Semantic search

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.

• * * RETRACTION AT 1700 ON 4/9/04 RUSSELL TO GOTT * * *

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)

The plant reported that there was a freon leak in the Drywell Chiller Room located in the Unit 2 Reactor Building. Maintenance was being performed on the refrigeration equipment relief valve when the gas began to escape into the room. There were six men working in the room and two were affected by the freon gas. They were treated on site and released. Oxygen content in the room was 20.5% and LEL for hydrocarbons was alarming at 19%. The drywell chiller room is secured and being cleared of the 1,200 pounds of freon gas. At this time an investigation is being made to determine the cause which may have been due to the relief valve opening or the workers removing the valve. The plant entered the NOUE due to a "toxic gas release". The NOUE will be terminated when the room is habitable. The NRC Resident Inspector was notified along with State and Local agencies.

• * * UPDATE PROVIDED BY FRANK GORLEY TO JEFF ROTTON AT 0303 EST ON 02/08/05 * * *

Licensee reported that the NOUE was terminated at 0245 EST after the freon had been successfully cleared from the Drywell Chiller Room. Licensee will notify the NRC Resident Inspector. Notified FEMA (Liggett), DHS (Knox), IRD (Wilson), R2DO (Payne), and NRR EO (Beckner).

On 4/2512005 at 12:00 p.m., EDT the Hatch Nuclear Plant's Technical Support Center (TSC) was removed from service for planned maintenance and equipment modifications. The maintenance activities require relocation of the control panel (1X75-C001) which contains the HVAC controls and annunciators as well as the radiation monitoring meter and annunciator. The loss of this equipment requires taking the TSC out of service. The maintenance activities are scheduled to take five weeks. The alternate TSC will remain operable and available, so no loss of TSC function will occur due to the maintenance activity. This event is reportable per 10CFR50.72 (b)(3)(xiii) as described in NUREG-1022, Rev. 2 since this work activity results in a loss of an emergency response facility for the duration of the evolution. The NRC Resident Inspector was notified and state and local notifications will be made.

• * * UPDATE ON 05/04/05 @ 1510 BY GREG JOHNSON TO GOULD * * *

The TSC was returned to service at 1100 EDT on 05/04/05. The NRC Resident Inspector was notified. The Reg 2 RDO (Lesser) was notified.

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.

• * * RETRACTION FROM SHIFT SUPERVISOR (TONY SPRING) TO ABRAMOVITZ AT 16:33 ON 6/6/2005 * * *

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).

• * * RETRACTION RESCINDED - S. BURTON TO M. RIPLEY 1524 EDT 06/08/05 * * *

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)

NUE declared on Hatch Unit 1 due to a fire on the main transformer lasting greater than 10 minutes. Unit 1 reactor scram on main turbine trip. Group 2 PCIV (Primary Containment Isolation Valves) isolation on low reactor water level (+3 inches). All group 2 valves closed as required. Fire extinguished at 1358 hours. Plant is stable. All control rods fully inserted on the reactor scram. The plant is in hot shutdown steaming through the turbine bypass valves to the main condenser. No safety systems are affected. Emergency electrical buses are on normal offsite power and emergency diesel generators are available if required. The onsite fire brigade responded to the fire and the fire suppression system for the main transformer functioned as expected. The licensee notified the NRC Resident Inspector.

• * * UPDATE FROM E. BURKETT TO W. GOTT AT 1950 ON 10/29/05 * * *

The licensee is still in a NUE. The fire is out. The fire reflashed at 1603 when the deluge system was secured to inspect the transformer. The deluge system was restarted immediately. Offsite fire support have responded. Cleanup efforts are in progress. The licensee notified the NRC Resident Inspector. Notified IRD (Blount), R2DO (Cahill), and NRR EO (J. Hannon).

• * * UPDATE FROM E. BURKETT TO P. SNYDER AT 2010 ON 10/29/05 * * *

The licensee is still in a NUE. This call is to add an offsite notification per 10 CFR 72.75 (b)(2) to another government agency. The site chemistry department notified the National Response Center (Coast Guard) at 1636 that the site had discharged an unknown quantity of oil into the Altamaha River and was implementing site spill control and countermeasures procedures. The licensee also notified the Georgia Environmental Protection Division. The licensee notified the NRC Resident Inspector. Notified IRD (Blount), NRR EO (J. Hannon), and R2DO(Cahill).

• * * UPDATE FROM E. BURKETT TO W. HUFFMAN AT 0021 EDT ON 10/30/05 * * *

The licensee reports that the fire is confirmed extinguished (i.e., no risk of reflash) based on long term monitoring of the transformer including thermography measurements. The Unit remains in an Unusual Event until assessment activities related to the transformer fire have been completed. The licensee notified the NRC Resident Inspector.

