Text

Enclosure 5 - WCAP-16643-NP, Revision 2, Kewaunee Power Station Heatup and Cooldown Limit Curves for Normal Operation
	ML101110128
Person / Time
Site:	Kewaunee
Issue date:	06/30/2009
From:	Burgos B, Rosier B; Dominion Energy Kewaunee, Westinghouse
To:	; Office of Nuclear Reactor Regulation
References
	10-047 WCAP-16643-NP, Rev 2
	Download: ML101110128 (67)
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Serial No.10-047 ENCLOSURE 5 LICENSE AMENDMENT 246 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS AND LOW TEMPERATURE OVERPRESSURE PROTECTION WCAP-116643-NP, Revision 2, Kewaunee Power Station Heatup and Cooldown Limit Curves for Normal Operation KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Westinghouse Non-Proprietary Class 3 WCAP-16643-NP June 20 Revision 2 Kewaunee Power Station Heatup and Cooldown Limit Curves for Normal Operation onWestinghouse 09

Westinghouse Non-Proprietary Class 3 WCAP-16643-NP, Revision 2 Kewaunee Power Station Heatup and Cooldown Limit Curves for Normal Operation B. A. Rosier*

B. N. Burgos June 2009 Approved:

Electronically Approved*

C. E. Meyer for P. C. Paesano, Manager Primary Component Asset Management

Electronically approved records are authenticated in the Electronic Document Management System.

Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355

I Westinghouse Non-Proprietary Class 3 ii PREFACE This report has been technically reviewed and verified by: F. C. Gift The Revision 1 scope of this report as indicated below has been verified by: E. J. Long The Revision 2 scope of this report as indicated below has been verified by: E. J. Long*

Electronically approved records are authenticated in the Electronic Document Management System.

RECORD OF REVISION Revision 0:

Revision 1:

Revision 2:

Original Issue Revision 1 is being issued to correct the data in Tables 5-1 and 5-3.

Revision 2 is being issued to correct Reference 3, and to properly format the headers and footers throughout.

P June 2009 Revision 2 WCAP-16643-N

Westinghouse Non-Proprietary Class 3 iii FOREWORD 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 Reactor Include the Use of a Master Curve-based Methodology for Reactor Pressure Vessel Integrity Assessment, 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 Safety Evaluation (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 10' 9 n/cm 2 (E>1.0 MeV).

Additionally, the removal and testing of the capsule with fluence equivalent to 60 years completes 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/cm2 (E>1.0 MeV), which closely approximates and bounds the EOLR vessel fluence.

Master Curve Fracture Toughness T0 Determination The methodology detailed in Appendix A of the NRCSE 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 To for weld wire heat 1 P3571, 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-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 iv 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 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 1 P3571. However, information regarding material properties was 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-16641-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-16641-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.

Heatup and Cooldown Limit Curves for Normal Operation WCAP-16643-NP provides the methodology and results of the generation of heatup and cooldown pressure temperature (PT) limit curves for normal operation of the Kewaunee Power Station reactor vessel. The PT curves were generated for 33 and 52.1 EFPY (EOL and EOLR) based on the latest available reactor vessel information and fluence data utilizing a 95.6% capacity factor.

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Westinghouse Non-Proprietary Class 3 V

The heatup and cooldown curves documented in this report were generated using the most limiting ART values and the NRC approved methodology documented in WCAP-14040-NP-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves." Specifically, the "axial-flaw" and "circ-flaw" methodologies of the 1998 ASME Code,Section XI through the 2000 Addenda was used, which makes use of the Klc methodology.

For weld wire heat 1P3571, the values of ART used corresponded to the EOL and EOLR values documented in WCAP-1 6609-NP, which utilized the approved NRC SE methodology for the Master Curve results.

The highest ART values for 52.1 EFPY were from the upper shell forging 123W250VA1 ("axial-flaw" orientation) and weld wire heat 1 P3571 ("circ-flaw" orientation). The final PT curves were a combination of both the "axial-flaw" and "circ-flaw" limiting ART values. The "circ-flaw" methodology from ASME Code Case N-588 is less restrictive than the standard "axial-flaw" methodology from the 1998 ASME Code,Section XI through 2000 Addenda. But due to low ART values for the forgings, the "circ-flaw" became limiting in various portions of the PT curves. In addition to the use of Code Case N-588, the PT curves also made use of the KI, methodology detailed in ASME Code Case N-640.

WCAP-16643-NP June 2009 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 vi TABLE OF CONTENTS L IS T O F T A B L E S......................................................................................................................................

v ii L IS T O F F IG U R E S......................................................................................................................................

ix E X E C U T IV E SU M M A R Y..........................................................................................................................

xi 1

IN T R O D U C T IO N...........................................................................................................................

I 2

FRACTURE TOUGHNESS PROPERTIES.................................................................................

3 3

CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS................. 8 4

CALCULATION OF ADJUSTED REFERENCE TEMPERATURE........................................ 12 5

HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES.................... 21 6

R E F E R E N C E S..............................................................................................................................

3 5 APPENDIX A THERMAL STRESS INTENSITY FACTORS (K,,)

APPENDIX B PRESSURE-TEMPERATURE LIMIT CURVES FOR 5F/hr HEATUP WITHOUT MARGINS APPENDIX C ENABLE TEMPERATURE CALCULATIONS FOR 33 AND 52.1 EFPY APPENDIX D GAMMA RAY DOSE TO BIOLOGICAL SHIELD AND ASSOCIATED NEUTRON FLUENCE PROJECTIONS WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 vii LIST OF TABLES Table 2-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the Kewaunee. Power Station Reactor Vessel Beltline Materials.......................................... 4 Table 2-2 Summary of the Best Estimate Cu and Ni Weight.Percent and Initial RTNDT Values for the Kewaunee Power Station Reactor Vessel Extended Beltline Materials.......................... 5 Table 2-3 Summary of the Initial RTNDT Values for the Kewaunee Power Station Closure Head and V essel F lan g es.....................................................................................................................

6 Table 2-4 Summary of the Kewaunee Power Station Reactor Vessel Beltline Material Chemistry F acto rs....................................................................

6 Table 2-5 Summary of the Kewvaunee Power Station Reactor Vessel Extended Beltline Material C hem istry F actors...............................................................................................................

7 Table 4-1 Summary of the Maximum Vessel Surface, 1/4T and 3/4T Fluence Values used for the Generation of the 33 and 52.1 EFPY Heatup/Cooldown Curves for the Beltline Materials 14 Table 4-2 Summary of the Vessel Surface, 1/4T and 3/4T Fluence Values used for the Generation of the 52.1 EFPY Heatup/Cooldown Curves for the Extended Beltline Materials............... 14 Table 4-3 Adjusted Reference Temperature Evaluation for 33 EFPY ART Calculations for the Beltline M aterials at the 1/4T Location........................................................................

15 Table 4-4 Adjusted Reference Temperature Evaluation for 52.1 EFPY ART Calculations for the Beltline M aterials at the 1/4T Location........................................................................

15 Table 4-5 Adjusted Reference Temperature Evaluation for 33 EFPY ART Calculations for the Beltline M aterials at the 3/4T Location......................................................................

16 Table 4-6 Adjusted Reference Temperature Evaluation for 52.1 EFPY ART Calculations for the Beltline Materials at the 3/4T Location...... *........................................

16 Table 4-7 Adjusted Reference Temperature Evaluation for 52.1 EFPY ART Calculations for the Extended Beltline M aterials at the 1/4T Location.........................................................

17 Table 4-8 Adjusted Reference Temperature Evaluation for 52.1 EFPY ART Calculations for the Extended Beltline M aterials at the 3/4T Location........................................................ 18 Table 4-9 1P3571 ART ValUes to be Used for Heat-up and Cool-down Curves at the 95.6%

Capacity Factor EOL RPV Maximum Fluence.....

a.......................... 19 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 viii LIST OF TABLES - (continued)

Table 4-10 IP3571 ART Values to be Used for Heat-up and Cool-down Curves at the 95.6%

Capacity Factor EOLR RPV Maximum Fluence.........................................................

19 Table 4-11 Summary of the Limiting 1/4T ART Values Used in the Generation of the Kewaunee Power Station Heatup/Cooldown Limit Curves..........................................................

19 Table 4-12 Summary of the Limiting 3/4T ART Values Used in the Generation of the Kewaunee Power Station Heatup/Cooldown Limit Curves.......................................

20 Table 5-1 33 EFPY Heatup Curve Data Points Using 1998 App. G Methodology (w/Klc, w/Flange Notch, w/Uncertainties for Instrumentation Errors, and w/AP Drop............................ 27 Table 5-2 33 EFPY Cooldown Curve Data Points Using 1998 App. G Methodology (w/Klc, w/Flange Notch, w/Uncertainties for Instrumentation Errors, and w/AP Drop........

28 Table 5-3

.52.1 EFPY Heatup Curve Data Points Using 1998 App. G Methodology (w/Kic, w/Flange Notch, w/Uncertainties for Instrumentation Errors, and w/AP Drop............................ 33 Table 5-4 52.1 EFPY Cooldown Curve Data Points Using 1998 App. G Methodology, (w/Kic, w/Flange Notch, w/Uncertainties for Instrumentation Errors, and w/AP Drop........

34 Table A-1 K1t Values for 100°F/hr Heatup Curve (w/o Margins for Instrument Errors).................. A-2 Table A-2 K1 t Values for 100°F/hr Cooldown Curve (w/o Margins for Instrument Errors)............ A-3 Table B-1 33 EFPY Pressure Temperature Data for a 5°F/hr Heatup Rate with NO Margins for Instrumentation Errors or Core Pressure Differential B-2 Table B-2 52.1 EFPY Pressure Temperature Data for a 5°F/hr Heatup Rate with NO Margins for Instrumentation Errors or Core Pressure Differential.....................................................

B-4 Table D-1 Cycle Average Relative Axial Power Distribution Used for Fluence Projections.......... D-8 Table D-2 Maximum Fast Neutron (E > 1.0 MeV) Projections for the Kewaunee Pressure Vessel D -9 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 ix LIST OF FIGURES Figure 5-1 Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 5°F/hr) Applicable for 33 EFPY Using 1998 App. G Methodology (w/K1 t) (with Margins for Instrumentation Errors and Inclusion of AP for Pressure Drop...................................

23 Figure 5-2 Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F/hr) Applicable for 33 EFPY Using 1998 App. G Methodology (w/Kit) (with Margins for Instrumentation Errors and Inclusion of AP for Pressure Drop..................... 24 Figure 5-3 Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr) Applicable for 33 EFPY Using 1998 App. G Methodology (w/Ki,) (with Margins for Instrumentation Errors and Inclusion of AP for Pressure Drop..................... 25 Figure 5-4 Kewaunee Power Station Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60 and 100°F/hr) Applicable for 33 EFPY Using 1998 App. G Methodology (w/K1,) (with Margins for Instrumentation Errors and Inclusion of AP for P ressu re D rop........................................................................................................................

2 6 Figure 5-5 Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 5°F/hr) Applicable for 52.1 EFPY Using 1998 App. G Methodology (w/K1,) (with Margins for Instrumentation Errors and Inclusion of AP for Pressure Drop..................... 29 Figure 5-6 Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F/hr) Applicable for 52.1 EFPY Using 1998 App. G Methodology (w/Ki)

(with Margins for Instrumentation Errors and Inclusion of AP for Pressure Drop..................... 30 Figure 5-7 Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr) Applicable for 52.1 EFPY Using 1998 App. G Methodology (w/Ki,) (with Margins for Instrumentation Errors and Inclusion of AP for Pressure Drop..................... 31 Figure 5-8 Kewaunee Power Station Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60'and 100°F/hr) Applicable for 52.1 EFPY Using 1998 App. G Methodology (w/Ki,) (with Margins for Instrumentation Errors and Inclusion of AP for P ressure D rop...........................

