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ENS 55303 | 11 June 2021 21:10:00 | Technical Specificiation Required Shutdown | At 1710 EDT on June 11, 2021, a Technical Specification required shutdown was initiated at Plant Hatch Unit 1. Technical Specification Condition 3.4.4.B unidentified LEAKAGE increase not within limits, was entered due to a greater than 2 gpm increase in unidentified LEAKAGE within the previous 24 hour period in MODE 1. This specification was entered on June 11, 2021, at 1615 EDT with a REQUIRED ACTION to restore leakage increase within limits within 4 hours. This REQUIRED ACTION could not be completed within the COMPLETION TIME; therefore, a Technical Specification required shutdown was initiated, and this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
Upon further review of the leakage rates, it was determined that at 1900 EDT on 6/11/2021 the drywell floor drain unidentified leakage increased greater than 2 gpm within the previous 24 hours while in MODE 1. Technical Specification (TS) 3.4.4.B was entered to reduce leakage increase to within limits within 4 hours. At 2000 EDT on 6/11/2021 unidentified leakage was reduced below the 2 gpm increase within the previous 24 hours due to actions taken to lower reactor power and pressure. Therefore, the TS required shutdown per TS 3.4.4.C was not applicable. Thus Event Report 55303 is being retracted. The NRC resident has been notified of the retraction. Notified R2DO (Miller). | Unidentified leakage |
ENS 52526 | 1 February 2017 21:05:00 | Unusual Event - Potential Security Event | Plant Hatch declared a notification of unusual event. Subsequent investigation determined the paraphernalia was related to plant drills. The Unusual Event was terminated at 1727 EST. The licensee notified the NRC Resident Inspector. Notified the DHS SWO, FEMA, NICC, and NNSA (via e-mail).
The event was retracted. The licensee notified the NRC Resident Inspector. Notified the R2DO (Shakur), NSIR (Holian), IRD MOC (Gott), DHS SWO, FEMA, NICC, and NNSA (via e-mail). | |
ENS 50662 | 9 December 2014 23:25:00 | Non-Functional Fire Barrier Discovered Affecting Both Safe Shutdown Paths | During an inspection of a fire penetration between Fire Area 1404, Switchgear Room 1G and Fire Area 1408, Switchgear Room 1F in the diesel generator building, the penetration was determined to be non-functional as a 3 hour fire barrier. In the event of a postulated fire in the affected areas, both safe shutdown paths on Unit 1 could be compromised. Given this information, the determination was made that this condition meets the reporting criteria of 10 CFR 50.72(b)(3)(ii)(B). Compensatory measures were established in accordance with the plant's Fire Hazard Analysis (FHA). The presence of the compensatory measures in addition to automatic fire detection in the fire areas ensure that the safe shutdown paths are preserved until the degraded condition can be repaired. (Condition report No.) CR904013 The licensee notified the NRC Resident Inspector
Further investigation revealed that after removing the outermost two inches of loose silicone foam material, and taking additional measurements there remained adequate silicone foam material to provide 10 inches of silicone foam sealing the penetration. Plant design shows that the wall in question is 18 inches thick and also that 9 inches of silicone foam is required in the penetration in order maintain a 3 hour fire rating for the wall/penetration. Since there is adequate foam in the penetration to maintain the 3 hour fire rating the penetration is fully functional. Based on this information, this penetration in its 'as found' state does NOT represent a condition that seriously degrades a principal safety barrier. As such this condition has been determined to no longer meet reporting requirement 10CFR50.72(b)(3) and is therefore NOT reportable. Based on this information the previous notification for Event 50662 is being retracted. The licensee notified the NRC Resident Inspector. Notified R2DO (Bartley) | Safe Shutdown Fire Barrier |
ENS 50588 | 2 November 2014 12:20:00 | Operations Support Center Ventilation System Out of Service | On November 2, 2014 EST, the bus providing power to the Health Physics emergency HVAC system for climate control tripped unexpectedly and has been out of service for greater than 30 minutes. The Health Physics emergency HVAC system for climate control is required for functionality of the Operations Support Center (OSC), which is a required emergency response facility (ERF). Actions to determine the cause of loss of 1R24-S030 Load Center and to return ERF to functional status are in progress with high priority. In the interim, the backup OSC remains fully functional and capable of providing the required support as defined in the Hatch Emergency Plan and emergency implementing procedures. This notification is being made in accordance with the plant's Technical Requirements Manual Specifications to make the notification within 8 hours in accordance with 10 CFR 50.72 (b)(3)(xiii) due to the unplanned loss of an emergency response facility. An update will be provided once the OSC has been restored to normal operation. The licensee notified the NRC Resident Inspector.
Power was restored to the 1R24-S030 Load Center at 1955 EST. At 2038 EST, power to the Health Physics emergency HVAC system was restored. The OSC was returned to service at 2120 EST. The licensee has notified the NRC Resident Inspector. Notified R2DO (Blamey).
