ML063250415

Analysis of Capsule T from the Dominion Energy Kewaunee Power Station Reactor Vessel Radiation Surveillance Program, WCAP-16641-NP, Revision 0

ML063250415
Person / Time
Site:	Kewaunee
Issue date:	10/31/2006
From:	Shaun Anderson, Burgos B, Carlson J, Conermann J Westinghouse
To:	Office of Nuclear Reactor Regulation
References
WCAP-16641-NP, Rev 0
Download: ML063250415 (194)

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ATTACHMENT 1 SUBMITTAL OF IRRADIATED REACTOR VESSEL SURVEILLANCE CAPSULE TEST RESULTS FOR CAPSULE T PER 10 CFR 50 APPENDIX H WCAP-16641-NP, REVISION 0, "ANALYSIS OF CAPSULE T FROM DOMINION ENERGY KEWAUNEE POWER STATION REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM," OCTOBER 2006 KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Westinghouse Non-Proprietary Class 3 WCAP-16641-NP October 2006 Revision 0 Analysis of Capsule T from Dominion Energy Kewaunee Power Station Reactor Vessel Radiation Surveillance Program O=Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16641-NP, Revision 0 Analysis of Capsule T from the Dominion Energy Kewaunee Power Station Reactor Vessel Radiation Surveillance Program B.N. Burgos

• J. Conermann S.L. Anderson October 2006 Approved:

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

• ElectronicallyApproved Records Are Authenticated in ihe ElectronicDocument Management System Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355

©2006 Westinghouse Electric Company LLC All Rights Reserved

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

R.G. Lott:

• ElectronicallyApproved

• ElectronicallyApproved Records Are Authenticated in ihe Electronic Document Management System

iv This page intentionally blank

V FORWARD The first application of the Master Curve approach for an irradiated reactor vessel weld metal was approved by the NRC for the Kewaunee Power Station (KPS) in 2001 (Safety Evaluation by the Office of NuclearReactor Include the Use of a Master Curve-basedMethodology for Reactor Pressure Vessel IntegrityAssessment, Docket No. 50-305, May 2001). Testing of the next surveillance capsule for KPS included the requirement to perform additional fracture toughness tests to help validate the previous Master Curve evaluation accepted by the NRC. Two reports have been prepared to describe the results and evaluation of the additional fracture toughness testing performed as part of the Capsule T evaluation.

Capsule Requirements In accordance with the NRC SE, removal and testing of one additional capsule at a fluence equivalent to End-of-License-Renewal (EOLR) for the vessel weld of concern would be acceptable for monitoring radiation damage. The currently evaluated fluence for EOLR is documented in WCAP- 16641 -NP, the Capsule T analysis report, where the value was determined to be 5.37 x 1O1 9 n/cm 2 (E>I.0 MeV).

Additionally, the removal and testing of the capsule with fluence equivalent to 60 yrs will complete the current KPS surveillance program requirements. In accordance with the SE requirement, Capsule T was removed at a calculated fluence of 5.62 x 1019 n/cm 2 (E>1.0 MeV), which closely approximates the EOLR vessel fluence.

Master Curve Fracture Toughness T0 Determination The methodology detailed in Appendix A of the NRC SE is the methodology accepted by the NRC. The licensee agreed to use this methodology for future Master Curve fracture toughness testing and to incorporate the results into the KPS licensing basis. All margin terms are defined in Appendix A.

Specific to the testing requirements, the NRC stated the following:

1. Use of ASTM E 1921-97 is acceptable,

2. The use of multi-temperature maximum likelihood methodology is currently not endorsed (since it was not included in the ASTM Standard).

It was acknowledged that the state of knowledge regarding specific technical topics associated with the Master Curve approach may be improved in the future. Additional conservatisms may be reduced or removed provided technical justification is made. The NRC recognized that it may reconsider its' position based on action within ASME Standards organizations and revisions to ASTM E 1921.

In establishing a valid measurement of T. for weld wire heat 1P3571, several sources for the test specimens were deemed acceptable:

1. Charpy V-notch (CVN) weld specimens,

2. Reconstituted specimens from the weld portion of untested CVN heat-affected-zone (HAZ) specimens, and/or

3. Reconstituted weld specimens from broken halves of the original, broken weld CVN specimens.

WCAP-1664 1-NP Rev. 0

vi All specimens for fracture toughness testing were to be single-edge bend, SE(B), geometry as defined in ASTM E 1921; these specimens when fatigue precracked and conforming to CVN size are generally referred to as precracked Charpy V-notch (PCVN) specimens. All of the information in paragraphs 11.1 through 11.2.3 of ASTM E1921-97 for Capsule T, Capsule S, the Maine Yankee Capsule A35, and any unirradiated specimens used for the current licensee submittal were required to be included in the final reports for Capsule T and the new Master Curve evaluation. Use of Code Case N-629 to define a suitable expression for calculating the RTTo parameter was considered acceptable.

The actual PCVN specimens utilized in determining the measurement of To for Capsule T were a fabricated from a combination of the original irradiated CVN weld specimens (eight total) along with the reconstitution of four unbroken CVN HAZ specimen portions to provide a total of twelve specimens.

Details concerning the testing of the PCVN specimens are documented in WCAP- 16609-NP (Master Curve Report) and WCAP-16641-NP (Capsule T Analysis). In accordance with NRC guidance, the methodology in Appendix A of the SE was used and presented in WCAP- 16609-NP. In addition to this methodology, a new methodology has been developed under International Atomic Energy Agency (IAEA) sponsorship which has been applied and documented in WCAP-16609-NP.

The actual test results are presented in WCAP-16641-NP and the analysis is described in WCAP-16609-NP.

Charpy V-Notch Testing and Analysis In accordance with the NRC SE, a full CVN curve was not required to be developed for the surveillance weld, heat 1P3571. However, information regarding material properties were still required to be estimated to include the transition temperatures at 30 ft-lb, 50 ft-lb, and 35 mils along with the drop in upper shelf energy (USE). Accordingly, the methodology used in determining these values was documented in WCAP-16641-NP. Reconstitution of specimens needed to determine material properties was to be performed in accordance with ASTM E 1253, as described in WCAP-1664 1-NP. For the forging and correlation monitor materials, full CVN curves were required and testing/analysis performed in accordance with ASTM E 185-82. CVN impact testing of the HAZ material was not required.

A full CVN curve was not developed for the surveillance weld, however, the transition temperature values representing 30 ft-lbs, 50 ft-lbs and 35 mils were determined using the methodology presented in WCAP-1664 1-NP. The drop in the surveillance weld USE was also documented as a part of this analysis. The test results for the forging and correlation monitor materials also were documented in WCAP- 16641-NP.

As indicated earlier, four of the HAZ CVN specimens were reconstituted and used as weld metal PCVN specimens to help determine the Master Curve To value for the surveillance weld.

WCAP-16641-NP Rev. 0

vii TABLE OF CONTENTS L IS T O F TA B LE S ....................................................................................................................................... ix L IST O F F IG U RE S ..................................................................................................................................... xi EX E C U T IV E SUM M A RY ........................................................................................................................ xiii I SUM M A RY O F RESU LTS .................................................................................................... 1-1 2 IN T RO D U CT ION ........................................................................................................................ 2-1 3 B A C K G RO U N D ......................................................................................................................... 3-1 4 D ESCRIPTIO N O F PR O G RA M .................................................................................................. 4-1 5 TESTING OF SPECIMENS FROM CAPSULE T ...................................................................... 5-1 5.1 O V E RV IE W .................................................................................................................... 5-1 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS (FORGING/SRM) ............................ 5-3 5.3 TENSILE TEST RESULTS (FORGING) ........................................................................ 5-5 5.4 MASTER CURVE TEST RESULTS (WELD) ................................................................ 5-6 5.5 WEDGE OPENING LOADING SPECIMEN TESTS .................................................... 5-7 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY ....................................................... 6-1 6.1 IN T RO D U C T ION ......................................................................................................... 6-1 6.2 DISCRETE ORDINATES ANALYSIS ........................................................................... 6-2 6.3 NEU TR O N D O SIM ETRY .............................................................................................. 6-6 6.4 CALCULATIONAL UNCERTAINTIES ......................................................................... 6-7 7 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE .................................................... 7-1 8 REF ERE N C E S ............................................................................................................................. 8-1 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS CREDIBILITY APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS APPENDIX C MASTER CURVE RESULTS FROM CAPSULE T AND PREVIOUS FRACTURE TOUGHNESS TESTS APPENDIX D CHARPY V-NOTCH PLOTS FOR EACH CAPSULE T MATERIAL APPENDIX E KPS SURVEILLANCE PROGRAM CREDIBILITY EVALUATION WCAP- 1664 1-NP Rev. 0

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ix LIST OF TABLES Table 4-1 Chemical Composition (wt %) of the KPS Reactor Vessel Surveillance Materials ......... 4-3 Table 4-2 Heat Treatment History of the KPS Reactor Vessel Beltline Region Surveillance M a te ria ls .......................................................................................................................... 4 -4 Table 4-3 Chemical Composition of the A533 Grade B, Class 1 ASTM Correlation Monitor Material (HSST 02) in the KPS Vessel Surveillance Program ......................................... 4-4 Table 4-4 Heat Treatment History of the A533 Grade B, Class 1 ASTM Correlation Monitor Material (HSST 02) in the KPS Vessel Surveillance Program ......................................... 4-5 Table 5-1 Charpy V-Notch Data for the Kewaunee Power Station Intermediate Shell Forging 122X208VA1 Irradiated to a Fluence of 5.62 x 10"9 n/cm2 (E > 1.0 MeV)

(Tangential O rientation) ................................................................................................... 5-8 Table 5-2 Charpy V-Notch Data for the Kewaunee Power Station Lower Shell Forging 123X167VAI Irradiated to a Fluence of 5.62 x l109 n/cm2 (E > 1.0 MeV)

(T angential O rientation) ................................................................................................... 5-9 Table 5-3 Charpy V-notch Data for the Kewaunee Power Station Surveillance Representative Weld Metal Irradiated to a Fluence of 5.62 x 10' 9 n/cm 2 (E> 1.0 MeV) ................................ 5-10 Table 5-4 Charpy V-notch Data for the Kewaunee Power Station Correlation Monitor Material Irradiated to a Fluence of 5.62 x 1019 n/cm 2 (E> 1.0 M eV) ........................................... 5-10 Table 5-5 Instrumented Charpy Impact Test Results for the Kewaunee Power Station Intermediate Shell Forging122X208VA I Irradiated to a Fluence of 5.62 x 1019 n/cm 2 (E> 1.0 MeV)

(T angential O rientation) ................................................................................................. 5-11 Table 5-6 Instrumented Charpy Impact Test Results for the Kewaunee Power Station Lower Shell Forging 123X167VA1 Irradiated to a Fluence of 5.62 x 10'9 n/cm 2 (E> 1.0 MeV)

(Tangential O rientation) ................................................................................................. 5-12 Table 5-7 Instrumented Charpy Impact Test Results for the Kewaunee Power Station Weld Metal IP3571 Irradiated to a Fluence of 5.62 x 10'9 n/cm2 (E> 1.0 MeV) .............................. 5-13 Table 5-8 Instrumented Charpy Impact Test Results for the Kewaunee Power Station Correlation Monitor Material Irradiated to a Fluence of 5.62 x 1019 n/cm 2 (E> 1.0MeV) ............... 5-13 Table 5-9 Effect of Irradiation to 5.62 x 1019 n/cm 2 (E> 1.0 MeV) on the Capsule T Toughness Properties of the Kewaunee Power Station Reactor Vessel Surveillance Materials ...... 5-14 WCAP-16641-NP Rev. 0

x LIST OF TABLES (Cont.)

Table 5-10 Comparison of the Kewaunee Power Station Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions .................................................................................................. 5-15 Table 5-11 Tensile Properties of the Kewaunee Capsule T Reactor Vessel Surveillance Materials Irradiated to 5.62 x 10' 9 n/cm2 (E> 1.0M eV) ................................................................ 5-17 Table 5-12 Previously Measured Tensile Properties for Kewaunee Weld Heat 1P3571 .................. 5-18 Table 5-13 Fracture Toughness Test Results from Capsule T .......................................................... 5-19 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures at The Surveillance C apsule C enter .......................................................................................... 6-9 Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface .......................................... 6-13 Table 6-3 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV) Within The Reactor Vessel W a ll ..................................................................................................................... 6 -17 Table 6-4 Relative Radial Distribution of Iron Atom Displacements (dpa) Within the Reactor Vessel Wa ll ................................................................................................................................ 6 - 17 Table 6-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from K ew au n ee ....................................................................................................................... 6 -18 Table 6-6 Calculated Surveillance Capsule Lead Factors .............................................................. 6-18 Table 6-7 Calculated Maximum Neutron Fluence (E > 1.0 MeV) for the Kewaunee Extended B eltline M aterials ....................................................................................................... 6-19 Table 6-8 Calculated Maximum Neutron and Gamma Ray Exposure of the Primary B io logical S h ield ....................................................................................................... 6-19 Table 7-1 Recommended Surveillance Capsule Withdrawal Schedule ............................................ 7-1 WCAP-16641-NP Rev. 0

xi LIST OF FIGURES Figure 4-1 Arrangement of Surveillance Capsules in the KPS Reactor Vessel ................................. 4-6 Figure 4-2 Capsule T Diagram Showing the Location of Specimens and Dosimeters ...................... 4-7 Figure 4-3 Dosimeter and Thermal Monitor Layout for KPS Capsule T .......................................... 4-9 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for KPS Reactor Vessel Intermediate Shell Forging 122X208VAI (Tangential Orientation) ................................................... 5-20 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for KPS Reactor Vessel Intermediate Shell Forging 122X208VA1 (Tangential Orientation) ................................................... 5-20 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for KPS Reactor Vessel Intermediate Shell Forging 122X208VA 1 (Tangential Orientation) ................................................... 5-21 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for KPS Reactor Vessel Lower Shell Forging 123X167VA1 (Tangential Orientation) ............................................................ 5-22 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for KPS Reactor Vessel Lower Shell Forging 123X167VA1 (Tangential Orientation) ............................................................ 5-22 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for KPS Reactor Vessel Lower Shell Forging 123X167VA1 (Tangential Orientation) ............................................................ 5-23 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for KPS Reactor Correlation Monitor M ateria l......................................................................................................................... 5 -2 4 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for KPS Correlation Monitor M a teria l. ....................................................................................................................... 5 -2 4 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for KPS Reactor Correlation Monitor M aterial ......... ............................................................................................ .. .... 5-25 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature Best-Fit through the Transition Region for K PS Reactor Vessel Weld M etal .............................................................................. 5-26 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature Best-Fit through the Transition Region for KPS Reactor Vessel Weld M etal .................................................................. 5-27 Figure 5-12 Charpy Impact Specimen Fracture Surfaces for Kewaunee Reactor Vessel Intermediate Shell Forging 122X208VA I (Tangential Orientation) ............................... 5-28 WCAP-16641-NP Rev. 0

xii LIST OF FIGURES (Cont.)

Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Kewaunee Reactor Vessel Lower Shell Forgingl23Xl67VA 1 (Tangential Orientation) ............................................................. 5-29 Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Kewaunee A553 Grade B Class 1 C orrelation M onitor Material ......................................................................................... 5-30 Figure 5-15 Charpy Impact Specimen Fracture Surfaces for Kewaunee Reactor Vessel Reconstituted Weld M etal 5-29 ............................................................................................................ 5-30 Figure 5-16 Tensile Properties for Kewaunee, Capsule T Reactor Vessel Intermediate Shell Forging 122X 208VA1 (Tangential O rientation) .......................................................................... 5-31 Figure 5-17 Tensile Properties for Kewaunee, Capsule T Reactor Vessel Lower Shell Forging 123X 167VA 1 (Tangential O rientation) .......................................................................... 5-32 Figure 5-18 Fractured Tensile Specimens from the Kewaunee, Capsule T Reactor Vessel Intermediate Shell Forging 122X208VA 1 (Tangential Orientation) ................................................... 5-33 Figure 5-19 Fractured Tensile Specimens from the Kewaunee, Capsule T Reactor Vessel Lower Shell Forging 123X167VA1 (Tangential Orientation) ............................................................ 5-34 Figure 5-20 Engineering Stress-Strain Curves for Kewaunee Reactor Vessel Intermediate Shell Forging 122X208VA1, Capsule T, Tensile Specimens P-13 and P-14 .......................... 5-35 Figure 5-21 Engineering Stress-Strain Curves for Kewaunee Reactor Vessel Intermediate Shell Forging 122X208VA1, Capsule T, Tensile Specimens P-15 and P-16 .......................... 5-36 Figure 5-22 Engineering Stress-Strain Curve for Kewaunee Reactor Vessel Intermediate Shell Forging 122X208VA 1, Capsule T, Tensile Specimen P-17 ......................................................... 5-37 Figure 5-23 Engineering Stress-Strain Curves for Kewaunee Reactor Vessel Lower Shell Forging 123X167VA1, Capsule T, Tensile Specimens S-10 and S-Il ........................................ 5-38 Figure 5-24 Engineering Stress-Strain Curves for Kewaunee Reactor Vessel Lower Shell Forging 123X 167VA 1, Capsule T, Tensile Specimens S- 12 and S-13 ........................................ 5-39 Figure 6-1 Kewaunee rO Reactor Geometry at the Core Midplane ................................................ 6-20 Figure 6-2 K ew aunee rz Reactor G eom etry ................................................................................... 6-21 WCAP-1664 1-NP Rev. 0

xiii EXECUTIVE

SUMMARY

The purpose of this report is to document the results of the testing of surveillance Capsule T from the Dominion Energy Kewaunee Power Station (KPS). Capsule T was removed at 24.6 EFPY and post irradiation mechanical tests of the Charpy V-notch and tensile specimens were performed. A fluence evaluation utilizing the recently released neutron transport and dosimetry cross-section libraries was derived from the ENDF/B-VI data-base. Capsule T received a fluence of 5.62 x 1019 n/cm2 (E>1.0 MeV) after irradiation to 24.6 EFPY The peak clad/base metal interface vessel fluence after 24.6 EFPY of plant operation was 2.60 x 1019 n/cm 2 (E>I.0 MeV).

This evaluation lead to the following conclusions: 1) Two out of the five measured 30 ft-lb shift in transition temperature values of the intermediate shell forging 122X208VA 1 (tangential orientation) are greater than the Regulatory Guide 1.99, Revision 2 [Reference 1], predictions. 2) Two out of the five measured 30 ft-lb shift in transition temperature values of the lower shell forging 123X167VA1 (tangential orientation) are greater than the Regulatory Guide 1.99, Revision 2, predictions. 3) Two of the five measured 30 ft-lb shifts in transition temperature values of the weld metal are greater than the Regulatory Guide 1.99, Revision 2, predictions. 4) The measured percent decrease in upper shelf energy for all the surveillance materials contained in the KPS surveillance program are less than the Regulatory Guide 1.99, Revision 2 predictions with the exception of one weld metal data point. 5) All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the life of the vessel (33 EFPY) as required by 10CFR50, Appendix G [Reference 2]. 6) The KPS surveillance data from the intermediate and lower shell forgings 122X208VAt and 123X167VA1 were found to be non-credible.

The KPS surveillance data from the weld metal was found to be credible. The credibility evaluation can be found in Appendix E.

Lastly, a brief description of the Charpy V-notch and Master Curve testing can be found in Section 1.

Removal and testing of Capsule T completes the requested action items discussed in the NRC Safety Evaluation from May 2001 (Safety Evaluation by the Office of Nuclear ReactorInclude the Use of a Master Curve-basedMethodology for Reactor Pressure Vessel IntegrityAssessment, Docket No. 50-305, May 2001) [Reference 22]. This report, along with WCAP-16609-NP [Reference 5], satisfies the agreement to use master curve technology and that the transmittal of the reports satisfies the utility requirements for surveillance capsule testing over the lifetime of the plant. One additional capsule remains in the vessel, but is outside the scope of the current licensing requirements.

WCAP-1664 1-NP Rev. 0

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1-1 1

SUMMARY

OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule T, the fifth capsule removed and tested from the KPS reactor pressure vessel, led to the following conclusions:

• Capsule T received an average fast neutron fluence (E> 1.0 MeV) of 5.62 x 1019 n/cm 2 after 24.6 effective full power years (EFPY) of plant operation.

" Irradiation of the reactor vessel intermediate shell forging 122X208VA 1 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (tangential orientation), resulted in an irradiated 30 ft-lb transition temperature of 65°F and an irradiated 50 ft-lb transition temperature of 110WE This results in a 30 ft-lb transition temperature increase of 90'F and a 50 ft-lb transition temperature increase of 95°F for the longitudinal oriented specimens. See Table 5-9.

• Irradiation of the reactor vessel lower shell forging 123X 167VA 1Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (tangential orientation), resulted in an irradiated 30 ft-lb transition temperature of 207F and an irradiated 50 ft-lb transition temperature of 50'F. This results in a 30 ft-lb transition temperature increase of 70'F and a 50 ft-lb transition temperature increase of 757F for the longitudinal oriented specimens. See Table 5-9.

• Irradiation of the correlation monitor material Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 2207F and an irradiated 50 ft-lb transition temperature of 245°F.

This results in a 30 ft-lb transition temperature increase of 1757F and a 50 ft-lb transition temperature increase of 165°F. See Table 5-9.

• Irradiation of the reactor vessel representative weld Charpy specimens reconstituted from the HAZ material resulted in an irradiated 30 ft-lb transition temperature of 22 l°F and an irradiated 50 ft-lb transition temperature of 285'F. This results in a 30 ft-lb transition temperature increase of 271F and a 50 ft-lb transition temperature increase of 295°F for the HAZ specimens. See Table 5-9.

" The average upper shelf energy of the intermediate shell forging 122X208VA 1 (tangential orientation) resulted in an average energy decrease of 21 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 139 ft-lb for the tangentially oriented specimens. See Table 5-9.

• The average upper shelf energy of the lower shell forging 123X 167VAI (tangential orientation) resulted in an average energy decrease of 13 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 144 ft-lb for the tangentially oriented specimens. See Table 5-9.

• The average upper shelf energy of the correlation monitor material Charpy specimens resulted in an average energy decrease of 32 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 91 ft-lb for the correlation monitor material specimens. See Table 5-9.

Rev. 00 Summary of Results WCAP- 1664 11-NP WCAP-1664 -NP Rev. Summary of Results

1-2

• The average upper shelf energy of the representative weld metal from reconstituted HAZ Charpy specimens resulted in an average energy decrease of 54 ft-lb after irradiation. An irradiated average upper shelf energy of 72 ft-lb for the representative weld metal from the reconstituted HAZ specimens was measured. See Table 5-9.

• A comparison, as presented in Table 5-10, of the KPS reactor vessel surveillance material test results with the Regulatory Guide 1.99, Revision 2 [Reference 1] predictions led to the following conclusions:

- Two out of the five measured 30 ft-lb shifts in transition temperature values of the intermediate shell forging 122X208VA 1 (tangential orientation) are greater than the Regulatory Guide 1.99, Revision 2, predictions.

- Two out of the five measured 30 ft-lb shifts in transition temperature values of the lower shell forging 123X1 67VA1 (tangential orientation) are greater than the Regulatory Guide 1.99, Revision 2, predictions.

- Three of the five measured 30 ft-lb shifts in transition temperature value of the weld metal are greater than the Regulatory Guide 1.99, Revision 2, predictions.

- All of the correlation monitor material 30 ft-lb shifts in transition temperature were greater than the Regulatory Guide 1.99, Revision 2, predictions.

- The measured percent decrease in upper shelf energy for all the surveillance materials contained in the KPS surveillance program are less than the Regulatory Guide 1.99, Revision 2 predictions with the exception of one weld measurement.

" All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the extended life of the vessel (33 EFPY) as required by 10CFR50, Appendix G

[Reference 2].

" The calculated end-of-license (EOL) (33 EFPY) neutron fluence (E> 1.0 MeV) at the core mid-plane for the KPS reactor vessel using the Regulatory Guide 1.99, Revision 2 attenuation formula (i.e., Equation #3 in the guide) are as follows:

Calculated: Vessel inner radius* = 3.44 x 10'9 n/cm2 Vessel 1/4 thickness = 2.329 x 10' 9 n/cm2 Vessel 3/4 thickness = 1.068 x 10'9 n/cm 2

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

WCAP-1664 1-NP Rev. 0 of Results Summary of Results WCAP- 1664 1-NP Rev. 0

2-1 2 INTRODUCTION This report presents the results of the examination of Capsule T, the fifth capsule removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Dominion Energy Kewaunee Power Station (KPS) reactor pressure vessel materials under actual operating conditions. Capsule T was withdrawn and tested as it relates to the NRC requirements of the Safety Evaluation dated May 21, 2001 [Reference 22]. The first application of the Master Curve approach for an irradiated reactor vessel weld metal was approved by the NRC for the Kewaunee Power Station (KPS) in 2001 (Safety Evaluation by the Office of Nuclear ReactorInclude the Use of a Master Curve-basedMethodology for ReactorPressure Vessel IntegrityAssessment, Docket No. 50-305, May 2001). The testing of Capsule T was performed as commitment to the NRC as a part of this Safety Evaluation and included the requirement to perform additional fracture toughness tests to help validate the previous Master Curve evaluation accepted by the NRC.

The surveillance program for the KPS reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the pre-irradiation mechanical properties of the reactor vessel materials are presented in WCAP-8107, Revision 0, "Wisconsin Public Service Corporation Kewaunee Nuclear Power Plant Reactor Vessel Radiation Surveillance Program." [Reference 3]. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E 185-73, "Recommended Practice for Surveillance Tests on Structural Materials for Nuclear Reactors" [Reference 4]. Capsule T was removed from the reactor after 24.6 EFPY of exposure and shipped to the Westinghouse Science and Technology Department Hot Cell Facility, where the post-irradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed for the forging, correlation monitor, and weld material CVN specimens. The testing details, as discussed in the Forward to this report, satisfy the requirements of ASTM E 185-82, as amended by the Safety Evaluation from May 2001 [Reference 22].

In addition to summarizing the testing of and the post-irradiation data obtained from surveillance Capsule T removed from the KPS reactor vessel and discussing the analysis of the data, the report also presents the results of the master curve testing for circumferential weld IP3571 that was documented in WCAP-16609-NP [Reference 5]. A full CVN curve was not developed for the surveillance weld, however, the transition temperature values representing 30 ft-lbs, 50 ft-lbs and 35 mils were determined along with the drop in Upper Shelf Energy (USE). Reconstitution of specimens needed to determine these material properties was performed in accordance with ASTM E 1253-99 [Reference 6].

WCAP-16641-NP Rev. 0 Introduction

2-2 This page intentionally blank WCAP-1664 1-NP Rev. 0 of Program Description of Program WCAP-16641-NP Rev. 0

3-1 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA533 Grade B-1 (base material of the KPS reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness during high-energy irradiation. The primary role of fracture mechanics testing of RPV steels is to provide estimates of the fracture toughness to be used in the analysis of reactor pressure vessel integrity. Therefore, analysis of the fracture behavior of the steel requires characterization of the toughness over a range of temperatures encompassing the ductile-to-brittle transition.

One such method for ensuring the integrity of reactor pressure vessels has been presented in "Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code [Reference 7]. The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RTNDT).

RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208 [Reference 8]) or the temperature 60'F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented perpendicular (transverse) to the major working direction of the plate. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (Ki, curve) which appears in Appendix G to the ASME Code [Reference 7].

The Kic curve is a lower bound of static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the K1c curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.

RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The changes in mechanical properties of a given reactor pressure vessel steel, due to irradiation, can be monitored by a reactor vessel surveillance program, such as the KPS reactor vessel radiation surveillance program [Reference 3], in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the initial RTNDT, along with a margin (M) to cover uncertainties, to adjust the RTNDT (ART) for radiation embrittlement. This ART (RTNDT initial + M + ARTNDT) is used to index the material to the K1, curve and, in turn, to set operating limits for the nuclear power plant that take into account the effects of irradiation on the reactor vessel materials.

In addition to the traditional RTNDT approach used in characterizing the fracture toughness of ferritic steels, the development of the Master Curve testing procedure has provided a powerful new tool in determining the fracture toughness of a material.

The development of elastic-plastic fracture mechanics techniques introduced a new fracture toughness measure, the J-integral, which can be measured in small specimens. The J-integral has been used both as WCAP- 1664 1-NP Rev. 0 Background

3-2 a measure of ductile fracture toughness and to estimate equivalent linear-elastic toughness values in the lower transition region. The Master Curve provides a framework for using the linear-elastic toughness estimates derived from the J-integral to characterize the fracture toughness in the ductile-to-brittle transition regime.

Ferritic steels used in RPV construction exhibit a characteristic transition from brittle behavior at low temperatures to ductile behavior at higher temperatures. Under normal operating conditions, a nuclear RPV should always be in the high toughness ductile region. The stresses in the RPV must be carefully controlled during heatup and cooldown to avoid brittle fracture. It is therefore important to characterize the temperature at which the ductile-to-brittle transition occurs in a pressure vessel steel. A complete characterization of the transition requires testing at multiple temperatures. The temperature range over which this transition occurs depends on two factors: the properties of the material and the loading conditions. There are numerous tests designed to characterize the ductile-to-brittle transition in ferritic steels. Each test presents a unique combination of specimen geometry and loading and therefore a ductile-to-brittle behavior that is specific to the test method. This test-specific behavior is generally described in terms of a characteristic transition temperature. While it is common to speak of the ductile-to-brittle transition temperature of a material, there is not a unique definition of this value. In practice, any definition of transition temperature must refer to the test procedure (e.g., Charpy V-notch 30 ft-lb transition, drop weight nil-ductility temperature, and Master Curve TO). In each test, the transition temperature describes the ductile-to-brittle transition in the material.