32 Figure B-I Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 5°F/hr) Applicable for 33 EFPY Using 1998 App. G Methodology (w/Kic) (WITHOUT Margins for Instrumentation Errors and WITHOUT Inclusion of AP for Pressure Drop).... B-3 Figure B-2 Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 5°F/hr) Applicable for 52.1 EFPY Using 1998 App. G Methodology (w/Ki,) (WITHOUT Margins for Instrumentation Errors and WITHOUT Inclusion of AP for Pressure Drop).... B-5 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 X

Figure D-1 Axial Maximum Integrated Gamma Ray Dose [rad] as a Function of Reactor Operating T im e D -3 Figure D-2 Cycle Average Relative Radial Power Distribution Used for Fluence Projections......... D-6 Figure D-3 Axially Averaged Pressure Vessel Flux per Unit Fuel Assembly Power 0 Degree Pressure Vessel Inner R adius - 1772 M W t...................................................................................

D -7 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 Xi EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure temperature (PT) limit curves for normal operation of the Kewaunee Power Station reactor vessel. The PT curves were "generated based on the latest available reactor vessel information and updated fluence values representing a 95.6% capacity factor.

The new Kewaunee heatup and cooldown pressure-temperature limit curves were generated using the limiting adjusted reference temperature (ART) values for 33 EFPY (EOL) and 52.1 EFPY (EOLR). The highest ART values for 33 EFPY were from the upper shell forging 123W250VA1 and intermediate shell forging 122X208VA1 ("axial-flaw" orientation) and weld wire heat 1P3571 ("circ-flaw" orientation).

The highest ART values for 52.1 EFPY were from the upper shell forging 123W250VA1 ("axial-flaw" orientation) and weld wire heat 1P3571 ("circ-flaw" orientation). The final PT curves were a combination of both the "axial-flaw" and "circ-flaw" limiting ART values. The "circ-flaw" methodology from ASME Code Case N-5 88 is less restrictive than the standard "axial-flaw" methodology from the 1998 ASME Code,Section XI through 2000 Addenda. But due to low ART values for the forgings, the "circ-flaw" became limiting in various portions of the PT curves. In addition to the use of Code Case N-588, the PT curves also made use of the Kic methodology detailed in ASME Code Case N-640. Both ASME Code Case N-588 and N-640 were joined together under ASME Code Case N-641 and incorporated into the,1998 ASME Code,Section XI, through 2000 Addenda.

The PT limit curves were generated for 33 and 52.1 EFPY using heatup rates of 5, 60 and I 00°F/hr, and cooldown rates of 0, 20, 40, 60 and 100°F/hr. The curves were developed with margins for instrumentation uncertainties and include a differential pressure drop (AP). The flange requirements of Appendix G also were evaluated as a part of the curves. The resulting heatup and cooldown PT limit curves can be found in Figures 5-1 through 5-8.

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1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. For the beltline and extended beltline region materials (excluding weld wire heat 1P357 1), the adjusted RTNDT of the limiting material of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTNDT, and adding a margin. The unirradiated RTNDT is designated as the higher of either the drop weight nil-ductility'transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion minus 60'F. The test specimens should be in the transverse orientation with length of the specimen running in the major working direction and the crack propagation normal to the major working direction.

RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT (IRTNDT). The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials" [Reference 1].

Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTNDT + margins for uncertainties) at the surface, 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface.

In addition to the beltline regions, materials that exceed 1 x 1017 n/cm2 (E>1.0 MeV) are subject to the guidelines provided in Appendix H of 10 CFR 50 [Reference 2]. In accordance with 10 CFR 50, Appendix H, any materials exceeding lx10 7 n/cm2 (E>I.0 MeV) should be identified and evaluated to observe the effects of including conservative estimates of transition temperature shift at the low fluences.

Reactor vessel materials not traditionally thought to be plant limiting due to low levels of neutron radiation must now be included and evaluated to determine the impact of accumulated maximum fluence through end-of-license-renewal (EOLR) corresponding to 5.37 x 10'9 n/cm2 (52.1 EFPY). WCAP-16685-P [Reference 3] contains the drawings that show the materials included in this extended beltline.

For weld wire heat 1P3571, the calculated values for ART corresponding to 33 and 52.1 EFPY were determined utilizing the NRC SE approved methodology for the Master Curve results [Reference 4] as documented in WCAP-16609-NP [Reference 5].

The heatup and cooldown curves documented in this report were generated using the most limiting ART values and the NRC approved methodology documented in WCAP-14040-NP-A, Revision 4 [Reference 6], "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves." Specifically, the "axial-flaw" and "circ-flaw" methodologies of the 1998 ASME Code,Section XI through the 2000 Addenda [Reference 7] was used, which makes use of the K1c methodology.

The purpose of this report is to present the calculations and the development of the Kewaunee Power Station heatup and cooldown curves for 33 and 52.1 EFPY. This report documents the calculated ART values and the development of the heatup and cooldown limit curves for normal operation. The pressure-WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 2

temperature (PT) curves herein were generated with instrumentation errors of 13'F and 30 psig.

Additionally, a pressure drop of 70 psig was incorporated into the curves to account for the pressure drop, across the core. The PT curves include a hydrostatic leak test limit curve from 2485 psig to 2000 psig, along with the pressure-temperature limits for the vessel flange region per the requirements of 10 CFR Part 50, Appendix G [Reference 8]. End-of-license (EOL) and EOLR fluence estimates for the KPS vessel were based on using a 95.6% capacity factor and resulted in values of 3.44 x 1019 n/cm 2 and 5.37 x 10' 9 n/cm 2, respectively (E > 1.0 MeV).

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2 FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic materials in the reactor coolant pressure boundary are determined in accordance with the NRC Standard Review Plan [Reference 9]. The beltline and extended beltline material properties of the Kewaunee Power Station reactor vessel are presented in Tables 2-1 and 2-2. The unirradiated RTNDT values for the closure head and vessel flange are documented in Table 2-3.

The Regulatory Guide 1.99, Revision 2 methodology used to develop the heatup and cooldown curves documented in this report is the same as that documented in WCAP-14040, Revision 4 [Reference 6].

The chemistry factors (CFs) were calculated using Regulatory Guide 1.99 Revision 2, Positions 1.1 and 2.1. Position 1.1 uses the Tables from the Reg. Guide along with the best estimate copper and nickel weight percents, which are presented in Tables 2-1 and 2-2. Position 2.1 uses the surveillance capsule data from all capsules withdrawn to date. Table 2-4 summarizes the Positions 1.1 and 2.1 CFs determined for the Kewaunee Power Station beltline materials. Table 2-5 summarizes the Position 1.1 CFs determined for the Kewaunee Power Station extended beltline materials.

It should be noted that in the calculations of Position 2.1 chemistry factors in Table 2-4, the ratio procedure described in Reference 1 was applied to account for chemistry differences between the vessel weld material and the surveillance weld material [Reference 10]. No temperature adjustments are required for the Kewaunee Power Station data since the surveillance capsule results correspond to the same temperature as the reactor. pressure vessel. There were no changes to CF for surveillance weld; the CF is defined in WCAP-15571, Rev. 1.

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TABLE 2-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the Kewaunee Power Station Reactor Vessel Beltline Materials In,,....

==!itiai* RTND Material Description Heat Cu t(t

/o)

Ni (wt %0) itial RTo*

Intermediate Shell Forging 122X208VA1 0.06 0.71 60 Lower Shell Forging 123X167VA1 0.06 0.75 20 Circumferential Weld 1P3571 0.287(a) 0.756(a)

-50 Notes:

(a)

WCAP-15074, Revision 1 [Reference 11].

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TABLE 2-2 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the Kewaunee Power Station Reactor Vessel Extended Beltline Materials Componenlt

-Naei 1

Hleat

'1Intal RTNDT Upper Shell Forging B6305 123W250VA1 0.12 0.71 60 21935 0.183 0.704

-56 Upper to Intermediate Shell 10-766 BOLH 0.04 1.01 10 Girth Weld IAGI 0.03 1.07 10 B6309-1 122W496VA1 0.13 0.68 0

Inlet Nozzles B6309-2 122W515VA1 0.13 0.75

-50 DOAJ 0.02 0.92 10 AOFJ 0.03 0.93 10 EOEJ 0.01 1.03 10 Inlet Nozzle to 1-769B &

Upper Shell Forging BOLH 0.04 1.01 10 Welds 1-769D CAFJ 0.03 1.01 10 EODJ 0.02 1.04 10 DBIJ 0.02 0.97 10 B6308-2 122X302VA2 0.13 0.72 0

Outlet Nozzles B6308-1 122X288VA2 0.13 0.84

-50 EODJ 0.02 1.04 10 CAFJ 0.03 1.01 10 AOFJ 0.03 0.93 10 EOEJ 0.01 1.03 10 Outlet Nozzle to 1-769A &

Upper Shell Forging BOBJ 0.02 0.91 10 Welds 1-769C LOBI 0.03 0.94 10 BOLH 0.04 1.01 10 DBIJ 0.02 0.97 10 FOIJ 0.03 0.94 10 WCAP-16643-NP June 2009 Revision 2

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TABLE 2-3 Summary of the Initial RTNDT Values for the Kewaunee Power Station Closure Head and Vessel Flanges Material Identification fIiiiti-l RT'ONDT(:FI Closure Head Flange

-60(a)

Vessel Flange 60(b)

Notes:

(a)

Certified Material Test Report Initial RTNDT values documented in Reference 12 for the replacement reactor vessel closure head.

(b)

From WCAP-14278, Revision 1 [Reference 13].

TABLE 2-4 Summary of the Kewaunee Power Station Reactor Vessel Beltline Material Chemistry Factors Material lieat Clieris Itry Factor (F

____________~

Posijtion11 oiio Intermediate Shell Forging 122X208VA1 37 34.50 Lower Shell Forging 123X167VA1 37 28.81 Circumferential Weld

1P3571 214 219.17

Actual CF used for calculations of ART is documented in WCAP-15074, Revision 1.

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Table 2-5 Summary of the Kewaunee Power Station Reactor Vessel Extended Beltline Material Chemistry Factors Compnp6eint W

MateriaDl IDHe.at Cu

.Ni.alcul-ted Chemistry (wt %)

(wt %V)

Factor ('F)~

Upper Shell Forging B6305 123W250VA1 0.12 0.71 84.65 21935 0.183 0.704 172.22 Upper to Intermediate Shell 10-766 BOLH 0.04 1.01 54 Girth Weld IAGI 0.03 1.07 41 B6309-1 122W496VA1 0.13 0.68 93 Inlet Nozzles B6309-2 122W515VA1 0.13 0.75 94.75 DOAJ 0.02 0.92 27 AOFJ 0.03 0.93 41 EOEJ 0.01 1.03 20 Inlet Nozzle to 1-769B &

Upper Shell Forging BOLH 0.04 1.01 54 Welds 1-769D CAFJ 0.03 1.01 41 EODJ 0.02 1.04 27 DBIJ 0.02 0.97 27 B6308-2 122X302VA2 0.13 0.72 94 Outlet Nozzles B6308-1 122X288VA2 0.13 0.84 96 EODJ 0.02 1.04 27 CAFJ 0.03 1.01 41 AOFJ 0.03 0.93 41 EOEJ 0.01 1.03 20 Outlet Nozzle to 1-769A &

Upper Shell Forging BOBJ 0.02 0.91 27 Welds 1-769C LOBI 0.03 0.94 41 BOLH 0.04 1.01 54 DBIJ 0.02 0.97 27 FOIJ 0.03 0.94 41 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 8

3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1 OVERALL APPROACH The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1j, for the metal temperature at that time. K1, is obtained from the reference fracture toughness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code [Reference 7].

The K1c curve is given by the following equation:

K1, = 33.2 + 20,734 *e[°° 0 2 (T-RTNDT)]

(I)

where, Kj,

=

reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This K1c curve is based on the lower bound of static critical K, values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, and SA-508-3 steelsand welds.

