Event Notification Report for March 09, 2000
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
03/08/2000 - 03/09/2000
EVENT NUMBERS
367763677736778367793678036781367823678336784
Power Reactor
Event Number: 36776
Facility: OCONEE
Region: 2 State: SC
Unit: [1] [2] [3]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-LP
NRC Notified By: MICHAEL HILL
HQ OPS Officer: DOUG WEAVER
Region: 2 State: SC
Unit: [1] [2] [3]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-LP
NRC Notified By: MICHAEL HILL
HQ OPS Officer: DOUG WEAVER
Notification Date: 03/09/2000
Notification Time: 11:16 [ET]
Event Date: 03/09/2000
Event Time: 10:35 [EST]
Last Update Date: 03/09/2000
Notification Time: 11:16 [ET]
Event Date: 03/09/2000
Event Time: 10:35 [EST]
Last Update Date: 03/09/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(1)(i)(A) - PLANT S/D REQD BY TS
10 CFR Section:
50.72(b)(1)(i)(A) - PLANT S/D REQD BY TS
Person (Organization):
EDWARD MCALPINE (R2)
EDWARD MCALPINE (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 |
| 3 | N | Y | 100 | Power Operation | 100 | Power Operation |
TECH SPEC REQUIRED SHUTDOWN - LOSS OF CONTROL ROOM AIR CONDITIONING
The licensee has two control room chillers. The 'A' chiller tripped because of low freon. The 'B' chiller was started and then tripped on two occasions. Based on the unreliability of the 'B' chiller, the licensee considered both chillers inoperable and entered Tech Spec 3.0.3 as a result. The licensee will start reducing power of all three units at 1 percent per hour at 1130. However they expect that they will have one chiller back in service prior to reaching the 2235, the time the units need to be in mode 3.
The licensee notified the NRC resident inspector.
* * * UPDATE ON 3/9/00 @ 1329 BY CLARKSON TO GOULD * * *
Actions were undertaken to restore the 'B' Chilled water train to operable status. At 1246 hours, on March 9, 2000, the 'B' Chilled Water train was returned to operable status and Technical Specifications 3.0.3 and 3.7.16, Condition E were exited. The shutdown of all three Oconee units was also halted at approximately 98% power and the units are being returned to 100%.
The NRC Resident Inspector was informed.
The Reg 2 RDO (McAlpine) was notified.
The licensee has two control room chillers. The 'A' chiller tripped because of low freon. The 'B' chiller was started and then tripped on two occasions. Based on the unreliability of the 'B' chiller, the licensee considered both chillers inoperable and entered Tech Spec 3.0.3 as a result. The licensee will start reducing power of all three units at 1 percent per hour at 1130. However they expect that they will have one chiller back in service prior to reaching the 2235, the time the units need to be in mode 3.
The licensee notified the NRC resident inspector.
* * * UPDATE ON 3/9/00 @ 1329 BY CLARKSON TO GOULD * * *
Actions were undertaken to restore the 'B' Chilled water train to operable status. At 1246 hours, on March 9, 2000, the 'B' Chilled Water train was returned to operable status and Technical Specifications 3.0.3 and 3.7.16, Condition E were exited. The shutdown of all three Oconee units was also halted at approximately 98% power and the units are being returned to 100%.
The NRC Resident Inspector was informed.
The Reg 2 RDO (McAlpine) was notified.
