Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 52817 | 21 June 2017 15:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Fire Event Could Adversely Impact Safe Shutdown Equipment | During the review of an electrical circuit coordination calculation to support an ongoing revision of the Fire Safe Shutdown Analysis (FSSA), a lack of appropriate circuit protection coordination was identified in the coordination of electrical protective devices on 118 VAC electrical panels operating in bypass mode of operation. One or more of these electrical panels could be lost for various 10 CFR Appendix R III.G.2 fires outside the Control Room at CPNPP (Comanche Peak Nuclear Power Plant) due to circuit coordination issues. This could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10 CFR 50 Appendix R. Immediate compensatory actions are being taken to establish (or confirm already existing) fire watches in the Fire Areas containing the associate circuits which can potentially jeopardize the FSSA. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. The NRC Resident Inspector has been notified.
On 06/21/2017 Comanche Peak reported an ENS Report (no. 52817) related to the potential loss of 118 VAC electrical panels operating in bypass mode of operation for various 10 CFR (50) Appendix R III.G.2 fires outside the Control Room due to circuit coordination issues. This could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10 CFR 50 Appendix R. Subsequent analysis by Engineering has determined that the affected panels remain coordinated for fire generated faults and can be credited as available by the Fire Safe Shutdown Analysis for a fire outside of the Control Room. Based on the above, the condition described in the ENS report no. 52817 is not considered to be an unanalyzed condition as described in 10 CFR 50.72(b)(3)(ii)(B). The licensee notified the NRC Resident Inspector. Notified R4DO (Azua). | ||
| ENS 52646 | 29 March 2017 00:57:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition for a Postulated Moderate Energy Line Break | On March 28, 2017 at approximately 1957 CDT, a condition was discovered whereby a postulated moderate-energy line break (MELB) involving three fire protection (FP) pipe segments in the Safeguards Building did not contain MELB shielding. It was subsequently determined a postulated crack in one of the affected FP piping sections could adversely affect circuitry associated with the cooling support system for the train A RHR (Residual Heat Removal) pump room, potentially causing the ventilation system to be unavailable to support operation of the train A RHR pump. This condition is not consistent with the CPNPP licensing basis for the protection of essential safe shutdown RHR equipment. At approximately 1957 CDT train A RHR was declared inoperable but available and the unit entered a seventy-two hour LCO (Limiting Condition for Operation) Action Statement per Technical Specification 3.5.2 B pending completion of mitigative actions. Since Unit 1 train B RHR system components and related supporting equipment have been periodically declared inoperable at various times in the last three years for surveillance testing or maintenance, given the MELB condition, both trains of RHR and or support equipment could have been inoperable and this represents an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B). At the time of discovery, train B RHR and support equipment were operable. Therefore, the identified condition is not reportable as a loss of safety function per 10 CFR 50.72(b)(3)(v). The Senior NRC Resident Inspector has been notified. Compensatory actions will include installing a spray shield on the affected cable trays.
On 03/29/2017 Comanche Peak reported an ENS Report (no. 52646) related to the identification of potential moderate-energy line break (MELB) considerations in the Safeguards Building and the potential for adverse interaction with specified Unit 1 electrical equipment. The specific interactions of concern were related to ventilation equipment which would support operation of the Unit 1 A RHR train and several segments of fire protection piping. Subsequent investigations by Engineering have determined: (1) all but one of the suspected potential interactions were determined to not be credible, i.e., the potential MELB would not result in an adverse interaction with the 'target' equipment, and (2) for the remaining potential interaction, an assessment of piping stresses determined there was not a credible MELB source in the affected piping segment and therefore there was not a potential for adverse interaction with the ventilation support equipment. Based on the above, the condition described in ENS report no. 52646 is not considered to be an un-analyzed condition as described in 10 CFR 50.72(b)(3)(ii)(B). The licensee informed the NRC Resident Inspector. Notified the R4DO (Groom). | ||
| ENS 52484 | 11 January 2017 21:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Teflon Found in Containment Spray Pump Components | On September 16, 2016, Comanche Peak reported an unanalyzed condition and potential loss of safety function per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v) related to Teflon (PTFE) installed in the pressure gauge diaphragm seal assemblies for all four of the Centrifugal Charging Pumps and both of the Positive Displacement Charging Pumps on Units 1 and 2 (EN#52244). On November 14, 2016, this event was subsequently retracted. On December 12, 2016, during the ongoing extent of condition review, Teflon was also found to be installed in the suction and discharge pressure gauge diaphragm seal assemblies for the Unit 1 and 2 Containment Spray Pumps. On January 11, 2017 at approximately 1500 CDT, the reportability evaluation determined that reasonable assurance did not exist that the Containment Spray system would have been able to fulfill its design function of removing heat from the containment environment without impacting the applicable dose limits. Teflon (PTFE) is a restricted material normally prohibited from use in contact with reactor coolant or in radiation environments. Teflon (PTFE) is not radiation tolerant and degrades in a radiation environment. The Teflon (PTFE) used in these diaphragm seal assemblies could fail during a postulated Loss of Coolant Accident (LOCA) which could cause the Containment Spray Pumps on Units 1 and 2 to be inoperable, and exceed system leakage limits. This could challenge dose limits and in plant post-accident accessibility. This represents an unanalyzed condition. The pressure gauges and diaphragm seals for all of the Unit 1 and 2 Containment Spray Pumps have been isolated and the Unit 1 and 2 Containment Spray Pumps are operable. The Teflon (PTFE) has likely existed in these diaphragm seals since initial plant licensing. Luminant Power is continuing to investigate the extent of this condition and potential repair techniques. The NRC Resident Inspector has been notified. | Containment Spray | |