3.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C* Kim + Kit < Kic (2)

where, Kim

=

stress intensity factor caused by membrane (pressure) stress Kit

=

stress intensity factor caused by the thermal gradients Ki

=

function of temperature relative to the RTNDT of the material C

=

2.0 for Level A and Level B service limits C

=

1.5 for hydrostatic and leak test conditions during which the reactor core is not critical WCAP-16643-NP June 2009 Revision 2

W6stinghouse Non-Proprietary Class 3 9

For membrane tension, the corresponding K1 for the postulated defect is:

Kim =

m. x (pRi/ t)

(3) where, Mm for an inside surface flaw is given by:

Mm

=

1.85 for ft < 2, Mm

=

0.926,I7 for 2* it *<3.464, Mm

=

3.21 for t > 3.464 Similarly, M, for an outside surface flaw is given by:

Mm

=

1.77 for It <2, Mm

=

0.893 r1 for 2*' ft - -<3.464, Mm

=

3.09 for rt > 3.464 and p = internal pressure, Ri = vessel inner radius, and t = vessel wall thickness.

For bending stress, the corresponding K, for the postulated defect is:

KIb

ý Mb

Maximum Stress, where Mb is two-thirds of Mm The maximum K1 produced by radial.thermal gradient for the postulated inside surface defect of Paragraph G-2120 is KI, = 0.953x10-3 x CR x t 2 5, where CR is the cooldown rate in 'F/hr. For a postulated outside surface defect, K1t = 0.753x10-3 x HU x t25, where HU is the heatup rate in 'F/hr.

The through-wall temperature difference associated with the maximum thermal K, can be determined from Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from Fig. G-2214-2 for the maximum thermal KI.

(a)

The maximum thermal K1 relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2).

(b)

Alternatively, the K, for radial thermal gradient can be calculated for a thermal stress distribution for any specified time during cooldown for a 1/4-thickness inside surfacedefect using the relationship:

Kit = (1.0359Co + 0.6322Ci + 0.4753C2 + 0.3855C3) * '

(4)

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 10 or similarly, KIT can be calculated during heatup for a 'A-thickness outside surface defect using the relationship:

Ki, = (1.043Co + 0.630C i + 0.481C 2 + 0.401C3) *

(5) where the coefficients Co, C1, C2 and C3 are determined from the thermal stress distribution at any.

specified time during the heatup or cooldown using the form:

oT(x) = Co + Ci(x / a) + C2(x / a) 2 + C3(x / a)3 (6) and x is a variable that represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.

Note, that Equations 3, 4 and 5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (PT) limit curves. No other changes were made to the OPERLIM computer code with regard to PT calculation methodology. Therefore, the PT curve methodology is unchanged from that described in WCAP-14040-NP-A, Revision 4 "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"

[Reference 6] Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.

At any time during the heatup or cooldown transient, KIc is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kit, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel Wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations.

From these relations, limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of KI, at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KIc exceeds Kit, the calculated allowable pressure during cooldown will be greater than the steady-state value.

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 I1I The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K1. for the 1/4T crack during heatup is lower than the KIc for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KI, values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

3.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS

.10 CFR Part 50, Appendix G [Reference 8] addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at,least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psig for Kewaunee), which is calculated to be 621 psig. The limiting unirradiated RTNDT of 60'F occurs in the vessel flange, so the minimum allowable temperature of this region is 180'F at pressures greater than 621 psig (without instrument uncertainties).

With instrument uncertainties of 13'F and 30 psig, and a differential pressure drop of 70 psig, the minimum allowable temperature of this region is 193°F at a pressure of 521 psig. This limit is shown in Figures 5-1 through 5-8 wherever applicable.

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 12 4

CALCULATION OF ADJUSTED REFERENCE TEMPERATURE 4.1 REGULATORY GUIDE 1.99, REVISION 2 APPROACH FOR CALCULATING ART From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART = Initial RTNDT + ARTNDT + Margin (7)

Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code. If measured Values of initial RTNDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

ARTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

ARTND_ CF

f(0.28- 0.0 log f)

(8)

To calculate ARTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence (based upon generic dpa trends) at the specific depth.

f(depth x) = fsurface

e (-0.24x)

(9) where x inches is the depth into the vessel wall measured from the vessel clad/base metal interface. The vessel beltline thickness is 6.5 inches. The resultant fluence is then used in Equation 8 to calculate the ARTNDT at the specific depth.

The calculated fast neutron fluence (E > 1.0 MeV) values at the inner surface of the KPS reactor vessel for 33 and 52.1.EFPY are shown in bold in Tables 4-1 and 4-2 for the beltline and extended beltline materials, respectively. These values were projected using ENDF/B-VI cross sections and are based on the results of the Capsule T radiation analysis and comply with Reg. Guide 1.190 [Reference 14]. The core power distribution chosen for future fluence projections was based on the continued use of low leakage fuel management including the consideration of potential variations in cycle to cycle design.

Note that the details of the fluence projections are discussed in WCAP-16641-NP [Reference 10]

(Capsule T analysis report).

These fluence data tabulations include fuel cycle specific calculated neutron exposures at the end of the twenty sixth fuel cycle (the last completed at KPS) as well as future projections to the end of Cycle 27 (the current operating cycle at the time the fluence analyses were performed) and for 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.

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 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 13 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 EOLR.

Margin is calculated as, M = 2 VC

+ CF2 The standard deviation for the initial RTNDT term (ai) is 0°F when the initial RTNDT is a measured value and 17°F when a generic value is available. The standard deviation for the ARTNDT term, CTa, is 17F for plates or forgings, and 8.5°F for plates or forgings when credible surveillance data is used. For welds, TA is equal to 28°F when surveillance capsule data is not used, and is 14'F (half the value) when credible surveillance capsule data is used. cTa does not have to exceed 0.5 times the mean value of ARTNDT.

Contained in Tables 4-3 through 4-8 are the Kewaunee 33 and 52.1 EFPY ART calculations used for generation of the heatup and cooldown curves.

4.2 MASTER CURVE APPROACH FOR CALCULATING ART FOR WELD HEAT 1P3571 The 1/4T and 3/4T ART values have previously been determined based on the Master Curve methodology presented in.WCAP-16609-NP for weld wire heat 1P3571. The results from WCAP-16609-NP will be used for the weld. This approach was previously approved by the NRC in Safety Evaluation by the Office of Nuclear Reactor Include the Use of a Master Curve-based Methodology for Reactor Pressure Vessel Integrity Assessment, Docket No. 50-305, May 2001 [Reference 4]. Tables 4-9 and 4-10 summarize the 1/4T and 3/4T ART values for weld wire heat 1P3571.

4.3 LIMITING ART VALUES FOR KPS PT LIMIT CURVES FOR NORMAL OPERATION Contained in Tables 4-11 and 4-12 is a summary of the limiting ART values used in the generation of the Kewaunee Power Station reactor vessel PT limit curves. The limiting materials for the "axial-flaw" methodology are Upper Shell Forging 125W250VA1 and Intermediate Shell Forging 122X208VA1. The limiting material for the "circ-flaw" methodology is Circumferential Weld 1P3571.

WCAP-16643-NP June 2009 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 14 Table 4-1 Summary of the Maximum Vessel Surface, 1/4T and 3/4T Fluence Values used for the Generation of the 33 and 52.1 EFPY Heatup/Cooldown Curves for the Beltline Materials TABLE 4-2 Summary of the Vessel Surface, I/4T and 3/4T Fluence Values used for the Generation of the 52..1 EFPY Heatup/Cooldown Curves for the Extended Beldline Materials (a) 52.1 EFPY Neutron Fluence (E> 1.0 MeV)

SMaterial~

[n/m2 jSurface IAT14j 3~/4T~

Intermediate Shell to Upper Shell Weld 5.33 x 101 3.609 x 10f8 1.654 x 1018 Upper Shell 5.33 x 10" 3.609 x 10" 1.654 x 10" RCS Inlet Nozzle to Upper Shell Weld 1.34 x 1017 0.907 x 10"7 0.416 x 1017 RCS Inlet Nozzle 1.20 x 1017 0.812 X.10 17 0.372 x 1017 RCS Outlet Nozzle to Upper Shell Weld 1.10 x 1017 0.745 x 1017 0.341 x 1017 Note:

(a) The outlet nozzles did not exceed the I x 1017 n/cm2 threshold (E > 1.0 MeV) at the clad/base metal-interface for 52.1 EFPY.

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 15 Westinghouse Non-Proprietary Class 3 15 Table 4-3 Adjusted Reference Temperature Evaluation for 33 EFPY ART Calculations for the Beltline Materials at the 1/4T Location l

CF 1/4T Fluence:

I T

/4 RTI I

<a1 )

aA MI 1/4TART<

MtraHet Pos (OF)

(x 109 /cm 2) 4F (OF)i (O

0F)

-(OF)(OF)

(OF)

Upper Shell Forging(a) 123W250VA1 1.1 84.65 0.2316 0.60502 51.22 60 0

17 34 145 1.1 37 2.329 1.2284 45.45 60 0

17 34 139 Intermediate Shell Forging 122X208VA 1 2.1 34.5 2.329 1.2284 42.38 60 0

17 34 136 1.1 37 2.329 1.2284 45.45 20 0

17 34 99 Lower Shell Forging 123X167VA1 2.1 28.81 2.329 1.2284 35.39 20 0

17 34 89 Note:

(a) The upper shelf forging was included here since the 33 EFPY fluence exceeded the I x 1017 n/cm 2 threshold (E > 1.0 MeV).

Table 4-4 Adjusted Reference Temperature Evaluation for 52.1 EFPY ART Calculations for the Beltline Materials at the 1/4T Location WCAP-1 6643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 16 Table 4-5 Adjusted Reference Temperature Evaluation for 33 EFPY ART Calculations for the Beltline Materials at the 3/4T Location Material Heat RG~ fCF 34TFuec 3/4T ARTNDT I

Y TA M

3/4T ART Pos

(

xOF)

~

m2~

F

'(OF) K(OF)

(F)

(OF)~ (OF)

(0 F)

Upper Shell Forging (a) 123W250VA1 1.1 84.65 0.1061 0.4288 36.30 60 0

17 34 130 1.1 37 1.068 1.0184 37.68 60 0

17 34 132 Intermediate Shell Forging 122X208VA1 2.1 34.5 1.068 1.0184 35.13 60 0

17 34 129 1.1 37 1.068 1.0184 37.68 20 0

17 34 92 Lower Shell Forging 123X167VA1 2.1 28.81 1.068 1.0184 29.34 20 0

17 34 83 Note:

(a) The upper shelf forging was included here as the 33 EFPY fluence for this material exceededthe I x 1017 n/cm 2 threshold (E > 1.0 MeV).