General Information or Other
Event Number: 36777
Rep Org: INSPECTION SERVICES COMPANY
Licensee: INSPECTION SERVICES COMPANY
Region: 3
City: PLYMOUTH State: MI
County:
License #: 21-26744-01
Agreement: N
Docket:
NRC Notified By: KRUZEL
HQ OPS Officer: CHAUNCEY GOULD
Licensee: INSPECTION SERVICES COMPANY
Region: 3
City: PLYMOUTH State: MI
County:
License #: 21-26744-01
Agreement: N
Docket:
NRC Notified By: KRUZEL
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 03/09/2000
Notification Time: 11:40 [ET]
Event Date: 03/09/2000
Event Time: 10:30 [EST]
Last Update Date: 03/09/2000
Notification Time: 11:40 [ET]
Event Date: 03/09/2000
Event Time: 10:30 [EST]
Last Update Date: 03/09/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
20.2201(a)(1)(i) - LOST/STOLEN LNM>1000X
10 CFR Section:
20.2201(a)(1)(i) - LOST/STOLEN LNM>1000X
Person (Organization):
ROGER LANKSBURY (R3)
JOHN HICKEY (NMSS)
ROGER LANKSBURY (R3)
JOHN HICKEY (NMSS)
STOLEN MOISTURE DENSITY GAUGE
INSPECTION SERVICES CO. REPORTED THAT A HUMBOLDT MOISTURE DENSITY GAUGE MODEL 5001 SERIAL # 2996 WAS STOLEN FROM THEIR PICKUP TRUCK WHILE IT WAS PARKED IN THEIR FACILITY PARKING LOT IN PLYMOUTH, MI. THE GAUGE WAS BOLTED TO THE TRUCK AT THE TIME OF THE THEFT. THE DEVICE CONTAINED 8 MILLICURIES OF Cs-137 AND 40 MILLICURIES OF Am-241. THE LOCAL POLICE WERE NOTIFIED.
INSPECTION SERVICES CO. REPORTED THAT A HUMBOLDT MOISTURE DENSITY GAUGE MODEL 5001 SERIAL # 2996 WAS STOLEN FROM THEIR PICKUP TRUCK WHILE IT WAS PARKED IN THEIR FACILITY PARKING LOT IN PLYMOUTH, MI. THE GAUGE WAS BOLTED TO THE TRUCK AT THE TIME OF THE THEFT. THE DEVICE CONTAINED 8 MILLICURIES OF Cs-137 AND 40 MILLICURIES OF Am-241. THE LOCAL POLICE WERE NOTIFIED.
General Information or Other
Event Number: 36778
Rep Org: PUBLIC SERVICE ELECTRIC AND GAS CO.
Licensee: BFGOODRICH COMPANY
Region: 1
City: HANCOCK'S BRIDGE State: NJ
County:
License #:
Agreement: N
Docket:
NRC Notified By: SUE SIMPSON
HQ OPS Officer: BOB STRANSKY
Licensee: BFGOODRICH COMPANY
Region: 1
City: HANCOCK'S BRIDGE State: NJ
County:
License #:
Agreement: N
Docket:
NRC Notified By: SUE SIMPSON
HQ OPS Officer: BOB STRANSKY
Notification Date: 03/09/2000
Notification Time: 12:15 [ET]
Event Date: 03/09/2000
Event Time: 00:00 [EST]
Last Update Date: 03/09/2000
Notification Time: 12:15 [ET]
Event Date: 03/09/2000
Event Time: 00:00 [EST]
Last Update Date: 03/09/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
Person (Organization):
GREG CRANSTON (R1)
VERN HODGE (via fax) (NRR)
GREG CRANSTON (R1)
VERN HODGE (via fax) (NRR)
10 CFR PART 21 REPORT
Replacement emergency diesel generator (EDG) air start motors were supplied to PSE&G with grease on the bendix drive shafts. This did not meet the requirements of the purchase order which stated: "Do not grease the bendix drive: as per the vendor manual this is to be lubricated with SAE 10 oil." Excessive grease on the bendix shaft could harden and prevent the drive pinion from engaging the ring gear on the EDG flywheel during starting.
The air start motors were supplied by:
BFGoodrich Company (formerly Coltec Industries)
Fairbanks Morse Engine Division
701 White Avenue
Beloit, WI 53511
Affected FMED Part #s: 21800136-1, 21800145, 21800146 and 21800147
Replacement emergency diesel generator (EDG) air start motors were supplied to PSE&G with grease on the bendix drive shafts. This did not meet the requirements of the purchase order which stated: "Do not grease the bendix drive: as per the vendor manual this is to be lubricated with SAE 10 oil." Excessive grease on the bendix shaft could harden and prevent the drive pinion from engaging the ring gear on the EDG flywheel during starting.
