Semantic search

Jump to navigation Jump to search
 Start dateReporting criterionTitleEvent descriptionSystemLER
ENS 5480630 July 2020 13:15:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition

EN Revision Imported Date : 8/18/2020 EMERGENCY PROCEDURE ERROR POTENTIALLY PREVENTING TIMELY COMPLETION OF EMERGENCY CORE COOLING SYSTEM RECIRCULATION ALIGNMENT At 0815 CDT on 7/30/2020, it was determined that a procedural error in emergency procedure ES1.3, Transfer to Cold Leg Recirculation, could delay realignment from emergency core cooling system (ECCS) injection phase to recirculation phase under lower plant operational modes. It is noted this scenario is postulated to occur only when the boron dilution mitigation system is operable in lower modes of operation as per Technical Specification 3.3.9 (required operable in Mode 2 (below P-6), 3, 4 and 5). Current plant conditions require this feature nonfunctional so this issue does not impact current plant conditions. This condition is not bounded by existing design and licensing documents; however, it poses no current impact to the health and safety of the public or plant personnel. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 8/17/2020 AT 1603 EDT FROM JOSH COPELAND TO KERBY SCALES * * *

Event Notification (EN) 54806, made on 7/30/2020, is being retracted because re-evaluation performed subsequent to the notification has demonstrated that the error in Emergency Operating Procedure ES1.3 would not have resulted in a condition outside of the current licensing basis analyses of record for the Callaway Plant. This re-evaluation addressed core effects, containment pressure-temperature and radiological consequences analyses, documented in the plant's corrective action program. The re-evaluation has led to the conclusion that the procedural error in ES1.3 would not have prevented any system required to be OPERABLE by the Technical Specifications from performing its specified safety functions. With all systems capable of performing their specified safety functions, the current licensing basis analyses of record for Callaway Plant remain valid and bounding. Based on these considerations, it has been determined that the condition reported in EN 54806 did not result in the plant being in an unanalyzed condition that significantly degraded plant safety. Consequently the condition did not meet the criteria for an 8-hour notification per 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades safety. The NRC Resident Inspector has been notified of this Event Notification retraction. Notified R4DO (Taylor)

ENS 5406112 May 2019 04:05:0010 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
Discovery of a Condition That Could Have Prevented Fulfillment of a Safety FunctionOn 5/11/19, Callaway Energy Center entered Mode 4 at 1217 (CDT). At 2305, the door from the Auxiliary Building to the RAM Storage building was found blocked open. This door is an Auxiliary Building pressure boundary for the Emergency Exhaust system. The Emergency Exhaust system is required in Modes 1,2,3,4, and during movement of irradiated fuel assemblies in the Fuel Building. The door was being blocked open with a large ramp. This rendered the Emergency Exhaust system not capable of performing its design safety function. LCO (Limiting Conditions for Operation) 3.7.13.B was entered, and preparations to move the ramp commenced. LCO 3.7.13.B is for two Emergency Exhaust trains being inoperable due to an inoperable auxiliary building boundary. The allowed outage time is 24 hrs. to restore the boundary to Operable. The door was closed and LCO 3.7.13.B was exited at 0111 on 5/12/19. This event is reportable per 10 CFR 50.72(b)(3)(v) 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 (C) control the release of radioactive material, or (D) mitigate the consequences of an accident. The NRC Senior Resident has been notified.
ENS 5375928 November 2018 06:00:0010 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Discovery of a Condition That Could Have Prevented Fulfillment of a Safety FunctionOn November 28, 2018, while performing an engineering review of the bases for environmental qualification (EQ) requirements for the Atmospheric Steam Dumps (ASDs), it was determined that applicable EQ requirements had not been applied to a key component of each of the ASDs. The result of this issue is that it the availability of the ASDs for a controlled plant cooldown following a postulated steam line break outside containment cannot be assured. Callaway is developing a compensatory action temporary plant modification to install insulation that will protect the affected ASD components from the post Main Steam Line Break temperature. This condition is reportable 10 CFR 50.72(b)(3)(v) 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 (B) remove residual heat, or (D) mitigate the consequences of an accident. The issue places the plant in a 24-hour Technical Specification (TS) Limiting Condition for Operations (LCO), 3.7.4. The licensee has notified the NRC Resident Inspector.
ENS 534853 July 2018 05:00:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionEn Revision Imported Date 8/1/2018

EN Revision Text: DISCOVERY OF AN UNANALYZED CONDITION THAT SIGNIFICANTLY DEGRADES PLANT SAFETY On July 3, 2018, while performing a review of Emergency Operating Procedures, a concern was identified regarding the potential for excessive loss of ultimate heat sink inventory (UHS) through the auxiliary feedwater (AFW) system mini-flow recirculation pathway. This condition would have the potential to prevent the ultimate heat sink from providing an adequate inventory of water for a 30-day mission time.

The normal water supply for the Callaway AFW system is the condensate storage tank (CST). The CST is a non-safety grade component. The safety-grade supply for AFW is the essential service water (ESW) system. The ESW system is supplied by the UHS. The UHS thermal performance analysis accounts for a loss of UHS inventory to the AFW system up until the point of the accident sequence that the AFW pumps would be secured. The analysis did not include an allowance for loss of UHS inventory through the AFW mini-flow recirculation pathway following the AFW pumps being secured. The EOP guidance that secures the AFW pumps does not isolate the mini-flow recirculation pathway.

Initial estimates indicate that loss of UHS inventory through the mini-flow recirculation pathway, if not isolated, would preclude the UHS from completing its 30-day mission time. This potential for depletion of the UHS placed the plant in an unanalyzed condition that significantly degraded safety.

Callaway has issued interim guidance to the on-shift personnel regarding this concern to ensure that the ultimate heat sink water level is maintained at a level that will be adequate to mitigate the potential loss of inventory.

This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades safety. The NRC Resident Inspectors have been notified of this condition.

  • * * RETRACTION ON 07/31/2018 AT 1430 EDT FROM LEE YOUNG TO ANDREW WAUGH * * *

Event Notification (EN) 53485, made on July 3, 2018, is being retracted because re-evaluation performed subsequent to the notification has demonstrated, based on actual plant equipment and environmental conditions, that the unanalyzed inventory losses previously reported by EN 53485 would not have depleted the UHS inventory to an unacceptable level during its 30-day mission time. The re-evaluation has led to the conclusion that the previously unanalyzed losses of UHS inventory would not have prevented the UHS from performing its specified safety functions and meeting its 30-day mission time requirements. With the UHS capable of performing its specified safety functions and meeting its 30-day mission time requirements, the systems supported by the UHS would have remained capable of performing their specified safety functions. Based on these considerations, it has been determined that the condition reported in EN 53485 did not result in the plant being in an unanalyzed condition that significantly degraded safety. Consequently, the condition did not meet the criteria for an 8-hour notification per 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades safety. The NRC Resident Inspector has been notified of the Event Notification retraction. Notified R4DO (Gaddy).