Table 4-6 Adjusted Reference Temperature Evaluation for 52.1 EFPY ART Calculations for the Beltline Materials at the 3/4T Location Ha' RG CF 3/4T' Fluence

~3/41' ART,,,

I al A

N1

3/4T ART,

-Mtra et Pos (OF)

(X1O' 9 n/C1 2Y,

~FF (O)

(0 F OF)((0 F

)

OU (0 F) 1.1 37 1.667 1.1408 42.21 60 0

17 34 136 Intermediate Shell Forging 122X208VA 1 2.1 34.5 1.667 1.1408 39.36 60 0

17 34 133 1.1 37 1.667 1.1408 42.21 20 0

17 34 96 Lower Shell Forging 123X 167VA1 2.1 28.81 1.8 21 2.1 28.81 1.667 1.1408 32.87 20 0

17 34 87 WCAP-1 6643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 17 Table 4-7 Adjusted Reference Temperature Evaluation for 52.1 EFPY ART Calculations for the Extended Beltline Materials at the 1/4T Location F

/TFluence I.AT~

MR a)A/ýNI 14TýART~

aterial I Heat CF (OF) ;jI12 9 oV Upper Shell Forging 123W250VA1 84.65 0.369 0.7245 61.33 60.0 0.0 17.00 34.0 155 UpperShelto 21935 172.22 0.369 0.7245 124.77

-56.0 17.0 28.00 65.5 134 Intermediate Shell Girth BOLH 54 0.369 0.7245 39.12 10.0 17.0 19.56 51.8 101 Weld IAGI 41 0.369 0.7245 29.70 10.0 17.0 14.85 45.1 85 122W496VA1 93 0.0812 0.3765 35.01 0.0 0.0 17.51 35.0 70 Inlet Nozzles 122W515VA1 94.75 0.0812 0.3765 35.67

-50.0 0.0 17.84 35.7 21 DOAJ 27 0.0907 0.3976 10.74 10.0 17.0 5.37 35.7 56 AOFJ 41 0.0907 0.3976 16.30 10.0 17.0 8.15 37.7 64 EOEJ 20 0.0907 0.3976 7.95 10.0 17.0 3.98 34.9 53 inlet Nozzle to Upper Shell BOLH 54 0.0907 0.3976 21.47 10.0 17.0 10-74 40.2 72 Forging Welds CAFJ 41 0.0907 0.3976 16.30 10.0 17.0 8.15 37.7 64 EODJ 27 0.0907 0.3976 10.74 10.0 17.0 5.37 35.7 56 DBIJ 27 0.0907 0.3976 10.74 10.0 17.0 5.37 35.7 56 EODJ 27 0.0745 0.3606 9.74 10.0 17.0 4.87 35.4 55 CAFJ 41 0.0745 0.3606 14.78 10.0 17.0 7.39 37.1 62 AOFJ 41 0.0745 0.3606 14.78 10.0 17.0 7.39 37.1 62 EOEJ 20 0.0745 0.3606 7.21 10.0 17.0 3.61 34.8 52 Outlet Nozzle to Upper BOBJ 27 0.0745 0.3606 9.74 10.0 17.0-4.87 35.4 55 Shell Forging Welds LOBI 41-0.0745 0.3606 14.78 10.0 17.0 7.39 37.1 62 BOLH 54 0.0745 0.3606 19.47 10.0 17.0 9.74 39.2 69 DBIJ 27 0.0745 0.3606 9.74 10.0 17.0 4.87 35.4 55 FOIJ 41 0.0745 0.3606 14.78 10.0 17.0 7.39 37.1 62 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 is Table 4-8 Adjusted Reference Temperature Evaluation for 52.1 EFPY ART Calculations for the Extended Beltline Materials at the 3/4T Location iMaterial Heat CF (OF) 34T Fluence ~3/41-ART,,,

I (F CF MAl 3/4TART 7kx1O'9 n/cm 2)

FF (OF)

(OF (F)<

(0[)~

(OF)

Upper Shell Forging 123W250VA1 84.65 0.1654 0.5250 44.44 60.0 0.0 17.00 34.0 138 21935 172.22 0.1654 0.5250 90.41

-56.0 17.0 28.00 65.5 100 Upper Shell to Intermediate Shell Girth BOLH 54 0.1654 0.5250 28.35 10.0 17.0 14.17 44.3 83 Weld IAGI 41 0.1654 0.5250 21.52 10.0 17.0 10.76 40.2 72 122W496VA1 93 0.0372 0.2486 23.12 0.0 0.0 11.56 23.1 46 Inlet Nozzles 122W515VA1 94.75 0.0372 0.2486 23.55

-50.0 0.0 11.78 23.6

-3 DOAJ 27 0.0416 0.2646 7.15 10.0 17.0 3.57 34.7 52 AOFJ 41 0.0416 0.2646 10.85 10.0 17.0 5.43 35.7 57 EOEJ 20 0.0416 0.2646 5.29 10.0 17.0 2.65 34.4 50 Inlet Nozzle to Upper Shell Ing We BOLH 54 0.0416 0.2646 14.29 10.0 17.0 7.15 36.9 61 Forging WeldsI CAFJ 41 0.0416 0.2646 10.85 10.0 17.0 5.43 35.7 57 EODJ 27 0.0416 0.2646 7.15 10.0 17.0 3.57 34.7 52 DBIJ 27 0.0416 0.2646 7.15 10.0 17.0 3.57 34.7 52 EODJ 27 0.0341 0.2365 6.39 10.0 17.0 3.19 34.6 51 CAFJ 41 0.0341 0.2365 9.70 10.0 17.0 4.85 35.4 55 AOFJ 41 0.0341 0.2365 9.70 10.0 17.0 4.85 35.4 55 EOEJ 20 0.0341 0.2365 4.73 10.0 17.0 2.37 34.3 49 Outlet Nozzle to Upper BOBJ 27 0.0341 0.2365 6.39 10.0 17.0 3.19 34.6 51 Shell Forging Welds LOBI 41 0.0341 0.2365 9.70 10.0 17.0 4.85 35.4 55 BOLH 54 0.0341 0.2365 12.77 10.0 17.0 6.39 36.3 59 DBIJ 27 0.0341 0.2365 6.39 10.0 17.0 3.19 34.6 51 FOIJ 41 0.0341 0.2365 9.70 10.0 17.0 4.85 35.4 55 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 19 Table 4-9 1P3571 ART Values to be Used for Heat-up and Cool-down Curves at the 95.6% Capacity Factor EOL RPV Maximum Fluence LoainFluence

~

Cale #1 Calc #2.

Caleic 1

hii9(XiO n/cmW)

(OF)

F2:.

(FF)

F)

Inside Surface 3.44 1.3227 261.7 276.8 285.2 274.6 1/4-T 2.33 1.2284 239.2 253.8 261.5 251.5 3/4-T 1.07 1.0183 189.1 202.3 208.8 200.1 Table 4-10 1P3571 ART Values to be Used for Heat-up and Cool-down Curves at the 95.6% Capacity Factor EOLR RPV Maximum Fluence rluenfceC

1C

2 ale NRC SE Location (at location)

F F Cc#1 ac Cac Average (X,109 In/CM 2 )

(OF)

(,A)~K

2a(O)

(F Inside Surface 5.37 1.4161 284.0 299.7 308.7 297.5 1/4-T 3.64 1.3352 264.7 279.9 288.4 277.7 3/4-T 1.67 1.1408 218.3 232.3 239.5 230.0 Table 4-11 Summary of the Limiting 1/4T ART Values Used in the Generation of the Kewaunee Power Station Heatup/Cooldown Limit Curves 1/4T Limiting "A'ial-Flaw" L

I4T Limiting 'Circ-Flaw,' ART LiiigMaterial~

(F) 0F) 33EIPY (EOL)

Upper Shell Forging 145 123W250VA1 Weld 1P3571 252 52.1 EFPY (EOLR)

~

Upper Shell Forging 155 123W250VA1 Weld 1P3571 278 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 20 Table 4-12 Summary of the Limiting 3/4T ART Values Used in the Generation of the Kewaunee Power Station Heatup/Cooldown Limit Curves Limiting Mateial 3AT1Limiting "Axial-Flaw" ART L/iftn 3/4CrcFlw A RT S33 EFPY (EOL)

Intermediate Shell Forging 122X208VA1 Weld 1P3571 200 52.1 EFPY (EOLR)

Upper Shell Forging 138 123W250VA1 Weld 1P3571 230 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 21 5

HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Sections 3 and 4 of this report. This approved methodology is also presented in WCAP-14040-NP-A, Revision 4.

Figures 5-1 through 5-3 present the limiting heatup curves with margins for possible instrumentation errors and AP drop using heatup rates of 5, 60 and 100°F/hr applicable for 33 EFPY with the "Flange-Notch" requirement using the "axial-flaw" and "circ-flaw" methodologies. These curves were generated using the 1998 ASME Code Section XI, Appendix G. Figure 5-4 presents the limiting cooldown curve with margins for possible instrumentation errors and AP drop using cooldown rates of 0, 20, 40, 60 and 100°F/hr applicable for 33 EFPY with the "Flange-Notch" requirement using the "axial-flaw" and "circ-flaw" methodologies. Again, this curve was generated using the 1998 ASME Code Section X1, Appendix G.

Figures 5-5 through 5-7 present the limiting heatup curves with margins for possible instrumentation errors and AP drop using heatup rates of 5, 60 and 100°F/hr applicable for 52.1 EFPY with the "Flange-Notch" requirement using the "axial-flaw" and "circ-flaw" methodologies. These curves were generated using the 1998 ASME Code Section XI, Appendix G. Figure 5-8 presents the limiting cooldown curve with margins for possible instrumentation errors and AP drop using cooldown rates of 0, 20, 40, 60 and 100lF/hr applicable for 52.1 EFPY with the "Flange-Notch" requirement using the "axial-flaw" and "circ-flaw" methodologies. Again, this curve was generated using the 1998 ASME Code Section X1, Appendix G.

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 5-1 through 5-8. This is in addition to other criteria, which must be met before the reactor is made critical, as discussed below in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 5-1 and 5-7 (heatup curves only). The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The goveming equation for the hydrostatic test is defined in the 1998 ASME Code Section X1, Appendix G as follows:

1.5 Kim < Kic

where, Kim is the stress intensity factor covered by membrane (pressure) stress, Ki, = 33.2 + 20.734 e[°0 2 (T-RTNDT)],

T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature.

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 22 The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in 10 CFR 50 Appendix G [Reference 8]. The pressure-temperature limits for core operation (except for low power physics tests) are that: 1) the reactor vessel must be at a temperature equal to or higher than-the minimum temperature required for the inservice hydrostatic test, and 2) the reactor vessel must be at least 40'F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 4 of this report. For the heatup and cooldown curves with margins for instrumentation errors and AP drop, the minimum temperatures for the in service hydrostatic leak tests for the Kewaunee Power Station reactor vessel at 33 EFPY is 229°F, and at 52.1 EFPY is 255°F. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40'F higher than the pressure-temperature limit curve constitutes the limit for core operation for the reactor vessel.

Figures 5-1 through 5-8 define all of the above limits fdr ensuring prevention of nori-ductile failure for the Kewaunee Power Station reactor vessel for 33 and 52.1 EFPY with the "Flange-Notch" requirement, with instrumentation uncertainties, and with AP drop. The data points used for developing the heatup and cooldown pressure-temperature limit curves shown in Figures 5-1 through 5-8 are presented in Tables 5-1 through 5-4.

The composite curves identified in Figures 5-1 through 5-8 are primarily controlled by the "circ-flaw" analyses. This means that the controlling material for the development of the limiting P-T curves is primarily the circumferential weld metal 1P3571. Tables 5-1 through 5-4 highlight data where the "axial-flaw" analyses become limiting.

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 23 Westinghouse Non-Proprietary Class 3 23 MATERIAL PROPERTY BASIS LIMITING MATERIAL: "axial-flaw" orientation - Upper Shell Forging (1/4T), Intermediate Shell Forging (3/4T)

"circ-flaw" orientation - Circumferential Weld 1 P3571 LIMITING ART VALUES AT 33 EFPY:

1/4T, 145°F (axial-flaw), 252°F (circ-flaw) 3/4T, 132°F (axial-flaw), 200'F (circ-flaw) 2500 2250 2000 1750 Fn 1500 T 1250

.2 1000 750 500 250 0 1. 1 -...4i4..

i-i--i--

1 n-1....

111 -

i.1 -

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 5-1 Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 51F/hr) Applicable for 33 EFPY Using 1998 App. G Methodology (w/K1,) (with Margins for Instrumentation Errors and Inclusion of AP for Pressure Drop)

WCAP-16643-NP June 2009 Revision 2

'Westinghouse Non-Proprietary Class 3 24 MATERIAL PROPERTY BASIS LIMITING MATERIAL: "axial-flaw" orientation - Upper Shell Forging (1/4T), Intermediate Shell Forging (3/4T)

"circ-flaw" orientation - Circumferential Weld 1 P3571 LIMITING ART VALUES AT 33 EFPY:

1/4T, 145°F (axial-flaw), 252°F (circ-flaw) 3/4T, 132°F (axial-flaw), 200'F (circ-flaw)

Version:5.2 Run:6072 Operlim.xls Version: 5.2 1 2500 2250 2000 1750 F 1500 U)

O 1250 0I

.2 1000 C.)

