ML111510227
| ML111510227 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 05/27/2011 |
| From: | Orf T Plant Licensing Branch II |
| To: | Wasik C Florida Power & Light Co |
| Orf, T J, NRR/DORL/301-415-2788 | |
| References | |
| Download: ML111510227 (3) | |
Text
From: Orf, Tracy Sent: Friday, May 27, 2011 2:46 PM To: 'Wasik, Chris' Cc: Abbott, Liz; 'Frehafer, Ken'
Subject:
St. Lucie Unit 1 EPU - request for additional information (Balance of Plant)
Dear Mr. Wasik,
By letter dated November 22, 2010 (Agencywide Documents Access and Management System Accession No. ML103560415) Florida Power & Light Company (the licensee) submitted a license amendment request for St. Lucie Unit 1.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's submittal and has concluded that additional information is required from the licensee in order for the NRC staff to complete their review. The questions below describe these requests for additional information (RAls).
The NRC requests that the licensee respond to these RAls within 30 days of the date of this e-mail. If the licensee concludes that more than 30 days are required to respond to the RAls, the licensee should request additional time, including a basis for why the extension is needed.
Please contact me at the number below or by e-mail if you have any questions on this issue or if you require additional time to submit your responses.
Sincerely, Tracy J. Orf, Project Manager St. Lucie Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Phone: (301) 415-2788 SBPB-1: 2.5.1.1.1-01, Flood Protection The NRCs acceptance criteria for flood protection are based on General Design Criteria (GDC) -2, which states that Structures, systems and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami and seiches without loss of capability to perform their safety functions. The licensee stated in its licensee amendment request (LAR) that GDC-2 addresses only external flooding analysis. The licensee referenced a submittal of an Individual Plant Examination (IPE) for St. Lucie, Units 1 and 2, to the staff in 1992 in response to the Generic Letter 88-20, Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR 50.54 (f). The IPE submittal addressed the internal flooding analysis for St. Lucie, Units 1 and 2.
In the LAR, the licensee stated that the internal flooding analysis in the St. Lucie IPE identified flood zones for safety-related areas and/or equipment, which could potentially contribute to the overall core damage frequency in the event of flooding. In particular, areas such as auxiliary
feedwater pumps and switchgear rooms were identified as part of the internal flooding analysis.
Although the licensee stated that the extended power uprate (EPU) would not affect the current internal flooding analysis for the safety-related areas as provided in the IPE, the licensee did not provide justification to address how each of the flood zones were assessed according to the IPE to make its determination.
Provide additional justification regarding internal flooding analysis for the flood zones under EPU conditions.
SBPB-2: 2.5.1.1.2-01, Equipment and Floor Drains The NRCs acceptance criteria for the equipment and floor drains (EFDS) are based on GDC-2 and GDC-4, in which they require the EFDS to be designed to withstand the effects of earthquakes and to be compatible with the environmental conditions (flooding) associated with normal operation, maintenance, testing, and postulated accidents (pipe failures and tank ruptures).
The licensee indicated in the LAR that the EPU will not impact the current seismic design of components in the EFDS, nor add any new equipment that would increase quantity of liquids or cause an inadvertent transfer of contaminated fluids to an uncontaminated drainage system.
However, the licensee does not provide further information in the LAR to justify its analysis of the EPU effects on the EFDS.
Provide additional information regarding the impact of the EPU on the EFDS to verify that the effects remain bounded by the current design capability of the EFDS.
SBPB-3: 2.5.1.2.2, Turbine Generator The NRCs acceptance criteria for the turbine generator are based on GDC-4, and relates to protection of systems, structures, and components (SSCs) important to safety from the effects of turbine missiles. The probability of turbine missile generation is minimized by providing a turbine overspeed protection system (with suitable redundancy) to maintain turbine speed within an acceptable range following certain transients. However, the EPU may increase the peak turbine speed by increasing the energy within the turbine system.
The licensee stated in the LAR that the two existing low pressure turbines will be replaced with new Siemens-supplied replacement turbines prior to the implementation of the EPU. The licensee also stated that the overall turbine control system will be upgraded by Westinghouse to improve reliability and maintainability. However, the licensee does not provide information of how the replacement Siemens rotors will comply with the Westinghouse-provided turbine control system to continue to meet its current design analysis for missile protection under EPU conditions.
Provide additional information regarding the compatibility of the new Siemens rotors with the Westinghouse turbine control system to justify its assessment of missile protection under EPU conditions.
SBPB-4: 2.5.4.1-01, Spent Fuel Pool Cooling and Cleanup System The NRCs acceptance criteria for the spent fuel pool cooling and cleanup (SFPCC) system are based on three regulatory factors: (1) GDC-5, which requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions; (2) GDC-44, which requires that a system with the capability to transfer heat loads from safety-related SSCs to a heat sink under both normal operating and accident conditions be provided; and (3) GDC-61, which requires that fuel storage systems be designed with residual heat removal (RHR) capability reflecting the importance to safety of decay heat removal, and measures to prevent a significant loss of fuel storage coolant inventory under accident conditions.