For RPV steels, the most commonly used tests are the nil-ductility drop weight test and the Charpy impact test. The characteristic temperature in the drop weight test is defined as the nil-ductility point at the low temperature end of the transition (NDT). For nuclear RPV steels, the Charpy V-notch transition is usually characterized by the temperature at a specific absorbed energy level (30 ft-lb or 50 ft-lb). However, Charpy V-notch tests may also be characterized in terms of the fracture appearance transition temperature (FATT) or the temperature at a specific level deformation (e.g., 35 mils lateral expansion).

Although it has long been recognized that fracture toughness, as defined in ASTM E 1820 [Reference 9],

undergoes a ductile-to-brittle transition (which is what is really needed for accurate integrity assessment),

a characteristic transition temperature for the fracture toughness has only recently been defined. The development of J-integral based techniques for measuring fracture toughness (Jc) in the transition region has allowed a much clearer definition of fracture toughness behavior in ferritic steels. Based on this experience, it has been observed that ferritic steels have a common temperature dependence of fracture toughness in the transition regime. It was this observation that led Wallin to the definition of a Master Curve [Reference 10] that allows the fracture toughness for any ferritic steel to be characterized solely in terms of a reference temperature, To, corresponding to a fracture toughness of 100 MPa-m12 (91 ksi-in"2 ).

ASTM E 1921 [Reference 11], which was originally adopted in 1997, provides a standard test method for the determination of T0. This reference temperature can be used to index the Master Curve or some bounding curve. This behavior is in sharp contrast to the Charpy V-notch behavior, where both the transition curve shape and characteristic temperature vary between materials and after irradiation.

The ductile-to-brittle transition behavior of a material may be characterized using any combination of the above mentioned tests. The availability of test specimens and the expense of performing the tests generally determine the particular test employed to characterize the material transition temperature.

While there are multiple measures of the ductile-to-brittle transition, the underlying mechanisms of deformation and fracture are clearly inter-related. For this reason, the various measures of transition temperature tend to be correlated. These correlations allow a determination from one test technique to be used to estimate characteristic transition temperatures for the remaining tests. While it is difficult to demonstrate a correlation between NDT and the Charpy V-notch related transition temperature, recent test Background WCAP- 1664 1-NP Rev. 0

3-3 results have indicated a reasonable correlation between Charpy V-notch transition temperatures and To.

This correlation is fundamental to all Charpy based procedures for estimating toughness values in current RPV integrity analysis.

Rev. 00 Background WCAP-16641-NP WCAP- 1664 1-NP Rev. Background

3-4 This page intentionally blank WCAP-1664 1-NP Rev. 0 Background WCAP- 1664 1-NP Rev. 0

4-1 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the KPS reactor pressure vessel core region (beltline) materials were inserted in the reactor vessel prior to initial plant start-up. The six capsules were positioned in the reactor vessel between the thermal shield and the vessel wall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core. The capsules contained specimens made from intermediate shell forging 122X208VA 1, lower shell forging 123X167VA1 and weld metal fabricated with 3/16-inch Mil B-4 weld filler wire, heat number 1P3571 and Linde 1092 flux, lot number 3958. This is the identical weld sire heat and flux as that used in the actual fabrication of the KPS vessel intermediate to lower shell girth weld seam, which has been the limiting beltline material in the KPS reactor pressure vessel.

Capsule T, the fifth capsule removed as a part of the surveillance program, was removed after 24.6 effective full power years (EFPY) of plant operation. This capsule contained Charpy V-notch impact specimens, tensile specimens and Wedge Opening Loading (WOL) specimens from the two shell ring forgings of the reactor vessel and associated weld metal, and Charpy V-notch impact specimens of the Heat Affected Zone (HAZ) and the ASTM correlation monitor material. All test specimens were machined form the 1/4 thickness location of the forgings after performing a simulated post-weld, stress-relieving treatment on the test material. The test specimens represent material taken at least one forging thickness from the quenched ends of the forging. Specimens were machined from weld and HAZ metal from a stress-relieved weldment joining the intermediate shell forging 122X208VA 1 and lower shell forging 123C167VA1. All HAZ specimens were obtained form the weld HAZ of the intermediate shell forging 122X208VA 1.

All base metal Charpy V-notch impact and tensile specimens were machined with the longitudinal axis of the specimen parallel to the principal working direction of the forgings. The notch of the forging Charpy specimens was machined such that the direction of crack propagation in the specimen was transverse to the working direction. This orientation is termed "tangential." Charpy V-notch and tensile specimens from the weld metal were oriented such that the longitudinal axis of the specimen was normal to the welding direction. The notch of the weld Charpy specimens was machined such that the direction of crack propagation in the specimen was in the welding direction.

The chemical composition and heat treatment of the unirradiated surveillance materials are presented in Tables 4-1 through 4-4. The data in Tables 4-1 through 4-4 were obtained from the unirradiated surveillance program, WCAP-8107 Appendices A and B, and from the evaluation of the weld metal 1P3571 documented in WCAP-15074 [Reference 12].

Capsule T contained dosimeter wires of pure iron, copper, nickel, and aluminum-0. 15 weight percent cobalt (cadmium-shielded and unshielded). In addition, cadmium shielded dosimeters of neptunium (NP237) and uranium (U2 38) were placed in the capsule to measure the integrated flux at specific neutron energy levels.

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

WCAP- 1664 1-NP Rev. 0 Description of Program

4-2 2.5% Ag, 97.5% Pb Melting Point: 579°F (304'C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point: 590'F (3 10°C)

The arrangement of the various testing specimens, dosimeters, and thermal monitors contained in Capsule T is shown in Figures 4-2 and 4-3.

Description of Program WCAP-16641-NP Rev. 0

4-3 Table 4-1 Chemical Composition (wt%) of the KPS Reactor Vessel Surveillance Materials Element Intermediate Shell Lower Shell Forging Weld Metal Forging 122X208VA1 123XI67VAl C 0.21 0.20 0.12 Si 0.25 0.28 0.20 Mo 0.58 0.58 0.48 Cu 0.06 0.06 0.219(a)

Ni 0.71 0.75 0.724(a)

Mn 0.69 0.79 1.37 Cr 0.40 0.35 0.090 V <0.02 <0.02 0.002 Co 0.011 0.012 <0.001 Sn 0.01 0.01 0.004 Ti <0.001 <0.001 <0.001 Zr 0.001 0.001 <0.001 As 0.001 0.004 0.004 Sb <0.001 0.001 0.001 S 0.011 0.009 0.011 P 0.01 0.01 0.016 Al 0.004 0.006 0.010 B <0.003 <0.003 <0.003 N2 0.006 0.010 0.012 Zn --- -- <0.001 Notes:

a. Based upon average results as described in report WCAP-15074, Revision 1.

WCAP-16641-NP Rev. 0 Description of Program

4-4 Table 4-2 Heat Treatment of the KPS Reactor Vessel Beltline Region Surveillance Materials Material Temperature ('F) Time (hrs) Coolant Austenitizing @ 1550 8 Water Quenched Intermediate Shell Forging 122X208VA1 Tempered @ 1230 14 Air Cooled Stress Relieved @ 1150 21 Furnace Cooled Austenitizing @ 1550 8 Water Quenched Lower Shell Forging 123X167VA1 Tempered @ 1230 14 Air Cooled Stress Relieved @ 1150 21 Furnace Cooled Weldment Stress Relieved @ 1150 19.25 hrs. Furnace Cooled Table 4-3 Chemical Composition of the A533 Grade B, Class 1 ASTM Correlation Monitor Material (HSST Plate 02) in the KPS Vessel Surveillance Program Element Chemical Analysis (wt%)

C 0.22 Mn 1.48 P 0.012 S 0.018 Si 0.25 Ni 0.68 Mo 0.52 Cu 0.14 WCAP-1664 1-NP Rev. 0 Description of Program Description of Program WCAP- 1664 1-NP Rev. 0

4-5 Table 4-4 Heat Treatment of the A533 Grade B, Class 1 ASTM Correlation Monitor Material (HSST Plate 02) in the KPS Vessel Surveillance Program Material Temperature (*F) Time (hrs) Coolant 1675 +/- 25 4 Air Cooled Correlation Monitor 1600 +/- 25 4 Water Quenched Material Tempered @ 1225 +/- 25 4 Furnace Cooled Stress Relieved @ 1150 +/- 25 40 Furnace Cooled to 600'F Description of Program WCAP- 1664 1-NP WCAP-1664 Rev. 00 1-NP Rev. Description of Program

4-6 2700 REACTOR VESSEL N"THERMAL SHIELD ig@ P) 10 FtsT Figure 4-1 Arrangement of Surveillance Capsules in the KPS Reactor Vessel Program of Program WCAP-1664 1-NP Rev. 0 Description of WCAP- 1664 1-NP Rev. 0

4-7 LEGEND: P - INTERMEDIATE SHELL FORGING 122X208VA1 (TANGENTIAL)

S - LOWER SHELL FORGING 123X167VA1 (TANGENTIAL)

W - WELD METAL (HEAT # 1P3571)

H - HEAT AFFECTED ZONE METAL R - CORRELATION MONITOR MATERIAL Tensile WOL WOL WOL Tensile WOL WOL WOL Tensile Char Char Char Char S-13 0-

] H [N ] S-48 W-32 S-6W3

-44 W-28 S-42 W-26 S-2s-l S-47 W-31 S-45 W-29 S-43 W-27 S-41 W-25 TOP OF VESSEL CENTER j Dosimeter "A" Dosimeter "B" Dosimeter "C" Dosimeter Block Char Char Char Char Char Char Charpy Charpy Charpy

[-40 H-32 S-38 H-30 P-48 H-28 PA6 H-26 P-44 R-32 P-42 R-30 P-40 R-28 P-38 R-26 P-37 R-25 C -39 I- S-37 H-29 P-47 H-27 P-45 H-25 P-43 R-31 P-41 R-29 P-39 R-27 P-15 P-15 Dosimeter "B" Tensile Tensile WOL WOL WOL Tensile P-I 4 P-I13 1 0 BOTTOM OF VESSEL Dosimeter "A" Figure 4-2 Capsule T Diagram Showing the Location of Specimens and Dosimeters WCAP-1664 Rev. 00 1-NP Rev. Description of Program WCAP-16641-NP Description of Program

4-8 This page intentionally blank WCAP-1664 1-NP Rev. 0 Description of Program WCAP-16641-NP Rev. 0

4-9 Dosimeter Block "89" Covers A(' '"

U-238 Sealed Cadmium Capsule . Oxide

- Np-237 Sealed Capsule Dosimeter and Thermal Monitors A1-0.15%Co & Cd Cu Ni 579TF Monitor 590' Monitor 579°F Monitor Fe -- - - - Fe- - - - - Fe Dosimeter "A" Dosimeter "B" Dosimeter "C" Figure 4-3 Dosimeter and Thermal Monitor Layout for KPS Capsule T WCAP-16641-NP Rev. 0 Description of Program

4-10 This page intentionally blank Description of Program WCAP-16641-NP Rev. 0

5-1 5 TESTING OF SPECIMENS FROM CAPSULE T 5.1 OVERVIEW The post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens was performed in the Remote Metallographic Facility (RMF) at the Westinghouse Science and Technology Department. Testing on charpy and tensile specimens was performed in accordance with 10CFR50, Appendices G and H [Reference 2], ASTM Specification E 185-82 [Reference 13], and Westinghouse Procedure RMF 8402, Revision 2 [Reference 14] as detailed by Westinghouse RMF Procedures 8102, Revision 3 [Reference 15], and 8103, Revision 2 [Reference 16].

Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-8107

[Reference 3]. No discrepancies were found.

Examination of the two low-melting point 5797F (304°C) and 590'F (31 0'C) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 579°F (304'C).

The Charpy impact tests were performed per ASTM Specification E23-02a [Reference 17] and RMF Procedure 8103 [Reference 16] on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy impact test machine is instrumented with an Instron Impulse instrumentation system, feeding information into an IBM compatible computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED). From the load-time curve (Appendix B), the load of general yielding (Pcy), the time to general yielding (TGY), the maximum load (Pm), and the time to maximum load (tM) can be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (PF), and the load at which fast fracture terminated is identified as the arrest load (PA).

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

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

ary= (Pay *L) /[B * ( W_a)2

• C ] (1) where: L = distance between the specimen supports in the impact machine B = the width of the specimen measured parallel to the notch W = height of the specimen, measured perpendicularly to the notch a = notch depth WCAP- 1664 1-NP Rev. 0 Testing of Specimens from Capsule T

5-2 The constant C is dependent on the notch flank angle (4), notch root radius (p) and the type of loading (i.e., pure bending or three-point bending). In three-point bending, for a Charpy specimen in which 4)=

450 and p = 0.0 10 inch, Equation 1 is valid with C = 1.21. Therefore, (for L = 4W),

ay = (PGy*L)/[B * (W a) "*1.21]=(3.305*PGy*W)/[B * (W-a)'] (2)

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

ay=33.3 *PcY (3) where cy is in units of psi and PGY is in units of lbs. The flow stress was calculated from the average of the yield and maximum loads, also using the three-point bend formula.

The symbol A in columns 5, 6, and 7 of Tables 5-5 through 5-8 is the cross-section area under the notch of the Charpy specimens:

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

Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM Specification E23-02a and A370-97a [Reference 18]. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

Tensile tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specification E8-04 [Reference 19] and E21-03 [Reference 20], and Procedure RMF 8102. All pull rods, grips, and pins were made of Inconel 718. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.

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

The extensometer gage length was 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-93

[Reference 21 ].

Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air. Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperatures. Chromel-Alumel thermocouples were positioned at the center and at each end of the gage section of a dummy specimen and in each tensile machine griper. In the test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower tensile machine griper and controller temperatures was developed over the range from room temperature to 550'F. During the actual testing, the grip temperatures were used to obtain desired specimen temperatures. Experiments have indicated that this method is accurate to +2°F.

The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were Testing of Specimens from Capsule T WCAP- 1664 1-NP Rev. 0

5-3 calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.

The fracture toughness of the surveillance specimens is measured in terms of the J-integral. ASTM E 1820 provides a general procedure for conducting J-integral tests. The Master Curve used in ASTM E 1921 describes the temperature dependence of the cleavage initiation toughness. A cleavage event is readily identifiable as a rapid, unstable crack advance. Cleavage initiation implies that this fracture instability must occur prior to the onset of significant stable tearing crack extension. For a known starting crack length, the elastic and plastic contributions to the J-integral can be determined by measuring the specimen load and the amount of plastic work applied to the specimen. The testing procedure requires unloading compliance measurements to demonstrate that stable tearing has not initiated prior to the cleavage failure. If cleavage initiation occurs and the measurement meets the various validity requirements of the testing standard, the value of the J-integral at the point of instability is defined as Jc.

The equivalent linear elastic plane-strain fracture toughness at cleavage instability is Kj,. The Master Curve describes the temperature dependence of Kjc.

The determination of the fracture toughness transition temperature, To, requires multiple measurements of Kjc. In the original testing program, fracture toughness measurements were conducted on irradiated weld specimens from KPS Capsule S and Maine Yankee Capsule A-35. Unirradiated archival material from the KPS surveillance weld was also tested. Fracture toughness values for the unirradiated Maine Yankee were provided by the utility. Additional fracture toughness tests were conducted as part of the surveillance testing program for KPS Capsule T.

The first application of the Master Curve approach for an irradiated reactor vessel weld metal (Heat #

1P3571) was approved by the NRC for the Kewaunee Power Station (KPS) in 2001 [Reference 22]. In accordance with the NRC SE, the transition temperature values representing 30 ft-lbs, 50 ft-lbs, 35 mils, and the USE needed to be determined without having to meet the requirement of developing a full Charpy V-notch curve. The original 8 weld specimens that were contained in Capsule T were used for Master Curve testing. It was decided that the HAZ specimens would be reconstituted in accordance with ASTM E 1253 with five specimens being made available for additional Master Curve testing and 3 specimens available for Charpy testing. The reconstituted specimens were taken from the weld portion of the HAZ specimens and were deemed representative of the weld material. The methodology used to determine the transition temperature properties and the drop in USE for the reconstituted HAZ specimens is discussed in Appendix D.

5.2 CHARPY V-NOTCH IMPACT TEST RESULTS The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule T, which received a fluence of 5.62 x 1019 n/cm 2 (E> 1.0 MeV) in 24.6 EFPY of operation, are presented in Tables 5-1 through 5-10 and are plotted in Figures 5-1 through 5-11.

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

WCAP- 1664 1-NP Rev. 0 Testing of Specimens from Capsule T

5-4 Irradiation of the reactor vessel intermediate shell forging 122X208VA 1 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (tangential orientation), resulted in an irradiated 30 ft-lb transition temperature of 65'F and an irradiated 50 ft-lb transition temperature of 11 0F. This results in a 30 ft-lb transition temperature increase of 90'F and a 50 ft-lb transition temperature increase of 95°F for the longitudinal oriented specimens. See Table 5-9.

Irradiation of the reactor vessel lower shell forging 123Xl67VA1Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (tangential orientation), resulted in an irradiated 30 ft-lb transition temperature of 20'F and an irradiated 50 ft-lb transition temperature of 50'F. This results in a 30 ft-lb transition temperature increase of 70'F and a 50 ft-lb transition temperature increase of 75°F for the longitudinal oriented specimens. See Table 5-9.

Irradiation of the correlation monitor material Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 220'F and an irradiated 50 ft-lb transition temperature of 245°F.

This results in a 30 ft-lb transition temperature increase of 175°F and a 50 ft-lb transition temperature increase of 165°F. See Table 5-9.

Irradiation of the reactor vessel representative weld Charpy specimens reconstituted from the HAZ material resulted in an irradiated 30 ft-lb transition temperature of 221IF and an irradiated 50 ft-lb transition temperature of 285°F. This results in a 30 ft-lb transition temperature increase of 271IF and a 50 ft-lb transition temperature increase of 295°F for the HAZ specimens. See Table 5-9.

The average upper shelf energy of the intermediate shell forging 122X208VA1 (tangential orientation) resulted in an average energy decrease of 21 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 139 ft-lb for the tangentially oriented specimens. See Table 5-9.

The average upper shelf energy of the lower shell forging 123X 167VA 1 (tangential orientation) resulted in an average energy decrease of 13 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 144 ft-lb for the tangentially oriented specimens. See Table 5-9.

The average upper shelf energy of the correlation monitor material Charpy specimens resulted in an average energy decrease of 32 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 91 ft-lb for the correlation monitor material specimens. See Table 5-9.

The average upper shelf energy of the representative weld metal from reconstituted HAZ Charpy specimens resulted in an average energy decrease of 54 ft-lb after irradiation. An irradiated average upper shelf energy of 72 ft-lb for the representative weld metal from the reconstituted HAZ specimens was measured. See Table 5-9.

A comparison, as presented in Table 5-10, of the KPS reactor vessel surveillance material test results with the Regulatory Guide 1.99, Revision 2 predictions led to the following conclusions:

Testing of Specimens from Capsule T WCAP- 1664 1-NP Rev. 0

5-5

- Two out of the five measured 30 ft-lb shifts in transition temperature values of the intermediate shell forging 122X208VAI (tangential orientation) are greater than the Regulatory Guide 1.99, Revision 2, predictions.

- Two out of the five measured 30 ft-lb shifts in transition temperature values of the lower shell forging 123X167VA 1 (tangential orientation) are greater than the Regulatory Guide 1.99, Revision 2, predictions.

- Three of the five measured 30 ft-lb shifts in transition temperature value of the weld metal are greater than the Regulatory Guide 1.99, Revision 2, predictions.

- All of the correlation monitor material 30 ft-lb shifts in transition temperature were greater than the Regulatory Guide 1.99, Revision 2, prediction.

- The measured percent decrease in upper shelf energy for all the surveillance materials contained in the KPS surveillance program are less than the Regulatory Guide 1.99, Revision 2 predictions with the exception of one weld measurement.

All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the extended life of the vessel (33 EFPY) as required by 10CFR50, Appendix G.

The fracture appearance of each irradiated Charpy specimen from the various surveillance Capsule T materials is shown in Figures 5-12 through 5-15 and shows an increasingly ductile or tougher appearance with increasing test temperature.

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

Appendix D presents the curve fits for the Charpy specimens from Capsule T along with a detailed explanation on how the best-fit line was drawn for the weld data where only three PCVN specimens were available.

5.3 TENSILE TEST RESULTS The results of the tensile tests performed on the various materials contained in Capsule T irradiated to 5.62 x 1019 n/cm 2 (E> 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated results as shown in Figures 5-16 and 5-17. Note that no tensile weld specimens were contained in Capsule T. A summary of previous tensile test results for weld heat 1P3571 is contained in Table 5-12.

The results of the tensile tests performed on the intermediate shell forging 122X208VA1 (tangential orientation) indicated that irradiation to 5.62 x 1019 n/cm 2 (E> 1.0 MeV) caused approximately a 12 to 13 ksi increase in the 0.2 percent offset yield strength and approximately a 10 to 12 ksi increase in the ultimate tensile strength when compared to unirradiated data (see Figure 5-16).

Testing of Specimens from Capsule T WCAP-1664 1-NP Rev.

WCAP-16641-NP Rev. 0 0 Testing of Specimens from Capsule T

5-6 The results of the tensile tests performed on the lower shell forging 123X167VA1 (tangential orientation) indicated that irradiation to 5.62 x loll n/cm 2 (E> 1.0 MeV) caused approximately a 13 to 15 ksi increase in the 0.2 percent offset yield strength and approximately a 13 to 14 ksi increase in the ultimate tensile strength when compared to unirradiated data (see Figure 5-17).

The fractured tensile specimens for the shell forgings 122X208VAI and 123X167VA1 are shown in Figures 5-18 and 5-19. The engineering stress-strain curves for the tensile tests are shown in Figures 5-20 through 5-24.

5.4 MASTER CURVE TEST RESULTS The determination of the fracture toughness transition temperature, To, requires multiple measurements of Kj,. In the original testing program, fracture toughness measurements were conducted on irradiated weld specimens from KPS Capsule S and Maine Yankee Capsule A-35. Unirradiated archival material from the KPS surveillance weld was also tested. Fracture toughness values for the unirradiated Maine Yankee were provided by the utility. Additional fracture toughness tests were conducted as part of the surveillance testing program for KPS Capsule T.

The initial testing was carried out using the eight weld Charpy specimens from Capsule T. The Charpy V-notches in the surveillance specimens were modified to produce sharp crack starters for the pre-cracks.

The specimens were side-grooved after pre-cracking to provide uniform crack fronts for testing. Note that specimens from the previous capsules and for the unirradiated condition were not side-grooved; use of side-grooved specimens is now recommended in ASTM E 1921, but side-grooving is not a requirement.

The minimum number of specimens required to determine To is set by ASTM El1921. However the standard also indicates that additional tests may be required to minimize the effect of material variability.

All eight of the weld Charpy V-notch specimens were converted to PCVN specimens and tested at 136°F.

The initial results verified previous observations of significant variability in data from weld 1P3571. For materials with significant variability, the testing of a minimum of twelve specimens is recommended by ASTM El1921 to minimize the effects of the scatter. In accordance with the recommendations of the standard, an additional four PCVN specimens were reconstituted using the weld portion of unbroken HAZ specimens. These additional weld specimens were also tested at 136°F as indicated in Table 5-13.

The ASTM Standard requires that all measurements be included in the To determination. The values for the measured Kjc are listed as well as the IT size adjusted values, KJc(IT).

A summary of all of the Master Curve testing data for 1P3571 along with photographs of the fracture surfaces are presented in Appendix C. Cleavage fracture points are indicated by the sharp drops in the load-displacement curves. The elastic contribution to the fracture toughness is determined by the load at failure, the plastic contribution is determined by the area under the load-displacement curve (plastic work) up to failure. Details of the fracture toughness analysis are discussed in WCAP-16609-NP, "Master Curve Assessment of the Kewaunee Power Station Weld Metal."

Testing of Specimens from Capsule T WCAP-16641-NP Rev. 0

5-7 The cleavage initiation point in any individual specimen is determined by the distribution of potential cleavage initiation sites. The corresponding distribution of cleavage initiation toughness values can be described by a characteristic Weibull distribution. The scatter in the measured toughness values for weld 1P3571 is broader than the normal width of Weibull distributions in the Master Curve. This scatter can be attributed to inhomogeneous microstructure. This inhomogeneity is evident in the fracture surface photographs in Appendix C, which show broad bands across the specimen. The width of these bands, which is comparable to the thickness of the specimens, is believed to be related to individual weld passes.

Variations in the density of cleavage initiation sites across the weld passes and between consecutive weld passes would explain the observed level of scatter..

5.5 WEDGE OPENING LOADING (WOL) SPECIMEN TESTS Per the surveillance capsule testing contract, the WOL specimens were not tested and are being stored at the Westinghouse Science and Technology Center Hot Cell facility.

Rev. 00 Testing of Specimens from Capsule T WCAP- 1664 11-NP WCAP-1664 -NP Rev. Testing Of Specimens from Capsule T

5-8 Table 5-1 Charpy V-notch Data for the Kewaunee Power Station Intermediate Shell Forging 122X208VA1 Irradiated to a Fluence of 5.62 x 1019 n/cm 2 (E> 1.0 MeV) (Tangential Orientation)

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

P47 -50 -46 6 8 7 0.18 0 P46 50 10 22 30 17 0.43 10 P45 75 24 31 42 23 0.58 15 P43 100 38 65 88 50 1.27 40 P40 125 52 43 58 35 0.89 30 P37 150 66 81 110 57 1.45 60 P44 175 79 77 104 55 1.40 55 P41 200 93 109 148 74 1.88 85 P48 250 121 119 161 85 2.16 90 P38 325 163 140 190 82 2.08 100 P39 350 177 141 191 85 2.16 100 P42 400 204 137 186 84 2.13 100 Testing of Specimens from Capsule T WCAP- 1664 1-NP Rev. 0

5-9 Table 5-2 Charpy V-notch Data for the Kewaunee Power Station Lower Shell Forging 123X167VA1 Irradiated to a Fluence of 5.62 x 1019 n/cm 2 (E> 1.0 MeV) (Tangential Orientation)

Temperature Impact Energy Lateral Expansion Shear 0

OF C ft-lbs Joules mils mm  %

-50 -46 9 12 8 0.20 0 0 -18 7 9 5 0.13 2 25 -4 15 20 11 0.28 5 50 10 67 91 52 1.32 20 75 24 62 84 44 1.12 20 100 38 124 168 77 1.96 90 125 52 88 119 61 1.55 55 150 66 97 132 67 1.70 65 175 79 139 188 89 2.26 100 250 121 142 193 86 2.18 100 300 149 147 199 85 2.16 100 350 177 146 198 83 2.11 100 Testing of Specimens from Capsule T WCAP-1664 WCAP-16641-NP Rev. 00 1-NP Rev. Testing of Specimens from Capsule T

5-10 Table 5-3 Charpy V-notch Data for the Kewaunee Power Station Surveillance Representative Weld Metal Irradiated to a Fluence of 5.62 x 10' 9 n/cm 2 (E> 1.0 MeV)

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

H26 190 88 13 18 13 0.33 25 H27 215 102 36 49 28 0.71 45 H32 400 204 72 98 59 1.50 100 Table 5-4 Charpy V-notch Data for the Kewaunee Power Station Correlation Monitor Material Irradiated to a Fluence of 5.62 x 1019 n/cm 2 (E> 1.0 MeV)

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

R31 100 38 4 5 7 0.18 5 R30 200 93 15 20 11 0.28 15 R28 225 107 43 58 29 0.74 40 R26 250 121 46 62 32 0.81 50 R27 275 135 68 92 53 1.35 70 R29 300 149 93 126 66 1.68 85 R32 400 204 93 126 73 1.85 95 R25 450 232 87 118 69 1.75 100 Testing of Specimens from Capsule T WCAP- 1664 1-NP Rev. 0

5-11 Table 5-5 Instrumented Charpy Impact Test Results for the Kewaunee Power Station Capsule T Intermediate Shell Forging 122X208VA1 Irradiated to a Fluence of 5.62 x 1019 n/cm 2 (E>1.0 MeV)

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

P47 -50 6 48 25 24 2816 0.13 2905 0.14 2905 0 94 95 P46 50 22 177 131 46 3246 0.14 4126 0.34 4121 0 108 123 P45 75 31 250 184 65 3496 0.14 4452 0.42 4437 0 116 132 P43 100 65 524 318 206 3370 0.15 4463 0.68 4229 0 112 130 P40 125 43 346 233 114 3225 0.15 4367 0.54 4352 0 107 126 P37 150 81 653 313 340 3231 0.15 4397 0.68 3980 580 108 127 P44 175 77 620 320 300 3234 0.19 4279 0.73 4021 671 108 125 P41 200 109 878 299 579 3191 0.15 4308 0.67 2935 1432 106 125 P48 250 119 959 298 661 2925 0.14 4176 0.68 2274 1466 97 118 P38 325 140 1128 305 823 3054 0.14 4314 0.68 N/A N/A 102 123 P39 350 141 1136 299 837 3052 0.15 4138 0.69 N/A N/A 102 120 P42 400 137 1104 292 812 2941 0.16 4111 0.69 N/A N/A 98 117 WCAP- 1664 1-NP Rev. 0 Testing of Specimens from Capsule T

5-12 Table 5-6 Instrumented Charpy Impact Test Results for Kewaunee Power Station Capsule T Lower Shell Forging 123X167VA1 Irradiated to a Fluence of 5.62 x 1019 n/cm 2 (E>1.0 MeV)