The air start motors were supplied by:
BFGoodrich Company (formerly Coltec Industries)
Fairbanks Morse Engine Division
701 White Avenue
Beloit, WI 53511
Affected FMED Part #s: 21800136-1, 21800145, 21800146 and 21800147
Power Reactor
Event Number: 36779
Facility: FT CALHOUN
Region: 4 State: NE
Unit: [1] [] []
RX Type: [1] CE
NRC Notified By: ERICK MATZKE
HQ OPS Officer: BOB STRANSKY
Region: 4 State: NE
Unit: [1] [] []
RX Type: [1] CE
NRC Notified By: ERICK MATZKE
HQ OPS Officer: BOB STRANSKY
Notification Date: 03/09/2000
Notification Time: 12:42 [ET]
Event Date: 03/09/2000
Event Time: 10:54 [CST]
Last Update Date: 03/09/2000
Notification Time: 12:42 [ET]
Event Date: 03/09/2000
Event Time: 10:54 [CST]
Last Update Date: 03/09/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(1)(ii)(B) - OUTSIDE DESIGN BASIS
10 CFR Section:
50.72(b)(1)(ii)(B) - OUTSIDE DESIGN BASIS
Person (Organization):
KRISS KENNEDY (R4)
KRISS KENNEDY (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 |
CONTROL ROOM DOSE RATE EXCEEDS STANDARD REVIEW PLAN LIMIT
"The Fort Calhoun Station is in the process of revising and reanalyzing the radiological consequences analysis for the plant as a result of corrective actions for LERs 1998-012 and 1998-013. In the process it has been determined that some of the assumptions made when the plant was constructed in the early 1970s, about 30 years ago, need to be validated. An important assumption that was brought into question was that the Control Room has zero (0) standard cubic feet per minute (SCFM) unfiltered inleakage. The control room envelope was recently tested and the unfiltered inleakage was determined to be very small at 8 SCFM. This level of leakage leads to a thyroid dose to the operators of about 32 rem. The Fort Calhoun Station USAR commits the station to NRC Standard Review Plan (SRP) section 6.4 which states that the thyroid dose to the operators shall be no higher than 30 rem. Higher thyroid doses have been approved for other plants and the dose projected certainly would not incapacitate an operator. The calculated dose is in excess of the current plant design basis.
"Plant actions to date have included putting in place an Safety Analysis for Operability (SAO) (similar to a Justification for Continued Operation [JCO]) to allow the plant to continue operating in this condition. The plant has been in communication with the NRC staff on the progress of this radiological consequences reanalysis."
The NRC resident inspector has been informed of this report by the licensee.
"The Fort Calhoun Station is in the process of revising and reanalyzing the radiological consequences analysis for the plant as a result of corrective actions for LERs 1998-012 and 1998-013. In the process it has been determined that some of the assumptions made when the plant was constructed in the early 1970s, about 30 years ago, need to be validated. An important assumption that was brought into question was that the Control Room has zero (0) standard cubic feet per minute (SCFM) unfiltered inleakage. The control room envelope was recently tested and the unfiltered inleakage was determined to be very small at 8 SCFM. This level of leakage leads to a thyroid dose to the operators of about 32 rem. The Fort Calhoun Station USAR commits the station to NRC Standard Review Plan (SRP) section 6.4 which states that the thyroid dose to the operators shall be no higher than 30 rem. Higher thyroid doses have been approved for other plants and the dose projected certainly would not incapacitate an operator. The calculated dose is in excess of the current plant design basis.
"Plant actions to date have included putting in place an Safety Analysis for Operability (SAO) (similar to a Justification for Continued Operation [JCO]) to allow the plant to continue operating in this condition. The plant has been in communication with the NRC staff on the progress of this radiological consequences reanalysis."
The NRC resident inspector has been informed of this report by the licensee.