| ENS 52244 | 15 September 2016 20:40:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Potential Degradation of Eccs Pump Pressure Indicators | During a review of commercial grade dedication records for a Unit 1 (Emergency Core Cooling System ECCS) Centrifugal Charging Pump discharge pressure gauge, it was identified that the process side of the diaphragm seal utilizes a Teflon (PTFE) gasket. Further review found Teflon (PTFE) to be installed in the pressure gauge seal assembly for all four of the Centrifugal Charging Pumps and both of the Positive Displacement Charging Pumps on Units 1 and 2. Teflon (PTFE) is a restricted material normally prohibited from use in contact with reactor coolant or in radiation environments. Teflon (PTFE) is not radiation tolerant and significantly degrades in a radiation environment. The Teflon (PTFE) used in these pressure gauges could fail during a LOCA (Loss of Coolant Accident) which could cause the (ECCS) Centrifugal Charging Pumps and both of the Positive Displacement Charging Pumps on Units 1 and 2 to be inoperable, and exceed system leakage limits. Excessive leakage from systems which would contain post-LOCA recirculation fluid would challenge onsite and offsite dose estimates and in-plant post-accident accessibility. This represents an unanalyzed condition. Currently, the pressure gauges for all four of the (ECCS) Centrifugal Charging Pumps and both of the Positive Displacement Charging Pumps on Units 1 and 2 have been isolated until this issue can be further evaluated. Luminant Power believes that the Teflon (PTFE) has existed in the pressure gauges since initial plant licensing. Luminant Power is currently investigating the extent of the condition and repair techniques. The NRC Resident Inspector has been notified.
On 09/16/2016, Comanche Peak reported an ENS Report (no. 52244) related to the identification of teflon-containing pressure-seal assemblies installed on the suction and discharge sides of the centrifugal charging pumps and on the suction side of the positive displacement pump. The technical concern was the potential for the teflon-containing assemblies to leak if subjected to post-LOCA recirculation fluid and associated radiation levels. Subsequent investigations by Engineering have determined: (1) the centrifugal charging pumps were operable for all postulated non-LOCA design bases events which required their operation and (2) for postulated LOCA scenarios which would involve radiation levels sufficient as to call into question the ability of the teflon-containing assemblies to maintain system pressure boundary, the ECCS function would be fulfilled in the event one or all of the charging pumps had to be removed from service (due to system leakage) and limiting (control room) doses would have remained below applicable regulatory limits. Based on the above, the condition described in ENS report no. 52244 is not considered to be an un-analyzed condition as described in10 CFR 50.72(b)(3)(ii)(B), nor is it considered to be a condition which could have led to a potential uncontrolled radiation release per 10CFR 50.72(b)(3)(v)(C), nor is it considered to be a condition which could have prevented fulfillment of a safety function under 10 CFR 50.72.(b)(3)(v)(D). The NRC Resident Inspector has been notified. Notified the R4DO (Azua). | ||
| ENS 52239 | 13 September 2016 22:30:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Involving Station Service Water Trains | Based on a walk down in the Service Water Intake Structure (SWIS) with the NRC Resident (Inspector), it was observed that a vertical section of 4 inch Fire Protection pipe that provides a normally pressurized source of fire water supply to the overhead sprinkler system in the SWIS is not Moderate Energy Line Break (MELB) shielded similar to the horizontal segment of the same line near the ceiling. In the event of a MELB crack along any portion of the unshielded pipe, the MELB has a potential impact to the function of any one of the 4 Service Water pumps. Only one train at a time would be affected during the event. This is due to the physical characteristics of the postulated MELB and the configuration/separation relative to the source line and target pumps and/or associated Motor Control Centers (MCCs) that support pump operation. Since the Service Water trains have been periodically declared inoperable at various times in the last three years for surveillance testing or maintenance, if the MELB were to have occurred during these times and affected the opposite train, then two Service Water trains could have been inoperable and this represents an unanalyzed condition. At the time of discovery, all four Service Water trains were operable, therefore, this condition is not reportable as a loss of safety function per 10 CFR 50.72(b)(3)(vi). Currently, Service Water Train B on each Unit has been declared inoperable per Technical Specification (TS) 3.7.8. This condition will be corrected within the 72-hour Completion Time of TS 3.7.8. Currently, Emergency Diesel Generator B on each Unit has been declared inoperable per Technical Specification (TS) 3.8.1. This condition will be corrected within the 72-hour Completion Time of TS 3.8.1. The NRC Resident Inspector was informed.