750 500 250 0

Acceptable Operation Criticality Limit based on inservice hydrostatic test temperature (229 F) for the service period up to 33 EFPY 0

50 100 150 200 250 300 350 40 Moderator Temperature (Deg. F) 0 450 500 550 Figure 5-2 Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 60'F/hr) Applicable for 33 EFPY Using 1998 App. G Methodology (w/Ki,) (with Margins for Instrumentation Errors and Inclusion of AP for Pressure Drop)

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 25 Westinghouse Non-Proprietary Class 3 25 MATERIAL PROPERTY BASIS LIMITING MATERIAL: "axial-flaw" orientation - Upper Shell Forging (1/4T), Intermediate Shell Forging (3/4T)

"circ-flaw" orientation - Circumferential Weld 1P3571 LIMITING ART VALUES AT 33 EFPY:

1/4T, 145°F (axial-flaw), 252°F (circ-flaw) 3/4T, 1327F (axial-flaw), 200'F (circ-flaw) 2500 2250 2000 1750 m1500 21250 0.

.21000 L) 750 500 250 0

0 50 100 150 200 250 300 350 400 Moderator Temperature (Deg. F) 450 500 550 Figure 5-3 Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 1001F/hr) Applicable for 33 EFPY Using 1998 App. G Methodology (w/K1 c) (with Margins for Instrumentation Errors and Inclusion of AP for Pressure Drop)

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 26 MATERIAL PROPERTY BASIS LIMITING MATERIAL: "axial-flaw" orientation - Upper Shell Forging (1/4T), Intermediate Shell Forging (3/4T)

"circ-flaw" orientation - Circumferential Weld 1P3571 LIMITING ART VALUES AT 33 EFPY:

1/4T, 145°F (axial-flaw), 252°F (circ-flaw) 3/4T, 1327F (axial-flaw), 2007F (circ-flaw)

FOperlim Version:5.2 Run:6072 Operlim.xls Version: 5.2 2500 2250 2000 1750 n 1500 U) 1250 1.2

.2 1000 M

750 500 250 0

0 50 100 150 200 250 300 350 400 450 Moderator Temperature (Deg. F) 500 550 Figure 5-4 Kewaunee Power Station Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60 and 1000F/hr) Applicable for 33 EFPY Using 1,998 App. G Methodology (w/K,,) (with Margins for Instrumentation Errors and Inclusion of AP for Pressure Drop)

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 27 TABLE 5-1 33 EFPY Heatup Curve Data Points Using 1998 App. G Methodology (w/KIc, w/Flange Notch, w/Uncertainties for Instrumentation Errors, and w/AP Drop)

S5'F/hr CriticalItIyh Limit 6 0'Ffifr Crtclt Limi 100'F/ibi>

criticaliity Limit T

P T

P p T

P T

P TJT.

F)__________________________

(F)

Psg psig)

F

( P~sig)

(0 F)

(psig)

('F (psig) 73 0

229 0

73 0

229 0

73 0

229 0

73 521 229 521 73 521 229 521 73 521 229 521 78 521 229 521 78 521 229 521 78 521 229 521 83 521 229 521 83 521 229 521 83 521 229 521 88 521 229 521 88 521 229 521 88 521 229 521 93 521 229 521 93 521 229 521 93 521 229 521 98 521 229 521 98 521 229 521 98 521 229 521 103 521 229 521 103 521 229 521 103 521 229 521 108 521 229 521 108 521 229 521 108 521 229 521 113 521 229 521 113 521 229 521 113 521 229 521 118 521 229 521 118 521 229 521 118 521 229 521 123 521 229 521 123 521 229 521 123 521 229 521 128 521 229 521 128 521 229 521 128 521 229 521 133 521 229 521 133 521 229 521 133 521 229 521 138 521 229 521 138 521 229 521 138 521 229 521 143 521 229 521 143 521 229 521 143 521 229

521, 148 521 229 521 148 521 229 521 148 521 229 521 153 521 229 521 153 521 229 521 153 521 229 521 158 521 229 521 158 521 229 521 158 521 229 521 163 521 229 521 163 521 229 521 163 521 229 521 168 521 229 521 168 521 229 521 168 521 229 521 173 521 229 521 173 521 229 521 173 521 229 521 178 521 229 521 178 521 229 521 178 521 229 521 183 521 229 521 183 521 229 521 183 521 229 521 188 521 233 521 188 521 233 521 188 521 233 521 193 521 233 1455 193 521 233 1363 193 521 233 1150 193 1455 238 1546 193 1363 238 1448 193 1150 238 1216 198 1546 243 1605 198 1448 243 1541 198 1216 243 1290 203 1605 248 1632 203 1541 248 1632 203 1290 248 1371 208 1632 253 1663 208 1632 253 1663 208 1371 253 1461 213 1663 258 1696 213 1663 258 1696 213 1461 258 1560 218 1696 263 1733 218 1696 263 1733 218 1560 263 1669 223 1733 268 1774 223 1733 268 1774 223 1669 268 1762 228 1774 273 1819 228 1774 273 1819 228 1762 273 1819 233 1819 278 1869 233 1819 278 1869 233 1819 278 1869 238 1869 283 1925 238 1869 283 1925 238 1869 283 1925 243 1925 288 1986 243 1925 288 1986 243 1925 288 1986 248 1986 293 2053 248 1986 293 2053 248 1986 293 2053 253 2053 298 2128 253 2053 298 2128 253 2053 298 2128 258 2128 303 2210 258 2128 303 2210 258 2128 303 2210 263 2210 308 2301 263 2210 308 2301 263 2210 308 2301 268 2301 313 2402 268 2301 313 2402 268 2301 313 2402 273 2402 317 2485 273 2402 317 2485 273 2402 317 2485 276.7 2485 276.7 2485 276.7 2485 Leak Test Limit T

143 229 P (psig) 2000 2485

Italicized and bolded numbers indicate the limiting values based on the "axial-flaw" orientation.

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 28 TABLE 5-2 33 EFPY Cooldown Curve Data Points Using 1998 App. G Methodology (w/Kic, w/Flange Notch, w/Uncertainties for Instrumentation Errorsj and w/AP Drop)

Steady State ~.

20?F/hr 4' 40/hr

'6 0?F/hr

~

2 OTh

'iT, P~

P,

'r P/1 1~~

11 ~

T P

73 0

73 521 73 521 73 521 73 521 73 521 78 521 78 521 78 521 78 521 78 521 83 521 83 521 83 521 83 521 83 521 88 521 88 521 88 521 88 521 88 521 93 521 93 521 93 521 93 521 93 521 98 521 98 521 98 521 98 521 98 521 103 521 103 521 103 521 103 521 103 521 108 521 108 521 108 521 108 521 108 521 113 521 113 521 113 521 113 521 113 521 118 521 118 521 118 521 118 521 118 521 123 521 123 521 123 521 123 521 123 521 128 521 128 521 128 521 128 521 128 521 133 521 133 521 133 521 133 521 133 521 138 521 138 521 138 521 138 521 138 521 143 521 143 521 143 521 143 521 143 521 148 521 148 521 148 521 148 521 148 521 153 521 153 521 153 521 153 521 153 521 158 521 158 521 158 521 158 521 158 521 163 521 163 521 163 521 163 521 163 521 168 521 168 521 168 521 168 521 168 521 173 521 173 521 173 521 173 521 173 521 178 521 178 521 178 521 178 521 178 521 183 521 183 521 183 521 183 521 183 521 188 521 188 521 188 521 188 521 188 521 193 521 193 521 193 521 193 521 193 521 193 521 193 521 193 521 193 521 193 521 193 1460 193 1460 193 1448 193 1393 193 1284 198 1551 198 1526 198 1473 198 1419 198 1313 203 1605 203 1553 203 1500 203 1448 203 1345 208 1632 208 1581 208 1531 208 1480 208 1381.

213 1663 213 1613 213 1564 213 1516 213 1421 218 1696 218' 1649 218 1602 218 1555 218 1465 223 1733 223 1688 223 1643 223 1599 223 1514 228 1774 228 1731 228 1689 228 1648 228 1569 233 1819 233 1779 233 1739 233 1701 233 1629 238 1869 238 1832 238 1795 238 1761 238 1696 243 1925 243 1890 243 1857 243 1826 243 1770 248 1986 248 1955 248 1926 248 1899 248 1853 253 2053 253 2027 253 2002 253 1980 253 1944 258 2128 258 2106 258 2086 258 2069 258 2045 263 2210 263 2193 263 2179 263 2168 263 2157 268 2301 268 2290 268 2282 268 2277 268 2277 273 2402 273 2397 273 2395 273 2395 273 2395 276.7 2485 276.7 2485 276.7 2485 276.7 2485 276.7 2485

Italicized and bolded numbers indicate the limiting values based on the "axial-flaw" orientation.

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 29 MATERIAL PROPERTY BASIS LIMITING MATERIAL:

"axial-flaw" orientation - Upper Shell Forging 123W250VAI "circ-flaw" orientation - Circumferential Weld 1P3571 LIMITING ART VALUES AT 52.1 EFPY:

1/4T, 1557F (axial-flaw), 2787F (circ-flaw) 3/4T, 138°F (axial-flaw), 230°F (circ-flaw) 2Operlim Version:5.2 Run:24339 Operlim.xls Version: 5.2 2500 225 it 2*1000....................................................

1750

.... Heatup Rate..

5 Deg. FIHr Critia Lmit I5 De. F/Hr 0

_ 15000 U)

U) 21250

-(Dpabe Acceptable OpeationOperation

.2 1000 750 Criticality Limit based on 500 inservice hydrostatic test t

temperature (255 F) for the service period up to 52 EFPY Bolt Up 250 t Temperature 0

0 50 100 150 200 250 300 350 40 Moderator Temperature (Deg. F)

'0 450 500 550 Figure 5-5

, Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 51F/hr) Applicable for 52.1 EFPY Using 1998 App. G Methodology (w/Kit) (with Margins for Instrumentation Errors and Inclusion of AP for Pressure Drop)

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 30 MATERIAL PROPERTY BASIS LIMITING MATERIAL:

"axial-flaw" orientation - Upper Shell Forging 123W250VAI "circ-flaw" orientation - Circumferential Weld 1P3571 LIMITING ART VALUES AT 52.1 EFPY:

1/4T, 155°F (axial-flaw), 278°F (circ-flaw) 3/4T, 138°F (axial-flaw), 230'F (circ-flaw) n Version:5.2 Run:24339 Operlim.xls Version: 5.2 2500 2250 2000 1750 F 1500 TP 1250

.2 1000 750 500 250 I

0 0

50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 5-6 Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 60'F/hr) Applicable for 52.1 EFPY Using 1998 App. G Methodology (w/K1,) (with Margins for Instrumentation Errors and Inclusion of AP for Pressure Drop)

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 31 MATERIAL PROPERTY BASIS LIMITING MATERIAL:

"axial-flaw" orientation - Upper Shell Forging 123W250VAI "circ-flaw" orientation - Circumferential Weld 1P3571 LIMITING ART VALUES AT 52.1 EFPY:

1/4T, 155'F (axial-flaw), 278°F (circ-flaw) 3/4T, 138°F (axial-flaw), 230'F (circ-flaw) n Version:5.2 Run:24339 Operlim.xls Version: 5.2 1 2500 2250 2000 1750 i1500 U-1250 CL 1000 750 500 2150 0

0 50 100 150 200 250 300 350 400 450 Moderator Temperature (Deg. F) 500 550 Figure 5-7 Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 100 0F/hr) Applicable for 52.1 EFPY Using 1998 App. G Methodology (w/K1 t) (with Margins for Instrumentation Errors and Inclusion of AP for Pressure Drop)

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 32 MATERIAL PROPERTY BASIS LIMITING MATERIAL:

"axial-flaw" orientation - Upper Shell Forging 123W250VA1 "circ-flaw" orientation - Circumferential Weld 1 P3571 LIMITING ART VALUES AT 52.1 EFPY:

1/4T, 155°F (axial-flaw), 278°F (circ-flaw) 3/4T, 138'F (axial-flaw), 230'F (circ-flaw) 2500

[Operlim Version:5.2 Run:24339 Operlimxls Version: 5.2]"

2250 2000

-+.

