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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56936 | 29 January 2024 17:02:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram | The following information was provided by the licensee via email: At approximately 1202 EST on 01/29/24, unit 2 experienced a reactor scram caused by a main turbine trip. Investigation is still ongoing. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All control rods were fully inserted. The licensee indicated that the turbine trip may have been caused by a power load imbalance, however the cause of the incident is under investigation. The scram was not complex. Decay heat is currently being removed thru bypass valves dumping to the main condenser. Initially unit 2 lost the use of the bypass valves due to lack of condenser vacuum. Unit 2 used the high pressure coolant injection (HPCI) system in the condenser storage tank (CST) to CST mode to remove decay heat. Residual heat removal was used to keep the torus cool. Condenser vacuum was regained and unit 2 is back to removing decay heat with the turbine bypass valves. There was no impact to unit 3. The licensee confirmed there was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The following information was provided by the licensee via email: Licensee adds 8-hour non-emergency 10 CFR 50.72(b)(3)(iv)(A) specified system actuation report to original 4-hour non-emergency 10 CFR 50.72(b)(2)(iv)(B) RPS Actuation report. At approximately 1202 EST on 01/29/24, unit 2 experienced a reactor scram by a main turbine trip. All control rods inserted. Reactor core isolation cooling system (RCIC) was manually initiated for level control. HPCI was manually initiated for pressure control. Primary containment isolation system (PCIS) Group II and III isolations occurred (specified system actuation). Investigation is ongoing. The NRC Resident Inspector has been notified. | High Pressure Coolant Injection Primary Containment Isolation System Reactor Core Isolation Cooling Residual Heat Removal Main Condenser Control Rod | |
| ENS 55899 | 16 May 2022 19:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Scram Due to Electrical Transients | The following information was provided by the licensee via fax: Unit 2 experienced multiple electrical transients resulting in a Group I Primary Containment Isolation Signal (PCIS) isolation and subsequent unit reactor scram. Low reactor water level during the automatic scram caused PCIS Group II and III isolation signals. Following the PCIS Group I isolation, all main steam lines isolated. All control rods inserted and all systems operated as designed. The following additional information was obtained from the licensee via phone in accordance with Headquarters Operations Officers Report Guidance: Peach Bottom Unit 2 automatically scrammed from 100 percent power due to an electrical transient and subsequent PCIS Group I isolation (Main Steam Isolation Valve closure). Unit 2 lost main feedwater due to the PCIS Group I isolation, however, all other systems responded as expected following the scram. High Pressure Coolant Injection is maintaining pressure control while Condensate Pumps are maintaining inventory. The unit is currently stable and in Mode 3. Peach Bottom Unit 3's Adjustable Speed Drives were impacted by the electrical transients and the unit reduced power to 98 percent power. The NRC Resident Inspector was notified. | Feedwater High Pressure Coolant Injection Main Steam Isolation Valve Primary containment Main Steam Line Control Rod | |
| ENS 55575 | 14 November 2021 10:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Trip Due to Lowering Main Condenser Vacuum | At 0525 EST, November 14, 2021, "Unit 2 was manually scammed by operations due to lowering main condenser vacuum. This resulted in PCIS (primary containment Isolation system) Group II/III isolation signals. All control rods inserted, and all systems operated as designed. Unit 3 is unaffected and remains at 100 percent power in Mode 1. The Resident Inspector was notified. | Primary Containment Isolation System Main Condenser Control Rod | |
