ML081620438

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Summary of Facility Changes, Tests and Experiments and Summary of Commitment Changes
ML081620438
Person / Time
Site: Kewaunee Dominion icon.png
Issue date: 06/02/2008
From: Wilson M
Dominion Energy Kewaunee
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
08-0293, LIC/MH/R1
Download: ML081620438 (17)
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Text

Dominion Energy Kewaunee, Inc.

N490 Highway 42, Kewaunee, WI 54216-9511 #Dominion JUN 02 2008 ATTN: Document Control Desk Serial No. 08-0293 U. S. Nuclear Regulatory Commission LIC/MH/R1 Washington, DC 20555-0001 Docket No.: 50-305 License No.: DPR-43 DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION

SUMMARY

OF FACILITY CHANGES. TESTS AND EXPERIMENTS AND

SUMMARY

OF COMMITMENT CHANGES Pursuant to 10CFR 50.59(d)(2), enclosed is a summary description of Facility Changes, Tests and Experiments evaluated in accordance with 10 CFR 50.59(c) and implemented at the Kewaunee Power Station during the last reporting period, which is defined as not to exceed 24 months.

A commitment change evaluation summary for those commitment changes that occurred during the last reporting period is also enclosed.

The enclosed summary encompasses all changes that occurred in both of the stated areas since our prior submittal of this information on June 4, 2007.

If you have questions or require additional information, please feel free to contact Ms.

Mary Jo Haese at 920-388-8277.

Very truly yours, Mic el J. Wilson Diretor Safety and Licensing, Kewaunee Power Station Commitments made by this letter: NONE 417~+

Serial No. 08-0293 Page 2 of 2 cc: Regional Administrator, Region III U. S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Ms. M. H. Chernoff Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-H4A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Kewaunee Power Station

Serial No. 08-0293 ATTACHMENT 1

SUMMARY

OF FACILITY CHANGES, TESTS AND EXPERIMENTS AND

SUMMARY

OF COMMITMENT CHANGES KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No. 08-0293 Attachment 1 Page 1 of 14 10CFR50.59 Evaluation 07-01-00 Activity Evaluated Procedure change to operate Control Room (CR) Air Conditioner (A/C) Chilled Water Pump "A" in Off/Manual Brief Description This activity is the change of the control of the CR A/C cooling coil supply from the automatic startup of Chilled Water Pump "A" (via the CR A/C Chiller Unit "A"), to the CR A/C "A" cooling requiring manual action (external to the CR boundary) to start the CR A/C Chilled Water Pump "A." Prior to this change, the plant design ensured that CR A/C would be automatically restored as power is restored to one or more of the safeguards electrical buses. An assessment of the manual actions necessary to restore CR A/C demonstrated that adequate time is available, and radiological conditions would allow, for the manual restoration of CR A/C Chilled Water Pump "A". It was therefore concluded that the action of manually restoring automatic operation of the CR A/C Chilled Water Pump "A" by manually placing the CR A/C Chilled Water Pump Breaker switch in "ON" is within the requirements for crediting the manual operator action in lieu of the automatic chiller operation. This conversion from automatic to manual actuation did not require prior NRC approval.

Reason for the Change Preventing the automatic start of the CR A/C Chilled Water Pump "A" during a Safety Injection (SI) sequence was necessary as a compensatory measure to address concerns that Safeguards 480V AC motor control center voltages may trend low under certain SI scenarios.

Summary All applicable questions within the 50.59 evaluation were answered "no." Question 8 was answered "N/A." This indicated that the modification could be implemented without prior NRC approval. It should be noted that a subsequent plant modification was implemented that allowed restoring the CR A/C Chilled Water Pump "A" to its normal automatic mode of operation.

10CFR50.59 Evaluation 07-03-00 This evaluation resulted in a license amendment request (LAR).