1750 I j tj Cooldown

~ 1500

~--.---~Rates F/Hr P 1 25 00 I -

tn

-100 120 Unacceptable

.2 1000 prto C-)

750 Acceptable Ap t ler o

7500-

~1 iBolt Up 250 Temperature 0

0 50 100 150 200 250 300 350 400 Moderator Temperature (Deg. F) 450 500 550 Figure 5-8 Kewaunee Power Station Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60 and 1000F/hr) Applicable for 52.1 EFPY Using 1998 App. G Methodology (w/Kit) (with Margins for Instrumentation Errors and Inclusion of AP for Pressure Drop)

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 33 TABLE 5-3 52.1 EFPY Heatup Curve Data Points Using 1998 App. G Methodology w/Kic, w/Flange Notch, w/Uncertainties for Instrumentation Errors, and w/AP Drop)

?rt Lm6 hr

1Criticality Lim, it lO h

iCriticafIity Litnit iiiC ritiCalltv Llmtnit 4 o

0 : r M

100,F/111-~tc l~y:E milii**...............

T P

.11T P

Th.

13 T

PK' T

P P

0(F)

(I

,:. (O ).(-

)

(.,

)*

.( 0T F)

(ps i(' *,;(psig)

(or) c(fsIg) i (Psig) 73 0

255 0

73 0

255 0

73 0

255 0

73 521 255 521 73 521 255 521 73 521 255 521 78 521 255 521 78 521 255 521 78 521 255 521 83 521 255 521 83

\\

521 255 521 83 521 255 521 88 521 255 521 88 521 255 521 88 521 255 521 93 521 255 521 93 521 255 521 93 521 255 521 98 521 255 521 98 521 255 521 98 521 255 521 103 521 255 521 103 521 255 521 103 521 255 521 108 521 255 521 108 521 255 521 108 521 255 521 113 521 255 521 113 521 255 521 113 521 255 521 118 521 255 521 118 521 255 521 118 521

°255 521 123 521 255 521 123 521 255 521 123 521 255 521 128 521 255 521 128 521 255 521 128 521 255 521 133 521 255 521 133 521 255 521 133 521 255 521 138 521 255 521 138 521 255 521 138 521 255 521 143 521 255 521 143 521 255 521 143 521 255 521 148 521 255 521 148 521 255 521 148 521 255 521 153 521 255 521 153 521 255 521 153 521 255 521 158 521 255 521 158 521 255 521 158 521 255 521 163 521 255 521 163 521 255 521 163 521 255 521 168 521 255 521 168 521 255 521 168 521 255 521 173 521 255 521 173 521 255 521 173

,521 255 521 178 521 255 521 178 521 255 521 178 521 255 521 183 521 255 521 183 521 255 521 183 521 255 521 188 521 255 521 188 521 255

521, 188 521 255 521 193 521 255 1300 193 521 255 1270 193 521 255 1076 193 1300 255 1374 193 1270 255 1345 193 1076 255 1135 198 1374 255 1455 198 1345 255 1419 198 1135 255 1200 203 1455 255 1515 203 1419 255 1496 203 1200 255 1272 208 1515 255 1533 208 1496 255 1533 208 1272 255 1351 213 1533 258 1553 213 1533 258 1553 213 1351 258 1395 218 1553 263 1575 218 1553 263 1575 218 1395 263 1426 223 1575 268 1600 223 1575 268 1600 223 1426 268 1459 228 1600 273' 1627 228 1600 273 1627 228 1459 273 1496 233 1627 278 1656 233 1627 278 1656 233 1496 278 1536 238 1656 283 1689 238 1656 283 1689 238 1536 283 1582 243 1689 288 1725 243 1689 288 1725 243 1582 288 1631 248 1725 293 1766 248 1725 293 1766 248 1631 293 1687 253 1766 298 1810 253 1766 298 1810 253 1687 298.

1748 258 1810 303 1859 258 1810 303 1859 258 1748 303 1815 263 1859 308 1913 263 1859 308 1913 263 1815 308 1889 268 1913 313 1973 268 1913 313 1973 268 1889 313 1971 273 1973 318 2039 273 1973 318 2039 273 1971 318 2039 278 2039 323 2112 278 2039 323 2112 278 2039 323 2112 283 2112 328 2193 283 2112 328 2193 283 2112 328.

2193 288 2193 333 2282 288 2193 333 2282 288 2193 333 2282 293 2282 338 2381 293 2282 338 2381 293 2282 338 2381 298 2381 342.8 2485 298 2381 342.8 2485' 298 2381 342.8 2485 302.8 2485 302.8 2485 302.8 2485 Leak Test Limit T ('F) 169 255 P (psig) 2000 2485_

Italicized and bolded numbers indicate the limiting values based on the "axial-flaw" orientation.

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 34 TABLE 5-4 52.1 EFPY Cooldown Curve Data Points Using 1998 App. G Methodology (w/Klc, w/Flange Notch, w/Uncertainties for Instrumentation Errors, and w/AP Drop)

'StuadyStatej 2OFhj O/j.

?~

~6O 0 F/hr I 0Fh T

P T

KP

~

F 1C T

~ 'P

(............

F)=*...

( sg :

('F)..

sg = :=: =(°F)

(psig)

F)*i:*ii!

(psg)==*

(OF) ii(sFg (psi"s)g (Psig)i~

73 0

73 521 73 521 73 521 73 521 73 521 78 521 78 521 78 521

78.

521 78 521 83 521 83 521 83 521 83 521 83 521 88 521 88 521 88 521 88 521 88 521 93 521 93 521 93 521 93 521 93 521

'-98 521 98 521 98 521 98 521 98 521 103 521 103 521 103 521 103 521 103 521 108 521 108 521 108 521 108 521 108 521 113 521 113 521 113 521 113 521 113 521 118 521 118 521 118 521 118 521 118 521 123 521 123 521 123 521 123 521 123 521 128 521 128 521 128 521 128 521 128 521 133 521 133 521 133 521 133 521 133 521 138 521 138 521 138 521 138 521 138 521 143 521 143 521 143 521 143 521 143 521 148 521 148 521 148 521 148 521 148 521 153 521 153 521 153 521 153 521' 153 521 158 521 158 521 158 521 158 521 158 521 163 521 163 521 163 521 163 521 163 521 168 521 168 521 168 521 168 521 168 521 173 521 173 521 173 521 173 521 173 521 178 521 178 521 178 521 178 521 178 521 183 521 183 521 183 521 183 521 183 521 188 521 188 521 188 521 188 521 188 521 193 521 193 521 193 521 193 521 193 521 193 1302 193 1302 193 1302 193 1290 193 1167 198 1377 198 1377 198 1365 198 1305 198 1184 203 1460 203 1440 203 1381 203 1322 203 1202 208 1515 208 1457 208 1399 208 1341 208 1223 213 1533 213 1476 213 1419 213 1362 213 1246 218 1553 218 1497 218 1441 218 1385 218 1272 223 1575 223 1520 223 1465 223 1410 223 1300 228 1600 228 1546 228 1492 228 1439 228 1332 233 1627 233 1574 233 1522 233 1470 233 1368 238 1656 238 1606 238 1555 238 1505 238 1407 243 1689 243 1640 243 1592 243 1544 243 1450 248 1725 248 1679 248 1632 248 1587 248 1499 253 1766 253 1721 253 1677 253 1634 253 1552 258 1810 258 1768 258 1727 258 1687 258 1612 263 1859 263 1820 263 1782 263 1746 263 1678 268 1913 268 1877 268 1843 268 1810 268 1751 273 1973 273 1941 273 1910 273 1882 273 1832

278, 2039 278 2011 278 1985 278 1961 278 1922 283 2112 283 2088 283 2067 283 2049 283 2022 288 2193 288 2174 288 2158 288 2146 288 2132 293 2282 293 2269 293 2259 293 2253 293 2253 298 2381 298 2374 298 2371 298 2371 298 2371 302.8 2485 302.8 2485 302.8 2485 302.8 2485 302.8 2485

Italicized and bolded numbers indicate the limiting values based on the "axial-flaw" orientation.

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 35 6

REFERENCES

1.

Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S.

Nuclear Regulatory Commission, May 1988.

2.

Code of Federal Regulations, 10 CFR Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.,

Federal Register, Volume 60, No. 243, dated December 19, 1995

3.

WCAP-16685-P, "Materials Documentation for the Kewaunee Power Station Extended Beltline,"

B. N. Burgos, April 2008.

4.

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

5.

WCAP-16609-NP, "Master Curve Assessment of Kewaunee Power Station Reactor Vessel Weld Metal," R. G. Lott, et. al., October 2006.

6.

WCAP-14040-NP-A, Revision 4, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," J.D. Andrachek, et al., May 2004.

\\

7.

Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, "Fracture Toughness Criteria for Protection Against Failure," Dated December 1998, through 2000 Addendum.

8.

Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements,"

U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

9.

"Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.

10.

WCAP-16641-NP, "Analysis of Capsule T from Dominion Energy Kewaunee Power Station Reactor Vessel Radiation Surveillance Program," B. N. Burgos, et.al., October 2006.

11.

WCAP-15074, Revision 1, "Evaluation of the 1P3571 Weld Metal from the Surveillance, Programs for Kewaunee and Main Yankee," B.N. Burgos, August 2006.

12.

Certified Material Test Report MHI-NMC-0172K for Kewaunee Power Station, "CMTR for Closure Head," dated 5/14/03.

13.

WCAP-14278, Revision 1, "Kewaunee Heatup and Cooldown Curves for Normal Operation,"

T.J. Laubham, et. al., September 1998.

14.

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.

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A Thermal Stress Intensity Factors (Kit)

The following pages contain the thermal stress intensity factors (Kit) for the maximum heatup and cooldown rates. The vessel radii to the 1/4T and 3/4T locations are as follows:

1/4T Radius = 67.781" 3/4T Radius = 71.031" WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 A-2 TABLE A-1 Kit Values for 100 0F/hr Heatup Curve (w/o Margins for Instrument Errors)

Vessel Temperature 14ThraStesVsl]inrtu 3/4 U ' Thrinal Stress Tem*np

((v, 1/41 Location for Intensitn Factor (Tne3/4 Ocatsn for 3/tensith Factor (0 F) 60 56.455

-0.952 55.144 0.524 65 59.695

-2.167 55.835 1.434 70 63.165

-3.048 57.327 2.169 75 66.942

-3.830 59.486 2.775 80 70.966

-4.423 62.167 3.261 85 75.134

-4.934 65.288 3.660 90 79.489

-5.331 68.754 3.983 95 83.928

-5.674 72.503 4.249 100 88.496

-5.943 76.474 4.467 105 93.112

-6.178 80.624 4.649 110 97.816

-6.364 84.919 4.800 115 102.549

-6.530 89.329 4.928 120 107.339

-6.662 93.834 5.035 125 112.148

-6.782 98.414 5.127 130 116.994

-6.879 103.056 5.205 135 121.853

-6.968 107.748 5.274 140 126.736

-7.042 112.480 5.333 145 131.627

-7.113 117.244 5.386 150 136.534

-7.172 122.034 5.434 155 141.447

-7.229 126.846 5.477 160 146.370

-7.278 131.675 5.516 165 151.297

-7.327 136.518 5.553 170 156.230

-7.370 141.372 5.587 175 161.167

-7.413 146.235 5.619 180 166.106

-7.452 151.106 5.650 185 171.049

-7.491 155.983 5.679 190 175.993

-7.527 160.865 5.707 195 180.939

-7.564 165.751 5.735 200 185.886

-7.599 170.639 5.762 205 190.835

-7.634 175.531 5.788 210 195.784

-7.667 180.424 5.814 215 200.735

-7.702 185.319 5.839 220 205.685

-7.734 190.216 5.865 225 210.637

-7.768 195.113 5.890 230 215.587

-7.800 200.012 5.915 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 A-3 TABLE A-2 Kit Values for 100 0F/hr Cooldown Curve (w/o Margins for Instrument Errors)

Vessel Temineratu re I 0 0'F/br Cooldown, Water~ (i, 1/~4T Location for

.1/1' Thiermal Str(,,,,

I Tllml.

lO 0'F/hr Coold]ow~n I tensity Fco 200.000 215.032 8.131 195.000 209.981 8.097 190.000 204.931 8.063 185.000 199.881 8.029 180.000 194.831 7.995 175.000 189.781 7.961 170.000 184.732 7.927 165.000 179.682 7.893 160.000 174.633 7.860 155.000 169.583 7.826 150.000 164.534 7.793 145.000 159.484 7.759 140.000 154.435 7.726 135.000 149.386 7.693 130.000 144.337 7.660 125.000 139.288 7.627 120.000 134.240 7.594 115.000 129.191 7.561 110.000 124.142 7.528 105.000 119.094

.7.495

/

100.000 114.045 7.463 95.000 108.997 7.430 90.000 103.949 7.397 85.000 98.901 7.365 80.000 93.852 7.333 75.000 88.804 7.300 70.000 83.757 7.268 65.000 78.709 7.236 60.000 73.663 7.203 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 B-1 APPENDIX B Pressure-Temperature Limit Curve for 51F/hr Heatup WITHOUT Margins The following pages contain the pressure-temperature limit curves for 33 and 52.1 EFPY representative of a 5°F/hr heatup rate with NO margins for instrumentation errors or differential pressure across the core.