In accordance with NUREG 1022, Revision 3, Supplement 1, the NRC endorsed NEI 13-01, 'Reportable Action Levels for Loss of Emergency Preparedness Capabilities,' such that 'if a licensee has a 'backup ERF' that is capable of performing the functions of the primary facility, the licensee's emergency assessment capability is not significantly impaired if the primary facility is not available.' Based on this information this condition is not reportable. Although NUREG 1022, Rev. 3, Supplement 1 did not require an event notification report to be made, an initial notification was generated based on requirements within licensee control documents. These documents have been updated to reflect the guidance provided in the supplement to NUREG 1022. This report is therefore being retracted in accordance with NUREG 1022, Rev. 3, Supplement 1 since a backup ERF remained fully functional and capable of providing the required support during the event. The licensee notified the NRC Resident Inspector. Notified R2DO (Hopper). | |
ENS 48795 | 28 February 2013 16:55:00 | Closed Cooling Water Isolation Valve Fails Local Leak Rate Test | On February 28, 2013, at 1155 EST, with the unit in Refueling Mode, a determination was made that reactor building closed cooling water (RBCCW) isolation valve (2P42-F051) exceeded its acceptance criteria for designed leakage when performing local leak rate testing. Diagnostic testing confirmed that all the leakage from its test boundary is going through this valve with an 'as found' leakage of >200,000 sccm at 32.87 psig. This valve is the outboard isolation barrier for that affected primary containment penetration with the inboard barrier being the RBCCW system itself as a 'closed' system. Previous practice is to conservatively include any leakage through this valve when performing as found leak rate tests as part of the primary containment leakage summary or as part of 0.6La. This is considered conservative since the RBCCW system inside containment is assumed to remain intact following a design basis accident (DBA) loss of coolant accident (LOCA). If the closed system remains intact there is no path for leakage to exit primary containment through this system. Since the past practice is to include the 'as found' leakage through this valve as part of 0.6La and since the 'as found' leakage would result in exceeding La, this condition is being considered a condition that results in the principal safety barriers being seriously degraded. This leakage would represent a loss of the containment function since the leak rate exceeded the Technical Specification limiting condition for operation (LCO) for primary containment. Further investigation is underway to determine if leakage through this single containment barrier is required to be included in the Appendix J primary containment leakage summary, since it is associated with a closed system. The licensee notified the NRC Resident Inspector.
Further investigation revealed an Appendix J exemption that was granted for the Hatch Unit 2 local leak rate test (LLRT) program in the 1978 time frame that specifically addressed the primary containment penetration that has primary containment isolation valve (PCIV) 2P42-F051 as its outboard barrier and the 'closed' system as its inboard barrier. The exemption recognizes that this system is designed to be intact and water filled post-LOCA, allows testing of the of the PCIV with water and states that the leakage through this PCIV is not included in the 0.6 La total. Since the Hatch RBCCW system supplying components inside primary containment is a 'closed' system and remains intact and water filled post-LOCA, there is no leakage path from primary containment through this RBCCW penetration. Since no leakage from primary containment can occur through this penetration in its 'as found' state, this condition does not represent a condition that seriously degrades a principal safety barrier. As such this condition has been determined to no longer meet reporting requirement 10CFR50.72(b)(3) and is therefore not reportable. Based on this information the previous notification is being retracted. The licensee notified the NRC Resident Inspector. Notified R2DO (O'Donohue). | Local Leak Rate Testing |
ENS 48772 | 20 February 2013 10:28:00 | Degraded Condition - Containment Penetration Fails Local Leak-Rate Testing | On February 20, 2013, at 0538 EST, local leak-rate testing (LLRT) of the 'A' feedwater check valves 2B21-F010A and 2B21-F077A revealed that neither valve would pressurize. Based on this information this line would not remain water filled post-LOCA and would result in the 'as found' minimum pathway leakage exceeding the limiting condition of operation (LCO) for Technical Specification 3.6.1.1. The cause for the LLRT failures will be determined and required corrective maintenance will be performed and valves successfully tested during the current refueling outage. The licensee notified the NRC Resident Inspector.
Subsequent investigation into the reported LLRT failure revealed that the initial LLRT performed on feedwater check valve 2B21-F010A was not considered an acceptable test, since that LLRT was not representative of the 'as found' condition of this check valve. The test volume for this valve had been slowly filled such that the check valve did not have the normal expected differential pressure across the valve disc to achieve normal check valve seating. After draining the test volume and refilling it by allowing the test volume to gravity fill from the reactor pressure vessel, the expected differential pressure across the valve disc occurred and seated the disc in such a way that it was more representative of the 'as found' condition for the check valve. An LLRT was then performed with a leakage of 50 accm (actual cubic centimeters per minute) against an acceptance criterion of 194 accm. No maintenance or operation of the check valve had occurred between the initial invalid test and the subsequent test performed with the disc in its 'as found' condition. An engineering evaluation was performed that documented the acceptability of using this means for establishing the test volume for feedwater check valves 2B21-F010A and 2B21-F010B for the 'A' and 'B' loops of feedwater, respectively. This engineering evaluation concluded that establishment of the required test volume in the manner described for primary containment penetration 9A satisfies the Hatch LLRT program requirements and that the leakage acceptance criterion for feedwater check valve 2B21-F010A in its 'as found' state was satisfied. The 2B21-F077A valve will be retested at a later date. Based on this information, the LLRT of this check valve in its 'as found' state was successful which actually resulted in successful minimum pathway leak rate test results for primary containment penetration 9A. These conclusive test results clearly indicated that the initial test results were incorrect and the 'as found' condition of this penetration isolation capability did not represent a significant degradation of a principal safety barrier as described in 10CFR50.72(b)(3)(ii)(A). For these reasons Notification # 48772 is being retracted. The licensee notified the NRC Resident Inspector. Notified the R2DO (McCoy). | |
ENS 46258 | 18 September 2010 20:27:00 | Unusual Event Declared Due to High-High Reactor Building Sump Level Alarm | This is a one-hour report for the discovery of a condition that met an Emergency Activation Level (EAL) for a Notification of an Unusual Event (NOUE). EAL HU-1 - Natural and Destructive Phenomena Affecting the Protected Area Threshold Value 6 states, 'Exceed Max Normal Operating Values specified in EOP 31 EO-EOP-014-1 SC - Secondary Containment Control Table 5 Secondary Containment Operating Water Levels.' Plant Hatch Unit 1 declared a NOUE on 9/18/10 at 1627 based on HU-1 Natural and Destructive Phenomena Affecting the Protected Area due to exceeding the Max Normal Operating Value for the Reactor Building NE Diagonal Floor Drain Sump. At 1615, both the Level High and Level High-High annunciators alarmed. Water has not overflowed the sump. However, due to its design, it is not possible to immediately visually confirm the sump level. The sump pumps were not running and would not run with their switches in START indicating either: (1) there is no water in the sump, (2) both pumps are OOC (Out of Commission), or (3) there is a problem in the sump level control system. Maintenance is assessing the situation. The licensee has notified the state and local agencies and will notify the NRC Resident Inspector.