Activity Evaluated Kewaunee Power Station (KPS) Calculation CAL-20776-SE-001 Rev. 2, "125/15 Ton Aux. Building Crane Trolley Structural and Seismic Design" Brief Description A structural and seismic analysis of the upgraded crane was performed as part of Design Change Request (DCR) 3629. This 50.59 evaluation covered the structural and

Serial No. 08-0293 Attachment 1 Page 2 of 14 seismic calculation that evaluated the ability of the bridge girders, bridge end trucks, and upgraded trolley to resist the static and dynamic loads specified in CMAA-70-2004 and ASME NOG-1 -2004. The KPS Updated Safety Analysis Report (USAR) requires the affected crane (as a Nuclear Safety Design Class I*Component) to be designed to Class I Design Basis Earthquake Loading (dynamic). The seismic analysis methodology for Class I structures and components is described in Appendix B to the KPS USAR and the Blume Report. The calculation revision employed the effects of the rolling of the crane trolley and bridge on their respective rails during the seismic event as a method to moderate the Design Basis Earthquake (DBE) design loads in a realistic, yet conservative manner. The use of rolling is not a methodology described (allowed) in the USAR for seismic analyses. Additionally, a 7 percent damped horizontal spectra was used per ASME NOG-1-2004. The USAR currently does not specifically define a horizontal damping factor for the crane.

Reason for the Change KPS (DCR) 3629 upgraded the Auxiliary Building Crane to make it single failure proof in accordance with NUREG 0612 and NUREG 0554. The calculation was revised to address the changes implemented under DCR 3629 and verify that the requirements for a single failure crane were met.

Summary Because the scope of the activity being evaluated was the affected calculation revision, all questions with the exception of question 8 were answered "N/A." Question 8 was answered "yes" because this activity involved allowances for the rolling of the crane trolley and bridge. This demonstrated the practice to be a departure from a method of evaluation described in the USAR used in establishing the design basis. On November 9, 2007, KPS submitted LAR 234 requesting that the NRC review and approve the allowances for the rolling of the crane trolley and bridge. On April 8, 2008, KPS withdrew LAR 234. KPS continues to develop the approach necessary to verify that the Auxiliary Building Crane meets the applicable criteria for single failure proof applications.

10CFR50.59 Evaluation 07-05-00 Activity Evaluated Scaffold Request MM2-07-084 Brief Description The referenced scaffold was erected to facilitate access to valves located in the KPS North Penetration Room. To simplify steps necessary to safely access valves located in the KPS North Penetration Room, KPS Installed the scaffold. Normal KPS Appendix R design requirements for the illumination of access and egress paths for local manual operation of plant equipment were not met with the scaffold installed. The KPS Appendix R Program allows the implementation of compensatory measures (e.g.

Serial No. 08-0293 Attachment 1 Page 3 of 14 establishing backup emergency lighting) in lieu of a fully conforming emergency lighting configuration.

Reason for the Change Scaffold was erected to facilitate safe operator access to valves located in the KPS North Penetration Room.

Summary All applicable questions within the 50.59 evaluation were answered "no." Question 8 was answered "N/A." The 50.59 evaluation concluded that the proposed scaffolding could be implemented without prior NRC approval, in part because appropriate portable lighting units were maintained available to, and would be obtained by, personnel tasked with the applicable Appendix R local manual actions.

10CFR50.59 Evaluation 07-05-01 Activity Evaluated DCR 3609 - Auxiliary Feedwater (AFW) Flow Control Modification Brief Description This modification upgraded the AFW system by replacing the AFW-2A/B Air Operated Valves with Motor Operated Valves (MOVs), replacing the AFW-1 OA/B MOVs (conversion from gate valves to globe valves), adding a turbine driven AFW pump low discharge pressure trip bypass in the control room, and replacing the AFW header flow elements (conversion from flow orifices to flow venturis). The modification also added AFW header drain valves.

