfaces - Weld Metal, C-6802-2/C-6802-3 (Wire Heat 90069).................................. 5-25 A

viii FRAMATOME ANP

BAW-2454 List of Figures (cont.)

Figure Pave 5-15.

Photographs of Tested Tension Test.Specimens and Corresponding Fracture Surfaces - Heat-Affected-Zone Material............................

5-26 5-16.

Charpy Impact Data for Irradiated Base Metal Plate C-6802-1, Transverse Orientation..............

5-27 5-17.

Charpy!Impact Data for Irradiated Weld Metal, C-6802-2/C-6802-3 (Wire Heat 90069)...........

5-28 5-18.

Charpy Impact Data for Irradiated Heat-Affected-Zone Material.

5-29 5-19.

Charpy Impact Data for Irradiated Standard Reference Material, HSST Plate 01.

5-30 5-20. Photographs of Charpy Impact Specimen Fracture Surfaces, Base Metal Plate C-6802-1, Transverse Orientation..........................

5-31 5-21.

Photographs of Charpy Impact Specimen Fracture Surfaces, Weld Metal, C-6802-2/C-6802-3 (Wire Heat 90069)

.......................... 5-32 5-22. Photographs of Charpy Impact SpecimenFracture Surfaces, Heat-Affected-Zone Material.

5-33 5-23.

Photographs of Charpy Impact Specimen Fracture Surfaces, Standard Reference Material, HSST Plate 01........................................

5-34 6-1.

San Onofre Unit 3 Core Overview........................................

6-8 6-2.

San Onofre Unit 3 2630 Dosimetry Locations........................................

6-8 B-1.

Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Base Metal Plate C-6802-1, Transverse Orientation Refitted Using Hyperbolic Tangent Curve-Fitting Method.........................................

B-1 6 B-2.

San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Base Metal Plate C-6802-1, Irradiated to 8 x 1018 n/cm2 (E>1.0 MeV) Transverse Orientation Refitted Using Hyperbolic Tangent Curve-Fitting Method........................................

B-17 B-3.

Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Base Metal Plate C-6802-1, Longitudinal Orientation Refitted Using Hyperbolic Tangent Curve-Fitting Method.........................................

B-18 B-4.

San Onofre Unit 3 97° Capsule Surveillance Charpy Impact Data for Base Metal Plate C-6802-1, Irradiated to 8 x 1018 n/cm2 (E>1.0 MeV) Longitudinal Orientation Refitted Using Hyperbolic Tangent Curve-Fitting Method.

B-I9 A

ix FRAMATOME ANP

BAW-2454 List of Fiaures (cont.)

Figure Page B-5.

Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069) Refitted Using Hyperbolic Tangent Curve-Fitting Method...........................................................

B-20 B-6.

San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069) Irradiated to 8 x 101S n/cm2 (E>1.0 MeV)

Refitted Using Hyperbolic Tangent Curve-Fitting Method........................................... B-21 B-7.

Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Heat-Affected-Zone Material Refitted Using Hyperbolic Tangent Curve-Fitting Method............................................................

B-22 B-8.

San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Heat-Affected-Zone Material, Irradiated to 8 x 1018 n/cm2 (E>1.0 MeV) Refitted Using Hyperbolic Tangent Curve-Fitting Method............................................................. B-23 B-9.

Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Standard Reference Material Refitted Using Hyperbolic Tangent Curve-Fitting Method............................................................

B-24 D-1.

Fluence Analysis Methodology for San Onofre Unit 3 Surveillance Capsule.............. D-10 E-1.

Location and Identification of Materials used in the Fabrication of the San Onofre Unit 3 Reactor Pressure Vessel.................................................. E-4 A

X FRAMATOME ANP

BAW-2454

1. Introduction This report presents the test and evaluation results of the second reactor vessel surveillance capsule (2630 Capsule) removed from the San Onofre Unit 3 reactor vessel during the I th refueling outage which began on January 6, 2003. The contents were evaluated after being irradiated in the San Onofre Unit 3 reactor as part of the reactor vessel surveillance program (RVSP) as documented in S-NLM-002, Revision 2(11. This report describes the testing and the post-irradiation results obtained from the 2630 capsule removed from San Onofre Unit 3 after receiving a maximum fast fluence of 2.471 x 1019 n/cm2 (E>1.0 MeV). Fluence values were calculated using Framatome ANP's NRC approved fluence methodology, which is fully Regulatory Guide 1.190 compliant.

These data are compared to the previous San Onofre Unit 3 RVSP results from the 970 capsule.12 This report meets the reporting requirements of I OCFR50, Appendix H and American Society for Testing and Materials (ASTM) Standard El85-82.131 The objective of the program is to monitor the effects of neutron irradiation on the mechanical properties of reactor vessel materials under actual plant operating conditions. The program was planned to monitor the effects of neutron-irradiation on the reactor vessel materials for the 40-year design life of the reactor pressure vessel. The San Onofre Unit 3 RVSP was designed and furnished by Combustion Engineering, Inc., based on ASTM Standard E 185-73.

A 1-1

~~FRAMATOME ANP

BAW-2454

2. Background

The ability of the reactor vessel to resist fracture is a primary factor in ensuring the safety of the primary coolant system in light water reactors. The reactor vessel beltline region is the most critical region of the vessel since it is exposed to the highest level of neutron-irradiation. The general effects of fast neutron-irradiation on the mechanical properties of low-alloy ferritic steels used in the fabrication of reactor vessels are well characterized and documented. These low-alloy ferritic steels exhibit an increase in ultimate and yield strength properties with a corresponding decrease in ductility after irradiation. The most significant mechanical property change in reactor vessel steels is the increase in the ductile-to-brittle transition temperature accompanied by a reduction in Charpy upper-shelf energy (CYUSE).

Code of Federal Regulation, Title 10, Part 50, (10 CFR 50) Appendix G, "Fracture Touighness Requirements," [4] specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of commercial light water reactors and provides specific guidelines for determining the pressure-temperature limitations for operation'of the RCPB. The fracture toughness and operational requirements are specified to provide adequate sAfety margins during normal operation, anticipated transients and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime.

The requirements of 10 CFR 50, Appendix G, became effective on August 16, 1973. These requirements are applicable to all boiling and pressurized water nuclear power reactors, including those under construction or in operation on the effective date.

10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program.Requirements, "['I defines the material surveillance program required to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of light water reactors resulting from exposure to neutron-irradiation and the thermal environment. Fracture toughness test data are obtained from material specimens contained in capsules that are periodically withdrawn from the reactor vessel. These data permit determination of the conditions under which the vessel can be operated with adequate safety margins against non-ductile fracture throughout its service life.

A method for guarding against non-ductile fracture in reactor vessels is described in Appendix G to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, A

2-1 FRAMATOME ANP

BAW-2454 Section III, "Nuclear Power Plant Cornponents" [61and Section XI, "Rulesfor Inservice Inspection

"[7J.

This method uses fracture mechanics concepts and the reference nil-ductility temperature, RTNDT, which is defined as the greater of the drop weight nil-ductility transition temperature (in accordance with ASTM E 208-8118]) or the temperature that is 60'F below that at which the material exhibits 50 ft-lbs and 35 mils lateral expansion. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KIR curve), which appears in Appendix G of ASME B&PV Code Section III and Section XI. The KIR curve is a lower bound of dynamic and crack arrest fracture toughness data obtained from several heats of pressure vessel steel. When a given material is indexed to the KIR curve, allowable stress intensity factors can be obtained for the material as a function of temperature. The operating limits can then be determined using these allowable stress intensity factors.

The RTNDT and the operating limits (pressure/temperature limits) of a nuclear power plant, are adjusted over the life of the plant to account for the effects of irradiation on the fracture toughness of the reactor vessel materials. Irradiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel are monitored by each plant's surveillance program. The surveillance capsules, as part of the surveillance program, contain prepared specimens of the reactor vessel materials, which are irradiated in the reactor vessel. A surveillance capsule is periodically removed from the operating nuclear reactor and the specimens are tested to determine the changes in mechanical properties. The increase in the Charpy V-notch 30 ft-lb temperature is added to the original RTNDT to adjust it for irradiation embrittlement. The adjusted RTNDT is used to index the material to the KIR curve which is used to set operating limits for the nuclear power plant. These new limits take into account the effects of irradiation on the reactor vessel materials.

10 CFR 50, Appendix G, also requires a minimum initial CVUSE of 75 ft-lbs for all beltline region materials unless it is demonstrated that lower values of upper-shelf fracture energy will provide an adequate margin of safety against fracture equivalent to those required by ASME Section XI, Appendix G. No action is required for a material that does not meet the initial 75 ft-lbs requirement if the irradiation embrittlement does not cause the CVUSE to drop below 50 ft-lbs. The regulations specify that if the CVUSE drops below 50 ft-lbs, it must be demonstrated, in a manner approved by the Office of Nuclear Reactor Regulation, that the lower values will provide adequate margins of safety.

A 2-2 FRAMATOME ANP

BAW-2454

3. Surveillance Program Description The reactor vessel surveillance program for San Onofre Unit 3 includes six capsules designed to monitor the effects of neutron and thermal environment on the materials of the reactor pressure vessel core region. The capsules, which were inserted into the reactor vessel before initial plant startup, were positioned inside the reactor vessel between the core support barrel and the vessel wall at the locations shown in Figure 3-1. S-NLM-00211 includes a full description of the capsule locations and design. The 2630 capsule was irradiated in the 263° position in the reactor vessel from the start of cycle I to the end of cycle 1.

The 263° capsule was removed during the cycle 11 refueling shutdown of San Onofre Unit 3. The capsule contained Charpy V-notch (CVN) impact test specimens fabricated from one heat of base metal plate (SA-533, Grade B, Class 1), heat-affected-zone (HAZ) material, a weld metal representative of the San Onofre Unit 3 reactor vessel beltline region intermediate and lower shell longitudinal welds, and a Standard Reference Material (SRM). The SRM is a standard heat of SA-533, Grade B, Class I made available by the NRC sponsored Heavy Section Steel Technology (HSST) program. The tensile test specimens were fabricated from the same base metal plate, HAZ, and weld metal. The number of specimens of each material contained in the 263° capsule is described in Table 3-1, and the location of the specimens within the capsule is shown in Figure 3-

2. The chemical compositions of the surveillance materials within the 2630 capsule, obtained from the RVSP baseline material report,191 are described in Table 3-2.

All base metal CVN and tensile specimens were machined from the 1/44-thickness (/4T) location of the plate material. The base metal specimens were oriented such that the longitudinal axis of each specimen was perpendicular to the principal working direction of the plate (transverse orientation).

The HAZ and weld metal specimens were oriented such that the longitudinal axis of each specimen was perpendicular to the weld seam. The CVN HAZ and weld metal specimens had the notch oriented parallel to the weld seam.

The 2630 capsule contained three dosimeter sets installed in the three tensile compartments of the capsule. Each dosimeter set consisted of cadmium shielded Al-Co, Ni, and Cu wires, as well as shielded U powder. Unshielded Al-Co, Ti, and Fe wires were also included, along with unshielded U powder.

A 3-1 FRAMATOME ANP

BAW-2454 Thermal monitors fabricated from four low-melting alloys were included in the capsule. The thermal monitors were sealed in quartz tubes and inserted in spacers located in Figure 3-2. The eutectic alloys and their melting points are listed below:t1' 80% Ag, 20% Sn 5.0% Ag, 5.0% Sn, 90.0% Pb 2.5% Ag, 97.5% Pb 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point 5360F Melting Point 5580F Melting Point 580'F Melting Point 590F A

FRAMATOME ANP 3-2

BAW-2454 Table 3-1. Test Specimens Contained in the 2630 San Onofre Unit 3 Capsule Number of Test Specimens Material Description T 1 Charpy Tension V-notch l

Base Metal Plate C-6802-1 Transverse 3

12 HAZ Metal 3

12 Weld Metal C-6802-2/C-6802-3 3

12 (Wire Heat 90069) l SRM*1 Total 9

48

  • Standard Reference Material: SA-533, Grade B, Class 1, HSST Plate OlMY I

A 3-3 FRAMATOME ANP

BAW-2454 Table 3-2. Chemical Composition of the 2630 San Onofre Unit 3 Capsule Surveillance Materials Chemical Composition, wt%

Base Metal Plate Weld Metal Element C-6802-1 C-6802-2/C-6802-3

______________________(W ire H eat 90069)

Si S

P Mn C

Cr Ni Mo V

Cb B

Co Cu Al W

Ti As Sn Zr N2 Sb Pb 0.23 0.014 0.008 1.38 0.24 0.07 0.57 0.58 0.005

<0.01 0.0005 0.010 0.05 0.033

<0.01

<0.01 0.009 0.005 0.002 0.010 0.39 0.009 0.004 1.54 0.12 0.05 0.08 0.55 0.005

<0.01 0.001 0.015 0.03 0.006 0.01

<0.01

< 0.001 0.002

< 0.001 0.006 0.0012

<0.001 A

FRAMATOME ANP 3-4

BAW-2454 Figure 3-1. Reactor Vessel Cross Section and Elevation Views Showing the Locations of RVSP Capsules in San Onofre Unit 3 Reactor Vessel r

7 O utlet Nozzle

//

180e I

104' Capsule

_.- Vessel

. Capsule' Assembly Core Support Barrel Plan View Elevational View A

FRAMATOME ANP 3-5

BAW-2454 Figure 3-2. Surveillance Capsule Assembly Showing Locations of Specimens and Monitors for the 2630 San Onofre Unit 3 Capsule III Lock Assembly Q

}

Wedge Couplin

  • Tensile -Monior.._.

Cornmpa rime nt Tensile -Monitor Compartment Tensile -Monitor Compartment 9 Assembly K

1K K

4%,

r Charpy Impact Compartments

  • Charpy Impact Compartments A

FRAMATOME ANP I

3-6

BAW-2454

4. Tests of Unirradiated Material Unirradiated material was evaluated for two purposes: (1) to establish baseline data to which irradiated properties data could be compared; and (2) to determine those material properties as required for compliance with 10 CFR 50, Appendices G and H.

Combustion Engineering, Inc., as part of the development of the San Onofre Unit 3 RVSP, performed the testing of the unirradiated surveillance material. The details of the testing procedures are described by Combustion Engineering, Inc.1'91 The unirradiated mechanical properties for the San Onofre Unit 3 RVSP materials are summarized in Appendices A and B of this report.

The original unirradiated Charpy V-notch impact data and the irradiated Charpy impact data for the 970 Capsule were evaluated based on hand-fit Charpy curves generated using engineering judgment.

These data were re-evaluated herein using a hyperbolic tangent curve-fitting program, and the results of the re-evaluation are presented in Appendix B. In addition, Appendix C contains a comparison of the Charpy V-notch shift results for each surveillance material: the fit presented in the 97° Capsule analysis report21 versus current hyperbolic tangent curve-fit.

A 4-1 FRAMATOME ANP

BAW-2454

5. Post-Irradiation Testing The post-irradiation testing of the tensile specimens, the CVN impact specimens, thermal monitors, and dosimeters for the 2630 San Onofre Unit 3 capsule was performed at the BWXT Services, Inc. in Lynchburg, Virginia.

5.1. Capsule Disassembly and Inventory After capsule disassembly, the contents of the 2630 capsule were inventoried and found to be free of corrosion and consistent with expectations. The capsule contained a total of 48 standard Charpy V-notch specimens, nine (9) tensile specimens, six (6) dosimetry blocks, and three (3) sets of temperature monitors.