Service water
Auxiliary Feedwater
ENS 533887 May 2018 18:35:0010 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor
10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Discovery of a Condition That Could Have Prevented Fulfillment of a Safety FunctionOn May 7, 2018, during an engineering review of mission time requirements for Technical Specification related equipment, a deficiency was discovered regarding the Emergency Operating Procedure (EOP) guidance for natural circulation cooldown with a stagnant loop. This condition could be the result of a postulated Main Steam Line Break with a loss of offsite power. During a natural circulation cooldown with a faulted steam generator, flow in the stagnant reactor coolant system (RCS) loop associated with the isolated faulted steam generator (SG) could stagnate and result in elevated temperatures in that loop. This becomes an issue when RCS depressurization to residual heat removal system (RHR) entry conditions is attempted. The liquid in the stagnant loop will flash to steam and prevent RCS depressurization. In this condition, the time required to complete the cooldown would be sufficiently long that the nitrogen accumulators associated with Callaway's atmospheric steam dumps and turbine driven auxiliary feedwater pump flow control valves would be exhausted. The atmospheric steam dumps and turbine driven auxiliary feedwater pump would not be capable of performing their specified safety functions of cooling the plant to entry conditions for RHR operation. This issue has been analyzed by Westinghouse in WCAP-16632-P. This WCAP determined that to prevent loop stagnation, the RCS cooldown rate in these conditions should be limited to a rate dependent on the temperature differential present in the active loops. The WCAP analysis was used to support a revision to the generic Emergency Response Guideline (ERG) for ES-0.2 "Natural Circulation Cooldown." Figure 1 in ES-0.2 provides a curve of the maximum allowable cooldown rate as a function of active loop temperature differential which is directly proportional to the level of core decay heat. At the time of discovery of this condition, Callaway's EOP structure did not ensure that the ES-0.2 guidance would be implemented for a natural circulation cooldown with a stagnant loop. Callaway has issued interim guidance to the on-shift personnel regarding this concern and is in the process of revising the applicable EOPs. This condition is reportable per 10 CFR 50.72(b)(3)(v) 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 (A) Shutdown the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) mitigate the consequences of an accident." The licensee notified the NRC Resident Inspector of this condition.Steam Generator
Reactor Coolant System
Auxiliary Feedwater
Residual Heat Removal
ENS 5322320 February 2018 18:25:0010 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor
10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat
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
All Three Auxiliary Feedwater Pumps Inoperable Due to Helb Door Being Open

At 1225 CST, all three Auxiliary Feedwater (AFW) pumps were declared inoperable at the Callaway Plant upon discovery that a door (DSK13311) credited for protection of equipment from the effects of a high-energy line break (HELB) hazard had come partially open due to vibration harmonics from the running turbine-driven Auxiliary Feedwater pump (TDAFP). Immediate investigation identified that play in the mechanism that holds the door closed had rendered it susceptible to movement from the vibration harmonics. The affected HELB door specifically protects safety-related instruments that provide a swap-over signal upon detection of a low suction pressure condition for the AFW pumps and thereby automatically effect a suction transfer for the AFW pumps from the condensate storage tank (normal/standby source) to the Essential Service Water (ESW) system (credited safety-related source). All of the AFW pump suction transfer instrument channels were declared inoperable. Per Technical Specifications (TS) 3.3.2, 'Engineered Safety Feature Actuation System (ESFAS) Instrumentation,' the applicable Condition(s) and Required Action(s) for inoperable AFW pump suction transfer instrumentation only addresses a single channel being inoperable. Thus, the condition of having all three instrument channels inoperable required entry into TS Limiting Condition for Operation (LCO) 3.0.3. At the same time, however, with the automatic suction transfer capability rendered inoperable, all three AFW pumps, i.e., the TDAFP and the 'A' and 'B' motor-driven AFW pumps, were declared inoperable. Although LCO 3.0.3 was applicable, entry into the Required Actions of LCO 3.0.3 was suspended per the Note attached to Required Action E.1 of TS 3.7.5, 'Auxiliary Feedwater (AFW) System,' which states, 'LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.' At 1336 CST, Operations took actions to prevent the TDAFP from running, so the remaining (AFW Pumps) could be returned to Operable status. Operations then declared the affected instrumentation and the 'A' and 'B' motor-driven AFW pumps Operable. This allowed LCO 3.0.3 and Conditions A, B, D, and E under TS 3.7.5 to be exited. With only the TDAFP inoperable, TS 3.7.5 Condition C and its Required Actions remain in effect. Due to the degraded HELB door rendering all three AFW pumps inoperable, the unidentified condition is being reported as an unanalyzed condition that significantly degraded plant safety (per 10 CFR 50.72(b)(3)(ii)(B)) as well as a condition that could have prevented the fulfillment of the safety functions of structures or systems that are needed to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, and mitigate the consequences of an accident (per 10 CFR 50.72(b)(3)(v)(A), (B), and (D), respectively). The NRC Senior Resident Inspector has been notified.

  • * * RETRACTION ON 4/4/2018 at 1109 EDT FROM JONATHAN LAUF TO DAVID AIRD * * *

Event Notification (EN) # 53223, made on 2/20/2018, is being retracted because new information has been obtained that negates the original basis for reporting the unanalyzed condition. Specifically, an evaluation of the HELB that is postulated to occur in the TDAFP room has determined that without crediting door DSK13311 for protection, the affected safety-related instruments would not be exposed to environmental conditions beyond their analyzed capability. This resulted in a conclusion that the unanalyzed condition of door DSK13311 being open did not prevent the affected safety-related instruments or their supported AFW pumps from performing their required safety functions to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, and/or mitigate the consequences of an accident, nor did it significantly degrade plant safety. Consequently, the condition did not meet the criteria for an 8-hour notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(A), (B), or (D). The licensee notified the NRC Resident Inspector. Notified R4DO (Drake).

Service water
Auxiliary Feedwater
ENS 5290515 August 2017 16:46:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionDiscovery of Non-Conforming Conditions During Tornado Hazards AnalysisOn August 15, 2017, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Callaway Plant identified a non-conforming condition in the plant design such that specific Technical Specification equipment is considered not to be adequately protected from tornado missiles. The recirculation lines for all three independent trains of Auxiliary Feedwater (AFW) connect to the Condensate Storage Tank (CST) inside the CST Valve House, which is not a tornado missile-resistant structure. The direct impact by a design basis missile could result in crimping of the recirculation lines, thereby creating the potential to cause damage to the Train A and B Motor-Driven Auxiliary Feedwater Pumps (MDAFPs) and the Turbine-Driven Auxiliary Feedwater Pump (TDAFP) by restricting recirculation flow to less than the design requirements. 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) 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 (A) shut down the reactor and maintain it in a safe shutdown condition, (B) remove residual heat, or (D) mitigate the consequences of an accident. These conditions are being addressed in accordance with NRC's Enforcement Guidance Memorandum EGM 15-002 and Interim Staff Guidance DSS-ISG-2016-01 (enforcement discretion and interim guidance documents). The NRC Resident Inspector has been notified.Auxiliary Feedwater05000483/LER-2017-002
ENS 5260713 March 2017 18:58:0010 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor
10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat
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 March 13, 2017, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Callaway Plant identified non-conforming conditions in the plant design such that specific Technical Specification equipment is considered to not be adequately protected from tornado missiles. The Emergency Fuel Oil Truck Connection Lines for both redundant Emergency Fuel Oil trains extend through the Plant South wall of the Diesel Generator Building structure where they may be exposed to design bases tornado missile impact. The direct impact by the design basis missile could result in damage to the fuel oil transfer lines, thereby preventing delivery of the fuel supply from the Emergency Fuel Oil buried storage tank to the Fuel Oil Day Tank. This condition affects the fuel supply to both supported Emergency Diesel Generator trains. 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) 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 (A) Shut down the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) 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 NRC Resident Inspector has been notified.

