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Information Notice No. 93-45: Degradation of Shutdown Cooling System Performance
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555 June 16, 1993 NRC INFORMATION NOTICE 93-45: DEGRADATION OF SHUTDOWN COOLING SYSTEM PERFORMANCE 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 to degradation of shutdown cooling system performance at the Oyster Creek Nuclear Generating Station resulting from inadequate operating procedures. 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 January 23, 1993, the Oyster Creek Nuclear Generating Station was in day 57 of the 14R refueling outage. The plant was in a normal cold-shutdown condition with the primary containment open and the reactor coolant system at about 43 C [110 F]. One reactor recirculation loop was in service, one recirculation loop was open, and the remaining three recirculation loops were either idled or isolated. Two shutdown cooling loops were operating with a combined flow rate of 11,735 liters per minute [3,100 gpm]. In order to support the planned activities, the licensee elected to secure operation of all recirculation pumps and lower reactor water level. This configuration would allow completion, in parallel, of recirculation pump maintenance activities and main steam isolation valve local leak rate testing. When the operators lowered the reactor water level to approximately 419 cm [165 inches] above the top of active fuel and secured the running recirculation pump, the shutdown cooling water took the least resistance path through the open recirculation loop, and the majority of flow bypassed the core (see Attachment 1). During the time that the planned activities were in progress, plant personnel did not realize that shutdown cooling system performance was degraded, allowing unmonitored heatup of the reactor core. On January 25, 1993, after completing the main steam isolation valve local leak rate testing, an operations engineer discovered that the reactor vessel metal temperature was about 109 C [228 F] at the mid-vessel point. Several 9306090036 . IN 93-45 June 16, 1993 Page 2 of 3 Technical Specifications prerequisites for exceeding 100 C [212 F] were not met, including the requirement to establish primary containment integrity. After discovery, operators took immediate measures to reduce reactor coolant system water temperature by first maximizing flow in the two operating shutdown cooling loops and then placing a third shutdown cooling loop in service. Additionally, the reactor water level was raised to approximately 508 cm [200 inches] above the top of active fuel and a recirculation pump was started. The immediate cause of the event was determined to be plant conditions established by a temporary procedure change to the shutdown cooling operating procedure that failed to provide sufficient forced flow through the reactor core to prevent thermal stratification. The temporary procedure change had been implemented to allow shutdown operations with all recirculation pumps secured and the nominal reactor water level less than the level at which spillover from the core region to the annulus is assured (approximately 470 cm [185 inches] above the top of active fuel). The change had been developed from a draft engineering evaluation, but did not include a requirement (stated in the body of the evaluation) to maintain a specified shutdown cooling flow of 22,712 liters per minute [6,000 gpm]. Instead, the usual shutdown cooling flow of about 11,735 liters per minute [3,100 gpm] was maintained which, due to the specified configuration, allowed the inadvertent reactor heatup. Further details can be found in Licensee Event Report 50-219/93-002 and NRC Augmented Inspection Team Inspection Report No. 50-219/93-80. Discussion This event at Oyster Creek indicated that the temporary procedure change represented a significant procedure revision and that the review process did not identify deficiencies in the procedure before it was implemented. The temporary procedure change had been developed from a draft engineering evaluation which concluded that two shutdown cooling loops in operation would adequately cool the core. An additional assumption in the body of the evaluation was that each loop of shutdown cooling would be operating at its design flow rate of 11,356 liters per minute [3,000 gpm], for a total flow rate of 22,712 liters per minute [6,000 gpm]. This flow rate would be adequate to induce spillover from the core region to the annulus at the reduced nominal water level. The temporary procedure change that implemented the engineering evaluation did not require a minimum flow rate of 22,712 liters per minute [6,000 gpm]. As a result, the operators following the deficient temporary procedure change placed two shutdown cooling loops in operation, but did not provide sufficient shutdown cooling water to the reactor core to maintain reactor vessel temperatures. In addition, the temporary procedure change contained no provision or guidance for monitoring available instruments to ensure that decay heat was being adequately removed from the reactor core. As part of the licensee corrective actions, the licensee reviewed the effectiveness of its safety review process concerning temporary changes and provided training on the subject of this event to all site personnel who perform technical and safety reviews. . IN 93-45 June 16, 1993 Page 3 of 3 This information notice requires no specific action or written response. If you have any questions 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: James S. Stewart, RI (215) 337-5240 Peter C. Wen, NRR (301) 504-2832 Attachments: 1. Figure 1, "Oyster Creek Shutdown Cooling System Flow With Open Recirculation Loop" 2. List of Recently Issued NRC Information Notices .
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