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

S45 -50 9 73 41 31 3743 0.14 4075 0.17 4075 0 125 130 S47 0 7 56 34 23 3520 0.15 3533 0.16 3533 0 117 117 S40 25 15 121 69 51 3629 0.14 4411 0.22 4404 0 121 134 S38 50 67 540 331 208 3534 0.15 4620 0.68 4208 0 118 136 S37 75 62 500 331 169 3516 0.15 4636 0.68 4427 0 117 136 S48 100 124 999 346 653 3658 0.16 4801 0.69 3029 1533 122 141 S43 125 88 709 316 393 3338 0.15 4499 0.67 3862 469 111 130 S44 150 97 782 316 466 3230 0.15 4416 0.69 3281 707 108 127 S46 175 139 1120 311 809 3290 0.15 4438 0.68 N/A N/A 110 129 S41 250 142 1144 302 843 3006 0.15 4306 0.68 N/A N/A 100 122 S39 300 147 1184 312 872 3101 0.14 4367 0.68 N/A N/A 103 124 S42 350 146 1176 297 879 3012 0.14 4254 0.67 N/A N/A 100 121 T WCAP-1664 1-NP Rev. 0 Testing of Specimens from Capsule T from Capsule WCAP- 1664 1-NP Rev. 0

5-13 Table 5-7 Instrumented Charpy Impact Test Results for the Kewaunee Power Station Weld Metal 1P3571 Irradiated to a Fluence of 5.62 x 1019 n/cm 2 (E>1.0 MeV)

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

H26 190 13 105 42 63 3586 0.16 3780 0.17 3780 556 119 123 H27 215 36 290 149 141 3321 0.14 4861 0.35 4680 1157 111 136 H32 400 72 580 216 364 3263 0.14 4563 0.48 N/A N/A 109 130 Table 5-8 Instrumented Charpy Impact Test Results for the Kewaunee Power Station Correlation Monitor Material Irradiated to a Fluence of 5.62 x 1019 n/cm 2 (E>1.0 MeV)

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

R31 100 4 32 17 15 2109 0.12 2132 0.13 2132 0 70 71 R30 200 15 121 63 58 3204 0.14 4027 0.21 4027 0 107 120 R28 225 43 346 224 123 3538 0.15 4719 0.49 4708 1361 118 137 R26 250 46 371 235 135 3555 0.15 4811 0.50 4801 1282 118 139 R27 275 68 548 239 309 3355 0.15 4541 0.53 4455 2143 112 131 R29 300 93 749 240 509 3398 0.18 4599 0.54 2369 1571 113 133 R32 400 93 749 234 515 3380 0.19 4496 0.54 3466 3015 113 131 R25 450 87 701 222 479 3287 0.16 4438 0.51 N/A N/A 109 129 WCAP- 1664 1-NP Rev. 0 Testing of Specimens from Capsule T

5-14 Table 5-9 Effect of Irradiation to 5.62 x 1019 n/cm 2 (E>1.0 MeV) on the Capsule T Toughness Properties of the Kewaunee Power Station Reactor Vessel Surveillance Materials Average 30 (ft-lb)(') Average 35 mil Lateral(b) Average 50 ft-lb('a) Average Energy Absorption(a)

Material Transition Temperature ( 0F) Expansion Temperature (*F) Transition Temperature (*F) at Full Shear (ft-lb)

Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AE Intermediate Shell Forging -25 65 90 -15 100 115 15 110 95 160 139 21 122X208VA I (Tangential)

Lower Shell Forging -50 20 70 -45 50 95 -25 50 75 157 144 13 123X167VAI (Tangential)

Reconstituted Representative -50 221 271 -35 249 284 -10 285 295 126 72 54 Weld Metal (Heat# 1P3571)

Correlation 45 220 175 60 245 185 80 245 165 123 91 32 Monitor Material

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

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

Testing of Specimens from Capsule T WCAP- 1664 1-NP Rev. 0

5-15 Table 5-10 Comparison of the Kewaunee Power Station Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy Fluence Temperature Shift Decrease Material Capsule (X 1019 n/cm 2, E > 1.0 MeV) Predicted Measured Predicted Measured (OF) (a) (OF) (b) (%) (a) (O/o)(C)

V 0.586 31.45 0 17 0 Intermediate Shell R 1.76 42.74 15 22 0 Forging P 2.61 46.51 25 24 2 122X208VA1 S 3.67 49.47 60 26 8 T 5.62 52.73 90 29 13 V 0.586 31.45 0 17 0 R 1.76 42.74 20 22 3 Lower Shell Forging P 2.61 46.51 20 24 0 123X167VA1 S 3.67 49.47 50 26 3 T 5.62 52.73 70 29 8 V 0.586 159.12 175 33 35 Surveillance R 1.76 216.22 235 42 38 Pora eldne P 2.61 235.31 230 Program Weld Metal 44 40 S 3.67 250.29 250 49 49 T 5.62 266.76 271 54 43 V 0.586 86.7 95 21 11 Correlation Monitor R 1.76 117.81 140 27 23 Matial P 2.61 128.21 155 Material 29 18 S 3.67 136.37 158 31 20 T 5.62 145.35 175 35 26 V 0.586 --- 80 --- 19 R 1.76 --- 150 --- 22 Heat Affected Zone P 2.61 --- 220 --- 24 Material S 3.67 --- 200 --- 23 T 5.62 (d) --- (d)

Notes:

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

(b) Calculated using measured Charpy data (See Appendix D).

(c) Values are based on the definition of USE given in ASTM E185-82 with the exception of the representative weld specimen reconstituted from HAZ material, where the USE was defined by a single point.

(d) Not required to be measured at a part of the NRC SE [Reference 22].

WCAP-16641-NP Rev. 00 Testing of Specimens from Capsule T WCAP- 1664 1-NP Rev. Testing of Specimens from Capsule T

5-16 This page intentionally blank Testing of Specimens from Capsule T WCAP- 1664 1-NP Rev. 0

5-17 2

Table 5-11 Tensile Properties of the Kewaunee Capsule T Reactor Vessel Surveillance Materials Irradiated to 5.62 x 1019 n/cm (E > 1.0 MeV) (a)

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

P-13 75 80.1 100.2 3.05 201.0 62.1 10.1 23.7 69 P-14 125 77.7 97.3 2.95 194.4 60.1 9.5 23.3 69 Intermediate Shell Forging P-15 200 74.4 93.2 2.83 206.4 57.6 8.9 22.4 72 122X208VA 1 P-16 300 70.8 90.7 2.85 196.2 58.1 9.0 21.9 70 P-17 550 68.8 92.1 3.05 189.9 62.1 9.0 21.2 67 S-10 75 82.5 102.4 3.08 211.7 62.6 10.1 24.2 70 Lower Shell S-11 125 80.5 100.3 3.10 198.5 63.2 8.9 21.8 68 Forging 123X167VA1 S-12 200 76.9 96.2 3.00 215.9 61.1 9.3 22.0 72 S-13 550 71.5 95.1 3.25 161.6 66.2 8.7 20.0 59 Notes:

(a) No weld data is reported since Capsule T did not contain weld tensile specimens.

Testing of Specimens from Capsule T Rev. 0 WCAP-16641-NP Rev. 0 Testing of Specimens from Capsule T

5-18 Table 5-12 Previously Measured Tensile Properties for Kewaunee Weld Heat 1P3571 Sample 0.2% Yield Ultimate Fracture Uniform Total Reduction Source Strength Strength Strength Elingation Elongation Number (ksi) (ksi) Load (kip) Stress (ksi) (ksi) (%) (%) in Area (%)

MY Capsule 3JK 300 107.0 118.2 4.5 193.7 91.7 10.8 19.5 53 A25 3J5 560 98.8 110.5 4.3 151.7 87.6 10.0 18.0 42 3JM 74 107.0 115.1 4.0 152.2 81.5 12.7 24.8 46 W253 3J7 300 93.7 105.9 3.9 171.8 79.5 12.0 21.6 54 3L4 560 87.6 105.9 4.2 174.6 85.6 10.5 17.9 51 3JT Room 102.2 116.9 --- .---

--- 24.2 52.1 A35 313 566 93.3 110.8 --- --- --- --- 21.8 42.8 3KJ(a) 650 --- 104.1 --- --- --- 18.2 44.0 MY Capsule 3JE 86 100.8 114.4 --- 200.5 82.3 15.1 26.0 58.9 W263 3J1 550 88.7 105.4 --- 162.6 87.6 10.1 17.9 46.1 3JB 560 84.0 103.3 --- 164.0 84.5 13.03 20.6 48.5 1 Room 69.0 84.9 -.--- --- 15.7 27.5 72.2 Kewaunee 2 Room 68.6 84.4 -.--- --- 15.4 26.5 72.3 Unirradiated 3 300 66.6 78.4 --- --- --- 10.2 21.9 69.9 Surveillance 4 300 62.7 76.8 --- --- --- 12.9 23.3 68.8 Weld 5 600 59.6 80.7 --- --- --- 13.8 22.7 65.0 6 600 60.1 81.0 --- --- --- 13.2 22.4 66.4 Kewaunee W3 250 97.8 109.9 3.80 188.6 77.5 12.0 22.5 59 Capsule R W4 550 87.6 103.8 3.98 175.5 80.9 12.0 19.2 54 Kewaunee W2 200 94.7 105.9 --- 74.3 188.5 10.4 22.0 61.6 Capsule V W1 550 83.5 102.0 --- 81.5 206.5 10.4 20.5 52.7 Kewaunee W5 200 101.3 112.0 4.00 192.9 81.5 10.5 21.8 58 Capsule S W6 550 91.7 105.9 4.00 176.2 81.5 9.8 19.7 54 Notes:

(a) Accidental pre-straining of specimen prior to testing resulted in unreliable yield strength measurement and questionable test results.

Testing of Specimens from Capsule T WCAP-16641-NP Rev. 0

5-19 Table 5-13 Fracture Toughness Test Results from Capsule T [Reference 51 Specimen Code Specimen Type Test Temperature KiC KJC(IT)

Specien_ CdeSpcime____ (OF) (ksi-in 12) (ksi-in 112 )

W25 PCVN 136 54.6 47.0 W26 PCVN 136 65.0 55.3 W27 PCVN 136 90.1 75.1 W28 PCVN 136 149.7 122.4 W29 PCVN 136 79.6 66.8 W30 PCVN 136 78.9 66.3 W31 PCVN 136 46.0 40.2 W32 PCVN 136 59.6 51.0 H25 Recon. PCVN 136 119.0 98.0 H29 Recon. PCVN 136 75.6 63.6 H30 Recon. PCVN 136 53.1 45.0 H31 Recon. PCVN 136 53.2 46.0 Testing of Specimens from Capsule T WCAP-16641-NP Rev. 0 WCAP-1664 1-NP Rev. 0 Testing of Specimens from Capsule T

5-20 KPS Forging 122X208VA1 (Tangential) 160 140 ___

120 -

100 /

80 -_ _ __

z> 60 40 20 ____________

-100 0 100 200 300 400 500 Temperature (Deg F)

Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for KPS Reactor Vessel Intermediate Shell Forging 122X208VA1 (Tangential Orientation)

KPS Forging 122X208VA1 (Tangential) 100 80 ___

U) 0 60 CL 0.

40 1 1003005

-100 0 100 200 300 400 500 Temperature (Deg F)

Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for KPS Reactor Vessel Intermediate Shell Forging 122X208VA1 (Tangential Orientation)

Testing of Specimens from Capsule T WCAP- 1664 1-NP Rev. 0

5-21 KPS Forging 122X208VA1 (Tangential) 100-90-80 70-60-S50 -

u0 40-30-20-10-0+-

-1O00 0 100 200 300 400 500 Temperature (Deg F)

Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for KPS Reactor Vessel Intermediate Shell Forging 122X208VA1 (Tangential Orientation)

Rev. 00 Testing of Specimens from Capsule T WCAP- 1664 11-NP WCAP-1664 -NP Rev. Testing of Specimens from Capsule T

5-22 KPS Forging 123X167VA1 (Tangential) 160 140 a_

120

• 100 80 z 60-40 20 .. .

-100 0 100 200 300 400 Temperature (Deg F)

Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for KPS Reactor Vessel Lower Shell Forging 123X167VAI (Tangential Orientation)

KPS Forging 123X167VA1 (Tangential) 100 80 0 40 6 _

W 40

" 20

-100 0 100 200 300 400 Temperature (Deg F)

Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for KPS Reactor Vessel Lower Shell Forging 123X167VA1 (Tangential Orientation)

Testing of Specimens from Capsule T WCAP- 1664 1-NP Rev. 0

5-23 KPS Forging 123X167VA1 (Tangential) 100' 90 __- -

80 70 /o

• 60o Mu 50 S40 30 20 100 0 10 200 300 400

-100 0100 200 300 400 Temperature (Deg F)

Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for KPS Reactor Vessel Lower Shell Forging 123X167VAI (Tangential Orientation)

Rev. 00 Testing of Specimens from Capsule T WCAP- 1664 11-NP WCAP-1664 -NP Rev. Testing of Specimens from Capsule T

5-24 KPS Correlation Monitor Material 100I 80 60 w" 40 z

0)20 0 100 200 300 400 500 Temperature (Deg F)

Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for KPS Reactor Vessel Correlation Monitor Material KPS Correlation Monitor Material 80-IS j60 -____

E 0

4400 wT r 20-0 0-0 100 200 300 400 500 Temperature (Deg F)

Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for KPS Reactor Vessel Correlation Monitor Material Testing of Specimens from Capsule T WCAP-16641-NP Rev. 0

5-25 KPS Correlation Monitor Material 100-80 _ ________

70 60 /

50 0 40_

30_

20 10 0 100 200 300 400 500 Temperature (Deg F)

Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for KPS Reactor Vessel Correlation Monitor Material WCAP- 1664 1-NP Rev. 0 Testing of Specimens from Capsule T

5-26 KPS Weld CVN Data 80 70

-"60 50 Q 40 zW30 U020 10 0

0 100 200 300 400 500 Temperature (Deg F)

Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature Best-Fit through the Transition Region for KPS Reactor Vessel Weld Metal Testing of Specimens from Capsule T WCAP- 1664 1-NP Rev. 0

5-27 KPS Weld Lateral Expansion Data 70 -. _________ - r r 601 -4 I

~50 50i40

• Measured Lateral uj30 S Expansion Data

~20 - - Slope = 0.308

_j 0 10 0 I J

I I I

I I I

7 I r

I 0 100 200 300 400 500 Temperature (Deg F)

Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature Best-Fit through the Transition Region for KPS Reactor Vessel Weld Metal WCAP- 1664 1-NP Rev.

Rev. 0 Testing of Specimens trom Lapsule I WCAP-16641-NP Testing Of Specimens from Capsule T

5-28 PZL/ -IN'IWI-1 P/lfh Ni%1 I- 1../1"% INVG I-'/1"4 I I li I° [JA11I) SI-P37. 150 0 F P44.175°F P41,200 0 F P48,250°F P38,325°F P39, 350-F Figure 5-12 Charpy Impact Specimen Fracture Surfaces of the Kewaunee Reactor Vessel Intermediate Shell Forging 122X208VA1 (Tangential Orientation).

WCAP-1664 1-NP Rev. 0 Testing of Testing Specimens from of Specimens Capsule T from Capsule T WCAP- 1664 1-NP Rev. 0

5-29 S44, 1M*J S46, 115-1 S41, 250-1

-,4/, -3ZU-T Figure 5-13 Charpy Impact Specimen Fracture Surfaces of the Kewaunee Reactor Vessel Lower Shell Forging 123X167VAI (Tangential Orientation).

Rev. 00 Testing of Specimens from Capsule T WCAP- 1664 1I-NP WCAP-1664 -N P Rev. Testing of Specimens from Capsule T

5-30 R26, 250'F R29, 300°F KiZ, 4UU-t Figure 5-14 Charpy Impact Specimen Fracture Surfaces of the Kewaunee A533 Grade B Class 1 Correlation Monitor Material.

H26, 190°F H27, 215°F H32, 400°F Figure 5-15 Charpy Impact Specimen Fracture Surfaces of the Kewaunee Reactor Vessel Reconstituted Weld Metal.

Testing of Specimens from Capsule T WCAP- 1664 1-NP Rev. 0

5-31 120.0 100.0 Ultimate Yield Strength A --

80.0-o0 0 00 o Y U) 60.0-0.2% Yield Strength 40.0-20.0-0.0 0 100 200 300 400 500 600 700 Temerature (Deg F)

Legend: A and o are Unirradiated Aand 9 are Irradiated to 5.37 x 1019 n/cm 2 (E > 1.0 MeV) 80 70 60 Reduction in Area 50 40

.4-U Total Elongation 30 20 U-10 Uniform Elongation 0

0 100 200 300 400 500 600 700 Temperature (Deg F)

Figure 5-16 Tensile Properties for Kewaunee, Capsule T Reactor Vessel Intermediate Shell Forging 122X208VA1 (Tangential Orientation).

WCAP- 1664 1-NP Rev. 0 Testing of Specimens from Capsule T

5-32 120.0 Ultimate Yield Strength 100.0 -

0A 80.0 -

an an 60.0 -

0I I-

.0~

0.2% Yield Strength U, 40.0 -

20.0 -

0.0 0 100 200 300 400 500 600 700 Temperature (Deg F)

Legend: A and o are Unirradiated Aand 9 are Irradiated to 5.37 x 1019 n/cm 2 (E > 1.0 MeV) 80.0 70.0 A 60.0 Reduction in Area F0:- 50.0 40.0 Total Elongation 30.0 20.0 10.0 Uniform Elongation 0.0 5 0 100 200 300 400 5O00 600 700 Temperature (Deg F)

Figure 5-17 Tensile Properties for Kewaunee, Capsule T Reactor Vessel Lower Shell Forging 123X167VA1 (Tangential Orientation).

Testing of Specimens from Capsule T WCAP-16641-NP Rev. 0

5-33 cl]r*- - n I 'I * - "'/COl' Stecimen P-14 Tested at 125VF Snecimen P- 15 Tested at 200'F S ecimen P-16 Tested at 300'F Specimen P-17 Tested at 550'F Figure 5-18 Fractured Tensile Specimens from the Kewaunee, Capsule T Reactor Vessel Intermediate Shell Forging 122X208VA1 (Tangential Orientation).

WCAP-16641-NP Rev. 0 Testing of Specimens from Capsule T

5-34 qnt-rimin Q.-1 1 T~ct~d nt ])';OF Specimen S-13 Tested at 550'F Figure 5-19 Fractured Tensile Specimens from the Kewaunee, Capsule T Reactor Vessel Lower Shell Forging 123XI67VA1 (Tangential Orientation).

Testing of Specimens from Capsule T WCAP- 1664 1-NP Rev. 0

5-35 KEWAUNEE T CAPSULE 100 80

,6 60 Ii)

I-U) 40 P-13 75"F 20 0

0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN KEWAUNEE T CAPSULE 100 80 60 U)

U) 40 P-14 125TF 20 0

0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN Figure 5-20 Engineering Stress-Strain Curves for Kewaunee Reactor Vessel Intermediate Shell Forging 122X208VAI, Capsule T, Tensile Specimens P-13 and P-14.

Rev. 00 Testing of Specimens from Capsule T WCAP-1664 WCAP- 1664 11-NP

-NP Rev. Testing of Specimens from Capsule T

5-36 KEWAUNEE T CAPSULE 100 80

( 60 UJ 40 200*F 20 0

0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN KEWAUNEE T CAPSULE 100 80 60 w

F-40 P-16 300*F 20 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN Figure 5-21 Engineering Stress-Strain Curves for Kewaunee Reactor Vessel Intermediate Shell Forging 122X208VA1, Capsule T, Tensile Specimens P-15 and P-16.

Testing of Specimens from Capsule T WCAP-16641-NP Rev. 0

5-37 KEWAUNEE T CAPSULE 100 80

, 60 U) 40 P-17 550"F 20 0

0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN Figure 5-22 Engineering Stress-Strain Curve for Kewaunee Reactor Vessel Intermediate Shell Forging 122X208VA1, Capsule T, Tensile Specimen P-17.

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5-38 KEWAUNEE T CAPSULE 100 80 60 C,,

LU 40 R-l fl 0

75 F 20 1 0

0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN KEWAUNEE T CAPSULE 100 80 6 80 40 S-11 20 125TF 20 0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN Figure 5-23 Engineering Stress-Strain Curves for Kewaunee Reactor Vessel Lower Shell Forging 123X167VA1, Capsule T, Tensile Specimens S-10 and S-11.

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5-39 KEWAUNEE T CAPSULE 100 80 60 40 S-12 200*F 20 0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN KEWAUNEE T CAPSULE 100 80 60 LU 40 S-13 550°F 20 0 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN Figure 5-24 Engineering Stress-Strain Curves for Kewaunee Reactor Vessel Lower Shell Forging 123X167VA1, Capsule T, Tensile Specimens S-12 and S-13.

Capsule TT from Capsule Specimens from of Specimens WCAP-1664 1-NP Rev. 0 Testing of Testing WCAP-16641-NP Rev. 0

5-40 This page intentionally blank Testing of Specimens from Capsule T WCAP- 1664 1-NP Rev. 0

6-1 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6.1 INTRODUCTION

This section describes a discrete ordinates S,, transport analysis performed for the Kewaunee reactor to determine the neutron radiation environment within the reactor pressure vessel and surveillance capsules.

In this analysis, fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements (dpa) were established on a plant and fuel cycle specific basis. An evaluation of the most recent dosimetry sensor set from Capsule T, withdrawn at the end of the twenty sixth plant operating cycle, is provided. In addition, to provide an up-to-date data base applicable to the Kewaunee reactor, the sensor sets from the previously withdrawn capsules (V, R, P and S) were re-analyzed using the current dosimetry evaluation methodology. The dosimetry evaluations for all capsules withdrawn to date are presented in Appendix A of this report. Comparisons of the results from the dosimetry evaluations with the analytical predictions served to validate the plant specific neutron transport calculations. These validated calculations subsequently formed the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 60 calendar years.

The use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for the development of damage trend curves as well as for the implementation of trend curve data to assess the condition of the vessel.

However, it has also been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves and improved accuracy in the evaluation of damage gradients through the reactor vessel wall.

Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, "Analysis and Interpretation of Light-Water Reactor Surveillance Results," [Reference 23] recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a database for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, "Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom." [Reference 24]

The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials."

All of the calculations and dosimetry evaluations described in this section and in Appendix A were based on the latest available nuclear cross-section data derived from ENDF/B-VI and made use of the latest available calculational tools. Furthermore, the neutron transport and dosimetry evaluation methodologies follow the guidance of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." Reference [25] The NRC approved neutron transport and dosimetry evaluation methodologies used for the Kewaunee application are described in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004 [Reference 26] and WCAP-16083-NP-A, "Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry, May 2006," [Reference 27] respectively.

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6-2 6.2 DISCRETE ORDINATES ANALYSIS A plan view of the Kewaunee reactor geometry at the core midplane is shown in Figure 4-1. Six irradiation capsules attached to the thermal shield are included in the reactor design to constitute the reactor vessel surveillance program. The capsules are located at azimuthal angles of 570, 670, 770, 2370, 247', and 2570 as shown in Figure 4-1. These full core positions correspond to the following octant symmetric locations represented in Figure 6-1: 130 from the core cardinal axes (for the 770 and 2570 capsule holder locations), 230 from the core cardinal axes (for the 670 and 247' capsule holder locations),

and 330 from the core cardinal axes (for the 570 and 2370 capsule holder locations). The stainless steel specimen containers are approximately 1-inch square and 64 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 5.33 feet of the 12-foot high reactor core.

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

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

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

z) q)(r, 0, z) = q(r, 0) * (p(r, T(r) where 4(r,0,z) is the synthesized three-dimensional neutron flux distribution, 4)(r,0) is the transport solution in r,0 geometry, ý(r,z) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and 4b(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the r,0 two-dimensional calculation. This synthesis procedure was carried out for each operating cycle at Kewaunee.

For the Kewaunee transport calculations, the eighth core r,0 model depicted in Figure 6-1 was utilized since the reactor is octant symmetric. The r,0 model included the core, the reactor intemals, the thermal shield, including explicit representations of surveillance capsules, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. This model formed the basis for the calculated results and enabled comparisons to the surveillance capsule dosimetry evaluations. In developing the analytical model, nominal design dimensions were employed for the various structural components. Likewise, water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full power operating conditions. The coolant densities were treated on a fuel cycle specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, et cetera. The geometric mesh description of the r,0 reactor model consisted Radiation Analysis and Neutron Dosimetry WCAP-1664 1-NP Rev. 0

6-3 of 148 radial by 105 azimuthal intervals. Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the r,0 calculations was set at a value of 0.001.

The rz model used for the Kewaunee calculations is shown in Figure 6-2 and extends radially from the centerline of the reactor core out to a location interior to the primary biological shield and over an axial span from an elevation below the lower core plate to above the upper core plate. As in the case of the r,0 model, nominal design dimensions and full power coolant densities were employed in the calculations. In this case, the homogenous core region was treated as an equivalent cylinder with a volume equal to that of the active core zone. The stainless steel former plates located between the core baffle and core barrel regions were also explicitly included in the model. The rz geometric mesh description consisted of 127 radial by 155 axial intervals. As in the case of the rO model, mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the rz calculations was also set at a value of 0.001.

The one-dimensional radial model used in the synthesis procedure consisted of the same 127 radial mesh intervals included in the rz model. Thus, radial synthesis factors could be determined on a meshwise basis throughout the entire geometry.

The power distributions used in the plant specific transport analysis were based on the individual core designs for each of the first twenty seven fuel cycles at Kewaunee. Specifically, the data utilized included cycle dependent fuel assembly initial enrichments, bumups, and axial power distributions. This information was used to develop spatial and energy dependent core source distributions averaged over each fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle averaged neutron flux, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was computed using an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and bumup history of each fuel assemblies. From these assembly dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

All of the transport calculations supporting this analysis were carried out using the DORT discrete ordinates code Version 3.1 [Reference 28] and the BUGLE-96 cross-section library [Reference 29]. The BUGLE-96 library provides a 67 group coupled neutron-gamma ray cross-section data set produced specifically for light water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P5 legendre expansion and angular discretization was modeled with an S1 6 order of angular quadrature. Energy and space dependent core power distributions, as well as system operating temperatures, were treated on a fuel cycle specific basis.

Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-8. In Table 6-1, the calculated exposure rates and integrated exposures, expressed in terms of both neutron fluence (E > 1.0 MeV) and dpa, are given at the radial and azimuthal center of the octant symmetric surveillance capsule positions, i.e., for the 130, 230, and 330 locations. These results, representative of the axial midplane of the active core, establish the calculated exposure of the surveillance capsules withdrawn to date as well as projected into the future. Similar information is provided in Table 6-2 for the reactor Rev. 00 Radiation Analysis and Neutron Dosimetry WCAP-1664 WCAP- 1664 11-NP

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6-4 vessel inner radius at four azimuthal locations. The vessel data given in Table 6-2 were taken at the clad/base metal interface, and thus, represent maximum calculated exposure levels on the vessel.

From the data provided in Table 6-2 it is noted that the peak clad/base metal interface vessel fluence (E > 1.0 MeV) at the end of the twenty sixth fuel cycle (i.e., after 24.6 effective full power years (EFPY) of plant operation) was 2.60x 10"9 n/cm 2.

Both calculated fluence (E > 1.0 MeV) and dpa data are provided in Table 6-1 and Table 6-2. These data tabulations include fuel cycle specific calculated neutron exposures at the end of the twenty sixth fuel cycle (the last completed at Kewaunee) as well as future projections to the end of Cycle 27 (the current operating cycle) and for several intervals extending to 60 calendar years of operation. The calculations account for a core power uprate from 1650 MWt to 1772 MWt that occurred during Cycle 26. The projections were based on the assumption that the core power distributions and associated plant operating characteristics from the Cycle 27 uprated core design were representative of future plant operation. The future projections are also based on the current reactor power level of 1772 MWt.

Radial gradient information applicable to fast (E > 1.0 MeV) neutron fluence and dpa are given in Tables 6-3 and 6-4, respectively. The data, based on the cumulative integrated exposures from Cycles 1 through 26, are presented on a relative basis for each exposure parameter at several azimuthal locations.

Exposure distributions through the vessel wall may be obtained by multiplying the calculated exposure at the vessel inner radius by the gradient data listed in Tables 6-3 and 6-4.

The calculated fast neutron exposures for the five surveillance capsules withdrawn from the Kewaunee reactor are provided in Table 6-5. These assigned neutron exposure levels are based on the plant and fuel cycle specific neutron transport calculations discussed in this section.

From the data provided in Table 6-5 it is noted that Capsule T received a fluence (E > 1.0 MeV) of 5.62x 1019 n/cm 2 after exposure through the end of the twenty sixth fuel cycle (i.e., after 24.6 EFPY of plant operation).

In addition to the Kewaunee surveillance capsule evaluations, the Master Curve Assessment of the Kewaunee pressure vessel weld metal [Reference 5] was supported by fracture toughness data derived from an accelerated capsule (A-35) withdrawn from the Maine Yankee reactor. In order to provide a consistent data base for use in this evaluation, the neutron exposure of the Maine Yankee accelerated capsule was re-calculated using current technology and ENDF/B-VI cross-sections. This re-calculation of the Maine Yankee accelerated capsule utilized the same methodology described earlier in this section. The results of the re-calculation of Maine Yankee Capsule A-35 indicated that the surveillance specimens had been irradiated for a period of 4.5 EFPY and had received a fluence (E > 1.0 MeV) of 6.11 x 10'9 n/cm 2.

Updated lead factors for the Kewaunee surveillance capsules are provided in Table 6-6. The capsule lead factor is defined as the ratio of the calculated fluence (E > 1.0 MeV) at the geometric center of the surveillance capsule to the corresponding maximum calculated fluence at the pressure vessel clad/base metal interface. In Table 6-6, the lead factors for capsules that have been withdrawn from the reactor (V, R, P, S, and T) were based on the calculated fluence values for the irradiation period corresponding to the time of withdrawal for the individual capsules. For the capsule remaining in the reactor (N), the lead Radiation Analysis and Neutron Dosimetry WCAP-16641-NP Rev. 0

6-5 factor corresponds to the calculated fluence values at the end of Cycle 26, the last completed fuel cycle at Kewaunee.