A 50.59 evaluation was required because the modification adversely affected several design functions of the affected valves. Specifically, the controls for AFW-2A, 2B, 10 A, and 1OB were adversely affected. The modified circuits contain switches, contacts, and relays that were not included in the pre-modification design. These new components result in new potential failure mechanisms, which could affect the ability to control AFW flow and pressure, or isolate a faulted Steam Generator. The evaluation assessed the proposed activity to replace the AFW-2A and AFW-2B Air Operated Valves with MOVs, and the modification to AFW-1 OA and AFW-1 OB that enabled their operation in modes not previously considered.

Reason for the Change The purpose of this modification was to eliminate local manual operator actions required by the applicable Operating License Condition, as described in paragraph 2.C.(8) of the KPS Operating License. The license condition states, "OperatorActions The auxiliary feedwater system local manual operatoractions as described in the License Amendment Request submitted May 5, 2005, and supplemented on June 9, 2005, shall be eliminated no later than completion of Kewaunee refueling outage R-29."

Serial No. 08-0293 Attachment 1 Page 4 of 14 Summary All applicable questions within the 50.59 evaluation were answered "no." Question 8 was answered "N/A." Therefore, the modification could be implemented without prior NRC approval.

10CFR50.59 Evaluation 07-06-00 Activity Evaluated Calculation HI-2073752, "WPMR Analysis of Freestanding Rack in North Pool of Kewaunee Power Station," in support of DCR 3633, "Modify Seismic Support Structure in North Spent Fuel Pool" Brief Description DCR 3633 involved removing the support frame from the area that will be used for loading spent fuel storage casks in the north spent fuel pool (SFP). The support frame is part of the support structure for all three spent fuel racks in the north SFP. Removal of the support frame and the modification of the west support platform involved removing the following previously existing components; the support frame, the tie down bolts that secured the northwest rack to the underlying platform, the bolted connections between the subject east and west support platforms, the seismic bracing between the west support platform and the north SFP walls, and the shear bars that were welded to the top surface of the existing platform beneath the northwest rack. With the removal of the subject restraints, the spent fuel rack in the northwest corner of the north SFP became freestanding. Freestanding fuel racks were previously part of the KPS design (racks located in the north end of the KPS fuel transfer canal). A 50.59 evaluation was required because this activity (Calculation HI-2073752, "WPMR Analysis of Freestanding Rack in North Pool of Kewaunee Power Station") involved a change in the method of evaluation that was used in supporting USAR analyses that demonstrated that intended design functions would be accomplished under design basis conditions including natural phenomena such as earthquakes.

Reason for the Change The inside dimensions of the support frame are smaller than the outside diameter of the spent fuel storage casks and, therefore, the frame removal was required prior to loading the first spent fuel storage cask.

Summary Because this activity was limited to a calculation, questions 1 through 7 of the 50.59 were marked as "N/A." Question 8 was answered "no" because the NRC had accepted the method employed within the calculation (using DYNARACK for the seismic analysis of the spent fuel racks in the canal pool) by the issuance of a Safety Evaluation Report (SER). The intended application of the methodology as approved in the SER was for evaluating spent fuel racks in another KPS SFP.

Serial No. 08-0293 Attachment 1 Page 5 of 14 10CFR50.59 Evaluation 08-01-00 Activity Evaluated The activity was the allowance of the software Electrical Transient Analysis Program (ETAP) Release 5.5.6N as a calculation tool to suplport the performance of electrical calculations for safety related systems, structures and components. This software product will replace the existing A_FAULT, CAPTOR and DAPPER (DOS) programs.

Brief Description The ETAP 5.5.6 software was authorized to be used to perform complex and/or repetitive electrical calculations and load flow modeling for designing, analyzing and simulating AC, DC, AC/DC, Single-Phase, Three-Phase, Balanced, Unbalanced, Loop and Radial electrical systems. The use of ETAP Release 5.5.6N software, instead of DAPPER, modifies the process used to show adequate safeguard bus voltages at the degraded grid relay setpoint (minus allowed variation) under the worst-case model, SI.

