W3F1-2004-0075, Reissue of Report BAW-2177, Analysis of Capsule W-97 Entergy Operations, Inc. Waterford Generating Station Unit 3 - Reactor Vessel Material Surveillance Program

Reissue of Report BAW-2177, Analysis of Capsule W-97 Entergy Operations, Inc. Waterford Generating Station Unit 3 - Reactor Vessel Material Surveillance Program

ML042710435
Person / Time
Site:	Waterford
Issue date:	09/13/2004
From:	Dodds R Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2004-0075 BAW-2177
Download: ML042710435 (131)

Text

Entergy Operations, Inc.

-Entergy 17265 River Road Killona, LA 70066 Tel 504 739 6650 W3F1-2004-0075 September 13, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Reissue of Report BAW-2177, "Analysis of Capsule W-97 Entergy Operations, Inc. Waterford Generating Station Unit 3 - Reactor Vessel Material Surveillance Program" Docket No. 50-382 License No. NPF-38 Waterford 3

Reference:

1. Letter Number W3F1 92-0369, dated November 25, 1992, Reactor Vessel Material Surveillance Program Requirements - Report of Test Results

2. Attachment to Letter Number W3F1 92-0369, BAW-2177, "Analysis of Capsule W-97 Entergy Operations, Inc. Waterford Generating Station Unit 3 - Reactor Vessel Material Surveillance Program,"

dated November 1992.

Dear Sir or Madam:

The purpose of this letter is to provide a revised copy of report BAW-2177, "Analysis of Capsule W-97 Entergy Operations, Inc. Waterford Generating Station Unit 3 - Reactor Vessel Material Surveillance Program," which contained an editorial error. This report was originally submitted as an attachment to Letter Number W3F1 92-0369, dated November 25, 1992, Reactor Vessel Material Surveillance Program Requirements - Report of Test Results.

This error was communicated to the Waterford 3 Senior Resident Inspector and NRC Project Manager. Following these discussions, Waterford 3 committed to submit a revised report which includes a correction of the editorial error.

This error was discovered during a recent review of report BAW-2177 by Westinghouse while assembling Charpy data for development of the new Charpy-irradiation correlation for the ASTM E900 standard. In review of the report by Areva, the vendor who provided the report, it was noted that the values in Table 5.6, uCharpy Impact Results for Capsule W-97 Weld Metal, 88114/0145, 6.47 x 1018 n/cm 2" were incorrect in Revision 0 of the report.

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W3Fl-2004-0075 Page 2 Figure 5-8, 'Charpy Impact Data for Irradiated Weld Metal, 88114/0145" had been plotted in Revision 0 using the correct data even though Table 5.6 had incorrect data. A "cut and paste" mistake had occurred from a previous draft.

Section 7.0 "Discussion of Capsule Results" and Table 7.3, "Observed vs. Predicted Changes for Capsule W-97 Irradiated Charpy Impact Properties - 6.47 x 1018 n/cm2 (E> 1 MEV)" used the correct data from Figure 5-8 for the comparison of observed vs. predicted property changes. Since the actual numbers used in the comparison of the transition temperature and upper shelf energy changes are correct, the conclusions that the calculated property changes are conservative relative to the observed properties therefore remain unchanged.

In summary, the error in Table 5.6 of report BAW-2177 has no impact on the conclusions contained in the report. The calculated reactor vessel material properties remain conservative in relation to the observed, via specimen testing, material properties.

There are no new commitments contained in this submittal.

Should you have questions regarding this report please contact Mrs. Stacie Fontenot at 504-739-6656.

Sincerely, R.A. Dodds Manager, Licensing RAD/STF/ssf

Attachment:

Report BAW-2177-01, dated February 2004, "Analysis of Capsule W-97 Entergy Operations, Inc. Waterford Generating Station Unit 3 - Reactor Vessel Material Surveillance Program"

W3F11-2004-0075 Page 3 cc: Mr. Bruce S. Mallett Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Waterford Steam Electric Station Unit 3 P.O. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Attn: Mr. N. Kalyanam Mail Stop O-07D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway ATTN: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn ATTN: N.S. Reynolds 1400 L Street, NW Washington, DC 20005-3502

e Attachment W3FI-2004-0075 Report BAW-2177-01, dated February 2004, "Analysis of Capsule W-97 Entergy Operations, Inc. Waterford Generating Station Unit 3 - Reactor Vessel Material Surveillance Program"

The B&W BAW-2177-01 Owners Grou February 2004 ANALYSIS OF CAPSULE W-97 ENTERGY OPERATIONS, INC.

WATERFORD GENERATING STATION, UNIT NO. 3

-- Reactor Vessel Material Surveillance Program --

At ARE VA

BANV-2177-01 February 2004 ANALYSIS OF CAPSULE W-97 ENTERGY OPERATIONS, INC.

WATERFORD GENERATING STATION, UNIT NO. 3

-- Reactor Vessel Material Surveillance Program --

by A. L. Lowe, Jr., PE R. E. Napolitano D. M. Spaar W. R. Stagg FRAMATOME ANP Document No. 77-2177-01 (See Section 10 for document signatures)

Framatome ANP, Inc.

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

FRAMATOME ANP

BAW-2177-01 RECORD OF REVISIONS Revision Description I Date 00 Original Release 11/92 01 Page 5-6 was replaced correcting the data in Table 5-6. All 2/04 subsequent reporting of the weld metal Charpy results were correct.

The conclusions are not affected. Page 10-2 was added containing the Revision 1 signatures. Both cover pages were replaced reflecting revision change.

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SUMMARY

This report describes the results of the examination of the first capsule (Capsule W-97) of the Entergy Operations, Inc., Waterford Generating Station, Unit No. 3 reactor vessel surveillance program. The objective of the program is to monitor the effects of neutron irradiation on the tensile and fracture toughness properties of the reactor vessel materials by the testing and evaluation of tension and Charpy impact specimens. The program was designed in accordance with the requirements of ASTM Specification E185-73.

The capsule received an average fast fluence of 6.47 x 1018 n/cm 2 (E > 1.0 MeV) and the predicted fast fluence for the reactor vessel T/4 location at the end of the fourth cycle is 2.74 x 1018 n/cm2 (E > 1 MeV). Based on the calculated fast flux at the vessel wall, an 80% load factor, and the planned fuel management, the projected fast fluence that the Waterford Generating Station, Unit No. 3 reactor pressure vessel inside surface will receive in 40 calendar years of operation is 3.69 x 1019 n/cm2 (E > 1 MeV) and the corresponding T/4 fluence is calculated to be 1.97 x 1019 n/cm2 (E > I MeV).

The results of the tension tests indicated that the materials exhibited normal behavior relative to neutron fluence exposure. The Charpy impact data results exhibited the characteristic shift to higher temperature for the 30 ft-lb transition temperature and a decrease in upper-shelf energy. These results demonstrated that the current techniques used for predicting the change in both the increase in the RTNDT and the decrease in upper-shelf properties due to irradiation are conservative.

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CONTENTS Page

1. INTRODUCTION . . .. . . . . . . . .. . . . . . . . . . . . . . . 1-1 .

2. BACKGROUND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1

3. SURVEILLANCE PROGRAM DESCRIPTION ................. . 3-1

4. PRE-IRRADIATION TESTS ................... 4-1 4.1. Tension Tests . . .. . . . . . . . .. . . . . . 4-1 4.2. Impact Tests ....... .. .. .... . 4-2

5. POST-IRRADIATION TESTING . . . . . . . . . . . . . . .... . . . 5-1 5.1. Visual Examination 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

6. NEUTRON FLUENCE .'....... ..... . . . . . . .6-1 6.1. Introduction . . ... . . . . . . . . . . . . 6-1 6.2. Vessel Fluence . . ... . . . . . . . . . . . . 6-4 6.3. Capsule Fluence . . . . . . . .. I. . . . . . . .6-5 6.4. Fluence Uncertainties . . . . .. . . . . . . . .6-6

7. DISCUSSION OF CAPSULE RESULTS-'. . ... . . . . .7-1 7.1. Pre-Irradiation Property Data . . . . ... . . 7-1 7.2. Irradiated Property Data . 7-1 7.2.1. Tensile Properties .,.. . . . . 7-1 7.2.2. Impact Properties . . . .. .

.7-2 7.3. Reactor Vessel Fracture Toughness . . 7-4 7.4. Operating Limitations . . .7-5 7.5. Pressurized Thermal Shock (PTS)'Evaluation . 7-5 7.6. Neutron Fluence Analysis . . . .7-5

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SUMMARY

OF RESULTS .. . . . . .. . . . . . ..... . . . . . . . . . 8-1

9. SURVEILLANCE CAPSULE REMOVAL SCHEDULE . . ... . . . ....... 9-1

10. CERTIFICATION . . .. . . . . ... . . . . . . . . . . 10-1

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Contents (Cont'd)

APPENDIXES Page A. Reactor Vessel Surveillance Program Background Data and Information . A-1 B. Pre-Irradiation Tensile Data .. B-i C. Pre-Irradiation Charpy Impact Data . . . . . . . . . . . . . . . ... C-1 L D. Fluence Analysis Methodology . . . . . . . . . . . . . . . . . . . . D-1 E. Capsule Dosimetry Data .E-1 F. Tension Test Stress-Strain Curves . . . . . . . . . . . . . . . . . . F-1 G. References .G-1 List of Tables Table 3-1. Specimens in Surveillance Capsule W-97 .3-2 3-2. Chemical Composition and Heat Treatment of Surveillance Materials . 3-3 5-1. Conditions of Thermal Monitors in Capsule W-97 . . . . . . . . . . 5-3 5-2. Tensile Properties of Caaesule W-97 Base Metal and Weld Metal Irradiated to 6.47 x 10' n/cm2 (E > 1 MeV) . . . . . . . . ... . . 5-4 5-3. Charpy Impact Results for Capsule W-97 Base Metal Lon2 itudinal (LT) Orientation, Heat No. M-1004-2, 6.47 x lo in/cm .. 5-5 5-4. Charpy Impact Results-for Capsule W-97 Base Metal Transverse (TL) Orientation, Heat No. M-1004-2, 6.47 x 1018 n/cm2 . . 5-5 l 5-5. Charpy Impact Results for Capsule W-97 Base Metal Heat-Affected Zone Material, Heat No. M-1004-2, 6.47 x 1O18 n/cm2 . . . . . . . . 5-6 5-6. Charpy Imact Results for Capsule W-97 Weld Metal, 88114/0145, 6.47 x 10 8 n/cm2 . . . . . . . . . . . . . . . . . . . . . . . . . 5-6 6-1. Surveillance Capsule Dosimeters .6-6 6-2. Waterford Unit 3 Reactor Vessel Fast Flux .6-7 6-3. Calculated Waterford Unit 3 Reactor Vessel Fluence. 6-8 6-4. Calculated Waterford Unit 3 Reactor Vessel DPA . . . . . . . . . . 6-9 6-5. Fluence, Flux, and DPA for 970 Surveillance Capsule . . .-. . . . . 6-9 6-6. Surveillance Capsule Measurements .6-10 1 6-7. Axial Power Data Affecting Flux . . . . . . . . . . . . . . . . . . 6-11 7-1. Comparison of Waterford Unit 3, Capsule W-97 Tension Test Results ..... . 7-6 7-2. Summary of Waterford Unit 3 Reactor Vessel Surveillance Capsules; Tensile Test Results . . . . . . . . . . . . 7-7 7-3. Observed Vs. Predicted Changes for Capsule W-97 Irradiated Charpy Impact Properties --6.47 x 1018 n/cm2 (E > 1 MeV) . . . . . . . . . 7-8 7-4. Evaluation of Reactor Vessel End-of-Life (32 EFPY) Fracture Toughness - Waterford Unit 3 ........ . .... . 7-9 7-5. Evaluation of Reactor Vessel End-of-Life (32 EFPY) Upper-Shelf J Energy -Waterford Unit 3 .7-10 7-6. Evaluation of Reactor Vessel End-of-Life Pressurized Thermal Shock Criterion - Waterford Unit 3 .7-11

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Tables (Cont'd)

Table Page A-I. Unirradiated Impact Properties and Residual Element Content Data of Beltline Region Materials Used for Selection of Surveillance Program Materials - Waterford Unit No. 3. . . . . A-3 A-2. Type and Quantity of'Specimens Contained in Each Irradiation Capsule Assembly . . ... . . . . . . . . A-4 B-i. Tensile Properties of Unirradiated Shell Plate Material, Heat No. M-1004-2, Longitudinal . . . . . . . . .8.... . . . . . . B-2 B-2.' Tensile Properties of Unirradiated Shell Plate, Material, Heat No. M-1004-2,-Transverse . . .. . ... . . . . . . B-2 B-3. Tensile Properties of Unirradiated Shell Plate, HAZ Material, Heat No. M-1004-2, Transverse. . .... . . . . . . B-3 B-4. Tensile Properties of Unirradiated Weld Metal 88114/0145. . B-3 C-1. Charpy Impact Data from Unirradiated Base Material, Longitudinal Orientation, Heat No. M-1004-2 . . . . . . . . .. . . . . . . . C-2 C-2. Charpy Impact Data from Unirradiated Base Material, Transverse Orientation, Heat No. M-1004-2 ... . . . . . . . . . . . . . . . C-2 C-3. Charpy Impact Data from Unirradiated Base Metal, Heat-Affected,-

Zone,. Heat No. M-1004-2 . . . . . . . C-3 C-4. Charpy Impact Data from Unirradiated Weld Metal, 88114/0145 . .. . C-3 D-1. Flux Normalization Factor for 97° Capsule . . . . . . . . . . . D-8 E-1. Detector Composition and Shielding .. . . . . . . . . ... . . . . E-2 E-2. Measured Specific Activities (Unadjusted) for Dosimeters in 97° Capsule.. . . . . . . . . ... . E-3 E-3. Dosimeter Activation Cross Sections, b/atom .E-4 List of FiQures Figure 3-1. Reactor Vessel Cross:Section Showing Location of Capsule W-97 in Waterford Unit 3 3-4 3-2. Typical Surveillance-Capsule Assembly Showing Location of Specimens and Monitors . .:- ... . . . . . .;. 3-5 3-3. Typical Surveillance Capsule Tensile - Monitor Compartment Assembly (Three per Capsule) ..... . . . . . . . . . . 3-6 3-4. Typical Surveillance Capsule Charpy Impact Compartment Assembly (Four. per Capsule) .. . ... 3-7 5-1. Photographs of Thermal Monitor Melt-Wire Capsules as Removed' From Surveillance Capsule ... 5-7 5-2. Photographs'of Tested Tension Test Specimens and Corresponding Fractured.Surfaces - Base Metal,.Transverse.Orientation. 5-8 5-3. Photographs of Tested Tension Test Specimens and Corresponding Fractured.Surfaces - Base Metal-Heat-Affected Zone . . . . . 5-9 IMEBCWV NUCCLEAR

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Figures (Cont'd)

Figure Page 1 5-4. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Weld Metal 88114/0145 . . . . . . . . . . . . 5-10 5-5. Charpy Impact Data for Irradiated Base Metal, Longitudinal Orientation, Heat No. M-1004-2 . . . . . . . . . . . . . . . . . . 5-11 5-6. Charpy Impact Data for Irradiated Base Metal, Transverse Orientation, Heat No. M-1004-2 .................