| ENS 53630 | 30 September 2018 04:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Scram Due to a Loss of Two Condensate Pumps | On Sunday, September 30, 2018, at 1130 EDT, an automatic scram was received on U3 following a loss of two condensate pumps. Following the reactor scram, water level lowered from normal level of 23" to below 1" which resulted in automatic Group II and Group III isolations. Reactor water level lowered to -48" which resulted in initiation of the High Pressure Coolant Injection and Reactor Core Isolation Cooling systems. Reactor water level and reactor pressure have been restored to their normal bands. All systems responded properly to the event. Unit 3 remains in Mode 3 with reactor pressure being controlled on the turbine bypass valves. The cause and details of the event are under investigation. This report is being made in accordance with 10 CFR 50.72(b)(2)(iv)(A), 10 CFR 50.72(b)(2)(iv)(B), and 10 CFR 50.72(b)(3)(iv)(A). All control rods inserted. Decay heat is being removed via the main condenser. The NRC Resident Inspector has been notified. A notification to the media and a press release were made. Unit 2 was unaffected and continues coastdown to refueling. | High Pressure Coolant Injection Reactor Core Isolation Cooling Main Condenser Control Rod | |
| ENS 47286 | 21 September 2011 14:20:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Relay Malfunction Resulted in a Valid Undervoltage Signal | On 9/21/11 at 10:20 a.m. during the performance of a simulated loss of offsite power testing off the E33 4KV emergency bus, an unplanned start of the E3 Emergency Diesel Generator (EDG) occurred due to a valid bus under voltage signal caused by a relay malfunction. The bus being tested was inoperable in support of the test and was not carrying any required safety system loads at the time of the event. The EDG was secured and troubleshooting initiated. The initial determination was the EDG actuation was from an invalid signal but following further review it was determined that the relay malfunction had caused an untimely bus transfer that resulted in a valid 4KV bus under voltage condition. The NRC Resident Inspector has been notified. | Emergency Diesel Generator | 05000277/LER-2011-003 |
| ENS 45348 | 14 September 2009 02:44:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Short Period During Plant Shutdown | During the scheduled shutdown to commence the Peach Bottom Unit 3 refueling outage, Unit 3 was manually shutdown using the mode switch in accordance with GP-3, 'Normal Plant Shutdown,' when reactor period lowered below 50 seconds as indicated on the WRNM (Wide Range Nuclear Monitoring) system. The Feedwater Startup Level Controller was in automatic set at 23" when a small addition of cold water added enough positive reactivity to cause reactor period to be less than 50 seconds. The shortest period observed was 44 seconds. The WRNM system RPS (Reactor Protective System) automatic SCRAM setpoint is 19 seconds. All control rods inserted into the core. Decay heat is being removed by shutdown cooling. The plant is continuing into its scheduled refueling outage. The licensee notified the NRC Resident Inspector. | Feedwater Shutdown Cooling Control Rod | |
| ENS 42889 | 7 October 2006 21:50:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Technical Specification Required Shutdown Due to Loss of Primary Containment Integrity | At 17:50, on 10/07/2006, the Peach Bottom Atomic Power Station identified a crack approximately 4 inches long on a Unit 2, Reactor Core Isolation Cooling (RCIC) test line, as the line penetrates the Suppression Pool of Primary Containment. The degraded piping represents a loss of primary containment integrity, placing one of the principle safety barriers in a 'Seriously Degraded' condition as defined by 10CFR50.72 (b)(3)(ii)(A). This condition required a Reactor Shutdown per the plants Technical Specifications (TS) (TS 3.6.1.1). Unit 2 was manually scrammed, at 20:16, in order to shutdown the reactor and place the unit in Mode 3 per the Technical Specifications. The TS required shutdown is reportable per 10CFR50.72(b)(2)(i). The unplanned reactor scram is reportable per 10CFR50.72(b)(2)(iv)(B). The reactor scram and resultant Emergency Safety Feature actuations were completed as required. In addition, the loss of primary containment integrity represents a condition that 'could have prevented the fulfillment of a safety function of a structure required to control the release of radioactive material' and/or 'mitigate the consequences of an accident.' This is reportable per 10CFR50.72(b)(3)(v)(C) and (D). Unit 2 is currently shutdown, Mode 3, with an RPV cooldown in progress, with plans to Enter Mode 4 by 02:00 on 10/08/06. All control rods fully inserted on the Manual Reactor Scram. The reactor is currently being fed from the condensate system with decay heat being removed to the condenser via the MSL drains. The electric plant is in a normal shutdown lineup. See EN # 42887 for related NOTICE OF UNUSUAL EVENT. Additional 10 CFR Section not listed above: 50.72(b)(3)(v)(D) ACCIDENT MITIGATION The licensee notified the NRC Resident Inspector. | Reactor Core Isolation Cooling Primary containment Control Rod | |