At 0400 EDT, the NOUE was terminated. The termination criteria was that the NE Diagonal Floor Drain Sump level was in its normal operating range with no abnormal sump in-leakage detected. Compensatory measures are in place to measure level in the sump. An investigation into the abnormal level indication is still in progress. The licensee will be notifying state and local authorities as well as the NRC Resident Inspector. Notified R2DO (Henson), IRD (Gott), NRR EO (Howe), DHS (Inser) and FEMA (Via).
On September 18, 2010, a notification, Event Number 46258, was submitted to the NRC to report a circumstance at Plant Hatch Unit 1 which appeared to be a 'High-High' reactor building sump level alarm. Such an alarm would meet the designated Emergency Activation Level (EAL) for a Notification of an Unusual Event (NOUE). On the day of the event, it was thought that receipt of both the 'High' and 'High-High' level annunciators for the reactor building northeast diagonal floor drain sump was indicative of the sump level having reached its maximum normal operating level, and a NOUE was made. During a subsequent Investigation, engineering personnel determined that the 'High' and 'High-High' level switches for this sump are wired such that both alarms are actuated concurrently when the 'High' level in the sump is reached at approximately 52 Inches. The 'High-High' sump level annunciation (the level at which the NOUE must be made) Is supposed to alarm at approximately 60 inches. Engineering personnel also determined that the level logic is currently improperly configured to clear both the 'High' and 'High-High' level alarms when the sump level rises to 60 inches. The initial 'High' and 'High-High' level alarms in the main control room were received at approximately 1615 EDT on September 18, 2010. At that time the sump level would have reached an actual level of approximately 51 inches (albeit with a false indication of approximately 60 inches due to the configuration error noted above) based on the setpoint at which this alarm occurs. The normal leakage into this sump at the time of this event was at a rate of approximately 1 inch per hour. The sump pumps were started at approximately 1800 EDT to pump down the sump. This action resulted in a maximum actual sump level of approximately 53.5 inches, assuming the normal leakage into the sump was occurring during this period. At no time did the sump level actually reach the 'High-High' sump level criteria of 60 inches, nor did the annunciated level clear prior to pumping down the sump which also demonstrated that the 60 inch criteria was not met. It is now apparent that the initial NOUE was a conservative report that was properly made based on the information available to the operators at the time. The subsequent investigation provided additional information regarding actual sump level conditions including information that both 'High' and 'High-High' alarms annunciate at the 'High' sump level and would have cleared had the 'High-High' level been reached. Since the 'High-High' sump maximum normal operating level of 60 inches was never reached during this event, the determination has been made that the EAL was not appropriate given the actual sump conditions, thereby making the event non-reportable and an NOUE unnecessary. Based on the preceding Information, Southern Nuclear Company (SNC) hereby provides notification that Event Number 46256 is retracted. The licensee will notify the NRC Resident Inspector. Notified R2DO(McCoy). | |
ENS 45850 | 16 April 2010 23:17:00 | Loss of Cooling Accident Signal Due to High Drywell Pressure Signal | On April 16, 2010 at 1917 hrs., Unit 1 received an ECCS (emergency core cooling system) loss of cooling accident (LOCA) signal on high drywell pressure. Based on plant data, drywell pressure reached a maximum pressure of approximately 1.25 psig, which is below the LOCA and RPS (reactor protection system) signal actuation pressure of 1.85 psig. At this time, the cause of the drywell pressure increase is under investigation. RPS logic did not initiate due to drywell pressure not reaching the actuation setpoint of 1.85 psig. Although the ECCS logic prematurely actuated, the signal is being treated as 'valid' for the ECCS actuation until further investigation is completed. All expected ECCS actions occurred as a result of the signal. The LOCA logic has been reset and all affected systems have been returned to normal or standby configuration. As a result of the LOCA system actuations, several cooling tower fans tripped and condenser vacuum began to decrease. Reactor power was reduced to approximately 86 percent as a result of decreasing condenser vacuum. Power is being maintained at approximately 86 - 88 percent at this time. There are no other plant issues or concerns at this time. The licensee notified the NRC Resident Inspector. According to the licensee, normal drywell operating pressure is .5 to 1.2 psig. Prior to the event, drywell pressure had been steady at approximately 1.0 psig. Current drywell pressure is .6 psig. According to the licensee, ECCS systems that started (but did not inject) included: Core Spray pumps, Residual Heat Removal pumps, High Pressure Coolant Injection pump, and the Diesel Generators. An event review team is assessing this event to determine the root cause.