  • * * RETRACTION ON 4/18/17 AT 1547 EDT FROM BRANDON LONG TO DONG PARK * * *

Event Notification EN # 52607, made on 03/13/2017, is being retracted because new information has been obtained that negates the originally reported condition. Specifically, subsequent to the Event Notification, an engineering analysis was performed which confirmed that a design basis missile strike on either of the unprotected truck connection lines would not result in damage to the extent that the affected fuel oil transfer lines would be prevented from delivering the fuel supply from the Emergency Fuel Oil buried storage tank to the Fuel Oil Day Tank. (The analysis showed that although bending / deformation of the lines would occur in response to the postulated missile strike, integrity of the lines would remain.) Based on the above, the unanalyzed condition did not prevent the fulfillment of the safety function of structures or systems that are needed to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, or mitigate the consequences of an accident, nor did it significantly degrade plant safety. Consequently, the condition does not meet the criteria for 8-hour notification that are provided in 10 CFR 50.72(b)(3)(ii)(B) or 10 CFR 50.72(b)(3)(v)(A), (B), or (D). As the condition does not require enforcement discretion, the provisions of EGM 15-002 and DSS-ISG-2016-01 (enforcement discretion and interim guidance documents) need not be invoked. However, the immediate and long-term compensatory actions that have been taken following discovery of the condition will remain in place until the condition is fully resolved. The NRC Resident Inspector has been informed of this Event Notification retraction. Notified R4DO (Azua).

Emergency Diesel Generator
ENS 5187420 April 2016 19:05:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionUnanalyzed Condition Due to Essential Service Water Pressure TransientRecent Operating Experience at Callaway has shown that the pressure transient experienced in the Essential Service Water (ESW) system during Engineered Safety Feature Actuation System (ESFAS) testing can result in gasket failure on the Control Room Air Conditioners rendering the units nonfunctional. Previously, this pressure transient was considered to be the result of the system alignment used to perform the surveillance test, which is not the same as the system lineup which would occur on an actual Loss of Offsite Power (LOOP) or Safety Injection Signal Design Basis Accident (DBA) event. However, on April 20th, 2016, Callaway received preliminary analysis results that predict the Control Room Air Conditioners would actually experience a greater pressure transient during a DBA than what is currently experienced during ESFAS testing. This condition could result in the Control Room Air Conditioners not being capable of performing their safety function following a DBA event, and challenge Control Room Habitability. Therefore, this condition meets the reporting criteria of 10 CFR 50.72(b)(3)(ii)(B). Based on current conditions (i.e., the plant is not in Power Operation), this condition does not present an immediate safety concern. The analysis of the pressure transient experienced by the ESW system during a postulated DBA event is preliminary and further evaluation of the analysis is ongoing. The NRC Resident Inspector has been notified.Service water05000483/LER-2016-001
ENS 5062519 November 2014 01:00:0010 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
Unanalyzed Condition Due to All Four Safety Injection Accumulators Inoperable

Following shift turnover from days to nights on 11/18/2014, it was discovered that all (4) of the Safety Injection (SI) Accumulator Outlet Isolation Valve breakers were unlocked and closed. At the time of discovery, 3 of the safety injection accumulator valves were open and 1 was closed for testing. At that time the plant was in MODE 3 at normal operating pressure and temperature. The plant had been performing RCS pressure isolation valve testing prior to shift turnover. The condition was discovered during testing of valves associated with the 'C' safety injection accumulator. After discovery of the condition, Operations directed that the 'A', 'B', and 'D' SI Accumulator Outlet Isolation Valve breakers be opened and locked. This action was completed by approximately 1930 (CST) on 11/18/2014.

The NRC Resident Inspector was notified. The plant entered T.S. 3.0.3 for approximately 30 minutes while restoring the 'A', 'B' and 'D' accumulators to operable (breakers opened and locked with their associated outlet valves open).

ENS 5047419 September 2014 16:34:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionUnanalyzed Condition Due to a Postulated Hot-Short Fire Event That Could Adversely Impact Safe Shutdown EquipmentFrom review of Event Notification 50468 made by Wolf Creek Nuclear Operating Company on 9/18/2014, which in turn was based on review of INPO Event Report 14-33, 'Direct Current Circuits Challenge Appendix R Fire Analysis,' it was determined that portions of the control circuits for the main turbine-generator direct-current (DC) Emergency Lube Oil Pump and the Emergency DC Seal Oil Pump at Callaway Plant are not properly fused to prevent overload and possible secondary fires. The review found that a fire at the motor starter cabinet in the turbine building could cause specific 'smart' hot shorts that could cause overheating of the control cable and result in secondary fires outside the turbine building, including the Control Building, thereby potentially affecting safe shutdown capability for the plant. Based on this information, it has been determined that this condition is unanalyzed, and on a conservative basis, is reportable per 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. As compensatory measures, hourly fire watches are in place in the affected areas of the Turbine Building and Control Building. These compensatory measures, in addition to automatic fire detection and suppression capability in these fire areas, ensure protection of the potentially affected equipment. The NRC Resident Inspector has been notified. The licensee continues to evaluate other control circuits to identify if this condition exists elsewhere.
ENS 494229 October 2013 20:00:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionPostulated Fire Event Could Result in a Hot Short That Could Adversely Impact Safe Shutdown EquipmentA review of industry operating experience regarding the impact of unfused Direct Current (DC) ammeter circuits in the control room has determined that the condition described below to be applicable to Callaway Nuclear Plant resulting in an unanalyzed condition with respect to 10 CFR 50 Appendix R analysis requirements. The original plant wiring design and associated analysis for the Class 1E Train B batteries and chargers (including the B Swing charger) control room ampere indications do not include overcurrent protection features to limit the fault current. In the postulated event, a fire in the control room could cause one of the ammeter wires to hot short to the ground plane; simultaneously, the fire causes another DC wire from the opposite polarity on the same battery to also hot short to the ground plane. This would cause a ground loop through the unprotected ammeter wiring. This event could result in excessive current flow (heating) in the ammeter wiring to the point of causing a secondary fire in the raceway system. The secondary fire could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10CFR50 Appendix R. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. Compensatory measures (fire watches) have been implemented for affected areas of the plant. The NRC Resident Inspector has been notified. Similar Events: EN #49411 and EN #4941905000483/LER-2013-009
ENS 478851 May 2012 18:00:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionFloor Drain Blockage Adversely Affects Asumptions of Pipe Break Analysis for Electrical Switghear Rooms

At 1300 on May 1, 2012, as a result of fire water flushing operations, it was observed that the floor drains in the 'A' and 'B' ESF (Engineered Safety Features) 4160 VAC switchgear rooms were draining extremely slow. Engineering was consulted and it was identified that the floor drains in these rooms are credited with preventing any water accumulation in these rooms as a result of internal flooding due to a pipe break. It is expected that the floor drains in the 'A' ESF switchgear room can drain approximately 134 gallons per minute (gpm) and the floor drains in the 'B' ESF switchgear room can drain approximately 208 gpm. With the floor drains partially blocked, a break in the 'A' Essential Service Water pipe in the 'B' ESF Switchgear Room would result in flood levels in the 'B' ESF Switchgear Room to exceed the maximum levels calculated in the current flooding analysis. The higher flood level may result in the inoperability of 'B' train Electrical Switchgear. The 'A' train Essential Service Water supplied equipment would be adversely affected due to the reduced flow. Consequently the pipe break would result in both ESF trains being adversely affected. Compensatory measures have been taken to restore system operability. The NRC Resident Inspector has been notified.