5-7. Charpy Impact Data for Irradiated Base Metal, Heat-Affected

. 5-12 I Zone, Heat No. M-1004-2. .. 5-13 5-8. Charpy Impact Data for Irradiated Weld Metal, 88114/0145. . . . . 5-14 1 5-9. Photographs of Charpy Impact Specimen Fracture Surfaces -

Base Metal, Longitudinal . . . . . 5-15 5-10. Photographs of Charpy Impact Specimen Fracture Surfaces -

Base Metal, Transverse ................. 5-16 1 5-11. Photographs of Charpy Impact Specimen Fracture Surfaces -

Base Metal, Heat-Affected Zone . . . . . ... . . . . . . . . . . . 5-17 5-12. Photographs of Charpy Impact Specimen Fracture Surfaces -

Weld Metal 88114/0145 . . . . . ... . . . . . . . . . . . . . . . . 5-18 6-1. General Fluence Determination Methodology . . . . . . . . . . . . . 6-2 6-2. Fast Flux, Fluence and DPA Distribution Through Reactor Vessel Wall. . . . . . . . ... . . 6-12 6-3. Azimuthal Flux and Fluence Distributions at Reactor Vessel Inside Surface ..... .... 6-13 6-4. Relative Axial Variation of E > 1 MeV Flux/Fluence . . . . . . . . 6-14 6-5. Radial Dimensions Used in Modeling Capsule and Pressure Vessel Regions . . . . . . . . . . . . . . . . . . . . . . . . . . 6-15 7-1. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Plate Material Longitudinal Orientation, Heat No. M-1004-2 . . 7-12 7-2. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Plate Material Transverse Orientation, Heat No. M-1004-2 . . . 7-13 j 7-3. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Base Metal, Heat-Affected-Zone, Heat No. M-1004-2 . . . . . . . 7-14 7-4. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Weld Metal 88114/0145 ................... 7-15 A-i. Location and Identification of Materials Used in the Fabrication of Waterford Unit 3 Reactor Pressure Vessel . . . . . . . . . . . . A-5 A-2. Location of Beltline Region Materials in Relationship to the Reactor Vessel Core ... . . . A-6 A-3. Location of Longitudinal Welds in Waterford Unit 3 Upper and Lower Shell Courses .. . A-7 I A-4. Location of Surveillance Capsule Irradiation Sites in l Waterford Unit 3. . ............. A-8 C-1. Charpy Impact Data from Unirradiated Base Metal (Plate), .8 Longitudinal Orientation, Heat No. M-1004-2 . . . . . . . . . . . . C-4 C-2. Charpy Impact Data from Unirradiated Base Metal (Plate), -

Transverse Orientation, Heat No. M-1004-2 . . . . . . . . . . . . . C-5 C-3. Charpy Impact Data from Unirradiated Heat-Affected-Zone Base Metal, Heat No. M-1004-2 .............. .... . C-6 NUCLEAR WSERVC COMPAN YCI

Figures (Cont'd)

Figure Page C-4. Charpy Impact Data for Unirradiated Weld Metal, 88114/0145 . . . . C-7 C-5. Charpy Impact Data for Unirradiated Correlation Monitor Material . C-8 D-1. Rationale for the Calculation of Dosimeter Activities and Neutron Flux in the Capsule .. ... D-9 D-2. Rationale for the Calculation of Neutron Flux in the Reactor Vessel ..... . . . . . . . . . . . . . . . . . . . . . D-10 D-3. Plan View Through Reactor Core Midplane (Reference R-e Calculation Model) . . . . . . . . . . . . D-11 F-I. Tension Test Stress-Strain Curve for Base Metal Plate Heat M-1004-2, Specimen No. 2L6, Tested at 7OF . . . . . . . . . . . . F-2 F-2. Tension Test Stress-Strain Curve for Base Metal Plate Heat M-1004-2, Specimen No. 2K5, Tested at 250F . . . . . . . . . . . . F-2 F-3. Tension Test Stress-Strain Curve for Base Metal Plate Heat M-1004-2, Specimen No. 2K2, Tested at 550F .F-3 F-4. Tension Test Stress-Strain Curve for Base Metal Heat-Affected Zone, Heat M-1004-2, Specimen No. 4K3, Tested at 70F . . . . . . . F-3 F-5. Tension Test Stress-Strain Curve for Base Metal Heat-Affected Zone, Heat M-1004-2, Specimen No. 4KK, Tested at 250F . . . . . . . F-4 F-6. Tension Test Stress-Strain Curve for Base Metal Heat-Affected Zone, Heat M-1004-2, Specimen No. 434, Tested at 550F . . . . . . . F-4 F-7. Tension Test Stress-Strain Curve for Weld Metal 88114/0145, Specimen No. 3JM, Tested at 70F .... . . . . . . . . . . . . . F-5 F-8. Tension Test Stress-Strain Curve for Weld Metal 88114/0145, Specimen No. 3KK, Tested at 250F .... . . . . . . . . . . . . . F-5 F-9. Tension Test Stress-Strain Curve for Weld Metal 88114/0145, Specimen No. 3KY, Tested at 550F .... . . . . . . . . . . . . . F-6

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1. INTRODUCTION This report describes the results of the examination of the first capsule (Capsule W-97) of the Entergy Operations, Inc., Waterford Generating Station, Unit No. 3 (Waterford Unit 3) reactor vessel material surveillance program (RVSP). The capsule was removed and evaluated after being irradiated in the Waterford Unit 3 reactor as part of the reactor vessel materials surveillance program (Combustion Engineering (C-E) Report C-NLM-0031). The capsule experienced a fluence of 6.47 x io18 n/cm2 (E > 1 MeV), which is the equivalent of approximately six effective full power years' (EFPY) operation of the Waterford Unit 3 reactor vessel inside surface.

The objective of the program is to monitor the effects of neutron irradiation on the tensile and impact properties of reactor pressure vessel materials under actual operating conditions. The surveillance program for Waterford Generating Station Unit No. 3 was designed and furnished by Combustion Engineering, Incorporated (C-E) as described inTR-C-MCS-0012 and conducted in accordance with IOCFR5O, Appendix H3. 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.

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2. BACKGROUND The ability of the reactor pressure vessel to resist.fracture is the primary factor in ensuring the safety of the primary system in light water-cooled reactors. The beltline region of the reactor vessel is the most critical region of the vessel because it is exposed to neutron irradiation. The general effects of fast neutron irradiation on the mechanical properties of low-alloy ferritic steels such as SA533, Grade B, used in the fabrication of the'Waterford Unit 3 reactor vessel, are well characterized and documented in the literature. The low-alloy ferritic steels used in the beltline region of reactor vessels 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 pressure vessel steels is the increase in temperature for the transition from brittle to ductile fracture accompanied by a reduction in the'Charpy upper-shelf energy value.

Appendix G to 10CFR50, "Fracture Toughness Requirements," 4 specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary .(RCPB) of water-cooled power reactors, and provides specific guidelines for determining the pressure-temperature limitations 'for operation of'the RCPB. The toughness and operational requirements are specified to provide adequate safety margins during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Although' the requirements of Appendix G to IOCFR50 became effective on August '16, 1973, "the requirements are 'applicable to all boiling and 'pressurized water-cooled nuclear power reactors, in6cli'di'ng those under'construction or in operation on the effective date.

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Appendix H to IOCFR50, "Reactor Vessel Materials Surveillance Program Requirements," 3 defines the material surveillance program required to monitor changes inthe fracture toughness properties of ferritic materials in the reactor vessel beltline region of water-cooled reactors resulting from exposure to neutron irradiation and the thermal environment. Fracture toughness test data are obtained from material specimens withdrawn periodically from the reactor vessel. These data will permit determination of the conditions under which the ij vessel can be operated with adequate safety margins against fracture throughout its service life. l A method for guarding against brittle fracture in reactor pressure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel (B&PV) Code, Section 1I1, "Nuclear Power Plant Components."5 This method utilizes fracture mechanics concepts and the reference nil-ductility temperature, RTNDT, which is defined as the greater of the drop weight nil-ductility transition temperature (per ASTM E-2088) or the temperature that is 60F 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 inAppendix G of ASME B&PV Code Section III. The KIR curve J is a lower bound of dynamic, static, and crack arrest fracture toughness results 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 this material as a function of temperature. Allowable operating limits can then ]

be determined using these allowable stress intensity factors.

The RTNDT and, in turn, the operating limits of a nuclear power plant, can be J adjusted to account for the effects of radiation on the properties of the reactor vessel materials. The radiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be monitored by a .

surveillance program in which a surveillance capsule containing prepared specimens of the reactor vessel materials is periodically removed from the I operating nuclear reactor and the specimens are tested. The increase in the Charpy V-notch 30 ft-lb temperature is added to the original RTNDT to adjust it for radiation embrittlement. This adjusted RTNDT is used to index the material to the KIR curve which, in turn, is used to set operating limits for the nuclear 2-2 15 lBCWNUCLEAR UJWSERVICE COMPANY

power plant. These new limits take into account the effects of irradiation on the reactor vessel materials.

Appendix G, 10CFR50, also requires a minimum Charpy V-notch upper-shelf energy 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 for deterioration as the result of neutron radiation. No action is required for a material that does not meet the 75 ft-lb requirement provided the irradiation deterioration does not cause the upper-shelf energy to drop below 50 ft-lbs. The regulations specify that ifthe upper-shelf energy drops below 50 ft-lbs, itmust be demonstrated in a manner approved by the Office of Nuclear Regulation that the lower values will provide adequate margins of safety.

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3. SURVEILLANCE PROGRAM DESCRIPTION The surveillance program for Waterford Unit 3 comprises six surveillance 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 near the inside wall of the reactor vessel at the locations shown in Figure 3-1.

The six capsules, designed to be placed in holders attached to the reactor vessel wall are positioned near the peak axial and azimuthal neutron flux. During the four cycles of operation, Capsule W-97 was irradiated in the 970 position adjacent to the reactor vessel wall as shown in Figure 3-1.

Capsule W-97 was removed during the fourth refueling'shutdown of Waterford Unit

3. The capsule contained Charpy V-notch impact test specimens fabricated from the one base metal (SA533, Grade B1) both longitudinal and transverse orienta-tion, one heat-affected-zone, and a weld metal. Tension test specimens were fabricated from the base metal, heat-affected-zone, and weld metal. The number of specimens of each material contained in the capsule are described in Table 3-1, and the location of the individual specimens within the capsule are described in Figures 3-2 through 3-4. The chemical composition and heat treatment of the surveillance material in Capsule W-97 are described in Table 3-2.

All plate and heat-affected-zone specimens were machined from the 1/4-thickness (1/4T) location of the plate material. Weld metal specimens were machined throughout the thickness of the weldment. Charpy V-notch and tension test specimens were cut from the surveillance material such that they were oriented with their longitudinal axes either parallel or perpendicular to the principal working direction.

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The neutron dosimeters contained in Capsule W-97 are as follows:

L Material Shielding Reaction Threshold Energy (Mev) Half-Life A Uranium None/Cd U238 (n,f) Sr90 0.7 28.0 years Sulfur None S (n,p) p 2.9 14.3 days Iron None Fe54 (n,p) Mn5 4 4.0 312.5 days Nickel Cd Ni58 (n,p) Co58 5.0 70.9 days Copper Cd Cu83 (n,a) Co80 7.0 5.27 years Titanium None Ti48 (n,p) Sc48 8.0 83.8 days Cobalt None/Cd Co5 9 (n,-y) Co80 Thermal 5.27 years Four thermal monitors of low-melting alloys were included in the W-97 capsule.

The eutectic alloys and their melting points are as follows:

Melting Alloy Composition, wt% Point, F 80.0 Au, 20.0 Sn 536 90.0 Pb, 5.0 Sn, 5.0 Ag 558 97.5 Pb, 2.5 Ag 580 97.5 Pb, 0.75 Sn, 1.75 Ag 590 Table 3-1. Specimens in Surveillance Capsule W-97 J Number of Test Specimens -I Material Description Tension CVN Impact Base Metal (M-1004-2)

Longitudinal 12 Transverse 3 12 Heat-Affected Zone 3 12 Weld Metal (88114/0145)

Total Per Capsule 3

9 12 48 IJ I

3-2 J WCW NUCLEAR IWSERVI/CE COMPA N Y

Table 3-2. Chemical Composition and Heat Treatment of Surveillance Materials Chemical Composition, w/o Heat No.(.) Weld Metal Element M-1004-2 88114/0145 b C 0.23 0.23 Mn 1.38 1.35 P 0.005 0.008 S 0.005 0.006 Si 0.23 0.16 Ni 0.58 0.22 Cr 0.01 0.05 Mo 0.57 0.57 Cu 0.03 0.04 Heat Treatment Heat No. Temp. F Time, h Cool ina Plate 1575+50 4 Water Quenched (M-1004-2) 1220+25 4 Furnace Cooled 1150+/-25 40 Furnace Cooled to 600F Weld Metal 1100-1175 40 1/2 Furnace Cooled to 600F (88114/0145)

(a)Chemical analysis by Combustion Engineering of surveillance program test plate.7 (b)Chemical analysis by Combustion Engineering of surveillance program test weld metal.'

3-3 15111 8 W N CLE AR IMWSE RVICE COMPANY

- - a-U L

Figure 3-1. Reactor Vessel Cross Section Showing Location L

of Caosule W-97 in Waterford Unit 3 L

1800 r .1 wOutlet Nozzle I1 <,,

L I

.1

.1 I

I I I1 I

j II 00.

I 3-4 I3WSERWVICE IIBSW NUCLEAR COMPANY

Figure 3-2. Typical Surveillance Capsule Assembly Showing Location of Snecimens and Monitors Lock Assembly rensile -Monitor-Compartment l1 Wedge Coupling Assembly Charpy Impact Compartments Tensile -Monitor Compartment -

Charpy Impact Compartments 3-5 NUCLEAR 13MERW1CEWCOMPANY

L Figure 3-3. Typical Surveillance Capsule Tensile - Monitor L

Compartment Assembly (Three

, Der CaDsule)

L Wedge Coupling - End Cap-. L Flux Spectrum Monitor Cadmium Shielded L

-Stainless Steel Tubing I..

'Cadmium Shield

'Threshold Detector .F I

Temperature Monitor-Low Melting Alloy I1 Housing I

Tensile Specimen Split Spacer- J.

Tensile Specimen Housing I'

I,

-I

-Rectangular Tubing Ix

-Wedge Coupling - End Cap 3-6 1115W VICE 13 WSER NUECLEAR CO MPAN Y

Figure 3-4. Typical Surveillance Capsule Charpy Impact ComDartment Assemblv (Four Der CaDsule)

Coupling - End Cap

'Charpy Impact Specimens SSpi acei Rectangular Tubing Wedge Coupling - End Cap 3-7 DI1BCW NUCLEAR UM SERVICE COMPANY

L Page Intentionally Left Blank I

1.

GWNUCLEA~

4. PRE-IRRADIATION TESTS Unirradiated material was evaluated for two purposes: (1) to establish a baseline of data to which irradiated properties data could be referenced; and (2) to determine those material properties to the extent practical from available material, as required for compliance with Appendixes G and H to 10CFR50.

The pre-irradiated specimens were tested by Combustion Engineering as part of the development of the Waterford Unit 3 surveillance program. The details of the testing procedures are described in C-E Report TR-C-MCS-0027 and are summarized here to provide continuity.

4.1. Tension Tests Tension test specimens were fabricated from the reactor vessel shell plate, HAZ metal, and weld metal. The specimens were 3.00 inches long with a reduced section 1.50 inches long by 0.250 inch in diameter. The tensile tests were performed using a Riehle universal screw testing machine with a maximum capacity of 30,000 lb and separate scale ranges between 50 lb and 30,000 lb. The machine is capable of constant cross head rate or constant strain rate operation.

Elevated temperature tests were performed in a 2-1/2" ID x 18" long high temperature tensile testing furnace with a temperature limit of 18007F. A Riehle high temperature, dual range extensometer was used for monitoring specimen elongation.

Tensile testing was conducted in accordance with ASTM E-8, "Tension Tests of Metallic Materials:"9 and/or Recommended Practice E-21, "Short-Time Elevated Temperature Tension Tests of Materials," 9 except as modified by Section 6.1 of Recommended Practice E-184, "Effects of High-Energy Radiation on the Mechanical Properties of Metallic Matereials." 10 4-1 ZIIBCWNCLEAtR 13 USBEANVCIC COMPANY

For each material type and/or condition, nine specimens in groups of three were tested at room temperature, 250 and 550F. All test data for the pre-irradiation tensile specimens are given in Appendix B.

L 4.2. Impact Tests L Charpy V-notch impact tests were conducted in accordance with the requirements of ASTM E23-7211 on a Model SI-1 BLH Sonntag Universal Impact Machine certified 1 to meet Watertown standards.'2 Test specimens were of the Charpy V-notch type, which were nominally 0.394 inch square and 2.165 inches long. 1 Impact test data for the unirradiated baseline reference materials are presented in Appendix C. Tables C-i through C-4 contain the basis data that are plotted l in Figures C-i through C-4. These data were replotted and re-evaluated to be consistent with the irradiated Charpy curves and evaluations. l 4-2 13JWSBERVICECOPN_

5. POST-IRRADIATION TESTING 5.1. Visual Examination and Inventory The capsule was inspected and photographed upon receipt and confirmed that the markings as those of Capsule W-97. The contents of the capsule were inventoried and found to be consistent with the'surveillance program report inventory. All specimens were visually examined and no signs of abnormalities were found. There was no evidence of rust or of the penetration of reactor coolant into the capsule.

5.2. Thermal Monitors Surveillance Capsule W-97 contained three temperature monitor holder blocks each containing four fusible alloys with different melting points. Each of the thermal monitors was inspected and the results are tabulated in Table 5-1.

Photographs of the monitors are shown in Figure 5-1.

From these data, it can be concluded that the irradiated specimens had been exposed to a maximum temperature no greater than 580F during the reactor vessel operating period. This is not 'significantly greater than the nominal inlet temperature of 550F, and is-considered acceptable.' However, the partly melted or slumped appearance of the 558F monitor is probably due to an irradiation induced creep mechanism 'and not the result of actual melting. This being the case, then the maximum temperature was no greater than 558F which is the most likely case. This behavior has been seen in other surveillance capsules. There appeared to be no signs of a significant temperature gradient along the capsule length.

5.3. Tension Test Results.

The results of the post-irradiation tension tests are presented in Table 5-2.

Tests were performed on specimens at room temperature; 250, and 550F. -They were tested on a 55,000-lb load capacity MTS servohydraulic computer-controlled universal test machine. All tests were run using stroke control with an initial 5-1 BUIIVBCWNUCLEANY

• IJJSSER VCE COMPANY

actuator travel rate of 0.005 inch per minute through yield point. Past specimen yielding an actuator speed of 0.040 inch per minute was used. A 4-pole extension t device with a strain gaged extensometer was used to determine the 0.2% yield point. Test conditions were in accordance with the applicable requirements of ASTM A370-77.13 For each material type and/or condition, specimens were tested t at room temperature, 250 and 550F. The data for both the heat-affect zone specimen and the weld metal specimen, tested at 250F, were lost because of a test it machine malfunction. The tension-compression load cell used had a certified accuracy of better than +0.5% of full scale (25,000 lb). Photographs of the I tension test specimen fractured surfaces are presented in Figures 5-2 through 5-4.

In general, the ultimate strength and yield strength of the material increased with a corresponding slight decrease inductility as compared to the unirradiated values; both effects were the result of neutron radiation damage. The type of behavior observed and the degree to which the material properties changed is I within the range of changes to be expected for the radiation environment to which the specimens were exposed. 1 The results of the pre-irradiation tension tests are presented in Appendix B.