Power Reactor
Event Number: 36780
Facility: SAINT LUCIE
Region: 2 State: FL
Unit: [1] [2] []
RX Type: [1] CE,[2] CE
NRC Notified By: CALVIN WARD
HQ OPS Officer: BOB STRANSKY
Region: 2 State: FL
Unit: [1] [2] []
RX Type: [1] CE,[2] CE
NRC Notified By: CALVIN WARD
HQ OPS Officer: BOB STRANSKY
Notification Date: 03/09/2000
Notification Time: 14:54 [ET]
Event Date: 03/09/2000
Event Time: 13:47 [EST]
Last Update Date: 03/09/2000
Notification Time: 14:54 [ET]
Event Date: 03/09/2000
Event Time: 13:47 [EST]
Last Update Date: 03/09/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(1)(ii)(B) - OUTSIDE DESIGN BASIS
10 CFR Section:
50.72(b)(1)(ii)(B) - OUTSIDE DESIGN BASIS
Person (Organization):
EDWARD MCALPINE (R2)
EDWARD MCALPINE (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 |
UNITS OUTSIDE OF DESIGN BASIS DUE TO 10 CFR PART 50, APPENDIX R ISSUE
Following is the abstract of an engineering document submitted to the NRC Operations Center:
"As part of the continuing resolution of fire protection program issues identified during a Follow-up Fire Protection Functional Inspection at St. Lucie in December 1999, a detailed review of Appendix R circuit separation and protection for the Units 1 & 2 containment was performed. This review considered the design and licensing basis for these features as described in the Unit 1 & 2 Updated Final Safety Analysis Reports (UFSARs), associated NRC Safety Evaluation Reports (SERs), Appendix R requirements and guidance provided in Generic Letter 86-10. The review has identified a number of separation issues for various safe shutdown circuits and/or components credited for post-fire safe shutdown operations following a fire in containment. The affected equipment is hi/lo pressure interface equipment along with control and indication equipment associated with the reactor coolant system."
The discrepancies noted in the report affect the power operated relief valves (PORVs), as well as pressurizer pressure and level instrumentation. The licensee has determined that the conditions do not affect operability of the affected equipment due to the low combustible material loading inside the containment and the low probability that a fire would occur inside the containment. The NRC resident inspector has been informed of this report by the licensee.
Following is the abstract of an engineering document submitted to the NRC Operations Center:
"As part of the continuing resolution of fire protection program issues identified during a Follow-up Fire Protection Functional Inspection at St. Lucie in December 1999, a detailed review of Appendix R circuit separation and protection for the Units 1 & 2 containment was performed. This review considered the design and licensing basis for these features as described in the Unit 1 & 2 Updated Final Safety Analysis Reports (UFSARs), associated NRC Safety Evaluation Reports (SERs), Appendix R requirements and guidance provided in Generic Letter 86-10. The review has identified a number of separation issues for various safe shutdown circuits and/or components credited for post-fire safe shutdown operations following a fire in containment. The affected equipment is hi/lo pressure interface equipment along with control and indication equipment associated with the reactor coolant system."
The discrepancies noted in the report affect the power operated relief valves (PORVs), as well as pressurizer pressure and level instrumentation. The licensee has determined that the conditions do not affect operability of the affected equipment due to the low combustible material loading inside the containment and the low probability that a fire would occur inside the containment. The NRC resident inspector has been informed of this report by the licensee.
Power Reactor
Event Number: 36781
Facility: DIABLO CANYON
Region: 4 State: CA
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: CHITWOOD
HQ OPS Officer: BOB STRANSKY
Region: 4 State: CA
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: CHITWOOD
HQ OPS Officer: BOB STRANSKY
Notification Date: 03/09/2000
Notification Time: 14:56 [ET]
Event Date: 03/09/2000
Event Time: 09:18 [PST]
Last Update Date: 03/09/2000
Notification Time: 14:56 [ET]
Event Date: 03/09/2000
Event Time: 09:18 [PST]
Last Update Date: 03/09/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(ii) - ESF ACTUATION
10 CFR Section:
50.72(b)(2)(ii) - ESF ACTUATION
Person (Organization):
KRISS KENNEDY (R4)
KRISS KENNEDY (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 |
CONTROL ROOM VENTILATION PLACED IN PRESSURIZATION MODE DUE TO SMOKE
At 0906 PST, rod drive motor/generator set 1-2 tripped due to overcurrent. Heavy smoke from burnt insulation was observed in the vicinity. In order to prevent the smoke from entering the control room, the shift foreman directed operators to manually place the control room ventilation system in pressurization mode (mode 4).
The NRC resident inspector will be informed of this report by the licensee.
At 0906 PST, rod drive motor/generator set 1-2 tripped due to overcurrent. Heavy smoke from burnt insulation was observed in the vicinity. In order to prevent the smoke from entering the control room, the shift foreman directed operators to manually place the control room ventilation system in pressurization mode (mode 4).
The NRC resident inspector will be informed of this report by the licensee.