In Table 6-7, fast neutron fluence (E > 1.0 MeV) projections are provided for all materials comprising the extended beltline region of the Kewaunee reactor pressure vessel. Fluence data are provided for the end of Cycle 26 (24.6 EFPY), for 40 calendar years of operation (33.0 EFPY), and for 60 calendar years of operation (52.1 EFPY). The information included in Table 7 shows that the maximum vessel exposures listed in Table 2 of this section occur on the intermediate shell base material with all other vessel materials experiencing a lower neutron exposure. The data in Table 6-7 also shows that through 40 years of operation the primary loop nozzles and associated welds remain below a fluence of 1.OE+ 17 n/cm 2, but the lower regions of the RCS inlet nozzle and both the inlet nozzle and outlet nozzle welds are projected to reach the 1.OE+ 17 n/cm 2 threshold during the 40 - 60 year operating period.

In Table 6-8 projections of maximum fast neutron fluence (E > 1.0 MeV) and gamma ray dose experienced by the primary biological shield are listed. The data are provided as a function of axial position relative to the active core midplane and extending from the bottom to the top of the active fuel.

The data traverses were taken along the 0' azimuth. The exposure data are tabulated for the end of Cycle 26 (24.6 EFPY), for 40 calendar years of operation (33.0 EFPY), and for 60 calendar years of operation (52.1 EFPY).

Neutron exposure projections beyond the end of Cycle 27 were based on an operating scenario that consisted of a series of 18 month operating cycles followed by a 25 day refueling outage. The reactor was considered to be operating at full power for the entire 18 month cycle. This full power period coupled with the 25 day refueling outage resulted in a net capacity factor of 95.6% with a total operating time of 33.0 EFPY at EOL and 52.1 EFPY at EOLE. Both of these operating times are including in the exposure projections tabulated in this section.

The neutron exposure projections were also based on continued use of low leakage fuel management. The specific radial power distribution used for the projections was as follows:

Relative Radial Power Distribution Used for Neutron Exposure Projections 7 8 9 10 11 12 13 G 0.887 1.113 1.265 1.203 1.186 1.235 0.420 H 1.121 1.223 1.208 1.280 1.239 1.145 0.342 I 1.269 1.208 1.284 1.193 1.231 0.563 J 1.208 1.282 1.188 1.215 1.118 0.348 K 1.192 1.246 1.238 1.134 0.487 L 1.242 1.155 0.568 0.353 M 0.421 0.345 WCAP- 1664 1-NP Rev. 0 Radiation Analysis and Neutron Dosimetry

6-6 These values represent the average relative radial power over the fuel cycle and can be estimated from the following equation:

EOCBurnup- BOCBurnup PAvgg CycleA verageBurnup where:

PAvg = Cycle Average Relative Power for an individual fuel assembly.

BOC Burnup = Beginning of Cycle Burnup for the individual fuel assembly.

EOC Burnup = End of Cycle Burnup for the individual fuel assembly.

Cycle Average Burnup = The Cycle Average Burnup for the Entire Core.

Actual future fuel cycle designs can be compared to the power distribution map shown above in order to evaluate actual performance compared to projections. To maintain the validity of the projections, the peripheral power distributions averaged over multiple cycles should be maintained at or below those shown above.

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

The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsule T, that was withdrawn from Kewaunee at the end of the twenty sixth fuel cycle, is summarized below.

Reaction Rates (rps/atom) M/C Reaction Measured Calculated Ratio 63Cu(n,a)60Co 4.76E- 17 4.92E- 17 0.97 54Fe(n,p) 54 Mn 5.20E-15 5.25E-15 0.99 58Ni(n,p)5 SCo 7.22E-15 7.20E-15 1.00 23 8U(n,p) 137Cs (Cd) 2.35E-14 2.52E-14 0.93 237 Np(n,f)137Cs (Cd) 2.11E-13 1.97E-13 1.07 Average: 0.99

% Standard Deviation: 5.2 Radiation Analysis and Neutron Dosimetry WCAP- 1664 1-NP Rev. 0

6-7 The measured-to-calculated (M/C) reaction rate ratios for the Capsule T threshold reactions range from 0.93 to 1.07, and the average M/C ratio is 0.99 +/- 5.2% (1G). This direct comparison falls well within the

+ 20% criterion specified in Regulatory Guide 1.190; furthermore, it is consistent with the full set of comparisons given in Appendix A for all measured dosimetry removed to date from the Kewaunee reactor. These comparisons validate the current analytical results described in Section 6.2; therefore, the calculations are deemed applicable for Kewaunee.

6.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Kewaunee surveillance capsule and reactor pressure vessel is based on the recommended approach provided in Regulatory Guide 1.190.

In particular, the qualification of the methodology was carried out in the following four stages:

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

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

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

4 - Comparisons of the plant specific calculations with all available dosimetry results from the Kewaunee surveillance program.

The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operational or geometric variables that impact power reactor calculations. The second phase of the qualification (H. B. Robinson comparisons) addressed uncertainties in these additional areas that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations.

The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant specific input parameters. The overall calculational uncertainty applicable to the Kewaunee analysis was established from results of these three phases of the methods qualification.

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

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6-8 The following summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference 26.

Capsule Vessel IR PCA Comparisons 3% 3%

H. B. Robinson Comparisons 3% 3%

Analytical Sensitivity Studies 10% 11%

Additional Uncertainty for Factors not Explicitly Evaluated 5% 5%

Net Calculational Uncertainty 12% 13%

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

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

The plant specific measurement comparisons described in Appendix A support these uncertainty assessments for Kewaunee. The least squares data comparisons for the five surveillance capsules withdrawn to date from the Kewaunee reactor indicate an average Adjusted to Calculated (A/C) ratio of 0.99 with a standard deviation of 4.1% and 1.00 with a standard deviation of 4.3% for fluence (E > 1.0 MeV) and iron atom displacements (dpa), respectively. This data comparison is not only self consistent for the Kewaunee reactor, but is in excellent agreement with the overall Westinghouse surveillance capsule dosimetry database which currently includes data evaluations for 125 sets of reactor dosimetry. The overall database comparisons indicate an average Adjusted to Calculated (A/C) ratio of 0.99 with a standard deviation of 7.1% and 1.00 with a standard deviation of 6.6% for fluence (E > 1.0 MeV) and iron atom displacements (dpa), respectively. These comparisons are similar to the Kewaunee specific results.

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6-9 Table 6-1 Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cumulative Cumulative Neutron Flux (E > 1.0 MeV)

Cycle Irradiation Irradiation [n/cm 2-s]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 130 230 330 1 4.06E+07 4.06E+07 1.3 1.44E+ 11 8.10E+10 7.64E+10 2 2.14E+07 6.20E+07 2.0 1.43E+11 8.64E+10 8.43E+10 3 3.24E+07 9.44E+07 3.0 9.14E+ 10 7.63E+10 7.90E+10 4 2.97E+07 1.24E+08 3.9 1.1OE+ll 7.19E+10 6.88E+10 5 2.18E+07 1.46E+08 4.6 1.11E+ll 7.25E+10 7.02E+10 6 2.44E+07 1.70E+08 5.4 1.08E+l I 7.70E+10 7.81E+10 7 2.56E+07 1.96E+08 6.2 1.02E+ 1I 6.91E+10 6.93E+10 8 2.49E+07 2.2 1E+08 7.0 1.05E+11 7.38E+10 7.32E+10 9 2.57E+07 2.47E+08 7.8 1.1OE+ll 7.33E+10 7.01E+10 10 2.33E+07 2.70E+08 8.6 1.02E+ 1I 7.11E+10 6.88E+10 11 2.7 1E+07 2.97E+08 9.4 9.99E+ 10 6.93E+10 6.58E+10 12 2.6 1E+07 3.23E+08 10.2 1.09E+ 11 6.87E+10 6.24E+10 13 2.8 1E+07 3.5 1E+08 11.1 1.08E+1 I 7.48E+10 7.22E+10 14 2.57E+07 3.77E+08 11.9 1.03E+ 11 7.26E+10 6.99E+10 15 2.68E+07 4.04E+08 12.8 1.05E+ 1I 7.30E+10 6.93E+10 16 2.75E+07 4.3 1E+08 13.7 8.33E+10 7.02E+10 7.06E+ 10 17 2.53E+07 4.56E+08 14.5 8.73E+10 7.05E+10 6.91E+10 18 2.67E+07 4.83E+08 15.3 8.47E+10 7.1IE+I0 7.38E+10 19 2.8 1E+07 5.11E+08 16.2 9.28E+10 7.15E+10 6.97E+ 10 20 2.57E+07 5.37E+08 17.0 9.24E+10 7.33E+10 7.47E+10 21 3.88E+07 5.76E+08 18.2 9.21E+10 7.09E+10 7.32E+10 22 3.96E+07 6.15E+08 19.5 1.0IE+11 7.21E+10 6.82E+10 23 4.08E+07 6.56E+08 20.8 9.68E+10 7.33E+10 7.29E+10 24 3.84E+07 6.95E+08 22.0 1.03E+l11 6.87E+10 6.64E+10 25 3.96E+07 7.34E+08 23.3 1.07E+ 11 7.12E+10 6.83E+10 26 4.27E+07 7.77E+08 24.6 1.O1E+l I 6.36E+10 6.07E+ 10 27 4.3 1E+07 8.20E+08 26.0 9.86E+10 6.15E+10 5.94E+ 10 Future 6.37E+07 8.84E+08 28.0 9.86E+10 6.15E+10 5.94E+ 10 Future 1.58E+08 1.04E+09 33.0 9.86E+10 6.15E+10 5.94E+10 Future 9.47E+07 1.14E+09 36.0 9.86E+10 6.15E+10 5.94E+ 10 Future 1.26E+08 1.26E+09 40.0 9.86E+10 6.15E+10 5.94E+ 10 Future 1.26E+08 1.39E+09 44.0 9.86E+10 6.15E+10 5.94E+10 Future 1.26E+08 1.51E+09 48.0 9.86E+10 6.15E+10 5.94E+10 Future 1.29E+08 i .64E+09 52.1 9.86E+10 6.15E+10 5.94E+10 Note: 1) Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

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6-10 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation [n/cm 2]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 13° 230 330 1 4.06E+07 4.06E+07 1.3 5.86E+18 3.29E+18 3.1OE+18 2 2.14E+07 6.20E+07 2.0 8.91E+18 5.14E+ 18 4.91E+18 3 3.24E+07 9.44E+07 3.0 1.19E+ 19 7.61E+18 7.47E+18 4 2.97E+07 1.24E+08 3.9 1.51E+19 9.74E+ 18 9.51E+18 5 2.18E+07 1.46E+08 4.6 1.76E+19 1.13E+ 19 L.10E+19 6 2.44E+07 1.70E+08 5.4 2.02E+19 1.32E+19 1.29E+19 7 2.56E+07 1.96E+08 6.2 2.28E+19 1.50E+ 19 1.47E+ 19 8 2.49E+07 2.21E+08 7.0 2.54E+ 19 1.68E+19 1.65E+19 9 2.57E+07 2.47E+08 7.8 2.82E+ 19 1.87E+19 1.83E+19 10 2.33E+07 2.70E+08 8.6 3.06E+ 19 2.04E+ 19 1.99E+19 11 2.7 1E+07 2.97E+08 9.4 3.33E+19 2.22E+ 19 2.17E+ 19 12 2.6 1E+07 3.23E+08 10.2 3.62E+ 19 2.40E+ 19 2.34E+19 13 2.8 1E+07 3.5 1E+08 11.1 3.92E+ 19 2.61 E+ 19 2.54E+ 19 14 2.57E+07 3.77E+08 11.9 4.18E+19 2.80E+ 19 2.72E+ 19 15 2.68E+07 4.04E+08 12.8 4.46E+19 2.99E+19 2.90E+19 16 2.75E+07 4.3 IE+08 13.7 4.69E+19 3.19E+19 3. 1OE+19 17 2.53E+07 4.56E+08 14.5 4.9 1E+19 3.37E+ 19 3.27E+ 19 18 2.67E+07 4.83E+08 15.3 5.14E+ 19 3.56E+19 3.47E+19 19 2.81 E+07 5.11E+08 16.2 5.40E+ 19 3.76E+19 3.67E+19 20 2.57E+07 5.37E+08 17.0 5.64E+ 19 3.95E+19 3.86E+19 21 3.88E+07 5.76E+08 18.2 6.OOE+19 4.22E+19 4.14E+19 22 3.96E+07 6.15E+08 19.5 6.40E+ 19 4.51 E+ 19 4.41E+19 23 4.08E+07 6.56E+08 20.8 6.79E+19 4.80E+19 4.71E+19 24 3.84E+07 6.95E+08 22.0 7.19E+19 5.07E+19 4.96E+19 25 3.96E+07 7.34E+08 23.3 7.61E+19 5.35E+19 5.23E+19 26 4.27E+07 7.77E+08 24.6 8.04E+19 5.62E+19 5.49E+19 27 4.3 1E+07 8.20E+08 26.0 8.47E+19 5.89E+19 5.75E+19 Future 6.37E+07 8.84E+08 28.0 9.10E+19 6.28E+19 6.13E+ 19 Future 1.58E+08 1.04E+09 33.0 1.07E+20 7.25E+19 7.07E+ 19 Future 9.47E+07 1.14E+09 36.0 1.16E+20 7.83E+19 7.63E+19 Future 1.26E+08 1.26E+09 40.0 1.28E+20 8.61 E+ 19 8.38E+19 Future 1.26E+08 1.39E+09 44.0 1.41E+20 9.38E+19 9.13E+19 Future 1.26E+08 1.51E+09 48.0 1.53E+20 1.02E+20 9.88E+19 Future 1.29E+08 1.64E+09 52.1 1.66E+20 1.1OE+20 1.06E+20 Note: 1) Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

WCAP-16641-NP Rev. 0 Radiation Analysis Radiation Neutron Dosimetry and Neutron Analysis and Dosimetry WCAP- 1664 1-NP Rev. 0

6-11 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cumulative Cumulative Iron Atom Displacement Rate Cycle Irradiation Irradiation _dpa/s]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 130 23- 330 1 4.06E+07 4.06E+07 1.3 2.63E-10 1.41E-10 1.34E-10 2 2.14E+07 6.20E+07 2.0 2.59E-10 1.50E-10 1.48E-10 3 3.24E+07 9.44E+07 3.0 1.65E-10 1.32E-10 1.38E-10 4 2.97E+07 1.24E+08 3.9 1.99E-10 1.25E-10 1.20E-10 5 2.18E+07 1.46E+08 4.6 2.01E-10 1.26E-10 1.23E-10 6 2.44E+07 1.70E+08 5.4 1.96E-10 1.33E-10 1.37E-10 7 2.56E+07 1.96E+08 6.2 1.84E-10 1.20E-10 1.21E-10 8 2.49E+07 2.2 1E+08 7.0 1.90E-10 1.28E-10 1.28E-10 9 2.57E+07 2.47E+08 7.8 1.98E-10 1.27E-10 1.23E-10 10 2.33E+07 2.70E+08 8.6 1.85E-10 1.23E-10 1.21E-10 11 2.7 1E+07 2.97E+08 9.4 1.81E-10 1.20E-10 1.15E-10 12 2.6 IE+07 3.23E+08 10.2 1.96E-10 1.19E-10 1.09E-10 13 2.8 IE+07 3.5 1E+08 11.1 1.95E-10 1.30E-10 1.27E-10 14 2.57E+07 3.77E+08 11.9 1.86E-10 1.26E-10 1.22E-10 15 2.68E+07 4.04E+08 12.8 1.90E-10 1.26E-10 1.21E-10 16 2.75E+07 4.3 1E+08 13.7 1.50E-10 1.21E-10 1.24E-10 17 2.53E+07 4.56E+08 14.5 1.57E-10 1.22E-10 1.21E-10 18 2.67E+07 4.83E+08 15.3 1.52E-10 1.23E-10 1.29E-10 19 2.8 1E+07 5.11E+08 16.2 1.67E-10 1.24E-10 1.22E-10 20 2.57E+07 5.37E+08 17.0 1.67E-10 1.27E-10 1.31E-10 21 3.88E+07 5.76E+08 18.2 1.66E-10 1.23E-10 1.28E-10 22 3.96E+07 6.15E+08 19.5 1.82E-10 1.25E-10 1.20E-10 23 4.08E+07 6.56E+08 20.8 1.75E-10 1.27E-10 1.28E-10 24 3.84E+07 6.95E+08 22.0 1.87E-10 1.19E-10 1.16E- 10 25 3.96E+07 7.34E+08 23.3 1.94E-10 1.23E-10 1.20E-10 26 4.27E+07 7.77E+08 24.6 1.82E-10 1.1OE-10 1.06E-10 27 4.3 1E+07 8.20E+08 26.0 1.78E-10 1.06E-10 1.04E- 10 Future 6.37E+07 8.84E+08 28.0 1.78E-10 1.06E-10 1.04E-10 Future 1.58E+08 1.04E+09 33.0 1.78E-10 1.06E-10 1.04E-10 Future 9.47E+07 1.14E+09 36.0 1.78E-10 1.06E-10 1.04E-10 Future 1.26E+08 1.26E+09 40.0 1.78E-10 1.06E-10 1.04E-10 Future 1.26E+08 1.39E+09 44.0 1.78E-10 1.06E-10 1.04E-10 Future 1.26E+08 1.51E+09 48.0 1.78E-10 1.06E-10 1.04E-10 Future 1.29E+08 1.64E+09 52.1 1.78E-10 1.06E-10 1.04E-10 Note: 1) Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

WCAP-16641-NP Rev. 00 Radiation Analysis and Neutron Dosimetry WCAP- 1664 1 -NP Rev. Radiation Analysis and Neutron Dosimetry

6-12 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cumulative Cumulative Iron Atom Displacements Cycle Irradiation Irradiation [dpa]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 13- 230 330 1 4.06E+07 4.06E+07 1.3 1.07E-02 5.72E-03 5.44E-03 2 2.14E+07 6.20E+07 2.0 1.62E-02 8.94E-03 8.6 1E-03 3 3.24E+07 9.44E+07 3.0 2.15E-02 1.32E-02 1.31E-02 4 2.97E+07 1.24E+08 3.9 2.74E-02 1.69E-02 1.67E-02 5 2.18E+07 1.46E+08 4.6 3.18E-02 1.97E-02 1.93E-02 6 2.44E+07 1.70E+08 5.4 3.66E-02 2.29E-02 2.27E-02 7 2.56E+07 1.96E+08 6.2 4.13E-02 2.60E-02 2.58E-02 8 2.49E+07 2.21E+08 7.0 4.60E-02 2.92E-02 2.90E-02 9 2.57E+07 2.47E+08 7.8 5.11E-02 3.24E-02 3.2 1E-02 10 2.33E+07 2.70E+08 8.6 5.54E-02 3.53E-02 3.49E-02 11 2.7 1E+07 2.97E+08 9.4 6.03E-02 3.85E-02 3.8 1E-02 12 2.6 1E+07 3.23E+08 10.2 6.55E-02 4.16E-02 4.09E-02 13 2.8 1E+07 3.5 1E+08 11.1 7.09E-02 4.53E-02 4.45E-02 14 2.57E+07 3.77E+08 11.9 7.57E-02 4.85E-02 4.76E-02 15 2.68E+07 4.04E+08 12.8 8.08E-02 5.19E-02 5.09E-02 16 2.75E+07 4.3 IE+08 13.7 8.49E-02 5.52E-02 5.43E-02 17 2.53E+07 4.56E+08 14.5 8.89E-02 5.83E-02 5.73E-02 18 2.67E+07 4.83E+08 15.3 9.30E-02 6.16E-02 6.08E-02 19 2.8 1E+07 5.11E+08 16.2 9.77E-02 6.51E-02 6.42E-02 20 2.57E+07 5.37E+08 17.0 1.02E-01 6.83E-02 6.76E-02 21 3.88E+07 5.76E+08 18.2 1.08E-01 7.3 1E-02 7.25E-02 22 3.96E+07 6.15E+08 19.5 1.16E-01 7.80E-02 7.73E-02 23 4.08E+07 6.56E+08 20.8 1.23E-0 I 8.32E-02 8.25E-02 24 3.84E+07 6.95E+08 22.0 1.30E-01 8.77E-02 8.69E-02 25 3.96E+07 7.34E+08 23.3 1.38E-01 9.26E-02 9.17E-02 26 4.27E+07 7.77E+08 24.6 1.45E-01 9.73E-02 9.62E-02 27 4.3 IE+07 8.20E+08 26.0 1.53E-01 1.02E-01 1.O1E-01 Future 6.37E+07 8.84E+08 28.0 1.64E-01 1.09E-01 1.07E-01 Future 1.58E+08 1.04E+09 33.0 1.93E-01 1.26E-01 1.24E-01 Future 9.47E+07 1.14E+09 36.0 2.09E-0 1 1.36E-01 1.34E-01 Future 1.26E+08 1.26E+09 40.0 2.32E-01 1.49E-0 I 1.47E-01 Future 1.26E+08 1.39E+09 44.0 2.54E-01 1.62E-01 1.60E-01 Future 1.26E+08 1.51E+09 48.0 2.77E-01 1.76E-01 1.73E-01 Future 1.29E+08 1.64E+09 52.1 3.OOE-01 1.90E-01 1.86E-01 Note: 1) Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Neutron Dosimetry and Neutron WCAP-1664 1-NP Rev. 0 Radiation Analysis Radiation Analysis and Dosimetry WCAP- 1664 1-NP Rev. 0

6-13 Table 6-2 Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Neutron Flux (E > 1.0 MeV)

Cycle Irradiation Irradiation [n/cm 2-s]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 150 300 450 1 4.06E+07 4.06E+07 1.3 4.76E+ 10 2.84E+10 1.88E+10 1.61E+10 2 2.14E+07 6.20E+07 2.0 4.61E+10 2.84E+ 10 2.06E+10 1.77E+10 3 3.24E+07 9.44E+07 3.0 3.13E+10 2.OOE+10 1.93E+10 1.66E+10 4 2.97E+07 1.24E+08 3.9 3.56E+10 2.25E+10 1.71E+10 1.45E+10 5 2.18E+07 1.46E+08 4.6 3.72E+ 10 2.28E+10 1.73E+10 1.51E+10 6 2.44E+07 1.70E+08 5.4 3.66E+10 2.26E+10 1.90E+10 1.68E+10 7 2.56E+07 1.96E+08 6.2 3.45E+10 2.I1E+10 1.70E+10 1.50E+10 8 2.49E+07 2.2 1E+08 7.0 3.51E+10 2.19E+10 1.79E+10 1.56E+10 9 2.57E+07 2.47E+08 7.8 3.77E+10 2.26E+10 1.74E+10 1.42E+10 10 2.33E+07 2.70E+08 8.6 3.40E+10 2.13E+10 1.70E+10 1.41E+10 11 2.7 1E+07 2.97E+08 9.4 3.32E+10 2.08E+10 1.64E+ 10 1.39E+10 12 2.6 1E+07 3.23E+08 10.2 3.72E+10 2.23E+10 1.57E+10 1.31 E+ 10 13 2.8 1E+07 3.5 1E+08 11.1 3.56E+10 2.24E+10 1.79E+10 1.47E+ 10 14 2.57E+07 3.77E+08 11.9 3.46E+10 2.16E+ 10 1.74E+10 1.42E+ 10 15 2.68E+07 4.04E+08 12.8 3.5 1E+10 2.19E+ 10 1.73E+10 1.39E+10 16 2.75E+07 4.3 1E+08 13.7 2.40E+10 1.82E+10 1.74E+10 1.43E+10 17 2.53E+07 4.56E+08 14.5 2.53E+10 1.89E+10 1.72E+10 1.41E+10 18 2.67E+07 4.83E+08 15.3 2.47E+ 10 1.84E+10 1.80E+10 1.54E+10 19 2.8 1E+07 5.11E+08 16.2 2.67E+10 1.99E+10 1.73E+10 1.43E+10 20 2.57E+07 5.37E+08 17.0 2.72E+10 1.98E+10 1.82E+10 1.60E+10 21 3.88E+07 5.76E+08 18.2 2.84E+10 1.96E+10 1.79E+10 1.56E+10 22 3.96E+07 6.15E+08 19.5 3.21E+10 2.12E+10 1.69E+10 1.36E+10 23 4.08E+07 6.56E+08 20.8 3.01E+10 2.06E+10 1.81E+10 1.53E+10 24 3.84E+07 6.95E+08 22.0 3.3 6E+ 10 2.14E+ 10 1.64E+10 1.42E+10 25 3.96E+07 7.34E+08 23.3 3.46E+ 10 2.22E+ 10 1.69E+10 1.41E+10 26 4.27E+07 7.77E+08 24.6 3.23E+10 2.06E+ 10 1.52E+10 1.36E+10 27 4.3 1E+07 8.20E+08 26.0 3.20E+ 10 2.03E+10 1.46E+10 1.39E+10 Future 6.37E+07 8.84E+08 28.0 3.20E+ 10 2.03E+10 1.48E+10 1.38E+10 Future 1.58E+08 1.04E+09 33.0 3.20E+ 10 2.03E+10 1.48E+10 1.38E+10 Future 9.47E+07 1.14E+09 36.0 3.20E+10 2.03E+10 1.48E+ 10 1.38E+10 Future 1.26E+08 1.26E+09 40.0 3.20E+10 2.03E+10 1.48E+10 1.38E+10 Future 1.26E+08 1.39E+09 44.0 3.20E+ 10 2.03E+10 1.48E+10 1.38E+10 Future 1.26E+08 1.51E+09 48.0 3.20E+10 2.03E+10 1.48E+10 1.38E+10 Future 1.29E+08 1.64E+09 52.1 3.20E+10 2.03E+10 1.48E+10 1.38E+10 WCAP-16641-NP Rev. 0 Radiation Analysis and Neutron Dosimetry

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

Cycle Irradiation Irradiation [n/cm2]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 150 300 450 1 4.06E+07 4.06E+07 1.3 1.93E+18 1.15E+18 7.64E+ 17 6.55E+17 2 2.14E+07 6.20E+07 2.0 2.92E+18 1.76E+18 1.20E+18 1.03E+18 3 3.24E+07 9.44E+07 3.0 3.93E+18 2.41E+18 1.83E+18 1.57E+18 4 2.97E+07 1.24E+08 3.9 4.99E+18 3.08E+18 2.34E+ 18 2.OOE+18 5 2.18E+07 1.46E+08 4.6 5.80E+ 18 3.57E+18 2.71E+18 2.33E+18 6 2.44E+07 1.70E+08 5.4 6.69E+18 4.12E+18 3.18E+ 18 2.74E+18 7 2.56E+07 1.96E+08 6.2 7.57E+18 4.66E+ 18 3.61E+18 3.13E+18 8 2.49E+07 2.21E+08 7.0 8.45E+18 5.21E+18 4.06E+ 18 3.52E+18 9 2.57E+07 2.47E+08 7.8 9.42E+18 5.79E+18 4.51E+18 3.88E+18 10 2.33E+07 2.70E+08 8.6 1.02E+19 6.29E+ 18 4.90E+ 18 4.21E+18 11 2.7 1E+07 2.97E+08 9.4 1.11E+19 6.85E+18 5.35E+18 4.59E+18 12 2.6 1E+07 3.23E+08 10.2 1.21E+19 7.43E+18 5.76E+ 18 4.93E+18 13 2.8 1E+07 3.5 1E+08 11.1 1.3 1E+19 8.06E+ 18 6.26E+ 18 5.34E+ 18 14 2.57E+07 3.77E+08 11.9 1.40E+ 19 8.61E+18 6.71E+18 5.71E+18 15 2.68E+07 4.04E+08 12.8 1.49E+ 19 9.20E+ 18 7.17E+ 18 6.08E+18 16 2.75E+07 4.31E+08 13.7 1.56E+19 9.70E+18 7.65E+18 6.47E+ 18 17 2.53E+07 4.56E+08 14.5 1.62E+19 1.02E+19 8.08E+18 6.83E+18 18 2.67E+07 4.83E+08 15.3 1.69E+19 1.07E+19 8.56E+18 7.24E+ 18 19 2.8 IE+07 5.11E+08 16.2 1.76E+19 1.12E+19 9.05E+ 18 7.64E+ 18 20 2.57E+07 5.37E+08 17.0 1.83E+19 1.17E+19 9.52E+18 8.05E+ 18 21 3.88E+07 5.76E+08 18.2 1.94E+19 1.25E+19 1.02E+19 8.65E+ 18 22 3.96E+07 6.15E+08 19.5 2.07E+ 19 1.33E+19 1.09E+ 19 9.19E+ 18 23 4.08E+07 6.56E+08 20.8 2.19E+19 1.42E+ 19 1.16E+19 9.82E+18 24 3.84E+07 6.95E+08 22.0 2.32E+19 1.50E+19 1.23E+19 1.04E+ 19 25 3.96E+07 7.34E+08 23.3 2.46E+ 19 1.59E+19 1.29E+19 1.09E+19 26 4.27E+07 7.77E+08 24.6 2.60E+ 19 1.68E+19 1.36E+19 1.15E+19 27 4.3 1E+07 8.20E+08 26.0 2.73E+19 1.76E+19 1.42E+ 19 1.21E+19 Future 6.37E+07 8.84E+08 28.0 2.94E+ 19 1.89E+19 1.51E+19 1.30E+19 Future 1.58E+08 1.04E+09 33.0 3.44E+ 19 2.21E+19 1.75E+19 1.52E+19 Future 9.47E+07 1.14E+09 36.0 3.75E+19 2.41E+19 1.89E+ 19 1.65E+19 Future 1.26E+08 1.26E+09 40.0 4.15E+19 2.66E+19 2.07E+19 1.82E+19 Future 1.26E+08 1.39E+09 44.0 4.56E+19 2.92E+19 2.26E+ 19 2.OOE+19 Future 1.26E+08 1.51E+09 48.0 4.96E+19 3.18E+19 2.45E+19 2.17E+19 Future 1.29E+08 1.64E+09 52.1 5.37E+19 3.44E+ 19 2.64E+19 2.35E+19 Radiation Analysis and Neutron Dosimetry WCAP- 1664 1-NP Rev. 0

