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| Entered date | Event description | |
|---|---|---|
| ENS 48525 | 20 November 2012 12:30:00 | Susquehanna Unit 2 discovered a condition that could have prevented the primary containment isolation valves for the reactor water cleanup (RWCU) system from automatically isolating on a high differential flow instrumentation signal. The RWCU system high differential flow signal was found to be indicating downscale due to an instrument failure. Both divisions of the RWCU high differential flow isolation logic utilize the same differential flow instrument loop. Thus, this single instrument failure would have prevented automatic isolation of the RWCU inboard and outboard primary containment isolation valves on a high differential flow signal. The other RWCU primary containment isolation instrumentation functions remained operable and the associated RWCU system primary containment isolation valves were capable of being remotely closed by the control room operators. At the time of discovery, Unit 2 was in Mode 2 due to an unplanned shutdown and all control rods had already been fully inserted as part of a soft shutdown sequence. The licensee has notified the NRC Resident Inspector.
Following the ENS report (EN 48525), Susquehanna determined that although the RWCU high differential flow isolation instrumentation would have prevented automatic isolation of the RWCU inboard and outboard primary containment isolation valves on a high differential flow signal, the RWCU high flow isolation instrumentation would detect a high flow condition and generate an isolation signal that would close the isolation valves. The RWCU high differential flow instrumentation is downstream of the RWCU pumps and it calculates the difference (delta) in flow between the inlet and the outlet of the RWCU heat exchangers. The SSES (Susquehanna Steam Electric Station) Technical Specification (TS) Bases Section 3.3.6.1 states that the RWCU Differential Flow signal is to detect a break in the RWCU system (pipe severance and separation). Engineering analysis determined that the RWCU pumps would run-out if a break occurs downstream of the pumps and the RWCU system flow rate would be approximately 1000 gpm, with one RWCU pump in operation. The flow rate would be higher for two RWCU pumps in operation. Therefore, RWCU isolation would occur from the RWCU Flow - High isolation signal due to a flow rate that is greater than 472 gpm (TS Table 3.3.6.1-1). This high flow isolation does not rely on the RWCU high differential flow instrumentation. The above analysis is consistent with the FSAR discussion in section 7.3.1.1a.2.4.1.9.3. Based on the above, Susquehanna has determined that the RWCU isolation function would still be completed if a pipe break occurred downstream of the RWCU heat exchangers and the RWCU high differential flow instrumentation is inoperable. Since there was no loss of safety function of structures or systems that are needed to control the release of radioactive material, this ENS report is retracted. The licensee will notify the NRC Resident Inspector. |
| ENS 48036 | 19 June 2012 19:48:00 | On June 19, 2012, Unit 1 was shut down due to increasing drywell unidentified leakage. At 1720 EDT on June 19, 2012, the leakage location was discovered to be a welded joint on the 'A' reactor recirculation piping where a 4 inch blank-flanged pipe for chemical decontamination connects to the 28 inch pipe. This constitutes pressure boundary leakage. The location is within the recirculation loop isolation valves on the suction side of the recirculation pump, therefore isolable from the reactor vessel. The reactor was in mode 3 at the time of discovery and LCO 3.4.4 requires entry into mode 4 within 36 hours. As such, the change in modes is required by Technical Specifications, therefore reportable within 4 hours under 10CFR50.72(b)(2)(i). In addition, the pressure boundary leakage at a welded joint is being reported as a degraded condition pursuant to 10CFR50.72(b)(3)(ii)A) and is therefore 8 hour reportable as well. The licensee notified the NRC Resident Inspector. A press release is planned by the licensee. |
| ENS 48092 | 11 July 2012 12:26:00 | At 0928 EDT on May 14, 2012, the 'B' Emergency Service Water (ESW) pump started during testing from the Unit 1 remote shutdown panel (RSP). The likely cause of the start was either human error in performing continuity checks or inadvertent contact with the manual start circuit in the RSP. Based on the likely cause, this was an invalid actuation of a system listed in 10 CFR 50.73(a)(2)(iv). As indicated in 10 CFR 50.73(a)(1), in the case of an invalid actuation reported under 10 CFR 50.73(a)(2)(iv) other than actuation of the reactor protection system (RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER. This 60-day telephone notification is being made to meet the reporting requirements instead of submitting an LER since the actuation was invalid and was not an RPS actuation with the reactor critical. The following additional information is being provided as specified in NUREG-1022: The specific train(s) and system(s) that were actuated: The 'B' ESW pump inadvertently started during testing from the Unit 1 remote shutdown panel. Whether each train actuation was complete or partial: This was a partial actuation (one of four ESW pumps). Whether or not the system started and functioned successfully: The 'B' ESW pump started successfully, operated properly, and continued running until manually secured via normal controls. At the time of the event, the licensee was performing a surveillance where control was shifted from the control room to the remote shutdown panel. The licensee notified the NRC Resident Inspector. |
| ENS 46135 | 28 July 2010 17:45:00 | A licensed operator was determined to have violated the licensee's Fitness for Duty Policy related to self-reporting a legal action. The employee's access to the Protected Area has been revoked. Contact the Headquarters Operations Officer for additional details. |