5.4. Chargy V-Notch Impact Test Results l The test results from the irradiated Charpy V-notch specimens of the reactor vessel beltline material are presented in Tables 5-3 through 5-6 and Figures 5-5 l through 5-8. Photographs of the Charpy specimen fracture surfaces are presented in Figures 5-9 through 5-12. The Charpy V-notch impact tests were conducted in accordance with the requirements of ASTM E23-88'4 on a Satec S1-1K impact tester certified to meet Watertown standards.'0 I The data show that the materials exhibited a sensitivity to irradiation within the values to be expected based on their chemical composition and the fluence to which they were exposed. Detailed discussion of the results are provided in Section 7.

The results of the pre-irradiation Charpy V-notch impact tests are given in Appendix C. l 5-2 1 3W5ERVICE NUCLEAR SE88BW COMPANY _

Table 5-1. Conditions of Thermal Monitors in Capsule W-97 Capsule Melt Post-Irradiation Segment Temperature 'Condition Al 536F Melted (Top) 558F Melted (slumped?)

580F Unmelted 590F Unmelted A4 536F Melted (Middle) 558F Melted (slumped?)

580F Unmelted.

590F Unmelted A7 536F Melted (Bottom) 558F Melted (slumped?)

580F Unmelted 590F Unmelted 5-3 OWSIEII NUCLEAR 1W IRWS ER VICE COMPANIY

Table 5-2. Tensile Properties of Capsule 1 -97 Base Metal and Weld Metal Irradiated to 6.47 x 10 n/cm2 (E > 1 MeV)**

Strength, psi Fracture Elongation. % Reduction Specimen Test Temp, Load, Stress, Strength, in Area, No. F Yield Ultimate 1bs Psi psi Uniform Total  %

Base Metal, M-1004-2. Transverse 2L6 70 70,400 92,600 3,097 173,000 63,100 11.7 26.2 63.5 2K5 250 65,500 85,800 2,834 175,300 57,700 10.2 23.1 67.1 2K2 550 63,500 90,000 2,994 162,900 61,000 10.2 23.0 62.5 Base Metal Heat-Affected Zone. M-1004-2 L,

4K3 70 69,500 93,5 00 2,844 184,700 57,500 7.0 20.3 68.9 4KK 4J4 550 69,600 91,0 00 2,913 194,700 - 59,300 6.4 18.5 69.5 Weld Metal 88114/0145 3JM 70 84,500 95,900 3,351 187,100 68,300 7.3 7.9* 63.5 3KK ---

3KY 550 74,000 93,200 2,766 187,700 56,400 7.9 22.6 70.0

• Results not valid - specimen necked and fractured outside extensometer gage length.

• • Stress-strain curves are presented in Appendix F.

I L_- - I - II I L .s t- e - .

Table 5-3. Charpy Impact Results for Capsule W-97 Base Metal Lon2 gitudinal (LT) Orientation, Heat No. M-1004-2, 6.47 x 1018 n/cm Test Impact Lateral Shear Specimen Temperature, Energy, Expansion, Fracture, ID F ft-lbs inch 14Y -50 '7.5 0.006 0 11D -25 24.0 0.021 0 13D 0 -22.5 0.025 10 133 20 36.0 0.032 10 15C 30 ' 84.5 0.061 40 12P 35 76.0 0.054 40 14C 50 - 72.5 0.057 40 15K 70 '90.0 0.069 80 132 100 113.0 0.075 70 11E 150 156.0* 0.093 100 14K 200 152. 0* 0.093 100 11Y 550 157.0 0.082 100

• Values used to determine upper-shelf energy value per ASTM E-185.15 Table 5-4. Charpy Impact Results for Capsule W-97 Base Metal Transverse (TL)

Orientation, Heat No. M-1004-2, 6.47 x 1018 n/cm2 Test Impact Lateral Shear Specimen Temperature, Energy, Expansion, Fracture, ID F ft-lbs inch 216 -50 7.0 0.006 0 214 -25 11.0 0.011 5 25A 0' 15.0 0.017 5 21C 10 36.0 0.031 10 22P 20 53.0 0.044 20 26K 35 54.0 0.047 40 25K 50 73.0 0.057 50 217 70 71.5 0.057 100 22L 100 88.5 0.070 70 245 150 121.5* 0.086 100 244 200 125.0* 0.086 100 22M 550 124.0 0.081 100

• Values used to determine upper-shelf energy value per ASTM E-185.'5 5-5 15111 SWNUCLEAA IM SIVERVICE COMPANY

BAW-2 177-01 I

Table 5-5. Charpy Impact Results for Capsule W-97 Base Metal Heat-Affected l Zone Material, Heat No. M-1004-2, 6.47 x 1018 n/cm2 a I

Test ImatLateral Shear Fracture, Specimen ID Temperature 0F mne Enry tls ay, Ebspansion, inch I

415 -100 21.0 0.014 5 46Y 455

-85

-65 35.0 53.5 0.026 0.036 25 15 I

45J -50 90.0 0.060 60 42C 43D 0

10 115.5 101.0 0.071 0.067 80 70 I

44U 20 121.0 0.073 85 45K 45Y 50 70 119.5 155.0*

0.077 0.083 85 100 I

41M 474 414 100 150 550 163.5*

150.0*

>240.0 0.090 0.071 100 100 I

• Values used to determine upper-shelf energy value per ASTM El 85.'5 L Table 5-6. Charpy Impact Results for Capsule W-97 Weld Metal, 88114/0145, 6.47 x 1018 n/cm2 L

I I

I I

L I.

• Values used to determine upper-shelf energy value per ASTM El 85.15 I A

I 5-6 FRAMATOME ANP I

III I . .

I I I I I I I I I I I I I I Figure 5-1. Photographs of Thermal Monitor Melt Wire Capsules as Removed From Surveillance Capsule en IE1 Bottom Group Middle Group Top Group ftY'

I Figure 5-2. Photographs of Tested Tension Test Specimens and Corresponding L

Fractured Surfaces - Base Metal. Transverse Orientation L

L Specimen 2L6 (70F)

L

,/A' Specimen 2K5 (250F)

A L

L Specimen 2K2 (550F)

L L

L L

L

'I L

Specimen 2L6 (70F) Specimen 2K5 (250F) Specimen 2K2 (550F) I L

5-8

/IIBC WNUCLE4R ISWSERVICE COMPANY I

I I .. .. . ., .f -.1, ,41. - . .

Specimen 4K3 No photograph Specimen 4KK (250F)

."F_, .,:

l S~pecimen 4J4 (550F)

No Photograph Specimen 4K3 (70F) Specimen 4KK (250F) Specimen 4J4 (550F) 5-9 O3WSEsR 6 596WNU)CLEARl ICE COMPANY y

U-L Figure 5-4. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Weld Metal 8R114/0145 L

L L

'I f,

.a -.1_ - , ;, - -., 1,:,. I.--.'.J Specimen 3dM (70F)

L Ii Specimen 3KK (250F) 1 I

12 Specimen 3KY (550F) I I

I I

I I

Specimen 3UM (70F) Specimen 3KK (250F) Specimen 3KY (550F)

I 151B8W NUCLEAR UWSEREVICE COMPANY 5-10 -l

Figure 5-5. Charpy Impact Data for Irradiated Base Metal, Longitudinal Orientation. Heat No. M-1004-2 100 75 U..

i'U 50 U,

25

' '0 0.10 1:: 0.08 22

=0.06

~.x 0.0 w 0.04 I-i

1. ! 0.02 z 0

• 0f

220 I I I I I I

- DATA

SUMMARY

200 TNDT TCv (35 MLE) +8F 180 - Tcv (50 FT-LB) +30F Tcv (30 FT-LB) +3F

= 160 - r._ -I oer- t.

PV-UQL %#%vu.

1I .wrft-lhc s w RTNDT0 O 140

.0.

0 Cn 120 _-

.0 6, 100 _-

w S

CL 80 -

E 60 _ S 40 -

MATERIAL SA-533ICIB1(L

' S~ FLUENCE 6.47 X 10' n/cm -

20 -

HEAT NO. M-1004-2 F . ..

I l

I l I I I ;

nLA'0

--100 0 100 200 - 300 400 500 600 Test Temperature, F 5-11 SEREVICE COMPANY

L Figure 5-6. Charpy Impact Data for Irradiated Base Metal, L Transverse Orientation. Heat No. M-1004-2 100 L 75

!U U-50 L

c 25 L 0

.C 0.10 II I I I I L

C* 0.08 0

R' 0.06 x

u.

w C 0.04 co

-J z

. 0.02 0

II I I I I

1 220

-DATA I

SUMMARY

. I .. I I I

200 TNOT ...

180 TCv (351MLE)

Tcv (50 FT-LB)

+14F

+32F I Tcv (30 FT-LB) +7F D 160 Cv-USE (AVG.) 123ft-lbs RTNDT ...

C 140 0

0L

.0 Ow120 1 4

100 w

Cl

0. 8 I

E 60 I 40 20 MATERIAL SA-533,CIB1(T FLUENCE 6.47 X 10 n/cmr I

HEAT NO. M-1004-2 n

-100 0 100 200 300 400 500 600 Test Temperature. F I

5-12 WWNU IWS"E'RVICE COMPANY .1

Figure 5-7. Charpy Impact Data for Irradiated Base Metal, Heat-Affected Zone. Heat No. M-1004-2 100 I

75 50 CO, 25 0

0.10 c' 0.08

• 0V t.

x 0.06 X 0.04

-j 2 0.02 o 4 z

0 L 220 . . .

l l I. I 0

-DATA

SUMMARY

200 _-

t 180 _- Tcv (35 MLE) -69F Tcv (50 rT-Le) -70F 0

-cTvy(30 FT-LB) 90F 160 Cv-USE (AVG.) 1561t-lb -

0

-0 RT __OT 140 F 0.

0 In 120 f 0

100 t CD a 0 80 i E

60 f 40 MATERIALSA-533,CIB (HAZ) 20 FLUENCE 6.47 X 10 n/cm HEAT NO. M-1004-2 I. I . I .I nL . l v

-100 0 100 200 300 400 500 600 Test Temperature, F 5-13 WSERVINCEUCOMPANY

L Figure 5-8. Charpv Impact Data for Irradiated Weld Metal. 88114/0145 100 _ ' L 75-50 _-

U-

,25 [

0 o0.10 1L C* 0.08 0

ao x

0.06 - I z 0.04 1

' 0.02 -

0 0

220 200

-DATA TNDT

SUMMARY

I Tcv (35 MLE) -38F 180 Tcv (50 Fr-LB) 35F Tcv (30 FT-LB) -44F I 160 Cv USE (AVG.) 143ft-1bs

.0 RTNOT ...

140 I

2 0.

0

.0 120 0 I U.' 100 80 0.

E 60 I 40 20 MATERIAL Weld Metal FLUENCE 6.47 X 10'8n/cm 2 I

7 o

-100 0 100 Il I 200 '

300 HEAT NO.

400 88114/0145 500 600 I

Test Temperature, F 5-14 9 NUCEA lBW BURwE'RWCE COMPAY I

I-- - - I ------ I --- I . I f- -- I------ I--- I , I I I I I I I I I I Fiourp 5-9, PhotoaraDhs of Charov Imnact Specimen Fracture Surfaces - Base Metal, Longitudinal Specimeiii 14Y (-5uF) Specimen lID (-25F) Specimen 13D (OF) Specimen 133 (20F) Specimen 15C (30F) Specimenl 12P (35F) to e%R W

-VE Q

R 133 1%

Specimen 132 (106F) Specimen IIE (150F) Specimen 14K (200F) Specimen 1lY (550F)

Specimen 14C (buF) Specimen 15K (70F)

Figure 5-10. Photographs of Charpy Impact Specimen Fracture Surfaces - Base Metal. Transverse l.

1_

Specipen 216 (-50F) Specimlen 214 (-25F) Specimen 25A (OF) SpeciInen 21C (10W) Specimen 22P (20F) Specioen 26K~(35F)

I Specimen 25K (50F) Specimen 217 JOF) Specimen 221i (550F)

- _ ;_ ;_ A_ ?0-- t- -- T - ro-- rm - - r~-

r--- , F- I --- I-- I--- t [. ---- I,- F -- (--- I- - I---- I---- I I- I I I I I Pinilro 9-1 1 Phntnnrnnh' n~f r~hnnv Tmnnr-f l~n~rirnn Fr'arttrvP -Iutifarp' - Raq6 Mptal 1pat-Affprtpd lann

. . --. - - . . .- ---. - 1 I .1l.r - - - - - - -I.- . . I -

L"1 I.j Specimen 415 (-IOOF) Specimen 46Y (-85F) Specimen 455 (-65F) Specimen 4WJ (-UFP) Specimen 42C (OF) Specinen 43D (lOF)

Mh eQ Speciraen 44U (20F) Specimen 45K (50F) Specimen 45Y (70F) Specimen 41M'(IUOF) Specimen 474 (150F) Specimen 414 (550F)

Fioure 5-12. PhotograDhs of Charny ImDact Snecimen Fracture Surfaces - Weld Metal 88114/0145 Ln c:

51mcitaell 35J j-tluf) Speciiacn 32T (-40F) Sptecimcn 31K1(-3bF) Speciiwcn 32.K (-20F) Spcciiwin 362 (-1!)K) Specimaen 34L (OK)

Specioeien 31T (2UF) Specime~n 35K (70F) Spec iihan 33C (1UUF) Specim~en 374 (201F) SPt~kedw 33B1(550F)

I * - =

6. -NEUTRON FLUENCE 6.1. Introduction The neutron fluence (time integral 'of flux)' is a quantitative way of expressing the cumulative exposure of a material to -apervading neutron flux over a specific period of time. Fast neutron fluence, defined as the fluence of neutrons having energies greater than I MeV, is the parameter that is presently used to correlate radiation induced changes in material properties. Accordingly, the fast fluence must be determined at two locations: (1)in the test specimens located in the surveillance capsule, and (2)in the wall of the reactor vessel. The former is used in developing the correlation between fast fluence and changes in the material properties of specimens, and the latter is used to ascertain the point of maximum fluence in the reactor vessel, the relative radial and azimuthal distribution *of the fluence, the fluence gradient through the reactor vessel wail, and the corresponding material properties.

The accurate determination of neutron flux is best accomplished through the.

simultaneous consideration of neutron dosimeter measurements and analytically derived flux spectra. Dosimeter measurements alone cannot be used to predict the fast fluence in the vessel wall or in the test specimens because (1)they cannot measure the fluence at the points of interest, and' (2) they provide only rudimentary information about the neutron energy spectrum.' Conversely, reliance on calculations 'alone to predict fast fluence is not prudent because of the length and complexity of the analytical procedures involved. In short,'

measurements and calculations are necessary complements of each other and together they provide assurance of accurate results.

Therefore, the determination 'of the fluence is accomplished using a combined analytical-empirical methodology which is outlined in Fig'ure' 6-1 and described in the following paragraphs. The details of the procedures and methods are presented in general terms in Appendix D and in BAW-1485P.'8' 6-1 15-11B6W NUCLEPAR 1; WSERVICE COMPANY

L Figure 6-1. General Fluence Determination Methodology L

Measurements of Neutron Analytical Determination of L Dosimeter Actitivies Dosimeter Activities and Neutron Flux I L Adjusted Energy Dependent Neutron L

Flux I

Neutron Fluence Reactor Operating History and Pre-dicted Future L

operation I

1 I

Analytical Determination of Dosimeter Activities and Neutron Flux I The analytical calculation of the space and energy dependent neutron flux in the test specimens and in the reactor vessel is performed with the two dimensional I discrete ordinates transport code, DOTIV.17 The calculations employ an angular quadrature of 48 sectors (S8), a third order LeGendre polynomial scattering I approximation (P3), the BUGLE cross section set18 with 47 neutron energy groups and a fixed distributed source corresponding to the time weighted average power distribution for the applicable irradiation period.

I In addition to the flux in the test specimens, the DOTIV calculation determines 1 the saturated specific activity of the various neutron dosimeters located in the surveillance capsule using the ENDF/B5 dosimeter reaction cross sections.)9 The saturated activity of each dosimeter is then adjusted by a factor which corrects l

for the fraction of saturation attained during the dosimeter's actual (finite) l l1 6-2 1 W!S'E'NVCEW MPy -l

ft.

irradiation history. Additional corrections are made to account for the following effects:

• Photon induced fissions in U.dosimeters (without this correction the results underestimate the measured activity).'

• Short half-life of isotopes produced in nickel, iron and titanium dosimeters (71 day Co-58, 312 day Mn-54 *and 84 day Sc-46 respectively).

(Without this correction, the results could be biased high or low depending on the long term versus short term power histories.)

Measurement of Neutron Dosimeter Activities The accuracy of neutron fluence predictions is improved ifthe calculated neutron flux is compared with neutron dosimeter measurements adjusted for the effects noted above. The neutron dosimeters located in the surveillance capsules are listed inTable 6-1. Both activation type and fission type dosimeters-were used.

The ratio of measured dosimeter activity to calculated dosimeter activity (M/C) is determined for each dosimeter, as discussed in Appendix D. These M/C ratios are evaluated on a'case-by-case basis to assess the dependability or veracity of each individual dosimeter response. After carefully evaluating all factors known to affect the calculations or the measurements, an average M/C ratio is calculated and defined'as the "normalization factor." The normalization factor is applied as an adjustment factor to the-DOT-calculated flux at all points of interest.