This is an update to Event Number 52239. On September 13, 2016 at 2228 EDT, Comanche Peak reported an unanalyzed condition involving station service water trains per 10CFR50.72(b)(3)(ii)(B). Specifically, the reported condition involved a vertical section of 4 inch Fire Protection pipe in the SWIS that was not adequately shielded for a Moderate Energy Line Break (MELB). In the event of a MELB crack along any portion of the unshielded pipe, the MELB had a potential impact to the function of any one of the 4 Service Water pumps. On October 6, 2016 at 1410 hours CDT, a section of eyewash station pipe in the Unit 2 Safeguards Building was identified as a result of extent of condition walkdowns that was not adequately shielded for a Moderate Energy Line Break (MELB). In the event of a MELB crack along any portion of this unshielded pipe, the MELB had the potential to impact Unit 2 Train B 480V Motor Control Center (MCC) 2EB2-1. This MCC provides power to Unit 2 Train B Emergency Core Cooling, Battery Charger, Containment Spray, and Containment Isolation Valve equipment. The affected eyewash station pipe was isolated shortly after it was discovered to not be adequately shielded for a MELB. Since 480V MCC 2EB1-1 and the Unit 2 Train A Emergency Core Cooling, Battery Charger, Containment Spray, and Containment Isolation Valve equipment trains have been periodically declared inoperable at various times in the last three years for surveillance testing or maintenance, if the MELB were to have occurred during these times and affected the opposite train, then 2EB1-1, 2EB2-1 and both trains of the Unit 2 Emergency Core Cooling, Battery Charger, Containment Spray, and Containment Isolation Valve equipment could have been inoperable and this represents an unanalyzed condition. At the time of discovery, 2EB1-1 and the Unit 2 Train A Emergency Core Cooling, Battery Charger, Containment Spray, and Containment Isolation Valve equipment was operable. Therefore, this condition is not reportable as a loss of safety function per 10 CFR 50.72(b)(3)(vi). The NRC Resident Inspector was informed. Notified R4DO (Werner).
This is an additional update to Event Number 52239. On September 13, 2016 at 2228 EDT and again on October 6, 2016 at 2009 EDT, Comanche Peak reported unanalyzed conditions involving Station Service Water System trains and a 480V Motor Control Center (MCC) per 10 CFR 50.72(b)(3)(ii)(B). The reported conditions involved sections of piping that were not adequately shielded for a Moderate Energy Line Break (MELB). In the event of a MELB crack along any portion of the unshielded piping, the MELB had a potential impact to the function of safety-related equipment in the Service Water Intake Structure and the Unit 2 Safeguards Building. On October 10, 2016 at 1708 CDT, as a result of ongoing extent of condition walkdowns, a section of fire protection pipe in the Unit 1 Safeguards Building was identified that was not adequately shielded for a MELB. In the event of a MELB crack along any portion of this unshielded pipe, the MELB had the potential to impact Unit 1 Train B Switchgear 1EA2, Unit 1 Train B 480V MCC 1EB4-2, and Unit 1 Train B Distribution Panel 1ED2-2. Only one of these power supplies at a time would be affected. 1EA2 provides 6.9KV electrical power to various Unit 1 Train B safety-related pumps, panels, sequencer, and transformers. 1EB4-2 provides 480V electrical power to various Unit 1 Train B safety-related pumps, valves, fans, panels, and transformers. 1ED2-2 provides 125VDC electrical power to EDG 1-02 channel 1 starting circuit. The affected fire protection pipe was isolated shortly after it was discovered to not be adequately shielded for a MELB. Since Unit 1 Train A Switchgear 1EA1, Unit 1 Train A 480V MCC 1EB3-2, and Unit 1 Train A Distribution Panel 1ED1-2 have been periodically declared inoperable at various times in the last three years for surveillance testing or maintenance, if the MELB were to have occurred during these times and affected the opposite train, then both trains of Unit 1 6.9KV power (1EA2 and 1EA1), both trains of Unit 1 480V power (1EB4-2 and 1EB3-2), and both trains of Unit 1 125VDC power (1ED2-2 and 1ED1-2) along with the safety-related equipment they supply could potentially have been inoperable and this represents an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B). At the time of discovery, none of the affected Train A equipment was inoperable. Therefore, this condition is not reportable as a loss of safety function per 10 CFR 50.72(b)(3)(v). The NRC Resident Inspector was informed. Notified R4DO (Werner).
