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Information Notice No. 93-28: Failure to Consider Loss of DC Bus in the Emergency Core Cooling System Evaluation May Lead to Nonconservative Analysis
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555 April 9, 1993 NRC INFORMATION NOTICE 93-28: FAILURE TO CONSIDER LOSS OF DC BUS IN THE EMERGENCY CORE COOLING SYSTEM EVALUATION MAY LEAD TO NONCONSERVATIVE ANALYSIS Addressees All holders of operating licenses or construction permits for nuclear power reactors. Purpose The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert addressees that a single failure of a 125-Vdc bus has been identified as potentially the worst case single failure in the emergency core cooling system (ECCS) evaluation for certain boiling water reactors (BWR-3 and BWR-4). It is expected that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice are not NRC requirements; therefore, no specific action or written response is required. Description of Circumstances On July 30, 1992, the Nebraska Public Power District, licensee for the Cooper Nuclear Station, notified the NRC that the worst case single failure in the ECCS had not been correctly identified in the analysis of the loss-of-coolant accident (LOCA) in conjunction with a loss of offsite power. To satisfy regulatory requirements, the most limiting single failure, which results in the most severe calculated consequences, must be considered in performing the LOCA analysis. The previous licensee analysis assumed that the worst case single failure was a failure of the low pressure coolant injection (LPCI) system injection valve in the ECCS train that is connected to one recirculation loop, concurrent with a pipe break in the other recirculation loop. In July 1992, the licensee recognized that failure of a 125-Vdc bus that provides the control power for the LPCI injection valve serving the unbroken recirculation loop is the worst single failure for this accident. The licensee discovered the problem while performing a plant design basis reconstitution. Discussion The low-pressure trains of the ECCS are shown conceptually in Figures 1 and 2. Trains A and B each include a core spray pump and two residual heat removal (RHR) pumps that function as LPCI pumps during a LOCA. The two RHR pumps in 9304050032. IN 93-28 April 9, 1993 Page 2 of 3 each train share common piping and a common injection valve to the associated recirculation loop. When offsite power is lost, two emergency diesel generators (EDGs) supply power to the pump motors and two 250-V batteries supply motive power to the valve motors. Two 125-V batteries provide control power to the valve motors and the output circuit breakers for the EDGs. In the previous ECCS analysis, as shown in Figure 1, the licensee assumed a guillotine break in one recirculation loop and failure of the LPCI injection valve for the other recirculation loop. If this event were to occur, coolant from RHR pumps B1 and B2 would not reach the reactor vessel because the coolant would flow out through the broken recirculation loop, and RHR pumps A1 and A2 would not pump coolant to the reactor vessel because of the failure of the LPCI injection valve to open. However, it was concluded that both core spray pumps would pump coolant to the reactor vessel and the temperature of fuel cladding would increase but remain less than the regulatory limit. The licensee has now determined that the worst single failure would be the failure of one of the two 125-Vdc buses. As shown in Figure 2, failure of the 125-Vdc bus for train A would prevent closure of the output breaker for EDG A. Motors for RHR pumps A1 and B2 and core spray pump A would not receive emergency power and would not supply coolant to the reactor vessel. Likewise, the LPCI injection valve for recirculation loop A would fail to open and RHR pump A2, which receives power from EDG B, would be unable to pump coolant to the reactor vessel. If the break were in recirculation loop B, then RHR pump B1 would pump coolant to the broken loop, leaving only core spray pump B to pump coolant to the reactor vessel, which is insufficient to perform the intended ECCS function. To avoid the possibility that the temperature of the fuel cladding might exceed the regulatory limit, the licensee reduced the reactor power level pending completion of modifications to correct the problem. The licensee took this action based on analyses by the General Electric Company. Although the combined probability of occurrence of a guillotine rupture, loss of offsite power, and failure of the 125-Vdc bus for the ECCS train serving the intact recirculation loop is very low, the licensee has modified the control power for the LPCI injection valves and recirculation pump discharge valves so that they are powered from the 250-Vdc buses, while leaving the 125-Vdc control power for the EDG output breakers unchanged. Related Generic Communications The General Electric Company submitted a report to the NRC as an attachment to a letter dated November 1, 1978, addressing the effect of a failure of a direct current power supply on BWR-3 and BWR-4 reactors. General Electric concluded that peak cladding temperature would be greater than previously anticipated for a small break LOCA, but less than the regulatory limit, and that the peak clad temperature would be unaffected for large breaks. The report indicated that at least two ECCS pumps would be available. . IN 93-28 April 9, 1993 Page 3 of 3 The NRC sent letters to licensees on April 25, 1980, requesting that addressees confirm the validity of the conclusions made by the General Electric report. The letters asked that responses include lists of ECCS equipment that would be available for breaks in the suction and discharge piping connected to pumps in the recirculation loops. Based on the recent analysis performed at Cooper, the conclusion in the General Electric report that at least two ECCS pumps would remain available may not be true for all scenarios involving direct current power supply failure. The NRC recently issued Information Notice 93-11, "Single Failure Vulnerability of Engineered Safety Features Actuation Systems," to alert licensees to a design deficiency identified at Millstone Nuclear Power Station, Unit 2, that causes a spurious engineered safety feature actuation when one train of DC electrical power is deenergized. In addition, in Generic Letter 89-18, "Systems Interactions in Nuclear Power Plants," the NRC highlighted concerns regarding actuation system designs, including electrical power system designs, that could cause adverse system interactions. This information notice requires no specific action or written response. If you have any question about the information in this notice, please contact one of the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager. ORIGINAL SIGNED BY Brian K. Grimes, Director Division of Operating Reactor Support Office of Nuclear Reactor Regulation Technical contacts: David L. Skeen, NRR (301) 504-1174 Elmo E. Collins, RIV (817) 860-8291 Attachments: 1. Figure 1: Low Pressure ECCS Trains... Figure 2: Break in Recirculation Loop... 2. List of Recently Issued Information Notices .
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