| ENS 41832 | 10 July 2005 07:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Turbine Trip Resulting in Rps Actuation and Automatic Reactor Scram | This non-emergency 4-hour and 8-hour report is pursuant to 10CFR50.72(b)(2) and 10CFR50.72(b)(3). On 7/10/05 at 0318 hours, Unit 2 experienced a turbine trip and subsequent reactor scram. An RPS actuation occurred, as expected, following the turbine trip. RPS responded normally and reactor power is 0%. Reactor water level went below 1 inch, as expected, following a reactor scram from 100% power. At a reactor level of less than 1 inch, PCIS Group 2 and 3 isolations occurred. Reactor water level has been restored above 1 inch to the normal level using the normal reactor feedwater system. All safety systems responded as designed. An investigation is in progress to determine the cause of the turbine trip and to initiate appropriate corrective actions. During the scram the RCS experienced a high pressure condition which caused both recirculation pumps to trip and 3 SRVs to lift and reseat. The licensee is presently depressurizing primary system in order to restart one of the recirculation pumps. Current RCS parameters: Pressure - 320 psig; Temperature - 428 F; and Level - +23 inches. All rods fully inserted. Prior to the turbine trip the licensee was conducting a routine weekly "Mechanical Trip Valve Test. The licensee notified the NRC Resident Inspector. | Feedwater | 05000277/LER-2005-001 05000278/LER-2005-003 |
| ENS 41277 | 22 December 2004 09:55:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Reactor Scram and Eccs Injection Following Opening of Turbine Bypass Valves | At approximately 04:55 on December 22, 2004, Unit 2 experienced a malfunction of Electro-Hydraulic Control (EHC) system resulting in opening of main turbine bypass valves and resultant loss of reactor pressure. The reactor automatically shutdown on RPS with the completion of a Group I isolation signal (Reactor pressure 850 prig and Reactor mode switch in RUN) resulting in a closure of the Main Steam Isolation Valves (MSIVs). Reactor level lowered to (ECCS) initiation set-point of -48 inches and High Pressure Coolant Injection (HPCI) system and Reactor Core Isolation Coolant (RCIC) system automatically initiated and restored level. When reactor level lowered below the 1 inch set-point, Group II and III Primary Containment Isolation System (PCIS) signals initiated. All Unit parameters are stable and RPS/PCIS/ECCS systems performed as designed. MSIVs remain closed. Reactor level and pressure are stable with HPCI and RCIC systems in control. Group I, II, and III isolations have been reset. The EHC malfunction is presently under investigation by Station Management. All systems functioned as required. The reactor water level is now at 23 inches and stable and the licensee is conducting a slow depressurization to Mode 4 to investigate the EHC system malfunction. The level transients experience during the scram would be expected with the closure of the MSIVs. The licensee has notified the NRC Resident Inspector. | High Pressure Coolant Injection Main Steam Isolation Valve Primary Containment Isolation System | |
| ENS 40537 | 22 February 2004 20:10:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Scram at Peach Bottom 2 Due to Decreasing Condenser Vacuum | Peach Bottom Unit 2 reactor was manually scrammed due to degrading main condenser vacuum. The reactor was manually scrammed prior to reaching the automatic scram setpoint. All plant systems responded as expected with no significant issues noted. A Group II and Group III Primary Containment Isolation was received due to reactor water level passing through 1 inch. All isolation systems responded as required and repositioned to their expected positions. The licensee also indicated that all control rods properly inserted into the core. The method of decay heat removal was using the main condenser. The licensee initiated a post scram review to identify and correct the source of degrading vacuum. The licensee also indicated the manual scram was initiated at 25 inches and lowering of condenser vacuum. The licensee notified the NRC Resident Inspector. | Primary containment Decay Heat Removal Main Condenser Control Rod | |