Neutron Fluence The determination of the neutron fluence from the time averaged flux requires only a simple multiplication by the time in EFPS (effective full-power seconds) over which the flux was averaged, i.e.

fin(AT) ESg4 §IjgAT where f11(AT) = Fluence at (i,j) accumulated over time T (n/cm2),

g = Energy'group index, 6-3

- IUSINIZCR UCLEA

L

= Time-average flux at (i,j) in energy group g, (n/cm 2-sec), L AT Irradiation time, EFPS. L Neutron fluence was calculated inthis analysis for the following components over the indicated operating time: L Test Specimens: Capsule irradiation time in EFPS Fluence Monitors: Capsule irradiation time in EFPS L Reactor Vessel: Vessel irradiation time in EFPS [

Reactor Vessel: Maximum point on inside surface extrapolated to 32 effective full power years The neutron exposure to the reactor vessel and the material surveillance L specimens was also determined in terms of the iron atom displacements per atom (DPA) of iron. The iron DPA is an exposure index giving the fraction of iron L atoms in an iron specimen which would be displaced during an irradiation. It is considered to be an appropriate damage exposure index since displacements of L atoms from their normal lattice sites is a primary source of neutron radiation damage. DPA was calculated based on the ASTM Standard E693-79 (reapproved 1985).20 A DPA cross section for iron is given in the ASTM Standard in 641 energy groups. DPA per second is determined by multiplying the cross section at a given energy by the neutron flux at that energy and integrating over energy.

DPA is then the integral of DPA per second over the time of the irradiation. In the DPA calculations reported herein, the ASTM DPA cross sections were first I collapsed to the 47 neutron group structure of BUGLE; the DPA was then determined by summing the group flux times the DPA cross section over the 47 energy groups L and multiplying by the time of the irradiation.

6.2. Vessel Fluence [

The maximum fluence (E > 1 MeV) exposure of the Waterford Unit 3 reactor vessel during Cycles I to 4 was determined to be 5.13 x 1018 n/cm 2 based on a maximum l neutron flux of 3.66 x 1010 n/cm2-s. The maximum fluence occurred at the clad-ding/vessel interface at an azimuthal location of approximately I degree from a major horizontal axis of the core (Figure 6-3). Cumulative DPA results were calculated at the quarter T positions and are presented in Table 6-4. 1 6-4 I3 186WNUCLEAR OMSERVICE COMPAFNY

Fluence data were extrapolated to 32 EFPY of operation based on two assumptions:

(1) the future fuel cycle operations do not differ significantly from the cycles 1 to 4 design, and (2) the latest calculated (or extrapolated) flux remains constant from EOC4 through 32 EFPY. The extrapolation was carried out from EOC 4 to 32 EFPY. The cycle averaged fluxes for future cycles are assumed to be the average flux experienced during cycles 1 to 4.

Fast fluence and DPA (displacements per atom) gradients relative to the inside surface of the vessel wall are shown in Figure 6-2. Reactor vessel neutron fluence lead factors, which are the 'ratio of the neutron flux at the clad interface to that in the vessel wall at the T/4, T/2 and 3T/4 locations, are 1.87, 4.03, and 9.17, respectively. DPA lead factors at the same locations are 1.62, 2.79, and 5.00, respectively. The relative fluence as.a function of azimuthal angle is shown in Figure 6-3. The peak average flux from cycles 1 to 4 occurred at about I degree with a corresponding value of 3.66 x 1010 n/cm 2-s.

The flux and fluence results were corrected using the final measured to calculated activity ratio (M/C) derived from the capsule (0.958) and were also corrected to account for an axial power peak (1.08). The M/C ratio is detailed in Appendix D. The axial fluence, which was normalized over the height of the core and assumed to be proportional to the axial power distributions in the peripheral assemblies, was averaged over cycles 1 to 4. Table 6-7 shows the nodal values used to obtain the axial factors. These values were based on time-averaged nodal values obtained from the customer. Figure 6-4 shows the axial flux variation, overlaid by an image of the capsule showing the axial factors in each dosimeter compartment.

6.3. Cansule Fluence The 970 capsule-was'irradiated in Waterford Unit 3 for Cycles :1 to 4, 4.44 EFPY, at a location 7 degrees off a major horizontal axis. The cumulative fast fluence at the center of the surveillance capsule was calculated to be 6.47 x 1018 n/cm2.

This fluence value represents an average value for the various locations in the capsule. It includes an axial peaking factor of 1.08 and a normalization factor of 0.958. The fluence is approximately 6% higher at the center of the charpy specimens closest to the core and approximately 6% lower at the center of the 6-5 11VBIW NUCLEAR JRWSERVICE COMPANY

M-L charpy specimens away from the core. Figure 6-5 shows a sketch of the capsule L

and pressure vessel, which includes the radial dimensions from the core center supplied by the customer, although the dimensions have been converted from the L inches in which the information was supplied to the centimeters which were used in the modelling. L 6.4. Fluence Uncertainties Surveillance capsules provide neutron dosimetry information as well as materials L data at various points during the lifetime of power reactors. The dosimetry results, measured-to-calculated ratios, obtained from numerous analyses utilizing L the same methodology provide a measure of confidence inthe analytical techniques and a benchmarking for the methodology used to determine vessel fluence. Table 6-6 presents a comparison of the results of fourteen surveillance capsule I-analyses which utilized B&W's methodology.

I Table 6-1. Surveillance CaDsule Dosimeters 1 Lower Energy Dosimeter Reactions(a)

Limit for Reaction. MeV Isotope Half-Life I

58 Ni (n,p) 58 Co 54Fe(np)54Mn 2.3 2.5 70.8 days 312.5 days I

63Cu(n,a). 0Co 3.7 5.27 years 1 45Ti (n,p)48Sc 83.81 days 1.9 238U(n, f)37cs 1.1 30.0 years I

59Co(n, y)60 Co thermal 5.27 years I (alReaction activities measured for capsule flux evaluation.

I l

I I

6-6 BUZBWNUCLA IOWSEARWl COMPANY l1

Table 6-2. Waterford Unit 3 Reactor Vessel Fast Flux Flux n/cm2 -s Fast Flux (E > 1 MeV), n/cm2-s (E > 0.1 MeV)

Inside Surface** TA -. Inside Surface Cycle (Max Location) T/4 3T/4 (Max Location)

Cycles 1 to 4 3.66E+10 1.96E+10 3.!99E+9 7.91E+10 5 EFPY 3.66E+10 1.96E+10* 3.!99E+9*

6 EFPY 3.66E+10 1.96E+10* 3.!99E+9*

7 EFPY 3.66E+10 1.96E+10* 3.!99E+9*

8 EFPY 3.66E+10 1.96E+10* 3.!99E+9*

16 EFPY 3.66E+10 1.96E+10* 3.!99E+9*

24 EFPY 3.66E+10 1.96E+10* 3.'99E+9*

32 EFPY 3.66E+10 1.96E+10* 3.'99E+9*

• Divide flux at inside surface by the appropriate lead factors on page 6-5 to obtain these T/4 and 3T/4 fast flux values.

• • Clad/Base metal interface at 221.54 cm from core center.

6-7 5108CW NUCLEAR RW SERVICE COMPANY

L-Table 6-3. Calculated Waterford Unit 3 Reactor Vessel Fluence L

Cummulative Fast Fluence. n/cm2 (E > I MeV)

Inside Surface"'" L Irradiation Time (Max Location) T/4 T/2 3T/4 End of Cycle 4 5.13E+18 2.74E+18 1.27E+18 5.59E+17 5 EFPY 5.76E+18 3.08E+18* 1.43E+18* 6.29E+17*

6 EFPY 6.92E+18 3.70E+18* 1.72E+18* 7.54E+17*

7 EFPY 8.07E+18 4.32E+18* 2.OOE+18* 8.80E+17*

8 EFPY 9.22E+18 4.93E+18* 2.29E+18* 1.01E+18*

16 EFPY 1.84E+19 9.86E+18* 4.58E+18* 2.01E+18*

24 EFPY 2.77E+19 1.48E+19* 6.87E+18* 3.02E+18*

32 EFPY 3.69E+19 1.97E+19* 9.15E+18* 4.02E+18*

• Calculated using these lead factors.

1.00 1.87 4.03 9.17 1 (a'Clad/Base metal interface at 221.54 cm from core center. I Conversion Factors Fluence (E > 1 MeV) 1.50E-21** 1.72E-21** 2. 16E-21** 2.73E-21** I to DPA.

• • Multiply fast fluence values (E > 1 MeV) in units of n/cm 2 by these factors to obtain the corresponding DPA values.

6-8

Tible 6-4. Calculated Waterford Unit 3 Reactor Vessel DPA DPA, Displacements/Atom (Total)

Cummulative Inside Surface"^'

Irradiation Time (Max Location) T/4 3T/4 End of Cycle 4 7.67E-3* 4.73 E-3* 2.76E-3* 1.53E-3*

5 EFPY 8.62E-3* 5.31E-3* 3.I E-3* 1.72E-3*

6 EFPY 1.03E-2* 6.38E-3* 3.72E-3* 2.06E-3*

7 EFPY 1.21E-2* 7.44E-3* 4.33E-3* 2.41E-3*

8 EFPY 1.38E-2* 8.50E-3* 4.95E-3* 2.75E-3*

16 EFPY 2.76E-2* 1.70E-2* 9.91E-3* 5. 50E-3*

24 EFPY 4.14E-2* 2.55E-2* 1 .49E-2* S. 25E-3*

32 EFPY 5.51E-2* 3.40E-2* 1.98E-2* 1.10E-2*

(aiClad/Base metal interface at 221.54 cm from core center.

• Calculated using these 1.00 1.62 2.79 5.00 lead factors Conversion Factors Fluence (E > 1 MeV) 1. 50E-21** 1.72E-21** 2.16E-21** 2.73E-21**

to DPA.

• • Fast fluence values (E > I MeY) in units of n/CM2 were multiplied by these factors to obtain the corresponding DPA values.

Table 6-5. Fluence. Flux. and DPA for 97 Surveillance Cansule E > 0.1 E > 1.0 MeV MeV Fl ux Fl uence, Flux, Capsule Irradiation Time n/cm~ n/cm2 DPA n/cm2 W-97 Cycles 1 to 4 4.62E+10 6.47E+18 9.25E-3 8.63E+10 (4.44 EFPY) 6-9 551RIWANUCOPAR IMWSEERVICE COMPANY

-A-Table 6-6. Surveillance Capsule Measurements IL Plant Capsule Measured/

Calculated L Arkansas One, Unit 1 Rancho Seco ANI-C RS1-F 1.04 1.03 L

Crystal River-3 CR3-F 0.99 L Oconee Unit 1 OC1-C 1.01 Oconee Unit 2 OC2-E 0.98 L Davis-Besse DBI-LG1 1.08 Crystal River-3 CR3-LGI 1.06 Oconee Unit 3 OC3-D 1.00 L

Davis-Besse TE1-D 1.03 St. Lucie W-83 1.08 L Shearon Harris U 0.88 Zion Unit 1 Y 1.11 Millstone Unit 2 W-104 0.99 Millstone Unit 2 W-97 0.94 I

Average M/C for 14 surveillance data points = 1.02 1 Sigma standard deviation of data base = 0.06 I

I I

I I

I 6-10 CLEAR SUREORWNCY COMPAN Y l1

.. Table 6-7. Axial Power Data-Affectina Flux Node Exposure* Relative EXD.

1 3.272 0.750 2 4.360 1.000 3 4.666 1.070 4 4.731 1.085 5 4.726 1.084 6 4.704 1.079 7 4.679 1.073 8 4.649 1.066 9 4.602 1.055 10 4.491 1.030 11 4.179 0.958 12 3.269 0.750 Avg 4.361

• These exposure values are based upon the nodal values for assemblies [9,1] and

[9,2] supplied by the customer and time-averaged for cycles 1 through 4.

6-11 SWUSERVICE COMPANY

z L

Figure 6-2. Fast Flux, Fluence and DPA Distribution L

Throuah Reactor Vessel Wall I 4- - L

-. LF(D)=1.62 L

LF(D) = 2.79 L

0.5 -221.54 cm Inside Surface Clad-Base Met L

Interface L

0.

a 0.3 227.02 cm I L j 0.2 j _ _-

L L

T12 a:

I~ 0.1 232.50 cm Ii

_1 A94.3___

I 3T14 71 I 0.05 -Flux/Fluence - DPA--------- L-9 LF(F)= 9.17 I

LF(F) Is Lead Factor for Fluence 243.45 cm I LF(D) Is Lead Factor for DPA Outside Surface 0.03 22 0 225 230 235 240 245 I Radial Position (cm)

I1 6-12 I

DI BSW NUCLEAR SWSE1RVICE COMPANY I1

i- - -- r Figure 6-3. Azimuthal Flux and Fluence Distributions at Reactor Vessel Inside Surface 1.1.._

1E 0.9 0.8 I, 0 0.6 0.6 0.5_

BY 0.4 0.3 0 10 20 30 40 50 Azirnuthal Position (degres)

Figure 6-4. Relative Axial Variation of E > I MeV Flux/Fluence 1.2 1.1 A

1 U

c a)

Plt it it 0E (U

0.9 0.8 0

eMlh

,:E 0.7 nz Mr- 0 an 20 40 60 80 100 120 140 160 N Distance from Bottom of Core (in.)

- -N -r r I r ~ I- - 1-

Figure 6-5. Radial Dimensions Used in Modeling CaEnnuI and Pressure Vessel Regions NOTE: Distances are from core center.

. . I. . . . . .

MATERIALS:

SS-304 I I . ,WATER

... ... .... ...... PV STEEL I . . . . .. ~ I. . . . .... . .

1111I 111111 A B C D E F G H I J K l M N PosrTIONI DISTANCE(cm) POSmON I DISTANCE(cm) POSmON DISTANCE(cm)

A 215.31 F 217.27 K 219.22 B 215.67 G 217.50 L 220.98 C 216.32 H 217.95 M 221.54 D 216.59 218.22 N 243.45 I

E - _____

217.04 218.87 J _______________ I __________ .1 6-15 1511B6WNUCLEAR INW SERVICE COMPANY

IL L

L L

L L

• L L

• L Page Intentionally Left Blank I,

I3IlBSW NUCLEAR WSERWVICE COMPANY

7. DISCUSSION OF CAPSULE RESULTS 7.1. Pre-Irradiation Propertv Data The weld metal and base metals were selected for inclusion in the surveillance program in accordance with the criteria in effect at the time the program was designed for Waterford Unit 3. The applicable selection criterion'was based on the unirradiated properties only. A review of the original unirradiated properties of the reactor vessel core beltline region materials-indicated no significant deviation from expected properties. Based on the design end-of-service peak neutron fluence value at the 1/4T vessel wall location and the copper content of the base metals, it was predicted that the end-of-service Charpy upper-shelf energy (USE) will not be below 50 ft-lb.

7.2. Irradiated Propertv Data 7.2.1. Tensile Properties Tables 7-1 and 7-2 compare irradiated properties from Capsule W-97 with the unirradiated tensile properties. At both room temperature and elevated temperature, the ultimate and yield strength changes in the base metal as a result of irradiation and the corresponding changes in ductility are within the limits observed for similar materials. There is some strengthening, as indicated by increases in ultimate and yield strengths and decreases inductility properties. The changes observed in the base metal are such as to be considered within acceptable limits. The changes, at both room temperature and 550F, in the properties of the base metal are equivalent to those observed for the weld metal, indicating a similar sensitivity of both the base metal and the weld metal to irradiation damage. In either case, the changes in tensile properties are insignificant relative to the analysis of the reactor vessel materials at this time period in the reactor vessel service life.

7-1 15111 BW NUCLEAR WWSER VICECOMPANY

1 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 of area. The most significant observation from these data is that the weld metal exhibited slightly greater sensitivity to neutron radiation than the base metal.

7.2.2. Impact ProDerties The behavior of the Charpy V-notch impact data is more significant to the calculation of the reactor system's operating limitations. Table 7-3 compares the observed changes in irradiated Charpy impact properties with the predicted changes. A comparison of the Charpy data curves are presented in Figures 7-1 through 7-4.

The 30 ft-lb transition temperature shift for the base metal in the longitudinal orientation is conservative compared to the value predicted using Regulatory Guide 1.99, Rev. 221 and when the margin is added the predicted value is very conservative. However, the 30 ft-lb transition temperature shift in the transverse orientation is not in good agreement with the predicted using Regulatory Guide 1.99, Rev. 2, and when the margin is added the predicted value equals the measured value without any margin for conservatism. It would be expected that these values for the longitudinal orientation would exhibit good agreement when it is considered that the data used to develop Regulatory Guide 1.99, Rev. 2, was taken from data obtained from longitudinal oriented specimens.

The transition temperature measurements at 30 ft-lbs for the weld metal is in relatively good agreement with the predicted shift using Regulatory Guide 1.99, Revision 2 but the predicted value is not conservative. The predicted shift being slightly under estimated indicates that the estimating technique based on the Regulatory Guide 1.99, Rev. 2, is not overly conservative for predicting the J 30 ft-lb transition temperature shift. Since the method requires that a margin be added to the calculated value to provide a conservative value, the final shift value using Regulatory Guide 1.99, Revision 2, is conservative, and future I

evaluations should be based on Position 2 when additional data are available which will help to account for some of the over-conservatism in the application of Regulatory Guide 1.99, Position 1.

7-2 BSW NUCLEAR I3 WSERVICE COMPANY

The data for the decrease in Charpy USE due to irradiation showed relatively good agreement with predicted values for the base metal. The weld metal decrease in Charpy USE was over predicted by 200 percent. However, the poor comparison of the measured weld metal data with the predicted value is to be expected in view of the lack of data for low copper-content materials at medium fluence values th'at were used to develop the estimating curves.