WCAP-1 6643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 B-2 Table B-1 33 EFPY Pressure Temperature Data for a 5°F/hr Heatup Rate with NO Margins for Instrumentation Errors or Core Pressure Differential 52F/1{r Heatiip ST P

(F

(ý _______

60 0

60 621 65 621 70 621 75 621 80 621 85 621 90 621 95 621 100 621 105 621 110 621 115 621 120 621 125 621 130 621 135 621 140 621 145 621 150 621 155 621 160 621 50gF/hr H~atup T

165 621 170 621 175 621 180 621 180 1555 185 1646 190 1705 195 1732 200 1763 205 1796 210 1833 215 1874 220 1919 225 1969 230 2025 235 2086 240 2153 245 2228 250 2310 255 2401 260 2502 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 B-3 2500 1

2250 2000

-T I

i 1750 Unacceptable cpb Operation I ° v)1500 I

9)1250

-K Rate -V

-i Acceptable Operation i 10001-A 7 5...

i,,

500 5

2 Boltup Temperture 250

~I-1 0.....

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure B-1 Kewaunee Power Station Reactor Coolant System Heatup Limitations (Heatup Rate of 5°F/hr) Applicable for 33 EFPY Using 1998 App. G Methodology (w/K1,)

(WITHOUT Margins for Instrumentation Errors and WITHOUT Inclusion of AP for Pressure Drop)

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 B-4 Table B-2 52.1 EFPY Pressure Temperature Data for a 5°F/hr Heatup Rate with NO Margins for Instrumentation Errors or Core Pressure Differential 5,F/hr Heatupi T

P 60 0

60 621 65 621 70 621 75 621 80 621 85 621 90 621 95 621 100 621 105 621 110 621 115 621 120 621 125 621 130 621 135 621 140 621 145 621 150 621 155 621 160 621 165 621 170 621 175 621 5'F/ir Heatup, TP 180 621 180 1400 185 1474 190 1555 195 1615 200 1633 205 1653 210 1675 215

1700, 220 1727 225 1756 230 1789 235 1825 240 1866 245 1910 250 1959 255 2013 260 2073 265 2139 270 2212 275 2293 280 2382 285 2481 I

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 B-5 Westinghouse Non-Proprietary Class 3 B-5 2500 2250 1750 Unacceptable 20 0 Rat

+Unjcceptable Operationp F.500BHtup 1250 60 F ]

a3.

Acceptable 0

1T I

o10 10 100005000 30 40 45 0

5 750 0

,° _ °..9.....

E Boltup

]Tempertur~e i

2 5 0..

J 60 F0 I.......

....!...........t.

0:

0 50 100 150 200 250 300 350 400 4*50 500 550 Moderator Temperature (Deg. F)

Figure B-2 Kewaunee Power Station Reactor-Coolant System Heatup Limitations (Heatup Rate of 5°F/hr) Applicable for 52.1 EFPY Using 1998 App. G Methodology (w/K],)

(WITHOUT Margins for Instrumentation Errors and WITHOUT Inclusion of AP for Pressure Drop)

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 C-1 Westinghouse Non-Proprietary Class 3 C-I APPENDIX C Enable Temperature Calculations for 33 and 52.1 EFPY Enable temperatures for 33 and 52.1 EFPY are calculated based on the methodology presented in the 1998 version of ASME Section XI (thru/2000 Addenda), Appendix G. This corresponds to the methodology documented for the development of the pressure-temperature limit curves.

Specifically, the enable temperature is equal to the greater of 200'F or the coolant temperature corresponding to a reactor vessel metal temperature less than RTNDT + 507.

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 C-2 The limiting case for enable temperature will be the circumferential flaw analysis since the ART values are largest for this analysis. This corresponds to the highest 1/4T ART value of 252°F for 33 EFPY and 2787F for 52.1 EFPY (see Table 4-11 of the main report)..

When utilizing the 33 EFPY pressure-temperature limit curves, the associated enable temperature is calculated to be 3170 F (without temperature instrumentation error).

When utilizing the 52.1 EFPY pressure-temperature limit curves, the associated enable temperature is calculated to be 343'F (without temperature instrumentation error).

It should be noted that the flange notch is currently limiting for all pressure-temperature limit curves for temperatures < 193°F. If the flange limitation is removed, the "axial-flaw" methodology for pressure-temperature limit curves would be limiting for temperatures < 193°F in the steady state condition.

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 D-1 APPENDIX D Gamma Ray Dose to Concrete Biological Shield and Associated Neutron Fluence Projections WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 D-2 Gamma Ray Dose in Concrete Biological Shield In WCAP-16641-NP, Revision 0, "Analysis of Capsule T from the Dominion Energy Kewaunee Power Station Reactor Vessel Radiation Surveillance Program," October 2006, a discussion of the maximum gamma ray dose experienced by the concrete biological shield was provided. In that document, axial data traverses extending from the bottom to the top of the active fuel along the 0 degree azimuth at the inner radius of the biological shield were given for several reactor operating times. The 0 degree traverse was intended to establish a bounding gamma ray dose profile for all of the materials comprising the shield.

From the data provided in WCAP-16641-NP, Revision 0, it was noted that, for an axial segment of the shield extending approximately +/- 5 feet from the midplane of the active fuel, the bounding gamma ray dose at the shield wall inner radius exceeded the concrete damage threshold of 1.OE+10 rad after approximately 33 effective full power years (EFPY) of operation. In order to provide further insight into the extent to which the concrete shield would be exposed to a gamma ray dose exceeding 1.OE+10 rad, the data traverses provided in WCAP-16641-NP, Revision 0 were expanded and are shown here.

In Figure D-1, distributions of integrated gamma ray dose applicable to 36, 44, 48, 52.1, and 54 EFPY of reactor operation are shown as a function of azimuthal angle and depth into the concrete material. The data were taken at the axial location of the maximum exposure near the core midplane and can be conservatively applied over a span of +/- 5 feet relative to the midplane of the active fuel. Due to reactor symmetry, the 0o-45' octant is representative of all eight octants comprising the plant geometry.

Relative to Figure D-1, it should be noted that areas shown in blue are representative of the wells cut into the biological shield to permit positioning of the power range, intermediate range, and source range ex-core detectors. These areas are essentially air filled and do not include concrete. Therefore, the gamma ray dose shown for these areas is not relevant to the potential degradation of the biological shield. For any given operating time, the integrated dose values shown in red font represent the areas of the shield wall that exceed the threshold dose of 1.OE+10 rad. At 36 EFPY these areas are confined to a relatively small portion of the shield geometry; while at times beyond 48 EFPY the entire circumference of the shield wall is impacted. For all cases, the impacted area is confined to a depth of eight inches or less from the inner surface of the shield wall. For each irradiation time, the integrated gamma ray dose values shown in black font represent areas of the biological shield wall that would be anticipated to experience an integrated gamma ray dose less than the damage threshold of 1.0E+1 0 rad throughout the total irradiation period.

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 D-3 Azimuth

[Deg.]

0 5

10 15 20 25 30 35 40 45 0

5 10 15 20 25 30 35 40 45 0

5 10 15 20 25 30 35 40 45 Distance From Biological Shield Inner Radius [In]

0 1.82E+1 0 1.80E+1 0 1.67E+10 1,52E+10 1.37E+10 1,24E+10 1.18E+10 1,.14E+10 1A.2E+ 10 1.11 E-41 0 1.~75E+10, 1.73E+10 1,61 E+1 0 1,.47E+10 1.32E+10 1,20E+10 1,14E+10 1 1 0E+ 10 1.0lE+1

1.07E+10 1.60E+10 1.49E+,10 1,36E+10 1.22E+10 1,11E+10 1,.06E+10 1.02E+10 9.*9+09 9.92E+09 2

4 6

54 EFPY 8

10 12 14 1.69E+1 0~

1.67E+1 0' 1.41 E+10 1,24E+10 1.10E+10 9.96E+09 9.47E+09 9.24E+09 1.03E+1 0 1.63E+:10 1.61 E+1 0 1,36E+10 1.20E+10 1,06E+10 9.63E+09 9.15E+09 8.93E+09

~9.91E+9~

9.97E+09~

.1.51 E+10 1i.49E+1 0 1,11Eý10 9.84E+09 8.92E+09 8.48E+09 8.28E+09 9.18E+d9

9. 23 E+09

~1.61 E+'1 0 1.58E+1OU 1,28E+10 1,10E+10 9.65E+09 8.72E+09 8.30E+09 8.25E+09 9.60E*09.

9.77E+09 1.53E+10 1,24E+10 1,06E+10 9.32E+09 8.43E+09 8.03E+09 7.98E+09

~9.44E+09 1.43E+10 1.41 E+10' 114E+ 10 9.84E+09 8.63E+09 7.81 E+09 7.43E+09 7.39E+09 8.60t+09 8.74Ei-09 1.54E+10 ~1.48E+1 07 1,44E+10 1,27E+10 1.15E+10 9.96E+09 9.55E+09 7.95E+09 8.25E+09 6.76E+09 7.45E+09 6.10E+09 7.11E+09 5.85E+09 7.25E+09 6.17E+09 8.30E+09 7.23E+09 9.32E+09 7 8.91 E+099 52.1 EFPY 1.48E+10 1 39E+/-10 1,11E+10 9.23E+09 7.97E+09 7.20E+09 6.87E+09 7.01 E+09 8.02E+09 9.01 E+09 1.43E+~1.

1.23E+10 9.62E+09 7.68E+09 6.54E+09 5.89E+09 5.66E+09 5.96E+09 6.99E+09 8.61 E+09 1.42E+10~

1,11E+10 8.35E+09 6.36E+09 5.32E+09 4.79E+09 4.63E+09 5.06E+09 6.31 E+09 8.39E+/-09 1 ý38E+1 0 1.07E+10 8.06E+09 6.14E+09 5.14E+09 4.63E+09 4,48E+09 4.89E+09 6.10E+09 8.11 E+09 1.27Et.10 9.91 E+09 7.45E+09 5.68E+09 4.76E+09 4.28E+09 4.15E+09 4.53E+09 5.65E+09 1,30E+10 9.54E+09 6.74E+09 4.90E+09 4.03E+09 3.62E+09 3.53E+09 4.01 E+09 5.32E+09 7.59E+09 1,25E+10 9.22E+09 6.51 E+09 4.73E+09 3.90E+09 3.50E+09 3.41 E+09 3.88E+09 5.15E+09 7.34E+09 1.16E+10 8.52E+09 6.01 E+09 4.38E+09 3.60E+09 3.24E+09 3.16E+09 3.59E+09 4.77E+09 6.79E+09 1,05E+10 7.93E+09 5.28E+09 3.68E+09 2:98E+09 2.67E+09 2.63E+09 3.09E+09 4.36E+09 6.13E+09 1,02E+10 7.66E+09 5.1OE+09 3.56E+09 2.88E+09 2.58E+09 2.54E+09 2.99E+09 4.22E+09 5.92E+09 9.39E+09 7.08E+09 4.71 E+09 3.29E+09 2.66E+09 2.39E+09 2.35E+09 2.77E+09 3.90E+09 5.49E+09 48 EFPY 1.37E+1 0 1.29E+10 1.02E+10 8.53E+09 7.38E+09 6.67E+09 6.37E+09 6.50E+09 7.43E+09

8. 35E +09'

'1.32E+10 1.13E+10 8.89E+09 7.10E+09 6.05E+09 5.46E+09 5.24E+09 5,52E+09 6.48E+09 7.9E+09~ ~7.51E+-0'9 Figure D-1 Axial Maximum Integrated Gamma Ray Dose [rad] as a Function of Reactor Operating Time WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 D-4 Azimuth Distance From Biological Shield Inner Radius [In]

[Deg.]