The licensee stated in the LAR that an instantaneous full core offload 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown, the spent fuel pool bulk temperature reached a maximum of approximately 125°F with two fuel pool cooling pumps in operation. The licensee later stated that during EPU conditions, a full core offload 140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> after shutdown would produce a bulk temperature of less than 150°F in the spent fuel pool. The licensee does not provide justification for why the analysis was performed at 140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> for EPU conditions as opposed to 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> for the current design capability of the spent fuel pool.
Provide additional information to justify the EPU analysis of full core offload 140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> after shutdown for the spent fuel pool as opposed to the current analysis of 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown.
SBPB-5: 2.5.4.3-01, Reactor Auxiliary Cooling Water Systems The NRCs acceptance criteria for the reactor auxiliary cooling water system (or component cooling water (CCW) as described for St. Lucie) are based on three regulatory factors: (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation
including flow instabilities and attendant loads (i.e., water hammer), maintenance, testing, and postulated accidents; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions; and (3) GDC-44, insofar as it requires that a system with the capability to transfer heat loads from safety-related SSCs to a heat sink under both normal operating and accident conditions be provided.
The licensee indicated in the LAR that calculations for potential for two-phase flow, void formation, and water hammer after EPU implementation were performed to compare each scenario to current design capability of the CCW. The licensee did not clearly state for each scenario how the EPU analysis was compared to the current design analysis to make the determination that the CCW will handle the water hammer effects of the EPU after a potential CCW restart.
Provide additional information to justify how the calculations for two-phase flow, void formation, and water hammer accounted for the EPU effects to determine that the current design analysis is still applicable after EPU implementation.
SBPB-6: 2.5.4.5-01, Auxiliary Feedwater Section 2.5.4.5.2.1 of Attachment 5 to the LAR and Section 10.5.2 of the St. Lucie Unit 1 FSAR describe that the AFW system consists of one greater than full flow capacity and two full flow capacity auxiliary feedwater pumps. Clarify what the term full capacity means in this context and address whether the definition was changed to support operation at EPU. Also, confirm that the assumed AFW system operation for the various design basis accident analyses and the calculated AFW success criteria for probabilistic analyses of corresponding initiating events was unchanged as a result of assumed change in thermal power supporting EPU. If confirmed, describe the extent that the power increase was accommodated by available margin as opposed to other changes, such as the proposed increase in the AFW initiation setpoint for low steam generator water level.
SBPB-7: 2.12-01, Power Ascension and Testing Plan The NRCs acceptance criteria for the proposed EPU test program are based on 10 CFR 50, Appendix B, Criterion XI, which requires establishment of a test program to demonstrate that SSCs will perform satisfactorily in service.
The licensee indicated in the LAR that equipment modifications are being made to the feedwater system to support EPU conditions and that the feedwater system response would be monitored during power ascension. The licensee also stated that the transient response for the feedwater system during EPU conditions was modeled using the CENTS computer code.
However, the staff is concerned that CENTS-modeled transient for the feedwater system may not reflect the actual transient response of the replacement feedwater pumps, along with its compatibility with the modified feedwater control system, during EPU conditions.
Discuss how the feedwater system will be assessed with actual transient testing prior to EPU implementation to confirm that the replacement feedwater pumps and feedwater control system will respond in a manner similar as the CENTS-modeled transient, or if actual transient testing will not be done, justify why it is not needed.
SBPB-8: 2.5.4.1-02, Spent Fuel Pool Cooling and Cleanup System
The NRCs acceptance criteria for the spent fuel pool cooling and cleanup (SFPCC) system are based on three regulatory factors: (1) GDC-5, which requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions; (2) GDC-44, which requires that a system with the capability to transfer heat loads from safety-related SSCs to a heat sink under both normal operating and accident conditions be provided; and (3) GDC-61, which requires that fuel storage systems be designed with residual heat removal (RHR) capability reflecting the importance to safety of decay heat removal, and measures to prevent a significant loss of fuel storage coolant inventory under accident conditions.
The licensee stated in Section 2.5.4.1 of the LAR that methods similar to what is described in the UFSAR Section 9.1.3.4 were used to perform spent fuel pool cooling analyses at EPU conditions. One of the assumptions described in the LAR for minimizing the heat transfer from the fuel pool is that the ambient air temperature in the vicinity of the fuel pool is assumed to be 110°F to minimize passive heat losses. However, in UFSAR Section 9.1.3.4, the licensee stated that the passive heat losses from the spent fuel pool surface to the Fuel Handling Building air assume the relative humidity of the building air is 100%. The LAR appears to introduce a revision to the spent fuel pool cooling analysis for St. Lucie Unit 1.
The staff requests that the licensee explain in detail how the methods used to perform spent fuel pool cooling analyses were changed. If new methods were introduced, such as modeling heat loss to the structure or environment, the staff also requests that the licensee provide the model and associated benchmarking or validation completed to establish the quality of the model.