Results from other surveillance capsules also indicate that RTNDT estimating curves have greater inaccuracies than originally thought. These inaccuracies are a function of a number of parameters related to the basicdata available at the time the estimating curves were established. These parameters may include inaccurate fluence values, inaccurate 'chemical composition values, and variations in data interpretation. The change in the regulations requiring the shift measurement to be based on the 30 ft-lb value has minimized the errors that resulted from using the 30 ft-lb data base to predict the shift behavior at 50 ft-lbs.

The design curves for predicting the shift will continue to be modified as more data become available; until that time, the design curves for predicting the RTNDT shift as given in Regulatory Guide 1.99, Revision 2, are considered adequate for predicting the RTNDT shift of those materials for which data are not available. These curves will be used to establish the pressure-temperature operational limitations for the irradiated portions of the reactor vessel until the time that improved prediction curves are developed and approved.

The relatively,poor agreement of the change in Charpy upper-shelf energy for the weld metal does support the conservatism of the prediction curves for low copper-content materials. However, for low copper-content base materials the predicted values are not conservative. Although the prediction curves are conservative for the weld metal in that they generally predict a larger decrease in upper-shelf energy than is'observed for a given'fluence and 'copper content, the conservatism can unduly restrict the operational limitations. These data support the contention that the upper-shelf energy drop curves will have to be revised as more reliable data become available; until that time the.design curves used to predict the decrease in upper-shelf energy of the controlling materials are considered conservative.

7-3 I3lUWSEREMIZBEWUNUCLEAR VICE COMPANY

7.3. Reactor Vessel Fracture Toughness An evaluation of the reactor vessel end-of-life fracture toughness was made and L the results are presented in Table 7-4.

The fracture toughness evaluation shows that the controlling base metal will have L a T/4 wall location end-of-life RTNDT of 69F based on Regulatory Guide 1.99, Revision 2, including a margin of 24F. The controlling weld metal will have a L T/4 wall location end-of-life RTNDT of 34F based on Regulatory Guide 1.99, Revision 2, including a margin of 52F. These predicted shifts may be excessive since data from the first surveillance capsule exhibited measured RTNDT values that are comparable to the Regulatory Guide mean values. It is estimated that L the end-of-life RTNDT shift for both the controlling base metal and weld metal will be significantly less than the value predicted using Regulatory Guide 1.99, Revision 2 because the use of future surveillance data will permit a reduction l in the applied margin. This reduced shift will permit the calculation of less restrictive pressure-temperature operating limitations than if Regulatory Guide L 1.99, Revision 2, was used.

An evaluation of the reactor vessel end-of-life upper-shelf energy for each of L the materials used in the reactor vessel fabrication was made and the results are presented in Table 7-5. This evaluation was made because the base metals used [

to fabricate the reactor vessel are characterized by upper-shelf energies measured only in the longitudinal orientation. Consequently, when adjusted for the transverse orientation are expected to be sensitive to neutron radiation damages and exhibit values significantly lower than the longitudinal value. The method used to evaluate the radiation induced decrease in upper-shelf energy is L the method defined in Regulatory Guide 1.99, Revision 2, which is the same procedure used in Revision 1. [

The method of Regulatory Guide 1.99, Revision 2, shows that the base metals used in the fabrication of the beltline region of the reactor vessel will have an [

upper-shelf energy greater than 50 ft-lbs through the 32 EFPY design life based on the T/4 wall location. Regulatory Guide 1.99 method also predicts an upper- [

shelf energy above 50 ft-lbs for the controlling base metal at the vessel inside wall. The weld metal upper-shelf energies unirradiated values are so high as to 7-4

^ZBWNUCLEA" I 1MSEFRVICE COMPANIYL

preclude any change of the values decreasing below 50 ft-lbs during the 32 EFPY design life. Based on the first surveillance capsule data, it is estimated that the controlling vessel base metal upper-shelf energy will remain above the required 50 ft-lbs during the vessel design life.

7.4. Operating Limitations The current normal pressure-temperature operating limitations are designed for operation through 8 EFPY. Based on the fluence calculations performed for Capsule W-97 and the results of the Charpy impact test results, the current operating limitations may be extended to 10.5 EFPY. However, any changes must be verified by confirmatory calculations and, in addition, any changes in the fuel cycle designs will require a review and possible verification for extension from the original 8 EFPY limit.

7.5. Pressurized Thermal Shock (PTS) Evaluation The pressurized thermal shock evaluation shown in Table 7-6 demonstrates that the Waterford Unit 3 reactor pressure vessel is well below the screening criterion limits and, therefore, need not take any additional corrective action as required by the regulation.

7.6. Neutron Fluence Analysis These new analyses calculated an end-of-life fluence value of 3.69 x 1019 n/cm2 (E > 1 MeV) at the reactor vessel inside surface peak location. The correspond-ing value for the vessel wall T/4 location is calculated to be 1.97 x 1019 n/cm 2 (E > 1 MeV). These values do not represent a reduction compared to the values calculated based on the design basis fluence values.

7-5 WSAIBWU NUCMPLNY

L Tnhla 7-1 Iu. -. . Comparison of Waterford Unit 3. Capsule W-97 Tension Test Results L

Room Temp Test Unirr** Irrad Elevated Temp Test*

Unirr** Irrad L Base Metal -- M-1004-2. Transverse Fluence, 1018 n/cm2 (E > 1 MeV) 0 6.47 0 6.47 L

Ultimate tensile strength, ksi 89.0 92.6 87.0 90.0 L 0.2% yield strength, ksi 68.1 70.4 64.5 63.5 Uniform elongation, % 11.0 11.7 9.9 10.2 L Total elongation, % 27.3 Reduction of area, % 68.2 26.2 63.5 22.3 65.3 23.0 62.5 L

Base Metal -- Heat-Affected Zone Fluence, 1018 n/cm 2 (E > 1 MeV) 0 6.47 0 6.47 Ultimate tensile strength, ksi 91.3 93.5 86.7 91.0 0.2% yield strength, ksi 68.2 69.5 60.2 69.6 Uniform elongation, % 6.8 7.0 6.6 6.4 Total elongation, % 21.3 20.3 20.3 18.5 Reduction of area, % 69.4 68.9 66.6 69.5 Weld Metal -- 88114/0145 Fluence, 1018 n/cm2 (E > 1 MeV) 0 6.47 0 6.47 Ultimate tensile strength, ksi 92.2 95.9 88.3 93.2 0.2% yield strength, ksi 81.0 84.5 72.2 74.0 Uniform elongation, % 9.6 7.3 9.2 7.9 Total elongation, % 27.7 23.7 22.6 Reduction of area, % 70.7 63.5 69.4 70.0

• Test temperature is 550F.

• • Average of the lower yield strength data in Appendix B.

• • • See footnote Table 5-2.

7-6 1.WU!"EG!VICVU OMANY I

[- I I I - I I I I I *1I I I I I III I I Table 7-2 Summarv of Waterford Unit 3 Reactor Vessel Surveillance Cansule Tensile Test Results Strength, ksi Ductility, %

Cap. Figence, Test Total Ly(a) Reduction Material I.D. 10 n/cm2 Temp, F. Ultimate V Yield W(a) El on. of Area ,W(a)

Base metal 0.00 71 89.0 -- 68.1 -- 27.3 -- 68.2 Transverse 550 87.0 64.5 -- 22.3 65.3 (M-1004-2)

W-97 6.47 70 92.6 + 4 70.4 + 3 26.2 -4 63.5 -7 550 90.0 + 3 63.5 - 2 23.0 +3 62.5 -4 Base metal - - 0.00 71 91.3 -- 68.2 -- 21.3 69.4 Heat-affected 550 86.7 -- 60.2 -- 20.3 -- 66.6 zone (M-1004-2)' W-97 6.47 70 93.5 + 2 69.5 + 2 20.3 - '5 68.9 - 1 1-4 1i 550 91.0 + 5 69.6 +16 18.5 - -9 69.5 +4 Weld metal -- 0.00 71 92.2 -- 81.0 --. 27.7 70.7 (88114/0145) 550 88.3 -- 72.2 - 23.7 69.4 W-97 6.47 70 95.9 + 4 84.5 + 4 (b 63.5 -10 550 93.2 + 6 74.0 + 2 22.6 -5 70.0 +1 "Change relative to unirradiated.

(b)See footnote Table 5-2.

ZO ISE Z

Table 7-3. Observed Vs. Predicted Changes for Capsule W-97 Irradiated Charpy Impact Properties - 6.47 x 10 n/cm 2 (E > 1 MeV)

Difference Predicted Per R.G. 1.99/2 Observed Without With Material Unirrad. Irrad. Diff. Marginsa) Margin(b)

Increase in 30 ft-lb Trans. Temp.. F Base Material (M-1004-2)

Longitudinal 0 +3 + 3 18 35 Transverse - 29 +7 +36 18 35 Heat-Affected Zone (M-1004-2) -106 -90 +16 18 35 Weld Metal (88114/0145) - 80 -44 +36 39 78 co Decrease in Charpy USE. ft-lb Base Material (M-1004-2)

Longitudinal 170 154 -16 N.A. 19(c)

Transverse 141 123 -18 N.A. 16"c Heat-Affected Zone (M-1004-2) 170 156 -14 N.A. 19(c)

Weld Metal (88114/0145) 156 143 -13 N.A. 26(cl Vj le 2m (a)Mean value per Regulatory Guide 1.99, Revision 2, May 1988.

1!

n;S (b)Mean value per Regulatory Guide 1.99, Revision 2, May 1988, plus margin.

n (C)Bounding value per Regulatory Guide 1.99, Revision 2, May 1988 (includes margin).

a I33 N.A. - Not applicable.

1%

r- - - -e ~ r~r r ~ r"r ~ r

I Ii- I I-- I I I I I I I I I I I I Table 7-4. Evaluation-of Reactor Vessel End-of-Life (32 EFPY) Fracture Touqhness - Waterford Unit 3 Material Chemical Estimated EOL Fluencelbe End-of-Life RTNDT, Fe4 Material Descriotion Composition, Inside T/4 Wall Fab. Mat'l. Reactor Vessel Heat C W/oop ei Surface Location Initial Inside T/4 Wall Code Beltline Location Number'l Type Copper Nickel n/cm 2

n/cm2 RTHOT' Surface Location M-1003-1 Intermed. Shell 56488-1 SA533, Gr. B 0.02 0.71 3.64E+19 1.95E+19 -30 +23 +17 M-1003-2 Intermed. Shell 56512-1 SA533, Gr. B 0.02 0.67 3.64E+19 1.95E+19 -50 + 3 -3 M-1003-3 Intermed. Shell 56484-1 SA533, Gr. B 0.02 0.70 3.64E+19 1.95E+19 -42 +11 + 5 M-1004-1 Lower Shell 57326-1 SA533, Gr. B 0.03 0.62 3.69E+19 1.97E+19 -15 +39 +32 M-1004-2 Lower Shell 57286-1 SA533, Gr. B 0.03 0.58 3.69E+19 1.97E+19 +22 +76 +69 1M-1004-3 Lower Shell 57359-1 SA533, Gr. B 0.03 0.62 3.69E+19 1.97E+19 -10 +44 +37 101-171 Mid. Circum. Weld WW88114/ ASA Weld/ 0.05 0.16 3.64F+19 1.95E+19 -70 +45 +34 FL0145 Linde 0091 101-124-A,-B,--C Intermed. Longit. Weld CE Lots MMA Weld/ 0.02 0.96 3.64E+19'd' 1 .95E+19'0 -60 +12 + 4 .

BOLA, HODA Type 8018 101-142-A,-B,-C Lower Longit. Weld WW83653/ ASA Weld/ 0.03 0.20 3.69E+19'"O 1 .97E+19'4 -80 +14 + 3

'0 FL3536 Linde 0091

"'Per Regulatory Guide 1.99, Revision 2, May 1988.

'"'Per Section 6 of this report using neutron transport calculation methods.

d'Materlals chemical compositions per response to Generic Letter 92-01.

Fluence value for longitudinal weld with maximum value.

"'Per response to Generic Letter 92-01.

Table 7-5. Evaluation of Reactor Vessel End-of-Life (32 EFPY) URper-Shelf Energy - Waterford Unit 3 Material Estimated Chemical EOL Fluencefw Estimated EOL-USE Estimated EFPY to Material Description Composition, Inside T/4 Wall Inltial Per RG 1.99/2"' 50 ft-lbs Fab. Mat'l. Reactor Vessel Heat w/opp USE Inside T/4 Wall Inside T/4 Wall Code Beltline Location NumbertI1 Type nLcm" nLcm2 fl:Ikbs Surface Location Surface Location M-1003-1 Intermed. Shell 56488-1 SA533, Gr. B 0.02 0.71 3.64E+19 1.95E+19 94))) 79 81 >32 >32 M-1003-2 Intermed. Shell 56512-1 SA533, Gr. B 0.02 0.67 3.64E+19 1.95E+19 97"') 81 84 >32 >32 M-1003-3 Intermed. Shell 56484-1 SA533, Cr. B 0.02 0.70 3.64E+19 1.95E+lg 76 78 >32 >32 M-1004-1 Lower Shell 57326-1 SA533, Gr. B 0.03 0.62 3.69E+19 1.97E+19 106"' 87 90 >32 >32 M-1004-2 Lower Shell 57286-1 SA533, Gr. B 0.03 0.58 3.69E+19 1.97E+19 94))) 78 80 >32 >32 M-1004-3 Lower Shell 57359-1 SA533, Gr. B 0.03 0.62 3.69E+19 1.97E+19 94u) 78 80 >32 >32 101-171 Mid. Circum. Weld WW88114/ ASA Weld/ 0.05 0.16 3.64E+19 1.95E+19 156k' 115 122 >32 >32 FL0145 Linde 0091 101-124-A,-B,-C Intermed. Longit. Weld CE Lots MMA Weld/ 0.02 0.96 3.64E+19'm~ 1.95E+1940 N.A. >50 >50 >32 >32 BOLA, HODA Type 8018 101-142-A,-B,-C Lower Longit. Weld WW83653/ ASA Weld/ 0.03 0.20 3.69E+19'd' 1.97E+19'4 N.A. >50' >50 >32 >32 FL3536 Linde 0091 "Per Regulatory Guide 1.99, Revision 2, May 1988.

INPer Section 6 of this report using neutron transport calculation methods.

"'Materials chemical compositions per response to Generic Letter 92-01.

WiFluence value for longitudinal weld with maximum value.

"'Per response to Generic Letter 92-01.

"'Based on 0.65 of longitudinal upper-shelf energy data.

'Estimate based on value given for the surveillance weld metal fabricated with the same weld wire and Linde 0091 weld flux.

t-.

Z3 -

S.

tnto 21h M

n;!

nn cJ P-R9 1%

_- I I I 0_

[-, - t I I - I I I I I III I I.I Table 7-6 Evaluation of Reactor Vessel End-of-Life Pressurized Thermal Shock Criterion - Waterford Unit 3 Materi al Estimated Chemical Inside PTS Eval uation, F11 Material Description Composition, Surface Inside Reactor Vessel Heat W/O EOL Fluence Initial Surface Scri ?ening Beltline Location Number"' Type Copper Nickel n/cm2 RTMOTO . RTT. Cril teria Intermed. Shell 56488-1 SA533, Gr. B 0.02 0.71 3.64E+19 -30 +31 2'70 Intermed. Shell 56512-1 SA533, Cr. B 0.02 0.67 3.64E+19 -50 +11 2'70 Intermed. Shell 56484-1 SA533, Gr. B 0.02 0.70 3.64E+19 -42 +19 2'70 Lower Shell 57326-1 SA533, Gr. B 0.03 0.62 3.69E+19 -15 +46 2:70 Lower Shell 57286-1 SA533, Gr. B 0.03 0.58 3.69E+19 +22 +83 270 Lower Shell 57359-1 SA533, Gr. B 0.03 0.62 3.69E+19 -10 +51 2270 Mid. Circum. Weld WW88114/ ASA Weld/ 0.05 0.16 3.64E+19 -70 +45 3300 FL0145 Linde 0091 Intermed. Longit. Weld CE Lots MMA Weld/ 0.02 0.96 3.64E+19'd1 -60 +32 270 BOLA/HODA Type 8018 Lower Longit. Weld WW83653/ ASA Weld/ 0.03 0.20 3.69E+19'd' -80 +23 270 t-4 FL3536 Linde 0091

"'Per IOCFRSO, Section 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events."

IbiPer Section 6 of this report using neutron transport calculation methods.

Mater/als chemical compositions per response to Generic Letter 92-01.

"dFluence value for longitudinal weld with maximum value.

"'Per response to Generic Letter 92-01.

bV3

a-L Figure 7-1.

• Comparison of Unirradiated and Irradiated L

Charpy Impact Data Curves for Plate Material 100 Lonqitudinal Orientation. Heat No. M-1004-2 L

C.)

a; 75 L 50 U,

25 L

0 0.10 L

d= 0.08 En a

L 21 uJ CZ 0.06 Z 0.04 L

-J z

u 0.02 L 0

220 I I I I I I L

200 _-

180 -

A = 16ft-lbs L

.: 160 H Unirradiated I

0.L 0

Cm 0

140 _

120 _

I cCa uJ L"

100 -

Fluence: 6.47 x 1o18 n/cm2 I

80 _-

I.

Cu 0.

E 60 -

3F 40 _

3F I.

20 MATERIAL SA-533.CIB1(L)

L.