| ENS 45930 | 15 May 2010 01:17:00 | At approximately 2301 hours EDT on May 14, 2010, Susquehanna Steam Electric Station Unit One reactor scrammed while performing a condensate pump trip test. The reactor operator placed the mode switch in shutdown when reactor water level reached +51 inches and rising. The main turbine tripped due to high reactor water level. All control rods inserted and both reactor recirculation pumps tripped. Reactor water level lowered to -30 inches causing Level 3 (+13 inches) isolations. The Operations crew restored reactor water level to the normal operating band using RCIC (Reactor Core Isolation Cooling) and subsequently the feedwater system. All isolations at this level occurred as expected. No steam relief valves opened. Pressure was controlled via turbine bypass valve operation. All safety systems operated as expected. The reactor is currently stable in Mode 3. An investigation into the cause of the shutdown is underway. Unit Two continued power operation. The NRC Resident Inspectors were notified. A press release will occur. The licensee was performing testing on the digital feedwater control system which was installed during their recent refueling outage when the loss of level control occurred. It appears that the control system did not respond fast enough to control water level. This resulted in the reactor operator inserting a manual scram at +51 inches prior to reaching the reactor automatic scram setpoint of +54 inches for water level. Currently, the plant is removing decay heat via main steam line drains to the condenser. The plant is in its normal shutdown electrical lineup with all safety equipment available. The licensee has notified the Pennsylvania Emergency Management Agency. |
| ENS 45767 | 15 March 2010 12:51:00 | Appendix J Local Leak Rate Testing has determined that Secondary Containment Bypass Leakage (SCBL) has been exceeded for Unit 1. During performance of leak rate test SE-159-045, the combined SCBL limit of 15 scfh (standard cubic feet per hour) for as-found minimum pathway was exceeded as specified in Tech Spec SR 3.6.1.3.11. Acceptance Criteria Test results were within Acceptance Criteria for the 10CFR50 Appendix J limits of 0.6 La (maximum allowed leakage rate). This event is being reported as a degraded or unanalyzed condition pursuant to 10CFR50.72(b)(3)(ii)(A). The RHR system containment spray penetration isolation valve was being tested when the failure occurred. The valve will be repaired and re-tested. The licensee has notified the NRC Resident Inspector. |
| ENS 45534 | 1 December 2009 21:53:00 | The following transmission was made as a voluntary notification at 1800 EST: On 12/01/09 at 1410 (EST), due to a Unit 2 cooling tower make-up supply line failure, a valve vault was flooded and overflowed at several thousands of gallons per minute flow. The local fire company was contacted to provide equipment assistance in pumping out the vault. The river water overflowing the vault entered nearby storm drains and a nearby building housing non-safety related equipment. There are no injuries or an emergency of any kind. The water entering the storm sewer does not constitute a reportable spill. Unit 2 reactor power was reduced to 80% to minimize cooling tower impact. The Pennsylvania Emergency Management Agency was notified at 1800 EST. Various other local and state agencies have been advised of the event. Subsequently, notifications were made to the following agencies: Salem Township Supervisor Chairman State Senator Baker State Rep. Boback Columbia County Commissioner Soberick Federal Affairs to notify US Rep. Kanjorski's staff. Although the impact of this make-up supply line failure to the environment is insignificant (Radiological levels are less than lower limit of detection), its occurrence coupled with subsequent notifications to aforementioned agencies is likely to cause heightened public or government concern. Thus, Susquehanna Steam Electric Station is making a four hour ENS notification pursuant to 10CFR50.72(b)(2)(xi). A press release is not planned at this time. The licensee has notified the NRC Resident Inspector. |
| ENS 43963 | 5 February 2008 21:07:00 | On February 5, 2008 EST at 1845 hours it was discovered that irradiated fuel moves had been performed during the previous shift with both Unit 1 and Unit 2 refuel floor high exhaust radiation monitors bypassed. The condition affected both Susquehanna Units. The radiation monitors are required to be operable for conditions noted in footnotes (a) and (b) in Technical Specification Tables 3.3.6.2-1 and 3.3.7.1-1 (i.e. operations with a potential for draining the reactor vessel, and during CORE ALTERATIONS and during movement of irradiated fuel assemblies in the secondary containment). The function of these instruments is to initiate systems that limit fission product release during and following certain postulated fuel handling accidents and to minimize the consequences of radioactive material in the control room environment. No movement of irradiated fuel assemblies was in progress when the issue was discovered. The event has been determined to be reportable within 8 hours under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D). The radiation monitors were bypassed on 1/31/08, as allowed, during a fuel pool activity NOT involving fuel movement. Approximately one hour of fuel movement occurred during the time the radiation monitors were bypassed. The oncoming shift manager identified the discrepancy during the shift turnover prior to assuming the shift. The licensee has notified the NRC Resident Inspector. |
| ENS 43918 | 18 January 2008 17:54:00 | At 1703 EST on 1/18/2008, Susquehanna LLC personnel became aware that a shipment received from GE Hitachi Nuclear Energy exceeded the allowable limit of 200 mr/hr contact dose rate. The external radiation limit of 200 mrem/hr was exceeded on one of the two boxes comprising the shipment. The limit per NDAP-QA-0648 is 200 mrem/hr on contact for a shipment type for a transport vehicle which is not designated exclusive use. The actual value was determined to be 350 mrem/hr, therefore reportable per the requirements of 10CFR20.1906(d)(2). The NRC Resident Inspector and the Shipper (GE Hitachi Nuclear Energy) were notified. The transport vehicle left Wilmington NC on 1/17/2008 at 1435 and was received by SSES (Susquehanna Steam Electric Station) on 1/18/2008 at 0800. There was no surface contamination noted on the shipment. The original survey completed prior to shipment noted the highest on contact dose rate was 170 mr/hr. This item is reportable under 50.72(b)(2)(xi) for offsite notification of an event of public interest. The boxes contained various pieces of equipment that GE uses to support refueling. The licensee has notified the NRC Resident Inspector and will be notifying the Pennsylvania Emergency Management Agency. |