5.2. Thermal Monitors The low-melting point (5360F, 5580F, 580'F and 590'F) eutectic alloys contained in the 263° capsule were encapsulated in quartz tubes. Each set of thermal monitors (capsule top, middle, and bottom) were photographed to reveal the shape of the monitors and examined for evidence of melting. In all cases the 5360F temperature monitor completely melted, and the 580'F and 590'F monitors did not show any evidence of melting. The 5580F temperature monitor melted in the top and middle compartments,' but not in the bottom compartment. Based on this examination, the maximum temperature that the test specimens near the bottom of the capsule were exposed to was between 5360F and 5581F and in the middle and top of the capsule the maximum temperature that the test specimens saw was between 5580F and 580 0F. Figures 5-1 through 5-3 contain photographs of the temperature monitors.

5.3. Tension Test Results The capsule contained a total of nine (9) specimens, three (3) specimens made of base metal-transverse orientation, HAZ, and weld metal. The specimens were of standard round type with a gage length of 1.0 inch and a nominal gage diameter of 0.250 inch. The tensile tests for each material were performed at: 1) room temperature, 2) 250'F, and 3) 550F. The results of the post-irradiation tension tests are presented in Table 5-1, and the stress-strain curves are presented in Figures 5-4 through 5-12. The tests were performed using a MTS Model 312 servohydraulic test machine using stroke control with an initial actuator travel rate of 0.0075 inch per minute.

A 5-1 FRAMATOME ANP

BAW-2454 Following specimen yielding, an actuator speed of 0.03 inch per minute was used. The tension testing was performed in accordance with the applicable requirements of ASTM Standard E 21-92.110] Photographs of the tension test specimen fractured surfaces are shown in Figures 5-13 through 5-15.

5.4. Charpy V-Notch Impact Test Results The Charpy V-notch impact testing was performed in accordance with the applicable requirements of ASTM Standard E 23-91.['1] Impact energy, lateral expansion, and percent shear fracture were measured at numerous test temperatures and recorded for each specimen. The impact energy was measured using a certified Satec S 1-1K Impact tester (traceable to NIST Standard) with a striker velocity of 16.96 ft/sec and 240 ft-lb of available energy. The lateral expansion was measured using a certified dial indicator. The specimen percent shear was estimated by video examination and comparison with the visual standards presented in ASTM Standard E 23-91.

The results of the Charpy V-notch impact testing are listed in Tables 5-2 through 5-5. The curves in Figures 5-16 through 5-19 were generated using a hyperbolic tangent curve-fitting program to produce the best-fit curve through the data. The symmetric hyperbolic tangent (TANH) function (test response, i.e., absorbed energy, lateral expansion, and percent shear fracture, "R," as a function of test temperature, "T") used to evaluate the surveillance data is as follows:[12 1 R= A+B*tanh[(T To)]

The Charpy V-notch data was entered with the coefficients A, B, TO, and C determined by the program by minimizing the sum of the errors squared (least-squares fit) in "R" of the data points about the fitted curve. Using these coefficients and the above TANH function, a smooth curve is generated through the data for interpretation of the material transition region behavior. The coefficients determined for irradiated materials in the 2630 capsule are shown in Table 5-6.

When performing the TANH fit for the absorbed energy, the lower shelf was fixed at 2.2 ft-lbs and the upper shelf was fixed at the upper shelf value defined by ASTM Standard El85-82 3]. In fitting the TANH function to the lateral expansion data, the lower shelf was fixed at I mil and the upper shelf was not fixed. For the percent shear fracture data, the lower shelf was fixed at 0% and the upper shelf at 100% shear in fitting the TANH function.

Photographs of the Charpy V-notch specimen fracture surfaces are presented in Figures 5-20 through 5-23.

A 5-2 FRAMATOME ANP

BAW-2454 All Charpy V-notch impact testing was performed using instrumentation to record a load-versus-time trace and energy-versus-time trace for each impact event. The instrumented data was recorded and is available upon request.

5.5. Hardness Testing Rockwell hardness measurements were performed on three specimens from each of three material categories (base metal, reference material, and weld). Rockwell hardness measurements could not be performed on the HAZ specimens due to the small size of the heat affected zone, so Knoop 500g microhardness measurements were alternatively conducted and the results converted to Rockwell B1131. Five measurements were performed on each specimen. The Rockwell tests were conducted in accordance with ASTM Standard E I8-971 4]. The Knoop microhardness measurements were conducted in accordance with ASTM Standard E 384-891")l.

The weld zones were verified using a macroetch, and the HAZ zones were verified by etching the polished metallographic samples. The results are tabulated in Tables 5-7 through 5-10.

5.6. Chemical Analysis Inductively Coupled Plasma (ICP) Atomic Emission Spectroscopy was performed on samples of weld (311B) and base metal (25Y) to determine the chemical composition. The samples were analyzed for manganese, molybdenum, phosphorous, sulfur, nickel, chromium, cobalt, vanadium, silicon, and copper. The results are tabulated in Table 5-1l.

5-3 FRAMATOME ANP

BAW-2454 Table 5-1. Tensile Properties of the 263° San Onofre Unit 3 Capsule Reactor Vessel Surveillance Materials, Irradiated to 2.471 x 10"9 n/cm2 (E>1.0 MeV)(a)

Tcst Strength Fracturc Properties Elongation Reduction Specimen Temp.

Yield Ultimate Load Stress Strength Uniform Total in Area Material No.

(°F)

(ksi)

Iki (b

(ki (si) 2J7 70 71.2 92.2 3179 177.5 64.8 11.2 24.0 63.5 Base Metal Plate C-6802-1 2L6 250 66.6 85.8 3038 179.0 61.9 10.2 21.9 65.4 Transverse 2KA 550 64.9 90.1 3491 150.3 71.1 9.5 18.7 52.7 3JL 70 66.0 94.7 3362 184.0 68.5 12.7 25.0 62.8 Weld Metal C-6802-2/C-6802-3 3KE 250 54.0 72.2 2473 151.8 50.4 9.2 20.3 66.8 (Wire Heat 90069) 3KC 550 53.4 75.8 2820 141.1 57.5 8.3 17.3 59.3 4KE 70 68.3 84.7 2884 150.9 58.8 9.7 NA(b) 61.1 HAZ Mctal 4KA 250 63.4 80.1 2775 155.0 56.5 9.0 NA(b) 63.5 C-6802-1 4J4 550 63.2 87.3 2823 157.6 57.5 9.8 21.8 63.5 (a). The fluence calculation is described in Section 6.0 of this report.

(b). Sarnple 4KE and 4KA failed outside the gage length; therefore the total elongation is not available.

A FRAMATOME ANP 54

BAW-2454 Table 5-2. Charpy V-Notch Impact Results for the 2630 San Onofre Unit 3 Capsule Base Metal Plate C-6802-1, Irradiated to 2.471 X:10 9 n/cm2 (E>1.O MeV)

Transverse Orientation Test Impact Lateral Shear Specimen Temperature,

Energy, Expansion,
Fracture, ID OF ft-lbs mils 214 0

6.5 0

0 21T 70 10.5 5

10 213 100

17.5 13 20 21E 125 14.5 15 15 24K 130 36 30 50 231 140 21.5 18 30, 254 150 63.5 51 45 21B 175 33.5 29 50 264 200 88.5*

71 100 21C 250 53*

46 85 262 280 97.5*

73 100 235 320 55*

55 100

  • Value used to determine upper-shelf energy (USE) in accordance with ASTM Standard E 185-82.[3]

A FRAMATOME ANP 5-5

BAW-2454 Table 5-3. Charpy V-Notch Impact Results for the 2630 San Onofre Unit 3 Capsule Weld Metal C-6802-2/C-6802-3, (Wire Hcat 90069),

Irradiated to 2.471 x 1019 n/cm2 (E>1.0 MeV)

Test Impact Lateral Shear Specimen Temperature,

Energy, Expansion,
Fracture, ID OF ft-lbs mils%

345

-50 10.0 2

0 31U 0

6.5 0

0 31L 25 11 7

20 33P 35 42 33 50 36D 45 42.5 36 65 321 60 13.5 10 5

35B 70 53.5 49 65 333 90 60.5 52 75 31P 100 27.5 26 50 36B 125 63.5*

50 100 36P 150 73*

65 100 32U 200 63.5*

60 95

  • Value used to dete rmine upper-shelf energy with ASTM Standard E 185-82.[3I (USE) in accordance A

FRAMATOME ANP 5-6

BAW-2454 Table 5-4. Charpy V-Notch Impact Results for the 2630 San Onofre Unit 3 Capsule Heat-Affected-Zone Material, Irradiated to 2.471 x 10? n/cm2 (E>1.0 MeV)

Test Impact Lateral Shear Specimen Temperature,.

Energy, Expansion,
Fracture, ID OF ft-lbs mils%

472

-50 2.5 0

0 47K 0

11.5 6

5 46U 25 9.5 9

5 433 45 31.5 24 45 416 70 13.5 8

30 45E 70 26.5 18 25 476 100 22 19 45 415 110 21.5 19 45 451 125 61.5*

47 100 432 150 68.5*

61 100

.44A 200 63.5*

55 100 452 250 83*

68 100

  • Value used to determine upper-shelf energy with ASTM Standard E 185-82.[3 (USE) in accordance A

FRAMATOME ANP 5-7

BAW-2454 Table 5-5. Charpy V-Notch Impact Results for the 2630 San Onofre Unit 3 Capsule Standard Reference Material, HSST Plate 01 Irradiated to 2.471 x 10 9 n/cm2 (E>1.0 MeV)

Test Impact Lateral Shear Specimen Temperature,

Energy, Expansion,
Fracture, ID OF ft-lbs mils B2U 70 6.5 3

0 B36 100 14.5 7

0 B24 125 24.5 15 10 B26 150 28.5 23 15 B27 160 34.5 24 35 B2B 175 36.5 28 35 B2E 200 43.5 36 55 B2A 225 54.5 47 70 B2M 250 69.5 57 90 B25 280 87.5*

65 100 B2L 320 92.5*

73 100 B2D 350 89*

67 100

  • Value used to dete rmine upper-shelf energy with ASTM Standard E 185-82.

(USE) in accordance A

FRAMATOME ANP 5-8

BAW-2454 Table 5-6. Hyperbolic Tangent Curve Fit Coefficients for the 2630 San Onofre Unit 3 Capsule Surveillance Materials Material Hyperbolic Tangent Curve Fit Coefficients Description JAbsorbed Energy Lateral Expansion PercentShearFracture Base Metal Plate A:

37.9 A:

30.8 A:

50.0 C-6802-1 B:

35.7 B:

29.8 B:

50.0 Transverse C:

53.4 C:

53.3 C:

65.1 l_______-______.__

To:

144.3 TO:

143.1 O.

TO:

156.0 Weld Metal A:

34.4 A:

31.5 A:

50.0 C-6802-2/C-6802-3 B:

32.2 B:

30.5 B:

50.0 (Wire Heat 90069)

C:

72.5 C:

71.3 C:

70.0 TO:

54.8 TO:

64.0 TO:

64.4 A:

35.7 A:

34.7 A:

50.0 HAZ Metal B:

33.5 B:

33.7 B:

50.0 C-6802-1 C:

68.6 C:

71.7 C:

58.2 TO:

101.6 TO:

116.8 TO:

92.0 Standard Reference A:

45.9 A:

37.2 A:

50.0 Material, HSST Plate 01 B:

43.7 B:

36.2 B:

50.0 C:

92.4 C:

90.3 C:

59.8 TO:' 192.0 TO:

196.7 TO:

192.3 A

FRAMATOME ANP 5-9

BAW-2454 Table 5-7. Rockwell Hardness B Measurements for the 2630 San Onofre Unit 3 Capsule Base Metal Plate C-6802-1, Irradiated to 2.471 x 1019 n/cm2 (E>1.0 MeV)

Specimen 21T I

Specimen 254 1

Specimen 264 91.3 92.2 93.3 91.8 92.2 93.8 92.3 93.5 93.4 91.5 93.1 94.2 92.8 92.7 92.5 Average 91.9 1

Average 92.7 l

Average 93.4 Table 5-8. Rockwell Hardness B Measurements for the 2630 San Onofre Unit 3 Capsule Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069),

Irradiated to 2.471 x 1019 n/cm2 (E>1.0 MeV)

Specimen 31U Specimen 35B I

Specimen 36D 88.5 92.2 93.5 88.3 94.9 92.5 89.8 94.9 96.8 86.7 93.5 94.4 87.7 93.5 95.7 Average 88.2 l

Average 93.8 l

Average 94.6 Table 5-9. Rockwell Hardness B Values for the 2630 San Onofre Unit 3 Capsule Heat-Affected-Zone Material, Irradiated to 2.471 x 1019 n/cm2 (E>1.0 MeV)

(Converted from Knoop)

[

Specimen 45E Specimen 432 1

Specimen 472 96.4 99 98.5 93.8 100 100 93.0 101 101.5 91.5 98.5 101.5 92.6 103 101

[

Average 93.5 l

Average 100.3 l

Average 100.5 A

FRAMATOME ANP 5-10

BAW-2454 Table 5-10. Rockwell Hardness B Measurements for the 2630 San Onofre Unit 3 Capsule Standard Reference Material, HSST Plate 01 Irradiated to 2.471 x 10'P n/cm2 (E>1.0 McV)

Specimen B2U l

Specimen B2E Specimen B26 96.8 98.9 97.2 97.8 97.3 96.2 97.9 98.1 97.8 98.4 98.1 97.8 97.6 98.4 98.9 Average 97.7 Average 98.2 Average 97.6 Table 5-11. Chemical Composition of the 2630 San Onofre Unit 3 Capsule Base Metal Plate C-6802-1 and Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069)

Element l

Base Metal l

Weld Material Manganese 1.25 l

1.27 Molybdenum 0.53 0.53 Phosphorous

<0.02

< 0.02 Sulfur 0.07 0.06 Nickel 0.54 0.11 Chromium 0.10 0.07 Cobalt 0.01 0.02 Vanadium

< 0.004

< 0.005 Silicon 0.27 0.39 Copper 0.06 0.03 A

FRAMATOME ANP 5-11

BAW-2454 Figure 5-1. Photographs of Thermal Monitors Removed from the San Onofre Unit 3 2630 Capsule Top Compartment A

FRAMATOME ANP 5-12

BAW-2454 Figure 5-2. Photographs of Thermal Monitors Removed from the San Onofre Unit 3 263° Capsule Middle Compartment i'j:x x,

A FRAMATOME ANP 5-13

BANV-2454 Figure 5-3. Photographs of Thermal Monitors Removed from the San Onofre Unit 3 2630 Capsule Bottom Compartment A

FRAMATOME ANP 5-14

2 Oct., 2003 File: 2J7 5

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Strength Yield:

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92232.

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Strength Tield:

UT5:

53954.

72210.

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2 Oct.. 2003 rQ File: 3KC c

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550 F( 287 C)

Ut Strength

'ield:

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Stren th Yield:

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BAW-2454 Figure 5-13. Photographs of Tested Tension Test Specimens and Corresponding Fracture Surfaces - Base Metal Plate C-6802-1, Transverse Orientation 2J7 70°F 2L6 2500F 2KA 550F A

FRAMATOME ANP 5-24

BAW-2454 Figure 5-14. Photographs of Tested Tension Test Specimens and Correspionding Fracture Surfaces - Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069) 3JL 70F 3KE 2500F 3KC 550F A

FRAMATOME ANP 5-25

BAW-2454 Figure 5-15. Photographs of Tested Tension Test Specimens and Corresponding Fracture Surfaces - Heat-Affected-Zone Material 1;tV~~~~~~~~~~~~~~~~~:

4KE 700F 4KA 20'F 414 550°F A

FRAMATOME ANP 5-26

BAW-2454 Figure 5-16. Charpy Impact Data for Irradiated Base Metal Plate C-6802-1, Transverse Orientation 100 I ;L 0

coQ 75 50 25

-100

-50 0

50 100 150 200 250 300 350 Temperature, F 400 E

0 co en x

( U In

.0 0)

'U U

(U E.