6-15 Table 6-2 cont'd Calculated Azimuthal Variation Of Fast Neutron Exposure Rates And Iron Atom Displacement Rates At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacement Rate Cycle Irradiation Irradiation [dpa/s] __450 Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 150 300 450 I 4.06E+07 4.06E+07 1.3 7.75E- 11 4.79E- 11 3.09E- 11 2.62E- 11 2 2.14E+07 6.20E+07 2.0 7.51E-11 4.79E-l1 3.38E-11 2.87E-11 3 3.24E+07 9.44E+07 3.0 5.09E-11 3.35E-11 3.15E-11 2.69E-11 4 2.97E+07 1.24E+08 3.9 5.80E-11 3.78E-I1 2.80E-11 2.35E-11 5 2.18E+07 1.46E+08 4.6 6.05E-11 3.83E-l1 2.83E-11 2.45E-11 6 2.44E+07 1.70E+08 5.4 5.92E-l1 3.80E-l1 3.12E-11 2.73E-11 7 2.56E+07 1.96E+08 6.2 5.63E-11 3.54E-11 2.78E-11 2.43E-11 8 2.49E+07 2.21E+08 7.0 5.74E-11 3.68E-11 2.94E-11 2.53E-11 9 2.57E+07 2.47E+08 7.8 6.lIE-1l 3.80E-11 2.86E-11 2.31E-11 10 2.33E+07 2.70E+08 8.6 5.58E-11 3.57E-11 2.79E-11 2.28E-11 11 2.71E+07 2.97E+08 9.4 5.39E-11 3.51E-11 2.69E-11 2.25E-11 12 2.61E+07 3.23E+08 10.2 6.05E-11 3.75E-11 2.58E-11 2.12E-11 13 2.81E+07 3.51E+08 11.1 5.80E-11 3.77E-11 2.94E-11 2.38E-11 14 2.57E+07 3.77E+08 11.9 5.60E-11 3.62E-11 2.84E-I1 2.30E-11 15 2.68E+07 4.04E+08 12.8 5.71E-11 3.66E-11 2.84E-I1 2.25E-11 16 2.75E+07 4.31E+08 13.7 3.93E-11 3.05E-11 2.84E-l1 2.30E-11 17 2.53E+07 4.56E+08 14.5 4.11E-11 3.16E-11 2.81E-l1 2.29E-11 18 2.67E+07 4.83E+08 15.3 4.01E-ll 3.07E-11 2.96E-11 2.51E-11 19 2.81E+07 5.IIE+08 16.2 4.38E-11 3.35E-I1 2.81E-11 2.31E-11 20 2.57E+07 5.37E+08 17.0 4.44E-11 3.31E-I1 3.OOE-11 2.57E-11 21 3.88E+07 5.76E+08 18.2 4.61E-11 3.30E-11 2.91E-11 2.53E-11 22 3.96E+07 6.15E+08 19.5 5.23E-11 3.56E-11 2.80E-11 2.20E-11 23 4.08E+07 6.56E+08 20.8 4.90E-11 3.43E-I1 2.94E-11 2.50E-11 24 3.84E+07 6.95E+08 22.0 5.47E-1I 3.59E-I1 2.68E-11 2.29E-11 25 3.96E+07 7.34E+08 23.3 5.63E- 11 3.71E-1l 2.80E- 11 2.27E-11 26 4.27E+07 7.77E+08 24.6 5.27E-11 3.49E-11 2.48E-11 2.22E-11 27 4.3 1E+07 8.20E+08 26.0 5.22E-11 3.43E-11 2.41E-11 2.23E-11 Future 6.37E+07 8.84E+08 28.0 5.21E-11 3.41E-11 2.42E-I1 2.24E-11 Future 1.58E+08 1.04E+09 33.0 5.21E-11 3.41E-11 2.42E-I1 2.24E-11 Future 9.47E+07 1.14E+09 36.0 5.21E-11 3.41E-11 2.42E-I1 2.24E-l1 Future 1.26E+08 1.26E+09 40.0 5.21E-11 3.41E-11 2.42E-11 2.24E-11 Future 1.26E+08 1.39E+09 44.0 5.21E-11 3.41E-11 2.42E-I1 2.24E-11 Future 1.26E+08 1.51E+09 48.0 5.21E-11 3.41E-11 2.42E-11 2.24E-11 Future 1.29E+08 1.64E+09 52.1 5.21E-11 3.41E-11 2.42E-11 2.24E-11 Rev. 00 Radiation Analysis and Neutron Dosimetry WCAP- 1664 11-NP WCAP-1664 -NP Rev. Radiation Analysis and Neutron Dosimetry

6-16 Table 6-2 cont'd Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacements Cycle Irradiation Irradiation [dpa]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 15° 30° 45_

1 4.06E+07 4.06E+07 1.3 3.15E-03 1.95E-03 1.26E-03 1.06E-03 2 2.14E+07 6.20E+07 2.0 4.76E-03 2.97E-03 1.98E-03 1.68E-03 3 3.24E+07 9.44E+07 3.0 6.40E-03 4.05E-03 3.OOE-03 2.55E-03 4 2.97E+07 1.24E+08 3.9 8.13E-03 5.18E-03 3.83E-03 3.25E-03 5 2.18E+07 1.46E+08 4.6 9.45E-03 6.01E-03 4.45E-03 3.78E-03 6 2.44E+07 1.70E+08 5.4 1.09E-02 6.94E-03 5.21E-03 4.45E-03 7 2.56E+07 1.96E+08 6.2 1.23E-02 7.85E-03 5.92E-03 5.07E-03 8 2.49E+07 2.21E+08 7.0 1.38E-02 8.76E-03 6.66E-03 5.70E-03 9 2.57E+07 2.47E+08 7.8 1.53E-02 9.74E-03 7.39E-03 6.29E-03 10 2.33E+07 2.70E+08 8.6 1.66E-02 1.06E-02 8.04E-03 6.82E-03 11 2.7 1E+07 2.97E+08 9.4 1.8 1E-02 1.15E-02 8.77E-03 7.43E-03 12 2.61E+07 3.23E+08 10.2 1.97E-02 1.25E-02 9.44E-03 7.99E-03 13 2.81E+07 3.51E+08 11.1 2.13E-02 1.36E-02 1.03E-02 8.65E-03 14 2.57E+07 3.77E+08 11.9 2.27E-02 1.45E-02 1.10E-02 9.25E-03 15 2.68E+07 4.04E+08 12.8 2.43E-02 1.55E-02 1.18E-02 9.85E-03 16 2.75E+07 4.31E+08 13.7 2.54E-02 1.63E-02 1.25E-02 1.05E-02 17 2.53E+07 4.56E+08 14.5 2.64E-02 1.71E-02 1.33E-02 1. 1IIE-02 18 2.67E+07 4.83E+08 15.3 2.75E-02 1.79E-02 1.40E-02 1. 17E-02 19 2.8 1E+07 5.11E+08 16.2 2.87E-02 1.89E-02 1.48E-02 1.24E-02 20 2.57E+07 5.37E+08 17.0 2.98E-02 1.97E-02 1.56E-02 1.30E-02 21 3.88E+07 5.76E+08 18.2 3.16E-02 2. 1OE-02 1.67E-02 1.40E-02 22 3.96E+07 6.15E+08 19.5 3.37E-02 2.24E-02 1.78E-02 1.49E-02 23 4.08E+07 6.56E+08 20.8 3.57E-02 2.38E-02 1.90E-02 1.59E-02 24 3.84E+07 6.95E+08 22.0 3.78E-02 2.52E-02 2.01E-02 1.68E-02 25 3.96E+07 7.34E+08 23.3 4.OOE-02 2.67E-02 2.12E-02 1.77E-02 26 4.27E+07 7.77E+08 24.6 4.23E-02 2.82E-02 2.22E-02 1.86E-02 27 4.3 1E+07 8.20E+08 26.0 4.45E-02 2.96E-02 2.33E-02 1.96E-02 Future 6.37E+07 8.84E+08 28.0 4.78E-02 3.18E-02 2.48E-02 2.10E-02 Future 1.58E+08 1.04E+09 33.0 5.6 1E-02 3.72E-02 2.87E-02 2.46E-02 Future 9.47E+07 1.14E+09 36.0 6. 1OE-02 4.04E-02 3.09E-02 2.67E-02 Future 1.26E+08 1.26E+09 40.0 6.76E-02 4.47E-02 3.40E-02 2.95E-02 Future 1.26E+08 1.39E+09 44.0 7.4 1E-02 4.90E-02 3.71E-02 3.23E-02 Future 1.26E+08 1.51E+09 48.0 8.07E-02 5.33E-02 4.01E-02 3.5 1E-02 Future 1.29E+08 1.64E+09 52.1 8.74E-02 5.77E-02 4.32E-02 3.80E-02 Dosimetry WCAP-1664 1-NP Rev. 0 Radiation and Neutron Analysis and Radiation Analysis Neutron Dosimetry WCAP- 1664 1-NP Rev. 0

6-17 Table 6-3 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV)

Within The Reactor Vessel Wall Radius Azimuthal Angle (cm) 00 150 300 450 168.04 1.000 1.000 1.000 1.000 172.25 0.638 0.645 0.634 0.646 176.46 0.364 0.375 0.364 0.374 180.66 0.200 0.211 0.203 0.209 184.87 0.103 0.116 0.112 0.114 Base Metal Inner Radius = 168.04 Base Metal 1/4T = 172.25 Base Metal 1/2T = 176.46 Base Metal 3/4T = 180.66 Base Metal Outer Radius = 184.87 Note: Relative radial distribution data are based on the cumulative integrated exposures from Cycles 1through 26.

Table 6-4 Relative Radial Distribution Of Iron Atom Displacements (dpa)

Within The Reactor Vessel Wall Radius Azimuthal Angle (cm) 00 150 300 450 168.04 1.000 1.000 1.000 1.000 172.25 0.707 0.724 0.710 0.718 176.46 0.472 0.497 0.481 0.487 180.66 0.304 0.334 0.321 0.322 184.87 0.180 0.214 0.207 0.206 Base Metal Inner Radius = 168.04 Base Metal 1/4T = 172.25 Base Metal 1/2T = 176.46 Base Metal 3/4T = 180.66 Base Metal Outer Radius = 184.87 Note: Relative radial distribution data are based on the cumulative integrated exposures from Cycles I through 26.

WCAP-16641-NP Rev. 0 Radiation Analysis and Neutron Dosimetry

6-18 Table 6-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Kewaunee Irradiation Time Fluence (E > 1.0 MeV) Iron Displacements Capsule [EFPY] [n/cm 2] [dpa]

V 1.3 5.86E+18 1.07E-02 R 4.6 1.76E+19 3.18E-02 P 11.1 2.61E+19 4.53E-02 S 16.2 3.67E+ 19 6.42E-02 T 24.6 5.62E+ 19 9.73E-02 Table 6-6 Calculated Surveillance Capsule Lead Factors Capsule ID And Location Status Lead Factor(a)

V (130) Withdrawn EOC 1 3.03 R (130) Withdrawn EOC 5 3.03 P (230) Withdrawn EOC 13 2.00 S (330) Withdrawn EOC 19 2.08 T (230) Withdrawn EOC 26 2.17 N (330) In Reactor 2.12 Note: (a) Lead factor for capsule remaining in the reactor is based on cycle specific exposure calculations through the last completed fuel cycle, i.e., Cycle 26.

Radiation Analysis and Neutron Dosimetry WCAP- 1664 1-NP Rev. 0

6-19 Table 6-7 Calculated Maximum Neutron Fluence (E > 1.0 MeV) for the Kewaunee Extended Beltline Materials Neutron Fluence (E > 1.0 MeV) [n/cm2 ]

24.6 EFPY 33.0 EFPY 52.1 EFPY Lower Shell to Lower Closure Head Weld < 1.00E+17 < 1.00E+17 < 1.00E+17 Lower Shell 2.57E+19 3.38E+19 5.22E+ 19 Lower Shell to Intermediate Shell Weld 2.47E+ 19 3.27E+19 5.07E+19 Intermediate Shell 2.60E+ 19 3.44E+ 19 5.37E+19 Intermediate Shell to Upper Shell Weld 2.58E+18 3.42E+ 18 5.33E+18 Upper Shell 2.58E+18 3.42E+ 18 5.33E+18 RCS Inlet Nozzle to Upper Shell Weld < 1.00E+17 < 1.00E+17 1.34E+17 RCS Inlet Nozzle < 1.00E+17 < 1.00E+17 1.20E+17 RCS Outlet Nozzle to Upper Shell Weld < 1.00E+17 < 1.00E+17 1.10E+17 RCS Outlet Nozzle < 1.00E+17 < 1.00E+17 < 1.00E+17 Safety Injection Nozzle < 1.00E+17 < 1.00E+17 < 1.00E+17 Vessel Support Bracket < 1.00E+17 < 1.00E+17 < 1.00E+17 Core Support Guide Lugs < 1.00E+17 < 1.00E+17 < 1.00E+17 Table 6-8 Calculated Maximum Neutron and Gamma Ray Exposure of the Primary Biological Shield Distance Neutron (E > 1.0 MeV) Fluence [n/cm 2] Gamma Ray Dose rad]

from Core Midplane

[ft] 24.6 EFPY 33.0 EFPY 52.1 EFPY 24.6 EFPY 33.0 EFPY 52.1 EFPY

-6.0 6.14E+17 7.86E+17 1.18E+18 2.58E+09 3.30E+09 4.93E+09

-5.5 9.57E+17 1.23E+18 1.86E+18 4.07E+09 5.23E+09 7.85E+09

-4.5 1.46E+18 1.90E+18 2.90E+18 6.29E+09 8.14E+09 1.24E+10

-3.5 1.69E+18 2.21E+18 3.40E+18 7.53E+09 9.81E+09 1.50E+10

-2.5 1.82E+18 2.39E+18 3.69E+18 8.17E+09 1.07E+10 1.64E+10

-1.5 1.85E+18 2.44E+18 3.79E+18 8.46E+09 1.11E+10 1.71E+10

-0.5 1.87E+18 2.48E+18 3.85E+18 8.59E+09 1.13E+10 1.75E+10 0.0 1.88E+18 2.49E+18 3.88E+18 8.61E+09 1.13E+10 1.75E+10 0.5 1.87E+18 2.49E+18 3.87E+18 8.59E+09 1.13E+10 1.75E+10 1.5 1.83E+18 2.43E+18 3.79E+18 8.42E+09 1.11E+10 1.72E+10 2.5 1.79E+18 2.37E+18 3.69E+18 8.11E+09 1.07E+10 1.65E+10 3.5 1.66E+18 2.19E+18 3.41E+18 7.44E+09 9.79E+09 1.51E+10 4.5 1.41E+18 1.85E+18 2.87E+18 6.25E+09 8.20E+09 1.26E+10 5.5 9.28E+17 1.21E+18 1.86E+18 4.34E+09 5.67E+09 8.67E+09 6.0 6.58E+17 8.57E+17 1.31E+18 3.26E+09 4.24E+09 6.46E+09 WCAP-16641-NP Rev. 0 Radiation Analysis and Neutron Dosimetry

6-20 Figure 6-1 Kewaunee r,0 Reactor Geometry at the Core Midplane oo P00

-- 0 ccE 0

S0 0 Radiation Analysis and Neutron Dosimetry WCAP-16641-NP Rev. 0

6-21 Figure 6-2 Kewaunee rz Reactor Geometry 4

40 80 120 160 200 240 R Axis (cm)

Rev. 00 Radiation Analysis and Neutron Dosimetry WCAP-16641-NP Rev.

WCAP-16641-NP Radiation Analysis and Neutron Dosimetry

6-22 This page intentionally blank Radiation Analysis and Neutron Dosimetry WCAP-1664 1-NP Rev. 0

7-1 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following surveillance capsule removal schedule meets the requirements of ASTM E 185-82

[Reference 13] and is recommended for future capsules to be removed from the KPS reactor vessel.

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

V 130 3.03 1.3 5.86E+18 R 130 3.03 4.6 1.76E+19 P 230 2.00 11.1 2.61E+19 S 330 2.08 16.2 3.67E+ 19 T 230 2.17 24.6 5.62E+19 N 330 2.12 (d) (d)

Notes:

(a) Updated in Capsule T dosimetry analysis.

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

(c) Actual plant evaluation calculated fluence.

(d) This capsule will reach a fluence of approximately 7.4 x 1019 (72 EFPY Peak Fluence) which occurs at 29.44 EFPY. It is recommended that this standby capsule be withdrawn between 29 and 30 EFPY and placed in storage. Since it is the last capsule to be removed, other means of monitoring vessel fluence should be implemented prior to its removal (e.g., ex-vessel dosimetry).

Surveillance Capsule Removal Schedule WCAP-1664 WCAP- Rev. 0 1-NP Rev.

1664 1-NP 0 Surveillance Capsule Removal Schedule

7-2 This page intentionally blank WCAP-1664 1-NP Rev. 0 Surveillance Capsule Removal Schedule Capsule Removal Schedule WCAP- 1664 1-NP Rev. 0

8-1 8 REFERENCES

1. Regulatory Guide 1.99, Revision 2, RadiationEmbrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May, t988.

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

3. WCAP-8107, Wisconsin PublicService CorporationKewaunee Nuclear PowerPlantReactor Vessel Radiation SurveillanceProgram,S.E. Yanichko, et al, April 1973.

4. ASTM E185-73, Annual Book of ASTM Standards, StandardPracticefor ConductingSurveillance Tests for Light- Water Cooled Nuclear PowerReactor Vessels.

5. WCAP- 16609-NP, Master Curve Assessment of the Dominion Energy Kewaunee Power Station Reactor Pressure Vessel Weld Metal, R.G. Lott, et al, dated August 2006.

6. ASTM E 1253-99, Annual Book of ASTM Standards, StandardGuide for Reconstitution of Irradiated Charpy-SizedSpecimen.

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

8. ASTM E208, Annual Book of Standards, Standard Test Method for ConductingDrop-Weight Test to DetermineNil-Ductility Transition Temperature of FerriticSteels.

9. ASTM E 1820, Annual Book of Standards, Standard Test Method for Measurement of Fracture Toughness.

10. K. Wallin, The Scatter in K 1, Results, Engineering Fracture Mechanics, Vol. 19, pp. 1085-1093, 1984.

11. ASTM E 1921, Annual Book of Standards, Standard Test Method for Determinationof Reference Temperature, To, for FerriticSteels in the TransitionRegion.

12. WCAP- 15074, Revision 1, Evaluationof the 1P3571 Weld Metal from the Surveillance Programsfor Kewaunee and Maine Yankee, W.L. Server, et al, dated August 2006.

13. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, StandardPracticefor ConductingSurveillance Tests for Light-Water Cooled NuclearPower Reactor Vessels.

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

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

References WCAP-16641-NP Rev. 00 WCAP-1664~-NP Rev. References

8-2

16. Procedure RMF 8103, CharpyImpact Testing, Revision 2.

17. ASTM E23-02a, Annual Book of ASTM Standards, StandardTest Method for Notched BarImpact Testing of Metallic Materials.

18. ASTM A370-97a, Annual Book of ASTM Standards, StandardTest Methods and Definitionsfor Mechanical Testing of Steel Products.

19. ASTM E8-04, Annual Book of ASTM Standards, StandardTest Methods for Tension Testing of Metallic Materials,ASTM.

20. ASTM E21-03 (1998), Annual Book of ASTM Standards, Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials,ASTM.

21. ASTM E83-93, Annual Book of ASTM Standards, StandardPracticefor Verification and Classificationof Extensometers.

22. Safety Evaluation by the Office of Nuclear Reactor, Kewaunee NuclearPowerPlant,Exemption from the Requirements of 10 CFR Part50, Appendix G, Appendix H, and Section 50.61, Docket No. 50-305, May 2001.

23. ASTM E853, Annual Book of ASTM Standards, StandardPracticefor Analysis and Interpretationof Light- Water Reactor Vessel Surveillance Results, E 706.

24. ASTM E693, Annual Book of ASTM Standards, StandardPracticefor CharacterizingNeutron Exposures in Iron andLow Alloy Steels in Terms of Displacementsper Atom, E 706.

25. Regulatory Guide RG- 1.190, Calculationaland DosimetryMethods for DeterminingPressure Vessel Neutron Fluence, U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

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

27. WCAP-16083-NP-A, Revision 0, Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water ReactorDosimetry May 2006.

28. RSICC Computer Code Collection CCC-650, DOORS 3.2, One, Two- and Three-Dimensional Discrete Ordinates Neutron/Photon TransportCode System, April 1998.

29. RSIC Data Library Collection DLC-185, BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications, March 1996 WCAP-1664 1-NP Rev. 0 References WCAP- 1664 1-NP Rev. 0

APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS Appendix A WCAP- 1664 11-NP WCAP-1664 Rev. 00

-NP Rev. Appendix A

This page intentionally blank Appendix A WCAP-1664 1-NP Rev. 0

A-1 A. I Neutron Dosimetry Comparisons of measured dosimetry results to both the calculated and least squares adjusted values for all surveillance capsules withdrawn from service to date at Kewaunee are described herein. The sensor sets from these capsules have been analyzed in accordance with the requirements of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence"[Reference A-1] using the NRC approved methodology described in WCAP-16083-NP-A, Revision 0, "Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry."[Reference A-2] One of the main purposes for presenting this material is to demonstrate that the overall measurements agree with the calculated and least squares adjusted values to within +/- 20% as specified by Regulatory Guide 1.190, thus serving to validate the calculated neutron exposures previously reported in Section 6.2 of this report.

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

Azimuthal Withdrawal Irradiation Capsule ID Location Time Time [EFPY]

V 130 End of Cycle 1 1.3 R 130 End of Cycle 5 4.6 P 230 End of Cycle 13 11.1 S 330 End of Cycle 19 16.2 T 230 End of Cycle 26 24.6 The azimuthal locations included in the above tabulation represent the first octant equivalent azimuthal angle of the geometric center of the respective surveillance capsules.

The passive neutron sensors included in the evaluations of Surveillance Capsules V, R, P, S, and T are summarized as follows:

WCAP-16641-NP Rev. 0 Appendix A

A-2 Reaction Sensor Material Of Interest Capsule V Capsule R Capsule P Capsule S Capsule T 63 Copper Cu(ncQ)60 Co X X X X X 4

Iron 1 Fe(np) 4aMn X X X X X Nickel 5SNi(n,p)5 SCo X X X X X 238 37 Uranium-238(a) U(n,f)1 Cs X X X X X 237 37 X X Neptunium-237(a) Np(n,f) 1 Cs X X 59 Cobalt-Aluminum(b) Co(n,7) 60Co X X X X X (a)The Uranium and Neptunium sensors were cadmium covered.

(b)The cobalt-aluminum sensors included both bare and cadmium-covered wire.

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

Since the construction of the surveillance capsules used in the Kewaunee reactor design places individual sensors at several radial locations within the materials test specimen array [Reference A-3], gradient corrections based on the plant specific neutron transport calculations were applied to the measured sensor reaction rates to index all of the measured data to the geometric center of the surveillance capsule. The applicable gradient corrections used to index the measured results to the capsule center (Radius = 158.35 cm) are summarized in the following tabulation.

Gradient Correction factor Sensor 130 Location 23' Location 330 Location Sensor Type Radius [cm] (V and R) (P and T) (S)

Copper 158.11 0.957 0.954 0.954 Nickel 158.11 0.955 0.953 0.951 Cd Covered Cobalt-Aluminum 158.11 0.952 0.955 0.952 Uranium-238 158.35 1.000 1.000 1.000 Neptunium-237 158.35 1.000 1.000 1.000 Iron 159.11 1.158 1.140 1.141 Bare Cobalt-Aluminum 159.11 0.984 0.988 0.985 The use of passive monitors such as those listed above does not yield a direct measure of the energy dependent neutron flux at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:

" the measured specific activity of each monitor,

" the physical characteristics of each monitor, Appendix A WCAP- 1664 1-NP Rev. 0

A-3

• the operating history of the reactor,

" the energy response of each monitor, and

" the neutron energy spectrum at the monitor location.

Results from the radiometric counting of the neutron sensors from Capsules V, R, P, S, and T are summarized in Table A-4. In all cases, the radiometric counting followed established ASTM procedures.

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

The irradiation history of the reactor over the irradiation periods experienced by Capsules V, R, P, S, and T was based on the monthly power generation of the Kewaunee reactor from initial criticality through the end of the respective dosimetry evaluation periods. For the sensor sets utilized in the surveillance capsules, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. The irradiation history applicable to Capsules V, R, P, S, and T is given in Table A-2.

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

A NoFYY- Pci [1-e- AJ[e- "I]

P ref where:

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

A = Measured specific activity (dps/gm).

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

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

Y = Number of product atoms produced per reaction.

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

Pref = Maximum or reference power level of the reactor (MW).

Cj= Calculated ratio of ý(E > 1.0 MeV) during irradiation periodj to the time weighted average 4(E > 1.0 MeV) over the entire irradiation period.

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

WCAP- 1664 1-NP Rev. 0 Appendix A

A-4 tj = Length of irradiation period j (sec).

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

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

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

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

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

Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 238U measurements to account for the presence of 235U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.

Corrections were also made to the 238U and 237Np sensor reaction rates to account for gamma ray induced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the Kewaunee fission sensor reaction rates are summarized as follows:

Correction Capsule V Capsule R Capsule P Capsule S Capsule T 235 U Impurity/Pu Build-in 0.861 0.817 0.789 0.754 0.690 238 U(y,f) 0.950 0.950 0.955 0.953 0.955 238 Net U Correction 0.819 0.776 0.753 0.718 0.659 237 Np(yf) 0.983 0.983 0.984 0.983 0.984 These factors were applied in a multiplicative fashion to the decay corrected uranium and neptunium fission sensor reaction rates.

Results of the sensor reaction rate determinations for Capsules V, R, P, S, and T are also provided in Table A-4. Along with the measured specific activities, decay corrected saturated specific activities and computed reaction rates for each sensor indexed to the radial center of the capsule are listed. The fission sensor reaction rates are listed both with and without the applied corrections for 238U impurities, plutonium build-in, and gamma ray induced fission effects.

Appendix A WCAP-1664 1-NP Rev. 0

A-5 A. 1.2 Least Squares Evaluation of Sensor Sets Least squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations resulting in a best estimate adjusted neutron energy spectrum with associated uncertainties. Best estimates adjusted values for key exposure parameters such as 4(E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least squares methods, as applied to surveillance capsule dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrum within their respective uncertainties. For example, R +/- 6 Ri =I (yig +/- 6 )(qg +/- 69) g relates a set of measured reaction rates, Ri, to a single neutron spectrum, 4 g, through the multigroup dosimeter reaction cross-section, cig, each with an uncertainty 6. The primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the least squares evaluation of the Kewaunee surveillance capsule dosimetry, the FERRET code[A- 2] was employed to combine the results of the plant specific neutron transport calculations and sensor set reaction rate measurements to determine adjusted values of exposure parameters (4(E > 1.0 MeV) and dpa) along with associated uncertainties for the five in-vessel capsules analyzed to date.

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

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

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

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

For the Kewaunee application, the calculated neutron spectra were obtained from the results of plant specific neutron transport calculations described in Section 6.2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section A. 1.1. The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section library [Reference A-4]. The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard E 1018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E 706 (IIB)".

The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum were input to the least squares procedure in the form of variances and covariances. The WCAP- 1664 1-NP Rev. 0 Appendix A

A-6 assignment of the input uncertainties followed the guidance provided in ASTM Standard E 944, "Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance."

The following provides a summary of the uncertainties associated with the least squares evaluation of the Kewaunee surveillance capsule sensor sets.

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

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

Reaction Uncertainty 63 Cu(n,(c) 6 Co 5%

4 54 1 Fe(n,p) Mn 5%

58Ni(n,p) 58 Co 5%

238 37 U(n,f)1 Cs 10%

23 7 Np(n,f) 13 7Cs 10%

59 6 Co(n,7) °Co 5%

These uncertainties are given at the ly level.

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

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

Appendix A WCAP-1664 1-NP Rev. 0

A-7 Reaction Uncertainty 63 Cu(n,ct) 60Co 4.08-4.16%

14Fe(n,p) 5 4 Mn 3.05-3.11%

58 Ni(n,p)58Co 4.49-4.56%

238 37 U(n,'1 Cs 0.54-0.64%

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

59 Co(nY)60 Co 0.79-3.59%

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

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

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

2 M gg, = R n + R

• R,* P g g 9 where Rn specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg and Rg, specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:

Pgg, =I 1O] 6 gg, + 0 e-H where 2

H - (g -g')

2y2 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range y (0 specifies the strength of the latter term). The value of 6 is 1.0 when g = g', and is 0.0 otherwise.