On April 16, 2010, Hatch Unit 1 received a LOCA ECCS initiation from a high drywell pressure signal. Based on the information available at that time, a notification was made to the NRC assuming the signal to be valid until further investigation could be completed. After further review, the determination has been made that the initiation signal originated from a faulted ATTS (Analog Transmitter Trip System) card and not from a valid high drywell pressure condition. Based on this information, this condition did not require an NRC notification in accordance with 10CFR50.72 and, as such, is being retracted through this updated response. The licensee notified the NRC Resident Inspector. Notified the R2DO (Hopper) | |
ENS 45775 | 18 March 2010 22:16:00 | Loss of Tone Alert Radios | At approximately 1816 EDT, loss of prompt notification system (tone alert radios) occurred. Security notified Jacksonville National Weather Service and Information Technology to investigate. Jacksonville National Weather Service determined a problem with their equipment and prompt notification system (tone alert radios) was returned to service at approximately 1845 eastern time. Security has notified state and local agencies of prompt notification system (tone alert radio) outage and return to service. The licensee has notified the NRC Resident Inspector. Licensee also notified State and local agencies. * * * RETRACTION FROM FRANK GORLEY TO PETE SNYDER AT 1535 ON 3/25/10 * * * On 3/18/10 at approximately 11:40 am we started receiving an 'off-air' alarm for the Plant Hatch NOAA Weather Radio. Initially we determined that a loss of the prompt notification system had occurred due to the alarm and consultation with Jacksonville National Weather Service, Ref. 10 CFR 50.72 (b)(3)(xiii). NOAA personnel subsequently provided information to Plant Hatch personnel at which time it was learned that even though 'off-air' alarms were received for the NOAA Weather Radio, the radio was never off the air. While NOAA personnel were troubleshooting the system the broadcast was switched over to a digital backup system that continued to have the capability to warn the public. There was a problem noted in the broadcast that was causing several seconds of silence in the broadcast signal. While on the digital backup, the notification capability was maintained. The broadcast had been returned to the primary system today. During the time these activities were underway, broadcast capability to the public was never lost. Based on this information, Plant Hatch is entering a Notification of Retraction for the 8 hr. report documented to the NRC Event Number 45775, as entered on 3/18/10. The licensee will notify the NRC Resident Inspector. Notified R2DO (Franke). | |
ENS 43884 | 6 January 2008 16:00:00 | Hpci Declared Inoperable Due to Failed Surveillance Test | During HPCI pump operability surveillance in preparation for a system outage, the system failed to achieve rated flow and pressure in the time required by procedure and Tech Specs. The procedural requirement is <49 seconds and the Tech Spec requirement is < 50 seconds. The system achieved rated flow and pressure in 54 seconds. The system outage has been delayed until troubleshooting plans can be developed and implemented. The licensee notified the NRC Resident Inspector.
The initial notification was made as a result of the failure of HPCI to meet its response time of 50 seconds as defined in the Technical Requirements Manual (TRM) due to a degraded component. Failure to meet the operability procedure requirements resulted in the high pressure coolant injection (HPCI) system being considered inoperable. Since HPCl is a single train system, its inoperability was the event that warranted a notification to meet the following reporting requirement: 10 CFR 50.72(b)(1)(v) Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems needed to: (D) Mitigate the consequences of an accident. An additional review and evaluation of the licensing basis was performed and demonstrated that the procedural required response time was set conservatively. As expected, the licensing basis does not require or assume any specific start time and HPCl is not credited in the accident analyses. The acceptance criterion contained in the TRM is within licensee control via the 10 CFR 50.59 process. Consequently, the TRM criterion was revised to 75 seconds to retain a value to assure continued monitoring and trending of HPCl performance in order to recognize and prevent continued performance degradation. Additionally, it is reasonable to conclude that HPCl would have completed its mission time of 4 hours despite the degraded condition of the EGR (Electronic Governor Remote) that caused the initial slower response time. This is based on the fact that HPCl started and ran at rated flow and pressure for approximately 39 minutes prior to shutdown with no problems identified. This removes any questions regarding its ability to restart and run based on demand during its mission time of 4 hours. An industry expert on this system and the engineer from the vendor that supports this system concurred with that conclusion. Since this subsequent review and evaluation determined that the slower response time did not render HPCl inoperable, no single train failure of HPCl occurred. The system was fully capable of performing its intended safety functions during the event timeline. Based on this information this notification serves to retract notification # 43884 made on 1/06/08. The licensee notified the Resident Inspector of this retraction. Notified the R2DO (Musser). | Time of Discovery Mission time |
ENS 43503 | 18 July 2007 16:20:00 | Unanalyzed Condition - Vent Space Less than Design Basis | During an inspection of the Unit 2 Reactor Building Steam Chase as a result of a similar situation being previously discovered on Unit 1, a storage gangbox was noted to be resting on one of the two hinged blowoff panels in the floor of the Steam Chase (elevation 130') and the hinged blowoff panels were determined to be restrained which would prevent their opening. The blowoff panels are designed to open to provide pressure and temperature relief between the torus room and 130 ft elevation of the reactor building for high energy steam line breaks. The 'as found' configuration of the blowoff panels hinder their capability to open, which constitutes a non-conforming and unanalyzed condition in that the vent area between the torus room and main steam chase in the reactor building is less than the area assumed in the analysis for a HPCI steam line break. As such, for a HPCl steam line break in the torus room, the short term pressure across the torus room ceiling would likely be greater than 2.3 psid, which is the maximum differential pressure stated in Chapter 15A of the Unit 2 FSAR. There is no known analyzed limit for the differential pressure between the torus room and the 130 ft elevation of the reactor building. The actual differential pressure given the 'as found' condition is not known at this time. Corrective actions have been taken to restore the assumed vent area between the torus room and the reactor building 130ft elevation. The gang box has been removed and both blow off panels no longer have restricted movement. The remaining 3' x 3' floor plug has also been removed, completely restoring the assumed vent area into compliance. Based on this information there is reasonable assurance that an adequate vent path currently exists such that the plant is no longer considered to be in a condition that significantly degrades plant safety. However, since the actual differential pressure given the 'as found' condition is not known at this time, the 'as found' condition as previously discussed is assumed to be an unanalyzed condition that represents a condition that significantly degraded plant safety; however, additional information is needed in order to more conclusively determine this. If more conclusive information is provided that indicates otherwise an update notification will follow. This was also reported under 10CFR50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector.