  • * * RETRACTION FROM KEITH DUNCAN TO JOHN KNOKE AT 1534 ON 05/31/12 * * *

On May 1, 2012, Callaway Plant made an ENS notification in accordance with 10 CFR 50.72(b)(3)(ii)(B) to report the discovery of partially blocked floor drains in the safety-related 4160 V switchgear rooms. At the time of the initial notification, preliminary information indicated that the partially blocked floor drains could have caused a postulated flooding event to adversely affect independent trains of safety-related equipment inside these rooms. Upon further analysis, Callaway Plant staff determined that the pipe break assumed in the flooding calculation of these rooms was overly conservative. Specifically, based on seismic qualifications, the guillotine break of Essential Service Water piping that was originally assumed is not required to be postulated. Instead, a much smaller, through-wall crack of fire protection system piping is the most severe break that must be postulated in the safety-related 4160 V switchgear rooms. An analysis of a postulated flood hazard in these rooms was performed based on the correct water source. Even if considering a complete blockage of the floor drains in these rooms, this analysis demonstrates that a postulated fire protection system piping crack would not have adversely affected safety-related equipment. Based on the results of this analysis, the partially-blocked floor drain condition described in EN 47885 did not meet the criteria for reportability as an unanalyzed condition that significantly degrades plant safety. Event Notification 47885 is hereby retracted. The licensee has notified the NRC Resident Inspector. Notified the R4DO (Greg Pick)

Service water
ENS 4778328 March 2012 20:00:0010 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
Degraded Condition Identified in the B Train Containment Cooler UnitsAt 1500 on March 28, 2012, Callaway Plant personnel discovered that the installation of a modification on the two 'B' train containment cooler units had inadvertently introduced a potential failure mechanism to the 'B' train containment coolers. Specifically, with the containment cooling fans initially in fast-speed operation, combined with certain initial plant conditions, thermal overload tripping of the coolers could occur. In such an event, during some postulated accidents, slow-speed restart of the containment coolers by the Load Shedding and Emergency Load Sequencing system could be prevented. As a result, the 'B' train containment coolers could be rendered unavailable for a portion of a postulated accident. Thus, the safety function of the 'B' train containment cooling fans cannot be assured when this degraded equipment condition is present and the containment cooling fans are run in fast-speed operation. This condition existed for the 'B' train containment cooling units since they were restored to service from maintenance at 0400 on March 15, 2012. Upon identification of this condition, the 'B' train containment cooling fans were switched from fast-speed to slow-speed operation and restored to operable status at 1515. This action precludes this degraded equipment condition from adversely affecting containment cooling fan function during an accident. Concurrent with this condition, the opposite train of containment coolers was removed from service for scheduled maintenance at 0505 on March 27, 2012. As a result, from 0505 on March 27, 2012 until 1515 on March 28, 2012, the safety function of the containment cooling system could not be assured for certain postulated accident conditions. The NRC Resident Inspector has been notified.
ENS 474269 November 2011 23:50:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionPotential Failure of High Density Polyethylene Piping in Esw SystemOn November 9, 2011 at 1715, Callaway Plant staff determined that a postulated design basis fire event in Fire Area C-1, (Control Building, elevation 1974, ESW Pipe Space, Room 3101) could result in failure of the High Density Polyethylene (HDPE) piping in the Essential Service Water (ESW) system. In 2008-2009 timeframe, Callaway Plant implemented a modification which replaced underground large bore carbon steel ESW piping with HDPE piping. Four short sections of this HDPE piping enter the Control Building and interface with steel piping in Room 3101. During the design of the modification, it was not recognized that a fire barrier should be installed to protect the HDPE piping from the consequences of a fire. As a result of the missing fire barrier, a postulated fire could cause a failure of one train of the large bore HDPE piping located within the fire area. The resultant pipe failure could lead to flooding in the fire area that could adversely affect both trains of ESW equipment required to achieve and maintain safe shutdown. An hourly fire watch has been imposed as a compensatory measure for this condition in accordance with the approved fire protection program. This condition is reported in accordance with 10 CFR 50.72 (b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. The NRC Resident Inspector has been notified.Service water
ENS 4708421 July 2011 16:00:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionPotential Adverse Effect on Esw Train During a Postulated Control Room FireOn July 21, 2011 at 1100 (CDT) Callaway Plant staff determined that a design deficiency could adversely affect the 'B' Train of Essential Service Water (ESW) in the event of a Control Room fire. 'B' Train is the credited train for completion of a post-fire safe shutdown as a result of a Control Room evacuation. As a result of this deficiency, normally closed valve EFHV0060 could spuriously open during a postulated control room fire. EFHV0060 is located on the ESW return line from the 'B' Component Cooling Water (CCW) heat exchanger. If EFHV0060 spuriously opened as a result of this postulated fire, the flow balance in the 'B' Train of the ESW system would be affected. In this scenario, cooling water flow to other essential components could be reduced to below the minimum requirements. A fire watch has been imposed as a compensatory measure for this condition. Additionally, EFHV0060 has been closed and de-energized to preclude spurious opening in the event of a postulated control room fire. This condition is reported in accordance with 10 CFR 50.72 (b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. The NRC Resident Inspector has been notified.Service water
ENS 4671531 March 2011 19:32:0010 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
Unanalyzed Condition Identified for Inoperable Rwst Level

In response to a condition identified in late 2010 concerning the control and removal of hazard barriers in the plant, a review of the basis and analysis for high energy line breaks (HELBs) and the barriers for protecting against such events has been underway at Callaway in accordance with the plant's corrective action program. While following up on a question from the NRC Resident Inspector, and as a result of an additional question from the Nuclear Oversight organization at Callaway, it was identified that non-safety piping located in the valve room associated with the Refueling Water Storage Tank (RWST) could potentially (make) all four RWST low water level pressure transmitters inoperable in the event of a malfunction of the non-safety piping concurrent with a design-basis loss-of-coolant accident (LOCA) and/or following a seismic event. The RWST water level transmitters (which are located in the RWST valve room) perform a safety-related function for the emergency core cooling system (ECCS) by automatically swapping suction sources for the ECCS during a LOCA from the RWST to the containment sumps when a low water level condition is reached in the RWST. These instrument channels are required to be OPERABLE in Modes 1, 2, 3 and 4 per Callaway Technical Specification 3.3.2, 'Engineered Safety Feature Actuation System (ESFAS) Instrumentation.' The subject non-safety piping delivers steam supplied by the Auxiliary Steam system to (and from) heaters surrounding the RWST for maintaining RWST contents above the minimum required temperature during winter conditions. The piping passes through the RWST valve room containing the noted RWST water level transmitters which were designed only for a mild environment. It has been identified, however, that the non-safety Auxiliary Steam piping constitutes a high energy line and that its failure could create harsh (hot and wet) conditions in the valve room to which the RWST water level instrumentation was not designed. Per the Callaway FSAR, where non-safety piping interfaces with safety-related piping or systems, the design must be such that failure of the non-safety piping does not adversely affect the safety function(s) of the interfacing safety-related piping or system (since non-safety piping may be assumed to malfunction in conjunction with a design-basis accident). In this case, and based on a conservative interpretation of the FSAR, if the non-safety piping in the RWST valve room is assumed to malfunction (i.e., break), a failure of the RWST instrumentation could occur, thereby preventing the ECCS suction swap over from occurring as required or assumed for LOCA mitigation. This condition required declaring all four RWST water level channels inoperable. In light of recognizing that the RWST water level instruments could be subject to a harsh environment when they were only designed for a mild environment, and could thus fail as a result, this condition represents an unanalyzed condition that significantly degrades plant safety. With regard to the impact on the required ECCS suction swap over function that requires the RWST water level channels to be operable, the inoperability of all four instrument channels is a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to shutdown the reactor and maintain it in a safe condition, remove residual heat, control the release of radioactive material, and mitigate the consequences of an accident. Upon declaring the RWST water level instrument channels inoperable, TS Limiting Condition for Operation (LCO) 3.0.3 was entered at time 1432 CDT on 3/31/2011. At 1634 CDT, the Auxiliary Steam system was isolated and depressurized. This removed the energy that could be released from a break in the non-safety piping, thereby restoring Operability for the RWST water level instruments. The NRC Senior Resident Inspector was notified.

