ML120100040
| ML120100040 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 03/28/2012 |
| From: | Beltz T Plant Licensing Branch III |
| To: | Heacock D Dominion Energy Kewaunee |
| beltz T, NRR/DORL/LPL3-1, 301-415-3049 | |
| References | |
| TAC ME6288 | |
| Download: ML120100040 (13) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555"()001 March 28, 2012 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Energy Kewaunee, Inc.
Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
KEWAUNEE POWER STATION -ISSUANCE OF AMENDMENT RE: LICENSE AMENDMENT REQUEST TO CHANGE THE CURRENT LICENSING BASIS FOR OPERATION OF SERVICE WATER FLOW TO COMPONENT COOLING HEAT EXCHANGERS (TAC NO. ME6288)
Dear Mr. Heacock:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 211 to Renewed Facility Operating License No. DPR-43 for the Kewaunee Power Station (KPS). The amendment revises the KPS current licensing basis in response to your application dated May 9,2011, as supplemented by letters dated June 30, and October 31,2011.
The amendment revises the KPS current licensing basis regarding the manner in which service water is supplied to the component cooling heat exchangers by the main return valves and the bypass flow control valves.
A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
A
~~
Terry A. Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-305
Enclosures:
- 1. Amendment No. 211 to License No. DPR-43
- 2. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555-0001 DOMINION ENERGY KEWAUNEE, INC.
DOCKET NO. 50-305 KEWAUNEE POWER STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 211 Renewed License No. DPR-43
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Dominion Energy Kewaunee. Inc. dated May 9, 2011, as supplemented by letters dated June 30, and October 31, 2011, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, by Amendment No. 211, the license is amended to authorize revision to the Updated Safety Analysis Report (USAR), as set forth in the application dated May 9, 2011, as supplemented by letters dated June 30, and October 31,2011. The licensee shall update the USAR to incorporate the licensing basis requirements related to the manner in which service water is supplied to the component cooling water heat exchangers as described in the licensee's application and supplements. and the NRC staff's safety evaluation attached to this amendment, and shall submit the revised description authorized by this amendment with the next update of the USAR.
- 2
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance. The USAR changes shall be implemented in the next periodic update of the USAR in accordance with 10 CFR 50.71(e).
FOR THE NUCLEAR REGULATORY COMMISSION Shawn A. Williams. Acting Chief Plant licensing Branch 111-1 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation Date of Issuance: March 28, 2012
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 211 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-43 DOMINION ENERGY KEWAUNEE, INC.
KEWAUNEE POWER STATION DOCKET NO. 50-305
1.0 INTRODUCTION
By application dated May 9,2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML11137A028), as supplemented by letters dated June 30, and October 31, 2011 (ADAMS Accession Nos. ML111820365 and ML113040362, respectively),
Dominion Energy Kewaunee, Inc. (DEK, the licensee), requested an amendment to Renewed Facility Operating License No. DPR-43, revising the Updated Safety Analysis Report (USAR) for the Kewaunee Power Station (KPS). The supplements dated June 30, and October 31,2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on November 1, 2011 (76 FR 67487).
The KPS service water system (SWS) currently supplies cooling water to the component cooling heat exchangers by the automatic opening of the 10-inch main service water (SW) return valves (SW-1300A and SW-1300B) upon receipt of a safety injection (SI) signal. The licensee deleted the automatic opening function of the main SW return valves upon receipt of an SI signal.
Instead, the 4-inch bypass flow control valves (SW-1306A and SW-13068) are opened automatically upon receipt of an SI signal to supply SW to the component cooling heat exchangers. By procedure, the main SW return valves would be remote manually operated at the start of the subsequent recirculation phase of loss-of-coolant accident (LOCA) mitigation.
This strategy was implemented in 2001 without prior NRC approval. The proposed license amendment request seeks NRC approval of the change. The licensee has been operating the facility using this component cooling strategy, in that the required changes are already being reflected in procedures, operator training, plant displays and controls, and in the simUlator.
2.0 REGULATORY EVALUATION
The regulatory requirements and guidance which the NRC staff considered in its review of the license amendment request are as follows:
Enclosure
- 2 Appendix A to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "General Design Criteria [GDC] for Nuclear Power Plants," Criterion 19 - Control room. The regulations in GDC-19 state, in part, that itA control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including toss-of-coolant accidents.... Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures."