A review of the 'as found' configuration of the plant was performed to determine if this configuration would still be bounded by the calculations that support the HELB analysis in the Unit 2 FSAR. The 'as found' configuration consisted of having one torus plug in place rather than open and the two hinged torus ceiling blow-off panels bolted shut instead of being free to open. This engineering review concluded that if the torus ceiling blow-off panels do not open and with only one torus plug open, the torus pressures will not exceed the current FSAR pressures. Additionally, the torus pressures were found to be acceptable as a result of the modeling of friction in the HPCI pipe break mass and energy releases. This being the case, for a HPCI steam line break in the torus room, the short term pressure across the torus room ceiling would be 1.93 psid for the 'as found' condition which is less than the maximum differential pressure of 2.27 psid as stated in Chapter 15A of the Unit 2 FSAR. It should be noted that corrective actions were taken upon discovery and that the assumed vent area between the torus room and the reactor building 130 ft elevation was restored shortly following discovery. The gang box was removed, the restraint on the blow-off panels removed and both blow-off panels were confirmed to have full range of motion to open if the conditions were present that would warrant that movement. Based on this review of the design calculations while taking the 'as found' conditions into consideration, the conclusion reached is that the nonconforming 'as found' conditions did not represent a condition that significantly degraded plant safety. For this reason this condition that was initially reported under 10CFR50.72(b)(3)(ii)(B) is being retracted. The licensee will notify the NRC Resident Inspector. R2DO(Shaeffer) notified. | Unanalyzed Condition |
ENS 43499 | 17 July 2007 12:00:00 | Unanalyzed Condition - Vent Space Less than Design Basis | During a review of the temporary repair of the steam line drain bypass line in the Unit 1 Reactor Building Steam Chase, two storage gangboxes were noted to be on the grated opening in the floor of the Steam Chase (elevation 129 ft). These grated openings are designed to be open to provide pressure and temperature relief between the steam chase and the torus room for high energy steam line breaks. Appendix N to the Unit 1 FSAR credits the openings for venting the steam chase to the torus room through the openings for a main steam line break, and for venting the torus room to the steam chase for a HPCI steam line break in the torus room. The most limiting event is the HPCI steam line break in the torus room and the vent path associated with that event. Original assumptions used in the calculation for the vent opening did not adequately account for the grating itself and for louvers installed in a previous plant modification. As a result the vent area was further reduced. Upon further review of the above condition, it has been determined that a non-conforming and unanalyzed condition exists In that the vent area between the torus room and main steam chase in the reactor building is less than the area assumed in the analysis, even without gangboxes covering a portion of the grating. As such, for a HPCI steam line break in the torus room, the short term pressure between the torus room and the corner rooms (diagonals) is greater than 2 psid, which is the stated limit in Appendix N of the Unit 1 FSAR. The corner rooms contain ECCS components in the RHR and core spray systems. Based on engineering judgment there is reasonable assurance that the present nonconforming condition does not prevent safety systems and structures from fulfilling their safety function. This is based on the following information: Structural Steel floor elevation platforms do not appear to have been credited in the structural design capability of the walls. These platforms should act to help maintain the wall intact with increased pressure. The increased pressure transient is a very short term transient, approximately 2-3 seconds in duration, after which the pressure will return to within 2 psid. It is expected that the wall would withstand this transient without degrading the performance of the low pressure ECCS systems or other structures and components. Lastly, the probability of occurrence of a steam leak leading to an instantaneous line break is very small. There is currently no report of steam leaks from the HPCI line, and although a probability evaluation has not been performed, it is likely that the probability of occurrence of such a break is very small. Thus, there is no known immediate threat that would prevent safety systems from performing their safety function. More detailed review is continuing at this point. Short term corrective action will be required to increase the open 'vent' area between the torus room and the reactor building 130 ft elevation and restore at least the assumed vent path from the torus room. This can be accomplished by removing the gangboxes over the vent area in the steam chase and/or completing a floor plug evaluation of vent area needed between the torus room and the reactor building 130 ft elevation which will restore compliance with the 2 psid criteria. Analysis is currently underway to assess the pressure and temperature effects on the safety related structures and equipment by these short term actions. Regarding reportability, based on engineering judgment as previously discussed, the unanalyzed condition does not represent a condition that significantly degraded plant safety; however, additional information is needed in order to more conclusively determine this. For this reason this condition is being conservatively reported under 10CFR50.72(b)(3)(ii)(B) until such time as more conclusive information is provided to make the final determination. The licensee notified the NRC Resident Inspector.