  • * * RETRACTION FROM ADAM SCHNITZ TO HOWIE CROUCH AT 1511 EDT ON 05/26/11 * * *

On March 31, 2011, event notification EN 46715 documented that a harsh environment from a postulated High Energy Line Break (HELB) in the Refueling Water Storage Tank (RWST) valve room could affect RWST level transmitters. These level transmitters provide RWST water level indication in the main control room, which is identified as a safe shutdown function in the Callaway FSAR. They also provide low RWST water level signals for effecting automatic swap over of suction sources for the Emergency Core Cooling System in the event of a loss-of-coolant accident (LOCA). This break may be postulated to occur on non-safety related auxiliary steam lines that run through the RWST valve room and on to the RWST heaters. This condition was initially reported both as an unanalyzed condition that significantly degraded plant safety and as a condition that could have prevented fulfillment of a safety function. When EN 46715 was reported, it was assumed that breaks were required to be postulated at any intermediate fitting, welded attachment, or valve on the subject auxiliary steam lines. Subsequent analysis shows that the sections of auxiliary steam piping in the RWST valve room are able to withstand safe shutdown earthquake (SSE) loadings and rupture loadings. For piping of this qualification, breaks at all intermediate fittings, welded attachments, and valves do not need to be postulated. Instead, line breaks are only required to be assumed at the terminal ends of the lines and at the locations specified for ASME Class 2 and 3 piping. None of these postulated break locations are located inside the RWST valve room, and a postulated auxiliary steam line break outside of the room would not adversely affect the RWST level transmitters. Since none of the postulated break locations are located inside the RWST valve room, there exists reasonable assurance that the RWST level transmitters would have remained capable of performing their safe shutdown function following a postulated break of the subject auxiliary steam lines. Further, there is no adverse effect on the assumed response to a postulated design basis LOCA since a hazard (such as a break in an auxiliary steam line) is not assumed to occur concurrently with the LOCA. Therefore, this condition does not meet the reporting requirements for an unanalyzed condition that significantly degraded plant safety or a condition that could have prevented fulfillment of a safety function. Event notification 46715 is hereby retracted. The NRC Senior Resident Inspector has been notified. Notified R4DO (Haire).

ENS 4669324 March 2011 04:54:0010 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
Pressure Transmitters Needed for Auxiliary Feedwater Suction Path Not Analyzed for Potential High Energy Line Break

While performing an extent of condition review of high energy line break (HELB) analyses, a detailed review of the auxiliary steam system was being performed. During this review, sections of pipe that run through rooms 1206/1207 in the Auxiliary Building were identified that have design ratings indicating that they could possibly be classified as high energy lines. The pipes were verified to have not been considered in the current HELB analyses. This condition affects pressure transmitters ALPT0037, 38, & 39 which are not qualified for operation in a harsh environment. These pressure transmitters provide the Auxiliary Feedwater Pump (AFW) Suction Transfer signal on low suction pressure from the non safety Condensate Storage Tank to the Safety Related supply (Essential Service Water). Technical Specification (TS) 3.3.2-6.h bases state: "since these detectors are in an area not affected by HELBs or high radiation, they will not experience any adverse environmental conditions and the Trip Setpoint reflects only steady state instrument uncertainties." Based upon the above bases, with the identified aux steam lines in service, the pressure transmitter's operability could not be assured. This represented an unanalyzed condition and had the potential to affect equipment used for accident mitigation. TS 3.0.3 was entered at time 2354 (CST) on 3/23/2011. At 0009 (CST) on 3/24/2011, Aux Steam valves FBV0158, FBV0I48, FAV0002, and FAV0003 were isolated, removing the HELB concern (TS 3.0.3 was exited at this time). These are the active feed (isolation valves) to the lines passing through the Aux Building Rooms 1206/1207. The licensee notified the NRC Resident Inspector.

  • * * RETRACTION AT 1525 ON 5/19/2011 FROM DAVID BONVILLIAM TO MARK ABRAMOVITZ * * *

On March 23, 2011, event notification EN 46693 documented that a harsh environment from a postulated High Energy Line Break (HELB) could affect pressure transmitter ALPT0037, 38 and 39. These pressure transmitters provide the Auxiliary Feedwater Pump suction transfer signal on low suction pressure from the Condensate Storage Tank to the safety-related water supply (Essential Service Water). This break was postulated to occur on auxiliary steam lines in Auxiliary Building rooms 1206 And 1207. This condition was initially reported both as an unanalyzed condition that significantly degraded plant safety and as a condition that could have prevented fulfillment of a safety function. When EN 46693 was reported, it was assumed that breaks were required to be postulated at any intermediate fitting, welded attachment, or valve on the subject auxiliary steam lines. Subsequent analysis shows that these sections of auxiliary steam piping are able to withstand safe shutdown earthquake (SSE) loadings and rupture loadings. For piping of this qualification, breaks at all intermediate fittings, welded attachments, and valves do not need to be postulated. Instead, line breaks are only required to be assumed at the terminal ends of the lines and at the locations specified for ASME Class 2 and 3 piping. None of these postulated break locations are located inside rooms 1206 and 1207. Analysis has been performed on these auxiliary steam lines for the remaining break locations that are required to be postulated. This analysis demonstrates reasonable assurance that safety related equipment, including pressure transmitters ALPT0037, 38 and 39, would have performed their safety functions following a postulated break of these auxiliary steam lines. Therefore, this condition does not meet the reporting requirements for an unanalyzed condition that significantly degraded plant safety or a condition that could have prevented fulfillment of a safety function. Event notification 46693 is hereby retracted. The NRC Resident Inspectors have been notified. Notified the R4DO (Shannon).

Service water
Auxiliary Feedwater
ENS 465977 February 2011 15:18:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionHigh Energy Line Break Analysis for Aux Steam to the Auxiliary Building

On December 15, 2009, Callaway Plant reported a condition in which valve FBV0146, an isolation valve on an auxiliary steam line in the Auxiliary Building, was found to be kept normally open. With FBV0146 open, the auxiliary steam line downstream of FBV0146 must be considered a high energy line. This configuration was not consistent with the analysis of record for High Energy Line Break (HELB) events. Valve FBV0146 was closed upon discovery of the condition. This condition was reported under EN # 45571 as an unanalyzed condition that significantly degrades plant safety. EN #45571 was then retracted on January 21, 2010 when analysis showed that no safety-related components would be rendered inoperable in a postulated HELB event due to the condition. Based on this analysis, FBV0146 was reopened. Subsequent review of this condition now shows that, with FCV0146 open, a harsh environment from a postulated HELB downstream of FBV0146 could be transmitted to other areas of the Auxiliary Building. This would occur via a flow path through door gaps and an Auxiliary Building elevator shaft. This flow path had not been considered by the previous analysis. The areas that could be affected by a postulated line break contain safe shutdown equipment (such as equipment for the Component Cooling Water system) that is not assumed to experience harsh conditions. Because of the potential impact on this equipment, this condition is considered to have met the criteria for reporting under 10 CFR 50.72(b)(3)(ii)(B). FBV0146 is now closed. This review was performed as part of the ongoing evaluation of HELB Program deficiencies described in Callaway Plant License Event Report (LER) 2010-009-00. The NRC Resident Inspector has been notified. The auxiliary steam line in the Auxiliary Building feeds non-safety related components.