The NRC requirements and review criteria that the staff considered to be most applicable include Section 9.1.1.2 of the KPS USAR, Engineered Safety Features Performance Capability, which states:
Criterion: Engineered Safety Features such as the emergency core cooling system and the containment heat removal system shall provide sufficient performance capability to accommodate the failure of any single active component without resulting in undue risk to the health and safety of the public (GDC 41).
Each of the auxiliary cooling systems which serve an emergency function provides sufficient capability in the emergency mode to accommodate any single failure of an active component and still function in a manner to avoid undue risk to the health and safety of the plant personnel and the public.
The SWS and component cooling system (CCS) provide COOling to the emergency core cooling systems and thus are important to preventing undue risk to the health and safety of the public.
10 CFR 50.120, "Training and qualification of nuclear power plant personnel" NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Revision 2, issued in March 2007 Chapter 13 addresses "Conduct of Operations," Specific sub-chapters considered in this review were Chapters 13.2.1, "Reactor Operator Requalification Program; Reactor Operator Training," and 13.5.2.1, "Operating and Emergency Operating Procedures,"
Chapter 18 provides review guidance for "Human Factors Engineering" NUREG-1764, "Guidance for the Review of Changes to Human Actions,"
Revision 1, published in September 2007 Generic Letter 82-33, "Supplement 1 to NUREG-0737 - Requirements for Emergency Response Capability," dated December 17,1982
- 3 NUREG-0700, "Human-System Interface Design Review Guidelines" Revision 2, published in May 2002 NUREG-0711, "Human Factors Engineering Program Review Model," Revision 2, published in February 2004 Information Notice 97-78, "Crediting of Operator Actions in Place of Automatic Actions and Modifications of Operator Actions, Including Response Times," dated October 23, 1997.
In accordance with the generic risk categories established in Appendix A to NUREG-1764, the task under review is involved in the SI sequence, and therefore, is considered "risk-important" due to the fact that it is required to prevent core uncovery when the decay heat removal function is lost. Because of its risk importance, the NRC staff in the Health Physics and Human Performance Branch of the Office of Nuclear Reactor Regulation (AHPB) performed a "Level One" review, Le., the most stringent of the graded reviews possible under the guidance of NUREG-1764.
Note: The AHPB staff performed an assessment of risk only for purposes of scoping its review and may conflict with the licensee's assessment of risk importance or that of other NRC branches, and should not be considered as an accurate assessment of risk when compared to other methods, especially those using plant-specific data and NRC-accepted methods of Probabilistic Risk Analysis and Human Reliability AnalYSis (PRAlHRA).
3.0 TECHNICAL EVALUATION
3.1 Balance of Plant Evaluation 3.1.1 Background The purpose of the SWS is to provide two trains of cooling water to redundant safety-related components including the component cooling heat exchangers, the containment fan coolers, diesel generators, the Auxiliary Building and Turbine Building fan coolers, control room conditioners, and the SI pump stuffing box and lube oil coolers. The SWS is designed such that during normal operation, flow to the component cooling heat exchangers is controlled by the automatic pOSitioning of the 4-inch bypass valves controlled by the component cooling temperature. The SWS also supplies various non safety-related loads.
After 2001, the SWS provided the additional flow to the component cooling heat exchangers during the SI phase of LOCA mitigation by automatically fully-opening the 4-inch bypass valves, instead of automatically opening the 10-inch return valves. By procedure, the 10-inch return valves would be remote manually opened at the beginning of the recirculation phase of LOCA mitigation.
The CCS removes heat from letdown flow to Chemical and Volume Control System and from various reactor coolant system (RCS) components during power operation. The CCS removes residual and sensible heat from the reactor coolant system during plant shutdown through the residual heat removal system. Since the heat is transferred from the CCS to the SWS, the CCS
-4 serves as an intermediate system between the RCS and the SWS and ensures that any leakage of radioactive fluid from the components being cooled is contained within the plant.
During the SI phase of LOCA mitigation, the CCS provides cooling for the containment spray pumps, safety injection pumps, reactor coolant pump seals and the residual heat removal (RHR) pumps. During the recirculation phase of LOCA mitigation, one component cooling pump and one component cooling heat exchanger accommodate the heat removal loads during recirculation. If either a component cooling pump or component cooling heat exchanger fails, the standby pump and the standby heat exchanger provide 100 percent backup.