100 80 60 40 20 0

-1 120 110 100 90 80 70 60 50 40 30 20 10 0

-1 IT39WLE.

150.7 F I

a S

0 -

a

=

)I C

D 0

100 200 Iemperature, F 300 400 T56.

163.3 F T3o:

132.4 F CvUSE:

73.5 ft-lbs

/@...'

10

-50 0

50 100 150 200 Temperature, F 250 300 350 400 A

FRAMATOME ANP 5-27

BAW-2454 Figure 5-17. Charpy Impact Data for Irradiated Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069) 100 L.

U 75 50 25 o i-d

-100

-50 u9 100 80 E

0 us 60-CL X

40 LU 20

-J 0

e

-100 0

50 100 150 200 250 Temperature, F 300 350 400 0

100 200 300 Temperature, F 400 100 90 80 T

h, 93.0 F T30:

44.8 F CvUSE:

66.7 f-lbs 0

4!Ii LU E

0.2 70 60 50 40

  • o 30 ------------

20 0

10 0

.100

-50 0

50 100 150 200 250 300 350 400 Temperature, F A

FRAMATOME ANP 5-28

t BAW-2454 Figure 5-18. Charpy Impact Data for Irradiated Heat-Affected-Zone Material U

6-(0

-100

-50 0

50 100 150 200 250 300 350 Temperature, F 400 x

-J 100 80 60 40 20 0

-1 C lT35MLE 117.5 F I

0*

5 ~0 S.....................

An. In

..~~~~~~~~~~~~~~~~~~~~~~

)O 0

100 200 Temperature, F 300

- 400 cm

.0 0) w U

0.

120 110 100 90 80 70 60 50 40 30 20 10 0

Tso:

133.0 F T30 89.9 F CvUSE:

69.1 ft-lbs S

I e...

so

)0

-50 0

50 100 150 200 250 I.

Temperature, F 300 350 400 A

FRAMATOME ANP 5-29

BANV-2454 Figure 5-19. Charpy Impact Data for Irradiated Standard Reference Material, HSST Plate 01 100 o;

75 2

50 c,

25 c

0

-100

-50 0

50 100 150 200 250 Temperature, F 300 350 400

' 100 6

80

.2 60

":L 40 Z.

20 0

IT35MLE 191.1 F

_..........~~~~

-100 0

100 200 Temperature, F 300 400 150 140 130 120 110 to 100

.0

="

80 0

LU 70 X

60 0.

E 50 40 30 20 10 0

T 200.6 F T3d-156.7 F CvUSE:

89.7 ft-lbs IS O

I...................................................................I............................

-100

-50 0

50 100 150 200 250 300 350 400 Temperature, F A

5-30 FRAMATOME ANP

BAW-2454 Figure 5-20. Photographs of Charpy Impact Specimen Fracture Surfaces, Base Metal Plate C-6802-1, Transverse Orientation a

214 0F 254150° 21T 707 21B 175F 213 100F 264 200 0F 21E 125TF 21C2506 F 24K 130WF 231 140F 262280TF 235 320F A

FRAMATOME ANP 5-31

BAW-2454 Figure 5-21. Photographs of Charpy Impact Specimen Fracture Surfaces, Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069) 345 -50OF 35B 700F 31U F

333 900F 31 L 25F 31P 1000F 33P 350F 36B 125TF 36D 450F 36P 150TF 321 60OF 32U 200F A

FRAMATOME ANP 5-32

BAW-2454 Figure 5-22. Photographs of Charpy Impact Specimen Fracture Surfaces, Heat-Affected-Zone Material 472 -50 0F 476 100' 47K F

415 110 F 46U 250F 451 125F 432 1500F 433 45W 416 700F 44A 200WF 45E 700 F 452 250T A

FRAMATOME ANP 5L33

BAW-2454 Figure 5-23. Photographs of Charpy Impact Specimen Fracture Surfaces, Standard Reference Material, HSST Plate 01 B2E 2000F B36 100F B2A 2250F B24 1250F B2M 2500F B26 150TF B25 280TF B27 160TF B2L 320 0F B2B 175 0F B2D 350 0F A

FRAMATOME ANP 5-34

BAW-2454

6. Neutron Fluence 6.1.

Introduction Over the last fifteen years, Framatome ANP has developed a calculational based fluence analysis methodology,1 16 1 that can be used to accurately predict the fast neutron fluence in the reactor vessel using surveillance capsule dosimetry or cavity 'dosimetry (or both) to verify the fluence predictions. This methodology was developed through a full-scale benchmark experiment that was performed at the Davis-Besse Unit 1 'reactor,1 161 and the methodology is described in detail in Appendix D. The results of the benchmark experiment demonstrated that the accuracy of a fluence analysis that employs the Framatome ANP methodology would be unbiased and have a precision well within the U.S Nuclear Regulatory Guide 1.190 limit of 20%.117 The Framatome ANP methodology was used t calculate the neutron fluence exposure to the 2630 capsule of the San Onofre Unit 3 nuclear reactor. The methodology was also used to estimate fluences on the inner surface of the reactor'vessel, as well as at specified locations in the pressure vessel. The fast neutron fluence (E>1 MeV) at each location was calculated in accordance with the requirements of U.S. Nuclear Regulatory Guide 1.190.

The energy-dependent flux on the capsule was used to determine the calculated activity of each dosimeter. Neutron transport calculations in two-dimensional geometry were used to obtain energy dependent flux distributions throughout the core. Reactor conditions were representative of an average over the cycles 1-8a, 8b-1Oa and cycle 10b-11 irradiation periods. These periods were separated to incorporate the effect on neutron fluence of water temperature variations that occurred after 356.728 EFPD of operation in cycle 8, and after 3.565 EFPD of cycle 10. A power uprate during cycle 11 -was also explicitly accounted for. Geometric detail was selected to explicitly'represent the dosimeter holder and the reactor vessel. A m6re detailed discussion of the calculational procedure is given in Appendix D. The calculated activities were adjusted for known biases (photofission, short-half-life, U-235 impurity,'and non-saturation), and compared to measured activities directly. It is noted that these measurements are not used in any way to determine the magnitude of the flux or the fluence. The measurements are used only to show that he calculational results are reasonable, and to show that the resujlts for SONGS Unit 3 cycles 1 -11 are consistent with the FANP benchmark database uncertainties.

6-1 FRAMATOME ANP

BAW-2454 6.2.

Fluence Results The San Onofre Unit 3 dosimeter holder is located in the downcomer region of the reactor, 70 off the major axis.

The dosimetry of the 2630 capsule was located in the reactor for a total irradiation time of 5431.36 effective full power days (EFPDs) for cycles 1-11. The rated thermal power for the first ten cycles was 3390 MWt, after which a 1.42% power uprate to 3438MWt occurred shortly into cycle 11.

The incident fast fluence (E>1.0 MeV) was calculated on the inner surface of the reactor vessel.

The layout of the reactor vessel is shown in Figure 6-1. The capsule is located between the core barrel and pressure vessel cladding, and is mounted on the pressure vessel wall, with the centerline of the capsule at 217.756 cm. The capsule is divided into 7 compartments, 3 of which contain dosimetry. The dimensions and compartment identification numbers are shown in Figure 6-2.

The three dosimeter compartments (1, 4, and 7) are labeled in accordance with their position in the holder, and dosimeters are labeled according to their holder relative position. For example, a Fe-54 dosimeter in compartment 1 would be designated as Top Fe-54. The dosimeters, which measure the fast fluence (E>1 MeV) inside of the San Onofre Unit 3 reactor, are listed in Table 6-1. The axial positions of each dosimeter are also listed relative to the bottom of the DORT model.

The fluence on the center of the capsule must be estimated for the center of the dosimetry capsule in order to allow for analysis of the Charpy and Tensile specimens. Doing this calculation for the cycle 1-11 analysis results in a maximum capsule fluence of 2.471 1E+19 n/cm2.

Flux estimates were also made on the inner surface of the reactor vessel and the vessel/clad interface. These estimates are of particular importance in determining the effect of neutron fluence on the properties of the vessel surface. The points of interest, and their respective three-dimensional coordinates, relative to the DORT origin, are shown in Table 6-2, with the inner surface radius given first.

The three-dimensionally synthesized fluxes at the inside surface of the vessel and vessel/clad interface are given in Table 6-3 for each point of interest over the cycle 1-8a, 8b-lOa, and Ob-1 irradiation periods. The azimuthal angles for the upper and lower shells shown in Table 6-3 were determined from the location of the maximum fluxes of Table 6-3.

A 6-2 FRAMATOME ANP

BAW-2454 Fluences for the vessel can also be extrapolated to longer time periods in order to estimate total fluences on the points of interest. This extrapolation is performed by assuming that the average fluence on the vessel for the extrapolated time is at equilibrium at the cycle 11 average fluence, which accounts for power uprate conditions which were completed during cycle 11. This assumption is acceptable since each cycle is an approximate equilibrium cycle, which is expected to continue throughout the lifetime of the plant, and the utility is expected to continue the use of a low leakage core design. End of life fluences are determined by taking the cumulative fluence and then extrapolating forward. The cumulative'fluence values for cycles 1-8a, 8b-lOa, and Ob-11 are shown in Table 6-4, along with the extrapolated EOL fluence, for the vessel wetted surface and Table 6-5 for the vessel/clad interface position. The end of life (32 EFPY) fluences' are calculated using the following formula:

F(EOL) = F(EOCI ) + (,

  • (tIoL (S) - trcx I (s))),

where F(EOL) is the fluence estimate at the end of life (32 EFPY), F(EOCI 1) is the fluence at the end of cycle 11, 4 I" is the flux for cycle 11, tEOL(S) is the total number of EFPS at the 32 EPFY end of life (1.0098E9 s), and tEOCI I(s) is the total number of EFPS accumulated through the end of cycle 11 (4.70997E8 s).

6.3.

Dosimetry Activity The ratio of the specified activities to the measured specific activities (C/M) is presented in Table 6-6 for cycles 1-1 1. In this table, overall average is the average C/M for the entire capsule.

6.4.

Capsule Lead Factors Lead factors between the vessel and capsule can be determined from the cumulative fluence values at EOC 11. This factor is determined by taking the cumulative fluence on the capsule at EOC 11, 2.471 lE+19 n/cm2 and dividing that value by the EOC 11 maximum vessel/clad interface fluence (Table 6-5), 1.9267E+1 9 n/cm2. Performing this calculation results in a lead factor for the capsule to vessel surface of 1.2825. Lead factors can also be determined for the 4 T and 3/4 T vessel positions using the methodology outlined in Reg Guide 1.99.[(8] The equation given for determining the attenuated fluence (E> 1.0 MeV, 10O9 n/cm2) is given in the Reg Guide as:

f=f5Uf (e 24x), where A

6-3 FRAMATOME ANP

BAW-2454 f is the fluence at the point desired, fsurf is the fluence on the wetted surface of the vessel, and x is the distance, in inches, of the desired position, as measured from the wetted surface. The thickness of the reactor vessel wall is 21.9075 cm (8.625 inches). The cladding thickness is 0.555625 cm (.21875 inches). Therefore, x for the /4 T position is 2.375 inches, and for the /4 T position, x is 6.6875 inches. Using the maximum fluence on the wetted surface at EOC 11 (Table 6-4), 2.01 E+19 n/cm2, the /4 T fluence is calculated to be 1.1373E+19 n/cm2, and the /4 T fluence is calculated to be 4.0401+18 n/cm2. Using these fluence values with the capsule fluence, the lead factor between the capsule and /4 T position is found to be 2.1727, while for the

/4 T position, the lead factor is found to be 6.1164.

A 6-4 FRAMATOME ANP

BAW-2454 Table 6-1. San Onofre Unit 3 Fast Fluence Dosimetry Dosimeter l

Can Axial position (cm)

Fe-54 Top 351.72 Fe-54 Mid 247.745 Fe-54 Bot 144.08 Cu-63 Top 354.26 Cu-63 Mid 250.28 Cu-63 Bot 146.62 Ni-58 Top 354.26 Ni-58 Mid 250.28 Ni-58 Bot 146.62 U-238

- Top 354.26 U-238 Mid 250.28 U-238 Bot 146.62 3D Coordinates for San Onofre Unit 3 Points of Interest Table 6-2.

Point of Interest J R position (cm) l 0 coordinate (°)

l Z coordinate (cm)

Intermediate Shell 220.98/221.536 Oto45 218.569 to 451 Lower Shell 220.98/221.536 0 to 45 0to 218.569 I.S. Max 220.98/221.536 0 to 45 0 to 451 14T 223.19 max max

'__3/4 227.51 max max

1 A

FRAMATOME ANP 6-5

BANV-2454 Table 6-3. 3D Synthesized Fluxes Vessel/Clad Interface Cycles 1-8a R (cm)

Theta ()

Z (cm) 3D Flux (n/cm2/s)

Intermediate Shell Max 221.5356 0.74111 281.5400 4.2385E+10 Lower Shell Max 221.5356 0.74111 214.5270 4.1243E+10 inside surface max 221.5356 0.74111 281.5400 4.2385E+10 Cycle 8b-lOa R (cm)

Theta (°)

Z (cm) 3D Flux (nlcm2/s)

Intermediate Shell Max 221.5356 0.74111 281.5400 3.7554E+10 Lower Shell Max 221.5356 0.74111 214.5270 3.6368E+10 Inside Surface Max 221.5356 0.74111 281.5400 3.7554E+10 Cycle lOb-l R (cm)

Theta (0)

Z (cm) 3D Flux (n/cm2 /s)

Internediate Shell Max 221.5356 0.74111 285.8500 3.8715E+10 Lower Shell Max 221.5356 0.74111 216.8570 3.7422E+10 Inside Surface Max 221.5356 0.74111 285.8500 3.8715E+10 Wetted Surface l

Cycles 1-8a R (cm)

Theta (0)

Z (cm) 3D Flux (n/cm2 /s)

Intermediate Shell Max 220.98 0.74111 285.3500 4.4299E+10 Lower Shell Max 220.98 0.74111 214.5270 4.3007E+10 Inside Surface Max 220.98 0.74111 285.3500 4.4299E+10 Cycle 8b-lOa R (cm)

Theta (0)

Z (cm) 3D Flux (n/cm2/s)

Intermediate Shell Max 220.98 0.74111 286.6000 3.9317E+10 Lower Shell Max 220.98 0.74111 214.5270 3.8043E+10 Inside Surface Max 220.98 0.74111 286.6000 3.9317E+10 Cycle 1Ob-11 R (cm) l Theta()

l Z (cm) l 3D Flux (n/cm2/s)

Intermediate Shell Max 220.98 0.741 11 286.6000 T 4.0162E+10 Lower Shell Max 220.98 0.74111 214.5270 j 3.8786E+10 inside surface max 220.98 0.74111 J 286.6000 l 4.0162E+10 A

FRAMATOME ANP 6-6

BAW-2454

.. Table 6-4. Cumulative Fluence Estimates at the Wetted Surface Flux Location Cumulative Fluence (n/cm2)

Extrapolated Fluence (n/cm2)

EOC 8a EOC 10a EOC 11 EFPY's 9.591 11.702 14.925 32 EFPY EFPY EFPY EFPY Intermediate Shell Max 1.3407E+19 1.603E+19 2.011 lE+19 4.1907E+19 Lower Shell Max 1.3016E+19 1.555E+19 1.9496E+19 4.0545E+19 Inside Surface Max 1.3407E+19 1.603E+19 2.011 1E+19 4.1907E+19 Table 6-5. Cumulative Fluence Estimates at the Vessel/Clad Interface Flux Location Cumulative Fluence (n/cm2)