  • * * UPDATE AT 1519 EDT ON 4/8/11 FROM KEITH DUNCAN TO S. SANDIN * * *

The licensee is retracting this report based on the following: On 02/07/2011, EN #46597 documented that a harsh environment from a postulated High Energy Line Break (HELB) could be transmitted to areas of the Auxiliary Building not qualified for harsh environments. This break was postulated to occur on an Auxiliary Steam line downstream of valve FBV0146 when FBV0146 is open. This condition was initially reported as an unanalyzed condition that significantly degraded plant safety. When EN #46597 was reported, the analysis of the Auxiliary Steam line included postulated break locations at any intermediate fitting, welded attachment, or valve. Subsequent analysis shows that this section of Auxiliary Steam piping is able to withstand safe shutdown earthquake (SSE) loadings and rupture loadings. For piping of this qualification, line breaks are only assumed to occur at terminal ends and at the locations specified for ASME Class 2 and 3 piping. Breaks at intermediate fittings, welded attachments, and valves are not required to be assumed. Postulated breaks of this Auxiliary Steam line at the locations described above have been analyzed. This analysis demonstrates reasonable assurance that safety-related equipment would have performed their safety functions following a postulated break of this Auxiliary Steam line. Therefore, this condition is not an unanalyzed condition that significantly degrades plant safety and does not meet the reporting requirements of 10 CFR 50.72(b)(3)(ii)(B). Event notification #46597, made on 02/07/2011, is hereby retracted. The NRC Resident Inspectors have been notified. Notified R4DO (O'Keefe).

ENS 457475 March 2010 19:25:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionUnanalyzed Condition - Steam Supply Valve Credited as Closed in Fsar Analysis Is Normally Open During Operations

Valve FBV0147, Boric Acid Batch Tank Auxiliary Steam Supply Isolation Valve, is credited with being closed in the Callaway FSAR. This eliminated the need to analyze lines FB-081-HBD and FB-082-HBD for high energy line breaks (HELB). However, FBV0147 was found to be kept normally open to allow steam service for the boric acid batching tank. This is contrary to the normal position assumed in the FSAR and HELB analyses. With valve FBV0147 open, the lines must considered high energy lines. The lines are in the auxiliary building and they traverse rooms containing several components including the flow transmitters for Residual Heat Removal (RHR) to train `A' accumulator injection supply header, RHR train `A' and 'B' SIS hot leg recirculation supply header, and several safety related auxiliary feedwater components. These instruments are used to provide indications during post-accident conditions. This configuration is therefore outside of current HELB analysis and potentially represents a condition that significantly degrades plant safety. The condition was identified to Operations at 0740. Valve FBV0147 was closed at 0810. At 1325 CST, the issue was determined to be reportable. The NRC Resident Inspector has been notified.

  • * * RETRACTION AT 1557 EDT ON 5/3/10 FROM HURT TO HUFFMAN * * *

On 03/05/2010, EN #45747 provided notification that valve FBV0147 was found to be kept normally open to allow steam service for the boric acid hatching tank. This configuration was not consistent with the normal position assumed in the FSAR and HELB analyses. An engineering evaluation was subsequently performed for the auxiliary steam inlet and outlet piping for the boric acid batching tank. This valuation identified four postulated break locations which have all been analyzed. The analyses determined that all safety-related components in the affected rooms would be able to perform their safety functions in the event of a line break. Pipe whip, jet impingement, compartment temperature and over pressurization, flooding, and internal missiles resulting from a postulated line break would not prevent any safety-related equipment from performing its design function. As supported by this evaluation, this condition does not meet the criteria for an unanalyzed condition that significantly degrades plant safety as stated in 10 CFR 50.72(b)(3)(ii)(B). Therefore, the notification made on 03/05/2010 is hereby retracted. The NRC Resident Inspector will be notified. R4DO (Farnholtz) notified.

Auxiliary Feedwater
Residual Heat Removal
ENS 4557115 December 2009 20:30:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionHelb Configuration Calculations Not Consistent with Previous Helb Analysis

During a review of a High-Energy Line Break (HELB) calculation, valve FBV0146 on a 3" Auxiliary Steam line, FB-093-HBD-3", was found to be kept normally open. This is not consistent with HELB analysis, which assumes flow in line FB-093-HBD-3" is restricted by an orifice plate or isolated by FBV0146. This HELB analysis classified FB-093-HBD-3" as a moderate-energy line downstream of FBV0146 as a result of line isolation. No orifice plate was installed, and line FB-093-HBD-3" downstream of the FBV0146 is considered a high-energy line if FBV0146 is open. A room containing line FB-093-HBD-3" downstream from FBV0146 includes several components, the most critical of which are two RCS (Reactor Coolant System) pressure transmitters in one train which are used to provide indications during post-accident conditions. This configuration is therefore outside of current HELB analysis and potentially represents a condition that significantly degrades plant safety. Valve FBV0146 was closed at 1522. The NRC resident inspectors have been notified.

  • * * RETRACTION FROM KEITH CRIBLEZ TO VINCE KLCO ON 1/21/2010 AT 1739 * * *

On 12/15/2009, EN #45571 provided notification that valve FBV0146 on Auxiliary Steam line FB-093-HBD-3 was found to be kept normally open. This configuration was not consistent with High Energy Line Break (HELB) analysis which assumes flow beyond this valve to be restricted or isolated. An engineering evaluation was subsequently performed for the rooms containing this line downstream of FBV0146. This evaluation has determined that all safety-related components in the affected rooms would be able to perform their safety functions in the event of a line break. Pipe whip, jet impingement, compartment over pressurization, flooding, and internal missiles resulting from a postulated line break would not prevent any safety-related equipment from performing its design function. As supported by this evaluation, this condition does not meet the criteria for an unanalyzed condition that significantly degrades plant safety as stated in 10CFR 50.72(b)(3)(ii)(B). Therefore, the notification made on 12/15/2009 is hereby retracted. The (NRC) Resident Inspectors have been notified. Notified the R4DO (Werner).

ENS 4324014 March 2007 19:35:0010 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Two Out of Three Auxiliary Feedwater Pumps Out of Service

A pinhole leak was discovered on B train Essential Service Water (ESW) system piping while preparing the pipe surface for non-destructive examination. Control room personnel were notified of the leak at 1435. B ESW was immediately declared inoperable. At the time of control room notification, surveillance testing on the Turbine Driven Auxiliary Feedwater Pump (TDAFP) was in progress. This surveillance testing made the TDAFP inoperable and non-functional. The surveillance activities were terminated and the TDAFP was returned to operable status at 1438. B ESW is the safety related water source for B train of auxiliary feedwater (AFW). For the three minute period between notification of the pinhole leak until the TDAFP was restored to operable status, there were two auxiliary feed pumps inoperable. This met the conditions for entry into T/S LCO Action 3.7.5.D which requires a plant shut down to Hot Standby within 6 hours. This action was exited when the TDAFP surveillance testing was terminated. Additionally, with 2 of 3 auxiliary feedwater pumps non-functional for 3 minutes, there was a condition which could have prevented fulfillment of a safety function for those 3 minutes. The licensee notified the NRC Resident Inspector.