WCAP-16641-NP Rev. 0 Appendix A

A-8 The set of parameters defining the input covariance matrix for the Kewaunee calculated spectra was as follows:

Flux Normalization Uncertainty (Rn) 15%

Flux Group Uncertainties (Rg, Rg,)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 29%

(E < 0.68 eV) 52%

Short Range Correlation (0)

(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Flux Group Correlation Range (7)

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 A. 1.3 Comparisons of Measurements and Calculations Results of the least squares evaluations of the dosimetry from the Kewaunee surveillance capsules withdrawn to date are provided in Tables A-5 and A-6. In Table A-5, measured, calculated, and adjusted values for sensor reaction rates are given for each capsule. Also provided in this tabulation are ratios of the measured reaction rates to both the calculated and least squares adjusted reaction rates. These ratios of M/C and M/A illustrate the consistency of the fit of the calculated neutron energy spectra to the measured reaction rates both before and after adjustment. In Table A-6, comparisons of the calculated and adjusted values of neutron flux (E > 1.0 MeV) and iron atom displacement rate are tabulated along with the A/C ratios observed for each of the capsules.

The data comparisons provided in Tables A-5 and A-6 show that the adjustments to the calculated spectra are relatively small and well within the assigned uncertainties for the calculated spectra, measured sensor Appendix A WCAP- 1664 1-NP Rev. 0

A-9 reaction rates, and dosimetry reaction cross-sections. Further, these results indicate that the use of the least squares evaluation results in a reduction in the uncertainties associated with the exposure of the surveillance capsules. From Section 6.4 of this report, it may be noted that the uncertainty associated with the unadjusted calculation of neutron fluence (E > 1.0 MeV) and iron atom displacements at the surveillance capsule locations is specified as 12% at the lcy level. From Table A-6, it is noted that the corresponding uncertainties associated with the least squares adjusted exposure parameters have been reduced to a range of from 6-7% for neutron flux (E > 1.0 MeV) and 7-8% for iron atom displacement rate. Again, the uncertainties from the least squares evaluation are at the I(Ylevel.

Further comparisons of the measurement results (from Tables A-5 and A-6) with calculations are given in Tables A-7 and A-8. These comparisons are given on two levels. In Table A-7, calculations of individual threshold sensor reaction rates are compared directly with the corresponding measurements. These threshold reaction rate comparisons provide a good evaluation of the accuracy of the fast neutron portion of the calculated energy spectra. In Table A-8, calculations of fast neutron exposure rates in terms of 4(E > 1.0 MeV) and dpa/s are compared with the adjusted values obtained from the least squares evaluation of the capsule dosimetry. These two levels of comparison yield consistent and similar results with all measurement-to-calculation comparisons falling well within the 20% limits specified as the acceptance criteria in Regulatory Guide 1.190.

In the case of the direct comparison of measured and calculated sensor reaction rates, the M/C comparisons for fast neutron reactions range from 0.84 to 1.17 for the 24 samples included in the data set.

The overall average M/C ratio for the entire set of Kewaunee data is 0.99 with an associated standard deviation of 7.6%.

In the comparisons of best estimate and calculated fast neutron exposure parameters, the corresponding A/C comparisons for the capsule data sets range from 0.93 to 1.03 for neutron flux (E > 1.0 MeV) and from 0.92 to 1.03 for iron atom displacement rate. The overall average A/C ratios for neutron flux (E > 1.0 MeV) and iron atom displacement rate are 0.99 with a standard deviation of 4.1% and 1.00 with a standard deviation of 4.3%, respectively.

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

Appendix A WCAP-1664 Rev. 0 1-NP Rev.

WCAP- 1664 1-NP 0 Appendix A

A-10 Table A- I Nuclear Parameters Used In The Evaluation Of Neutron Sensors Target Fission Atom 90% Response Monitor Reaction of Range Product Yield Fraction (MeV) Half-life (%)

Material Interest 63 Copper Cu (n,a) 0.6917 4.8-11.9 5.271 y 54 Iron Fe (n,p) 0.0585 2.1 -8.5 312.3 d Nickel 58 0.6808 1.6-8.4 70.82 d Ni (n,p) 238 Uranium-238 U (n,f) 1.0000 1.3-7.1 30.07 y 6.02 23 7 Neptunium-237 Np (n,f) 1.0000 0.4-4.6 30.07 y 6.17 59 Cobalt-Aluminum Co (n,y) 0.0015 non-threshold 5.271 y Note: The 90% response range is defined such that, in the neutron spectrum characteristic of the Kewaunee surveillance capsules, approximately 90% of the sensor response is due to neutrons in the energy range specified with approximately 5% of the total response due to neutrons with energies below the lower limit and 5% of the total response due to neutrons with energies above the upper limit.

Appendix A WCAP- 1664 1-NP Rev. 0

A-11 Table A-2 Monthly Thermal Generation During The First Twenty Six Fuel Cycles Of The Kewaunee Reactor Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 1974 4 121412 1977 4 999717 1980 4 1175492 1974 5 600420 1977 5 1188560 1980 5 349621 1974 6 864909 1977 6 1156447 1980 6 137929 1974 7 776871 1977 7 1201271 1980 7 1186866 1974 8 1172100 1977 8 1049588 1980 8 1125341 1974 9 655831 1977 9 1175016 1980 9 946334 1974 10 317628 1977 10 1207031 1980 10 1181104 1974 11 752190 1977 11 1175653 1980 11 1183962 1974 12 907703 1977 12 1194097 1980 12 1188454 1975 1 781842 1978 1 1218479 1981 1 1223429 1975 2 928272 1978 2 1098398 1981 2 1077953 1975 3 1096956 1978 3 1203881 1981 3 1136912 1975 4 827400 1978 4 811877 1981 4 777154 1975 5 874521 1978 5 72462 1981 5 0 1975 6 777805 1978 6 1016869 1981 6 777459 1975 7 842326 1978 7 1129811 1981 7 1204230 1975 8 1175609 1978 8 1182936 1981 8 1222464 1975 9 604163 1978 9 1137041 1981 9 1161892 1975 10 916973 1978 10 1208085 1981 10 1170754 1975 11 823299 1978 11 1129396 1981 11 1148183 1975 12 1171599 1978 12 1200591 1981 12 1219280 1976 1 1106166 1979 1 1212033 1982 1 1203474 1976 2 501575 1979 2 1047590 1982 2 1075185 1976 3 0 1979 3 1135718 1982 3 1224649 1976 4 377039 1979 4 1151567 1982 4 326008 1976 5 685126 1979 5 979438 1982 5 208610 1976 6 1171872 1979 6 0 1982 6 1163350 1976 7 1180703 1979 7 0 1982 7 1218118 1976 8 1190960 1979 8 740156 1982 8 1222402 1976 9 1060865 1979 9 1126633 1982 9 1180708 1976 10 1198895 1979 10 1184602 1982 10 1214805 1976 11 1145875 1979 11 1167457 1982 11 1178287 1976 12 1187141 1979 12 1202648 1982 12 1077471 1977 1 606787 1980 1 713133 1983 1 1218455 1977 2 0 1980 2 1124815 1983 2 1102759 1977 3 179775 1980 3 1223050 1983 3 649484 WCAP- 16641-NP Rev. 0 Appendix A

A-12 Table A-2 cont'd Monthly Thermal Generation During The First Twenty Six Fuel Cycles Of The Kewaunee Reactor Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 1983 4 0 1986 4 267903 1989 4 537116 1983 5 526179 1986 5 1191848 1989 5 1221793 1983 6 1174626 1986 6 1176926 1989 6 902394 1983 7 1159011 1986 7 1218187 1989 7 1221968 1983 8 1223940 1986 8 1179835 1989 8 1221653 1983 9 1184200 1986 9 1180207 1989 9 1182431 1983 10 1224069 1986 10 1202091 1989 10 1222999 1983 11 1183857 1986 11 1179133 1989 11 1181532 1983 12 1224505 1986 12 1215853 1989 12 1182173 1984 1 1225897 1987 1 1215680 1990 1 1217978 1984 2 1139329 1987 2 924655 1990 2 1104275 1984 3 532963 1987 3 0 1990 3 69012 1984 4 0 1987 4 959357 1990 4 347514 1984 5 837414 1987 5 1185857 1990 5 1211562 1984 6 1144593 1987 6 1146829 1990 6 1172482 1984 7 1199335 1987 7 1183725 1990 7 1224388 1984 8 1222297 1987 8 1219047 1990 8 1224227 1984 9 1183431 1987 9 1168054 1990 9 1182733 1984 10 1205018 1987 10 1219759 1990 10 1224609 1984 11 1183532 1987 11 1176994 1990 11 1182652 1984 12 1222229 1987 12 1215302 1990 12 1208996 1985 1 1224279 1988 1 1217528 1991 1 1220109 1985 2 271048 1988 2 1134763 1991 2 1100105 1985 3 0 1988 3 57885 1991 3 285385 1985 4 682721 1988 4 478699 1991 4 0 1985 5 1212434 1988 5 1195654 1991 5 694766 1985 6 1179472 1988 6 1180894 1991 6 1180240 1985 7 1218713 1988 7 1196631 1991 7 1210778 1985 8 1148512 1988 8 1067560 1991 8 1211741 1985 9 1178061 1988 9 1088933 1991 9 1171047 1985 10 1222974 1988 10 1221808 1991 10 1147457 1985 11 1121094 1988 11 1169480 1991 11 1182454 1985 12 1181301 1988 12 1219313 1991 12 1211606 1986 1 1217881 1989 1 1204791 1992 1 1222525 1986 2 1067018 1989 2 755230 1992 2 1143608 1986 3 0 1989 3 0 1992 3 225433 Appendix A Appenix AWCAP-1664 1-NP Rev. 0

A-13 Table A-2 cont'd Monthly Thermal Generation During The First Twenty Six Fuel Cycles Of The Kewaunee Reactor Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 1992 4 351235 1995 4 65 1998 4 1143027 1992 5 1226398 1995 5 388395 1998 5 1182185 1992 6 1185833 1995 6 1131317 1998 6 1125155 1992 7 1227301 1995 7 1185361 1998 7 1181037 1992 8 1228782 1995 8 1187206 1998 8 1180476 1992 9 1063233 1995 9 1107004 1998 9 1142017 1992 10 1229699 1995 10 1199601 1998 10 595143 1992 11 1113009 1995 11 1169068 1998 11 50510 1992 12 1227368 1995 12 1208346 1998 12 1075135 1993 1 1136925 1996 I 1208794 1999 1 1166738 1993 2 1085208 1996 2 1130772 1999 2 1070798 1993 3 156579 1996 3 1171857 1999 3 1194219 1993 4 381383 1996 4 1130838 1999 4 1147356 1993 5 1227217 1996 5 1205079 1999 5 1193822 1993 6 754286 1996 6 1166067 1999 6 1133072 1993 7 1226890 1996 7 1204827 1999 7 1191624 1993 8 1226800 1996 8 1174752 1999 8 1192858 1993 9 1184271 1996 9 598870 1999 9 1151703 1993 10 1227961 1996 10 0 1999 10 1193859 1993 11 1186431 1996 11 0 1999 11 1153786 1993 12 1223457 1996 12 0 1999 12 1191329 1994 1 929429 1997 1 0 2000 1 1190621 1994 2 1105268 1997 2 0 2000 2 1111904 1994 3 1200731 1997 3 0 2000 3 1190979 1994 4 22616 1997 4 0 2000 4 796780 1994 5 751549 1997 5 4975 2000 5 0 1994 6 1187510 1997 6 26 2000 6 832708 1994 7 1225867 1997 7 1164539 2000 7 1156171 1994 8 1225305 1997 8 1179559 2000 8 1163425 1994 9 1187543 1997 9 1149530 2000 9 1137281 1994 10 1228790 1997 10 1189032 2000 10 1177799 1994 11 1185472 1997 11 1148938 2000 11 1091925 1994 12 1227116 1997 12 1178977 2000 12 1176511 1995 1 1227125 1998 1 1190492 2001 1 1180239 1995 2 1108344 1998 2 720791 2001 2 1067797 1995 3 1084328 1998 3 1183303 2001 3 1179343 WCAP-16641-NP Rev. 0 Appendix A

A-14 Table A-2 cont'd Monthly Thermal Generation During The First Twenty Six Fuel Cycles Of The Kewaunee Reactor Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 2001 4 1142622 2002 7 1226420 2003 10 1242270 2001 5 1161973 2002 8 1223596 2003 11 1203864 2001 6 1073786 2002 9 1187230 2003 12 1222505 2001 7 1176575 2002 10 1227937 2004 1 603941 2001 8 1176245 2002 11 1187472 2004 2 1120517 2001 9 865416 2002 12 1216854 2004 3 1199719 2001 10 0 2003 1 1226903 2004 4 1273278 2001 11 0 2003 2 1059368 2004 5 1317279 2001 12 950489 2003 3 1225395 2004 6 1271435 2002 1 1225438 2003 4 127791 2004 7 1313665 2002 2 1107933 2003 5 720780 2004 8 1306672 2002 3 1213191 2003 6 1186466 2004 9 1274881 2002 4 1185520 2003 7 1225660 2004 10 321012 2002 5 771167 2003 8 1244020 2002 6 1186480 2003 9 1203912 WCAP-16641-NP Rev. 0 Appendix AA Appendix WCAP- 1664 1-NP Rev. 0

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

Cycle [EFPS] Capsule V Capsule R Capsule P Capsule S Capsule T I 4.06E+07 1.44E+ ll 1.44E+ 11 8.10E+10 7.64E+ 10 8.10E+10 2 2.14E+07 1.43E+1l 8.64E+10 8.43E+10 8.64E+10 3 3.24E+07 9.14E+10 7.63E+10 7.90E+10 7.63E+10 4 2.97E+07 1.1OE+11 7.19E+10 6.88E+10 7.19E+10 5 2.18E+07 1.11E+ll 7.25E+10 7.02E+10 7.25E+10 6 2.44E+07 7.70E+10 7.81E+10 7.70E+10 7 2.56E+07 6.91E+10 6.93E+10 6.91E+10 8 2.49E+07 7.38E+10 7.32E+10 7.38E+10 9 2.57E+07 7.33E+10 7.01E+10 7.33E+10 10 2.33E+07 7.11E+10 6.88E+10 7.11E+10 11 2.71E+07 6.93E+10 6.58E+10 6.93E+10 12 2.61E+07 6.87E+10 6.24E+10 6.87E+10 13 2.81E+07 7.48E+10 7.22E+10 7.48E+10 14 2.57E+07 6.99E+10 7.26E+10 15 2.68E+07 6.93E+10 7.30E+10 16 2.75E+07 7.06E+10 7.02E+10 17 2.53E+07 6.91E+10 7.05E+10 18 2.67E+07 7.38E+10 7.11E+10 19 2.81E+07 6.97E+10 7.15E+10 20 2.57E+07 7.33E+10 21 3.88E+07 7.09E+10 22 3.96E+07 7.21E+10 23 4.08E+07 7.33E+10 24 3.84E+07 6.87E+10 25 3.96E+07 7.12E+ 10 26 4.27E+07 6.36E+10 Average 1.44E+11 1.20E+l1 7.44E+10 7.17E+10 7.22E+10 Rev. 00 Appendix A WCAP-16641-NP WCAP- 1664 1-N P Rev. Appendix A

A-16 Table A-3 (continued)

Calculated Cj Factors at the Surveillance Capsule Center Core Midplane Elevation Fuel Cycle Length Ci Cycle [EFPS] Capsule V Capsule R Capsule P Capsule S Capsule T I 4.06E+07 1.000 1.200 1.089 1.066 1.122 2 2.14E+07 1.185 1.162 1.176 1.197 3 3.24E+07 0.760 1.025 1.101 1.056 4 2.97E+07 0.911 0.967 0.959 0.997 5 2.18E+07 0.924 0.975 0.979 1.005 6 2.44E+07 1.035 1.089 1.067 7 2.56E+07 0.929 0.967 0.958 8 2.49E+07 0.992 1.020 1.022 9 2.57E+07 0.985 0.978 1.015 10 2.33E+07 0.956 0.959 0.985 11 2.71E+07 0.931 0.917 0.959 12 2.61E+07 0.924 0.870 0.952 13 2.81E+07 1.006 1.007 1.036 14 2.57E+07 0.975 1.006 15 2.68E+07 0.966 1.011 16 2.75E+07 0.984 0.972 17 2.53E+07 0.964 0.976 18 2.67E+07 1.029 0.985 19 2.81E+07 0.972 0.991 20 2.57E+07 1.015 21 3.88E+07 0.982 22 3.96E+07 0.999 23 4.08E+07 1.015 24 3.84E+07 0.951 25 3.96E+07 0.987 26 4.27E+07 0.835 Appendix A WCAP-16641-NP Rev. 0

A-17 Table A-4 Measured Sensor Activities And Reaction Rates Surveillance Capsule V Gradient Corrected Measured Saturated Reaction Activity Activity Rate Reaction Location (dps/g) (dps/g) (rps/atom) 63 Cu (n,a) 60Co Top-Middle 6.89E+04 4.85E+05 7.08E-17 Bottom-Middle 7.83E+04 5.51 E+05 8.05E-17 Average 7.56E-17 54 Fe (n,p) 54 Mn Top 1.93E+06 5.03E+06 9.24E- 15 Top-Middle 1.76E+06 4.59E+06 8.43E-15 Middle 1.84E+06 4.80E+06 8.81E-15 Bottom-Middle 1.82E+06 4.75E+06 8.71E-15 Bottom 1.98E+06 5.16E+06 9.48E- 15 Average 8.93E-15 58Ni (n,p) 5"Co Middle 1.17E+07 8.26E+07 1.13E-14 Average 1.13E-14 23 SU (n,f) 37 1 Cs (Cd) Middle 2.47E+05 8.60E+06 5.65E-14 Including 235U, 239 Pu, and y fission corrections: 4.62E-14 237 Np (n,f) 13 7Cs (Cd) Middle 2.18E+06 7.59E+07 4.84E-13 Including y fission corrections: 4.76E-13 59Co (n,y) 6

°Co Top 2.27E+07 1.60E+08 1.03E- 11 Average 1.03E-11 59Co (n,7) 6 0Co (Cd) Top 1.01E+07 7.11E+07 4.42E-12 Bottom 1.02E+07 7.18E+07 4.46E- 12 Average 4.44E-12 Notes: 1) Measured specific activities are indexed to a counting date of August 11, 1976.

23

2) The average 8U (n,f) reaction rate of 4.62E-14 includes a correction factor of 0.861 to account for plutonium build-in and an additional factor of 0.950 to account for photo-fission effects in the sensor.

3) The average 237Np (n,f) reaction rate of 4.76E-13 includes a correction factor of 0.983 to account for photo-fission effects in the sensor.

4) Reaction rates referenced to the Cycle I Rated Reactor Power of 1650 Mwt.

Appendix A WCAP-16641-NP WCA-P- 1664 1-NP Rev. Rev. 00 Appendix A

A-18 Table A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule R Gradient Corrected Measured Saturated Reaction Activity Activity Rate Reaction Location (dps/g) (dps/g) (rps/atom) 63 Cu (n,ct) 60 Co Top-Middle 1.65E+05 4.20E+05 6.13E-17 Average 6.13E-17 14Fe (n,p) 54 Mn Top 2.25E+06 4.33E+06 7.95E-15 Top-Middle 2.09E+06 4.02E+06 7.39E-15 Middle 2.11E+06 4.06E+06 7.46E-15 Bottom 2.32E+06 4.47E+06 8.20E-15 Average 7.75E-15 58Ni (n,p) "SCo Middle 1.31E+07 1.08E-14 7.87E+07 Average 1.08E-14 238U (n,f) 13 7Cs (Cd) Middle 7.99E+05 8.11E+06 5.33E-14 239 Including 2 35U, Pu, and y fission corrections: 4.13E-14 237 Np (n,f) 13 7Cs (Cd) Middle See Note 3 Including y fission corrections:

59Co (nY) 60 Co Top 5.31E+07 1.35E+08 8.67E-12 Bottom 5.57E+07 1.42E+08 9.1OE-12 Average 8.88E-12 59 60 Co (nY) Co (Cd) Top 2.63E+07 6.69E+07 4.16E-12 Bottom 2.36E+07 6.OOE+07 3.73E-12 Average 3.94E-12 Notes: 1) Measured specific activities are indexed to a counting date of October 21, 1980.

23

2) The average 8U (n,f) reaction rate of 4.13E-14 includes a correction factor of 0.817 to account for plutonium build-in and an additional factor of 0.950 to account for photo-fission effects in the sensor.

3) The neptunium sensor contained in Capsule R produced erroneous results and was rejected from further consideration.

4) Reaction Rates referenced to the Cycles 1-5 Average Rated Reactor Power of 1650 MWt.

Appendix A WCAP-16641-NP Rev. 0

A-19 Table A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule P Gradient Corrected Measured Saturated Reaction Activity Activity Rate Reaction Location (dps/g) (dps/a) (rps/atom) 63 60 Cu (n, o) Co Top-Middle 2.12E+05 3.19E+05 4.65E-17 Bottom-Middle 2.44E+05 3.67E+05 5.35E-17 Average 5.OOE-17 54 54 Fe (n,p) Mn Top 2.74E+06 3.68E+06 6.65E-15 Top-Middle 1.95E+06 2.62E+06 4.73E-15 Middle 2.11E+06 2.83E+06 5.12E-15 Bottom-Middle 2.19E+06 2.94E+06 5.3 1E-15 Bottom 2.33E+06 3.13E+06 5.65E-15 Average 5.49E-15 58Ni (n,p) 5"Co Middle 2.44E+07 5.03E+07 6.86E-15 Average 6.86E-15 23 8U (n,f) 137Cs (Cd) Middle 1.13E+06 5.16E+06 3.39E-14 235 239 Including U, Pu, and y fission corrections: 2.55E-14 237 Np (n,f) 13 7 Cs (Cd) Middle 8.28E+06 3.78E+07 2.41E-13 Including y fission corrections: 2.37E-13

' 9Co (ny) 60Co Top 4.20E+07 6.32E+07 4.08E-12 Bottom 5.01E+07 7.54E+07 4.86E-12 Average 4.47E-12 59 60 Co (nY) Co (Cd) Top 1.79E+07 2.70E+07 1.68E-12 Bottom 2.04E+07 3.07E+07 1.91E-12 Average 1.80E-12 Notes: 1) Measured specific activities are indexed to a counting date of May 12, 1988.

23

2) The average 8U (n,f) reaction rate of 2.55E-14 includes a correction factor of 0.789 to account for plutonium build-in and an additional factor of 0.955 to account for photo-fission effects in the sensor.

237

3) The average Np (n,f) reaction rate of 2.37E-13 includes a correction factor of 0.984 to account for photo-fission effects in the sensor.

4) Reaction Rates referenced to the Cycles 1-13 Average Rated Reactor Power of 1650 MWt.

WCAP- 1664 1 -NP Rev. 0 Appendix A

A-20 Table A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule S Gradient Corrected Measured Saturated Reaction Activity Activity Rate Reaction Location (dps/gz) (dps/gz) (Ms/atom) 63Cu (n,co) 6 0 Co Top-Middle 2.15E+05 3.03E+05 4.40E- 17 Bottom-Middle 2.38E+05 3.35E+05 4.87E-17 Average 4.64E-17 14Fe (n,p) 54Mn Top 1.53E+06 2.74E+06 4.96E-15 Top-Middle 1.39E+06 2.49E+06 4.50E-15 Middle 1.52E+06 2.72E+06 4.92E-15 Bottom-Middle 1.49E+06 2.67E+06 4.83E-15 Average 4.80E-15 5t Ni (np) 58Co Middle 6.92E+06 5.14E+07 6.99E-15 Average 6.99E-15 2 38 U (n,f) 13 7Cs (Cd) Middle 1.3 1E+06 4.42E+06 2.90E-14 U, 239 Pu, 23 5 Including and y fission corrections: 2.09E-14 23 7 Np (n,f) 37 1 Cs (Cd) Middle 7.90E+06 2.67E+07 1.70E-13 Including y fission corrections: 1.67E-13 59 Co (nY) 60 Co Top 4.38E+07 6.16E+07 3.96E-12 Bottom 4.48E+07 6.30E+07 4.05E-12 Average 4.01E-12 59 60 Co (n,y) Co (Cd) Top 2.14E+07 3.01E+07 1.87E-12 Bottom 2. 1OE+07 2.95E+07 1.84E-12 Average 1.85E-12 Notes: 1) Measured specific activities are indexed to a counting date of October 13, 1994.

2) The average 238U (n,f) reaction rate of 2.09E-14 includes a correction factor of 0.754 to account for plutonium build-in and an additional factor of 0.953 to account for photo-fission effects in the sensor.

237

3) The average Np (n,f) reaction rate of 1.67E-13 includes a correction factor of 0.983 to account for photo-fission effects in the sensor.

4) Reaction Rates referenced to the Cycles 1-19 Average Rated Reactor Power of 1650 MWt.

Appendix A WCAP- 1664 1-NP Rev. 0

A-21 Table A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule T Gradient Corrected Measured Saturated Reaction Activity Activity Rate Reaction Location (d2s/g) (dps/a) (rps/atom) 63Cu (nct) 60 Co Top-Middle 2.08E+05 3.07E+05 4.47E-17 Bottom-Middle 2.35E+05 3.47E+05 5.05E-17 Average 4.76E-17 54 Fe (n,p) 54 Mn Top 9.21E+05 2.99E+06 5.41E-15 Top-Middle 8.05E+05 2.62E+06 4.73E-15 Middle 8.89E+05 2.89E+06 5.23E-15 Bottom-Middle 8.58E+05 2.79E+06 5.04E- 15 Bottom 9.5 IE+05 3.09E+06 5.59E- 15 Average 5.20E-15 5 58 8Ni (n,p) Co Middle 5.85E+05 5.30E+07 7.22E-15 Average 7.22E-15 23 8U (n,f) 137Cs (Cd) Middle 2.16E+06 5.42E+06 3.56E-14 239 Including 2 3 5 U, Pu, and y fission corrections: 2.35E-14 23 7 Np (n,f) 137 Cs (Cd) Middle 1.34E+07 3.36E+07 2.15E-13 Including y fission corrections: 2.11E-13 59Co (nY) 60 Co Top 4.45E+07 6.57E+07 4.22E-12 Average 4.22E-12 59Co (nY) 60Co (Cd) Top 2.35E+07 3.47E+07 2.16E-12 Bottom 2.16E+07 3.19E+07 1.98E-12 Average 2.07E-12 Notes: 1) Measured specific activities are indexed to a counting date of December 28, 2005.

38

2) The average 2 U (n,f) reaction rate of 2.35E-14 includes a correction factor of 0.690 to account for plutonium build-in and an additional factor of 0.955 to account for photo-fission effects in the sensor.

3) The average 237Np (n,f) reaction rate of 2.11 E- 13 includes a correction factor of 0.984 to account for photo-fission effects in the sensor.

4) Reaction Rates referenced to the Cycles 1-26 Average Rated Reactor Power of 1653 MWt. This lifetime average core power accounts for an uprate from 1650 MWt to 1772 MWt during Cycle 26.

WCAP- 1664 1-NP Rev. 0 Appendix A

A-22 Table A-5 Comparison of Measured, Calculated, and Adjusted Reaction Rates At The Surveillance Capsule Center Capsule V Reaction Rate [rps/atom]

Reaction Measured Calculated Adjusted M/C M/A 63 6 Cu(n,CC) 0CO 7.58E-17 7.37E-17 7.42E-17 1.03 1.02 54Fe(n,p) 54 Mn 8.93E-15 8.95E-15 8.78E-15 1.00 1.02 58Ni(n,p) 5 SCo 1.13E- 14 1.25E-14 1.19E- 14 0.90 0.95 238U(nf)137Cs (Cd) 4.62E-14 4.79E-14 4.68E-14 0.96 0.99 2 7 3 Np(n,f)137Cs (Cd) 4.76E-13 4.17E-13 4.44E-13 1.14 1.07 59 6 Co(ny) 0Co 1.03E-11 1.05E-11 1.03E-11 0.98 1.00 59Co(n,y) 60Co (Cd) 4.44E-12 4.17E-12 4.43E-12 1.06 1.00 Note: See Section A. 1.2 for details describing the Adjusted (A) reaction rates.

Capsule R Reaction Rate [rps/atom]

Reaction Measured Calculated Adjusted M/C M/A 63Cu(nct) 6 0CO 6.13E-17 6.40E-17 6.24E-17 0.96 0.98 54Fe(n,p) 54 Mn 7.75E-15 7.60E-15 7.68E-15 1.02 1.01 5S8Ni(n,p)5SCo 1.08E-14 1.06E-14 1.07E-14 1.02 1.01 238 U(n,0137 Cs (Cd) 4.13E- 14 4.02E-14 4.12E-14 1.03 1.00 59Co(ny)60Co 8.88E-12 8.67E-12 8.90E-12 1.02 1.00 59Co(ny)60Co (Cd) 3.94E-12 3.43E-12 3.92E- 12 1.15 1.01 Note: See Section A. 1.2 for details describing the Adjusted (A) reaction rates.

Capsule P Reaction Rate [rps/atom]

Reaction Measured Calculated Adjusted M/C M/A 63 6 Cu(n,at) 0CO 5.OOE-17 5.04E-17 4.98E-17 0.99 1.00 54Fe(n,p) 54 Mn 5.49E-15 5.36E-15 5.35E-15 1.02 1.03 58Ni(n,p) 5SCo 6.86E- 15 7.36E-15 7.16E- 15 0.93 0.96 238 U(n,f)137Cs (Cd) 2.55E-14 2.60E-14 2.60E-14 0.98 0.98 237 Np(n,f)137Cs (Cd) 2.37E-13 2.03E-13 2.21E-13 1.17 1.07 59Co(n,y) 6 0Co 4.47E-12 4.69E-12 4.48E-12 0.95 1.00 59Co(n,y)6°Co (Cd) 1.80E-12 1.77E-12 1.80E-12 1.02 1.00 Note: See Section A. 1.2 for details describing the Adjusted (A) reaction rates.