The DAPPER model, as an element of the method of evaluation (MOE), is specified in the licensing basis correspondence, and the DAPPER model demonstrated acceptable analysis as compared to test data (field measurements) to comply with regulations.

Based on the benchmarking between DAPPER and ETAP Release 5.5.6N (with respect to actual field test measurements), the results of the new analysis are essentially the same. The two programs use similar iterative matrix mathematical methods and the results have been shown to be the same within the acceptance criteria of position #4 of Power (Systems) Branch Technical Position-1 (PSB-1). Therefore, because of the similarity in application method and results, the use of ETAP Release 5.5.6N software instead of DAPPER was judged to not result in a departure from a method of evaluation described in the USAR used in establishing the design bases.

Reason for the Change This change was made to meet the requirements of the Current Licensing Basis (CLB) for the KPS relative to software used for load flow modeling and complex and/or repetitive electrical calculations.

Summary Because this activity was limited to a calculation, questions 1 through 7 of the 50.59 evaluation were marked as "N/A." Question 8 was answered "no."

10CFR50.59 Evaluation 08-03-00 Activity Evaluated Degraded Grid and Loss of Voltage Relay Replacement via a DCR Brief Description The modification replaced the Loss of Voltage Relays (LVR) and Degraded Voltage Relays (DVR) that are a part of the protection scheme for KPS ESF Buses 1-5 and 1-6.

Serial No. 08-0293 Attachment 1 Page 6 of 14 The original electro-mechanical devices are no longer in production. A replacement solid-state electronic relay was selected for use in safety related applications. The new relays for both buses are located in the Relay Room, rather than in the respective Diesel Generator Rooms (current location). They also require 125V DC control power to ensure functionality. The DC control power is supplied from the DC circuitry in the existing voltage restoration panels. This activity required a 50.59 evaluation because in the modified configuration a loss of the 125V DC control power to the new LVRs and DVRs results in loss of relay function. Additionally, the new LVRs and DVRs have failure mechanisms different from the original relays.

Reason for the Change This change was made to accommodate the fact that the original relays are no longer in production and spares are difficult to obtain. The new relay models are currently in production and are readily available.

Summary All applicable questions within the 50.59 evaluation were answered "no." Despite the introduction of new failure mechanisms, it was shown that there was not a malfunction with a different result created. Additionally, question 8 was answered "N/A."

IOCFR50.59 Evaluation 08-03-01 Activity Evaluated The 50.59 evaluation was developed to reflect changes to the LVR and DVR setpoint values from those considered in the original evaluation (discussed in the pervious sections of this report).

Brief Description The changed setpoint limits implemented coincident with the replacement of the DVRs and LVRs were the focus of this evaluation. The nominal setpoint for the LVRs and DVRs was changed, however the maximum and minimum voltage setpoint limits for the new relays are within the range provided in Technical Specifications (TS) Table TS 3.5-

1. The relay response time delays remain below the maximum time delays permitted in Table TS 3.5-1.

Reason for the Change This change was made to reflect the changes to the LVR and DVR set-point values in the previous subject evaluation (08-03-00).

Summary All applicable questions within the 50.59 evaluation were answered "no." Question 8 was answered "N/A."

Serial No. 08-0293 Attachment 1 Page 7 of 14 10CFR50.59 Evaluation 08-04-00 Activity Evaluated Calculation No. C11723, Rev. 0, titled: "125VDC Battery BRA-101 and BRB-101 sizing, voltage drop, short circuit and charger sizing" Brief Description Calculation No. C11723, Rev. 0, titled: "125VDC Battery BRA-1 01 and BRB-101 sizing, voltage drop, short circuit and charger sizing"; contains changes in the MOE for the safeguards battery and charging sizing. This is an initial issue of Calculation No.