Upon further review of the above 'as found' conditions, it has been determined that there are existing conservatisms in the current analysis which bound the flow restriction caused by the gang boxes on the grating. The evaluation concluded that the gang boxes found on the grated opening in the floor of the steam chase would not increase the pressures in the Unit 1 reactor building as a result of HELB conditions. Thus the pressure between the torus room and the corner rooms (diagonals) which is limited to 2 psid as stated limit in Appendix N of the Unit 1 FSAR is not affected. In addition, an additional open floor plug (the 3 ft by 3 ft floor plugs between Elevation 130 and the torus room below found to be covered by a hinged metal plate) is acceptable since it causes less differential pressure across reactor building compartments during the HELB's evaluated. The results of this additional review confirmed the original engineering judgment that there was reasonable assurance that the as found nonconforming condition did not prevent safety systems and structures from fulfilling their safety function. Short term corrective actions were completed upon discovery of he 'as found' condition to further increase the open 'vent' area between the torus room and the reactor building 130 ft elevation and restore at least the assumed vent path from the torus room. This was accomplished by removing the gang boxes over the vent area in the steam chase. Based on this review of the design calculations white taking the 'as found' conditions into consideration, the conclusion reached is that the nonconforming 'as found' conditions did not represent a condition that significantly degraded plant safety. For this reason this condition that was initially reported under 10CFR50.72(b)(3)(ii)(B) is being retracted. The licensee will notify the NRC Resident Inspector. R2DO (Shaeffer) notified. | Unanalyzed Condition |
ENS 42914 | 17 October 2006 15:21:00 | Accident Mitigation - Hpci System Inoperable | Unit HPCI system declared inoperable. During performance of the quarterly surveillance, HPCI pump operability, the HPCI system was secured when the "HPCI Turbine Oil Pressure Low" alarm was received and confirmed. Turbine governor end bearing oil pressure was 2 PSIG. The alarm setpoint is 6 PSIG and procedure limit is 10-12 PSIG. Pressure adjust valve was throttled open to raise pressure to 11 PSIG. The HPCI system was not immediately declared inoperable since an evaluation was being performed to determine if 2 PSIG turbine bearing oil pressure was adequate. Evaluation by the vendor will not be complete until 10/18/06. At 1655 hours, HPCI declared operable after a successful run with adequate oil pressure. HPCI is a single train system. The licensee notified the NRC Resident Inspector.
Original report was entered in error on 10/17/06 with Unit 2 versus Unit 1. Changed EN #42914 to accurately reflect the affected unit (Unit 1).
EN #42914 was submitted by Southern Nuclear Operating Company based upon a conservative decision to declare the HPCI system inoperable pending further evaluation to support its operability. Southern Nuclear Operating Company retracts EN #42914 based on the following discussion. During a subsequent review of the parameters by the HPCI Turbine Vendor, Dresser-Rand, and site engineering it was concluded that the HPCI system would have been capable of performing its intended safety function with the lower turbine governor end bearing oil pressure. During the operation of the system, the visual local indication was approximately 2.5 PSIG oil pressure at the governor end bearing. A review of the data showed that with a governor end oil pressure of the procedural minimal of 10 PSIG, the predicted oil flow rate would be 1.08 gpm with a minimum film thickness of 0.48 mils and a maximum bearing temperature of 228 deg F. With a degraded oil pressure of 2.5 PSIG, the predicted oil flow rate would be 0.54 gpm with a minimum film thickness of 0.46 mils and a maximum bearing temperature of 233 deg F. Based on the calculated data, the turbine governor end bearing would have performed satisfactorily for at least 8 hours at an oil pressure of 2.5 PSIG. Using the design basis success criteria, HPCI operation is successful if the system can inject water through the core Feedwater line for a total of 4 hours over a 24 hour period. The 4 hour mission time for HPCI is based on the design basis - if HPCI fails, it is backed up by the Automatic Depressurization System (ADS) in combination with Core Spray and Low Pressure Coolant Injection. The HPCI system is not credited for long term injection or late injection. Although the oil flow rate was reduced by 50% and the minimum film thickness reduced by 4%, the bearing temperatures were predicted to only increase a maximum of 5 deg F. Supporting this conclusion is the fact that the bearing was not damaged during the operation with low oil pressure when the turbine was run for 9 minutes at 2.5 PSIG governor end bearing oil pressure. The licensee will notify the NRC Resident Inspector. | Mission time |
ENS 42372 | 23 February 2006 23:50:00 | Degraded Condition of Shroud Tie Rods | While performing Unit One (In-Vessel Visual Inspection) IVVI examination of the four Shroud Tie Rods (Upper Support Horizontal Support Surface) the following results were reported: Tie Rod at 135 degree location: Crack-like indications beginning at the inner corner on both sides of the left support and extend to two thirds of the way to the outer corner with full penetration. Tie Rod at 225 degree location: Crack-like indication beginning at the inner corner on one side of the left support and extending a small portion of the way toward the outer corner. The indication is similar to that described for the shroud tie rod in the 135 degree location except that it is much less pronounced and is only on one side. Tie Rod at 45 degree location: No apparent indications present. Tie Rod at 315 degree location: No apparent indications present. One of the design criteria of the Shroud Tie Rods is to maintain zero separation between the shroud horizontal welds at 100% uprated power, assuming all of the horizontal welds (H1 thru H8) are fully cracked. The Shroud Tie Rods are also designed to maintain structural integrity of the shroud during all design basis accidents and transients. These findings bring into question the ability of these shroud tie rods to have performed their design function with the reactor in operation. This condition constitutes a serious degradation of a principal safety barrier had the unit been operating. The reactor is presently shutdown and the condition discovered does not represent an immediate safety concern for Unit 1. The extent of this condition is believed to be limited to Unit 1, since Unit 2 core shroud tie rods are made of different materials and installed in a different configuration. The licensee notified the NRC Resident Inspector.