WCAP-1664 1-NP Rev. 0 Appendix AA WCAP- 1664 1-NP Rev. 0

A-23 Table A-5 (continued)

Comparison of Measured, Calculated, and Adjusted Reaction Rates At The Surveillance Capsule Center Capsule S Reaction Reacti o Rate

[rns/atoml

-Rate

/ ato Reaction Measured Calculated Adjusted M/C M/A 63Cu(n,a) 6 0CO 4.64E-17 4.47E-17 4.60E-17 1.04 1.01 54Fe(n,p)5 4 Mn 4.80E-15 4.93E-15 4.84E-15 0.97 0.99 58Ni(n,p)5SCo 6.99E-15 6.80E-15 6.75E-15 1.03 1.04 238U(n,f)137Cs (Cd) 2.09E-14 2.47E-14 2.33E-14 0.85 0.90 237Np(n,f)13 7Cs (Cd) 1.67E-13 1.99E-13 1.75E-13 0.84 0.95 59Co(n,y)60Co 4.01E-12 4.61E-12 4.03E-12 0.87 1.00 59Co(n,'y)60Co (Cd) 1.85E-12 1.79E-12 1.84E-12 1.03 1.01 Note: See Section A. 1.2 for details describing the Adjusted (A) reaction rates.

Capsule T Reaction Rate [rps/atom]

Reaction Measured Calculated Adjusted M/C M/A 63 Cu(n,cL) 6 0CO 4.76E- 17 4.92E- 17 4.80E- 17 0.97 0.99 54Fe(n,p) 54 Mn 5.20E- 15 5.25E-15 5.19E- 15 0.99 1.00 58Ni(n,p) 58 Co 7.22E- 15 7.20E- 15 7.15E- 15 1.00 1.01 2 38 U(nf)137Cs (Cd) 2.35E-14 2.52E-14 2.51 E- 14 0.93 0.94 237Np(n,f)137 Cs (Cd) 2.11E-13 1.97E-13 2.04E- 13 1.07 1.03 59Co(n,y) 60Co 4.22E-12 4.52E-12 4.25E-12 0.93 0.99 59Co(n,y)60Co (Cd) 2.07E-12 2.05E-12 2.05E-12 1.21 1.01 Note: See Section A. 1.2 for details describing the Adjusted (A) reaction rates.

Appendix A WCAP- 1664 11-NP WCAP-1664 Rev. 00

-NP Rev. Appendix A

A-24 Table A-6 Comparison of Calculated and Adjusted Exposure Rates At The Surveillance Capsule Center

ý(E > 1.0 Me [n/cm 2-s]

Uncertainty Capsule ID Calculated Adjusted (1o) A/C V 1.45E+ 11 1.42E+ 11 6% 0.98 R 1.21E+lI1 1.25E+11 7% 1.03 p 7.46E+10 7.55E+10 6% 1.01 S 7.20E+ 10 6.67E+ 10 6% 0.93 T 7.24E+ 10 7.24E+ 10 6% 1.00 Note: Calculated results are based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period and are the average neutron exposure rates over the irradiation period for each capsule. See Section A. 1.2 for details describing the adjusted exposure rates.

Iron Atom Displacement Rate [dpa/s]

Uncertainty Capsule ID Calculated Adjusted (ic7) BE/C V 2.58E-10 2.58E-10 7% 1.00 R 2.14E-10 2.21E-10 8% 1.03 p 1.27E-l0 1.29E-10 7% 1.02 S 1.23E-10 1.14E-10 7% 0.92 T 1.23E-10 1.24E-10 7% 1.00 Note: Calculated results are based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period and are the average neutron exposure rate over the irradiation period for each capsule. See Section A. 1.2 for details describing the adjusted exposure rates.

Appendix A WCAP-1664 1-NP Rev. 0

A-25 Table A-7 Comparison of Measured/Calculated (MYC) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions M/C Ratio Reaction Capsule V Capsule R Capsule P Capsule S Capsule T 63 4

Cu(n,a) 64°Co 1.03 0.96 0.99 1.04 0.97 1 Fe(n,p)1 Mn 1.00 1.02 1.02 0.97 0.99 5

238 SNi(n,p)5 8Co 0.90 1.02 0.93 1.03 1.00 U(n,p)137Cs (Cd) 0.96 1.03 0.98 0.85 0.93 237 Np(n,D137 Cs (Cd) 1.14 1.17 0.84 1.07 Average 1.01 1.01 1.02 0.94 0.99

% Standard Deviation 8.7 3.2 8.8 10.2 5.2 Note: The overall average M/C ratio for the set of 24 sensor measurements is 0.99 with an associated standard deviation of 7.6%.

Table A-8 Comparison of Adjusted/Calculated (A/C) Exposure Rate Ratios BE/C Ratio Capsule ID 4i(E > 1.0 MeV) dpa/s V 0.98 1.00 R 1.03 1.03 P 1.01 1.02 S 0.93 0.92 T 1.00 1.00 Average 0.99 1.00

% Standard Deviation 4.1 4.3 Appendix A WCAP-16641-NP Rev. 0 WCAP-1664 1-NP Rev. 0 Appendix A

A-26 Appendix A References A-1. Regulatory Guide RG- 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

A-2. WCAP-16083-NP-A, Revision 0, "Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry," May 2006.

A.3. WCAP-8107, Revision 0, "Wisconsin Public Service Corp. Kewaunee Nuclear Power Plant Reactor Vessel Radiation Surveillance Program," April 1973.

A-4. RSIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section Compendium", July 1994 Appendix A WCAP-16641-NP Rev. 0

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

• Specimen prefix "P" denotes intermediate shell forging 122X208VA I, Tangential Orientation

• Specimen prefix "S"denotes lower shell forging 123X167VA1, Tangential Orientation

" Specimen prefix "R" denotes A533 Grade B Class 1 Standard Reference Material

• Specimen prefix "H" denotes reconstituted Weld Material Appendix B WCAP-16641-NP WCAP- Rev. 00 16641 -NP Rev. Appendix B

This page intentionally blank WCAP-16641-NP Rev. 0 Appendix B Appendix B WCAP- 1664 1-NP Rev. 0

B-1 5m000 4000.00 3000.00 2000.00 1000.00 0.054 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tir*-I (irs)

P47, F 5000.00 4000.00 3000.00 2000.00 1000.00 n 0ni 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-i (ms)

P46, 50°F WCAP-1664 1-NP Rev. 0 Appendix B

B-2 5000.00 4000.00 3000.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1(ms)

P45, 75°F 5000.00 4000.00 3000.00 2000.00 1000.00 n~(i

/

0,00 1.00 2.00 3.00 4.00 5.00 6.00 Time-I (ms)

P43, 100 0 F WCAP-1664 1-NP Rev. 0 Appendix B Appendix B WCAP- 1664 1 -NP Rev. 0

B-3 5000.

4000.

3000.

2000.

1000.

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

P40, 125°F

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

P37, 150°F WCAP- 1664 1-NP Rev. 0 Appendix B

B-4 3000.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-i (ms)

P44,1750 F 5000.00 4000.00 3000.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (m)

P41, 200°F WCAP-1664 1-NP Rev. 0 B

Appendix B WCAP- 1664 1-NP Rev. 0

B-5 5000.00 4000.00 300000 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

P48, 250°F 5000.00 4000.00

- 3000.00 2000.00 0.00 ',

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

P38, 325°F Rev. 00 Appendix B WCAP-16641-NP WCAP- 1664 1-NP Rev. Appendix B

B-6 B-6 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-I (ms)

P39, 350 0 F 5000.

4000.

7 3000.

0, 2000.

1000.

0.00 1f00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

P42, 400°F Appendix B WCAP- 1664 1-NP Rev. 0

B-7

.5000.00 4000,00!

400000 1000,00 0,00 3000.00 0.00 1.00 2.00 3.00 4.00 5.00 5.00 Time-i (ins)

S45, F 5=0000 A

40M.00

'7 3M0000 2 -0000 1000.00" 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

S47, O°F Appendix B WCAP-1664 WCAP- Rev. 00 1-NP Rev.

1664 1-NP Appendix B

B-8 500.00 40.00 3=0.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time.1 (ms)

S40, 25-F 5000.00 4000.00 3000.00 2000.00 1000.001 A

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

S38, 50°F Appendix B WCAP-16641-NP Rev. 0

B-9 5000.00 4000.00 3000.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tmie-1(ms)

S37, 75°F 5000.00 4000.00

- 3000.00 2000 .00 1000.00 O.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1(ms)

S48, 100°F Rev. 00 Appendix B WCAP-16641-NP WCAP- 1664 1-NP Rev. Appendix B

B-10 5000.00 4000.00

,300.00 2000.00 1000.00 0.001 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-i (ms)

S43, 1250 F 5000.00 4000.00 3000.00 2000.00 1000.00 -

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

S44, 150 0 F Appendix B WCAP- 1664 1-NP Rev. 0

B-11 4M.00 4000.00

\

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

S46, 175-F 5000.00 4000.00 3000.00 2000.00 1000.00 3.0405060 0.00 . , * * *

• t ' *.

0.00 1.0 2.00 3.00 4.00 5.00 6.00 Tme-1 (mos)

S41, 250°F Appendix B WCAP-16641-NP WCAP- Rev. 00 1664 1-NP Rev. Appendix B

B-12 5000.00 4000.00 3=.00 2000.00 1000.00 0.00 0.

Time-1(ms)

S39, 300 0 F 5000.00 4000.00 3000.00 2000.00 1000.00*

0.00 1.00 2.00 3.00 4.O0 5.00 6.00 lime-1 (ms)

S42, 350 0 F WCAP-1664 1-NP Rev. 0 Appendix B WCAP-16641-NP Rev. 0

B-13 50.00 4000.00 3O.00 2000.00 1000.00 0.00 1.00 2.00 3.00 4.00 500 6 in Time-1 (ms)

R31, 100°F 5000.00 4000.00 3000.00 2000.00 1000.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

R30, 200°F Appendix B WCAP- 1664 11-NP WCAP-1664 Rev. 00

-NP Rev. Appendix B

B-14 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

R28, 225-F 4000.00 3000.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-I (ns)

R26, 250°F Appendix B WCAP-1664 1-NP Rev. 0

B-15 500000

.00 200000 1000"00 0.00 0.00 1.00 2.00 3-00 4.00 5.00 6.00 Time-i (ms)

R275, 275-F 5000.00 4000.00 30O0.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (i )

R29, 300-F Appendix B WCAP-1664 WCAP-16641-NP Rev. 0 1-NP Rev. 0 Appendix B

B-16 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-I (ms)

R32, 4000 F 5000.00 4000.00 3000.00 2000.00 1000.00 3.00 Time-i (ms)

R25, 450°F Appendix B WCAP-1664 1-NP Rev. 0

B-17 sm000 400000 30000*

2000.00 1000.00 L

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

H26, 190°F O.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-i (ms)

H27,215°F WCAP- 16641-NP Rev. 0 Appendix B

B-18 5000.00.

4000.00 3000.00" 2000.00 1000.00-Time-1 (ms)

H32, 4000 F Appendix B WCAP-16641-NP Rev. 0

APPENDIX C MASTER CURVE RESULTS FROM CAPSULE T AND PREVIOUS FRACTURE TOUGHNESS TESTS WCAP- 1664 1-NP Rev. 0 Appendix C

This page intentionally blank WCAP-1664 1-NP Rev. 0 C

Appendix C WCAP-16641-NP Rev. 0

C-1 Table C-1 KPS Capsule T Fracture Toughness Test Results SPECIMEN ID W25 W26 I W27 W28 W29I W30 W31 W32 H25 1 H29 H30 H31 Pmax (lbs)-determined from 934 918 1120 1158 1014 1214 816 913 1008 723 588 819 test LL Compliance (mils/lb) 0.0055 0.0045 0.0045 0.0048 0.0055 0.0044 0.0050 0.0060 0.0057 0.0080 0.0090 0.0053 Total Area (in-lbs) 2.95 3.56 7.35 22.15 6.19 6.51 1.86 3.40 13.20 4.80 2.18 2.94 PlasticArea (in-lbs), calculated 0.58 1.65 4.53 18.93 3.38 3.29 0.21 0.92 10.28 2.72 0.62 1.16 Ao, Initial Crack length, (in.) 0.2068 0.2110 0.2040 0.2020 0.2076 0.1907 0.2094 0.2138 0.2140 0.2356 0.2470 0.2010 bo, Remaining Ligament (in.) 0.1872 0.1829 0.1900 0.1920 0.1864 0.2033 0.1846 0.1802 0.1800 0.1584 0.1470 0.1930 Ao/W, Crack/width Ratio 0.5249 0.5357 0.5178 0.5127 0.5268 0.4840 0.5314 0.5427 0.5431 0.5980 0.6269 0.5102 f(Ao/W) 2.8863 2.9923 2.8193 2.7731 2.9046 2.5322 2.9495 3.0641 3.0691 3.7423 4.1973 2.7504 Ke (ksi-inA0.5) 48.85 49.79 57.19 58.18 53.33 55.67 43.58 50.69 56.04 48.99 44.68 40.81 Je (in-lbs/in,)!]000 0.0749 0.0778 0.1026 0.1062 0.0893 0.0973 0.0596 0.0806 0.0985 0.0753 0.0626 0.0523 Jp (in-lbs/in2)/l000 0.0186 0.0547 0.1518 0.5966 0.1096 0.0980 0.0068 0.0308 0.3456 0.1039 0.0257 0.0364 Jt (in-lbs/in2)/l000 0.0935 0.1325 0.2545 0.7029 0.1988 0.1952 0.0664 0.1114 0.4442 0.1792 0.0883 0.0887 KJC (ksi-inA0.5) 54.6 65.0 90.1 149.7 79.6 78.9 46.0 59.6 119.0 75.6 53.1 53.2 KJC(1TAdjusted) - (ksi-inAO.5) 47.0 55.3 75.1 122.4 66.8 66.3 40.2 51.0 98.0 63.6 45.8 45.9 WCAP-1664 1-NP Rev. 0 Appendix C

C-2 This page intentionally blank WCAP-16641-NP Rev. 0 Appendix C Appendix C WCAP- 1664 1-NP Rev. 0

C-3 Specimen W-25 SPECIMEN ID W25 Pmax (Ibs) - determined from test 934 Test Temperature 136 0 F LL Compliance (mils/Ib) 0.0055 Total Area (in-lbs) 2.95 PlasticArea (in-lbs). calculated 0.58 Ao, Initial Crack length, (in.) 0.2068 bo, Remaining Ligament (in.) 0.1872 Ao/W Crack/width Ratio 0.5249 f(Ao/W) 2.8863 Ke (ksi-inAO.5) 48.85 Je (in-lbsiin2)/l000 0.0749 Jp (in-lbs/in2)/l000 0.0186 Jt (in-lbsiin2)/l000 0.0935 KJC (ksi-inA0.5) 54.6 KJC(ITAdjusted) - (ksi-inA0.5) 47.0 B Bnet W Span Y.S. T.S.

(in.) (in.) (in.) (in.) (ksi) (ksi) 0.394 0.314 0.394 1.576 107 111 Precracking/Straightness Kmax %of ao %of B 16.8 4.11% 2.16% 1 Load Rate: 0.71 ksi-in 112/sec 3 4W 2w 1 U25 WCAP-16641-NP Rev. 0 Appendix C

C-4 SPECIMEN ID W26 Specimen W-26 Pmax (lbs) - determinedfromn test 918 Test Temperature 136oF LL Compliance (mils/lb) 0.0045 Total Area (in-lbs) 3.56 PlasticArea (in-lbs), calculated 1.65 Ao. Initial Crack length. (in.) 0.211063 bo, Remaining Ligament (in.) 0.1829 Ao/1W Crack/width Ratio 0.5357 f(Ao/W) 2.9923 Ke (ksi-inA0.5) 49.79 Je (in-lbs/in2)/l 000 0.0778 Jp (in-lbs/in2)/1000 0.0547 Jt (in-lbs/in2)/1000 0.1325 KJC (ksi-inA0.5) 65.0 KJC(1TAdjusted) - (ksi-inAO.5) 55.3 B Bnet W Span Y.S. T.S.

(in.) (in.) (in.) (in.) (ksi) (ksi) 0.394 0.314 0.394 1.576 107 111 Precracking/Straightness Kmax %of a. %of B 23.7 5.45% 2.92%

Load Rate: 0.73 ksi-in 112/sec

• 0oo

-od -* DIpW.-mI Appendix C WCAP-16641-NP Rev. 0

C-5 SPECIMEN ID W27 Specimen W-27 Pmax (lbs) - determined from test 1120 Test Temperature 136oF LL Compliance (mils/lb) 0.0045 Total Area (in-lbs) 7.35 PlasticArea (in-lbs), calculated 4.53 Ao, Initial Crack length, (in.) 0.2040 bo, Remaining Ligament (in.) 0.1900 Ao/W, Crack/width Ratio 0.5178 f(Ao/14W 2.8193 Ke (ksi-inA0.5) 57.19 2

Je (in-lbs/in )/l000 0.1026 Jp (in-lbs/in2)/l000 0.1518 Jt (in-lbs/in2)/1000 0.2545 KJC (ksi-inA0.5) 90.1 KIC(I TAdjusted) - (ksi-inAO.5) 75.1 B Bnet W Span Y.S. T.S.

(in.) (in.) (in.) (in.) (ksi) (ksi) 0.394 0.314 10.394 1 1.576 107 111 Precracking/Straightness Kmax %of a. %of B IW 22.2 3.67% 1.90%

Load Rate: 0.74 ksi-inl/2 /sec 2.0

.0U ý M WCAP-16641-NP Rev. 0 Appendix C

C-6 Specimen W-28 Test Temperature 136oF SPECIMEN ID W28 Pmax (lbs) - determined from test 1158 LL Compliance (mils/lb) 0.0048 Total Area (in-lbs) 22.15 PlasticArea (in-lbs), calculated 18.93 Ao, Initial Crack length, (in.) 0.2020 bo, Remaining Ligament (in.) 0.1920 Ao/W Crack/width Ratio 0.5127 f(Ao/W) 2.7731 Ke (ksi-inAO.5) 58.18 Je (in-lbs/in2)/1000 0.1062 Jp (in-lbs/in2)/l000 0.5966 Jt (in-lbs/in2)/l000 0.7029 KJC (ksi-inA0.5) 149.7 KJC(1 TAdjusted) - (ksi-hA0.5) 122.4 B Bnet W Span Y.S. T.S.

(in.) (in.) (in.) (in.) (ksi) (ksi) 0.394 0.314 0.394 1.576 107 111 Precracking/Straightness s...

Straight -0 Kmax 1400 22.1 3.96%

Load Rate: 0.74 ksi-inl/ 2/sec 4WO 0 02 02 L..a U-. Ohpo.n,.

WCAP-1664 1-NP Rev. 0 Appendix C Appendix C WCAP-16641-NP Rev. 0

C-7 Specimen W-29 SPECIMEN ID W29 Test Temperature 136 0 F Pmax (lbs)- determined from test 1014 LL Compliance (mils/lb) 0.0055 Total Area (in-lbs) 6.19 PlasticArea (in-lbs), calculated 3.38 Ao, Initial Crack length, (in.) 0.2076 bo, Remaining Ligament (in.) 0.1864 Ao/W Crack/width Ratio 0.5268 f(Ao/W) 2.9046 Ke (ksi-inA0.5) 53.33 Je (in-lbs/in2)/l 000 0.0893 Jp (in-lbs/in2)/1000 0.1096 Jt (in-lbs/in2)/1000 0.1988 KJC (ksi-inA0.5) 79.6 KJC(IT Adjusted) - (ksi-inA0.5) 66.8 B Bnet W Span Y.S. T.S.

(in.) (in.) (in.) (in.) (ksi) (ksi) 0.394 0.314 0.394 1.576 107 111 2WOO Precracking/Straightness Iwo Kmax %of ao %of B 22.9 4.10% 2.16%

Load Rate: 0.73 ksi-inl/2/sec 000 400 zoo 0¸05 LoadUm Di..placi n...

Appendix C WCAP-1664 WCAP- Rev. 00 1-NP Rev.

1664 1-NP Appendix C

C-8 Specimen W-30 SPECIMEN ID W30 Test Temperature 136 OF Pmax (ibs) - determinedfrom test 1214 LL Compliance (mils/lb) 0.0044 Total Area (in-lbs) 6.51 PlasticArea (in-lbs), calculated 3.29 Ao, Initial Crack length, (in.) 0.1907 bo, RemainingLigament (in.) 0.2033 Ao/1W, Crack/width Ratio 0.4840 f(Ao1W) 2.5322 Ke (ksi-inA0.5) 55.67 Je (in-lbs/in2)/1i000 0.0973 Jp (in-lbs/in2)/l000 0.0980 Jt (in-lbs/in2)/l000 0.1952 KJC (ksi-inAO.5) 78.9 KJC(IT Adjusted) - (ksi-inA0.5) 66.3 B Bnet W Span Y.S. T.S.

(in.) (in.) (in.) (in.) (ksi) (ksi) 0.394 0.314 0.394 1.576 107 111 Precracking/Straightness Kmax %of a. %of B 19.9 7.08% 3.43%

12 Load Rate: 0.69 ksi-inl/ 2 /sec LoanUi- Displboemat WCAP-16641-NP Rev. 0 Appendix C Appendix C WCAP- 1664 1-NP Rev. 0

C-9 Specimen W-31 Test Temperature 136 0 F SPECIMEN ID W31 Pmax (lbs) - determined from test 816 LL Compliance (mils/lb) 0.0050 Total Area (in-lbs) 1.86 PlasticArea (in-lbs), calculated 0.21 Ao, Initial Crack length, (in.) 0.2094 bo, RemainingLigament (in.) 0.1846 Ao/W, Crack/width Ratio 0.5314 f(Ao/W9 2.9495 Ke (ksi-inA0.5) 43.58 Je (in-lbs/in2)/I000 0.0596 Jp (in-lbs/in2)/l000 0.0068 Jt (in-lbs/in2)il000 0.0664 KJC (ksi-inA0.5) 46.0 KJC(1T Adjusted) - (ksi-in,0.5) 40.2 B Bnet W Span Y.S. T.S.

(in.) (in.) (in.) (in.) (ksi) (ksi) 0.394 0.314 0.394 1.576 107 111 Precracking/Straightness Kmax %of ao %of B 19.6 2,87% 1.52%

-l 11 Load Rate: 0.7391 ksi-in 2/sec 6OO Pop-in: 4OO Ci/Co = 1.32 >1.029 Pop-in terminated test record.

Load UneDtpI~ce=t WCAP- 16641-NP Rev. 0 Appendix C

C-10 Specimen W-32 Test Temperature 136 0 F SPECIMEN ID W32 Pmax (Ibs) - determined from test 913 LL Compliance (mils/lb) 0.0060 Total Area (in-lbs) 3.40 PlasticArea (in-lbs), calculated 0.92 Ao, Initial Crack length, (in.) 0.2138 bo, Remaining Ligament (in.) 0.1802 Ao/1W, Crack/width Ratio 0.5427 f(Ao/9) 3.0641 Ke (ksi-inAO.5) 50.69 2

Je (in-lbs/in )/l000 0.0806 Jp (in-lbs/in2)/l&00 0.0308 Jt (in-lbs/in2)/l&00 0.1114 KJC (ksi-inAO.5) 59.6 KJC(1TAdjusted) - (ksi-inAO.5) 51.0 B Bnet W Span Y.S. T.S.

(in.) (in.) (in.) (in.) (ksi) (ksi) 0.394 0.314 0.394 1.576 107 111 2WO Precracking/Straightness Kmax %of a. %of B 23.9 3.51% 1.90% MO W.

.0 Load Rate: 0.73 ksi-in"12/sec

.0 400

.0 0 U2 0 025 Losa 001 4.~c~,~

WCAP-16641-NP Rev. 0 Appendix C C WCAP- 1664 1-NP Rev. 0

C-11 Specimen H-25 SPECIMEN ID H25 Pmax (ibs) - determined from test 1008 Test Temperature LL Compliance (mils/lb) 0.0057 136 0 F Total Area (in-lbs) 13.20 PlasticArea (in-lbs), calculated 10.28 Ao, Initial Crack length, (in.) 0.2140 bo, Remaining Ligament (in.) 0.1800 Ao/IW, Crack/width Ratio 0.5431 f(Ao/W 3.0691 Ke (ksi-inA0.5) 56.04 2

Je (in-lbs/in)/l 000 0.0985 Jp (in-lbs/in2)/l000 0.3456 It (in-lbs/in2)/1000 0.4442 KJC (ksi-inA0.5) 119.0 KJC(1T Adjusted) - (ksi-inAO.5) 98.0 B Bnet W Span Y.S. T.S.

(in.) (in.) (in.) (in.) (ksi) (ksi) 0.394 0.314 0.394 1.576 107 111 Precracking/Straightness Kmax %of ao %of B ,

18.7 5.15% 2.79%

Load Rate: 0.36 ksi-inl/ 2/sec r.o 4.0 OW5 WCAP- 1664 1-NP Rev. 0 Appendix C

C-12 Specimen H-29 Test Temperature 136 0 F SPECIMEN ID H29 Pmax (Ibs) - determined from test 723 LL Compliance (mils/lb) 0.0080 Total Area (in-lbs) 4.80 PlasticArea (in-lbs), calculated 2.72 Ao, Initial Crack length, (in.) 0.2356 bo, Remaining Ligament (in.) 0.1584 Ao/*, Crack/widthRatio 0.5980 f(Ao1W) 3.7423 Ke (ksi-inAO.5) 48.99 2

Je (in-lbs/jn)1 000 0.0753 Jp (in-lbs/in2)/l 000 0.1039 Jt (in-lbsiin2)/l000 0.1792 KJC (ksi-inA0.5) 75.6 KJC(1TAdjusted) - (ksi-jin0.5) 63.6 B Bnet W Span Y.S. T.S.

(in.) (in.) (in.) (in.) (ksi) (ksi) 0.394 0.314 0.394 1.576 107 111 Precracking/Straightness Kmax %of a. %of B 11C 19.1 7.22% 4.31%

Load Rate: 0.38 ksi-inl/ 2/sec L-~ U-.iI.~0 Appendix C WCAP-1664 1-NP Rev. 0

C-13 Specimen H-30 Test Temperature 136 0 F SPECIMEN ID H30 Pmax (lbs) - determined from test 588 LL Compliance (inils/lb) 0.0090 Total Area (in-lbs) 2.18 PlasticArea (in-lbs). calculated 0.62 Ao, Initial Crack length, (in.) 0.2470 bo. Remaining Ligament (in.) 0.1470 Ao/W Crack/width Ratio 0.6269 f(Ao/W) 4.1973 Ke (ksi-inAO,5) 44.68 2

Je (in-lbsiin)/l000 0.0626 Jp (in-lbs/in2)l000 0.0257 Jt (in-lbs/in2)/l000 0.0883 KJC (ksi-inA.5) 53.1 KJC(l TAdjusted) - (ksi-inA0.5) 45.8 B Bnet W Span Y.S. T.S.

(in.) (in.) (in.) (in.) (ksi) (ksi) 0.394 0.314 0.394 1.576 107 111 20OO Precracking/Straightness Kmax %of ao %of B ..

13.2 3.44% 2.16% .

SR2000 Load Rate: 0.37 ksi-in l/2sec o D20 5 Appendix C WCAP-16641-NP WCAP- Rev. 00 1664 1-NP Rev. Appendix C

C-14 SPECIMEN ID H31 Specimen H-31 Pmax (lbs) - determinedfrom test 819 Test Temperature 136 0 F LL Compliance (mils/lb) 0.0053 Total Area (in-lbs) 2.94 PlasticArea (in-lbs), calculated 1.16 Ao, Initial Crack length, (in.) 0.2010 bo, Remaining Ligament (in.) 0.1930 Ao/W, Crack/width Ratio 0.5102 f(Ao/1W) 2.7504 Ke (ksi-inAO.5) 40.81 Je (in-lbs/in2 )/1000 0.0523 Jp (in-lbs/in2)/l000 0.0364 Jt (in-lbs/in2)/l000 0.0887 KJC (ksi-inA0.5) 53.2 KJCO(TAdjusted) - (ksi-inA0.5) 45.9 B Bnet W Span Y.S. T.S.

(in.) (in.) (in.) (in.) (ksi) (ksi) 0.394 0.314 0.394 1.576 107 111 Precracking/Straightness low 1.

Kmax %of ao %of B low 16.4 1.99% 1.02%

12.

low Load Rate: 0.34 ksi-inl/2/sec A Pop-in:

Ci/Co = 1.032 >1.029 200 Pop-in terminated test record.

Load*nmOl*l*

WCAP-16641-NP Rev. 0 Appendix C C WCAP- 1664 1-NP Rev. 0

C-15 Table C-2 Unirradiated Pre-cracked Charpy Specimens Tested at -200'F SPECIMEN ID wps201 wps202 wps203 wps204* Iwps205 1wps206 wps207 Iwps208 Wps209 ps210 Pmax (lbs) - determined from test 1341 1253 1550 1261 1250 1274 1253 1364 1374 1468 LL Compliance (mils/lb) 0.0062 0.0081 0.0071 0.0115 0.0081 0.0068 0.0087 0.0059 0.0079 0.0059 Total Area (in-lbs) 15.40 6.92 9.49 16.60 7.52 6.12 7.76 8.56 10.65 9.66 PlasticArea (in-Ibs), calculated 9.87 0.58 0.97 7.43 1.21 0.57 0.95 3.12 3.17 3.33 Ao, InitialCrack length, (in.) 0.2080 0.2060 0.1988 0.1988 0.2100 0.2030 0.2100 0.2050 0.2030 0.1988 bo, RemainingLigament (in.) 0.1860 0.1880 0.1953 0.1953 0.1840 0.1910 0.1840 0.1890 0.1910 0.1953 Ao/*W, Crack/width Ratio 0.528 0.523 0.504 0.504 0.533 0.515 0.533 0.520 0.515 0.504 f(Ao/W) 2.915 2.867 2.700 2.700 2.965 2.796 2.965 2.843 2.796 2.700 Ke (ksi-inAO.5) 63.24 58.09 67.70 55.08 59.95 57.61 60.10 62.71 62.14 64.12 Je (in-lbs/in) 0.1333 0.1125 0.1528 0.1011 0.1198 0.1106 0.1204 0.1311 0.1287 0.1371 Jp (in-lbs/in2) 0.2558 0.0149 0.0240 0.1834 0.0317 0.0145 0.0250 0.0795 0.0800 0.0822 Jt (in-lbs/in2) 0.3891 0.1274 0.1768 0.2845 0.1515 0.1251 0.1454 0.2106 0.2087 0.2193 KJC (ksi-inAO.5) 108.0 61.8 72.8 92.4 67.4 61.3 66.1 79.5 79.1 81.1 KJC(1TAdjusted) - (ksi-inA0.5) 89.4 52.8 61.5 77.0 57.2 52.3 56.1 66.8 66.5 68.0

• Invalid precracks Appendix C WCAP-1664 1-NP Rev.