C11723. The calculation establishes and documents certain elements of the basis for the safety related 125V DC power system using the ETAP DC software package. The safety related 125V DC system model contains the elements required for battery sizing, DC system voltage drop, short circuit analyses and battery charger sizing. The application of the new calculation method did not result in a departure from a MOE described in the USAR used in establishing the design bases or in the safety analyses, because the change gave results that are more conservative, or essentially the same as those results obtained by the methodology described in the USAR.

Reason for the Change A design modification replaced the original C&D Model LCR-19 cells for battery BRA-101 and BRB-1 01 with new C&D Model LCR-25 cells. Calculation No. C11723, Rev. 0, titled: "125VDC Battery BRA-101 and BRB-101 sizing, voltage drop, short circuit and charger sizing" was performed to support that modification.

Summary Because this activity was limited to a calculation, questions 1 through 7 of the 50.59 were marked as "N/A." Question 8 was answered "no" because the activity does not result in a departure from a MOE described in the USAR used in establishing the design bases or the safety analyses, because the new method gave results that are more conservative, or essentially the same as those results obtained by the methodology described in the USAR.

10CFR50.59 Evaluation 08-05-00 Activity Evaluated The activity is the approval and issuance of the initial revision of Calculation No.

C11450, titled: "Auxiliary Power System Modeling and Analysis". This evaluation covers the change in the software from DAPPER to ETAP and the associated evaluation of methodology analysis elements.

Brief Description This evaluation was performed as a result of changing from a USAR-described MOE analysis element. In particular, the manner in which motor efficiencies are used in

Serial No. 08-0293 Attachment 1 Page 8 of 14 calculating EDG load has changed. The new method used (employing revised analysis elements for motor efficiencies) is less prescriptive than that described in the USAR.

Still, the method results are within the accuracy of the overall calculation and thus are essentially the same. Therefore, the activity did not involve a departure from a method described in the USAR.

Reason for the Change This change was made to better characterize the loading on the Emergency Diesel Generator (EDG) systems.

Summary Because this activity was limited to a calculation, questions 1 through 7 of the 50.59 evaluation were marked as "N/A." The new method used (employing revised analysis elements for motor efficiencies) is less prescriptive than that described in the USAR.

Still, the method results are within the accuracy of the overall calculation and thus are essentially the same. Therefore, the activity does not result in a departure from a method described in the USAR, and question 8 was answered "no."

10CFR50.59 Evaluation 08-06-00 Activity Evaluated This activity involved a procedure change for addressing the loss of a single Fuel Oil Transfer Pump (FOTP).

Brief Description The procedure change established the steps and controls necessary to open the FOTP discharge cross connect valves to enable either FOTP to feed either set of EDG day tanks. However, this capability is described in the KPS USAR and was, therefore, prescreened out. Options made available in the new procedure included: 1) using the operable FOTP to feed one or both pairs of EDG day Tanks, and/or 2) using a temporary pump to transfer fuel oil from one Fuel Oil Storage Tank (FOST) to the other FOST and its associated operable FOTP. The activity assessed under the 50.59 evaluation also included accepting a non-functioning siphon line connecting the two EDG FOSTs, 'as-is.' The temporary pump assembly utilizes a positive displacement pump located at ground level to transfer fuel oil between the 1A and 1 B EDG FOSTs.

The two tanks fill locations would be used as the transfer point. The temporary pump will have local manual controls and has provision to receive power from a safeguards electrical bus.

Reason for the Change This change was made to support continued compliance with requirements of the CLB for the KPS relative to EDG fuel oil onsite storage requirements.

Serial No. 08-0293 Attachment 1 Page 9 of 14 Summary All applicable questions within the 50.59 evaluation were answered "no." Question 8 was answered "N/A." Therefore, the modification could be implemented without prior NRC approval.