U-100 I

0 100 I

200 I

300 HEAT NO.

I 400 M-1004-2 I

500 600 I

Test Temperature, F I

7-12 MYIESWENUCLEAR 13 W SERVICE COMPANY l

Figure 7-2. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Plate Material Transverse Orientation. Heat No. M-1004-2 100

Q 75

=

'u-M 50 A..

25 0

0.10

!: oE 0.08

(i

. M

a. 0.06

, LU _

00.04 c' 0.02 0

0 220 I I I I I 200 _-

i80 _

• .0 160 _

140 _- Unirradiated .

a.

• 0 120 a.

100 Fluence: 6.47 x 1018 n/cm2 80 _

60 _

32F 40 _

36F 20 _- MATERIAL SA-533.CIB1i(T HEAT NO. M-1004-2

_ I l l l l n

U- 0

-100 0 100 . 200 300 400 500 600 Test Temperature, F 7-13 13WSBEVWINUCECANY

-3L-L Figure 7-3. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Base Metal, Heat-Affected-Zone. Heat No. M-1004-2 100 75 C-,

L. 50 C,z 25 0

/

0.10 CF 0

0.08 CL 0.06 C;

0.04

-j 0

0.02 z

0 220 I I I I I 200 180

/A = 14ft-lbs

.0 160 Unirradiated CF 140 0.

0

.0 120 ci 100 w

LU tu 80 Fluence: 6.47 x 1018 n/cm2 03.

E 60 18F 40 16F 20 MATERIAL SA-533.CIB1 (HAZ) -

HEAT NO. M-1004-2 I I I I I 0

-200 -100 0 100 200 300 400 500 Test Temperature, F 7-14

Figure 7-4. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Weld Metal 88114/0145 100

- 75 -

Unirradiated

• 50 - Fluence: 6.47 x 1018 n/cm2

.~25 co U,

• 0.10 CE0.08 Unirradiated

_ 10.0 LU

~ / X tFluence: 6.47 x 1018 n/cm 2 0.04 3F

.~0.02 0

,, 1601 _ _

220 200 -

1807-

.~160 -

40 Unirradiated lI36 Iq 0:F 140-co 120-C- 100 Fluence: 6.47 x 1018 n/cm 80 E

6032 40 20 MATERIAL WELD METAL HEAT NO. 8811410145 0-200 .-100 0 100 200 300 400 500 Test Temperature, F 7-15

Page Intentionally Left Blank J111W WNUCLEAR I3SWSER VICE COMPANY

8.

SUMMARY

OF RESULTS The analysis of the reactor vessel material contained in the first surveillance capsule (Capsule W-97) removed for evaluation as part of the Waterford Generating Station Unit No. 3 Reactor Vessel Surveillance Program, led to the following conclusions:

1. The capsule received an average fast fluence of 6.47 x 1018 n/cm2 (E >

1.0 MeV). The predicted fast fluence for the reactor vessel T/4 location at the end of the fourth fuel cycle is 2.74 x 1018 n/cm2 (E >

1 MeV).

2. The fast fluence of 6.47 x 1018 n/cm2 (E > 1 MeV) increased the RTNDT of the capsule reactor vessel core region shell materials by a maximum of 40F.

3. Based on the calculated fast flux at the vessel wall, an 80% load factor and the planned fuel management, the projected fast fluence that the Waterford Generating Station Unit No. 3 reactor pressure vessel inside surface will receive in 40 calendar year's operation is 3.69 x 1019 n/cm2 (E > I MeV).

4. The increase in the RTNDT for the transverse oriented shell plate material was in poor agreement with that predicted by the currently used design curves of RTNDT versus fluence (i.e., Regulatory Guide 1.99, Revision 2).

5. The increase in the RTNDT for the weld metal was in good agreement with that predicted.

6. Neither the base metal nor the weld metal upper-shelf energies at the T/4 location, based on surveillance capsule results, are predicted to decrease below 50 ft-lbs prior to 32 EFPY.

7. The current techniques (i.e., Regulatory Guide 1.99, Revision 2) used to predict the change in the base metal and the weld metal RTNDT proper-ties due to irradiation are conservative except for the base metal transverse properties.

8. The current techniques (i.e., Regulatory Guide 1.99, Revision 2) used to predict the change in the base metal and the weld metal Charpy 8-1 B NUCLEAR RI'5E1RV E 5JW SERVICE COMPANY

upper-shelf properties due to irradiation are in good agreement with the base metal and conservative for the weld metal.

9. The analysis of the neutron dosimeters demonstrated that the analytical techniques used to predict the neutron flux and fluence were accurate.

1 1

8-2

• SUERMI"C'E CMPANY]

9. SURVEILLANCE CAPSULE REMOVAL SCHEDULE Based on the post-irradiation test results of Capsule W-97 and the recommended withdrawal schedule of Table 1 of E18515 the following schedule is recommended for the examination of the remaining capsules in the Waterford Generating Station, Unit No. 3 RVSP:

Evaluation Schedule (a)

Capsul e Location of Lead Removal Expected Capsule Identification Capsul esfa) Factorb Time Fluence (n/cm2 )(,I W-83 830 1.26 15 EFPY 2.19 x 10' W-104 1040 0.81 Spareld) (2.99 x 10 9)

W-263 2630 1.26 26 EFPY 3.69 x 0"'

W-277 2770 1.26 Spare(d) (4.65 x 1019)

W-284 2840 0.81 Spare d) (2.99 x 1019)

Reference reactor vessel irradiation locations, Figure 3-1.

(b)The factor by which the capsule fluence leads the vessels maximum inner wall fluence.

("Estimated fluence values based on current fuel cycle designs.

(d)Spare capsule to be irradiated and available for an intermediate evaluation, if data needed, to support licensing requirements or provide data for license renewal. Capsule withdrawal at 32 EFPY will have estimated fluence as defined in brackets (.

9-1 1 ABW NUCLEAR l WSERNVCE COMPAN Y

10. CERTIFICATION The specimens were tested, and the data obtained from Entergy Operations, Inc.,

Waterford Generating Station, Unit No. 3, reactor vessel surveillance Capsule W-97 were evaluated using accepted techniques and established standard methods and procedures in accordance with the requirements of 10CFR50, Appendixes G and H.

This report has been reviewed for technical content and accuracy.

///fl/92 M.-J. D#en (Material Analysis) Date M&SA Unit 114 1r/

L. Petrush (Fluence Analysis) I Date Performance Analysis Unit Verification of independent review.

9~VQscalovl Po 1E Rm4Qp III 1-447--"

k. E. Moore, Manager Date M&SA Unit This report is approved for- release.

T. L. Baldwin, P.E. Date Program Manager 10-1 IMSIZEEIN&YEV2005ANY

BAW-2177-01 ll REVISION 1 CERTIFICATION L

The revision to the document is technically accurate and conforms to accepted techniques, established standard methods and procedures in accordance with the L

requirements of 10 CFR 50, Appendices G and H.

L L

J. B. Hall/ (Materials Analysis) Date Materials & Structural Analysis Unit L

This report has been reviewed for technical content and accuracy.

L L

B. R. Grambau (Materials Analysis) Date Materials & Structural Analysis Unit L

Verification of independent review.

L A D. McKim, Manager Date I

Materials & Structural Analysis Unit I

This report is approved for release.

• SFW 2&1/S I

v. R. Gray Date Program Manaae *1 I

I I

I A

10-2 FRAMATOME ANP I

APPENDIX A Reactor Vessel Surveillance Program Background Data and Information A-1 IWS ERMVICE COMPANY

1. Material Selection Data L The data used to select the materials for the specimens in the surveillance L program, in accordance with E185-73, are shown in Table A-1. The locations of these materials within the reactor vessel are shown in Figures A-1 through A-4.

2. Definition of Beltline ReQion The beltline region of Waterford Unit 3 was defined in accordance with the [

definition given in ASTM E185-73.

3. Cansule Identification L The capsules used in the Waterford Unit 3 surveillance program are identified below by identification, location, and original target fluence.' L Capsule Removal Capsule Identification Capsule Location(

Approximate Refueling Target Fluence, n/cm2 L 1 W-97 970 7 6.0 x 10'8 2 W-104 1040 19 1.6 x 10'9 L 3 W-284 2840 30 2.5 x109 L 4 W-263 2630 Standby ---

5 W-277 2770 Standby --- I 6 W-83 830 Standby ---

4. Specimens Per Surveillance Capsule The type and quantity of each material contained in each surveillance capsule is L shown in Table A-2.

A-2 I WIERNE COMPANY

I - I- -- I-- [" - - I-----, I - I- I I I I I I I I I Table A-1. Unirradiated Impact Properties and Residual Element Content Data of Beltline Region Materials Used for Selection of Surveillance Program Materials - Waterford Unit No. 321.22.23 Charpy Impact Data, Longitudinal Fabricator Material 30 50 35 Material Ident., Beltline Drop wt ft-lb, ft-lb, MLE, USE, RTNOT Chemistry. wt%

Code Heat No. Material Type Region Location TNDT, F F F F ft-lb F Cu Ni P S M-1003-1 56488-1 SA533, Gr. B Intermed. Shell -30 -30 -10 -10 144 -30 0.02 0.71 0.004 0.010 M-1003-2 56512-1 SA533, Gr. B Intermed. Shell -50 -55 -12 -15 149 -50 0.02 0.67 0.006 0.007 M-1003-3 56484-1 SA533, Gr. B Intermed. Shell -50 -22 - 2 -10 138 -42 0.02 0.70 0.007 0.009 M-1004-1 57326-1 SA533, Gr. B Lower Shell -50 +10 +25 +20 163 -15 0.03 0.62 0.006 0.008 7, 1 M-1004-2 57286-1 SAS33, Gr. B Lower Shell -20 +37 +62 +55 144 22 0.03 0.58 0.005 0.005 M-1004-3 57359-1 SA533, Gr. B Lower Shell -50 +12 +30 +25 145 -10 0.03 0.62 0.007 .0.007 C',I -70 --- --- --- --- -70 0.05 0.16 0.008 0.008 101-171 88114/0145 ASA Weld/Linde 0091 Middle Circum.

101-124-A,-B,-C BOLA/HODA MMA Weld/Type 8018 Intermed. Longit. -60 --- --- --- --- -60 0.02 0.96 0.010 0.016 101-142-A, -B, -C 83653/3536 ASA Weld/Linde 0091 Lower Longit. -80 --- --- -- --- -80 0.03 0.20 0.007 0.009 ft'

Table A-2. TYDe and Quantity of Specimens Contained in Each Irradiation Capsule Assembly Base Metal Weld Metal Correl.

Target (Heat No M-1004-2) (88114/0145) c) HAZ (Heat Material b)

Capsule Fl uence(&) Impact No. M-1004-2) Total Specimens Location (n/cm2) L T Tensile Impact Tensile Impact Tensile Impact Impact Tensile Vessel 970 6.0 x 1018 12 12 3 12 3 12 3 48 9 Vessel 104° 1.6 x 10 -- 12 3 12 3 12 3 12 48 9 Vessel 2840 2.5 x 1019 12 12 3 12 3 12 3 48 9 Vessel 2630 Standby -- 12 3 12 3 12 3 12 48 9 Vessel 2770 Standby 12 12 3 12 3 12 3 48 9 I:P Vessel 830 Standby 12 12 3 12 3 12 3 48 9 TOTALS 48 72 18 72 18 72 18 24 288 54 (a)Adjusted to nearest value attainable during scheduled refueling.

(Reference material correlation monitors.

"c'Weld wire/weld flux lot combination.

V3 L = Longitudinal En. T = Transverse MR, a!

- P

Figure A-1. Location and Identification of Materials Used in the Fabrication of Waterford Unit 3 Reactor Pressure Vessel REACTOR VESSEL BELTLINE MATERIALS

-NOT SHOWN INTERMEDIATE SHELL WELD SEAM No. 101-124B WELD SEAM No. 101-124C PLATE No. M-1003-2 LOWER SHELL

- # WELD SEAM No. 101-142B WELD SEAM No. 101 -142C PLATE No. M-1004-2 42" ID OUTLET J 30" ID INLET NOZZLE NOZZLE UPPER TO INTERMEDIATE INTERMEDIATE SHELL SHELL GIRTH SEAM LONGITUDINAL WELD WELD No. SEAM No. 101-124A INTERMEDIATE SHELL - INTERMEDIATE SHELL PLATE No. M-1003-3 PLATE No. M-1003-1 INTERMEDIATE-TO-LOWER SHELL GIRTH SEAM WELD No. 101-171 LOWER SHELL PLATE No. M-1004-1 LOWER SHELL PLATE -

No. M-1004-3 LOWER SHELL LONGITUDINAL WELD SEAM No. 101-142A A-5 WEESW NUCLEAR WSERVICE COMPAA/Y

a-L Figure A-2. Location of Beltline Region Materials in L

Relationship to the Reactor Vessel Core L

L L

I I

co I a) LO II I

> a . I I

I I1 I

• = Centerline of Weld I1 I

A-6 5WWNUCAAN .1

Figure A-3. Location of Longitudinal Welds in Waterford Unit 3 Upper and Lower Shell Courses 0O M-1 003-2 101-124C

-101-124A 90 270 M-1 003-3 101-124B 180 0

101-142C

-101-142A

- 90 270 M-1004-2 101-142B D04-1 180

.. ...-. .A-7 IIBWNUCLE4R JRWSEER VICE CO0mMPAN

3a-I Figure A-4. Location of Surveillance Capsule Irradiation Sites in Waterford Unit 3 180° t C ' Outlet Nozzle

//' I\ 1.

Vessel 1 1040 \Inl

\NozzleI

{ /\ // / Core Shroud i ,A

> \A/ eg~Core Support Barrel He9 k / _ N \<' essel Vess Reactor Vessel 2630 1 i / . . a essel'1

'V830 e~ssel ti 2770r Vessel 284 I

II 00 A-8 I3WSERVICE III WNUCLE4R COMPANY

APPENDIX B Pre-Irradiation Tensile Data B-1 I WSERVICE COMPANY

L ii Table B-1. Tensile Properties of Unirradiated Shell Plate I

Material, Heat No. M-1004-2. Loncitudinal Test Reduction Elongation, I Specimen Temp, Strength. ksi Fracture Fracture. ksi of Area, Total/Unif.

No. F Yield* Ultimate Load. lb Strength Stress  % I I-1J2 71 68.6/66.7 88.5 2640 53.9 189 71.4 29/11.3 IJi 1K2 71 71 70.4/67.4 70.0/68.6 88.4 90.1 2700 2700 55.1 55.1 180 193 69.4 71.4 27/11.3 30/11.7 I IJA 250 63.7/63.1 82.5 2640 53.9 176 69.4 24/ 9.2 1K3 250 66.1/64.9 84.1 2640 53.9 176 69.4 24/ 9.3 I 1JL 250 63.7/63.1 83.3 2700 55.1 180 69.4 26/ 9.3 1J6 WJC 550 550 63.7/----

63.1/----

85.5 85.9 2700 2700 55.1 55.1 193 208 71.4 73.4 23/ 9.3 26/ 9.8 I

1J3 550 62.5/ ---- 85.6 2760 56.3 173 67.3 25/10.2 I

• Lower and upper yield strengths.

I Table B-2. Tensile Properties of Unirradiated Shell Plate Material. Heat No. M-1004-2. Transverse I

Test Reduction Elongation, Specimen No.

Temp, F

Strength. ksi Yield* Ultimate Fracture Fracture, ksi Load. lb Strength Stress of Area, Total/Unif. I 2KC 71 69.2/68.6 89.7 2880 58.8 192 69.4 27/10.8 I 2KT 71 68.4/67.2 88.4 2820 58.8 188 70.0 29/11.3 2KB 2KD 71 250 69.8/68.6 65.5/64.3 89.0 83.7 2940 2700 60.0 55.1 196 180 65.3 69.4 26/10.8 23/ 9.7 I.

2JE 2L2 250 250 64.6/64.6 64.9/64.3 83.9 82.3 2820 2940 57.6 60.0 188 163 69.4 63.3 21/ 9.3 23/ 9.3 I 2J7 550 64.9/ ---- 87.2 2880 58.8 169 65.3 23/10.2 2KP 550 63.7/ --- 86.9 3000 61.2 188 67.3 22/ 9.8 I1 2KU 550 64.9/ --- 87.0 2880 58.8 160 63.3 22/ 9.8 l1

• Lower and upper yield strengths.

-I I

B-2 I

Table B-3. Tensile Properties of Unirradiated Shell Plate HAZ Material. Heat No. M-1004-2. Transverse Test Reduction Elongation, Specimen Temp, Strength, ksi Fracture Fracture, ksi of Area, Total/Unif.

No. F Yield* Ultimate Load, lb Strength Stress 4KT 71 71.0/68.6 91.0 2820 57.6 188 69.4 22/ 7.3 4JJ 71 68.6/68.0 90.9 2820 57.6 188 69.4 21/ 7.0 4K4 71 69.8/68.0 91.9 2820 57.6 188 69.4 21/ 6.2 4KP 250 64.3/63.7 84.3 2640 53.9 176 69.4 21/ 5.4 4J5 250 63.1/63.1 84.0 2640 53.9 165 67.3 19/ 5.8 4KE 250 63.7/63.7 84.8 2640 53.9 176 69.4 21/ 5.4 4JT 550 60.6/60.0 86.3 2820 57.6 166 65.3 21/ 6.7 4JE 550 63.1/61.8 86.9 2820 57.6 176 67.3 20/ 6.8 4J4 550 60.0/58.7 86.9 2820 57.6 176 67.3 20/ 6.3

• Lower and upper yield strengths.