This is an update to Event Number 52239. On November 17, 2016 at 0730 CST, during ongoing extent of condition walkdowns in the Boric Acid Transfer Pump Area of the Auxiliary Building, two pressurized fire protection pipe segments were identified that did not contain Moderate Energy Line Break (MELB) shielding. In the event of a MELB crack along the unshielded portion of these pipes, the MELB had the potential to impact Unit 1 Train B 480V Motor Control Center (MCC) 1 EB4-1. This MCC provides 480V electrical power to various Unit 1 Train B safety-related pumps, valves, fans, battery chargers, and transformers. At 0743 CST, Technical Specification 3.8.9 Condition A was entered for one AC electrical power distribution subsystem inoperable. At 1021 CST, MCC 1 EB4-1 was declared Operable after MELB shielding was installed on the affected fire protection lines. Since Unit 1 Train A 480V MCC 1 EB3-1 and the associated Unit 1 Train A safety-related pumps, valves, fans, battery chargers, and transformers have been periodically declared inoperable at various times in the last three years for surveillance testing or maintenance, given the MELB condition, 1 EB4-1, 1 EB3-1 and both trains of the Unit 1 safety-related pumps, valves, fans, battery chargers, and transformers they supply could have been inoperable and this represents an unanalyzed condition. At the time of discovery, 1 EB3-1 and the associated Unit 1 Train A safety-related pumps, valves, fans, battery chargers, and transformers were operable. Therefore, this condition is not reportable as a loss of safety function per 10 CFR 50.72(b)(3)(v). The NRC Resident Inspector was informed. Notified R4DO (Azua).
This is an update to Event Number 52239. On December 5, 2016 during ongoing extent of condition walk downs in the Auxiliary Building, pressurized fire protection pipe segments (a flange and a pipe elbow) were identified which did not contain Moderate Energy Line Break (MELB) shielding. In the event of a MELB crack along the un-shielded portion of the pipes, a MELB had the potential to impact Unit 2 Train B 480V Motor Control Center (MCC) 2EB4-1. This MCC provides 480V electrical power to various Unit 2 Train B safety-related pumps, valves, fans, battery chargers, and transformers. At approximately 1355 CST Technical Specification 3.8.9 Condition A was entered for one AC electrical power distribution subsystem inoperable. At 1459 CST, MCC 2EB4-1 was declared Operable after MELB shielding was installed on the affected fire protection line locations. Since Unit 2 Train A 480V MCC 2EB3-1 and the associated Unit 2 Train A safety-related pumps, valves, fans, battery chargers, and transformers have been periodically declared inoperable at various times in the last three years for surveillance testing or maintenance, given the MELB condition, 2EB4-1 , 2EB3-1 and both trains of the Unit 2 safety-related pumps, valves, fans, battery chargers, and transformers they supply could have been inoperable and this represents an un-analyzed condition. At the time of discovery, 2EB3-1 and the associated Unit 2 Train A safety-related pumps, valves, fans, battery chargers, and transformers were operable. Therefore, this condition is not reportable as a loss of safety function per 10CFR 50.72(b)(3)(v). The NRC Resident Inspector was informed. Notified R4DO (Gaddy).