• * * UPDATE FROM E. BURKETT TO W. HUFFMAN AT 0105 EDT ON 10/30/05 * * *

Based on the results of assessment activities related to the transformer fire, the licensee has exited the Unusual Event at 0050. The licensee notified the NRC Resident Inspector. The R2DO (Cahill), EO (Hannon), IRD (Blount), DHS (Doyle), and FEMA (Bisco) have been notified.

This is a non-emergency Event Notification made in accordance with 10 CFR 20.2201(a)(1)(ii) to inform the NRC of a nuclear material accountability discrepancy amounting to, in the aggregate, a portion of a spent fuel rod used at Plant Hatch (HNP). In the process of reviewing records and physically verifying the contents of the spent fuel pool (SFP) as a part of activities associated with SNC's (Southern Nuclear Company) response to Bulletin 2005-01, SNC has identified discrepancies between fuel segments located in the SFP and segments indicated in plant records. The segments originated in the early 1980s during fuel reconstitution and inspection activities. The discrepancies call into question the location of segments of single spent fuel rods in each of three bundles. Characterization of three segments located in the SFP provided rod serial numbers which, in turn, were used to determine the bundles from which these segments originated. These bundles were then inspected, and the length of fuel in the location corresponding to each rod segment's intended location was determined. The aggregate in-bundle length found in these rod locations was combined with the lengths of the segments located in the SFP and compared to the design active fuel length of the three rods. This comparison results in a material discrepancy of approximately 55 inches, based on the length measurements. In addition to this discrepancy, historical records indicate two segments (totaling 13 inches of fuel length), which may not be within the inventory of segments identified to date in the SFP. When this amount is added to the length associated with the three rod locations, a discrepancy of approximately 68 inches results. When the planned supplemental inspection of select bundles and SFP locations is completed and photographs are evaluated to aid in the determination of special nuclear material present, this 68 inch estimate may increase or decrease. On June 16, 2005 SNC formed a team to identify and characterize material in the SFPs at Hatch in order to account for special nuclear material (SNM) at the level of detail requested by Bulletin 2005-01. A work scope was established and specialized resources were contracted to support the work activities. During the performance of these work activities, a number of items of interest were characterized as being either SNM or non-SNM items. On October 28, 2005, SNC submitted the interim status report to NRC. On November 4, 2005 the Hatch Plant Review Board (PRB) reviewed the SNM Issue Resolution Team's assessment of data produced by the records searches and physical cataloging of SNM in the SFPs. Based on that review, the PRB concurred that a discrepancy exists in material accounting for a portion of a spent fuel rod in each of three bundles and records, as noted above, which in the aggregate approximates 68 inches of fuel rod length. This length is equivalent to about 45% of the length of one intact fuel rod. Further physical examination of the SFP will include additional examination of SFP floor areas that have not been examined to date and selected fuel bundle inspections. The SFP floor areas are limited to a small number of locations that are under equipment or objects stored on the SFP floor. This expanded work scope is expected to be completed by December 15, 2005. Based on the nature of the fuel rod segments and radiation monitoring, a high degree of confidence exists that the segments are in a restricted area of the plant or otherwise under the control of a licensed facility such that the public health and safety has not been adversely affected. In addition, there is no evidence of theft or diversion. This notification satisfies the 30-day notification requirement of 10 CFR 20.2201(a)(1)(ii). A subsequent written report will be made in accordance with 10 CFR 20.2201(b). The licensee has informed the NRC Resident Inspector regarding the discrepancies. SNC will be making a press release describing the current status of this issue. Accordingly, this notification also satisfies the 4-hour notification requirement of 10 CFR 50.72(b)(2)(xi) with respect to issuance of the press release associated with this issue.