3.1.2 Evaluation Prior to 2001, the SWS was designed such that upon an SI signal, the maximum SWS flow was automatically provided to the CCS heat exchangers by automatic opening of the 10-inch main return valves (SW-1300A and SW-13008). These valves are located downstream of the CCS heat exchangers. These valves provide maximum SWS flow during SI to the CCS heat exchangers, whose major cooling loads do not occur until the recirculation phase of LOCA mitigation. A plant modification was performed in 2001, changing the automatic feature of SI to fully open SW-1306A and SW-13068, instead of opening SW-1300A and SW-13008. Valves SW-1306A and SW-13068 are 4-inch valves, and referred to as the SWS bypass flow control valves. These valves provide a bypass around the return valves (SW-1300A and SW-13008).
The 2001 plant modification also changed plant procedures, requiring remote manual operation of SW-1300A and SW-13008 from the control room at the start of recirculation phase of LOCA mitigation to provide maximum SWS flow to the CCS heat exchangers. The licensee performed this modification to maximize the SWS flow to the containment fan coils units during the injection phase of LOCA mitigation.
During a recent inspection, the NRC staff questioned the sufficiency of the original 10 CFR 50.59 evaluation that was performed by the licensee for the plant modification in 2001, because the licensee replaced an automatic action with a manual operator action. The licensee had revised Emergency Operating Procedure (EOP) ES 1.3, "Transfer to Containment Sump Recirculation," to add steps that*required the operators to manually open the SW-1300A & 8 valves prior to initiating the transfer to containment sump recirculation. The 10 CFR 50.59 evaluation failed to adequately address how this new permanent manual action did not increase the probability for the malfunction of equipment important to safety, previously evaluated in the USAR. Therefore, the licensee submitted license amendment request (LAR) 252 to obtain NRC approval for the new permanent manual action.
The NRC staff reviewed the licensee's LAR and questioned the adequacy of SWS flow through the 4-inch valves during the SI phase of LOCA mitigation, because during SI the CCS heat load changes in that the SI pumps, RHR pumps and containment spray pumps, which require CCS cooling, all start. The licensee did not provide justification in the LAR for the reduced SWS flow to the CCS heat exchangers when SI is initiated. Therefore, the staff conducted a conference call with the licensee to explain that justification of flow through the 4-inch bypass valves (SW 1306A and SW-13068) would be required in order for the NRC staff to start review of the LAR and perform the safety evaluation. The licensee responded in a letter dated June 30,2011, stating that when it performed the plant modification in 2001, they assessed the adequacy of the cooling capabilities of SWS flow through the 4-inch valves during the SI phase of LOCA
-5 mitigation and determined that the cooling flow was adequate. The licensee explained that their calculations show that the heat load on the CCS heat exchangers during SI is less than one fourth the heat load during recirculation; while the SWS flow rate to the CCS heat exchangers through the 4-inch valves SW 1306A & B (1120 gallons per minute (gpm)/1223 gpm) during SI are greater than one half the SWS required flow (2000 gpm to the CCS heat exchangers during the recirculation phase of LOCA mitigation. Thus, flow through the 4-inch valves during SI injection exceeds the flow necessary for satisfactory cooling. The licensee further justified their explanation by stating that the CCS heat loads during SI are less than the CCS heat loads during normal operation because the non essential loads, i.e. letdown heat exchanger, are automatically isolated. Since the 4-inch valves allow satisfactory flow during normal operation, the flow during SI is proven adequate by virtue that the heat load during SI is less than during normal operation.
The NRC staff noted that the licensee stated that their calculations showed that approximately 1200 gpm flowed through each the 4-inch throttle valves, SW 1306A & B. The staff notes that 1200 gpm flow through a 4-inch pipe has a flow velocity greater than 30 ftlsec and that calculated flow through a 4-inch throttle valve could be as high as 50 ftlsec. Therefore, the staff requested that the licensee discuss the validity of the large flow velocities through the 4-inch valves and to address vibration, cavitation and choke flow through these valves, in an e-mail Request for Additional Information dated September 15, 2011 (ADAMS Accession No. ML11259A013). The licensee responded in a letter dated October 31, 2011 and stated that calculations show that cavitation and choke flow do not occur under the flow conditions described above [> 1200 gpm (1300 gpm) through a 4-inch full ported globe valve]. The licensee also reported that results of the service water testing performed in 2001 showed that no excessive vibration or noise was identified at flow rates greater than 1300 gpm through SW 1306A and SW-1306B. Therefore, the staff was satisfied that flow through SW-1306A and B was adequate during SI and that the flow did not cause vibration, cavitation or choked flow.