This is an update to Event Number 52239. On December 22, 2016 at approximately 1046 (CST) during ongoing extent of condition walk downs in the common Auxiliary Building (AB) corridor room (X-179), several normally pressurized Waste Processing (WP) pipe segments and one Vent & Drain (VD) segment which are greater than 1" nominal pipe diameter, did not contain MELB shielding. In the event of a MELB crack along the unshielded portion of these pipes, a MELB could have had the potential to impact Unit 1, Train B 480V Motor Control Center (MCC) 1EB4-1. This MCC provides 480V electrical power to various Unit 1 Train B safety-related pumps, valves, fans, battery chargers, and transformers. Prior to the field walkdown, the subject WP and VD line segments were either isolated and depressurized (WP lines) and/or the AB sump discharges realigned (VD) such that the subject lines would pose no threat to the MCC 1EB4-1 if confirmed that shielding is required. As such, the identified condition does not adversely affect operability of 1EB4-1 and entry into a Technical Specification action statement was not required. Field activities continue to install MELB shielding in the affected locations. Since Unit 1 Train A 480V MCC 1EB3-1 and the associated Unit 1 Train A safety-related pumps, valves, fans, battery chargers, and transformers have been periodically declared inoperable at various times in the last three years for surveillance testing or maintenance, given the MELB condition, 1EB4-1, 1EB3-1 and both trains of the Unit 1 safety-related pumps, valves, fans, battery chargers, and transformers they supply could have been inoperable and this represents an unanalyzed condition. At the time of discovery, 1EB3-1 and the associated Unit 1 Train A safety-related pumps, valves, fans, battery chargers, and transformers were operable. Therefore, this condition is not reportable as a loss of safety function per 10CFR 50.72(b)(3)(v). The NRC Resident Inspector was informed. Notified R4DO (Hay). | Service water Emergency Diesel Generator Containment Spray | 05000445/LER-2016-002 |
| ENS 52217 | 1 September 2016 15:25:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition on Turbine Driven Auxiliary Feedwater Pump | During a review of ongoing analyses related to postulated tornado missiles, a question was raised about the sentinel valve on the turbine driven auxiliary feedwater pump (TDAFW). The sentinel valve is designed as a warning system on steam equipment to warn personnel of increased back pressure. The valve is not an ASME component and its operation is not required to support TDAFW operation. The draft analysis predicts the TDAFW exhaust stack could be partially crimped by a tornado missile and the resultant back pressure on the turbine would increase to approximately 40 psi. This is higher than the set point for the sentinel valve (nominally 27 and 29 psi for Units 1 and 2, respectively). Therefore, in a design basis tornado with a design basis tornado missile striking the TDAFW exhaust stack, and in a condition where the TDAFW is demanded to run, the sentinel valve is expected to lift and allow steam to flow into the room. Vendor correspondence indicates that at approximately 40 psi the sentinel valve will conservatively pass 600 lbm/hr. Thus, it is conservatively considered operation of the TDAFW under such conditions would create an adverse steam environment which would be beyond that which the TDAFW pump has been analyzed to operate. Actions planned to alleviate the above condition would eliminate the potential for adverse environmental conditions. The steam supplies to the TDAFW have been isolated to affect repairs, which are expected to be limited to removal of the sentinel valve from each Unit and installation of a plug. Said activities are expected to be completed within the Allowed Out-of-Service Time (AOT) of the TDAFW of seventy-two hours per Technical Specification 3.7.5. NRC Resident Inspector has been informed.
On 09/01/2016 Comanche Peak reported an ENS Report (no. 52217) related to unanalyzed conditions related to the sentinel valve on the turbine driven auxiliary feedwater pump (TDAFW) during postulated tornado-based scenarios and non-tornado based scenarios. The technical concern was the potential for the sentinel valve to release steam into the TDAFW room and result in adverse environmental conditions within the room and potentially external to the TDAFW room for both tornado-based and non-tornado based scenarios. Subsequent investigations by Engineering have determined the sentinel valve would not be demanded to open during tornado based scenarios and would not result in adverse environmental conditions internal or external to the TDAFW room in any design bases scenario. The licensee has notified the NRC Resident Inspector. Notified the R4DO (Farnholtz). | Auxiliary Feedwater | |
| ENS 51751 | 23 February 2016 23:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Involving Normally Open Battery Room Doors | At CPNPP (Comanche Peak Nuclear Power Plant), eyewash stations are located just outside of the Class 1E battery rooms. The battery room doors are normally open and if a MELB (Moderate Energy Line Break) occurred on the demineralized water line connected to the eyewash station, the water could potentially spray onto the Class 1E safety related batteries. If this occurred, an electrical short could potentially cause a loss of both the batteries and the associated battery chargers. This condition has been conservatively determined to be reportable per 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. Currently, the demineralized water lines on the battery room eyewash stations for both Units 1 and 2 have been isolated, therefore, all safety related equipment is currently operable. Comanche Peak Engineering is performing a past operability review of this condition. The NRC Resident Inspector has been notified.