Retraction of NRC Event # 42372: After further review and evaluation it has been determined that the eight hour call made February 23, 2006 per the guidance of 50.72(b)(3)(ii)(A) should be retracted. On 02/23/2006 at approximately 2135 EST, Unit 1 was in the refuel mode. During that time routine inspection of the Reactor Vessel Shroud Restraint Tie Rod Assemblies was in progress. The inspections revealed two cracks on the 135 degree assembly and one crack on the 225 degree assembly. In addition the mechanical preload on the 315 degree assembly was found to be below the design value. These assemblies were originally installed as a mechanical replacement of the horizontal shroud welds. An evaluation was performed of the as-found condition that considered the effects of the cracks on the upper supports and the reduced mechanical preload on the 315 degree assembly. The results of the analysis showed that sufficient compression existed for the tie rod assemblies to be considered operable in the as found condition. Inspection and evaluation of the horizontal shroud welds (original plant design) further determined that sufficient intact weld ligament existed to ensure the shroud design function was maintained without relying on the tie rod assemblies. Therefore, the structural integrity of the shroud was and is maintained for normal operation as well as all design basis accidents and transients. The licensee notified the NRC Resident Inspector. Notified R2 DO (M. Lesser) | |
ENS 41725 | 23 May 2005 21:47:00 | Manual Scram Due to Condenser Hotwell Chemistry | Based on increasing conductivity in the reactor vessel and condenser hotwell, a power reduction was initiated from 100 percent power. A manual scram was inserted at 57 percent RTP and 49 percent Core Flow based on Chemistry recommendations due to sulfates and chlorides in the hotwell. Following the scram a reduction in reactor water level to -28 inches resulted in a Primary Containment Group 2 Isolation (ESF) occurring. All isolations and systems responded as expected. Current plant status is Hot Shutdown with plans to proceed to cold shutdown. All control rods fully inserted and decay heat is being removed with the bypass valves into the condenser. The licensee notified the NRC Resident Inspector.
After further review and evaluation it has been determined that the four hour call made May 23, 2005 per the guidance of 50.72(b)(2)(iv)(B) should be retracted. A review of the event with respect to NUREG 1022 Revision 2 determined that: The manual scram was part of a pre-planned sequence to shut the plant down due to an equipment problem. The manual scram was part of a pre-planned sequence. The guidance to scram the reactor was established by the plant's Abnormal Operating Procedure addressing a condenser tube leak and was part of a preplanned sequence to prevent future equipment and component failures. The Manual Scram was not inserted to protect the plant against an event that presented a challenge to an FSAR analyzed event. In other words, this was not an Anticipated Operational Occurrence, an Accident, or a Special Event as defined in section 15.1.3 of the Unit 2 FSAR. Rather it was part of a plan to shutdown the reactor to protect against future potential equipment problems due to out of limits chemistry parameters. Further justification is provided by the fact that the manual scram was not initiated in anticipation of an automatic scram. Per NUREG 1022 Rev. 2: 'The staff also considers intentional manual actions, in which one or more system components are actuated in response to actual plant conditions resulting from equipment failure or human error, to be reportable because such actions would usually mitigate the consequences of a significant event. This position is consistent with the statement that the commission is interested in events where a system was needed to mitigate the consequences of the event.' However, the reporting requirement itself indicates that actuations that result from pre-planned sequences are not reportable. An example is provided in the NUREG of an equipment problem involving the loss of recirc pumps. In this example it is stated that: 'Even though the reactor scram was in response to an existing written procedure, this event does not involve a preplanned sequence because the loss of the recirc pumps and the resultant off-normal procedure entry were event driven, not pre-planned.' This is similar to our event, however, in the NUREG example, the reactor is scrammed to protect against the possibility of a stability event and stability is an FSAR analyzed event. In our case we were shutting down for chemistry reasons, not an FSAR type event. It is concluded that when the RPS is used to shutdown the reactor as part of a plan for the resolution of equipment problems, and the RPS is not needed to mitigate the consequences of an FSAR analyzed event, i.e., one which threatens a fission product boundary (i.e., fuel cladding, RCPB, primary and secondary containments), the RPS actuation is not reportable under 50.72(b)(2)(iv)(B). The licensee notified the NRC Resident Inspector. Notified R2DO (Haag).
On May 23, 2005 a four hour report was made per the guidance of 50.72(b)(2)(iv)(B), 'Any event or condition that results in an actuation of the RPS when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' This was made per event # 41725. The report was made within the four hour time frame of 10 CFR 50.72(b)(2). The four hour report for event # 41725 was retracted on June 6, 2005. After further consideration, the retraction made on June 6, 2005 is being cancelled and the original report re-instated. The licensee notified the NRC Resident Inspector. Notified R2 DO (K. Landis) | Anticipated operational occurrence Fuel cladding |
ENS 41103 | 8 October 2004 05:50:00 | High Pressure Coolant Injection (Hpci) System Declared Inoperable | On 10/08/04 on Unit Two, the HPCI Valve Operability was being performed. During the course of this evolution the suction path was transferred from the Condensate Storage Tank (CST) to the Suppression Pool. When the HPCI System was aligned to the Suppression Pool the Suction Pressure decreased from 25.5 psig to 1.5 psig. With HPCI aligned to the suppression pool and with suction pressure less than 14 psig the HPCI System was declared INOPERABLE. Investigation continues as to the cause of the low suction pressure. Preliminarily it is suspected that the Suppression Pool suction path was not adequately filled and vented following a recent tag out of that suction path for maintenance inspection activities. Investigation continues. The Licensee notified the NRC Resident Inspector.