| ENS 40198 | 25 September 2003 04:00:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.73(a)(1), Submit an LER | Initiation of a Technical Specification Required Shutdown at Peach Bottom 2 | On 09/25/03 @ 2210, while restoring reactor vessel instrumentation to service on Unit 3, an invalid ECCS actuation, (reactor vessel lo-lo-lo) occurred which caused all four (4) EDG to automatically start. All eight (8) 4 KV buses remained supplied from offsite sources. The Unit 3 ECCS pump auto starts were previously defeated per plant procedure and were not required to be operable per Tech Specs. The invalid ECCS initiation signal caused the offsite power Loss-of-Power instrumentation setpoints to transfer to the degraded voltage LOCA setpoints. With the degraded voltage LOCA setpoints initiated, the 4 KV E-bus fast transfer capability on a degraded voltage-NON-LOCA condition is defeated. With the inability to fast transfer the 4 KV E-buses on a degraded voltage NON-LOCA condition, both offsite sources are inoperable. Prior to restoring the reactor vessel level instrumentation, the E-2 diesel generator was inoperable. With both offsite source.; and one EDG inoperable, LCO 3.8.1 Required Action H.1 requires an entry into LCO 3.0.3 for Unit 2. With Unit 3 in MODE 5, LCO 3.0.3 is not applicable. After one (1) hour, actions were taken to initiate a plant shutdown per LCO 3.0.3. No negative reactivity was added as the preparations were administrative in nature. At 09/26/03 @ 0006, the invalid ECCS initiation was removed. And all offsite Loss-of-Power instrumentation has been returned to OPERABLE status. Following the automatic start of all four (4) EDGs, E-1, E-3, and E-4 were successfully shutdown per plant procedures. E-2 could not be shutdown per plant procedures and was shutdown locally. The cause of the inability to shutdown E-2 per plant procedures is under investigation. E-2 EDG remains inoperable. The NRC Resident Inspector was notified by the licensee.
This 60-day optional report, as allowed by 10 CFR 50.73(a)(1), is being made under the reporting requirement in 10CFR50.73(a)(2)(iv)(A) to describe an unplanned, invalid actuation of a specified system, specifically the Units 2 and 3 Emergency Diesel Generators (EDGs).' 'On 9/25/03 at 2210 while restoring reactor vessel instrumentation to service on Unit 3, an invalid EDG start actuation signal on Reactor Vessel Lo-Lo-Lo level occurred which caused all four Emergency Diesel Generators to automatically start. Investigation into the cause of the event identified that the sequence of operating instrument valves associated with the instrument rack for level transmitters LT 72A and LT 72C resulted in the invalid reactor vessel lo-lo-lo reactor water level signal and subsequent EDG logic actuation. The EDGs initiated as expected for the given conditions. Off-site power was not affected and continued to supply power to the emergency busses. Once the reason for the EDG initiation was determined, the E-1, E-3, and E-4 EDGs were shut down per plant procedures from the Main Control Room by approximately 2255 hours. The E-2 EDG could not be shut down normally from the Main Control Room control switch. It was later shut down at the local control panel in the E-2 EDG bay at about 0030 hours on 9/26/03. Repairs to the E-2 EDG control switch were completed in accordance with the Corrective Action Program (CR 177605).' 'Notification of the event to the NRC was initially made at 0125 on 9/26/03 (EN # 40198). Since the initial report, it was determined that a Technical Specification LCO 3.0.3 condition did not exist since the off-site sources were operable. This event has been entered into the site-specific corrective action program for resolution (CR 1776 10). The licensee has informed the NRC Resident Inspector. | Emergency Diesel Generator |