On February 23, 2016 at 2045 (EST), Comanche Peak reported an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B). Specifically, the reported condition involved eyewash stations that are located just outside of the Class 1E battery rooms. The battery room doors are normally open and if a Moderate Energy Line Break (MELB) occurred on the demineralized water line connected to the eyewash station, the water could potentially spray onto the Class 1E safety related batteries. If this occurred, an electrical short could have potentially caused a loss of both the batteries and the associated battery chargers. This condition was conservatively determined to be reportable per 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. The demineralized water lines on the battery room eyewash stations for both Units 1 and 2 were isolated, and Comanche Peak Engineering initiated a past operability evaluation of this condition. The past operability evaluation has been completed and shows that there are no operability concerns regarding a MELB impact on the Class 1E batteries, DC bus or Class 1E battery chargers. Therefore, Comanche Peak requests that the February 23, 2016, 10 CFR 50.72(b)(3)(ii)(B) reportable event for Units 1 & 2 be retracted. The NRC Resident Inspector has been notified. Notified the R4DO (Whitten). | ||
| ENS 50835 | 19 February 2015 22:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Involving Normally Closed Watertight Doors | During Main Steam Safety Valve testing conducted prior to refueling outages, normally closed watertight doors are opened in support of the testing. If a postulated one square foot non-mechanistic crack were to occur within the Break Exclusion Area during the test, safety related equipment located just outside of these doors could be adversely affected. With these watertight doors open, compliance with the Comanche Peak licensing basis may not be assured. This condition has been conservatively determined to be reportable as an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B). Currently, the watertight doors on both Units 1 and 2 are closed, therefore, all safety related equipment is currently operable. Comanche Peak Engineering is performing a review of the original Comanche Peak licensing basis regarding the non-mechanistic crack event to determine what equipment impacts are required to be assessed. The NRC Resident Inspector has been notified. | Main Steam Safety Valve | |
| ENS 50043 | 18 April 2014 17:52:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Cable Routing Unanalyzed for Fire Safe Shutdown Barrier | While implementing a multiple spurious operations (MSO's) modification during a refueling outage, it was identified that a fire safe shutdown cable routing location may have been in question. On 04/18/2014 at 1252 CDT, it was determined that this cable was routed through a cable tray that was not designed to have a fire safe shutdown barrier. This created an unanalyzed condition that significantly degrades plant safety, consistent with NUREG 1022, Rev. 3 guidance. This cable is associated with a Motor Operated Valve (MOV) used to isolate the containment sump from the Residual Heat Removal (RHR) pump suction during normal system operations. Compensatory measures in this area were already in effect for the resolution of MSO's scenarios. The compensatory measures assure the Systems, Structures, and Components (SSC's) associated with this cable remain operable. The licensee has notified the NRC Resident Inspector. | Residual Heat Removal | |
| ENS 49419 | 8 October 2013 17:40:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Hot Short Fire Event That Could Adversely Impact Safe Shutdown Equipment | A review of industry operating experience regarding the impact of unfused Direct Current (DC) ammeter circuits in the control room has determined the described condition to be applicable to Comanche Peak Nuclear Power Plant resulting in a potentially unanalyzed condition with respect to 10CFR50 Appendix R analysis requirements. The original plant wiring design and associated analysis for the Class 1 E batteries 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 could 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 reportable in accordance with 10CFR50.72(b)(3)(ii)(B) as an unanalyzed condition. Compensatory measures (fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Resident Inspector. See related Event #49411. | 05000395/LER-2013-005 05000445/LER-2013-002 | |
| ENS 47713 | 1 March 2012 23:05:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Battery Room Ventilation System Degraded | While Units 1 and 2 were in Mode 1, operating at 100% power, an issue was identified with the doors for the safety related battery rooms and their normal position. Several doors to the battery rooms were found to be held open via electromagnetic door devices. These doors are an integral part of their battery room exhaust system. As designed, there is no uninterruptible power to the door mechanisms, all the doors are expected to close in the event of a Loss of Offsite Power. The closure of the doors to the battery rooms, being an integral part of the exhaust system, would disrupt ventilation in the battery rooms and could allow hydrogen to increase to unacceptable levels during Loss of Offsite Power events. Compensatory measures have been taken to secure the doors open to maintain the hydrogen purging function before required limits are reached. The battery room ventilation system remains functional supporting operability of the batteries. Luminant Power determined this issue to be reportable as an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B). The licensee will notify the NRC Resident Inspector. | ||