Fuel Cycle Facility
Event Number: 36782
Facility: PORTSMOUTH GASEOUS DIFFUSION PLANT
Region: 3 State: OH
Unit: [] [] []
RX Type: URANIUM ENRICHMENT FACILITY
Comments: 2 DEMOCRACY CENTER
6903 ROCKLEDGE DRIVE
BETHESDA, MD 20817 (301)564-3200
NRC Notified By: JEFF CASTLE
HQ OPS Officer: BOB STRANSKY
Region: 3 State: OH
Unit: [] [] []
RX Type: URANIUM ENRICHMENT FACILITY
Comments: 2 DEMOCRACY CENTER
6903 ROCKLEDGE DRIVE
BETHESDA, MD 20817 (301)564-3200
NRC Notified By: JEFF CASTLE
HQ OPS Officer: BOB STRANSKY
Notification Date: 03/09/2000
Notification Time: 17:04 [ET]
Event Date: 03/09/2000
Event Time: 03:32 [EST]
Last Update Date: 03/11/2000
Notification Time: 17:04 [ET]
Event Date: 03/09/2000
Event Time: 03:32 [EST]
Last Update Date: 03/11/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
ROGER LANKSBURY (R3)
THOMAS ESSIG (NMSS)
ROGER LANKSBURY (R3)
THOMAS ESSIG (NMSS)
24-HOUR NRC BULLETIN 91-01 REPORT
"At 0332 hours on 03/09/2000, it was identified that an error had been made in connecting a cell calibration test buggy, covered by NCSA-PLANT069, at X-330 building cell 31-3-9. The cell was connected to the HI DAT (HI DATUM) port instead of the PROCESS inlet port. The NCSA identifies installed chemical traps on the test buggy as a passive design characteristic relied upon to prevent an accumulation of uranium in the vacuum pump oil. The installed traps are also a control contingency in preventing back flow of vacuum pump oil to the cell manifold. Connection of the process system to the HI DAT port bypassed these chemical traps and provided a direct flow path between the process gas system and vacuum pump.
"Establishing a connection between the vacuum pump and cell process gas manifold constitutes a loss of one control of the double contingency control principle. The cell calibration buggy was disconnected and it was determined that no oil from the vacuum pump had migrated through the buggy to the cell manifold. The potential for a criticality to occur is precluded based on the amount of oil contained in the vacuum pump and by the assay of the material at that point of the cascade. Pull compliance with NCSA-PLANT069 was regained when the calibration buggy was disconnected from the cell manifold.
"There was no loss of hazardous/radioactive material or radioactive/radiological contamination exposure as a result of this event.
"SAFETY SIGNIFICANCE OF EVENTS:
"This event has a low safety significance. Due to operator error, the cell was connected to the wrong inlet port This allowed the possibility that process gas could bypass the chemical trap(s) and then collect in the oil reservoir of the vacuum pump. The oil reservoir (limited to <= one quart) is sized such that it is safe for 100% enriched material. The process gas that may have entered the oil reservoir of the pump is approximately 2.25% enriched. There is insufficient oil In the vacuum pump for a criticality to occur. Additionally the test buggy has been disconnected from the cell. Thus, there is no possibility of adding additional uranium to the oil reservoir of the vacuum pump.
"POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO(S) OF HOW CRITICALITY COULD OCCUR):
"For a criticality to occur, sufficient UF6 would need to collect in the oil reservoir of the vacuum pump and the vacuum pump would need to be replaced with a different model such that the oil reservoir is large enough (greater than 3 quarts) for a criticality to occur.
"CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.):
"This NCSA relied on preventing the accumulation of process gas from collecting in the vacuum pump oil by placing a chemical trap upstream of the vacuum pump. This control was lost. The second control was to limit the amount of oil in the vacuum pump to less than 1 quart, which is less than the minimum volume of oil required for a criticality at 100% U235.
"ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS LIMIT AND % WORST CASE OF CRITICAL MASS):
"At this time it is not known if any UF6 reached the oil reservoir for the vacuum pump. Enrichment in cell 31-3-9 is estimated to be 2.25% U-235. If process gas reached the oil reservoir, it would be in the form of UF4/oil mixture.
"NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES
"Nuclear criticality safety is maintained by two controls, the first Is the placement of chemical traps upstream of the vacuum pump. This control was lost. The second control was on the allowed volume of oil in the reservoir of the vacuum pump. This control was maintained throughout the event. Thus, one control relied on for double contingency was lost."
The NRC resident inspector has been informed of this notification.