Table B-4. Tensile Pronerties of Unirradiated Weld Metal 88114/0145 Test Reduction Elongation, Specimen Temp, Strength. ksi Fracture Fracture. ksi of Area, Total/Unif.

_ No. F Yield* Ultimate Load, lb Strength Stress 3KE 71 85.7/82.0 92.9 2760 56.3 184 69.4 27/ 9.3 3J3 71 84.5/80.2 91.6 2760 56.3 197 71.4 27/ 9.3 3K1 71 84.5/80.8 92.0 2700 55.1 193 71.4 29/10.2 3L2 250 79.6/75.9 86.9 2700 55.1 180 69.4 22/ 7.7 3J5 250 80.2/75.9 87.2 2760 56.3 197 71.4 21/ 7.8 3JC 250 79.6/73.5 85.4 2580 52.6 184 71.4 23/ 7.3 3K4 550 72.2/ ---- 88.8 2580 52.6 172 69.4 24/ 9.5 3KA 550 72.2/---- 88.2 2640 53.9 189 71.4 24/ 9.3 3KU 550 72.2/---- 87.9 2820 57.6 176 67.3 23/ 8.8

- *Lower and upper yield strengths.

B-3 111BCWVNUCLEAR IWS ER VICE COMPANIY

iL L

L L

L L

Page Intentionally Left Blank L

3WN NURy L

APPENDIX C Pre-Irradiation Charpy Impact Data C-1 INIIBEW NICUCLEAR IJWSERUICE COMPANY

U L

Table C-1. Charpy Impact Data From Unirradiated Base Material, L

Lonaitudinal Orientation. Heat No. M.-1004-2 Absorbed Lateral Shear L Specimen Test Temp, Energy, Expagsion, Fracture, ID F ft-lb 10 in.

L 13T -80 7.0 3 0 156 152 15J

• -40

-40 0

12.0 11.5 14.5 10 10 14 10 10 15 L

123 0 48.5 41 20 112 lip 40 40 86.0 105.0 65 78 40 65 L

153 80 107.0 72 75 11A 12D 14P 80 120 120 130.0 131.0 147.0 90 89 92 80 85 90 L

147 1iM 137 140 160 160 158.0 169.5 177.5 91 94 90 100 100 100 L

127 210 168.5 95 100 126 210 175.0 90 100 L Table C-2. Charpy Impact Data From Unirradiated Base Material, Transverse Orientation. Heat No. M-1004-2 L

Specimen Test Temp, Absorbed Energy, Lateral Expagsion, Shear Fracture, l

ID F ft-lb 10 in.

264 -80 9.0 7 0 I

21L -60 10.0 6 0 22E 23B

-40

-40 20.5 28.5 18 23 10 10 L 25P 0 44.0 37 20 26B 21K 23L 40 40 0 65.5 68.5 73.0 50 55 57 25 40 60 I

26A 80 118.0 82 75 23A 212 80 120 130.0 123.5 77 81 75 90 I

255 120 141.0 90 100 21P 254 160 160 136.0 138.5 88 92 100 100 I 25T 210 143.5 88 100 23C 210 145.0 90 100 I

I C-2 f*;I6WWNUCLER

/WERNIE COMPANY I

Table C-3. Charpy Impact Data from Unirradiated Base Metal, Heat-Affected Zone. Heat No. M-1004-2 Absorbed Lateral Shear Specimen Test Temp, Energy, Expansion, Fracture, ID F ft-lb 10 in.

417 -150 6.5 9 0 44C -135 10.5 8 0 451 -120 21.5 17 10 42J - 80 44.5 33 30 46T - 80 76.5 50 45 45P - 40 113.5 73 70 443 - 40 116.5 77 60 466 0 118.0 77 75 46D 0 138.0 88 80 42M 40 126.0 84 75 47T 40 162.0 88 100 45C 80 163.5 91 100 43M 80 177.0 84 100 41K 120 152.5 90 100 45D 120 183.5 89 100 427 160 164.5 87 100 41A 160 183.5 89 100 Table C-4. Charnv Imnact Data from Unirradiated WPMd Metal 8R114/0145 Absorbed Lateral Shear Specimen Test Temp, Energy,. Expansion, Fracture, ID F ft-lb 10 in.

354 -180 3.5 4 0 316 -150 5.5 3 0 36A -120 8.0 6 10 36D - 80 13.5 12 20 31D - 80 45.5 36 30 32A - 40 83.5 61 50 313 - 40 96.5 69 75 357 0 122.5 80 80 341 0 130.5 95 85 31A 40 142.0 95 - 90 344 40 149.0 97 100 33T 80 146.0 94 100 32M 80 158.0 96 100 37L 120 155.5 96 100 32B 120 162.5 97 100 343 160 148.5 94 100 36J 160 171.0 94 100 C-3 SIUIEB5WNUCLEAR IRWSERI/CE COMPANY

AL-I Figure C-1. Charpy Impact Data From Unirradiated Base Metal I

(Plate). Lonqitudinal Orientation. Heat No. M-1004-2 100 C) 75 U

U-50 cu CD U,

25 0

0.10 0

C= 0.08 0

C,'

R. 0.06 LU X 0.04

-J 2 0.02 0

0 220 II I l I I I

- DATA

SUMMARY

200 - TNDT OF TCV (35 MLE) +12F 1801 - TCY (50 FT-LB) +27F i TCV (30 FT-LB) OF a

160 1- Cv-USE (AVG.) 170ft-lbs RTNIT OF 140 co

.0 120 I

>1 Cl 100I LU Ct 80 I E

60 1 40 I MATERIAL SA-533.CIB1(L) 20 1 FLUENCE None -

HEAT NO. M-1004-2 I I I n

-100 0 100 200 300 400 500 600 Test Temperature, F I

C-4

/I3!&~W NUCUA SM S"ERVI CE COMPANY I

Figure C-2. Charpy Impact Data From Unirradiated Base Metal (Plate), Transverse Orientation. Heat No. M-1004-2 100 I.

75 C,,a 50 25 0

0.10

cF 0.08 0

23 0.06 LU w 0.04

-I 2 0.02 0

z 0

220 I I

-DATA

SUMMARY

200 [- TNDT -20F TCV (35 MLE) -13F 180 1- Tcv (50 FT-LB) OF Tcv (30 FT-LB) -29F 160 - CV-USE (AVG.) 141ft-lbs RTNDT -20F 140 [

-0 S 120 [

.0 C) 1001-CD)

CC 80 [

0.E 60 I 40 [

MATERIAL SA-533,CIB1 (T) 20 FLUENCE None HEAT NO. M-1004-2 I ____

U-100 0 0 100 . 200 300 400 500 600 Test Temperature, F C-5 13W0W NUCLEAR IMWSERVICE COMPANY

I Figure C-3. Charpy Impact Data From Unirradiated Heat-Affected-Zone Base Metal, Heat No. M-1004-2 100 75 U- 50 Cu 25 0

0.10 C~

0.08 0

.(A 0.

x 0.06 LU 0.04

-J 0.02 0

z 0

220 _

I I I I I I

-DATA

SUMMARY

200 - TNOT -50F TCv (35 MLE) -87F 0 6 180 TCV (50 Fr-LB) -88F TCv (30 FT-LB) 106F 160 CV-USE (AVG.) 170ft-lbs

• 0

.0 RTNOT -50F /

e 140 0 a

I 0.

120 0 00 EU 100 C

80 E

60 0 40 MATERIALSA-5 3 3 .CIBI (HAZ) 20 FLUENCE None -

HEAT NO. M-1004-2 I I I I I 0

- I200 .100 0 100 200 300 400 50(

Test Temperature. F C-6 3W!! W NUCUAR RUWSVERVICE COMPANY

Figure C-4. I. Charpy Impact Data From Unirradiated Weld Metal. 88114/0145 100 75

. 50

' 25 0

0.10 C: 0.08 G

x 0.06 LU 3 0.04 2 0.02 2

z I

220 I I I I I I

- DATA

SUMMARY

200 TNDT -80F TCV (35 MLE) -69F 180 Tcv (50 FT-LB) -67F TcV (30 FrTLB) _ 80F 0 s 160 Cv-USE (AVG.) 156ft-lbs 0A RTNDT -80F

• 0 c 140 0

° 120 100 mu t'Y 80 E

60 40 MATERIAL Weld Metal 20 FLUENCE None -

HEAT NO. 88114/0145 I I I I 0,- I a00 I

• 100 0 100 200 300 400 500 Test Temperature, F C-7 UJSB WNUCLEAR WSER VICE COMPANY

L L

L L

I-L L

L L

Page Intentionally Left Blank I I

I I3 W B6W NUCLEAR SERV/ICE COMPANY

APPENDIX D Fluence Analysis Methodology D-1 IRW SEVICE COMPANY

1. Analytical Method A semi-empirical method is used to calculate the capsule and vessel flux. The [

method employs explicit modeling of the reactor vessel and internals and uses an average core power distribution in the discrete ordinates transport code DOTIV, l version 4.3. DOTIV calculates the energy and space dependent neutron flux for the specific reactor under consideration. This semi-empirical method is conven- L iently outlined in Figures D-1 (capsule flux) and D-2 (vessel flux).

The two-dimensional transport code DOTIV was used to calculate the energy- and space-dependent neutron flux at all points of interest in the reactor system.

DOTIV uses the discrete ordinates method of solution of the Boltzmann transport equation and has multi-group and asymmetric scattering capability. The reference L calculational model is an R-E geometric representation of a plan view through the reactor core midplane which includes the core, core liner, coolant, core barrel,,

thermal shield, pressure vessel, and concrete. The material and geometry model, represented in Figure 0-3, uses one-eighth core symmetry. In order to include reasonable geometric detail within the computer memory limitations, the code parameters are specified as P3 order of scattering, S8 quadrature, and 47 energy groups. The P3 order of scattering adequately describes the predominately forward scattering of neutrons observed in the deep penetration of steel and water media, as demonstrated by the close agreement between measured and calculated dosimeter activities. The SB symmetric quadrature has generally produced accurate results in discrete ordinates solutions for similar problems, and is used routinely in the B&W R-O DOT analyses.

Flux generation in the core was represented by a fixed distributed source which the code derived based on a combined 235U and 239Pu fission spectrum, the input relative power distribution, and a normalization factor to adjust flux level to the desired power density.

Geometrical Configuration For modeling purposes, the actual geometrical configuration was divided into three parts, as shown in Figure D-3. The first part, Model "A," was used to l generate the energy-dependent angular flux at the inner boundary of Model "B,"

which began at the inner surface of the core barrel. Model A included a detailed D-2 I3 W5MEEMVICE BVW NUCLOAR COMPANY

representation of the core baffle (or liner) in R-0 geometry that has been checked for both metal thickness and total metal volume to ensure that the DOT approximation to the actual geometry was as accurate as possible for these two very important parameters. The second, Model B, contained an explicit represen-tation of both surveillance capsules and associated components for the'applicable time periods. The B&W Owners Group's Flux Perturbation Experiment" verified that the surveillance capsule must be explicitly included in the DOT models used for capsule and vessel flux calculations in order to obtain the desired accuracy.

Detailed explicit modeling of the capsule, capsule holder tube, and internal components were therefore incorporated into the DOT calculational models. The third, Model "C," was similar to Model B except that no capsule was included.

Model C was used indetermining the vessel flux in quadrants that did not contain a surveillance capsule; typically these quadrants contain the azimuthal flux peak on the inside surface of the reactor vessel.

An overlap region of approximately 52.95 cm was specified between Model A and Models B or C. The width of this overlap region, which was fixed by the placement of the Model A vacuum boundary and the Model B boundary source, was determined by an iterative process that resulted in close agreement between the overlap region flux as predicted by Models'A and B or C. The outer boundary was placed sufficiently far into the concrete shield (cavity wall) that the use of a "vacuum' boundary condition did not cause a perturbation in the flux at the points of interest.

MacroscoDic Cross Sections Macroscopic cross sections, required for transport analyses, were obtained with the mixing code GIP. Nominal compositions were used for the structural metals.

Coolant compositions were determined using the average boron concentration over a fuel cycle and the bulk temperature of the region. The core region was a homogeneous mixture of fuel, fuel cladding, structure, and coolant.

The cross-section library presently used is the (47-neutron group and 20-gamma group) BUGLE coupled set. The dosimeter reaction cross sections are based on the ENDF/B5 library,-and are listed -in Table E-3. The measured and calculated dosimeters activities are compared injTable D-1. -

D-3 1 WSiHWC UCOMPANY

IL Distributed Source The neutron population in the core during full power operation is a function of neutron energy, space, and time. The time dependence was accounted for in the analysis by calculating the time-weighted average neutron source, i.e. the neutron source corresponding to the time-weighted average power distribution.

L The effects of the other two independent variables, energy and space, were accounted for by using a finite but appropriately large number of discrete iU intervals in energy and space. In each of these intervals the neutron source was assumed to be invariant and independent of all other variables. The space and I energy dependent source function can be considered as the product of a discretely expressed "spatial function" and a magnitude coefficient, i.e. L SVijg= [ PD] X (RPDijXg] L magnitude spatial t where:

Svrg - Energy-and space-dependent neutron source, n/cc-sec, l v/K Fission neutron production rate, n/w-sec, PD = Average power density in core, w/cc, L RPDU = Relative power density at interval (i j), unitless, Xg= Fission spectrum, fraction of fission neutrons having energy in group "g," l i - Radial coordinate index, -

j = Azimuthal coordinate index, g Energy group index.

The spatial dependence of the flux is directly related to the RPD. Even though l the entire (eighth-core symmetric) RPD was modeled in the analysis, only the peripheral fuel assemblies contributed significantly to the ex-core flux. The l axial average RPD distribution is calculated on a quarter-core symmetric basis -

for the entire capsule irradiation period. The time-weighted average RPD ]

D-4 I3 E WBSWNUCLEAR SERVICE COMPANY

distribution is used to generate the normalized space and energy dependency of the neutron source. Calculations for the energy and space dependent, time-averaged flux were performed for the midpoint of each DOT interval throughout the model'. Since the reference model calculation'produced fluxes in the R-e plane that averaged over the core height, an axial correction factor was-required to adjust these fluxes to the capsule elevation. This factor was calculated to be 1.08.

A pin-by-pin RPD was provided by the customer, and was subsequently used to produce the source for use in the DOTIV code.

1.1. Capsule Flux and Fluence Calculation As discussed above, the DOTIV code was used to explicitly model the capsule assemblies and to calculate the neutron flux as a function of energy within the capsules. The calculated fluxes were used in the following equation to obtain calculated activities for comparison with the measured data. The calculated activity for reaction product Di, in (pCi/gm) is:

3.Xl4) E D (3.7xlO')An E E Un(E) ¢~(E) P.F (I-e-i) e-A~Ji where:

N = Avogadro's number, An= Atomic weight of target material n, f,= Either weight fraction of target isotope in n-th material or the fission yield of the desired isotope, an(E) = Group-averaged cross sections for material n (listed in Table E-3)

+(E) = Group averaged fluxes calculated by DOTIV analysis, Fj = Fraction of full power during j-th time interval, tj A; = Decay constant of the ith isotope, T Sum of total irradiation time, i.e., residual time in reactor, and the wait time between reactor shutdown and counting times, D-5 W NUCLEAR 1WWSERVICE COMPANY

1 Tj = Cumulative time from reactor startup to end of j-th time period.

t,= Length of the j-th time period Adjustments were made to the calculated dosimeter activities to correct for the effects listed below:

238 Photofission adjustments to U dosimeter activities Axial correction factor to adjust for axial power distribution After making these adjustments the calculated dosimeter activities were used with the corresponding measured activities to obtain the measured to calculated activity ratios or flux normalization factors:

C= DI (measured)

Di (calculated) -

These normalization factors were evaluated, averaged, and then used to adjust the calculated test specimen flux and fluence for each capsule to be consistent with I the dosimeter measurements. The flux normalization factors are given in Table D-1. Note that the Co-60 dosimeters are typically not used in the determination of the final normalization factor to be applied to the calculated flux due to the fact that they do not respond in the regions of interest, E > 1.0 MeV and E > 0.1 MeV, and the thermal region is not accurately calculated in the DOT analysis.

2. Vessel Fluence ExtraDolation For past core cycles, fluence values in the pressure vessel were calculated as described above. Extrapolation to future cycles was required to predict the useful vessel life. Two time periods were considered in the extrapolation: 1) l operation to date for which vessel fluence has been calculated, 2) future fuel cycles which no analyses exist.

For the Waterford Unit 3 analysis, time period 1 was through cycle 4, and time period 2 covered cycles from the end of cycle 4 through 32 EFPY. The flux and l fluence for time period 2 was estimated by assuming that the flux at the inside surface of the pressure vessel (PVIS) for future cycles was the same as that l calculated for cycles 1 through 4. This was a reasonable assumption because the l

D-6

• gZGBE NUCLEAR

• ZWSERVICE COMPANIY

first four cycles were similar in that fresh fuel was loaded in the peripheral locations in each of the cycles.

It was found in the Waterford Unit 3 analysis and is shown in figure 6-3 that the peak fluence'at the PVIS occurred at approximately 1 degree off the major axis for cycles-l to 4. For this reason, the flux used to extrapolate from EOC 4 to 32 EFPY was the flux calculated at 1 degree. Future analyses will ascertain the actual effects.