Commitment Change Evaluation Summary Document(s) Evaluated:

1. Letter from C. W. Lambert (NMC) to Document Control Desk (NRC), "Kewaunee Improvement Initiatives - Commitments," dated March 18, 2005.
2. Letter from E. S. Grecheck (DEK) to Document Control Desk (NRC), "Endorsement and Adoption of Licensing Actions," dated September 15, 2005.
3. Letter from W. R. Matthews (DEK) to Document Control Desk (NRC), "Update on Improvement Initiatives," dated November 14, 2005.
4. Letter from W. R. Matthews (DEK) to Document Control Desk (NRC), "Closure of Improvement Initiatives to Corrective Action Program," dated December 5, 2006.

Brief

Description:

The commitment was to provide the necessary infrastructure and tools required to execute and reinforce the Picture of Excellence. The implementation steps included establishing monthly Picture of Excellence Review Group (PERG) meetings to foster accountability for implementation of the site Excellence Plans.

The implementation of monthly Picture of Excellence Review Group (PERG) meetings to foster accountability for implementation of the site Excellence Plans has been revised in that the site Excellence Plans are currently reviewed during Station Leadership Meetings and Station Management Alignment Meetings.

Bases for change: The implementation of monthly PERG meetings to foster accountability for implementation of the site Excellence Plans has been revised in that the site Excellence Plans are currently reviewed during Station Leadership Meetings and the Station Management Alignment Meetings. The Leadership meetings are a means of achieving alignment and assuring accountability to our Excellence Plan objective and actions. These meetings occur frequently.

Summary: It has been determined that a separate group, i.e. PERG, was not necessary for review of the Site Excellence Plans since the attendees at the Leadership meetings include Directors, Managers, and Supervisors at KPS. The review of the Site Excellence Plans has been incorporated into the Leadership meetings described above.

These meetings are currently scheduled weekly, however, they will be no less frequent than monthly.

Serial No. 08-0293 Attachment 1 Page 10 of 14 Commitment Change Evaluation Summary Document(s) Evaluated:

1. Letter from C. W. Lambert (NMC) to Document Control Desk (NRC), "Kewaunee Improvement Initiatives - Commitments," dated March 18, 2005.
2. Letter from E. S. Grecheck (DEK) to Document Control Desk (NRC), "Endorsement and Adoption of Licensing Actions," dated September 15, 2005.
3. Letter from W. R. Matthews (DEK) to Document Control Desk (NRC), "Update on Improvement Initiatives," dated November 14, 2005.
4. Letter from W. R. Matthews (DEK) to Document Control Desk (NRC), "Closure of Improvement Initiatives to Corrective Action Program," dated December 5, 2006.

Brief

Description:

The commitment was to complete AC electrical models and calculations to provide clear bases for safety related settings and loads. Calculations were to be completed, validated and issued by the end of the 1 st Quarter 2007 but the completion was delayed to March 31, 2008.

Bases for change: The overall Electrical Calculation Project was expected for completion by December 31, 2007 as originally discussed with the NRC. The calculations identified in this commitment were completed by March 31, 2008.

There were significant delays with obtaining the motor data from the vendors (many of the motor models needed to be recreated by the vendors and it took one year longer than anticipated). Also, delays were incurred because of quality issues with the initial products submitted by the contractor.

Summary: The calculations identified in this commitment that were expected for completion by end of the first quarter as originally discussed with the NRC were completed by March 31, 2008.

Commitment Change Evaluation Summary Document(s) Evaluated:

1. Letter from J. G. Partlow (NRC) to All Holders of Operating Licenses or Construction Permits for Nuclear Power Plants, "Service Water System Problems Affecting Safety-Related Equipment (Generic Letter 89-13)," dated July 18, 1989.
2. Letter from C. R. Steinhardt (WPSC) to Document Control Desk, "Response to Generic Letter 89-13", dated January 29, 1990.
3. Letter from M. J. Davis (NRC) to K.H. Evers (WPSC), "NRC Generic Letter (GL) 89-13, 'Service Water System Problems Affecting Safety-Related Equipment,' dated July 18, 1989", dated February 14, 1990.