* * * UPDATE ON 03/11/00 AT 0128 ET FROM ERIC SPAETH TAKEN BY MACKINNON * * *
It was discovered that the internal volume of the oil reservoir in the vacuum pump was greater than the NCSA control limit of less than or equal to one quart. The actual volume of the reservoir is approximately one and one third quart (The manufacturer recommends that the vacuum pump oil level should be 1.33 quarts). At the time of the initial notification, it was not known that the oil reservoir was greater than the NCSA controlled limit. However, this volume is still within the three quart volume analyzed in the NCSE for the normal case condition. Therefore, the safety significance of this event remains low. The Certificate holder said that the NCSA limit for oil will be increased.
CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.):
The second control parameter was also lost which was to limit the oil level in the vacuum pump to less than one quart.
NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES:
The second control was lost (oil level greater than one quart) when it was discovered that the volume of the oil reservoir exceeded the allowed volume in the NCSA. Even though the volume of the reservoir exceeded the allowed amount it was still within the three quart limit analyzed in the NCSE for the normal case condition.
R3DO (Roger Lanksbury) and NMSS (Tom Essig) notified.
The NRC Resident Inspector was notified of this event update by the certificate holder.
"At 0332 hours on 03/09/2000, it was identified that an error had been made in connecting a cell calibration test buggy, covered by NCSA-PLANT069, at X-330 building cell 31-3-9. The cell was connected to the HI DAT (HI DATUM) port instead of the PROCESS inlet port. The NCSA identifies installed chemical traps on the test buggy as a passive design characteristic relied upon to prevent an accumulation of uranium in the vacuum pump oil. The installed traps are also a control contingency in preventing back flow of vacuum pump oil to the cell manifold. Connection of the process system to the HI DAT port bypassed these chemical traps and provided a direct flow path between the process gas system and vacuum pump.
"Establishing a connection between the vacuum pump and cell process gas manifold constitutes a loss of one control of the double contingency control principle. The cell calibration buggy was disconnected and it was determined that no oil from the vacuum pump had migrated through the buggy to the cell manifold. The potential for a criticality to occur is precluded based on the amount of oil contained in the vacuum pump and by the assay of the material at that point of the cascade. Pull compliance with NCSA-PLANT069 was regained when the calibration buggy was disconnected from the cell manifold.
"There was no loss of hazardous/radioactive material or radioactive/radiological contamination exposure as a result of this event.
"SAFETY SIGNIFICANCE OF EVENTS:
"This event has a low safety significance. Due to operator error, the cell was connected to the wrong inlet port This allowed the possibility that process gas could bypass the chemical trap(s) and then collect in the oil reservoir of the vacuum pump. The oil reservoir (limited to <= one quart) is sized such that it is safe for 100% enriched material. The process gas that may have entered the oil reservoir of the pump is approximately 2.25% enriched. There is insufficient oil In the vacuum pump for a criticality to occur. Additionally the test buggy has been disconnected from the cell. Thus, there is no possibility of adding additional uranium to the oil reservoir of the vacuum pump.
"POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO(S) OF HOW CRITICALITY COULD OCCUR):
"For a criticality to occur, sufficient UF6 would need to collect in the oil reservoir of the vacuum pump and the vacuum pump would need to be replaced with a different model such that the oil reservoir is large enough (greater than 3 quarts) for a criticality to occur.
"CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.):
"This NCSA relied on preventing the accumulation of process gas from collecting in the vacuum pump oil by placing a chemical trap upstream of the vacuum pump. This control was lost. The second control was to limit the amount of oil in the vacuum pump to less than 1 quart, which is less than the minimum volume of oil required for a criticality at 100% U235.
"ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS LIMIT AND % WORST CASE OF CRITICAL MASS):
"At this time it is not known if any UF6 reached the oil reservoir for the vacuum pump. Enrichment in cell 31-3-9 is estimated to be 2.25% U-235. If process gas reached the oil reservoir, it would be in the form of UF4/oil mixture.
"NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES
"Nuclear criticality safety is maintained by two controls, the first Is the placement of chemical traps upstream of the vacuum pump. This control was lost. The second control was on the allowed volume of oil in the reservoir of the vacuum pump. This control was maintained throughout the event. Thus, one control relied on for double contingency was lost."
The NRC resident inspector has been informed of this notification.