| ENS 47357 | 20 October 2011 01:02:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Due to Pressure Instruments Containing Aluminum Material | With Unit 1 in Mode 4 and Unit 2 in Mode 1 an issue was identified with the body material of existing installed pressure instruments for both the Personnel and Emergency Airlocks of both units. The pressure instruments were determined to have an aluminum body which is not suited for safety related use in containment. Aluminum is a restricted/limited material in containment because it is not compatible with accident conditions and has failures with multiple adverse effects. Due to this condition, the pressure instruments would potentially lose pressure integrity during a LOCA with containment spray actuation. These pressure instruments are located inside containment and are connected to tubing that penetrates the airlock barrel. In event of a failure of any pressure instrument the integrity of the airlock would be compromised. The containment air locks form part of the containment pressure boundary and, as such, a loss of pressure boundary integrity would no longer meet general design criteria. Compensatory measures have been taken to prevent a failure of the airlock integrity due to containment spray actuation and at this time the airlock is operable. Luminant power determined this to be reportable at 2002 on 10/19/11 per 50.72(b)(3)(ii)(B) Comanche Peak Units 1 and 2 being in an unanalyzed condition that significantly degrades plant safety. The licensee notified the NRC Resident Inspector. | Containment Spray | 05000445/LER-2011-003 |
| ENS 47352 | 18 October 2011 20:45:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Aluminum Valve Discovered Installed in Containment | While Unit 1 was in Mode 5, for the 15th refueling outage, an issue was identified with the valve material of an existing installed airlock hydraulic system valve. The valve was determined to have an aluminum body which is not suitable for safety related use in containment. The Unit 1 airlock hydraulic system penetrates the containment pressure boundary. The airlock hydraulic system achieves containment integrity by being a closed system under GDC-57. A loss of pressure boundary integrity would no longer meet General Design Criteria 57 (GDC-57) for a closed system. Aluminum is a restricted/limited material in containment because it is not compatible with accident conditions and has failures with multiple adverse effects. Due to this condition, the valve would potentially lose pressure integrity during a LOCA with containment spray actuation. Compensatory measures have been taken to prevent containment spray from affecting this valve and at this time, the airlock is operable. Luminant Power determined this issue to be reportable at 1545 (CDT) on 10/18 per 50.72(b)(3)(ii)(B) Comanche peak Unit 1 being in an unanalyzed condition that significantly degrades plant safety. The Unit 2 airlock is a different design and this condition does not apply to Unit 2. This material was installed during original construction and discovered during a licensee self-assessment. The licensee will notify the NRC Resident Inspector. | Containment Spray | 05000445/LER-2011-003 |
| ENS 47086 | 21 July 2011 16:53:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Vulnerability from a Potential Control Room Fire on "a" Safeguards Bus | A potential scenario has been identified that has not been analyzed in the Comanche Peak Nuclear Power Plant Fire Safe Shutdown Analysis (FSSA). This situation is described below for Unit 1, but also applies to Unit 2. Listed below is the configuration for 1EA1. The basic configuration is typical for 1EA2, 2EA1 and 2EA2 as well (ref. E1-0001) - Safeguard Bus 1EA1 is a 6.9 kV switchgear with a Main-Tie-Main configuration. - The normal lineup has one feeder breaker closed, the other feeder breaker open and the tie breaker closed. - 1EA1 receives normal power from the secondary side of Startup Transformer XST2 through breaker 1EA1-1. - 1EA1 receives alternate power from the secondary side of Startup Transformer XST1 through breaker 1EA1-2. - 1EA1 receives emergency power from diesel generator 1EG1 through breaker 1EG1. - An alternate source of power for 1EA1 is also available from Train C through breaker 1EA1-3. The control wiring for the 1EA1-1 circuit breaker contains the following attributes that are important to understand the issue: - Switch 43/1EA1-1, located in the Shutdown Transfer Panel (STP) is used to transfer control of 1EA1-1 from the Control Room (1-CB-11 switch CS-1 EA1-1) to the Hot Shutdown Panel (HSP) switch CS-1 EA-1 L. (Ref. E1-0031-01&02) - There is a trip circuit fuse located in the 6.9 kV switchgear compartment for 1EA1-1 for control of the trip circuit when the breaker is controlled at 1CB-11 and a separate fuse for the trip circuit when the breaker is controlled by the HSP. - The trip circuit for 1EA1-1 has a contact routed through the Control Room to the Solid State Protection System (SSPS) Cabinet. This contact is in the trip circuit when control is from the Control Room or when control is from the HSP. (Ref. E1-0031-01) The scenario is based on a fire in the Control Room. If the fire in the Control Room causes a ground in the wiring routed in the Control Room to the SSPS cabinet, the fuse for the 1E1-1 trip circuit would open. In the event of a fire in the Control Room, control of the plant is transferred to the HSP. When the control of breaker 1EA1-1 is transferred to the HSP and the ground condition in the SSPS wiring still exists, the second 1EA1-1 trip circuit would open. At this time there would be no way to remotely trip open 1EA1-1. The breaker could still be tripped mechanically at the breaker. Therefore, if 1EA1-1 is closed, 1EA1 could remain energized if off-site power is available. If off-site power is not available, but 1EA1-1 