• * * UPDATE AT 15:48 ON 8/21/2006 FROM FRANK GORLEY TO ABRAMOVITZ * * *

This is an update of non-emergency Event Notification 42135 that was previously made on November 10, 2005, in accordance with 10CFR 20.2201. This non-emergency Event Notification is made in accordance with 10 CFR 74.11 and informs the NRC of the loss of special nuclear material (SNM) from the historic breakage of several fuel rods used at Plant Hatch (HNP) in the early 1980s amounting to, in the aggregate, approximately 18 inches of a spent fuel rod. This amount is based on the available information, potentially affected by incomplete historic documentation. While reviewing records and physically verifying the contents of the spent fuel pool (SFP) in 2005 and 2006 associated with its response to Bulletin 2005-01, SNC identified discrepancies between fuel rod segments located in the SFP and segments indicated in plant records. This discrepancy was the subject of Event Notification 42135 and LER 2005-003 transmitted by letter NL-05-2262 dated 12/09/2005, including Rev. 1 of the LER, dated 04/14/2006 (transmitted by letter NL-06-0689). Additional locations in and around the fuel racks and additional fuel bundles were inspected between November 11, 2005, and July 21, 2006. Fuel vendor disorientation and plant SNM offsite shipping records reviews and quantity reconciliations were also recently completed. Based an the results of this expanded work scope, SNC concluded that some SNM material has been lost. This material either resides in some unidentified location in the SFP, resides in the bottom of the SFP as particles or small pieces, or was inadvertently shipped to a licensed low level waste processing facility. One SNM fragment, referred to as Item 30, was dropped during the physical activities in the pools and has not yet been recovered. This 4-1/2-inch fuel segment had been characterized and quantified and was dropped in the SFP during handling. A search was performed to look for item 30, but the intact segment was not found. During this search, a cladding segment with no appreciable SNM inside was located and identified as Item 32. It may be a portion of Item 30. Item 30's fuel length of 4-1/2 inches is included in the total characterized as lost. Based on the nature of the fuel rod segments, fragments, pellets, pellet chips, and small particles, and the barrier provided by in-plant radiation monitoring Instrumentation, a high degree of confidence exists that the lost SNM is either still in the SFP or was inadvertently shipped offsite to a licensed low level waste processing facility. Throughout its investigation and review, SNC has identified no evidence to indicate the possibility of theft or diversion of the missing quantity of SNM material. This notification satisfies the one-hour notification requirement of 10 CFR 74.11 (b). A subsequent written report will be made in accordance with 10CFR74.11(c). The licensee has informed the NRC Resident Inspector regarding the discrepancies and the conclusion regarding the lost material. SNC will be making a press release describing the current status of this issue. Accordingly, this notification also satisfies the 4-hour notification requirement of 10 CFR 50.72(b)(2)(xi) with respect to issuance of the press release associated with this issue. Notified the R2DO (Lesser), PAO (Brenner), and NRR EO (Jung).

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 FROM E. BURKETTE TO M. RIPLEY 0859 EDT 04/21/06 * * *

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)

The Unusual Event was declared due to a fire lasting greater than 10 minutes (after discovery) within the protected area. The fire was located on the Isophase Bus Duct near the Main Transformer. A load reduction to 80% rated thermal power is in progress on Unit 1. The fire was extinguished at 0450 with dry chemicals extinguishers." The licensee believes the fire was on cabling in the Isophase bus duct due to overheating. The licensee has been monitoring hot spots on the duct cabling for several days. The fire did not apparently impact any other systems and observers are at the location monitoring for any change in conditions. No Tech Spec Limiting Conditions of Operations resulted from the fire and there is no impact on Unit 2 operations (which is currently at 100%). Termination criteria will be based on management judgment and safety assessment.

• * * UPDATE FROM P. UNDERWOOD TO M. RIPLEY 0727 EDT 05/01/06 * * *

Unusual Event terminated at 0655 (EDT). Fire extinguished. Reactor power reduction continuing. Management evaluating continued operation. The unit is currently at 77% power and plans are to reduce power to 60% pending their evaluation of continued operation. The licensee will notify the NRC Resident Inspector. Notified R2 DO (K. Landis), NRR EO (M.J. Ross-Lee), IRD Manager (P. Wilson), DHS (S. York), and FEMA ( M. Roland).

During routine weekly operation of High Pressure Coolant Injection (HPCI) Auxiliary Oil Pump (AOP), 2E41C002-3, the pump displayed indications of excessive motor current after the pump had been inservice for approximately 45 minutes. The pump was secured and, following review of electrical diagrams and consultation with Electrical Maintenance, the operating current of the AOP was checked and determined to be excessive. The AOP was declared inoperable, with the AOP inoperable, the HPCI system cannot be considered operable. The HPCI System is a single train ECCS system. Investigation into the cause of the high motor current is ongoing. All other Emergency Core Cooling Systems are fully operable including Reactor Core Isolation Cooling (RCIC). The NRC Resident Inspector was notified of this event by the licensee.