* * * UPDATE ON 03/11/00 AT 0128 ET FROM ERIC SPAETH TAKEN BY MACKINNON * * *
It was discovered that the internal volume of the oil reservoir in the vacuum pump was greater than the NCSA control limit of less than or equal to one quart. The actual volume of the reservoir is approximately one and one third quart (The manufacturer recommends that the vacuum pump oil level should be 1.33 quarts). At the time of the initial notification, it was not known that the oil reservoir was greater than the NCSA controlled limit. However, this volume is still within the three quart volume analyzed in the NCSE for the normal case condition. Therefore, the safety significance of this event remains low. The Certificate holder said that the NCSA limit for oil will be increased.
CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.):
The second control parameter was also lost which was to limit the oil level in the vacuum pump to less than one quart.
NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES:
The second control was lost (oil level greater than one quart) when it was discovered that the volume of the oil reservoir exceeded the allowed volume in the NCSA. Even though the volume of the reservoir exceeded the allowed amount it was still within the three quart limit analyzed in the NCSE for the normal case condition.
R3DO (Roger Lanksbury) and NMSS (Tom Essig) notified.
The NRC Resident Inspector was notified of this event update by the certificate holder.
Power Reactor
Event Number: 36783
Facility: NORTH ANNA
Region: 2 State: VA
Unit: [1] [2] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: RICHARD WESLEY
HQ OPS Officer: BOB STRANSKY
Region: 2 State: VA
Unit: [1] [2] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: RICHARD WESLEY
HQ OPS Officer: BOB STRANSKY
Notification Date: 03/10/2000
Notification Time: 13:01 [ET]
Event Date: 03/09/2000
Event Time: 14:30 [EST]
Last Update Date: 03/10/2000
Notification Time: 13:01 [ET]
Event Date: 03/09/2000
Event Time: 14:30 [EST]
Last Update Date: 03/10/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(vi) - OFFSITE NOTIFICATION
10 CFR Section:
50.72(b)(2)(vi) - OFFSITE NOTIFICATION
Person (Organization):
EDWARD MCALPINE (R2)
EDWARD MCALPINE (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 |
NOTIFICATION TO FEDERAL ENERGY REGULATORY COMMISSION
The licensee notified the Federal Energy Regulatory Commission (FERC) regarding the inoperability of the spillway diesel generator. The diesel generator was inoperable while batteries were being replaced. The diesel generator is now fully operable.
The NRC resident inspector has been informed of this notification by the licensee.
The licensee notified the Federal Energy Regulatory Commission (FERC) regarding the inoperability of the spillway diesel generator. The diesel generator was inoperable while batteries were being replaced. The diesel generator is now fully operable.
The NRC resident inspector has been informed of this notification by the licensee.
Power Reactor
Event Number: 36784
Facility: CATAWBA
Region: 2 State: SC
Unit: [1] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: DON BRADLEY
HQ OPS Officer: BOB STRANSKY
Region: 2 State: SC
Unit: [1] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: DON BRADLEY
HQ OPS Officer: BOB STRANSKY
Notification Date: 03/10/2000
Notification Time: 17:58 [ET]
Event Date: 03/09/2000
Event Time: 22:57 [EST]
Last Update Date: 03/10/2000
Notification Time: 17:58 [ET]
Event Date: 03/09/2000
Event Time: 22:57 [EST]
Last Update Date: 03/10/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
EDWARD MCALPINE (R2)
EDWARD MCALPINE (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 | 95 | Power Operation | 95 | Power Operation |
COURTESY NOTIFICATION REGARDING ONSITE FATALITY
At 2357 on 3/9/2000, a Duke temporary employee was pronounced dead at the Piedmont Medical Center. The individual had collapsed at the site, and attempts to resuscitate the individual were unsuccessful. The licensee made a courtesy notification to the Occupational Safety and Health Administration (OSHA) regarding this occurrence.
The NRC resident inspector has been informed of this notification by the licensee.
At 2357 on 3/9/2000, a Duke temporary employee was pronounced dead at the Piedmont Medical Center. The individual had collapsed at the site, and attempts to resuscitate the individual were unsuccessful. The licensee made a courtesy notification to the Occupational Safety and Health Administration (OSHA) regarding this occurrence.
The NRC resident inspector has been informed of this notification by the licensee.