remains closed, bus 1EA1 would remain electrically connected to XST2. As part of the transfer of control from the Control Room to the HSP, operators start up the diesel generator and close 1EG1 to place 1EA1 loads on the generator. If 1EG1 is closed and 1EA1-1 breaker is still closed with off-site power available, the generator will be immediately connected to grid power through XST2 without synchronizing. If 1EG1 is closed and 1EA1-1 breaker is still closed with off-site power not available, the generator will be immediately connected to XST2 and attempt to energize XST2. This large current draw associated with energizing XST2 would likely stall and damage diesel generator 1EG1. Compensatory Action is being implemented through procedure revisions to preclude damage to the diesel generator during this scenario. The licensee will notify the NRC Resident Inspector. * * * RETRACTION FROM TOM RUCKER TO PETE SNYDER AT 1713 ON 9/8/11 * * * At 1939 central daylight time on July 21, 2011, Luminant Power notified the NRC (Event No. 47086) of a Unanalyzed Condition per 50.72(b)(3)(ii)(B) regarding the vulnerability from a potential control room fire on 'A' Safeguards bus. The event report described a portion of a cable running from the Hot Shutdown Panel (HSP) to the Solid State Protection System (SSPS) cabinet in the Control Room (CR) that was not protected from a CR fire scenario for which a worst case Control Room/Cable Spreading Room fire induced short could result in the 1EA1-1 circuit breaker, connecting the 345 kV Startup Transformer to the grid, not tripping and damaging the EDG. Upon further review it has been determined that the Fire Safe Shutdown Analysis (FSSA) modeled that cable in the analysis since CPNPP began commercial operation. Additionally, the FSSA accounted for the fire-induced circuit ground on this cable and one of the specified manual actions is to trip the 1EA1-1 circuit breaker to assure the EDG would load the 'A' Safeguard bus to support the required fire scenario. The EDG output breaker is designed to auto close once the 1EA1-1 circuit breaker is tripped open. The previous version (prior to compensatory actions) of ABN-803A/B, the procedure used should a fire occur in the Control Room, directed the Reactor Operator to trip 1EA1-1 circuit breaker once the diesel was verified running, then ensure the EDG breaker closed. There is no procedural step directing personnel to manually close the EDG breaker. Since, in this scenario, indication for the 1EA1-1 circuit breaker at the Hot Shutdown Panel would be lost, it would be expected that the RO would direct the RRO to verify 1EA1-1 circuit breaker position. Based on the above, Luminant Power has concluded that the FSSA adequately modeled the plant and procedures were written which would not direct any action that would have caused the condition stated in the CR description. Based on the above, the conclusion is that CPNPP did not have an unanalyzed condition that significantly degraded plant safety per 50.72 (b)(3)(ii)(B) regarding the vulnerability from a potential control room fire on 'A' Safeguards bus. Therefore, this event is retracted. The licensee will notify the NRC Resident Inspector. Notified R4DO (Lantz). | ||
| ENS 43562 | 13 August 2007 14:00:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Inadequate Fire Protection on Safety Chilled Water System Electrical Cables | At 0900 on July 30 2007, an Engineer noted during the review of a revision to the Comanche Peak Fire Safe Shutdown Analysis that a cable associated with the control circuitry for Train B of the Safety Chilled Water System may not be adequately protected from a potential fire. By design, electrical control cables for Trains A and B of the Safety Chilled Water System are located in the same fire zone. The original design specified that the Train B electrical control cables in this zone were to be protected with fire barrier material (thermolag). However, in this case the fire barrier material was found to be missing from the Train B electrical control cables. Upon discovery of this condition, a fire impairment was implemented for the affected fire zone. Engineering performed an evaluation of this condition and at 0900 on August 13, 2007 concluded that if a fire occurred in the affected fire zone, the required degree of separation for redundant safe shutdown trains was inadequate (i.e. both A and B trains were affected) and this would adversely affect the control circuitry and potentially prevent the Unit 1 Safety Chilled Water System from performing its intended safety function. The Unit 1 Safety Chilled Water Systems safety function at Comanche Peak is to remove heat dissipated from engineering safety features equipment and to maintain ambient temperatures in rooms containing safety related equipment below maximum design temperatures. This condition is similar to an example given in NUREG 1022, Rev. 2, Section 3.2.4 for an unanalyzed condition that significantly affects plant safety (fire barrier missing such that the required degree of separation for redundant safe shutdown trains is lacking). Therefore, this condition is reportable per 10CFR50.72(b)(3)(ii)(B), 'The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' The licensee notified the NRC Resident Inspector.
CPNPP is retracting Event Notification 43562 based on the following: Further review of this issue by Engineering has determined that the required degree of separation for redundant safe shutdown trains was adequate and the Unit 1 Safety Chilled Water System was capable of performing its intended safety function. Therefore, this condition is not reportable per 10CFR50.72(b)(3)(ii)(B), 'The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' CPNPP has informed the NRC Resident Inspector. R4DO (R. Nease) notified. |