Serial No. 08-0293 Attachment 1 Page 11 of 14

4. Letter from J. G. Partlow (NRC) to All Holders of Operating Licenses or Construction Permits for Nuclear Power Plants, "Service Water System Problems Affecting Safety-Related Equipment (Generic Letter 89-13, Supplement 1)", dated April 4, 1990.
5. Letter from G. C. Wright (NRC) to K.W. Evers (WPSC) in reference to a meeting between the NRC and WPSC staff, dated July 10, 1990.
6. Letter from C. R. Steinhardt (WPSC) to Document Control Desk, "Implementation of Generic Letter 89-13 Recommended Actions," dated October 21, 1991.
7. Letter from A. G. Hansen (NRC) to C.A.Schrock (WPSC), "NRC Generic Letter (GL) 89-13, 'Service Water System Problems Affecting Safety-Related Equipment,'

dated November 21, 1991.

Brief

Description:

In NRC Letter from C.R. Steinhardt (WPSC) to Document Control Desk, "Implementation of Generic Letter 89-13 Recommended Actions," dated October 21, 1991, which addressed the implementation of Action Item II of Generic Letter 89-13, the following commitment was made: "Assurance of continued operability of the ... D/G jacket water coolers will therefore be provided by a combination of SW temperature rise monitoring and periodic inspection and cleaning. Frequency of cleaning will be as shown necessary by inspection results."

The commitment change revises the commitment to read, "Assurance of continued operability of the D/G jacket water coolers will be provided by periodic inspection and cleaning of the coolers. The frequency of cleaning will be as shown necessary by inspection results."

Bases for change: Per Action Item II of Generic Letter 89-13 periodic inspection and cleaning of a heat exchanger is one of the NRC-designated methods for providing assurance that the heat exchanger will perform its safety related function.

The original commitment for SW temperature rise monitoring was not an effective monitoring method for a diesel generator jacket water heat exchanger. Temperature rise monitoring on the service water side verifies a heat load for a given flow at a given inlet temperature. This does not verify the capability of the heat exchanger to remove the required heat load at the design limiting combination of service water flow and temperature unless the temperature rise is measured at the limiting combination of inlet flow and temperature. The shell temperature will remain essentially constant as fouling accumulates because the diesel generator modulating bypass valve will divert more and more flow to the shell side as the cooler loses effectiveness.

Summary: Based upon the large volume of historical diesel generator operation data, typical temperatures for jacket water out of the engine are well below 180 degrees F and are usually in the area of 170 degrees F. This history supports a conclusion that the diesel generator jacket water coolers continue to perform within their design. Proto-

Serial No. 08-0293 Attachment 1 Page 12 of 14 HX analysis demonstrates that the coolers perform per design and are capable of removing the required heat during post accident conditions.

Commitment Change Evaluation Summary Document(s) Evaluated:

1. Letter from C. W. Lambert (NMC) to Document Control Desk (NRC), "Kewaunee Improvement Initiatives - Commitments," dated March 18, 2005.
2. Letter from E. S. Grecheck (DEK) to Document Control Desk (NRC), "Endorsement and Adoption of Licensing Actions," dated September 15, 2005.
3. Letter from W. R. Matthews (DEK) to Document Control Desk (NRC), "Update on Improvement Initiatives," dated November 14, 2005.
4. Letter from W. R. Matthews (DEK) to Document Control Desk (NRC), "Closure of Improvement Initiatives to Corrective Action Program," dated December 5, 2006.

Brief

Description:

The commitment was to complete DC electrical models and calculations to provide clear bases for safety related settings and loads. Calculations were to be completed, validated and issued by the end of the 2007 but the completion was delayed to March 31, 2008.

Bases for change: The overall Electrical Calculation Project was expected for completion by December 31, 2007 as originally discussed with the NRC. The calculations identified in this commitment were completed by March 31, 2008.

There were significant delays with obtaining the motor data from the vendors (many of the motor models needed to be recreated by the vendors and it took one year longer than anticipated). Also, delays were incurred because of quality issues with the initial products submitted by the contractor.

