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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 51958 | 25 May 2016 18:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Discovery of Non-Conforming Conditions During Tornado Hazards Analysis | On May 25, 2016, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Byron Station identified non-conforming conditions in the plant design such that specific TS equipment on both units is considered to not be adequately protected from tornado missiles. Each of the following reportable conditions is postulated by tornado missiles entering openings through the wall that separates the Auxiliary Building and the Turbine Building: The following items could be impacted by tornado missiles entering the 0A and 0B (common) MCR (main control room) turbine building intake openings: Main control room pressurization, main control room ductwork and dampers, chilled water to the VC (control room ventilation) coils, main control room radiation monitors. The following items could be impacted by tornado missiles entering the Division 11 and 21 MEER (miscellaneous electrical equipment rooms) rooms: Exhaust from the MEER (affects room cooling and MCR pressure), exhaust from a battery room, conduits and cabinets associated with the battery chargers and DC bus. The following items could be impacted by tornado missiles entering the Division 12 and 22 MEER rooms: Exhaust from the MEER (affects room cooling and MCR pressure), MEER supply fan and ductwork, battery room exhaust, and the instrument inverter cabinets. The RWST (refueling water storage tank) roof access opening Bilco hatch is fabricated from sheet metal that is not designed to prevent all postulated tornado missiles from entering the tank. The tank pressure boundary is 24" thick concrete and is designed to withstand an external tornado missile impact. Thus a missile that enters the tank will not adversely impact the tank pressure boundary. The following items could be impacted by tornado missiles entering the RWST roof access: The 6" RWST recirculation pipe, 3" overflow pipe, and 24" suction pipe. This piping is located inside the tank and they are approximately 130 degrees around the tank away from the hatch opening. This condition creates a potential LOSF (loss of safety function) with the Byron Essential Service Water Cooling Towers (UHS) (ultimate heat sink) with the discovery that the power and control cables to four of eight cooling tower fans can be damaged by tornado missiles penetrating through wall openings. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v)(D) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. These conditions are being addressed in accordance with EGM 15-002 and DSS-ISG-2016-01 (enforcement discretion and interim guidance documents). The licensee has notified the NRC Resident Inspector. | Service water | |
| ENS 51772 | 7 March 2016 10:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Involving Diesel Driven Auxiliary Feedwater Pump Air Intakes | The Auxiliary Feedwater (AF) system at Byron automatically supplies feedwater to the Steam Generators (SG) to remove decay heat from the Reactor Coolant System following a loss of normal feedwater supply. The AF System consists of a motor driven pump (A) and a diesel driven pump (B) configured into two trains for each unit. Each pump provides 100% of the required AF capacity to the SGs as assumed in the accident analysis. One pump at full flow is sufficient to remove decay heat and cool the unit to Residual Heat Removal (RHR) entry conditions. The diesel driven AF pump is powered from an independent diesel whose combustion air intake is located in the Seismic Category II (non-seismically qualified) Turbine Building but the diesel and pump are located in the Seismic Category I (seismically qualified) Auxiliary Building. During the ongoing NRC Component Design Basis Inspection at the sister Braidwood Station, inspectors asked about the acceptability of the diesel combustion air intake being located in the non-seismic Turbine Building. During the review of available documentation related to the AF diesel engine combustion air intake, it was identified that the documentation did not support operation of the diesel with High Energy Line Break (HELB) environmental conditions in the Turbine Building. This has been reviewed and determined to be applicable to Byron Station Units 1 and 2. Specifically, prior evaluations did not account for air displacement by steam release during the event. After running different models for the Turbine Building HELB, diesel driven AF pump operability was supported for all but the Main Feedwater (FW) HELB. For the FW HELB, the best air density obtained failed to remain above the required levels deemed acceptable for engine operation and remained suppressed for extended periods of time. Additional efforts to qualify the FW piping in the Turbine Building for an Operating Basis Earthquake (OBE) to eliminate this piping from HELB considerations were not successful. This condition applies to both Units 1 and 2 but does not affect the motor driven AF pumps. This event does not constitute a loss of safety function at the point of discovery because the Byron opposite train motor driven AF pumps were operable on both Units 1 and 2. This event is reportable per 10 CFR 50.72(b)(3)(ii)(B) for 'any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' The licensee has notified the NRC Resident Inspector. The licensee entered a 72-hour Action Statement and engineering is analyzing the issue. | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Residual Heat Removal | 05000454/LER-2016-001 |
| ENS 51335 | 20 August 2015 22:55:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Condition That Could Prevent Pressurizer Porv Block Valves from Operating | On 8/20/2015 at 1755 (CDT), a design flaw was discovered with the pressurizer power operated relief valve (PZR PORV) block valve control circuitry. Specifically, the circuit deficiency for which a design basis fire in the Main Control Room (MCR) or cable spreading room could prevent the PZR PORV block valves from being closed from the local control switch at their associated motor control center (MCC). Engineering has reviewed this issue and determined that a potential fire induced ground in the MCR or cable spreading room could clear the associated control power fuses which would prevent the block valves from operating at the local control switch. These valves are considered to form a High/Low pressure interface which requires postulating a proper polarity DC cable to cable fault. Engineering has reviewed the circuit design and cable routing associated with PORVs 1(2)RY455A and 1(2)RY456 and determined that their associated cables are routed with other DC circuit cables in the MCR control board and cable spreading room raceways, such that this postulated fault could potentially cause spurious opening of one of the PORVs even after the control power fuses have been removed as directed by the station abnormal operating procedures for control room inaccessibility. This identified block valve circuit deficiency prevents the credited safe shutdown action of locally closing the block valves to mitigate the spurious operation of a PORV. Hourly fire watches of the affected MCR and cable spreading room fire zones have been implemented. In addition, the MCR is continuously staffed and the affected cable spreading room fire zones are equipped with detection and automatic suppression. This event is being reported under 10CFR50.72(b)(3)(ii)(B) for 'Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. The licensee has notified the NRC Resident Inspector.