  • * * UPDATE AT 1555 EDT ON 5/4/07 FROM KEITH DUNCAN TO S. SANDIN * * *

The licensee provided the following information as the basis for retracting this report: On March 14, 2007, (Event Number 43240) Callaway Plant reported a condition that, at the time, was believed to be a condition which could have prevented fulfillment of a safety function. At that time the 'A' motor driven auxiliary feedwater pump (MDAPP) was operable, the turbine driven auxiliary feedwater pump was not functional because of a surveillance test in progress. The 'B' MDAFP was presumed to be non-functional because of a pinhole leak in the 'B' train essential service water (ESW) system piping. 'B' ESW is the safety related water source for the 'B' MDAFP. Subsequent inspection, non-destructive examination, analysis and evaluation of the 'B' train ESW piping determined that the structural integrity of the pipe was retained. 'B' ESW pump was able to provide the required flow to the train. 'B' train ESW was functional with the pinhole leak. With the 'B' ESW train functional, the 'B' train of auxiliary feedwater had its emergency water source. The auxiliary feedwater system would have been able to fulfill its safety function. This event is not reportable per 10CFR50.72(b)(3)(v). The licensee will inform the NRC Resident Inspector. Notified R4DO (O'Keefe).

Service water
Auxiliary Feedwater
ENS 4245430 March 2006 22:50:0010 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an AccidentInadequate Operator Response Time for Component Cooling Water System Realignment During a Large Break Loca

At 1650 on March 30, 2006, a concern was identified where the operators in the training simulator could not complete realignment of the component cooling water (CCW) flow to the residual heat removal (RHR) heat exchanger in a timely manner under certain accident scenarios. This could result in exceeding the maximum design temperature of the CCW system. In addition, assumptions made in the containment pressure and temperature analysis following a large break loss of coolant accident (LOCA) are non-conservative with respect to when CCW flow to the RHR heat exchangers is manually established in accordance with emergency operating procedures. Callaway plant FSAR indicates CCW system flow is manually aligned to the RHR heat exchangers prior to the recirculation phase of emergency core cooling system (ECCS). If the automatic transfer of the RHR pumps to cold leg recirculation, which happens at the Lo-Lo-1 level of the refueling water storage tank (RWST), occurs before CCW flow has been manually aligned to the RHR heat exchanger, containment sump water at temperatures up to 270F can be circulated through the RHR heat exchanger without CCW flow on the other side of the heat exchanger. The CCW side of the heat exchanger would contain stagnant water. This water can heat up quickly with no established flow and exceed the design rated temperature of the system. Recent simulator scenarios of large break LOCAs have shown that the CCW alignment is not reached before the Lo-Lo-1 RWST alarm level is reached. The CCW alignment is completed as part of procedure ES-1.3, Transfer to Cold Leg Recirculation. A review of two large break LOCA scenarios completed on 3-20-06 show that it takes between 1:00 and 1:30 minutes to initiate the step to align CCW to the RHR heat exchangers and takes between 3:00 and 4:30 minutes to complete the alignment. In addition to CCW system temperature concerns, an assumption that CCW flow is established to the RHR heat exchanger prior to reaching the Lo-Lo-1 level in the RWST is made in the containment temperature and pressure response analyses. As a result, a failure to establish CCW flow to the RHR switchover would result in an adverse impact on the inputs used in the Licensing Bases Containment Analysis. However, preliminary sensitivity runs using containment analyses codes indicate that post-peak temperature and pressure are not significantly affected by this issue. Actions taken: 1650 Declared both trains of CCW inoperable. Declared both trains of ECCS inoperable and entered Technical Specification 3.0.3 1710 Both trains of CCW aligned with flow through the RHR heat exchangers 1711 Exited Technical Specification 3.0.3 The licensee notified the NRC Resident Inspector.

  • * * RETRACTION PROVIDED BY GREG BRADLEY TO JEFF ROTTON AT 1747 EDT ON 05/22/06 * * *

The purpose of this notification is to retract a previous notification made on 3/30/06 (EN# 42454). That report was made per 50.72(b)(3)(v)(D) - Accident Mitigation. An engineering evaluation has determined the RHR and CCW systems would have fulfilled their safety functions had they been necessary to respond to an event. Since the safety functions would have been performed there are no applicable reporting criteria under 50.72 or 50.73 and Event Notification 42454 is retracted. The NRC Resident Inspector will be notified. Notified R4DO (Shaffer).

Residual Heat Removal
ENS 4225713 January 2006 14:30:0010 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
Pressurizer Porv Stroke Times Exceed Analysis AssumptionsThe analyses for the Callaway Cold Overpressure Mitigating System (COMS) assumes an opening time and delay time for the pressurizer Power Operated Relief Valves (PORV)s. Evaluations performed by Callaway engineering personnel have determined that PORV stroke times measured during surveillance testing do not account for all of the delay times credited in the Design Bases COMS Analyses of Record. Further reviews determined the allowed delay times could not be met by the control loop. This results in potentially non-conservative PORV lifts settings relative to those specified in the PTLR for satisfying Technical Specification (TS) 3.4.1.2.a and 3.4.12.c. At the time the determination was made, the plant was in Mode 1. Technical Specification 3.4.12, COMS, is applicable in Mode 4 with RCS temperature less than or equal to 275 F, Mode 5 and Mode 6 with the head on the reactor vessel. The TS allows other methods of providing cold overpressure protection (ex. Residual Heat Removal system suction relief valves). Remedial actions have been initiated to ensure the COMS TS function is maintained when required, pending completion of corrective actions. The licensee notified the NRC Resident Inspector.Residual Heat Removal
ENS 422424 January 2006 19:45:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionUnanalyzed Condition Related to Emergency Diesel Fuel Oil Tank Vortex Formation

At 1345, 1/4/06, Callaway Nuclear Plant completed a preliminary evaluation of the potential for vortex formation at the suction of the Emergency Diesel Generator fuel oil transfer pumps. This evaluation was performed in response to industry operating experience concerning an NRC finding identified during a utility inspection. The preliminary conclusion was that the existing Technical Specification level requirements for the Emergency Diesel Generator underground fuel oil storage tanks may be non-conservative and should be increased by an additional 489 gallons. This will result in a six day fuel oil volume requirement of 69,746 gallons and a seven day fuel oil volume requirement of 80,816 gallons. A review of indicated fuel oil storage tank levels for the last three years was conducted and it was determined that there was sufficient fuel oil contained within the underground storage tanks to satisfy the new, higher volume requirements. In those instances where the volume would have been below the new requirements, this condition was already being tracked in the Callaway Plant Equipment Out of Service Log (EOSL) to satisfy associated Technical Specification requirements. It was conservatively decided to institute administrative controls utilizing plant procedural controls and establish new, elevated fuel oil tank level limits which will ensure vortex formation can not occur. Once the evaluation results are finalized, a review of all actions taken to date will be performed to ensure all required actions have been identified and appropriate measures taken to ensure compliance. The licensee notified the NRC Resident Inspector.

  • * * RETRACTION PROVIDED BY SHIFT MANAGER (BRADLEY) TO ROTTON AT 1539 ON 02/16/06 * * *

The evaluation discussed in the original notification has been completed. Administrative minimum limits for fuel oil level established to maintain compliance with the original Technical Specification Emergency Diesel Generator underground fuel oil volume requirements were in fact sufficient to ensure the prevention of vortex formation at the suction of the fuel oil transfer pumps. The fuel oil transfer subsystem remained capable of performing its safety function; therefore, there are no applicable reporting criteria under 50.72 or 50.73 and Event Notification 42242 is retracted. The licensee notified the NRC Resident Inspector. Notified R4DO (Graves).

