| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
08/30/2007 - 08/31/2007 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 43499 | Facility: HATCH
Region: 2 State: GA
Unit: [1] [ ] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: STEVE BRUNSON
HQ OPS Officer: JASON KOZAL | Notification Date: 07/17/2007
Notification Time: 15:21 [ET]
Event Date: 07/17/2007
Event Time: 08:00 [EDT]
Last Update Date: 08/30/2007 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
MARK LESSER (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text 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.
* * * 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. |
| !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 43503 | Facility: HATCH
Region: 2 State: GA
Unit: [ ] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: STEVE BRUNSON
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 07/18/2007
Notification Time: 18:43 [ET]
Event Date: 07/18/2007
Event Time: 12:20 [EDT]
Last Update Date: 08/30/2007 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD
50.72(b)(3)(v)(B) - POT RHR INOP
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
| Person (Organization):
MARK LESSER (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text 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.
* * * RETRACTION FROM GORLEY TO HUFFMAN AT 1435 EDT ON 8/30/07 * * *
"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. |
| !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 43562 | Facility: COMANCHE PEAK
Region: 4 State: TX
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: EUGENE SKELTON
HQ OPS Officer: JOHN KNOKE | Notification Date: 08/13/2007
Notification Time: 17:04 [ET]
Event Date: 08/13/2007
Event Time: 09:00 [CDT]
Last Update Date: 08/30/2007 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
| Person (Organization):
ANTHONY GODY (R4)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text 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.
* * * RETRACTION ON 08/30/07 AT 1249 EDT FROM RAUL MATINEZ TO MACKINNON * * *
"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. |
| General Information or Other | Event Number: 43570 | Rep Org: MISSISSIPPI DIV OF RAD HEALTH
Licensee: PEPSI BOTTLING GROUP
Region: 4
City: CANTON State: MS
County:
License #: WN-R0831
Agreement: Y
Docket:
NRC Notified By: BOBBY SMITH
HQ OPS Officer: JASON KOZAL | Notification Date: 08/15/2007
Notification Time: 18:06 [ET]
Event Date: 08/15/2007
Event Time: [CDT]
Last Update Date: 08/30/2007 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
LINDA HOWELL (R4)
JOSEPH HOLONICH (FSME)
| Event Text AGREEMENT STATE REPORT - INADVERTENT SHIPMENT OF RADIOACTIVE MATERIAL
The State provided the following information via email:
"DRH received notification on 8-15-07 from the Washington Dept. of Health, about the possibility of two (2) Industrial Dynamic Models FT-12 (Device Serial # 101942/Source # 318) and FT-50 (Device Serial No. 116158/Source # 3109) each containing 100 millicuries of Americium 241, being [inadvertently] delivered to a company in Mississippi. Washington DOH asked MS DRH to assist. DRH sent inspectors to the Canton, Mississippi location and verified that the sources were in both devices. DRH called Washington DOH and reported that the devices were indeed at that location.
"DRH called contractor QSA Global for assistance with disposal. QSA will deal directly with company that shipped devices to Mississippi.
"DRH confirmed both sources were in [the] devices with survey measurements of 0.7 Mr/hr at [the] surface."
Mississippi report number - MS 07003
* * * UPDATE FROM STATE OF MISSISSIPPI (SMITH) TO HUFFMAN AT 1735 EDT ON 8/30/07 * * *
The State provided the following information via email:
"The Americium sources (2 sources/100 millicuries each) in the Industrial Dynamic Filtec gauges were picked up for disposal by licensed contractor on 8-30-07. DRH will also notify State of Washington DOH about disposal of sources."