WCAP-16641-NP Rev. 0 Appendix C

C-16 Table C-3 Unirradiated Reconstituted Pre-cracked Charpy Specimens Tested at -200"F SPECIMEN ID rkwl rkw3 rkw6 rkw7 I rkw8 I rkwlO rkwl I rkw2 Pmax (lbs) - determinedfrom test 1255 1378 1472 1421 1582 1468 1422 1556 LL Compliance (mils/lb) 0.0045 0.0049 0.0065 0.0059 0.0043 0.0043 0.0057 0.0040 Total Area (in-lbs) 10.12 8.02 6.01 8.54 11.56 14.12 6.32 19.90 PlasticArea (in-lbs), calculated 6.58 3.37 0.00 2.59 6.18 9.49 0.56 15.01 Ao, Initial Cracklength, (in.) 0.2040 0.1970 0.1880 0.1940 0.1870 0.1920 0.1910 0.1920 bo, Remaining Ligament (in.) 0.1900 0.1970 0.2060 0.2000 0.2070 0.2020 0.2030 0.2020 Ao/1W Crack/width Ratio 0.5178 0.5000 0.4772 0.4924 0.4746 0.4873 0.4848 0.4873 f(Ao/W) 2.819 2.663 2.480 2.599 2.460 2.558 2.538 2.558 Ke (ksi-inAO.5) 57.22 59.32 59.02 59.72 62.97 60.73 58.37 64.38 2

Je (in-lbs/in) 0.1091 0.1173 0.1161 0.1189 0.1322 0.1229 0.1136 0.1382 Jp (in-lbs/in2) 0.1669 0.0825 0.0000 0.0624 0.1439 0.2266 0.0133 0.3584 Jt(in-lbs/in2) 0.2761 0.1998 0.1161 0.1813 0.2761 0.3495 0.1269 0.4966 KJC (ksi-inAO.5) 91.0 77.4 59.0 73.7 91.0 102.4 61.7 122.1 KJC(1TAdjusted) - (ksi-inA0.5) 75.9 65.1 50.5 62.2 75.9 84.9 52.7 100.5 Appendix C WCAP-16641-NP Rev. 0

C-17 Table C-4 Unirradiated 1/2 T Compact Tension Specimens Tested at -187"F SPECIMEN ID wpsl01 wpsl02 wpsl03 wpsl04 wpsl05 wpsl06 Pmax (lbs)- determinedfrorn test 3869 2867 3750 3445 3433 2416 LL Compliance (mils/Ib) 0.0026 0.0029 0.0025 0.0029 0.0027 0.0024 Total Area (in-ibs) 24.02 13.69 22.25 17.49 17.13 7.13 PlasticArea (in-ibs), calculated 4.56 1.77 5.00 0.41 1.14 0.04 Ao, Initial Crack length, (in.) 0.5200 0.5280 0.5230 0.5250 0.5220 0.5170 bo, Remaining Ligament (in.) 0.4800 0.4720 0.4770 0.4750 0.4780 0.4830 Ao/*W, Crack/width Ratio 0.5200 0.5280 0.5230 0.5250 0.5220 0.5170 f(Ao/W) 10.134 10.400 10.233 10.299 10.200 10.038 Ke (ksi-inAO.5) 78.42 59.63 76.74 70.96 70.03 48.50 Je (in-lbs/lir) 0.2050 0.1185 0.1963 0.1678 0.1635 0.0784 Jp (in-lbs/in2) 0.0427 0.0169 0.0471 0.0039 0.0108 0.0004 Jt (in-lbs/in2) 0.2477 0.1354 0.2434 0.1717 0.1742 0.0788 KJC (ksi-inA0.5) 86.2 63.7 85.5 71.8 72.3 48.6 KJC(ITAdjusted) - (ksi-inAO.5) 75.4 56.5 74.8 63.3 63.7 43.8 Appendix C WCAP-16641-NP Rev. 00 WCAP-16641-NP Rev. Appendix C

C-18 Table C-5 Irradiated Pre-cracked Capsule S Charpy Specimens Tested at 136"F SPECIMEN ID T w24 W19 h17 h18 w23 I h20* h19 w17 h21 Pmax (Ibs) - determinedfrom test 1514 1349 1433 1285 1490 1319 1273 1323 1076 LL Compliance (mils/lb) 0.0069 0.0070 0.0071 0.0073 0.0071 0.0071 0.0071 0.0071 0.0082 Total Area (in-lbs) 15.80 8.49 25.90 20.36 23.03 29.60 9.97 15.57 15.09 PlasticArea (in-Ibs), calculated 7.90 2.12 18.61 14.34 15.13 23.42 4.19 9.35 10.33 Ao, Initial Crack length, (in.) 0.1920 0.1920 0.1960 0.2030 0.1990 0.2000 0.1970 0.1970 0.2100 bo, Remaining Ligament (in.) 0.2020 0.2020 0.1980 0.1910 0.1950 0.1940 0.1970 0.1970 0.1840 Ao/1W Crack/width Ratio 0.4873 0.4873 0.4975 0.5152 0.5051 0.5076 0.5000 0.5000 0.5330 f(Ao/W) 2.5585 2.5585 2.6412 2.7960 2.7059 2.7280 2.6625 2.6625 2.9652 Ke (ksi-inAO.5) 62.65 55.84 61.20 58.10 65.20 58.20 54.81 56.97 51.60 Je (in-lbs/in2) 0.1308 0.1039 0.1248 0.1125 0.1417 0.1129 0.1001 0.1082 0.0888 Jp (in-lbs/in2) 0.1887 0.0505 0.4534 0.3620 0.3742 0.5821 0.1025 0.2289 0.2707 Jt (in-lbs/in2) 0.3195 0.1545 0.5782 0.4745 0.5159 0.6950 0.2026 0.3371 0.3594 KJC (ksi-inA0.5) 97.9 68.1 131.7 119.3 124.4 144.4 78.0 100.6 103.8 KJC(1TAdjusted) - (ksiinA0.5) 81.3 57.7 108.1 98.3 102.3 118.2 65.6 83.5 86.1 C WCAP-16641-NP Rev. 0 Appendix C WCAP- 1664 1-NP Rev. 0

C-19 Table C-6 Irradiated Pre-cracked Capsule S Charpy Specimens Tested at 59"F SPECIMEN ID [ W21 [W20 W22 Pmax (lbs) - determined from test 1100 1220 1330 CLL (mils/lb) - determinedfrom initial unloadingsof test 0.0071 0.0071 0.0071 Total Area beneath Load-Disp Curve 3.77 3.77 3.85 PlasticArea (in-ibs), calculated 0.00 0.00 -2.43 Ao, Initial Crack length, (in.) 0.2 0.2 0.2 bo, Remaining Ligament (in.) 0.194 0.194 0.194 Ao/W, Crack/width Ratio 0.508 0.508 0.508 f(Ao/W) 2.728 2.728 2.728 Ke 48.54 53.83 58.68 Je 0.0785 0.0966 0.1148 Jp 0.0000 0.0000 -0.0604 Jt 0.0785 0.0966 0.0544 KJC (ksi-inAO.5) 48.5 53.8 40.4 KJC(ITAdjusted) - (ksi-inAO.5) 42.2 46.4 35.8 Appendix C WCAP-16641-NP Rev. 0 WCAP- 1664 1-NP Rev. 0 Appendix C

C-20 Table C-7 Irradiated Precracked Maine Yankee Specimens Tested at 210"F SPECIMEN ID 322 36a 313 1 371a 33u 375 1 371b I 37ua Pmax (lbs) - determined from test 1482 1220 1373 1570 1259 1436 1457 1722 LL Compliance (mils/lb) 0.0065 0.0071 0.0071 0.0059 0.0073 0.0071 0.0060 0.0064 TotalArea (in-lbs) 9.96 5.28 14.77 35.00 7.26 10.57 9.52 15.82 PlasticArea (in-lbs), calculated 2.82 0.00 8.07 27.70 1.48 3.26 3.18 6.35 Ao, Initial Cracklength, (in.) 0.1840 0.2020 0.1930 0.1860 0.2020 0.1960 0.1900 0.1970 bo, RemainingLigament (in.) 0.2100 0.1920 0.2010 0.2080 0.1920 0.1980 0.2040 0.1970 Ao/*W Crack/widthRatio 0.4670 0.5127 0.4898 0.4721 0.5127 0.4975 0.4822 0.5000 f(AoiW) 2.4042 2.7731 2.5788 2.4415 2.7731 2.6412 2.5185 2.6625 Ke (ksi-inAO.5) 57.63 54.72 57.27 62.00 56.47 61.34 59.35 74.15 Je (in-lbs/in2) 0.1107 0.0998 0.1093 0.1281 0.1063 0.1254 0.1174 0.1833 Jp (in-lbs/in2) 0.0648 -0.0001 0.1936 0.6423 0.0372 0.0794 0.0752 0.1553 Jt(in-lbs/in2) 0.1755 0.0997 0.3029 0.7704 0.1435 0.2048 0.1927 0.3386 KIC (ksi-inA0.5) 72.6 54.7 95.3 152.0 65.6 78.4 76.0 100.8 KJC(1TAdjusted) - (ksi-inAO.5) 61.3 47.1 79.3 124.2 55.8 65.9 64.0 83.6 WCAP-1664 1-NP Rev. 0 C

Appendix C WCA-P- 1664 1-NP Rev. 0

C-2 1 Table C-8 Unirradiated Maine Yankee Weld Specimen Data SPECIMEN ID C04-4 [ C04-5 C04-2 C04-7 C04-8 [ C04-3 [C04-6 Pmax(lbs)-determinedfrom test 1070 1110 1230 1150 1170 1180 1180 LL Compliance (mils/lb) 0.0090 0.0090 0.0089 0.0095 0.0096 0.0088 0.0089 Total Area (in-lbs) 7.01 7.98 12.93 12.29 13.42 13.61 14.09 PlasticArea (in-lbs), calculated 1.86 2.44 6.20 6.00 6.84 7.49 7.89 Ao, Initial Cracklength, (in.) 0.2031 0.2040 0.2003 0.2084 0.2032 0.2060 0.2045 bo, Remaining Ligament (in.) 0.1909 0.1900 0.1937 0.1856 0.1908 0.1880 0.1895 Ao/W Crack/width Ratio 0.5155 0.5178 0.5084 0.5289 0.5157 0.5228 0.5190 f(A o/1W) 2.7984 2.8193 2.7347 2.9252 2.8007 2.8667 2.8310 Ke (ksi-inAO.5) 49.14 51.35 55.20 55.20 53.77 55.51 54.82 Je (in-lbs/in)/lO00 0.0792 0.0865 0.0999 0.0999 0.0948 0.1011 0.0986 Jp (in-lbs/in2)/lQ00 0.0469 0.0618 0.1543 0.1560 0.1730 0.1920 0.2008 Jt (in-lbs/in2)/lO00 0.126 0.148 0.254 0.256 0.268 0.293 0.299 KJC (ksi-in'0.5) 62.0 67.2 88.0 88.3 90.4 94.5 95.5 KJC(1TAdjusted) - (ksi-inA0.5) 52.9 57.1 73.5 73.8 75.4 78.7 79.5 Appendix C WCAP-16641-NP Rev. 0 WCA-P-16641-NP Rev. 0 Appendix C

C-22 This page intentionally blank WCAP-16641-NP Rev. 0 Appendix C Appendix C WCAP- 1664 1-NP Rev. 0

APPENDIX D CHARPY V-NOTCH PLOTS FOR THE KPS CAPSULE T SURVEILLANCE MATERIALS Appendix D 16641-NP WCAP- 1664 Rev. 00 1-NP Rev. Appendix D

This page intentionally blank Appendix D WCAP-16641-NP Rev. 0

D-1 A summary of the Upper Shelf Energy values for all of the capsule materials for the Kewaunee Power Station are shown in Table D-1. The definition for Upper Shelf Energy (USE) is given in ASTM E185-82, Section 4.18, and reads as follows:

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

Westinghouse typically reports the average of all Charpy data > 95% shear as the USE. In some instances, there may be data deemed 'out of family' and are removed from the determination of the USE based on engineering judgement. For the weld material, only one data point exists from Capsule T, so the upper shelf will be set to that value of 72 ft-lbs.

Table D-1 Summary of the Upper Shelf Energy Values For Kewaunee Power Station Capsule Analyses Material Unirradiated Capsule V Capsule R Capsule P Capsule S Capsule T Intermediate Shell Forging 160 ft-lbs 180 ft-lbs 160 ft-lbs 157 ft-lbs 148 ft-lbs 139 ft-lbs 122X208VA1 (Tangential)

Lower Shell Forging 157 ft-lbs 178 ft-lbs 153 ft-lbs 159 ft-lbs 152 ft-lbs 144 ft-lbs 123X167VAI (Tangential)

Weld Metal 126 ft-lbs 82 ft-lbs 78 ft-lbs 76 ft-lbs 64 ft-lbs 72 ft-lbs (Heat #

1P3571)

Heat Affected 180 ft-lbs 145 ft-lbs 141 ft-lbs 136 ft-lbs 139 ft-lbs Zone Material Correlation 123 ft-lbs 109 ft-lbs 95 ft-lbs 101 ft-lbs 98 ft-lbs 91 ft-lbs Monitor Material Westinghouse typically reports the average of all Charpy data > 95% shear as the USE. In some instances, there may be data deemed 'out of family' and are removed from the determination of the USE based on engineering WCAP- 1664 1-NP Rev. 0 Appendix D

D-2 Weld Metal There were three data points generated for the surveillance weld based on the reconstitution of the HAZ charpy specimens. The specimens were reconstituted in accordance with ASTM E 1253 [Reference D-1]

with the weld portion of the HAZ specimen used to represent the weld metal (see Figure 1 from ASTM E 1253-99 shown below).

,EE- 10 law~

(Note a)

HAZ

- 5 ina

//// Reconatifution Weld

///

k End Tab End Tab I

Central Tont Section (Not&A)

NOTE A-No plastic deformation from previous testing is permitted in the region between the HAZs caused by the reconstitution welds (the central test section).

NOTE B-Temperature during welding shall not exceed the irradiation tempera-ture.

FIG. I Schematic of a Reconstituted Charpy Specimen WCAP-1664 1-NP Rev. 0 Appendix DD WCAP- 1664 1-NP Rev. 0

D-3 The actual values obtained from the Charpy testing of the HAZ reconstituted specimens are shown below

[Reference D-2]:

Charpy Energy Lateral Test Temperature Specimen ID (OF) (ft-lbs) Expansion  % Shear (mils)

H26 190 13 13 25 H27 215 36 28 45 H32 400 72 59 100 With only three data points available, using a tanh fit to the data is not an option. In order to best estimate the T 30 and T 50 energy values through the transition regions, a review of the data from Capsule S will be done to see how the material behaved. Typically, through engineering experience, the slope through the transition region will decrease with increased fluence. If the slope through transition region of Capsule S data is a known quantity, we can then approximate the anticipated slope through the Capsule T data. This will allow estimated T 30 and T5 0 values to be determined.

The approach taken to determine the T 30 and T5 0 values for Capsule T is described below:

1. Determine the slope of the transition region for Capsule S weld charpy data (this is the latest data set available from Kewaunee) based on the T 30 and T5 0 values previously determined.

2. Based on engineering experience, the slope decrease with increases in fluence. Therefore, based on engineering judgement, the slope determined for the Capsule S transition region will be decreased by 10%.

3. Apply the slope from the Capsule S data (minus 10%) to the Charpy data obtained from the Capsule T testing and establish a best fit through the transition region.

4. Determine the value of T 30 and T50 based on the best fit through the transition region.

Using this approach, the slope through the transition region for Capsule S data was determined by calculating the slope going from the T 30 value to the T50 value. Shown below:

Measured T 30 = 200F Measured T50 = 258F Slope = (50-30)/(258/200) = 0.345 A decrease of 10% would results in a slope of 0.3108 through the transition region of the Capsule T data.

WCAP- 1664 1-NP Rev. 0 Appendix D

D-4 Taking this ratio and applying this slope through the transition region of the Capsule T weld data results in the fit shown in Figure D-1. Note the sloped line was shifted left and right until the line represented what was considered to be the best fit through the data.

KPS Weld CVN Data 80 70 60 50 LU 40 z 30 20 10 0

0 100 200 300 400 500 Temperature (Deg F)

Figure D-1 Based on the results of the best fit line through the transition region, the T 30 value for the weld material will be approximated to 2210F. The T5 0 value for the weld material will be approximated to be 285 0 F.

A similar approach was taken for the lateral expansion data where the data from Capsule S was used to determine the slope through the Capsule T data (minus 10%).

Based on the results from the Capsule S report, there is a measured value of T35 and from the chart in Reference D-2, a second value can be determined. This data is shown below.

From Table 5-13 [Reference D-3]

Appendix D WCAP- 1664 1-NP Rev. 0

D-5 T35 = 238°F From Figure 5-4 [Reference D-3], an approximated value was:

@ 200'F, LE = 22 mils Slope = (35-22)/(238-200) = 0.342 A 10% decrease results in a lateral expansion transition region slope value of 0.308 for Capsule T. This value was then used to approximate a best fit line with the results plotted in Figure D-2.

KPS Weld Lateral Expansion Data 70 60 S50

• 40 w 30

'D

-J 20 10 0

0 100 200 300 400 500 Temperature (Deg F)

Figure D-2 Based on the results of the best fit line through the transition region, the T35 mils value for the weld material will be approximated to 2490F.

WCAP- 16641-NP Rev. 0 Appendix D

D-6 Forging and Correlation Monitor Materials For the remaining materials within the capsule, a generic fit to the data, as was done in previous WCAPs, was performed to determine the material properties.

For the KPS Forging 122X208VA1 material, the data was plotted onto a chart with a fitted curve drawn through the data to represent a best-fit for the Charpy Energy, Lateral Expansion and Shear Data. Shown in the following Figures are the results for this fit.

KPS Forging 122X208VA1 (Tangential) 160 -

140 _ _ _ _ _

40 100- 0_0020_30_0 500 80 C

20 ____________

-100 0 100 200 300 400 500 Temperature (Deg F)

From the fitted curve, the resulting T 30 and T5 0 values for Charpy Energy are:

T0= 65*F To= 110'F Appendix D WCAP-16641-NP Rev. 0

D-7 From the fitted curve, the resulting T35 value for lateral Expansion is:

T35mihs = 100°F KPS Forging 122X208VA1 (Tangential) 100 - _ _- _

70 -- _

60 S50 __ __ _____ ____

C 40 30_

20 -_______

-100 0 100 200 300 400 500 Temperature (Deg F)

Appendix D WCAP- 1664 11-NP WCAP-1664 Rev. 00

-NP Rev. Appendix D

D-8 For the KPS Forging 123X 167VAI material, the data was plotted onto a chart with a fitted curve drawn through the data to represent a best-fit for the Charpy Energy, Lateral Expansion and Shear Data. Shown in the following Figures are the results for this fit.

KPS Forging 123X167VA1 (Tangential) 160-140 l -- ,,,,.--

120 __

S1001

.0 LU z 60__

40 20 _ *

-100 0 100 200 300 400 Temperature (Deg F)

From the fitted curve, the resulting T30 and T5 0 values for Charpy Energy are:

T 30 = 20-F T 5o = 50°F KPS Forging 123X167VA1 (Tangential) 100 80-____

o 60-_____ _ _ _ _ _ _

C 0.

Lu 40-M

-120

-100 0 100 200 300 400 Temperature (Deg F)

From the fitted curve, the resulting T35 value for lateral Expansion is:

T35mils = 50OF WCAP-1664 1-NP Rev. 0 Appendix D Appendix D WCAP- 1664 1-NP Rev. 0

D-9 KPS Forging 123X167VA1 (Tangential)

U)

-100 100 200 300 400 Temperature (Deg F)

Rev. 00 Appendix D WCAP- 1664 11-NP WCAP-1664 -NP Rev. Appendix D

D-1O For the KPS Forging Correlation Monitor material, the data was plotted onto a chart with a fitted curve drawn through the data to represent a best-fit for the Charpy Energy, Lateral Expansion and Shear Data.

Shown in the following Figures are the results for this fit.

KPS Correlation Monitor Material 100-80-

-rS60 40 z

20 _

0 L 4 0 100 200 300 400 500 Temperature (Deg F)

From the fitted curve, the resulting T 30 and T50 values for Charpy Energy are:

T0= 220'F T 5 0 =245'F KPS Correlation Monitor Material 80 60

.2 20 0-0 100 200 300 400 500 Temperature (Deg F)

From the fitted curve, the resulting T35 value for Lateral Expansion is:

T35mils = 245 0F Appendix D WCAP- 1664 1-NP Rev. 0

D-11 KPS Correlation Monitor Material 100 9080 70 60- -

40.

80 ._ . . . . . . .__ _ _ _ _ _ _ _

30..

20 10 0 100 200 300 400 500 Temperature (Deg F)

Appendix D References D-1 ASTM E 1253-99, Annual Book of Standards, StandardGuide for Reconstitution of Irradiated Charpy-Sized Specimen.

D-2 STC Report STD-MCE-06-43, Charpy V-notch and Tensile Test Results of the Dominion Nuclear Operating Company Kewaunee Nuclear Plant Unit No. I Reactor Vessel Radiation Surveillance Program, Capsule T, R.G. Lott, et al, October 2006.

D-3 WCAP-14279, Revision 1, Evaluation of Capsulesfrom the Kewaunee and Capsule A-35 from the Maine Yankee Nuclear PlantReactor Vessel Radiation Surveillance Programs,C.C. Kim, et al, September 1998.

Appendix D WCAP-1664 WCAP- Rev. 0 1-NP Rev.

1664 1-NP 0 Appendix D

D-12 This page intentionally blank Appendix D WCAP- 1664 1-NP Rev. 0

APPENDIX E KEWAUNEE POWER STATION SURVEILLANCE PROGRAM CREDIBILITY EVALUATION Appendix E WCAP- 1664 11-NP WCAP-1664 Rev. 00

-NP Rev. Appendix E

This page intentionally blank WCAP-1664 1-NP Rev. 0 Appendix EE WCAP- 1664 1-NP Rev. 0

E-1 INTRODUCTION:

Regulatory Guide 1.99, Revision 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

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

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

EVALUATION:

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

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

...the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that arepredicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiationdamage.

The KPS reactor vessel consists of the following beltline region materials:

Intermediate shell forging 122X208VA1 Lower Shell forging 123X 167VA 1 Circumferential Weld 1P3571 All of these beltline materials were contained in the surveillance program. Based on this discussion, Criterion I is met for the KPS reactor vessel.

Appendix E WCAP-1664 WCAP- 1664 11-NP Rev. 00

-NP Rev. Appendix E

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

Based on engineering judgment, the scatter in the data presented in these plots is small enough to permit the determination of the 30 ft-lb temperature and USE of the KPS surveillance materials. Hence, the KPS surveillance program meets this criterion. This holds true for the weld in that a 30 ft-lb temperature could be determined and an USE established even with only three data points.

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of ARTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 171F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82.

The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for these data and to determine if the scatter of ARTNDT values about this line is less than 287F for welds and less than 17°F for the plate.

The KPS intermediate and lower shell forgings and surveillance weld will be evaluated for credibility.

Table E- I contains the calculation of chemistry factors for the KPS reactor vessel beltline materials contained in the surveillance program. These chemistry factors are calculated per Regulatory Guide 1.99, Revision 2, Position 2.1. [Note that when applyingsurveillance weld data to determine the vessel CFper position 2.1, an adjustment called the "RatioProcedure"is applied. This "Ratio" is not required when evaluating the credibilityof the surveillance weld data since we want to determine the CF for the surveillance weld so thatpredictedsurveillance weld shifts can be measured againstmeasuredshifts.]

WCAP-1664 1-NP Rev. 0 Appendix EE Appendix WCAP- 1664 1-NP Rev. 0

E-3 Table E- 1 Chemistry Factor Determination for the Kewaunee Power Station Surveillance Materials Material Capsule Capsule f~a) FF(b) ARTNDT(c) FF*ARTNDT FF2 V 0.586 0.850 0 0 0.723 R 1.76 1.155 15 17.33 1.334 Intermediate Shell P 2.61 1.257 25 31.43 1.580 Forging S 3.67 1.337 60 80.22 1.788 122X208VAi T 5.62 1.425 90 128.25 2.031 SUM 257.23 7.456 2

CF = I(FF

• ARTNDT) +- X(FF ) = (257.23) + (7.456) = 34.50 V 0.586 0.850 0 0 0.723 R 1.76 1.155 20 23.1 1.334 Lower Shell P 2.61 1.257 20 25.14 1.580 Forging S 3.67 1.337 50 66.85 1.788 123X167VA1 T 5.62 1.425 70 99.75 2.031 SUM: 214.84 7.456 CF = Y(FF

• ARTNDT) + Z(FF 2) = (214.84) + (7.456) = 28.81 V 0.586 0.850 175 148.75 0.723 R 1.76 1.155 235 271.43 1.334 Circumferential P 2.61 1.257 230 289.11 1.580 Weld S 3.67 1.337 250 334.25 1.788 lP3571 T 5.62 1.425 271 386.18 2.031 SUM: 1430.04 7.456 CF = I(FF

• ARTNDT) + E(FF 2) = (1430.04) + (7.456) = 191.8 The scatter of ARTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table D-2 Appendix E WCAP- 1664 11-NP WCAP-1664 Rev. 00

-NP Rev. Appendix E

E-4 Table E-2.

Calculation and Comparison of ARTNDT Values for Kewaunee Power Station Surveillance Materials Predicted Measured Scatter <17 0 F Material CF Capsule Capsule FF ART.DT ARTNDT ARTNDT (Base Metals)

(OF) Fluence (OF) (OF) (OF) <28 0 F

(__F)__ (_F) ( _F (Weld)

Intermediate 34.5 V 0.586 0.850 29.3 0 29.3 N Shell Forging 34.5 R 1.76 1.155 39.9 15 24.9 N 122X208VA1 34.5 P 2.61 1.257 43.4 25 18.4 N 34.5 S 3.67 1.337 46.1 60 -13.9 Y 34.5 T 5.62 1.425 49.2 90 -40.8 N Lower Shell 28.81 V 0.586 0.850 24.5 0 24.5 N Forging 28.81 R 1.76 1.155 33.3 20 13.3 N 123X167VAI 28.81 P 2.61 1.257 36.2 20 16.2 N 28.81 S 3.67 1.337 38.5 50 -11.5 Y 28.81 T 5.62 1.425 41.0 70 -29.0 Y Surveillance 191.8 V 0.586 0.850 163.03 175 -12 Y Weld 191.8 R 1.76 1.155 221.53 235 -13 Y 1P3571 191.8 P 2.61 1.257 241.09 230 11 Y 191.8 S 3.67 1.337 256.44 250 6 Y 191.8 T 5.62 1.425 273.32 271 2 Y These results in Table E-2 indicate that data points fall outside the 1ly of 17'F scatter band for both forgings. Therefore, the forging surveillance data are deemed not-credible per the third criterion. No weld data points fall outside the +/- Ia of 28'F scatter band for the surveillance weld data; therefore, the weld data are deemed credible per the third criterion.

Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 251F.

The capsule specimens are located in the reactor between the fuel and the vessel wall opposite the center of the core. The test capsules are in baskets attached to the thermal shield. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens are subjected to equivalent operating conditions such that the temperatures will not differ by more than 25°F. Hence, this criterion is met.

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

The KPS surveillance program does contain correlation monitor material. NUREG/CR-6413, ORNL/TM-13 133 contains a plot of residual versus fast fluence for the correlation monitor material (Figure 11 in the report). The measured versus predicted reference temperature shifts are presented in Table E-3 for the Kewaunee Correlation Monitor Material.

Appendix E WCAP- 1664 1-NP Rev. 0

E-5 Table E-3 Predicted versus Measured ARTNDT Values for the Kewaunee Correlation Monitor Material Predicted Measured 0 0 Maeil Material (CFF) Capsule Fluence(F)

Cpule FF ARTNDT F ARTNDT (OF) (OF)

Correlation 102 V 0.586 0.850 86.7 95 Monitor 102 R 1.76 1.155 117.81 140 102 P 2.61 1.257 128.21 155 102 S 3.67 1.337 136.37 158 102 T 5.62 1.425 145.35 175 Table E-3 shows a 2oy uncertainty of less than 50'F, which is the allowable scatter in NUREG/CR-6413, ORNL/TM-13133. Hence, this criterion is met.

Conclusion Based on the preceding responses to all five criteria of Regulatory Guide 1.99, Revision 2, the plant forging surveillance data is deemed "non-credible," the weld surveillance data is deemed "credible."

WCAP-1664 1-NP Rev. 0 Appendix E

E-6 This page intentionally blank Appendix E WCAP- 1664 1-NP Rev. 0