Emergency Diesel Generator
ENS 4187728 July 2005 02:25:0010 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an AccidentTwo Trains of Control Room Emergency Ventilation Systems (Crevs) Inoperable and Not Restored within 24 Hours

At 2125 on 7/27/05, Door 33021 (B Engineering Safety Feature Switchgear to B Emergency Diesel Generator) (B ESF Switchgear to B EDG) was found not to be latched. Reviewing history; the door was first discovered not to latch at 1215 on 7/26/05, by Security. The Control Room was notified at 2125 on 7/27/05 by an Equipment Operator, who found the door unlatched. Door was subsequently latched closed at 2155 on 7/27/05. Due to this door not being able to be verified latched, T/S LCO 3.7.10.B should have been entered at 1215 on 7/26/05. This renders 2 trains of Control Room Emergency Ventilation Systems (CREVS) inoperable, and if not restored within 24 hours, a plant shutdown is required; being in Mode 3 (Hot Standby) in 6 hours and Mode 5 (Cold Shutdown) in 36 hours. The plant should have been in Mode 3 at 1815 on 7/27/05. This time was not met. As stated previously, the door was verified to be latched at 2155 on 7/27/05. A plant shutdown is not being made due to the LCO 3.7.10.B being satisfied at 2155 on 7/27/05. Door 33021 (B ESF Switchgear to B Emergency Diesel Generator) was repaired at 0022 on 7/28/05. This issue has been entered in the licensee corrective action program. The NRC Resident Inspector was notified of this event by the licensee.

  • * * RETRACTION FROM R. REIDMEYER TO M. RIPLEY 1419 EDT 09/09/05 * * *

Upon further review, it was concluded that this event is not reportable. The design functions of this door are pressure boundary and fire protection. Based upon the following criteria, this event was determined to be not reportable: 1) Pressure boundary: Actual duration of door inoperability did not result in a violation of Control Room Emergency Ventilation System Technical Specification Action completion time limits. 2) Fire protection: Only one fire suppression system was impacted and the inoperability of a fire protection suppression system for a single area is not reportable with regards to the Fire Protection Program. The loss of one fire suppression system was bounded by Callaway licensing basis. The licensee will notify the NRC Resident Inspector. Notified R4 DO (Powers)

Emergency Diesel Generator
Control Room Emergency Ventilation
ENS 4132612 January 2005 19:30:00Other Unspec Reqmnt
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
24 Hour Condition of License Report Regarding Halon System Actuator Port Connection Error

At 1330 on January 12, 2005, station personnel identified an error in connection of pilot lines to the manual-pneumatic actuator on halon bottles required for fire suppression. The vendor was contacted to confirm the configuration. The vendor indicated that the halon bottles would not properly discharge if the pilot lines were not properly connected. The system engineer inspected the halon systems. It was determined that five of six fire areas protected by halon systems were affected. Fire watches were implemented for the affected fire areas. Affected areas: A-27, Load Center/MG set Room, main - correct, reserve - 1 valve correct/1 valve incorrect A-17, South Electrical Penetration Room, main - correct, reserve - incorrect A-18, North Electrical Penetration Room, Main - correct, reserve - correct C-9, ESF Switchgear room 1*, main - incorrect, reserve - incorrect C-10, ESF Switchgear room 2*, main - incorrect, reserve - incorrect C-27, Control room cable trenches/chases**, bottle 1 - correct, bottle 2 - incorrect The main bank is sufficient to suppress a fire in a fire area.

  • One halon system protects both of the fire areas.
    • One halon bottle will provide general area coverage. The second bottle ensures sufficient halon concentration for upper portions of the cable chases in the control room.

The design and licensing basis for the fire protection system does not require consideration of a fire in more than one fire area at a time. No degraded fire barriers between the above fire areas were identified which would have allowed a fire to affect more than one of the fire areas at a time. Repairs were immediately initiated to correct the condition. As of 2010 CST, the repairs have been completed for the affected fire areas and restored to operable status. The licensee notified the NRC Resident Inspector.

  • * * RETRACTION FROM H. BRADLEY TO W. GOTT AT 1225 ON 2/23/05 * * *

Investigation - Informational tests conducted by the Vendor (Chemetron) and witnessed by Wolf Creek, Callaway, and NRC personnel on January 26, 2005 determined that the Halon systems would have properly actuated in the as-found incorrect configuration (port 'A' and 'B' connections reversed). The only identified difference in the actuation sequence between the tests conducted in the incorrect configuration versus the correct configuration is a delay of less than 2 seconds from the time the solenoid received the discharge signal until the first cylinder actuated. There is no regulatory or National Fire Protection Association standard or guideline that places a time requirement on this interval. This very slight time delay would have had no effect on the designed function of the Halon suppression system to extinguish a fire. Additional details are provided in the Chemetron report, 'Report on Actuation Arrangements for Halon Extinguishing System Units,' (Correspondence ULNRC 05-121) that includes the test procedure and results. Halon system function is to establish sufficient halon concentration for sufficient time to suppress a fire. This capability was not lost with the delay in actuation. Regulatory Evaluation - Guidance for reporting to the criterion of 10 CFR 50.73(a)(2)(ii) is provided in section 3.2.4 of NUREG 1022 rev 2, 'Event Reporting Guidelines 10 CFR50.72 and 50.73.' This guidance states that an LER is required for a seriously degraded principal safety barrier or an unanalyzed condition that significantly degrades plant safety. Operating License Condition 2.C(5)(c) states the following: The Operating Corporation shall maintain in effect all provisions of the approved fire protection program as described in the SNUPPS Final Safety Analysis Report for the facility through Revision 15, the Callaway site addendum through Revision 8, and as approved in the SER through Supplement 4, subject to provisions d below. Conclusion: - Based upon the information provided, the Halon suppression system would have operated to extinguish a fire. This condition is not considered reportable to the requirements of 10 CFR 50.72(b)(3)(ii)(B), 10 CFR 50.73(a)(2)(ii), nor is it a violation of the Operating License Condition 2.C(5)(c). Consistent with this conclusion, ENS notification number 41326 for this event is to be retracted. The licensee notified the NRC Resident Inspector.

ENS 403733 December 2003 18:30:0010 CFR 50.72(b)(3)(ii)(B), Unanalyzed ConditionEmergency Procedure Deficiency Causes Unanalyzed Condition

While reviewing operator emergency response times contained in Callaway Plant's Final Safety Analysis Report (FSAR), it was determined that emergency procedure E-0 did not contain specific guidance for actions to be taken when one train of Control Room Emergency Ventilation System (CREVS) failed to properly operate. In FSAR Chapter 15A, the limiting single failure analyzed for the CREVS is the failure of a filtration fan within one train of CREVS. In this accident analysis scenario, a Control Room Filtration Unit fan fails and the train must be secured to prevent inadequately filtered Control Building air from being introduced into the Control Room. If the train is not isolated within 30 minutes, postulated dose to Control Room staff could potentially exceed GDC 19 limits. While procedure E-0 addressed identifying faulted CREVS equipment and an attempted restoration of the faulted equipment, it did not contain sufficient guidance to ensure the Control Room staff would isolate the faulted train of CREVS if the equipment restoration attempt failed. A revision to procedure E-0 has been issued to correct this procedural deficiency. The licensee has notified the NRC Resident Inspector.

  • * * * RETRACTION FROM E. HENSON TO M. RIPLEY 1425 ET 2/2/04 * * * *

This notification is being retracted. Further evaluations concluded that a local area radiation monitor would have alerted the Control Room staff to a developing adverse condition in sufficient time for operators to have identified and isolated the faulted CREVS train prior to exceeding regulatory dose limits. This event does not represent an unanalyzed condition reportable per 10CFR50.72(b)(3)(ii)(B)." The NRC Resident Inspector was notified of this retraction by the licensee. Notified R4 DO (A. Gody)

Control Room Emergency Ventilation