The technical evaluation of adding an additional manual action to the LOCA mitigation procedure was evaluated by NRC staff in the Health Physics and Human Performance Branch, as discussed in Section 3.2 of this safety evaluation.
3.1.3 Summary Based on the above, the NRC staff concludes that the licensee will continue to meet the requirements of KPS USAR Section 9.1.1.2, "Engineered Safety Features Performance Capability," in that the SWS will continue to provide satisfactory cooling to the CCS during both the SI and recirculation phases of LOCA mitigation.
3.2 Human Performance Evaluation 3.2.1 Description of Operator Action(s) Added/Changed/Deleted A design change in 2001 removed the automatic open feature from valves SW-1300AlB (initiated by an SI signal). Because these valves must be open to provide SW cooling to the component cooling heat exchangers when initiating post-accident containment sump recirculation, the appropriate EOP (ES-1.3, "Transfer to Containment Sump Recirculation") was revised to make this a manual action. In Procedure ES-1.3,
-6 operators are directed to establish component cooling flow to RHR heat exchangers. In Step 9.a, operators are directed to ensure both component cooling pumps are running.
In Step 9.b, operators are directed to open both component cooling heat exchanger outlet valves (SW-1300A for the "A" heat exchanger; SW-1300B for the "B" heat exchanger). In Step 9.c, operators are directed to open both component cooling-to-RHR heat exchanger valves.
3.2.2 Operating Experience Review At the time that this design change was developed, no applicable precedent was identified by the licensee. The NRC staff searched the ADAMS database for similar design changes submitted for approval over the last twenty years, and found that no similar license amendment requests had been submitted for NRC approval.
Therefore, the NRC staff agrees that no significant operating experience was available in 2001 and that none is currently available for, or applicable to, the proposed design.
3.2.3 Functional Requirements Analysis and Function Allocation Because the operator actions associated with the proposed change are simple, are part of an existing task sequence in an approved procedure, and do not add significant workload, a full re-analysis of the functional requirements analysis and function allocation were not necessary. The licensee's engineering analysis was sufficient to identify the control and display requirements.
The NRC staff concludes that the licensee's approach is acceptable based on the fact that, by adding this manual action, there are no significant changes in display and control requirements or operator workload.
3.2.4 Task Analysis Because this operator action is simple, the only aspect requiring re-analysis was the confirmation of time constraints for the action sequence. The licensee's review did not identify any changes that would appreciably impact the time for initiating transition to containment sump recirculation in response to a loss of decay heat removal event. No issues were identified that could have added to the workload of operators in a manner that would prevent them from timely injection of SW into the component cooling heat exchangers.
The NRC staff concludes that full revision of the licensee's task analysis is not necessary, because the actions associated with this proposed change are simple and easy enough to be addressed by the licensee's engineering analysis. Additionally, the licensee has validated that all operating crews can complete the required tasks within the time constraints established in the accident analyses.
3.2.5 Staffing Based on the similarity of the new control interface to the old (changes to labeling only),
and the simplicity of operation, no new or additional staff are required nor are there any new or additional qualifications required to perform the action sequence within the time constraints established.
The NRC staff concludes that no additional staffing or qualifications, or changes thereto, are required and finds this human performance aspect of the license amendment request to be acceptable.
3.2.S Probabilistic Risk and Human Reliability Analyses The licensee chose not to submit a risk-informed application using PRAlHRA and, therefore, did not identify any additional human reliability insights that might be applicable to operator performance. However, because a probabilistic basis for plant changes is not strictly required, this approach is acceptable to the NRC staff.
3.2.7 Human-System Interface Design The licensee identified only two human-system interface design changes required to support the change from automatic to manual opening of valves SW-1300A and SW 1300B. First, annunciator light panels to indicate SW bypass valves (SW-130SA and SW-130SB) being open were added to the SI status panel. Valves SW-130SA and SW-1306B are those that took over the "auto-open on SI" function from valves SW 1300A and SW-1300B. This will allow operators to confirm this new automatic action quickly and easily. Second, the control switch nameplates for valves SW 1300A and SW-1300B were changed to remove the center position "AUTO" identifier.