Extrapolated Fluence (n/cm2)

EOC 8a EOC 10a EOC 11 EFPY's 9.591 11.702 14.925 32 EFPY EFPY EFPY

- EFPY Intermediate Shell Max 1.2828E+19 1.5330E+19 1.9267E+19 4.0278E+19 Lower Shell Max 1.2482E+19 1.4905E+19 1.8711E+19 3.9020E+19 Inside Surface Max 1.2828E+19 1.5330E+19 1.9267E+19 4.0278E+19 Table 6-6. C/MI Ratios Cycles 1-11 Dosimeter Calculated I

Measured C

Dosimeter l

("i Cii/g)

(PCi/g)

Capsule Average Top Fe 1523.768 1580.000 Middle Fe 1464.528 1446.000 Bottom Fe 1500.490 1427.000 Top Ni 1959.008 2108.000 Middle Ni 1873.381 1782.000 Bottom Ni 1929.600 1994.000 0.93636 Top Cu 12.048 14.890 Middle Cu 11.888 14.170 Bottom Cu 12.172 16.520 Top Sh U238 26.091 28.470 Middle Sh U238 25.847 27.230 Bottom Sh U238 26.490 26.260 A

FRAMATOME ANP 6-7

BAW-2454 Figure 6-1. San Onofre Unit 3 Core Overview Figure 6-2. San Onofre Unit 3 2630 Dosimetry Locations 35.40 cm I

2

__ 33.496 cm 3

35.084 cm 4 *t----- 33.338 cm 35.56 cm iS 0

5 34.7663 cm 6

3 8.656 cm --

~ 7_

Dwin not to scale A

FRAMATOME ANP 6-8

BAW-2454

7. Discussion of Capsule Results 7.1. Chemical Composition Data In addition to the 2630 capsule weld and base metal chemical analysis of broken Charpy specimens, chemical analyses results were reported in the San Onofre Unit 3 RVSP baseline and 970 capsule reports. These analyses were performed on the unirradiated surveillance materials and broken Charpy specimens. The chemical compositions of the base and weld metals are presented in Tables 7-1 and 7-2, respectively. The similarity of the chemical contents provides evidence that the specimens are of the same heat of material, with the exception of silicon, where the measurement from the 970 capsule is significantly lower than the other two measurements. The mean copper and nickel contents for the San Onofre Unit 3 RVSP base metal plate and veld metal represent the best-estimate chemical contents for these surveillance materials.

7.2. Unirradiated Material Property Data The base metal and weld metal were selected for inclusion in the San Onofre Unit 3 surveillance program in accordance with the regulations in effect at the time the program was designed. The applicable selection criterion was based on the unirradiated properties of the San Onofre Unit 3 reactor vessel beltline region materials only.

The unirradiated mechanical properties for the San Onofre Unit 3 RVSP materials are summarized in Appendices A and B of this report. The original unirradiated Charpy impact data were evaluated based on hand-fit Charpy curves generated using engineering judgement.[9 1 These data were re-plotted and re-evaluated herein using a hyperbolic tangent curve-fitting program in order to be consistent with the 263° capsule Charpy curves. Appendix C contains a comparison of the Charpy V-notch shift results for each surveillance material: original-fit versus current hyperbolic tangent curve-fit for the unirradiated and 970 capsule data. The hyperbolic tangent curve fitting procedure used herein is consistent with current industry practice for fitting Charpy impact data.[191 7.3. Irradiated Property Data In addition to the 2630 capsule mechanical test data, surveillance data are also available from the 970 San Onofre Unit 3 RVSP capsule. Westinghouse performed the testing and evaluation for the 970 capsule. 1 21 A

7-1 FRAMATOME ANP

BAW-2454 7.3.1. Tensile Properties Table 7-3 compares the irradiated and unirradiated tensile properties. Review of the surveillance tensile test data indicates that the ultimate strength and yield strength changes in the base metal plate as a result of irradiation and the corresponding changes in ductility are within the ranges observed for similar irradiated materials. The changes in tensile properties for the surveillance weld metal, as a result of irradiation, are also within the observed ranges for similar irradiated materials. The general behavior of the tensile properties as a function of neutron irradiation is an increase in both ultimate and yield strength and a decrease in ductility as measured by both total elongation and reduction in area.

7.3.2. Impact Properties Tables 74 and 7-5 compare the measured changes in irradiated Charpy V-notch impact properties from the 2630 capsule with the predicted changes in accordance with Regulatory Guide 1.99, Revision 2.118]

None of the measured 30 ft-lb transition temperature shifts for the materials in the 263° capsule are within one standard deviation of the shift predicted using Regulatory Guide 1.99, Revision 2, Position 1. (See Table 7-4). The measured shift for the base metal plate and HAZ is greater than the shift predicted using Regulatory Guide 1.99, Revision 2, Position 1.1 plus 2aA. The measured shift for the weld metal is greater than the shift predicted using Regulatory Guide 1.99, Revision 2, Position 1. plus I A, but within Regulatory Guide 1.99, Revision 2, prediction plus/minus 2a,.

The measured shift for the SRM is less than the shift predicted using Regulatory Guide 1.99, Revision 2, Position 1.1 minus I aA, but within Regulatory Guide 1.99, Revision 2, prediction plus/minus 2 aA.

The measured upper-shelf energies for the San Onofre Unit 3 2630 capsule surveillance materials do not fall below the required 50 ft-lb limit. The measured percent decrease in CvUSE for the surveillance base metal plate, weld metal, HAZ, and SRM are in good agreement with the values predicted using Regulatory Guide 1.99, Revision 2 (see Table 7-5).

The radiation-induced changes in toughness of the San Onofre Unit 3 surveillance materials are summarized in Table 7-6. The original irradiated Charpy impact data for the 970 capsule were evaluated based on hand fit curves.12 1 These data were re-plotted and re-evaluated in Appendix B using a hyperbolic tangent curve-fitting program to be consistent with the 2630 capsule Charpy curves. In addition, Appendix C contains a comparison of the Charpy V-notch shift results for each surveillance material (hand fit versus current hyperbolic tangent curve-fit).

A 7-2 FRAMATOME ANP

BAW-2454 7.3.3. Hardness No baseline hardness readings are available. Hardness values were reported in the 970 capsule report for the base, weld and HAZ samples. The increase in hardness with irradiation is as expected for these materials (See Table 7-7).

7.4. Surveillance Capsule Withdrawal Schedule The withdrawal schedule listed in Table 7-8 meets the requirements of ASTM El 8 5-82.

7.5 Predicted Adjusted Reference Temperatures The limiting material with respect to the adjusted reference temperature (ART) in the San Onofre beltline region is the plate C6802-1, which is the plate contained in the surveillance program. A, description of the San Onofre Unit 3 reactor vessel beltline materials is contained in Appendix E.

Results from plant specific surveillance programs must be integrated into the RTpTS estimate if the surveillance data have been deemed credible as judged by the following criteria (the credibility criteria are the same for 10 CFR 50.61 and Regulatory Guide 1.99, Revision 2):

Criterion 1.

Materials in the surveillance capsules must be those which arc the controlling materials with regard to radiation embrittlement.

Surveillance data available include the following materials:

Base Metal Plate C6802-1: Heat No. C9195-2 Weld Metal: Wire Heat No. 90069 All these heats of material lie within the reactor vessel beltline region of the San Onofre Unit 3 reactor vessel. Therefore, these materials could be controlling with regard to radiation embrittlement.

Criterion 2.

Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions must be small enough to permit the determination of the 30 ft-lb temperature unambiguously.

The Charpy V-notch data from 97° and 2630 capsule data for the plate C6802-1 permit the reasonable determination of the 30 ft-lb temperatures. However, for the 263°.capsule weld data the 30 ft-lb temperature is ambiguous.

A 7-3 FRAMATOME ANP

BAW-2454 Criterion 3.

Where there are to or more sets of surveillance data from one reactor, the scatter of ARTNDT values must be less than 280F for welds and 17'F for base metal. Even if the fluence range in the capsule fluence is large (two or more orders of magnitude), the scatter may not exceed tvice those values.

The scatter of the measured ARTNDT values for the available surveillance data are presented in Table 7-9. The scatter of the measured ARTNDT values is less than 171F for base metal Plate C6802-1, therefore the surveillance data is credible. The weld was not evaluated since the 30 ft-lb transition temperature is ambiguous (See Figure 5-17) and it is not the limiting material.

Criterion 4.

The irradiation temperature of the Charpy specimens in the capsule must equal the vessel wall temperature at the cladding/base metal interface within

+/-250F.

The San Onofre Unit 3 capsules are positioned inside the reactor vessel between the core support barrel and the vessel wall. Therefore, the irradiation temperature of the San Onofre Unit 3 capsule specimens is considered to be within 250F of the reactor vessel inside surface cold-leg temperature.

Criterion 5.

The surveillance data for the correlation monitor material in the capsule; if present, must fall within the scatter band of the database for the material.

The San Onofre Unit 3 2630 Capsule included Heavy Section Steel Technology (HSST) Plate 01 correlation monitor material. The measured value is 1.3 standard deviations below the predicted value. The measured shift is low, but not unreasonable. The measured shift has good agreement with other industry data.

Based on the above evaluation of the five criteria for assessing credibility that is called out in 10 CFR 50.61 and Regulatory Guide 1.99, Revision 2, the San Onofre Unit 3 surveillance base metal plate C6802-1 are credible, while the weld material heat 90069 is not credible.

The chemistry factor from the Regulatory Guide 1.99, Revision 2, Table 2 is 37.0, however, the surveillance data shows a greater credible shift than the Regulatory Guide 1.99, Revision 2 prediction. The best fit slope to the 970 and 263° capsule data yields a chemistry factor of 71.7 (See Table 7-9). Since the surveillance data is credible, the cA margin term can be cut in half yielding 8.50F. The 20 and 32 EFPY prediction for the vessel/clad interface pressurized thermal A

7-4 FRAMATOME ANP

BAW-2454 shock reference temperature and the /4 T and 3/4 T adjusted reference temperatures are shown in Table 7-10.

7.6 Predicted Decrease in Charpy Upper Shelf Energy The curves in Figure 2 in Regulatory Guide 1.99, Revision 2 are straight lines described by the following equations given in NUREG/CR-5799 12 0 The equation of the upper-bound curve for the Charpy upper shelf energy (USE) prediction is:

AUSE(%) = 42.39f 0-1502 wherefis fluence in units of 1019 n/cm2. The inside surface maximum fluence with the Regulatory Guide 1.99, Revision 2 attenuation to the /4 T vall location was conservatively used for all locations. Refer to Table 7-10 for the maximum /4 T wall fluence.

The equation of the lower curves for base metal is:

AUSE(%) = (I OCU + 9)f 0.2368 where Cu is wt% Cu. The equation of the lower curves for weld metal is:

AUSE(%) = (lOOCu + 14)f 02368 Using the above equations, the reactor vessel beltline material predicted USE is calculated as follows:

Predicted USE = Initial USE *

- AUSE(%)]

Results of application of the above equations to the San Onofre Unit 3 beltline materials are presented in Table 7-11.

For the following reasons, the surveillance data was not used to predict the USE for the surveillance materials contained in the reactor vessel:

A 7-5 FRAMATOME ANP

BAW-2454 the measured drop in USE values for the 263° surveillance capsule materials irradiated to about the same fluence as the T wall peak fluence were smaller than or close to the Regulatory Guide 1.99, Revision 2 prediction (See Table 7-5), and the USE values are not close to the 10 CFR 50 Appendix G 50 ft-lb limit.

A 7-6 FRAMATOME ANP

BAW-2454 Table 7-1.

Chemical Composition Data for San Onofre Unit 3 Reactor Vessel Surveillance Base Metal Plate C-6802-1 Element l_____________ -Composition

(% Weight)

Element_____I Baseline 9 l

970

[

2630 l

Averagea).

Copper 0.05 0.058 0.06 0.06 Nickel 0.57 0.575 0.54 0.56 Phosphorous 0.008 0.009

< 0.02 0.009 Chromium 0.07.

0.101 0.10 0.09 Molybdenum 0.58 0.531 0.53 0.55 Vanadium 0.005 0.014

< 0.004 0.010 Manganese 1.38 1.307 1.25 1.31 Cobalt 0.010

< 0.010 0.01 0.01 Sulfur

.0.014 0.0135 0.07 0.03 Silicon 0.23 0.048 0.27 0.18 (a).

Measurements below the detection limit were not included in the average.

Table 7-2.

Chemical Composition Data for San Onofre Unit 3 Reactor Vessel Surveillance Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069)

Element Composition (% Weight) l lemen l

BaselineT9J 970 (b) 2630 l

Averagea)

Copper 0.03 0.033 0.03 0.03 Nickel 0.08 0.101 0.11 0.10 Phosphorous 0.004 0.013

< 0.02 0.009 Chromium 0.05 0.062 0.07 0.06 Molybdenum 0.55 0.554 0.53 0.54 Vanadium 0.005 0.015

< 0.005 0.010 Manganese 1.54 1.447 1.27 1.42 Cobalt 0.015 0.011 0.02 0.02 Sulphur 0.009 0.0075 0.06--

0.03 Silicon 0.39 -

0.383-

.0.39 0.39 (a).

Measurements below the detection limit were not included in the average.

(b).

Average of two specimens.

7-7 FRAMATOME ANP

BAW-2454 Table 7-3. Summary of San Onofre Unit 3 Reactor Vessel Surveillance Capsules Tensile Test Results Strcngth, ksi Ductii ty_ %

Fluence, Test 1

1 1

Total 1 Reduction Material 1019 n/cm2 Temp., F Yield l

/Ca) l J

(a)

Elong.

l

/

l of Area l

(_)ll Base Metal Plate 0.00 71 68.11 90.111) 27kb) 68( )

C-6802-1 250 5 9.2(b) 83.6(b) 23(b) 66(b)

(Transverse Orientation) 550 5 8.6(b) 87.2(b) 24(b) 63(b) 0.8 74 71.3 4.6 91.7 1.7 23

-14 70 4

200 67.7 14.4 86.6 3.6 23

-1 72 9

550 60.1 2.6 86.6

-0.7 22

-7 70 11 2.471 70 71.2 4.5 92.2 2.3 24

-10 64

-6 250 66.6 12.6 85.8 2.7 22

-6 65

-1 550 64.9 10.8 90.1 3.3 19

-21 53

-17 Weld Metal 0.00 71 52.4()

65 C-6802-2/C-6802-3 250 48.6(b) 73.41-b) 23(b) 67(b) l (Wire Heat 90069) 550 55.5(b) 85.8(b) 20(b) 60(b) 0.8 74 51.4

-1.8 75.4

-10.7 20

-9 69 6

250 62.6 28.7 93.7 27.7 20

-l4 70 4

550 54.5

-1.7 94.7 10.4 23 15 64 7

2.471 70 66.0 26.0 94.7 12.2 25 14 63

-4 250 54.0 11.0 72.2

-1.6 20

-13 67 0

550 53.4

-3.7 75.8

-11.6 17

-14 59

-1 (a) Change relative to unirradiated material property.

(b) Mean value of available test data.