Summary: The calculations identified in this commitment that were expected for completion by December 31, 2007 as originally discussed with the NRC were completed by March 31, 2008.

Commitment Change Evaluation Summary Document(s) Evaluated:

1. Letter from C. E. Rossi (NRC) to All Holders of Operating Licenses or Construction Permits for Nuclear Power Plants, "NRC Bulletin No. 88-04: Potential Safety-Related Pump Loss," dated May 5, 1988.
2. Letter from D. C. Hintz (WPSC) to Document Control Desk, "Follow-up Response to NRC Bulletin 88-04: Potential Safety Related Pump Loss," dated January 31, 1989.

Serial No. 08-0293 Attachment 1 Page 13 of 14

3. Letter from J. G. Giitter (NRC) to C. R. Steinhardt (WPSC), "Response to NRC Bulletin 88-04," dated February 10, 1989.
4. Letter from C. R. Steinhardt (WPSC) to Document Control Desk, "Status of WPSC Evaluation to Increase Safety Injection Pump Recirculation Flow in Response to NRC Bulletin No. 88-04: Potential Safety Related Pump Loss," dated April 26, 1991.
5. Letter from C. R. Steinhardt (WPSC) to Document Control Desk, "Status of WPSC Evaluation to Increase Safety Injection Pump Recirculation Flow in Response to NRC Bulletin No. 88-04: Potential Safety Related Pump Loss," dated February 26, 1993.
6. Letter from C. A. Schrock (WPSC) to Document Control Desk, "Status of WPSC Evaluation to Increase Safety Injection Pump Recirculation Flow in Response to NRC Bulletin No. 88-04: Potential Safety Related Pump Loss," dated April 23, 1991.
7. Letter from C. R. Steinhardt (WPSC) to Document Control Desk, "Status of WPSC Evaluation to Increase Safety Injection Pump Recirculation Flow in Response to NRC Bulletin No. 88-04: Potential Safety Related Pump Loss," dated April 30, 1993.
8. Letter from C. R. Steinhardt (WPSC) to Document Control Desk, "Close-out to NRC Bulletin No. 88-04: Potential Safety Related Pump Loss," dated November 8, 1994.

Brief

Description:

In response to NRC Bulletin 88-04, KPS committed to performing disassembled inspections of the SI pumps every 15 years to ensure no damage is occurring as a result of operation on mini-flow.

SI Pump 1A was due for its 15-year inspection in September 2008. Therefore, it was scheduled for the Refueling Outage 29 in spring of 2008. An evaluation was accomplished that resulted in being able to extend the disassembled inspection of SI Pump 1A to once every 16.5 years.

Bases for change: SI Pumps A and B were inspected after 12 years and 14 years of operation, respectively. No evidence of damage due to minimum flow recirculation was found. If any damage were occurring to the pumps due to recirculation, some evidence of it would have been detected on the impeller eyes, discharge vanes, or volutes during these inspections.

Pump performance monitoring has not shown any degradation in the performance of either SI pump. Axial vibrations, a key indicator of flow instability at low flow rates, have been low and steady throughout pump life. In addition, operation of the pumps on mini-flow has always been smooth and quiet. There is no evidence of high vibration, noise, or axial shuttling.

The operational history of the SI pumps was reviewed for the period since their last inspection. This historical operation was extrapolated out to 16.5 years. The total number of starts and hours of operation in this period is not excessive, even when conservative assumptions are used for the remaining operating cycles before the next inspection.

Serial No. 08-0293 Attachment 1 Page 14 of 14 The pump vendor reviewed and concurred with the evaluation performed for extension of the inspection frequency.

Summary: Based on review of the maintenance and operational history of the SI pumps, the frequency of disassembled inspections may be extended to 16.5 years. At the end of this period, the pumps will still be capable of providing flow under design and licensing basis conditions, including operation on minimum flow recirculation for 34.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

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