R4DO (Nease) and FSME (McConnell) notified. |
| Power Reactor | Event Number: 43607 | Facility: HATCH
Region: 2 State: GA
Unit: [1] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: FRANK GORLEY
HQ OPS Officer: JEFF ROTTON | Notification Date: 08/30/2007
Notification Time: 09:18 [ET]
Event Date: 08/30/2007
Event Time: 08:58 [EDT]
Last Update Date: 08/30/2007 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
SCOTT SHAEFFER (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 100 | Power Operation | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text PLANNED ELECTRICAL MAINTENANCE REMOVES TSC HVAC FROM SERVICE
"Planned preventive maintenance activities are being performed today (August 30, 2007) on the Hatch Nuclear Plant's Normal Supply Breaker (1C) to the Diesel Generator Building Motor Control Center (MCC) (MPL # 1R24-S026). This MCC feeds the Technical Support Center (TSC) 480V AC distribution panel (MPL# 1R25-S102) which supplies power to the TSC HVAC. The work activities affecting the TSC are planned to be performed and completed expeditiously within one work shift (<12 hours). During this work activity the TSC HVAC system will be removed from service. If an emergency condition occurs that requires activation of the Technical Support Center, during the time these work activities are being performed, it will take no more than four hours to return the equipment back to functional status, dependent on the stage of the work activity at the time the emergency occurs. Plans are to utilize the TSC for any declared emergency during the time these work activities are being performed as long as radiological conditions allow. Procedure 73EP-EIP-063-0. Technical Support Center Activation 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 primary TSC.
"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. |
| Power Reactor | Event Number: 43608 | Facility: GINNA
Region: 1 State: NY
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP
NRC Notified By: ROY GILLOW
HQ OPS Officer: JEFF ROTTON | Notification Date: 08/30/2007
Notification Time: 11:08 [ET]
Event Date: 08/30/2007
Event Time: [EDT]
Last Update Date: 08/30/2007 | Emergency Class: NON EMERGENCY
10 CFR Section:
26.73 - FITNESS FOR DUTY
| Person (Organization):
ART BURRITT (R1)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text POTENTIAL FITNESS FOR DUTY SAMPLE COLLECTION ERROR
The licensee is conservatively making this report although no specific FFD impacts are indicated. A non-licensed employee responsible for FFD sample collection did not meet station reliability expectations and his access to the site has been terminated. Contact the Headquarters Operations Officer for additional details.
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 43610 | Facility: VERMONT YANKEE
Region: 1 State: VT
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: M. PLETCHER
HQ OPS Officer: JOHN MacKINNON | Notification Date: 08/30/2007
Notification Time: 17:16 [ET]
Event Date: 08/30/2007
Event Time: 15:13 [EDT]
Last Update Date: 08/30/2007 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
ART BURRITT (R1)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | A/R | Y | 62 | Power Operation | 0 | Hot Shutdown | Event Text AUTOMATIC REACTOR SCRAM ON TURBINE STOP VALVE CLOSURE
Event Description:
"Reactor scram (4 hr notification) automatic scram
"Primary containment isolation of Groups 2,3,4, and 5 due to RPV Level < low level setpoint (<127") due to reactor scram. (8 hr notification).
Actions Taken (reference applicable Technical Specifications):
"Implemented OT 3100 (Reactor Scram Procedure) EOP-1 (RPV Control). Placed the plant in a stable condition and implemented OP 0109, Plant Restoration. "
The NRC Resident Inspector was notified of this event by the licensee.
Reactor was initially at approximately 63% power due to cooling tower damage which occurred more than a week ago. License was performing a surveillance test of the # 2 turbine stop valve. The valve was shut per the surveillance test procedure but they were unable to open the valve. Personnel were in the heater bay and mechanical assistance was applied to open the valve. The valve opened quickly at which point the licensee received a turbine stop valve closure signal which generated an automatic reactor scram. All rods fully inserted into the core. Reactor vessel water level decreased below 127 inches, due to the reactor scram, which caused primary containment isolation of groups 2,3,4 and 5. Reactor vessel water began to increase because Reactor feedwater pumps "A" & "B" were still operating. Reactor feedwater pump "B" was secured. When reactor vessel water increased to 173 inches, high level alarm, reactor feedwater pump "A" automatically tripped. Highest reactor vessel water level increased to was approximately 179 inches. No SRV's opened. All Emergency Core Cooling Systems, EDGs are fully operable if needed and the electrical grid is stable. Reactor vessel water level is being maintained using a reactor feedwater pump. Only other anomaly was that for some unknown reason automatic pressure control went to mechanical pressure control during the transient. Licensee is investigating the event. | |