A FRAMATOME ANP 7-8

BAW-2454 Table 7-4. Measured vs. Predicted 30 ft-lb Transition Temperature Changes for 2630 San Onofre Unit 3 Capsule Surveillance Materials - 2.471 x 1019 n/cm2 Measured 30 ft-lb Transition 30 ft-lb Transition Temperature Shift Predicted in Temperature,F Accordance With Regulatory Guide 1.99, Revision 2 l

l

~~~~Chemistry Material Unirradiated Irradiated l Difference Factor l ARTNDTC), F j A F l ARTNDT - a, F J ARTNDT+CA, F Base Metal Plate C-6802-1 36 132 96 37.0()

46.0 17 29.0 63.0 (Transverse Orientation)

Weld Metal, C-6802-2/C-6802-3

-27 45 72 28.5(a) 35.4 28 7.4 63.4 (Wire Heat 90069)

Heat-Affect-Zone Material

-13 90 103 37.0()

46.0 17 29.0 63.0 C-6802-i Plate 01 15

.11237b 677 1.10 Standr frencISTe 1 5 157 142 131.7()

163.7 1 7 146.7 180.7 (a) Chemistry factor based on mean copper and nickel contents as shown in Tables 7-1 and 7-2.

(b) Chemistry factor based on copper and nickel contents as shown in NUREG/CR-655 1.11 (c) ARTNDT = Chemistry Factor

  • fluence factor (using the 2630 capsule fluence).

A FRAMATOME ANP 7-9

BAW-2454 Table 7-5. Measured vs. Predicted Upper-Shelf Energy Decreases for the 2630 San Onofre Unit 3 Capsule Surveillance Materials - 2.471 x 1019 n/cm2

% Decrease Predicted Measured Upper-Shelf Energy, ft-lb In Accordance With Regulatory Guide 1.99, Rev. 2 Material Unirradiated Irradiated

% Decrease Figure 2()

Base Metal Plate C-6802-1 92 74 19.6 18.6(b)

(Transverse Orientation)

Weld Metal, C-6802-2/C-6802-3 8 1 67 16.3 21.1(C)

(Wire Heat 90069)

Heat-Affect-Zone Material 85 69 18.8 18.6(b)

C -6 802-I1_

Standard Reference Material, HSST Plate 01 134 90 32.8 32.7(d)

(a) Calculated using equation reported in NUREG/CR-5799.12 °0 (b) Based on mean copper content as shown in Table 7-1.

(c) Based on mean copper content as shown in Table 7-2.

(d) Based on mean copper content as shown in NUREG/CR-655 I.1 91 A

FRAMATOME ANP 7-10

BAW-2454 Table 7-6. Summary of San Onofre Unit 3 Reactor Vessel Surveillance Capsules Charpy Impact Test Results (B ased on Tanh Reevaluation) l I

Measured Measured Transition Temperature Upper-Shelf

Fluence, ACv30,
ACv5O, Energy, Material Capsule l09 n/cm2 F

F ft-lb

% Decrease Base Metal Plate Baseline 92 C-6 802-1 (Transverse Orientation) 97 0.8 58 81 76 17.4 263 2.471 96 94 74 19.6 Weld Metal, Baseline 81 C-6802-2/C-6802-3Baeie-- (Wire Heat 90069) 97 0.8 30 34 69 14.8 263 2.471 72 80 67 17.3 Heat-Affected Zone Material Baseline 85 C-6802-1

'97 0.8 57 58 74 12.9 263 2.471 103 72 69 18.8 Standard Reference Material, Baseline 134 HSST Plate 01 263 2.471 142 159 90 32.8 A

FRAMATOME ANP 7-11

BAW-2454 Table 7-7. Hardness Data for San Onofre Unit 3 Reactor Vessel Surveillance Materials Average Rockwell Hardness B Scale Element 970 2630 relative to 970 Base Metal Plate 86.2 92.7

+ 8 %

C-6802-1 (Transverse)

Weld Metal, C-6802-21C-6802-3 89.8 92.2

+ 3 %

(Wire Heat 90069)

Heat-Affected Zone 88.5 98.1

+ 11 %

Material, C-6802-1 Standard Reference 97.8 Material, HSST Plate 01 Table 7-8. Surveillance Capsule Withdrawal Schedule Capsule -

Estimated Location Capsule Lead Removal Fluence (Degrees)

Factor (EFPY)

(x 1019 n/cm2 )

97 1.21 4.33 0.8 263 1.28 14.9 2.471 83(a) 1.28 25.0 4.03 277(a) 1.28 Standby 104 0.86 Standby 284 0.86 Standby ta)Either the 830 or the after 25.0 EFPY with Standby.

2770 Capsule can be withdrawn the remaining capsule serving as A

FRAMATOME ANP 7-12

BAW-2454 Table 7-9. Credibility Assessment for San Onofre Unit 3 Reactor Vessel Beltline Limiting Plate C-6802-1 Predicted (Measured -.

Fluence Measured ARTNDT from Predicted)

Measured Capsule (x1

9)

Fluence

ARTNDT, Best Fit Line(a),
ARTNDT, ARTNDT
  • ff ff' n/cm2 Factor (ff)

F F

F 97 0.8 0.937 58 T 67.2

-9.2 54.4 0.879 263 2.471 1.243 96 j

89.1 6.9 119.4 1.546 (a) Predicted ARTNDT = (Slope best fit) * (Fluence Factor) and Slope best it = I(ARTNDT

  • ff) / Z(fff) = 71.7 0F) 0 7-13 A

FRAMATOME ANP

BAW-2454 Table 7-10. Predicted Adjusted Reference Temperature for the Uprated Peak Fluence in San Onofre Unit 3 Reactor Vessel Beltline Limiting Plate C-6802-1 Peak Fluence Location Extrapolated Chmsty Initial EFPY (neiae ll)

Uprated Fluence, Fluence tr RTNDT

Margin, ARTNDT,
ARTNDT,

_________erediteShel)xl 019 n/cm' Factor jFactor(c)

F F

F(a)

F (b)

Inside Wetted Surface 2.659(d) 20 Vessel/Clad Interface 2.551(e) 1.251 71.7

+40 17 89.7 146.7 l/4T Location 1.504(h) 1.113 71.7

+40 17 79.8 136.8 3/4T Location 0.534(")

0.825 71.7

+40 17 59.2 116.2 Inside Wetted Surface 4.191(0 9

32 Vessel/Clad Interface

4. 028(g) 1.358 71.7

+40 17 97.4 154.4 l/4T Location 2.370(11) 1.233 71.7

+40 17 88.4 145.4 3/4T Location 0.842(h) 0.952 71.7

+40 17 68.3 125.3 (a)

ARTNDT = Chemistry Factor

  • Fluence Factor.

(b)

ARTNDT = Initial RTNDT + ARTNDT + Margin.

(c)

Calculated per RG 1.99 Rev. 2 position 2; best fit of the surveillance data (Section 7.5).

(d)

Interpolated from information in Table 6-4 (e)

Interpolated from information in Table 6-5 (f) from Table 6-4 (g) from Table 6-5 (h)

Attenuation calculated per RG 1.99 Rev. 2;['81 see Section 6.2 for equation and dimensions used.

A 7-14 FRAMATOME ANP

BAW-2454 Table 7-11. Predicted Charpy Upper Shelf Energy at 32 EFPY for the San Onofre Unit 3 Reactor Vcssel Beltline Region Materials at the 1/4T Location with Power Uprate*

Material Description Copper Initial Predicted Predicted Reactor Vessel Matl.

lHeat Composition CvUSE, CvUSE Decrease in Beltline Location Ident.

l Number wt%

ft-lbs ft-lbs CvUSE (%)

Inter. Shell Long. Weld 2-203A 83650 0.04 136 106 22.1 Inter. Shell Long. Weld 2-203B 83650 0.05 136 104 23.3 Inter. Shell Long. Weld 2-203C 83650 0.04 136 106 22.1 Lower Shell Long. Weld 3-203A 88114 0.04 161 125 22.1 Lower Shell Long. Weld 3-203B 88114 0.04 161 125 22.1 Lower Shell Long. Weld 3-203C 88114 0.04 161 125 22.1 Upper/Inter. Shell Girth Weld 8-203 88118 0.05 125 96 23.3 Lower/Inter. Shell Girth Weld 9-203 90069 0.06 123 93 24.5 90144 0.05 91 70 23.3 Intermediate Shell C-6802-1 C9195-2 0.06 95 78 18.4 Intermediate Shell C-6802-2 C9218-2 0.04 115 97 15.9 Intermediate Shell C-6802-3 C9195-1 0.06 105 86 18.4 Lower Shell C-6802-4 C9220-1 0.05 118 98 17.2 Lower Shell C-6802-5 C9218-1 0.04 116 98 15.9 Lower Shell C-6802-6 B3388-1 0.06 92 75 18.4

  • Calculated at the /4T peak reactor vessel fluence of 2.370 x10' 9 n/cm2.

A FRAMATOME ANP 7-15

BAW-2454

8. Summary of Results The analysis of the reactor vessel material contained in the second surveillance capsule, the 2630 capsule, removed for evaluation as part of the San Onofre Unit 3 Reactor Vessel Surveillance Program, led to the following conclusions:
1. The capsule received a maximum fast neutron fluence of 2.471 x 10 9 n/cm2 (E > 1.0 MeV).
2. Based on cycle 11 flux (including power uprated conditions) extrapolated to 32 EFPY, the projected peak fast fluence of the San Onofre Unit 3 reactor vessel beltline region clad/vessel interface is 4.028 x 10'9 n/cm2 (E > 1.0 MeV).
3. The 30 ft-lb transition temperature for the Base Metal Plate C-6802-1, in the transverse orientation, increased 961F after irradiation to 2.471 x 10'9 n/cm2 (E > 1.0 MeV),

which is higher than predicted by Regulatory Guide 1.99, Revision 2, therefore the surveillance data was used to establish the 32 EFPY adjusted reference temperature prediction for this beltline limiting material. In addition, the CVUSE for this material decreased 20%.

4. The 30 ft-lb transition temperature for the Weld Metal, C-6802-2/C-6803-3 (Heat 90069), increased 720F after irradiation to 2.471 x 10'9 n/cm2 (E > 1.0 MeV), which is higher than predicted by Regulatory Guide 1.99, Revision 2. In addition, the CUSE for this material decreased an expected 16%.
5. The 30 ft-lb transition temperature for the heat affected zone increased 1031F after irradiation to 2.471 x 10'9 n/cm2 (E > 1.0 MeV), which is higher than predicted by Regulatory Guide 1.99, Revision 2. In addition, the CUSE for this material decreased an expected 19%.
6. The 30 ft-lb transition temperature for the Standard Reference Material, HSST Plate 01, increased 1420F after irradiation to 2.471 x 10'9 n/cm2 (E > 1.0 MeV), which is A

8-1 FRAMATOME ANP

BAW-2454 lower than predicted by Regulatory Guide 1.99, Revision 2, however, it is in agreement with other industry data. In addition, the CUSE for this material decreased an expected 33%.

7. The measured upper-shelf energies for the San Onofre Unit 3 2630 capsule surveillance materials do not fall below the required 50 ft-lbs limit after irradiation to 2.471 x 10'9 n/cm2 (E > 1.0 MeV).
8. None of the San Onofre Unit 3 beltline materials are predicted to fall below the 50 ft-lb limit required in 10 CFR 50 Appendix G before 32 EFPY.

A 8-2 FRAMATOME ANP

BAW-2454

9. Certification The specimens obtained from the Southern California Edison Company's San Onofre Unit 3 reactor vessel surveillance capsule (the 2630 capsule) were tested and evaluated using accepted techniques and established standard methods and procedures in accordance with the requirements of 10 CFR 50.61, 10 CFR 50 Appendix G, and 10 CFR 50 Appendix H.

J. Bfi~all (Material Analysis)

Date Materials & Structural Analysis Unit JTW. Newman, 3i. (Fluence Analysis)

Date Effluence and Radiation Analysis Unit This report has been reviewed for technical content and aca 7, 247-,

//?,-/Loy Gram aterial Analysis)

Date Materials & Structural Analysis Unit J. N. Byard (Fluen6e Analysis)

Date Fluence and Radiation Analysis Unit Verification of independent review.

A. D. McKim, Manager Date Materials & Structural Analysis Unit This report is approved for release.

,t,6 *&A9L91Q/

I-22-O'/

K. E. Moore Date Program Manager A

9-1 FRAMATOME ANP

BAW-2454

10. References
1. "Program for Irradiation Surveillance ofSan Onofre Reactor Vessel Materials, San Onofre Nuclear Generating Station, Units 2 and 3, " S-NLM-002. Revision 2, Combustion Engineering, Inc., Windsor, Connecticut, October 1974.
2. E. Terek, E. P. Lippincott, and A. Madeyski, "Analysis of the Southern California Edison Company San Onofre Unit 3 Reactor Vessel Surveillance Capsule Removedfroin the 970 Location, " Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, March 1991.
3. ASTM Standard E 185, "Standard Recommended Practice for Surveillance Testsfor Aruclear Reactor Vessels, " American Society for Testing and Materials, Philadelphia, Pennsylvania.
4. Code of Federal Regulation, Title 10, Part 50, "Domestic Licensing of Production and Utilization Facilities, " Appendix G. Fracture Toughness Requirements.
5. Code of Federal Regulation, Title 10, Part 50, "Domestic Licensing of Production and Utilization Facilities, " Appendix H, Reactor Vessel Material Surveillance Program Requirements.
6. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, "Nuclear Power Plant Components, "Appendix G. Protection Against Nonductile Failure, 1989 Edition.
7. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components, " Appendix G. Fracture Toughness Criteria for Protection Against Failure, 1989 Edition.
8. ASTM Standard E 208-81, "Methodfor Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, " American Society for Testing and Materials, Philadelphia, Pennsylvania.
9. A. Ragl, "Southern California Edison San Onofre Unit 3 Evaluation of Baseline Specimens Reactor Vessel Materials Irradiation Surveillance Program, " TR-S-MCM-004, Combustion Engineering, Inc., Windsor, Connecticut, November 1979.
10. ASTM Standard E 21-92, "Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials, " American Society for Testing and Materials, Philadelphia, Pennsylvania.

A-

  • - Available from Framatome ANP, Lynchburg,iVirginia.

FRAMATOME ANP 10-1

BAW-2454

11. ASTM Standard E 23-91, "Standard Test Methods for Notched Bar Impact Testing of Mfetallic Mfaterials, " American Society for Testing and Materials, Philadelphia, Pennsylvania.
12. W. Oldfield, "Curve Fitting Impact Test Data: A Statistical Procedure," ASTM Standardization News, November 1975.
13. "M1etals Handbook Ninth Edition, " Volume 8, Mechanical Testing, American Society for Metals, Metals Park, Ohio, 1985.
14. ASTM Standard E 18-97, "Standard Test Methods for Rockwell Hardness and Rockwell Superficial Hardness of Metallic laterials, " American Society for Testing and Materials, Philadelphia, Pennsylvania.
15. ASTM Standard E 384-89, "Standard Test Methodfor Microhardness of Materials,"

American Society for Testing and Materials, Philadelphia, Pennsylvania.

16. J.R. Worsham, et al., "Fluence and Uncertainty Aethodologies," BAW-2241 P-A. Revision 1, Framatome ANP, Lynchburg, Virginia, April 1999.*
17. U.S. Nuclear Regulatory Commission, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," Regulatory Guide 1.190, March 2001.
18. U.S. Nuclear Regulatory Commission, "Radiation Embrittlement of Reactor Vessel Materials, " Regulatory Guide 1.99. Revision 2, May 1988.
19. E. D. Eason, J. E. Wright, and G. R. Odette, "Improved Embrittlement Correlationsfor Reactor Vessel Steels," NUREG/CR-655 1, November 1998.
20. R. D. Cheverton, T. L. Dickson, J. G. Merkle, and R. K. Nanstad, "Review of Reactor Pressure Vessel Evaluation Report for Yankee Row Nuclear Power Station," NUREG/CR-5799, March 1992.
21. Letter to U.S. NRC Document Control Desk from W. C. Marsh, SCE "Revision to Supplemental Response to Generic Letter 92-01, Revision 1, "Reactor vessel Structural integrity, I OCFR505() " San Onofre Nuclear Generating Station Units 2 and 3, " June 22, 1994.