• • • UPDATE FROM A. DISMUKE TO J. KNOKE AT 0933 ON 07/21/06 ***

Retraction of NRC Event # 42657: After further review and evaluation it has been determined that the eight hour call made June 21, 2006 per the guidance of 50.72(b)(3)(v)(D) should be retracted. On 06/21/2006 at approximately 0015 EDT, Unit 2 was at 100 percent Rated Thermal Power. During routine weekly operation of the High Pressure Coolant Injection (HPCI) Auxiliary Oil Pump, 2E41-C002-3, the pump displayed indications of excessive motor current after the pump had been in-service for approximately 45 minutes. Investigation revealed the running amps to be 46 amps with nameplate data running amps shown as 27 amps. An evaluation was performed for the as-found condition that considered the cause and effects of the increased running amps on the ability of the auxiliary oil pump to perform its design function. Specifically, the effect of a shunt resistor short to open was reviewed. Areas reviewed for impact were motor speed, system over pressurization, motor insulation, Environmental Qualification, and motor service life. The results of the evaluation showed that significant margin existed to ensure the auxiliary oil pump design function was maintained. Therefore, the auxiliary oil pump operability was maintained and HPCI operability was also maintained. The HPCI system was immediately removed from service using normal plant procedures; a work order initiated, and the existing motor was replaced to ensure continued long term reliability. The licensee notified the NRC Resident Inspector. Notified R2DO (Ernstes)

On 9/27/2006, the HVAC for the Hatch Nuclear Plant's Technical Support Center (TSC) was removed from service for planned preventive maintenance and inspections and testing activities. These work activities are planned to be performed and completed within a 12 hour work shift. During the time these activities are being performed, the TSC air handling unit, TSC condensing unit, TSC filter train and the fan unit for the TSC filter train will not be available for operation. As such, the TSC HVAC will be rendered non-functional during the performance of this work activity. If an emergency condition requiring activation of the TSC occurs during the time these work activities are being performed, then contingency plans call for utilization of the TSC as long as radiological conditions allow. The site Technical Support Center activation procedure provides instructions to direct TSC management to the Control Room and TSC support personnel to the Simulator Building to continue TSC activities if it is necessary to relocate from the TSC so that TSC functions can be continued. This event is reportable per 10CFR50.72 (b)(3)(xiii) as described in NUREG-1022, Rev. 2 since this work activity affects an emergency response facility for the duration of the evolution. The licensee will notify the NRC Resident Inspector.

• * * UPDATE RECEIVED FROM ANDY DISMUKE TO JOE O'HARA AT 1014 ON 9/28/06 * * *

The maintenance outage affecting the normal power supply for the TSC HVAC was not completed yesterday (September 27, 2006) as originally scheduled. The work schedule called for the normal power supply to the TSC HVAC to be returned to service within 12 hours; however, problems occurred in the functional test of the associated bus that provides the normal power supply to the TSC HVAC. During the functional test, the bus successfully swapped from normal supply to alternate supply, however the swap back to normal was unsuccessful. When the alternate breaker was tripped per procedure, the normal breaker indicated closed (green and amber lights extinguish, red light illuminated as expected, but the motor control center (MCC) was not energized. Maintenance has subsequently reenergized the MCC from alternate supply and TSC HVAC is now functional as of 0808 today (September 28, 2006). The total out of service time for the TSC HVAC was approximately 29 hours, 21 minutes. A repair plan is being developed to make necessary repairs to the breaker for the normal supply. Once repairs are completed, the breaker will be re-installed so that the MCC can be re-energized from the normal supply. This repair activity will require the TSC HVAC to be removed from service again in order to switch back to the normal power supply. An update will be provided when this evolution is expected to occur. This event is reportable per 10CFR50.72(b)(3)(xiii) as described in NUREG-1022 Rev 2 since this work activity affects an emergency response facility for the duration of the evolution. The licensee will notify the NRC Resident Inspector. The R2DO(Collins) has been notified.

• * * UPDATE ON 9/28/06 AT 15:02 FROM HATCH (DISMUKE) TO ABRAMOVITZ

On 9/28/06, the HVAC for the Hatch Nuclear Plant's Technical Support Center (TSC) is being removed from service to install the normal supply breaker in the MCC that supplies power to the TSC HVAC following breaker repairs. Estimated time for the TSC HVAC being out of service is 6 -8 hours. An update will be provided when the TSC HVAC is returned to service. The licensee will notify the NRC Resident Inspector. Notified the R2DO (Collins).

• * * UPDATE 9/29/06 AT 0004 ET FROM HATCH (F. GORLEY) TO M. RIPLEY

On 09/28/06 at 2329 ET, the HVAC System for the TSC was returned to functional status following the replacement of the normal supply breaker to the MCC and energizing the MCC that supplies the TSC HVAC. Walk-down of the HVAC System for proper operation was completed at 2335 ET. The licensee will notify the NRC Resident Inspector. Notified the R2DO (Ayres).

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.

• * * RETRACTION FROM GORLEY TO HUFFMAN AT 1435 EDT ON 8/30/07 * * *

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.