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Potential Loss of Spent Fuel Pool Cooling after a Loss-Of-Coolant Accident or a Loss of Offsite Power
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555 October 7, 1993 Information Notice No. 93-83: POTENTIAL LOSS OF SPENT FUEL POOL COOLING FOLLOWING A LOSS OF COOLANT ACCIDENT (LOCA) Addressees All holders of operating licenses or construction permits for boiling water reactors (BWR). Purpose The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert addressees to an issue the NRC is evaluating concerning the potential loss of spent fuel pool (SFP) cooling following a LOCA. It is expected that recipients will review the information for applicability to their facilities and consider any appropriate actions. However, suggestions contained in this information notice are not NRC requirements; therefore, no specific action or written response is required. Description of Circumstances On November 27, 1992, a 10 CFR Part 21 notification was filed to notify the NRC of concerns regarding the potential effects of a loss of SFP cooling coincident with a LOCA at Susquehanna Steam Electric Station (SSES). Since the initial submittal, additional submittals dated December 14, 1992, and January 2, March 31, August 13, and October 1, 1993, have been made regarding the concerns. In response to these concerns, Pennsylvania Power and Light Company, the licensee for SSES, has made submittals to the NRC dated May 24, July 6, and August 16, 1993. The licensee met with the NRC on March 18 and July 8, 1993. The NRC is currently evaluating the 10 CFR Part 21 notification and subsequent information. Discussion Units 1 and 2 at SSES are BWRs with Mark II containments designed by the General Electric Company. The SFPs for each unit are located above each reactor in a reactor building common area. The two SFPs communicate through a common cask storage pit when the path is not isolated by gates. The SFP cooling systems for Units 1 and 2, as described in the updated final safety 9310070169. IN 93-83 October 7, 1993 Page 2 of 3 analysis report (UFSAR), are non-seismic Category I, non-Class 1E powered, and Quality Group C systems. The SFP cooling system for each unit consists of three parallel heat exchangers, cooled by non-Class 1E service water, and three pumps. During normal operation, the water temperature of the SFP is maintained below 52�C [125�F]. Makeup water to accommodate for evaporation and small leakage losses from the SFP is normally supplied by the condensate transfer system. During refueling outages, the residual heat removal (RHR) system is designed to provide supplemental cooling to the SFP. The RHR system is connected to the SFP by manually operating valves in the reactor building. The RHR system cools the SFP using seismic Category I piping and can be isolated from the non-seismic SFP cooling systems. The seismic Category I emergency service water system also is available to provide makeup water for evaporative losses. This system also requires the operation of manual valves in the pool area. A LOCA coincident with a loss of SFP cooling could potentially limit recovery actions. A LOCA in one unit may restrict access to the reactor building for that unit. The transfer of steam or radioactive materials through the heating, ventilation, and air conditioning systems also may restrict access to the adjacent reactor building. Because entry to the reactor building is necessary to provide a method of SFP cooling or makeup water addition when the normal SFP cooling and make-up systems are inoperable, a delay in accessing the reactor building may result in the SFP water boiling. The submitted information identified the following concerns: Potential loss of normal SFP cooling and makeup water systems as a result of piping stresses caused by LOCA-induced hydrodynamic effects in the reactor buildings. Potential inability to align emergency methods of SFP cooling and makeup water addition under post-LOCA conditions. Potential loss of safety-related equipment as a result of the temperature and steam effects of SFP water boiling within the reactor building. Potential loss of safety-related equipment as a result of flooding from condensation of water vapor created by boiling the SFP water. Adequacy of instrumentation to monitor SFP temperature and level. Acceptability of the source term used to predict accessibility to the SFP area and reactor building. Consideration of SFP heat loads in the design basis for the ultimate heat sink. . IN 93-83 October 7, 1993 Page 3 of 3 Consideration that a loss of offsite power may last longer than 24 hours. The NRC staff is evaluating these concerns and the licensee's actions as they relate to the safe operation of SSES. The NRC staff also is evaluating the safety significance of the concerns and their generic applicability to other BWRs. This information notice requires no specific action or written response. If you have any questions regarding the information in this notice, please contact one of the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation project manager. /S/'D BY AECHAFFEE FOR/ Brian K. Grimes, Director Division of Operating Reactor Support Office of Nuclear Reactor Regulation Technical contacts: David H. Shum, NRR (301) 504-2860 George Hubbard, NRR (301) 504-2870 Attachment: List of Recently Issued Information Notices .
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