A

FRAMATOME ANP 10-2

BAW-2454 APPENDIX A Unirradiated and Irradiated Tensile Data for the San Onofre Unit 3 RVSP Materials A

FRAMATOME ANP A-1

BAW-2454 Table A-1. Tensile Properties of Unirradiated Base Metal Plate C-6802-1, Transverse Orientation Specimen Test Strength, Ksi Elongtion, %

Reduction No.

Temp. (F) 0.2% Yield or Ultimate Unifor n Total of Area, %

Upper/Lower U

2J6 71 66.6/64.6 88.8 9.8 27 68.4 2JT 71 69.2/68.3 90.9 8.8 26 67.3 2JY 71 68.6/61.2 90.7 9.3 27 66.9 2K2 250 63.7/63.3 84.1 8.0 23 65.9 2K7 250 64.0/63.7 84.1 7.1 23 65.9 2KK 250 49.8/48.6 82.5 6.8 24 65.7 2KI 550 59.4/58.5 87.1 7.6 23 63.0 2J4 550 58.7/56.2 88.5 8.0 25 63.0 2J3 550 57.6/56.1 86.0 8.6 23 64.0 Table A-2. Tensile Properties of Unirradiated Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069)

Specimen Test Strength, Ksi Elongation, %

Reduction No.

Temp. (F) 0.2% Yield or Ultimate Uniforn Total of Area, %

Upper/Lower 3K7 71 52.5/51.0 87.8 9.7 21 62.9 3KJ 71 51.4/51.3 76.3 9.6 22 69.4 3JM 71 53.2/51.7 89.2 9.7 23 63.3 3JU 250 49.6/48.8 68.1 8.0 24 67.6 3JK 250 51.4/50.2 83.9 11.4 23 63.7 3K2 250 44.9/44.3 68.2 8.0 23 69.8 3JJ 550 60.1 95.1 19 53.0 3KU 550 47.0/44.1 74.2 7.7 20 65.9 3K5 550 59.3 88.0 8.6 21 60.8

  • Extensometer arms slipped on sample.

A FRAMATOME ANP A-2

BAW-2454 Table A-3. Tensile Properties of Unirradiated Heat Affected Zone Metal Specimen Test Strength, Ksi Elongation, %

Reduction No.

j Temp. (F)'

0.2% Yield or Ultimate Uniform Total of Area, %

Upper/Lower l 4J6 71 64.4/60.9 86.8 9.9 25 66.4 4J2 71 64.9/63.3 87.8 11.6 27 62.9 4KY 71 63.0/61.0 82.0 4K2 250 58.9/57.1 75.1 7.7 20 59.3 4K6 250 58.6/57.1 75.1 7.7 20 60.0 4KL 250 58.4/57.1 75.6 8.3 20 58.9 4JC 550 54.0/52.7 84.5 9.2 25 64.4 4J5 550 54.1/52.2 82.8 8.4 23 66.5 4JK 550 54.9/54.4 82.4:

8.8 21 67.2

  • Sample broke outside gage length.

Table A-4. Tensile Properties of Base Metal Plate C-6802-1, Irradiated to'8 x 1018 n/cm2 (E> 1.0 MeV)

Transverse Orientation

-Specimen Test 1 Strength, Ksi Elongation, %

Reduction.

No.

JTemp.-(F)_1 0.2% Yield I Ultimate Uniform l Total ofArea,%

2KE 74 71.3 91.7 12.0 23 70 2KC 200 67.7 86.6 10.5 23 72 2K3 550 60.1 86.6 10.5 22 70 A

FRAMATOME ANP A-3

BAW-2454 Table A-5. Tensile Properties of Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069)

Irradiated to 8 x 1018 n/cm2 (E> 1.0 MeV)

Specimen l Test l

Strength, Ksi Elongation, %

Reduction No.

l Temp. (F) l0.2% Yield l Ultimate l Uniform l Total of Area,%

3JC 74 51.4 75.4 11.4 20 69 3JE 250 62.6 93.7 10.5 20 70 3J4 550 54.5 94.7 13.5 23 64 Table A-6. Tensile Properties of Heat Affected Zone Metal Irradiated to 8 x 1018 n/cm2 (E> 1.0 MeV)

Specimen Test l

Strength, Ksi Elongation, %

Reduction No.

l Temp. (F) l_0.2% Yield __Ultimate _[Uniform Total l of Area,%

4JD 74 67.2 87.6 12.0 25 69 4JK 175 62.1 77.9 12.0 25 64 4JA 550 60.1 84.5 12.0 25 68 A

FRAMATOME ANP A-4

BAW-2454 APPENDIX B Unirradiated and Irradiated Charpy V-Notch Impact Surveillancc Data for the San Onofrc Unit 3 RVSP Materials Using Hyperbolic Tangent Curve-Fitting Method B-1 AM B-i1 FRA MATO ME AN P

BAW-2454 Table B-i. Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Base Metal Plate C-6802-1, Transverse Orientation Test Impact l Lateral Shear Specimen Temp.

Energy Expansion Fracture No.

(F)

(ft-lbs)

(mils)

(%)

252 23E 26J 22P 26D 22K 263 23Y 25M 24T 25E 267 23A 221 25Y 23P 255 21K 22A

-40 0

0 40 40 80 80 100 100 100 120 120 160 160 160 210 210 250 250 13 14 23 18 52 25 57 42 74 88 85 106 61 63 113 96 109 80 87 12 12 24 17 45 25 50 40 60 65 73 80 60 58 81 74 82 71 73 0

10 0

10 20 20 40 30 40 50 50 80 60 70 90 90 100 100 100 A

FRAMATOME ANP B-2

BAW-2454 Table B-2. San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Base Metal Plate C-6802-1, Irradiated to 8 x 1018 n/cm2 (E>1.0 MeV)

Transverse Orientation Test Impact Lateral Shear Specimen Temp.

Energy Expansion Fracture No.

(F)

(ft-lbs)

(mils)

(%)

25U 0

19 18 10 23K 25 16 11 10 21A 50 9

12 10 23L 75 21 20 15 22B 100 25 25 20 25J 115 55 48 45 25L 130 50 44 50 247 150 33 33 50 223 165 36 37 65 25C 200 92 73 100 245 225 62 59 100 23M 250 73 77 100 A

FRAMATOME ANP B-3

BAW-2454 Table B-3. Hyperbolic Tangent Curve Fit Coefficients for San Onofre Unit 3, Base Metal Plate C-6802-1, Transverse Orientation Hyperbolic Tangent Curve Fit Coefficients Absorbed Energy l

Lateral Expansion Percent Shear Fracture Unirradiated A:

47.1 A:

38.8 A:

50.0 B:

44.9 B:

37.8 B:

50.0 C:

71.3 C:

84.7 C:

77.5 TO:

64.9 TO:

59.7 TO: 110.9 970 Capsule A:

38.9 A:

57.5 A:

50.0 B:

36.7 B:

56.5 B:

50.0 C:

100.7 C:

169.6 C:

67.9 TO: 119.5 TO: 195.2 TO: 134.3 A

FRAMATOME ANP B-4

BAW-2454 Table B-4. Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Base Metal Plate C-6802-1 Longitudinal Orientation Test Impact Lateral Shear Specimen Temp.

Energy Expansion Fracture No.

(F) J (ft-lbs)

(mils)

(%)

151

0.

6.5 5

0 146 40 15 13 10 127 40 16 15

'10 11U 80 33.5 30

30 13M 80 39.5 36 30 14C 120 36 37 40 126 120 45 41 40 14B 160 51 47 60 13U 160 83 70 80 13E 210 66 62 70 14T 210 96 81 80 IIL 250 90

'80 100 1IC 250 93 87 100 A

FRAMATOME ANP B-6

BAW-2454 Table B-5. San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Base Metal Plate C-6802-1 Irradiated to 8 x 1018 n/cm2 (E>1.0 ?leV)

Longitudinal Orientation Test Impact Lateral Shear Specimen Temp.

Energy Expansion Fracture No.

(F)

(ft-lbs)

(mils)

(%)

147 50 6

5 5

153 75 22 15 15 15K 100 29 24 25 113 125 35 32 30 14Y 150 32 26 30 15B 165 45 43 45 14A 175 30 27 45 14E 200 46 46 55 lIT 225 75 66 80 12K 250 45 48 80 12L 250 48 46 95 14M 275 90 76 100 A

FRAMATOME ANP B-6

BAW-2454 Table B-6. Hyperbolic Tangent Curve Fit Coefficients for San Onofre Unit 3, Base Metal Plate C-6802-1 Longitudinal Orientation Hyperbolic Tangent Curve Fit Coefficients AbsorbedEnergy Lateral Expansion i Percent Shear Fracture Unirradiated A:

47.6 A:

47.0

'A:

50.0 B:

45.4 B:

46.0 B:

50.0 C:

94.9 C:

116.2 C:

93.3 TO: 120.0 TO: 133.1 TO: 131.4 970 Capsule A:

35.6 A:

48.5 A:

50.0 B:

33.4 B:

47.5 B:

50.0 C:

112.8 C:

164.5 C:

90.0 TO: 145.1 TO: 213.9 TO: 174.2

 I A

FRAMATOME ANP B-7

BAW-2454 Table B-7. Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069)

Test Impact Lateral Shear Specimen Temp.

Energy Expansion Fracture No.

(F)

(ft-lbs)

(mils)

(%)

31Y 37C 32C 36M 36C 36T 37E 3A2 35Y 34M 376 33K 316 344 32K 31C 35M 372 34E

-80

-80

-40

-40 0

0 40 40 80 80 120 120 160 160 210 210 250 250 250 5

21 17 34 36 48 58 63 72 80 71 84 64 97 80 97 71 78 84 5

19 18 31 34 45 53 56 65 82 65 79 67 90 85 84 69 73 82 0

20 20 30 30 40 40 60 80 100 80 90 70 100 90 90 100 100 100 B-8 A

FRAMATOME ANP

BAW-2454 Table B-8. San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069)

Irradiated to 8 x 1018 n/cm2 (E>1.0 MeV)

Test Impact Lateral-Shear Specimen Temp.

Energy Expansion Fracture No.

(F)

(ft-lbs)

(mils)(%)

363

-50 23 20 15 33L

-25 27 25 20 31D 0

5 6

5 37B 10 42 34 60 367 25 42 44 65 331 50 44 44 70 37U 60 61 57 95 371 80 65 59 95 365 105 67 60 100 34P 150 75 73 100 33D 190 77 76 100 377 225 63 58 100 A

FRAMATOME ANP B-9

BAW-2454 Table B-9. Hyperbolic Tangent Curve Fit Coefficients for San Onofre Unit 3 Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069)

Hyperbolic Tangent Curve Fit Coefficients Absorbed Energy l

Lateral Expansion [ Percent Shear Fracture Unirradiated A:

41.4 A:

40.1 A:

50.0 B:

39.2 B:

39.1 B:

50.0 C:

76.7 C:

86.2 C:

100.7 TO:

-4.4 TO:

-0.3 TO:

25.1 970 Capsule A:

35.8 A:

35.5 A:

50.0 B:

33.6 B:

34.5 B:

50.0 C:

70.6 C:

79.9 C:

44.5 TO:

14.5 TO:

19.8 TO:

16.6 A

FRAMATOME ANP B-10

BAW-2454 Table B-10. Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Heat-Affected-Zone Material Test Impact Lateral Shear Specimen Temp.

Energy Expansion Fracture No (F) -

(b-lbs)

-(mils)

(0/)

47L

-80 l

6.

4 0.

45B

-80 6

4 0

42Y

-40 30 26 10 436

-40 30 26 10 45J 0

30 25 20 45Y 0

46 37 30 42J 40 33 34 20 42K 40 35 32 20 42U 80 73 48 80 454 80 79 63 50 471 120 42 41 60 475 120 50 49 70 43C 160 61 62 80 43K 160 87 72 90 46K 210 57 57 70 444 210 105 73 100 46A 250 63 66 100 42T 250 75 68 100 46Y 250 98 80 100 A

FRAMATOME ANP B-11

BAW-2454 Table B-11. San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Heat-Affected-Zone Material, Irradiated to 8 x 10' n/cm2 (E>1.0 MeV)

Test Impact Lateral Shear Specimen Temp.

Energy Expansion Fracture No.

(F)

(fi-lbs)

(mils) ll 441

-75 9

8 5

42L

-50 18 16 15 45C

-20 3

4 5

43P 0

26 23 35 464 25 35 37 45 41C 60 33 33 50 47E 95 43 44 75 46T 125 41 38 70 45M 145 45 44 80 437 165 63 61 100 45P 200 84 68 100 434 250 75 65 100 A

FRAMATOME ANP B-12

BAW-2454

  • ,Table B-12. Hyperbolic Tangent Curve Fit Coefficients for San Onofre

-~Unit 3 Heat-Affected-Zone Material I l Hyperbolic Tangent Curve Fit Coefficients Weld Metal Absorbed Energy Lateral Expansion Percent Shear Fracture Unirradiated A:

43.7 A:

37.3 A:

50.0 B:

41.5 B:

36.3 B:

50.0 C:

148.9 C:

145.1 C:

103.2 l__________

TO:

37.9 TO:

38.3 TO:

76.3 970 Capsule A:

38.1 A:

38.9

A:

50.0 B:

35.9 B:

37.9 B:

50.0 C:

129.7 C:

164.7 C:

101.3 l_____-:___

TO:

73.6 TO:

88.0 TO:

51.6 A

FRAMATOME ANP B-13

BAW-2454 Table B-13. Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Standard Reference Material Test Impact Lateral Shear Specimen Temp.

Energy Expansion Fracture No.

(F)

(ft-lbs)

(mils)

J l B3L

-80 4.5 2

0 B41

-40 10 8

0 B3D 0

16 16 10 B3K 0

18 17 10 B3J 40 58 44 20 B3Y 40 61 49 20 B3M 80 70 57 40 B3C 80 77 60 40 B3U 120 107 70 70 B42 120 118 80 70 B3B 160 132 92 100 B43 160 140 93 100 B3E 210 128 87 100 B3T 210 134 87 100 A

FRAMATOME ANP B-14

BAW-2454 Table B-14. Hyperbolic Tangent Curve Fit Coefficients for San Onofre Unit 3 Standard Reference Material 1

Hyperbolic Tangent Curve Fit Coefficients Weld Metal Absorbed Energy l

Lateral Expansion T Percent Shear Fracture Unirradiated A:

67.9 A:

46.1 A:

50.0 B:

65.7 B:

45.1 B:

50.0 C:

71.6 C:

77.6 C:

61.3 TO:

61.7 TO:

50.7 TO:

88.7 A

FRAMATOME ANP B-15

BAW-2454 Figure B-i. Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Base Metal Plate C-6802-1 Transverse Orientation

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -

100 at 75 2

50 U.

0 25 co 0

-100

-50 0

50 100 150 200 250 300 350 400 Temperature, F 100 _

6 80-o W 60-

,:5 40 2

20 120 110 100 90 8

80 0

c 60

_w o

50.

0.

E 40 JT3 eE:

51.1 F I

0 0

0 I

D 0

100 200 Temperature, F 300 400 TSo:

69.5 F IT

36.

F I

0 CvUSE:

92.0 ft-lbs V0 0~~~~~

0~~~

  • /

0

  • /

0 A* --

30 20 10 0

- C10

-50 0

50 100 150 200 250 300 350 400 Temperature, F A

FRAMATOME ANP B-16

BAW-2454 Figure B-2. San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Base Metal Plate C-6802-1 Irradiated to.8 x 1018 n/cm2 (E>1.0 MeN)

Transverse Orientation

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -

100 c'

75 0

E! 50 U'.

c 25 U)

-1 e 100 e 80 0

60

. 40 w

2 20

-1 00

-50 0

50 100

- 150, 200 250 300 350 400 Temperature, F IT35MLE:

123.8 F 0

00 0

100 200 Temperature, F 300 400 120 110 100 90 U) 80 70 cn C

60 0

50-E 40 30 20 10

  • 0 T50 150.8 F T30:

94.5 F CvUSE:

75.7 f-lbs 0

0

0.

/

0

.100

.50 0

50 100 150

.200 250 300 350 400 Temperature, F A

B-17 FRAMATOME ANP

BAW-2454 Figure B-3. Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Base Metal Plate C-6802-1 Loigiitudinal Orientation

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -

ion.

a; a

E 0.

eK uJ

.t E2 C

w UL EU 75 -

50 -

25 -

0 0

u

-1t a

I I

I I

I I

I I

0O

-50 0

50 100 150 200 250 300 350 40C Temperature, F 10U 80 60 40 20

-1(

120 110 100 90 80 70 60 50 40 30 20 10 0

-1 IT35,x:

102.2 F 00 I

I 0

0 100 200 Temperature, F 300 400 T50:

125.1 F CvUSE:

93.0 -lbs 2

)0

-50 0

50 100 150 200 250 300 350 400 Temperature, F A

FRAMATOME ANP B-18

BAW-2454 Figure B-4. San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Base Metal Plate C-6802-1.

-Irradiated to 8 x 10 8 n/cm2 (E>1.0 McV)

Longitudinal Orientation

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -

100-e75 2L.

25 ~

~

~

00

-100

-50 0

50 100 150 200 250 300 350 400 Temperature, F 100

'- 80 2

CL

£ 60 x

40 Lu E

20

-A0 I T35MLE:

165.8 F 0

I

-100 0

100 200 Temperature, F 300 400 120 110 100 90 80 70 60 w

r 50 E

40 30 20 10 0

T50:

197.2 F T30:

126.0 F CvUSE:

69.0 ft-lbs 0

-100

-50 0

50 100

- 150 200 250 300 350 Temperature, F 400 A

FRAMATOME ANP

- B-19

BAW-2454 Figure B-5. Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069)

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -

co 2

to

-100

-50 0

50 100 150 200 250 300 350 400 Temperature, F

.0 100 c

80

.f2 coe 60 C

40 e

20 3

O i

-11.7 F l

0 I

I

-1 w

  • 100 0

100 200 Temperature, F 300 400 100 90 0

0 T[0:

12.7 F T3 :

-27.4 F CvUSE:

80.6 -lbs 80 I-70 -

0 w

'A a

0.

0 60 50....

40 -

S 30 20 -

  • 10 0

/

0 0

-100

-50 0

50 100 150 200 250 300 350 400 Temperature, F A

FRAMATOME ANP B-20

BAW-2454 Figure B-6. San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Weld Metal C-6802-2/C-6802-3 (Wire Heat 90069)

Irradiated to 8 x 1018 n/cm2 (E>1.0 McV)

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -

100 75 0

a 50 L.

c 25 U) 0 4

-11

"' 100 eF 80 0

2 60 M

0.u 40 E1 20 3

0 i

00

-50 0

50 100

. 150 200 Temperature, F 250 300 350 400 T35MtE:

18.5 F P_~

0~~~

0~~~~~~

00 0

100 200 Temperature, F 300 400 100 90 80 70 in

.n Ui i

r.

E 60 1 50 Tso:

46.3 F T3:

2.2 F CvUSE:

69.4 ft-lbs 0~~~~

........................ /

40 1 30 20 1 10 0

-100

-50 0

50 100 150 200 250 300 350 400 Temperature, F A

FRAMATOME ANP B-21

BAW-2454 Figure B-7. Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Heat-Affected-Zone Material

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -

100 a'

EF 75 50 25 O

A

-100

-50 0

50 100 150 200 250 300 350 400 Temperature, F

.R2 100 80 c

60 EL 40 w

e 20

'I O T~~~ 5 ~~~

0~

0 ~

0 0

I, 29.2F I

0 0

-100 0

100 200 Temperature, F 300 400 120 110 100 90 w

80 70 El 60 w

U 50 Ca E

40 30 20 10 0

TSO:

60.6 F T30:

-13.2 F 0

CvUSE:

85.3 ft-lbs S~~~~~

/~~~ *

.. 7

-100

-50 0

50 100 150 200 250 300 350 400 Temperature, F A

B-22 FRAMATOME ANP

BAW-2454 Figure B-S. San Onofre Unit 3 970 Capsule Surveillance Charpy Impact Data for Heat-Affected-Zone Material, Irradiated to 8 x 1018 n/cm2 (E>1.0 MeYV)

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -

U)

-100

-50 0

50 100 150 200 250 300 350 Temperature, F 400

.2 a,-

'U 0.a' w

a' 100 80 E 60 E 40 [

20

-100

[TssmLE:

70.F S

. A 0

100 200 Temperature, F 300 400 120 110 100 90

'0 80 70 41 60

'j 50*

to 0.

E40 T5: 118.3 F T30:

438 F CvUSE: 4.0 ftlbs 0

0 0

0 30 -

20 -.

10 01

-100

-50 0

50 100 150 200 250 300 350 400 Temperature, F A

FRAMATOME ANP B-23

BAW-2454 Figure B-9. Unirradiated Surveillance Charpy V-Notch Impact Data for San Onofre Unit 3, Standard Reference Material

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -

100 a;

754 50 U.

25U

-100

-50 0

50 100 150 200 250 300 350 400 Temperature, F 2 100 80 e

60 40 2

20 O

JT3smL:

31.3 F.3I

-100 0

100 200 Temperature, F 300 400 150 140 130 120 110 Da 100

2 w

90

='

80 u

70 e

60 E

50 40 30 20 10 0

0 S

-100

-50 0

50 100 150 200 250 300 350 400 Temperature, F A

B-24 FRAMATOME ANP

BAW-2454 APPENDIX C Charpy V-Notch Shift Comparison:

Original Hand Curve Fit vs. Hyperbolic Tangent Curve Fit C-1 FRAMATOME ANP

BAW-2454 Table C-1. Comparison of Curve Fit Transition Temperature Shifts for San Onofre Unit 3 Surveillance Material, Base Metal Plate C-6802-1, Transverse Orientation 30 ft-lb Transition Temperature Fluence Hand Curve Fit Hyperbolic Tangent Curve Fit Casue (x1019 n/cm2)I FAv.

FSIfF Capsule (E>1.0 MeV)

Avg., IF Shift, FAvg.,

F Shift, F Unirradiated 50 36 970 0.8 105 55 95 50 ft-lb Transition Temperature (xF19 n/cm 2)Hand Curve Fit Hyperbolic Tangent Curve Fit Capsule (E>1.0 MeV)

Avg., F Shift, FAvg.,

F Shift, OF Unirradiated 93 70 970 0.8 160 67 151 81 35 MLE Transition Temperature Fluence Hand Curve Fit Hyperbolic Tangent Curve Fit (xlOt 9 n/cm2)

Capsule

(>1.0 MeV)

Avg., 0F Shift, OF Avg., OF l

Shift, -F Unirradiated 85 51 970 0.8 133 48 124 J

73 A

FRAMATOME ANP C-2

BAW-2454 Table C-2. Comparison of Curve Fit Transition Temperature Shifts for San Onofre Unit 3 Surveillance Material,

'Base Metal PlatetC-6802-1, Longitudinal Orientation 30 ft-lb Transition Temperature Flue19 n/cen2)

Hand Curve Fit Hyperbolic Tangent Curve Fit (xlO'9 n/cm2)

Capsule (E>1.0 MeV)

Avg., 0F [

Shift, 0F Avg., F l Shift, 0F Unirradiated 80 I8 970 0.8 130 50 126 45 50 ft-lb Transition Temperature Capsu (x1019 n/cm2)

'Hand Curve Fit Hyperbolic Tangent Curve Fit Capsule (E>1.0 MeV)

Avg., 0FShift, F

Avg., F Shift, 0F Unirradiaied 125 J-125 970 0.8

'170 45 197 72 35 MLE Transition Temperature Fluence Hand Curve Fit Hyperbolic'Tangent Curve Fit Capsule (E>1.0 MeV)

Avg.,

7Sf1 Unirradiated 1

0 f

102

'l 6

970

'0.8 155 55 j

166 64 A

FRAMATOME ANP C-3

BAW-2454 Table C-3. Comparison of Curve Fit Transition Temperature Shifts for San Onofre Unit 3 Surveillance Material, Weld Metal C-6802-21C-6802-3 (Wire Heat 90069) 30 ft-lb Transition Temperature Fluence lHand Curve Fit Hyperbolic Tangent Curve Fit (xlO'0 n/cm2) 1 Capsule (E>1.0 MeV)

Avg., F J Shift, 0F Avg., 0F Shift, F l

Unirradiated

-27 l

-27 970 0.8 5

32 2

29 50 ft-lb Transition Temperature Fluence Hand Curve Fit Hyperbolic Tangent Curve Fit Capsule (xlO '9 n/cm2) j 0FAg,,

Capsule I (E>1.0 MeV)

Avg., 0F

Shift, 0

F Avg., F Shift, F Unirradiated 20 13 970 0.8 45 25 46 33 35 MLE Transition Temperature Fluence Hand Curve Fit Hyperbolic Tangent Curve Fit (x 1O' 9 n/c mn 2 )

Capsule (E>1.0 MeV)

Avg., OF Shift, OF Avg. OF l

Shift, F Unirradiated

-12

-12 970 J

0.8 25 37 19 31 A

FRAMATOME ANP C-4

BAW-2454 Table C-4. Comparison of Curve Fit Transition Temperature Shifts for San Onofre Unit 3 Surveillance Material, Heat-Affected-Zone Material 30 ft-lb Transition Temperature Fluence Hand Curve Fit Hyperbolic Tangent Curve Fit (xl IO "ntcm')

Capsule (E>1.0 MeV)

Avg., °F Shift, °F Avg., OF Shift, °F Unirradiated

-10

-13 970 0.8 35 45 44 57 50 ft-lb Transition Temperature Fluence Hand Curve Fit Hyperbolic Tangent Curve Fit (xlI 0 "n/cm2)

Capsule (E>1.0 MeV)

Avg., °F Shift, OF Avg., OF Shift, °F Unirradiated 50 61 l

970 0.8 120 J

70 J

118 57 35 MLE Transition Temperature (X1 ncM2)

Hand Curve Fit Hyperbolic Tangent Curve Fit Capsule (E>1.0 MeV)

Avg., °F J Shift, OF j

Avg., F Shift, °F Unirradiated 20 29 970 0.8 50 J

30 1

71 J

42 A

FRAMATOME ANP C-5

BAW-2454 APPENDIX D Fluence Analysis Methodology i

! I D-1 A

FRAMATOME ANP

BAW-2454 The primary tool used in the determination of the flux and fluence exposure to the surveillance capsule dosimeters is the two-dimensional discrete ordinates transport code DORT.[D 3]

The San Onofre Unit 3 capsule was located at 7.00 (off of the major axis) for cycles I through

11. The power distributions in the 11 irradiation cycles indicated that the distribution along the Z-axis was separable into a single channel. This means that within the core region the three-dimensional (R, 0, Z) neutron flux can be considered to be separable functions of R, 0 and Z.

The integral over the Z-axis provides the parameters for the RO DORT model. Beyond the core, the neutron flux is a multi-channel function along the R-axis.

However, the three-dimensional neutron flux can be considered to be separable functions of R, Z and 0.

The integral over the 0-axis provides the parameters for the RZ DORT model. Beyond the core, the material symmetry in both 0 and Z provide the means of determining the weighted parameters. Within the core region, an explicit three-dimensional model provides the means of performing the respective integrals. Therefore, the cycle 1-1 1 irradiations can be modeled using the standard FANP synthesis procedures. [D11 Figure D-1 depicts the analytical procedure that is used to determine the fluence accumulated over each irradiation period. As shown in the figure, the analysis is divided into seven tasks: (I) generation of the neutron source, (2) development of the DORT geometry models, (3) calculation of the macroscopic material cross sections, (4) synthesis of the results, and (5-7) estimation of the calculational bias, the calculational uncertainty, and the final fluence. Each of these tasks is discussed in greater detail in the following sections.

D.1.

Generation of the Neutron Source The time-averaged space and energy-dependent neutron sources for cycles 1-1 1 were calculated using the SORREL[D4 1 code. The effects of bumup on the spatial distribution of the neutron source were accounted for by calculating the cycle average fission spectrum for each fissile isotope on an assembly-by-assembly basis, and by determining the cycle-average specific neutron emission rate. This data was then used with the normalized time weighted average pin-by-pin relative power density (RPD) distribution to determine the space and energy-dependent neutron source. The azimuthally averaged, time averaged axial power shape in the peripheral assemblies was used with the fission spectrum of the peripheral assemblies to determine the neutron source for the axial DORT run. These two neutron source distributions were input to DORT as indicated in Figure D-1. Three separate sources (1-8a, 8b-lOa and Ob-1) were developed in order to account for two changes in Tin that occurred during cycles 8 and 10. A power uprate that occurred in cycle 11 was also accounted for during the synthesis procedure.

A D-2 FRAMATOME ANP

BAW-2454 D.2.

Development of the Geometrical Models The system geometry models for the mid-plane (R, 0) DORT were developed using standard FANP interval size and configuration guidelines. 'The RO model for the cycles 1-8a, 8b-lOa, and Ob-1 analysis extends radially from the center of the core to the outer surface of the pressure vessel, and azimuthally from the major axis to 45°. The axial model extended from 35 cm below the active core region to 35 cm above the active core region. The geometrical models either met or exceeded all guidance criteria concerning interval size that are provided in Reg Guide 1.190.[D-2] In all cases, cold dimensions were used. The geometry models were input to the DORT code as indicated in Figure D-l. These models will be used in all subsequent fluence analyses that may be performed by Framatome ANP for San Onofre Unit 3.

D.3.

Calculation of Macroscopic Material Cross Sections In accordance with Reg Guide 1.190, the BUGLE-931 5 cross section library was used. The GIP code D 6] was used to calculate the macroscopic energy-dependent cross sections for all materials used in the analysis - from the core out through the pressure vessel and from core plate to core plate. The ENDF/B-VI dosimeter reaction cross sections were used to generate the response functions that were used to calculate the DORT-calculated "saturated" specific activities.

D.4.

DORT Analyses The cross sections, geometry, and appropriate source were combined to create a set of DORT models (RO and RZ) for the cycles 1-8a, 8b-lOa and 1Ob-I 1 analyses. Each RO DORT run utilized a cross section Legendre expansion of three (P3), seventy directions (S o), with the appropriate boundary conditions. The RZ models used a cross section Legendre expansion of three (P3), seventy directions (Slo), with the appropriate boundary conditions. A theta-weighted flux extrapolation model was used, and all other requirements of Reg Guide 1.190 that relate to the various DORT parameters were either met or exceeded for all DORT runs.

D.5.

Synthesized Three Dimensional Results The DORT analyses produce two sets of two-dimensional flux distributions, one for a vertical cylinder and one for the radial plane for each set of dosimetry.. The vertical cylinder, which will be referred to as the RZ plane, is defined as the plane bounded 35 cm above and below the active core region and radially by the center of the core the outside surface of the reactor pressure vessel. The horizontal plane, referred to as the RO plane, is defined as the plane bounded radially by the center of the core and the outside surface of the pressure vessel, and azimuthally by the A

D-3 FRAMATOME ANP

BAW-2454 major axis and the adjacent 450 radius. The vessel flux, however, varies significantly in all three cylindrical-coordinate directions (R, 0, Z). This means that if a point of interest is outside the boundaries of both the R-Z DORT and the R-0 DORT, the true flux cannot be determined from either DORT run. Under the assumption that the three-dimensional flux is a separable function,D I both two-dimensional data sets were mathematically combined to estimate the flux at all three-dimensional points (R, 0, Z) of interest. The synthesis procedure outlined in Reg Guide 1.190 is identical to the basis used for the Framatome ANP flux-synthesis process.

D.6.

Calculated Activities and Measured Activities The calculated activities for each dosimeter type "d" for each irradiation period were determined using the following equation:

G Cd =Z(r)XRFg x BdxNSF (1) g.I where Cd calculated specific activity for dosimeter "d" in ~iCi of product isotope per gram of target isotope I g(rd.)

three dimensional flux for dosimeter "d" at position Yd for energy group "g" RF dosimeter response function for dosimeter "d" and energy group "g" Bd bias correction factors for dosimeter "d" NSF non-saturation correction factor (NSF).

For this analysis, three separate sets of activities will be calculated; therefore a combination of the calculated activities must be performed. The EOC 11 total calculated activity, C"1I, will be accomplished using the equation, for dosimeter "i",

(C EOCII

= (Cd(I-8a) ) + (Cd(8b-10a))I + (Cd(IOb-I )

(2)

Each calculated activity in equation 2, Cd(cycle), is calculated using equation 1, however each set of data, i.e. 1-8a, 8b-lOa and lOb-1I, will be calculated using a cycle specific NSF(cycle) factor.

A D-4 FRAMATOME ANP

BAW-2454 The bias correction factors (Bd) in the specific activity calculation above are listed in Table D-l.

A photofission factor was applied to correct for the fact that some of the '"Cs atoms present in the dosimeter were produced by (y, f) reactions and were not accounted for in DORT analysis.

Likewise, an impurity factor was included to account for U-235 content in the U-238 dosimetry.

The short half life was insignificant and therefore was not applied.

D.7.

C/M Ratios The following explanations will define the meanings of the terms "measurements" (M) and "calculations" (C) as used in this analysis: [D-11 Measurements: The meaning of the term "measurements" as used by Framatome ANP is the measurement of the physical quantity of the dosimeter (specific activity) that responded to the neutron fluence, not to the "measured fluence." For the example of an iron dosimeter, a reference to the measurements means the specific activity of 54Mn in giCi /g, which is the product isotope of the dosimeter reaction:

5 4Fe+n-e 5 4Mn+p+

Calculations: The calculational methodology produces two primary results - the calculated dosimeter activities and the neutron flux at all points of interest. The meaning of the term "calculations" as used by Framatome ANP is the calculated dosimeter activity. The calculated activities are determined in such a way that they are directly comparable to the measurement values, but without recourse to the measurements. That is, the calculated values are determined by the DORT calculation and are directly comparable to the measurement'values. ENDF/B-VI based dosimeter reaction cross sections[D 7.and response functions were used in determining the calculated values for each individual dosimeter. In summary, it should be stressed that the calculation values in the Framatome ANP approachlD4, are independent of the measurement values.

D.8.

Uncertainty The San Onofre Unit 3 cycles I through 11 fluence predictions are based on the methodology described in the Framatome ANP "Fluence and Uncertainty Methodologies" topical report, BAW-2241P-A. The time-averaged fluxes, and thereby the fluences throughout the reactor and vessel, are calculated with the DORT discrete ordinates computer code using three-dimensional A

D-5 FRAMATOME ANP

BAW-2454 synthesis methods. The basic theory for synthesis is described in Section 3.0 of the topical and the DORT three-dimensional synthesis results are the bases for the fluence predictions using the Framatome ANP "Semi-Analytical" (calculational) methodology.

The uncertainties in the San Onofre Unit 3 fluence values have been evaluated to ensure that the greater than 1.0 MeV calculated fluence values are accurate (with no discernible bias) and have a mean deviation that is consistent with the Framatome ANP benchmark database of uncertainties.

Consistency between the fluence uncertainties in the updated calculations for San Onofre Unit 3 cycles 1-1 1 and those in the Framatome ANP benchmark database ensures that the vessel fluence predictions are consistent with the 10 CFR 50.61, Pressurized Thermal Shock (PTS) screening criteria and the Regulatory Guide 1.9 9[D-81 embrittlement evaluations.

The verification of the fluence uncertainty for the San Onofre Unit 3 reactor includes:

estimating the uncertainties in the cycles 1 through 11 dosimetry measurements, estimating the uncertainties in the cycles 1 through 11 benchmark comparison of calculations to measurements, estimating the uncertainties in the cycles through 11 pressure vessel fluence, and determining if the specific measurement and benchmark uncertainties for cycles 1-11 are consistent with the Framatome ANP database of generic uncertainties in the measurements and calculations.

The embrittlement evaluations in Regulatory Guide 1.99 and those in 10 CFR 50.61 for the PTS screening criteria apply a margin term to the reference temperatures. The margin term includes the product of a confidence factor of 2.0 and the mean embrittlement standard deviation. The factor of 2.0 implies a very high level of confidence in the fluence uncertainty as well as the uncertainty in the other variables contributing to the embrittlement. The dosimeter measurements from the San Onofre Unit 3 analysis would not directly support this high level of confidence. However, the dosimeter measurement uncertainties are consistent with the Framatome ANP database. Therefore, the calculational uncertainties in the updated fluence predictions for San Onofre Unit 3 are supported by 728 additional dosimeter measurements and thirty-nine benchmark comparisons of calculations to measurements as shown in Appendix A of the topical. The calculational uncertainties are also supported by the fluence sensitivity evaluation of the uncertainties in the physical and operational parameters, which are included in the vessel fluence uncertainty.[D I] The dosimetry measurements and benchmarks, as well as the fluence sensitivity analyses in the topical are sufficient to support a 95 percent confidence level, with a confidence factor of +/-2.0, in the fluence results from the "Semi-Analytical" methodology.

A D-6 FRAMATOME ANP

BAW-2454 The Framatome ANP generic uncertainty in the dosimetry measurements has been determined to be unbiased and has an estimated standard deviation of 7.0 percent for the qualified set of dosimeters. The San Onofre Unit 3 cycles 1-11 dosimetry measurement uncertainties were evaluated to determine if any biases were evident and to estimate the standard deviation. The dosimetry measurements were found to be appropriately calibrated to standards traceable to the National Institute of Standards and Technology and are thereby unbiased by definition. The mean measurement uncertainties associated with cycles 1-1 1 are as follows:

0a~ 6.22%.

This value was determined from Equation 7.6 in the topicalED-I and indicates that there is consistency with the Framatome ANP database. Consequently, when the Framatome ANP database is updated, the San Onofre Unit 3 cycles 1-1 1 dosimetry measurement uncertainties may be combined with the other 728 dosimeters. Since the cycles 1-1 1 measurements are consistent with the Framatome ANP database, it is estimated that the San Onofre Unit 3 dosimeter measurement uncertainty may be represented by the Framatome ANP database standard deviation of 7.0 percent. Based on the Framatome ANP database, there appears to be a 95 percent level of confidence that 95 percent of the San Onofre Unit 3 dosimetry measurements, for fluence reactions above 1.0 MeV, are within +/-14.2 percent of the true values.

The Framatome ANP generic uncertainty for benchmark comparisons of dosimetry calculations relative to the measurements indicates that any benchmark bias in the greater than 1.0 MeV results is too small to be uniquely identified. The estimated standard deviation between the calculations and measurements is 9.9 percent. This implies that the root mean square deviation between the Framatome ANP calculations of the San Onofre Unit 3 dosimetry and the measurements should be approximately 9.9 percent in general and bounded by +20.04 percent for a 95 percent confidence interval with thirty-nine independent benchmarks.

The weighted mean values of the ratio of calculated dosimeter activities to measurements (C/M) for cycles 1-11 have been statistically evaluated using Equation 7.15 from the topical. The deviation in the benchmark comparisons is as follows:

Ac = 6.363%.

tA D-7 FRAMATOME ANP

BAW-2454 This deviation indicates that the benchmark comparisons are consistent with the Framatome ANP database. Consequently, when the Framatome ANP database is updated, the cycles1-11 benchmark uncertainties may be included with the other thirty-nine benchmark uncertainties in the topical. The consistency between the cycles 1-11 benchmark uncertainties and those in the Framatome ANP database indicates that the San Onofre Unit 3 fluence calculations for cycles 1-11 have no discernible bias in the greater than 1.0 MeV fluence values. In addition, the consistency indicates that the fluence values can be represented by the Framatome ANP reference set which includes a calculational standard deviation of 7.0 percent at dosimetry locations. See Table D-2.

A D-8 FRAMATOME ANP

BAW-2454

- Table D-1. Bias Correction Factors Dosimeter Type I

Bias Activation Short Half Life Fiso Photofission

. Impurities Table D-2. Calculational Fluence Uncertainties Uncertainty (%)

Standard Deviation j 95% / 95% Confidence Typc of Calculation (6) l

( +/-2)

Capsule

-7.0 14.2 Pressure Vessel 10.0 20.0 (maximum location) 10.0 20.0 Pressure Vessel (extrapolation)1.428 A

FRAMATOME ANP D-9

BAW-2454 Figure D-1. Fluence Analysis Methodology for San Onofre Unit 3 Surveillance Capsule A

FRAMATOME ANP D-10

BAW-2454 D.9.

References D-1.

Worsham, J.R., et al., "Fluence and Uncertainty Methodologies," BAW-2241 P-A.

Revision 1 Framatome ANP, Lynchburg, Virginia, April 1999.

D-2.

U.S. Nuclear Regulatory Commission Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.

D-3.

Rutherford, M. A., N. M. Hassan, et. al., Eds., "DORT, Two Dimensional Discrete Ordinates Transport Cod,." BWNT-TM-107, Framatome ANP, Lynchburg, Virginia, May 1995.*

D-4.

Hassler, L. A., and N. M. Hassan, "SORREL, DOT Input Generation Code User's Manual," NPGD-TM-427. Revision 10, Framatome ANP, Lynchburg, Virginia, May 2001.

D-5.

Ingersoll, D. T., et. al., "B UGLE-93, Production and Testing of the ITA4MIA-B6 Fine Group and the BUGLE-93 Broad Grozp Neutron/photon Cross-Section Libraries Derivedfrom ENDF/B-VI Nuclear Data," ORNL-DLC-175 Radiation Safety Information Computational Center, Oak Ridge National Laboratory, Oak Ridge, Tennessee, April 1994.

D-6.

Hassler, L. A. and N. M. Hassan, "GIP User 's Manualfor B&WY Version, Group Organized Cross Section Input Program," NPGD-TM-456. Revision 11, Framatome ANP, Lynchburg, Virginia, August 1994.

D-7.

Worsham, J. R., "BUGLE-93 Response Functions," 32-1232719-00, Revision 0, Framatome ANP, Lynchburg, Virginia, June 1995.

D-8.

U.S. Nuclear Regulatory Commission, "Radiation Embrittlement of Reactor Vessel Materials," Regulatorv Guide 1.99. Revision 2, May 1998.

A

BAW-2454 APPENDIX E Reactor Vessel Surveillance Program Background Data and Information I:

A FRAMATOME ANP E-1

BAW-2454 E.1.

San Onofre Unit 3 Reactor Pressure Vessel The San Onofre Unit 3 reactor vessel was fabricated by Combustion Engineering, Inc. (CE). The San Onofre Unit 3 reactor vessel beltline region consists of two shells (intermediate and lower),

containing three base metal plates each with associated longitudinal welds and two circumferential weld seams. Table E-1 presents a description of the San Onofre Unit 3 reactor vessel beltline materials including their copper and nickel chemical contents and their unirradiated RTNDT values. 1211 The locations of the materials within the reactor vessel beltline region are shown in Figure E-1. 2"I The heat treatment for the plate materials consisted of austenitization at 1575 +/- 50'F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; water quenched and tempered at 1225 3 250F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. For ASME Code qualification, the plates were stress relieved at 1150 +/- 250F for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> and then were furnace cooled to 600'F at a rate of 1000F/hour. The actual time at temperature for a specific weld or plate in the vessel depended upon the sequence of vessel fabrication; intermediate and final stress relief times were selected such that the total did not exceed 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> for any particular portion of the vessel.

Longitudinal weld seams would see stress relief times near the 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> maximum, while the closing girth weld in the beltline region would see not more than approximately half this amount.

All of the testing of plate materials was performed on pieces with essentially an identical heat treatment as the actual reactor vessel. The surveillance weldment received a final 41-hour and 45-minute stress relief at IOOF to I 1500F.121]

E.2.

Surveillance Material Selection Data The materials that were selected for the surveillance program were in accordance with an ASTM E 185 Standard prior to 1974. The San Onofre Unit 3 RVSP capsules include the predicted limiting reactor vessel beltline plate C-6802-1, heat no. C9195-2 at the time of the development of the surveillance program.111 The surveillance weld used in the San Onofre Unit 3 RVSP was fabricated using the wire heat 90069 which is the same wire heat used for the lower to intermediate shell girth weld in the reactor vessel. 211 A

E-2 FRAMATOME ANP

BAW-2454 Table E-1. Description of the San Onofre Unit 3 Reactor Vessel Beltline Region Materials Chemical Chemical Initial Toughness Properties Beltline Material Material Material Composition Region Location Identification Type Heat No.

Cu, Ni,

NDT, RTNDT,
USE, wt%

wt%

F F

ft-lbs Inter. Shell Long. Weld 2-203A Subarc Weld 83650 0.04 0.17

-40

-40 136; Inter. Shell Long. Weld 2-203B Subarc Weld 83650 0.05 0.21

-40

-40 136 Inter. Shell Long. Weld 2-203C Subarc Weld 83650 0.04 0.08

-40

-40 136 Lower Shell Long. Weld 3-203A Subarc Weld 88114 0.04 0.21

-70

-70 161 Lower Shell Long. Weld 3-203B Subarc Weld 88114

0.04 0.19

-70

-70 161 Lower Shell Long. Weld 3-203C.

Subarc Weld 88114

'0.04 0.21

-70 161 Upper/Inter. Shell Girth 8-203 Subarc Weld 88118 0.05 0.17

-70

-70 125 Weld Lower/Inter. Shell Girth 9-203

-Subarc Weld -

90069-0.06 0.04

-60:

-60 123 Weld 90144 0.05 0.04

-50

-50 91-Intermediate Shell C-6802-1 A 533B-1 C9195-2 0.06 0.58'

-20

+40 95 Intermediate Shell C-6802-2 A 533B-1 C9218-2 0.04 0.57

-20

+10 115 Intermediate Shell C-6802-3 A 533B-1 C9195-1 0.06 0.58

-10

+20 105 Lower Shell C-6802-4 A 533B-1 C9220-1 0.05 0.56

-30

+10 118' Lower Shell C-6802-5 A 533B-1 t9218-1 0.04 0.55 0

+10 116 Lower Shell C-6802-6 A 533B-1 B3388-1 0.06 0.62

-40

+20 92.

A FRAMATOME ANP E-3

BAW-2454 Figure E-1. Location and Identification of Materials used in the Fabrication of the San Onofre Unit 3 Reactor Pressure Vessel

.1a Welsttr Va et Sone Weld em rio sihew r

A FRAMATOME ANP E4