The NRC staff concludes that these changes are appropriate and acceptable to support the conversion of SW-1300A and SW-1300B from automatic to manual control. Additiona"y, no changes were required to the Safety Parameter Display System, assuring that the general plant overview display will be retained with no negative effect.
3.2.8 Procedure Design In the proposed design, the bypass flow control valves (instead of the SW main return valves) automatically supply required cooling water flow to the component cooling heat exchangers upon receipt of an SI signal. The SW main return valves are only needed to be opened during the subsequent recirculation phase of SI for loss-of-coolant accident mitigation. Therefore, the licensee revised EOP ES-1.3 to add a step (Step 9) to establish flow through the SW main return valves, thereby increasing flow to the RHR heat exchangers as part of the transfer 'from the injection phase of SI to the containment sump recirculation phase.
- 8 Based on the procedure changes described by the licensee, the NRC staff concludes that appropriate revisions to EOPs have been made in support of the proposed plant modification.
3.2.9 Training Program Design Licensed operator training was revised to incorporate the revised procedure for transitioning to the containment sump recirculation phase. A plant modification was implemented to revise the simulator to match the main control room, including modification of the modeling of SW-1300A{B) and SW-1306A{B) with respect to SI signal inputs, as well as the SI active status panel windows 0404 (SW1306A AND CC6A OPEN) and 0408 (SW1306B AND CC6B OPEN).
Since the plant modification was completed in 2001, the annual licensed operator requalification simulator and cycle exams have resulted in no failures related to meeting the time requirements for this task. Acceptable results were also obtained for the time critical portion of the annual job performance measures that were conducted on the licensed operators.
Based on the licensee's statements that operators have been trained and successfully tested on their ability to transfer to containment sump recirculation with the procedures, instruments, and time now available, the NRC staff concludes that the revisions to the training program to be acceptable.
3.2.10 Human Factors Verification and Validation (V&V)
In order to verify the correctness and validate the effectiveness of the proposed plant modification, all plant Operations crews have been observed performing a transfer to containment sump recirculation on the plant-specific simulator. No failures were observed.
Procedures GNP-05.16.05, "Process Control for Emergency Operating Procedures and EOP Background Documents," and GNP-05.16.06, "Validation of Time Dependent Operator Actions," are processes that the licensee uses to verify and validate the ability of operators to accomplish these critical actions within the time available, and with the instrumentation, training, and procedures available to them.
Based on the existence of controlled processes for V&V of operator manual actions, and the successful testing of relevant members of the Operations staff population, the NRC staff concludes that the DEK approach to Human Factors V&V to be acceptable.
3.2.11 Human Performance Monitoring Strategy Based on the administrative protections afforded by the licensee's procedure control program against inadvertent change, and by the periodic re-validation provided by Just In-Time training prior to each refueling outage, the NRC staff finds the DEK long-term monitoring strategy to be acceptable.
- 9 3.2.12 Summary Based on the statements provided by DEK (Le., that only minor changes to procedures, training, and the human interface design are required to support the proposed license amendment request, and that appropriate administrative controls are being applied to procedures, training, and human interface design to prevent inadvertent changes in the future), the NRC staff concludes that the proposed license amendment request is acceptable from a the human performance perspective.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Wisconsin State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluent that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration, and there has been no public comment on such a finding (76 FR 67487). Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: George Lapinsky, NRR Gerard Purciarello, NRR Date of issuance: March 28, 2012
Mr. David A. Heacock March 28, 2012 President and Chief Nuclear Officer Dominion Energy Kewaunee, Inc.
Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
KEWAUNEE POWER STATION -ISSUANCE OF AMENDMENT RE: LICENSE AMENDMENT REQUEST TO CHANGE THE CURRENT LICENSING BASIS FOR OPERATION OF SERVICE WATER FLOW TO COMPONENT COOLING HEAT EXCHANGERS (TAC NO. ME6288)
Dear Mr. Heacock:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 211 to Renewed Facility Operating License No. DPR-43 for the Kewaunee Power Station (KPS). The amendment revises the KPS current licensing basis in response to your application dated May 9, 2011, as supplemented by letters dated June 3D, and October 31, 2011.
The amendment revises the KPS current licensing basis regarding the manner in which service water is supplied to the component cooling heat exchangers by the main return valves and the bypass flow control valves.
A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely, IRA!
Terry A. Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-305
Enclosures:
- 1. Amendment No. 211 to License No. DPR-43
- 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
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