The following information was obtained from the licensee via facsimile: The Unit 2 HPCI system was considered inoperable, since the technical information that would conclusively support its continued operability given the condition encountered could not be assembled within the time constraints of the reporting requirements. Subsequent to the event the system was confirmed be properly filled and vented with a negligible amount of air vented in the process. It was determined that this small amount of air was introduced to the suction piping as a result of an inspection activity performed for the HPCI suction check valve prior to the event. A limited amount of air remained in the torus suction piping causing the decrease in suction experienced during the event. Engineering reviewed the implications of the low suction pressure on the ability of the HPCI system to perform its safety function given the design of the system and the suction sources available. In each case Engineering was able to conclusively determine that the HPCI system would not have tripped due to low suction pressure had it received an automatic initiation signal and was actually operable during the time frame that Operations had conservatively treated the system as inoperable. Additionally, the effect of the trapped air being entrained in the pump suction was also analyzed, and the conclusion reached was that the air would not have prevented the pump's proper performance. Based on this information, the event reported on 10/08/2004 is not reportable. The licensee has notified the NRC Resident Inspector. The Headquarters Operations Officer notified R2DO (Bonser). | |
ENS 40824 | 17 June 2004 09:21:00 | High Pressure Coolant Injection (Hpci) Declared Inoperable | While changing out a light bulb in the Unit 1 HPCI room, the surveillance operator noticed that the bolts for the HPCI pump discharge check valve, 1E41-F005, bonnet on pressure seal were loose. All six bolts could be turned by hand. HPCI was declared inoperable until an investigation is performed. The bolts were properly torqued and HPCI pump run is complete and sat. Licensee entered Technical Specification 3.5.1 (14 day Limiting Condition of Operation). All other Emergency Core Cooling Systems are fully operable including the Reactor Core Isolation Cooling System. The NRC Resident Inspector was notified of this event by the licensee.
The following information was obtained from the licensee via facsimile: A subsequent investigation revealed that a hot torque of the Unit 1 HPCI pump discharge check valve (1E41-F005) was performed during the 165 psig pump operability surveillance test which was performed following reassembly during the Unit 1 refueling outage. During this surveillance, the pump discharge pressure and thus internal pressure on the pressure seal cover, is maintained between 265 and 305 psig. However, during the rated pressure run, the pump discharge pressure achieved is required to be greater than or equal to 1135 psig. This difference in internal pressure (1135-305 = 830 psig minimum), acting on the bottom of the pressure seal cover, forced the pressure seal cover to move upward toward the cover retainer, compressing the pressure seal more tightly. As the pressure seal cover moved toward the cover retainer, the retainer bolts also moved upward, but the cover retainer remained stationary, due to gravity. Therefore, the retainer bolts were no longer torqued against the retainer cover, creating the 'as found' condition. It is site and vendor (Flowserve) experience that once the pressure seal cover is wedged upward, sufficient friction exists between the pressure seal cover, the pressure seal, and the valve body to prevent the pressure seal cover from relaxing the sealing force on the pressure seal once the valve internal (system) pressure is removed. Furthermore, no gap existed between the head of any of the retainer bolts and the retainer cover. The sealing function of the pressure seal was never lost, and the valve would have performed its design function while the retainer bolts were in the 'as found' 'finger tight' condition. Therefore, the valve remained operable at all times when the HPCI system was required to be operable following the Unit 1 refueling outage. This was further substantiated by the fact that no leakage was observed during the rated pressure pump operability run on 3/14/04. Following discovery of the 'finger tight' condition, the retainer bolts were cold torqued to the appropriate value (370 ft-lbs) by Maintenance personnel on 6/17/04. A HPCI pump surveillance was then performed and a hot torque of 370 ft-lbs was performed immediately after the system was shutdown. Tampering was considered as a possible cause for the loosened bolts, but no evidence could be found to support these bolts being loosened intentionally. All evidence available suggests that the bolts were loosened by internal pressure, which is consistent with vendor experience. Based on the above information this event is not reportable, and this notification serves to withdraw the previous notification made on 6/17/2004. The licensee has notified the NRC Resident Inspector. The Headquarters Operations Officer notified R2DO (Bonser). | |
ENS 40649 | 6 April 2004 04:10:00 | Momentary Loss of Noaa Weather Radio System | The Plant Site detected a momentary loss of the NOAA Weather Radio System from 00:10 am on 04/06/04 to 00:14 am on 04/06/04. During this time interval of inoperability for this system, a major loss of off site notification capability is considered. Site Emergency Planning personnel were notified and will investigate this momentary loss of system capability. The NOAA Weather Radio System is currently in service and functioning properly. The licensee will notify the NRC Resident Inspector.
After further investigation, a determination has been made that there was no major loss of offsite notification capability of the Hatch Prompt Notification System (NOAA Weather Radio System). The momentary losses reported were caused by testing of the primary audio feed line utilizing a tone generator while investigating the loss of this feed. The secondary audio feed from Jacksonville NWS remained operable during the time of testing. The Licensee notified the NRC Resident Inspector. Notified R2DO (Decker) | |
ENS 40647 | 5 April 2004 23:16:00 | Brief Loss of Noaa Weather Radio Capability | Site experienced a momentary loss of the NOAA Weather Radio System from 1916 to 1918 EST. During the time that this system was out of service, a major loss of offsite communication capability is considered. Site emergency planning personnel were notified and will investigate this momentary loss. The system is in service and functioning properly. The licensee notified both state/local agencies and will inform the NRC Resident Inspector.
After further investigation, a determination has been made that there was no major loss of offsite notification capability of the Hatch Prompt Notification System (NOAA Weather Radio System). The momentary losses reported were caused by testing of the primary audio feed line utilizing a tone generator while investigating the loss of this feed. The secondary audio feed from Jacksonville NWS remained operable during the time of testing. The Licensee notified the NRC Resident Inspector. Notified R2DO (Decker) |