0 2

4 6

8 10 12 14 44 EFPY 0

1.49E510 1.39E+1O0 1.32E+10 1.26E+10 1.22E[t101.1t7E+10 1,07E+10 8.64E+09 5

1.47E

101.37E510 1,30E1*.19E+10 1.04E+10 9.12E+09 7.83E+09 6.51E+09 10 1.37E+10 1.16E+10 1.05'5E+10 9.41E+09 8.18E+09 6.86E+09 5.53E+09 4.33E+09 15 1,25E+10 1,02E+10 9.06E+09 7.85E+09 6.54E+09 5.23E+09 4.03E+09 3.03E+09 20 1.13E+10 9.06E+09 7.95E+09 6.79E+09 5.57E+09 4.38E+09 3.32E+09 2.45E+09 25 1,03E+10 8.23E+09 7.20E+09 6.15E+09 5.03E+09 3.95E+09 2.99E+09 2.20E+09 30 9.78E+09 7.82E+09 6.86E+09 5.87E+09 4.83E+09 3.82E+09 2.92E+09 2.17E+09 35 9.39E+09 7.63E+09 6.81E+09 5.99E+09 5.09E+09 4.18E+09 3.31E+09 2.56E+09 40 9*21 E+09 8.47E+09 7.93E+09 6.85E+09 5.97E+09 5.21E+09 4.40E+09 3.60E+09 45

~9.15E+~09 8.51 E+09 8.06E+09 7.70E+09 7.35E+09 6.93E+09 6.26E+09 5.06E+09 36 EFPY 0

1.36E,10 1,27E+10 1*20E+10 1.1*5E+10 1.11E+*10 1.07E+10 9.73E+09 7.88E+09 5

1.5E+10 1.25E+1001.18E+10.

08E+10 9.50E+09 8.32E+09 7.15E+09 5.94E+09 10 1,25E+10 1,06E+10 9.59E+09 8.59E+09 7.47E+09 6.26E+09 5.05E+09 3.95E+09 15 1.14E+10 9.32E+09 8.28E+09 7.17E+09 5.97E+09 4.78E+09 3.68E+09 2.77E+09.

20 1,03E+10 8.29E+09 7.27E+09 6.21E+09 5.09E+09 4.01E+09 3.04E+09 2.24E+09 25 9.40E+09 7.53E+09 6.59E+09 5.63E+09 4.61E+09 3.62E+09 2.73E+09 2.01E+09 30 8.96E+09 7.17E+09 6.28E+09 5.38E+09 4.43E+09 3.50E+09 2.67E+09 1.99E+09 35 8.60E+09 6.99E+09 6.24E+09 5.49E+09 4.66E+09 3.83E+09 3.03E+09 2.34E+09 40

j4-E'd09 7.75E+0 7:26ýE+09 6.27E+09 5.47E+09 4.77E+09 4.02E+09 3.30E+09 45

8.37E+09

7.79E+09 7.38E+09 7.064EL09.673E+09 6.34E+09' 5.73E+09 4.63E+09 Figure D-1 (Continued)

Axial Maximum Integrated Gamma Ray Dose Irad] as a Function of Reactor Operating Time Neutron Fluence Projections The neutron exposure projections beyond the end of Cycle 27 that were used in pressure vessel integrity assessments for the Kewaunee reactor were based on an NRC approved methodology that follows the guidance and meets the requirements of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001. The accepted fluence methodology described in Regulatory Guide 1.190 is based on a best estimate rather than a conservative approach. In particular, the guide states that "The methodology presented is intended as a best estimate, rather than a bounding or conservative fluence determination. For example, in the RTPTS correlation called for in 10CFR 50.61, the best estimate fluence is used to calculate the shift in RTPTS. Uncertainty in the shift prediction (e.g., from uncertainty in the fluence, chemistry factor, or shift correlation) has been included separately in an explicit margin term."

WCAP-16643-NP K

June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 D-5 For the Kewaunee reactor, the best estimate fluence projections were determined by first completing a fuel cycle specific analysis for all cycle designs completed at the time of the analysis (Cycles 1 through

27) and then projecting into the future based, on an anticipated mode of operation for the reactor. In this case, the future projections were based on the assumption that the cycle average radial core power distribution for Cycle 27 and a cycle average axial power distribution calculated for an equilibrium uprate cycle would be applicable for future operation. The power uprate from 1650 MWt to 1772 MWt was assumed to occur during Cycle 26.

The neutron exposure projections beyond the end of Cycle 27 were also based on an operating scenario.,

that consisted of a series of 18 month operating cycles each 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 corresponding total operating times of 33.0 EFPY at EOL and 52.1 EFPY at EOLE.

Depictions of the specific radial and axial power distributions used for the fluence projections are given in Figure D-2 and Table D-1, respectively. In regard to Figure D-2 and Table D-1, the relative power distribution data are intended to represent an average over the fuel cycle rather than a snapshot at any particular time during the cycle. Results of the fluence projections applicable to the maximum exposure location for the Kewaunee reactor vessel are provided in Table D-2. The neutron fluence values included in Table D-2 form the basis for the current integrity evaluations performed for the Kewaunee pressure vessel. The maximum neutron flux associated with the projections given in Table D-2 was 3.20E+10 n/cm'-s.

The linkage between the neutron fluence accrued by the pressure vessel and the corresponding number of effective full power years (EFPY) of operation, also included in Table D-2, is dependent on the continued use of core power distributions similar to the average distributions used to formulate the projections and on the continued applicability of the assumed 95.6% reactor capacity factor. Both of these factors are part of the overall uncertainty associated with the fluence projections.

To assure optimum reactor operation, it is desirable to provide as much flexibility as possible to the core designer without compromising the pressure vessel integrity evaluations. That is, as much leeway as possible should be provided to the designer regarding loading pattern options. An example of a fuel cycle option that might be considered for the long term could be the use of Gd versus IFBA's as burnable absorbers. An example of a short term option could be an individual cycle re-design due to fuel damage or other unforeseen circumstances. Either of these examples could have a small cycle dependent impact that may either increase or decrease the fluence relative to projections.

From a regulatory viewpoint, the key requirement relative to neutron exposure is that the fluence limit used in the vessel integrity evaluations should not be exceeded. One way to accomplish this end is to restrict the power generation in key locations along the periphery of the reactor core. Since the neutron fluence at the pressure vessel wall is dominated by the power generation in those key locations, this approach tends to assure that the rate of fluence accumulation (flux), as well as the total fluence, are never exceeded throughout the reactor lifetime. While straightforward in concept, this approach may be undesirable because it restricts the loading pattern options available to the core designer and does not preclude the necessity to monitor the actual vessel exposure.

WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 D-6 Placing a restriction on peripheral power generation, in effect, limits the rate of fluence accumulation (flux) for each vessel material. In actuality, it is important to note that it is the total fluence that sets the operational limit not the rate of fluence accumulation. It is acceptable for the plant specific fluence at any point in time to be above or below the projection line as long as margin remains between the actual fluence and the fluence limit used in the pressure vessel integrity evaluations.

An alternative to restricting the rate of fluence accumulation to the vessel can be achieved using the data provided in Figure D-3. In Figure D-3, the maximum neutron flux at the pressure vessel wall per unit fuel assembly power is depicted for the two octants spanning the azimuthal location of the maximum fluence (0 Degrees). The data provided in Figure D-3 include the effects of symmetry along the 45 degree azimuths defining the two octants and, therefore, are sufficient to define the total impact on the vessel fluence at the 0 degree location.

The data from Figure D-3, when multiplied by a given core radial power distribution, summed over the two octants, and then multiplied by the maximum axial peaking factor, provide a good approximati6n of the maximum neutron flux at the pressure vessel inner radius for the design cycle in question. The resultant flux can then be compared with the projection flux of 3.20E+10 n/cm 2-s to detennine if the design under consideration results in fluence accumulation above or below the projection trend-line. The approach can be applied over several fuel cycles in 'order to trend actual core designs against the fluence projections.

This approach should be treated as a good approximation to the determination of pressure vessel neutron exposure that provides a useful guide for the core designer, but should not be construed as a detailed fluence analysis. While this approximate approach is valid for tracking trends versus projections, detailed fluence updates should be performed periodically in order to determine the actual best estimate vessel exposure as required by Regulatory Guide 1.190.

Relative Assembly Power 3

4 5

6 7

8 9

10 11 G

0.887 H

1.223 1.121 1.223 I

1.284 1.208 1.269 1.208 1.284 J

1.215 1.188 1.282 1.208 1.282 1.188 1.215 K

0.487 1.134 1.238 1.246 1.192 1.246 1.238 1.134 0.487 L

0.353 0.568 1.155 1.242 1.155 0.568 0.353 M

0.345 0.421 0.345 Figure D-2 Cycle Average Relative Radial Power Distribution Used for Fluence Projections WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 D-7 Westinghouse Non-Proprietary Class 3.

D-7 G

H J

K L

M Neutron Flux (E > 1.0 MeV) [n/cm 2-s]

3 4

5 6

7 8

3.17E+05 5.05E+05 9.62E+05 5.05E+05 9.82E+05 3.41E+06 4.32E+06 3.41E+06 9.

2.14E+06 7.78E+06 1.95E+07 2.18E+07 1.95E+07 7.

2E+06 1.40E+07 5.09E+07 1.49E+08 1.86E+08 1.49E+08 5.

9.41E+07 3.83E+08 1.32E+09 1.88E+09 1.32E+09 3.

1.52E+10 2.46E+10 1.52E+10 9

10 11 2.82 82E+05 78E+06 09E+07 83E+08 2.14E+06 1.40E+07 2.82E+06 9.41 E+07 V

Figure D-3 Axially Averaged Pressure Vessel Flux per Unit Fuel Assembly Power 0 Degree Pressure Vessel Inner Radius - 1772 MWt WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 D-8 Westinghouse Non-Proprietary Class 3 D-8 Table D-1 Cycle Average Relative Axial Power Distribution Used for Fluence Projections Fraction of Core Height 0.00 0.02 0.06 0.10 0.15 0.19 0.23 0.27 0.31 0.36 0.40 0.44 0.48 0.52 0.57 0.61 0.65 0.69 0.73 0.78 0.82 0.86 0.90 0.94 0.99 1.00 F(z) 0.23 0.41 0.78 0.93 0.99 1.03 1.06 1.08 1.10 1.12 1.13 1.15 1.15 1.16 1.16 1.15 1.14 1.13 1.11 1.09 1.06.

1.02 0.93 0.77 0.42 0.30 WCAP-16643-NP June 2009 Revision 2

Westinghouse Non-Proprietary Class 3 D-9 Table D-2, Maximum Fast Neutron (E > 1.0 MeV) Projections for the Kewaunee Pressure Vessel Cumulative Operating Neutron Time Fluence Cycle

[EFPY]

[n/cm2]

EOC 26 24.6 2.60E+19 EOC 27 26.0 2.73E+19 Future 28.0 2.94E+19 Future 33.0 3.44E+19 Future 36.0 3.75E+19 Future 40.0 4.15E+19 Future 44.0 4.56E+19 Future 48.0 4.96E+19 Future 52.1 5.37E+19 WCAP-16643-NP June 2009 Revision 2

ML101110128

Text

SUMMARY

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