During the extent of condition review, an additional design deficiency was identified with respect to the PZR PORV and PZR PORV Block valves control circuitry. Specifically, the current mitigating strategy for removing PZR PORV control power fuses does not adequately prevent a PZR PORV from spuriously opening due to fire induced hot short. Furthermore, local actions to close the associated PZR PORV block valve at the motor control center (MCC) may not be effective because the MCC may not have electrical power during the design basis fire. Therefore, the credited safe shutdown action to remove the PZR PORV control power fuses does not prevent the PZR PORV from spuriously opening during design basis fires in some of the upper and lower cable spreading room fire zones. The affected Fire Zones are the same upper and lower spreading rooms previously identified and fire watches of the affected areas remain in place. The NRC Resident Inspector has been notified. Notified the R3DO (Skokowski). | 05000454/LER-2015-004 | |
| ENS 46708 | 30 March 2011 01:00:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Potential Voiding in Auxiliary Feedwater Alternate Suction Line | The design of the Auxiliary Feedwater (AF) system is for the AF pumps to normally take suction from the condensate storage tank. If the condensate storage tank is not available, the essential service water system provides the alternate supply. Due to the AF system suction piping and valve configuration, a voided section of pipe could exist in the portion that isolates the condensate storage tank supply from the essential service water supply. A preliminary vendor analysis has determined that the void fraction to reach the pump in a dynamic scenario exceeds the acceptance criteria for AF pump operability. Based on past operation in this configuration, the event is being reported as a unanalyzed condition that significantly degrades plant safety and a condition that could have prevented the fulfillment of a safety function (i.e., remove residual heat) under 10CFR50.72(b)(3)(ii) and 10CFR50.72(b)(3)(v). Further review of the void model and pump performance characteristics are planned. In 2011, prior to the completion of this analysis. The void was refilled and verified full for both trains at Byron U1 and U2. Unit 1 is defueled. This condition affects both 'A' and 'B' trains of auxiliary feedwater for both Unit 1 and Unit 2. The NRC Resident Inspector has been notified. | Service water Auxiliary Feedwater | |
| ENS 46416 | 12 November 2010 19:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Inaccurate Information Provided in License Amendment Request | At 1300, on November 12, 2010, Exelon Generation Company LLC concluded that inaccurate information contained in the PRA technical bases for a 1987 License Amendment Request (LAR) for Byron and Braidwood Stations would have potentially impacted the acceptability of the LAR by the NRC. The LAR was to extend Allowed Outage Times (AOT) from 72 hours to 7 days for several systems, to include the Component Cooling (CC) and Residual Heat Removal (RH) Systems. The original design intent of the CC system was that each unit has two independent CC pumps and a fifth pump (U0) CC pump could be used as an operable spare for any of the unit specific pumps. This is how CC was modeled in the PRA technical justification for the 1987 LAR. However, a piping configuration design flaw was recently evaluated in that the U0 CC pump could not be considered an operable spare for either unit's B pumps was not modeled in the PRA. During the evaluation to assess the potential significance of this CC design flaw on the PRA justification for the 1987 LAR, another potentially significant discrepancy was discovered in that it appears the operational practice to always split CC trains after a design basis LOCA was not modeled correctly in the RH analysis. Administrative controls have been put in place to restrict the AOT for the CC pumps and RH trains to the pre-LAR timeframe of 72 hours pending the permanent corrective actions. In addition, administrative controls have been put in place to prohibit the U0 CC pump from being an operable spare for either unit's B trains. This event is being reported as an unanalyzed condition that significantly degrades plant safety under 10 CFR 50.72(b)(3)(ii). The NRC Resident Inspectors have been notified | Residual Heat Removal |