D-7 11511BCW1'JUCLEAR

WS ERVICE COMPANY IMEINPYCO A

Table D-1. Flux Normalization Factor for 97lCapsule Measured Calculated Flux Dosimeter Activity, (a) Activity, (bi Normalization(c)

Reaction uCiJ/q UCi/ Factor 58Ni(n,p)5 Co 1835.0 1900.7 0.965 46Ti(n,p)48Sc 365.9 342.6 1.067 54Fe(n,p) 54 Mn 1427.0 1442.0 0.990 53Cu(n,a)"0Co 8.377 8.847 0.947 238U(n,f) 37Cs . 157 d 9.957 0.819 8

"Co(n,y)80 Cowal 4.380E+5 3.409E+5 1.286 59Co(n,g) eCoin 4.860E+4 5.643E+4 0.862 Averaged: 0.958(°}

(alAverage of three dosimeter wires, except for U powder capsules.

(b)Each listed activity was determined as the average of three calculated activities.

Ic)Average of measured to calculated activity ratios for each dosimeter type.

(diThe U dosimeters were powder capsules, three shielded and three bare. Since the U-238 dosimeter response in the thermal range is negligible, the shielding should have virtually no effect on the response of the dosimeters. The three shielded U samples showed good agreement with the calculations. However, of the three bare U-238 dosimeters, one was unrecoverable and one gave unreasonable results. Therefore, the U values in this table are for four dosimeters, three shielded and one bare, averaged together.

"'Bare dosimeters.

I0Cd-shielded dosimeters.

")Average of all dosimeters except Co.

D-8 3 VWINUCLE4 I3JWSEE&WCE COMPANY

Figure D-1. Rationale for the Calculation of Dosimeter Activities and Neutron Flux in the CaDsule ENDF/B4 Cross Sections Geometry & Quadrature Power Dis 13n-ENDF/85 Dosimeter Reacl for Model A DOT buttons SLnce Uon Cross Sections Capsule Iriser-t1on DDosmt Poe Radial Pov ActiviA t ShapsApp IG(eometry &

Quadrature Modlel OT4.3 Angula Flu odol B At Barre DDosimeter +. . feapsurod AAcil .ctit

. .Dosimeter Axial Ac~thltl Correction

.ormalizatic

.E .factor-

.CapsulWe FlxMCRatio D 9:

1U11EVNCWNUCLEAR IR WSERVICE COMPAIIY

ap Figure 0-2. Rationale for the Calculation of L

Neutron Flux in the Reactor Vessel

L

!L L

L L

Ii Ii I

I D-10 WNUCYE4Am I

F- - - - - II --- -

I I I I 'I I I --- ,-- I  ! - I-- I r 5 1 Figure D-3. Plan View Throuqh Reactor Core Midnlane (Reference R-9 Calculation Model) 0

~I."

thI M~

M~OdW A Amodel WC

APPENDIX E Capsule Dosimetry Data E-1 RMS8BCW NUCLEAR IZWSER VICE COMPANY

Table E-1 lists the characteristics of the neutron dosimeters. Tables E-2 and E-3 show the measured activity per gram of target material (i.e., per gram of uranium, nickel, etc.) for each capsule's dosimeters. Activation cross sections I for the various materials were flux-weighted with the 235U fission spectrum shown in Table E-4. I1 Table E-1. Detector Comnosition and Shieldina I Detector Material Shielding Reaction 58 Ni I

Ni Wire Cd (n,p) 58 Co Co Wire Bare 59Co(n, y)6 Co I Co Wire Cd 59Co(n,y)6 Co Fe Wire Bare 54Fe(n,p) 54Mn I Cu Wire Bare 63Cu(n,a)60 Co 238U(n, f)' 37Cs I

U308 Cd U3 08 Bare 238 1U(n, f) 13 7 Cs I 4 "Ti (n,p) 4 8 Sc Ti Wire Cd I

I I

.1

.1 I

E-2 13UBWNUCLEAI ZJWERVICE COMPANY I.

Table E-2. Measured Specific Activities (Unadjusted) for Dosimeters in 970 Capsule Dosimeter Activity,

-(uCi/qm of Target)

Detector Material Dosimeter Reaction Upper Center Lower Ni Wire 58Ni (n,p) 58 Co 1808 .1768 1929 59Co (n Iy)eoCo Co Wire(b) 4.357E+5 4.895E+5 3.887E+5 Co Wire(sh) 5gCo(n, y)eCo 5.163E+4 4.442E+4 4.976E+4 54Fe(n,p) 54Mn Fe Wire 1462 1374 1445 Cu Wire Cu(n,a)"0Co 8.519 7.907 8.706 U Powder(sh) 23SU (nf)137Cs 7.946 7.941 8.424 23 8 U Powder(b) U(nf)13 7Cs 16.00 -- 8.317 48Ti (n,p) 48Sc Ti Wire 332.3 354.9 410.5 E-3 I3W SCW NUCLEAR I:WSERVICE COMPANY

L Table E-3. Dosimeter Activation Cross Sections. b/atom(a)

I Group Upper No. Energy (eV) 40Ti (n,p) 23 sU(n, f) 54Fe(n,p) 58Ni(n,p) 63Cu(n, a) 59Co(n, y) I 1 1.733+7 2.407-1 1.215+0 2.803-1 3.215-1 3.641-2 7.968-4 I 2 1.419+7 2.667-1 1.033+0 4.260-1 4.980-1 4.535-2 8.380-4 3 1.221+7 2.600-1 9.851-1 4.728-1 5.734-1 5.360-2 7.633-4 4 1.000+7 2.356-1 9.933-1 4.769-1 5.971-1 3.842-2 6.978-4 I

5 8.607+6 2.043-1 9.898-1 4.759-1 5.988-1 1.926-2 9.431-4 6 7.408+6 1.555-1 8.240-1 4.687-1 5.845-1 9.389-3 2.214-3 I 7 6.065+6 9.645-2 5.588-1 4.266-1 5.141-1 2.956-3 2.455-3 8 4.966+6 3.766-2 5.452-1 3.041-1 3.847-1 4.568-4 2.871-3 I 9 3.679+5 5.573-3 5.292-1 1.998-1 2.424-1 3.600-5 3.269-3 10 3.012+6 4.747-4 5.282-1 1.371-1 1.674-1 5.844-6 3.523-3 I 11 2.723+6 6.816-6 5.365-1 8.061-2 1.232-1 1.692-6 3.772-3 12 2.466+6 1.100-6 5.398-1 5.715-2 9.340-2 6.645-7 3.938-3 13 2.365+6 3.770-7 5.404-1 5.134-2 8.278-2 4.712-7 4.006-3 I

14 2.346+6 3.427-7 5.410-1 4.564-2 7.227-2 3.305-7 4.090-3 15 2.231+6 2.326-7 5.358-1 2.892-2 4.600-2 1.124-7 4.337-3 I 16 1.921+6 8.518-8 4.799-1 8.181-3 2.440-2 1.500-8 4.931-3 17 18 1.653+6 0.000-0 0.000-0 3.154-1 2.933-3 1.206-2 0.000-0 6.222-3 I 1.353+6 4.480-2 6.824-4 3.758-3 0.000-0 8.205-3 19 20 1.003+6 8.209+5 0.000-0 0.000-0 1.296-2 3.820-3 5.308-5 4.367-6 1.362-3 1.156-3 0.000-0 0.000-0 7.473-3 6.519-3 I

21 22 7.427+5 6.081+5 0.000-0 0.000-0 1.553-3 6.233-4 6.842-7 1.097-7 9.891-4 7.958-4 0.000-0 0.000-0 6.905-3 7.598-3 I

0.000-0 23 24 4.393+5 2.688+5 0.000-0 2.846-4 1.635-4 8.051-8 5.615-8 6.086-4 4.483-4 0.000-0 0.000-0 9.233-3 8.724-3 I 25 2.972+5 0.000-0 1.001-4 3.448-8 3.058-4 0.000-0 1.058-2 26 1.835+5 0.000-0 7.720-5 1.197-8 1.577-4 0.000-0 1.322-2 I1 27 1.111+5 0.000-0 6.115-5 0.000-0 6.464-5 0.000-0 1.780-2 28 29 6.738+4 4.087+4 0.000-0 0.000-0 6.174-5 6.984-5 0.000-0 0.000-0 7.780-6 0.000-0 0.000-0 0.000-0 3.155-2 3.211-2

.1 30 3.183+4 0.000-0 7.894-5 0.000-0 0.000-0 0.000-0 3.892-2 I

I E-4 I

Table E-3. Dosimeter Activation Cross Sections, b/atom"a) (Cont'd)

Group Upper No. Energy (eV) 4"Ti (n,p) 238U(n,f) 54 Fe(n,p) 58 Ni(np) 8 3CU(na) '9Co (n, )

31 2.606+4 0.000-0 8.361-5 0.000-0 0.000-0 0.000-0 9.668-2 32 2.418+4 0.000-0 8.624-5 0.000-0 0.000-0 0.000-0 3.587-2 33 2.188+4 0.000-0 9.269-5 0.000-0 0.000-0 0.000-0 5.816-2 34 1.505+4 0.000-0 9.681-5 0.000-0 0.000-0 0.000-0 9.916-2 35 7.108+3 0.000-0 3.211-5 0.000-0 0.000-0 0.000-0 1.906-1 36 3.355+3 0.000-0 3.380-9 0.000-0 0.000-0 0.000-0 4.447-2 37 1.585+3 0.000-0 8.094-4 0.000-0 0.000-0 0.000-0 2.462-2 38 4.540+2 0.000-0 1.279-5 0.000-0 0.000-0 0.000-0 2.424-1 39 2.145+2 0.000-0 1.857-3 0.000-0 0.000-0 0.000-0 7.332+1 40 1.013+2 0.000-0 2.814-5 0.000-0 0.000-0 0.000-0 2.782+0 41 3.727+1 0.000-0 1.518-4 0.000-0 0.000-0 0.000-0 1.730+0 42 1.068+1 0.000-0 7.968-5 0.000-0 0.000-0 0.000-0 2.361+0 43 5.044+0 0.000-0 5.481-7 0.000-0 0.000-0 0.000-0 3.533+0 44 1.855+0 0.000-0 5.600-7 0.000-0 0.000-0 0.000-0 5.344+0 45 8.764-1 0.000-0 1.100-6 0.000-0 0.000-0 0.000-0 7.722+0 46 4.140-1 0.000-0 2.000-6 0.000-0 0.000-0 0.000-0 1.464+1 47 1.000-1 0.000-0 4.300-6 0.000-0 0.000-0 0.000-0 2.922+1

"'ENDF/B5 values that have been flux weighted (over BUGLE energy groups) based on a 235U fission spectrum in the fast energy range plus a l/E shape in the intermediate energy range.

E-5 5IMS1SVWINUCLEAR lJW SERVICE COMPANY

-APPENDIX F Tension Test Stress-Strain Curves F-1 13WSEWVICE COMPANY

i~

Figure F-1. Tension Test Stress-Strain Curve for Base Metal Plate Ii Heat M-1004-2. Snecimen No. 2L6. Tested at 70F 110.

Specimen: 2L6 Strength Yield: 70432.

Test Temp.: 70 F( 21 C) 700.

I

• UTS: 92605.

BB. 600.

0 0

500. 2

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I 0.

0. Gao . 0O .12 . 18 . 24 . 30 Figure F-2.

Engineering Strain Tension Test Stress-Strain Curve for Base Metal Plate Heat M-1004-2. Snerimen Nno PM Tested at 25OF I

100.

Specimen: 2K5 Strength Test Temp.: 250 F( 121 C)

I Yield: 65484.

80.

- UTS: 85778. 800.

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I Engineering Strain F-2 9WNUCLE411 IMSHE'RVICE COMPMY

Figure F-3. Tension Test Stress-Strain Curve for Base Metal Plate Heat M--1004-2. SDecimen No. 2K2. Tested at 550F Test Temp.: 550 F( 287 C) 100.

80.

0.

I0 50. 0; L

aU)

C L- 0.

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20. :

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0. 00 . 06 .12 . 18 .24 . 30 Engineering Strain Figure F-4. Tension Test Stress-Strain Curve for Base Metal Heat-Affected Zone.'Heat M-1004-2. Specimen No. 4K3. Tested at 70F' 1 Specimen: 4K3 Test Temp.: 70 F( 21 C)

Strength Yield: 69495. 1700.

1 UTS: 93506.I Be. 6Z0_o 0

(L Soo :2 500.

0 65.

0 0 L

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0. 00 . 05 .12 .18 .24 . 30 Engineering Strain F-3 13WS BWNUCLANR P COMPANY WSERVICE

11 Figure F-5. Tension Test Stress-Strain Curve for Base Metal Heat-Affected I

Zone. Heat M4-1004-2. Sinocimn No. UK. Tptpd at 750F I

i I

NO DATA - SEE SECTION 5.3 I

I I

II.

I Figure F-6. Tension Test Stress-Strain Curve for Base Metal Heat-Affected Zone. Heat M-1004-2. Snpcimen No. 4.14. Tested at ;5;F 1 10.

Specimen: 4J4 Strength Test Temp.: 550 F( 287 C)

I Yield: 69622. 700.

Be. I UTS: 91017.

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F-4 I3lWBSWNUCLEAR SERVICE COMPANY I1

Figure F-7. Tension Test Stress-Strain Curve for Weld Metal 88114/0145, Specimen No. 3JM, Tested at 70F Specimen: 3JM Test Temp.: 70 F( 21 C) 1 10.

Strength Yield: 84493. 700.

- UTS: 95852.

Be. -4600.

a IL

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0

66. 06 0

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0 0

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. 12 Engineering Strain Figure F-8. Tension Test Stress-Strain Curve for Weld Metal 88114/0145, Specimen No. 3KK. Tested at 250F NO DATA - SEE SECTION 5.3 F-5 111VBSW NUCLEAR I jWSERVICE COMPANY

IL Figure F-9. Tension Test Stress-Strain Curve for Weld Metal 88114/0145. Specimen No. 3KYE Tested at 550F I 110.

Specimen: 3KY Strength Test Temp.: 550 F( 287 C) 1J Yield: 74018. 700.

UTS: 93170.

Be. 8.

CK

' 44. *_30 400.

(Di 0

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22.I44 .00 l 100.

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I F-6 I3WBSWNUCLE4P

• MSERVICE COMPANY

APPENDIX G References G-1 1W SERVICE 13 UIMUW NUCLEAR COMPANY

1

1. Program for Irradiation Surveillance of Waterford Unit Three Reactor Vessel Materials, C-NLM-003, Revision 1, October 30, 1974. j

2. Summary Report on Manufacture of Test Specimens and Assembly of Capsules for Irradiation Surveillance of Waterford Unit 3 Reactor Vessel Materials, TR-C-MCS-001, Combustion Engineering, Inc., Windsor, Connecticut.

3. Code of Federal Regulation, Title 10, Part 50, Domestic Licensing of J Production and Utilization Facilities, Appendix H, Reactor Vessel Material Surveillance Program Requirements.

4. Code of Federal Regulation, Title 10, Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix G, Fracture Toughness Requirements.

5. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel ]

Code,Section III, Nuclear Power Plant Components, Appendix G, Protection Against Nonductile Failure (G-2000).

6. ASTM Standard E208, "Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.

7. A. G. Ragl, et al., Louisiana Power and Light Waterford Steam Electric Station Unit No. 3, Evaluation of Baseline Specimens, Reactor Vessel Materials Irradiation Surveillance Program, TR-C-MCS-002, Combustion Engineering, Inc., Windsor, Connecticut, August 1977. J

8. ASTM Standard E8, "Standard Methods of Tension Testing of Metallic Materi-als," in ASTM Standards, American Society for Testing and Materials, -

Philadelphia, PA.

9. ASTM Standard E21, "Standard Recommended Practice for Elevated Temperature ]

Tension Tests of Metallic Materials," inASTM Standards, American Society for Testing and Materials, Philadelphia, PA. l

10. ASTM Standard E184, "Standard Practice for Effects of High-Energy Neutron Radiation on the Mechanical Properties of Metallic Materials," in ASTM I Standards, American Society for Testing and Materials, Philadelphia, PA.

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11. ASTM Designation E23-72, "Method for Notched Bar Impact Testing of Metallic Materials," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.

12. Standardized Specimens for Certification of Charpy Impact Specimens from the Army Materials and Mechanics Research Center, Watertown, Mass. 02172, Attn:

DRXMR-MQ.

13. ASTM Designation A370-77, "Methods and Definitions for Mechanical Testing of Steel Products," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.

14. ASTM Designation E23-86, "Methods for Notched Bar Impact Testing of Metallic Materials," in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.

15. ASTM Designation E185-XX (to be released), Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels, in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.

16. S. Q. King, Pressure Vessel Fluence Analysis for 177-FA Reactors, BAW-1485P.

Revision 1, Babcock & Wilcox, Lynchburg, VA, March 1988.

17. B&W's Version of DOTIV Version 4.3, Filepoint 2A4, "One- and Two-Dimensional Transport Code System," Oak Ridge National Laboratory, Distributed by the Radiation Shielding Information Center as CC-429, November 1, 1983.

18. "Bugle - 80 Coupled 47 Neutron, 20 Gamma-Ray, P3 , Cross Section Library for LWR Shielding Calculations," Radiation Information Shielding Center, DLC-75.

19. Dosimeter File ENDF/B5 Tape 531, distributed March 1984, National Neutron Data Center, Brookhaven National Laboratory, Upton, Long Island, NY.

20. American Society of Testing Materials, Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements Per Atom (DPA), E693-79, (Reapprov-ed 1985).

21. U.S. Nuclear Regulatory Commission, Radiation Damage to Reactor Vessel Material, Regulatorv Guide 1.99. Revision 2, May 1988.

G-3 55185W NUCLEAR MUSBEE VICE COMPANY

22. Code of Federal Regulations, Title 10, Part 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.

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