Event Notification Report for January 21, 2000
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
01/20/2000 - 01/21/2000
EVENT NUMBERS
366103661236613366143661536627
Power Reactor
Event Number: 36610
Facility: POINT BEACH
Region: 3 State: WI
Unit: [1] [] []
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: JOHN SELL
HQ OPS Officer: STEVE SANDIN
Region: 3 State: WI
Unit: [1] [] []
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: JOHN SELL
HQ OPS Officer: STEVE SANDIN
Notification Date: 01/21/2000
Notification Time: 04:21 [ET]
Event Date: 01/21/2000
Event Time: 02:39 [CST]
Last Update Date: 01/24/2000
Notification Time: 04:21 [ET]
Event Date: 01/21/2000
Event Time: 02:39 [CST]
Last Update Date: 01/24/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(1)(i)(A) - PLANT S/D REQD BY TS 50.72(b)(2)(ii) - RPS ACTUATION
10 CFR Section:
50.72(b)(1)(i)(A) - PLANT S/D REQD BY TS 50.72(b)(2)(ii) - RPS ACTUATION
Person (Organization):
DAVID HILLS (R3)
DAVID HILLS (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | M/R | Y | 40 | Power Operation | 0 | Hot Standby |
UNIT 1 MANUALLY TRIPPED FROM 40% POWER DUE TO A DECREASING FOREBAY LEVEL
"UNIT 1 WAS MANUALLY TRIPPED FROM 40% POWER AFTER AN ANTICIPATORY POWER REDUCTION AT 0239 CST DUE TO DECREASING LEVEL IN THE CIRC. WATER INLET FOREBAY. DECREASING LEVEL WAS APPARENTLY DUE TO ICING OF THE INTAKE STRUCTURE. UNIT 1 CIRC. WATER SYSTEM HAS BEEN SECURED AND FOREBAY LEVEL HAS INCREASED. THE UNIT 1 CIRC. WATER SYSTEM WAS SECURED TO MAINTAIN OPERABILITY OF THE SERVICE WATER SYSTEM. UNIT 2 REMAINS AT 100% POWER AND IS STABLE."
FOREBAY LEVEL HAD DECREASED TO -11 FT FROM THE NORMAL OPERATING -8 FT. THE -11 FT CORRESPONDS TO THE MINIMUM DESIGN LIMIT FOR THE SERVICE WATER SYSTEM. CURRENTLY, LEVEL IS -6 FT 4 INCHES. STEAM GENERATORS USING THE ATMOSPHERIC DUMPS ARE PROVIDING THE HEAT SINK FOR DECAY HEAT REMOVAL SINCE THE CONDENSER IS NOT IN SERVICE. ALL CONTROL RODS FULLY INSERTED FOLLOWING THE MANUAL TRIP.
THE LICENSEE WILL INFORM THE NRC RESIDENT INSPECTOR.
* * * UPDATE ON 01/21/00 @ 1339 EST BY KRAUSE TO GOULD * * *
THE LICENSEE UPDATED THIS EVENT BY ALSO CLASSIFYING IT AS A RPS ACTUATION DUE TO THE MANUAL REACTOR TRIP.
THE NRC RESIDENT INSPECTOR WILL BE NOTIFIED.
THE R3DO (HILLS) WAS NOTIFIED.
* * * UPDATE ON 01/24/00 @ 1201 EST BY KRAUSE TO JONES * * *
The licensee performed an engineering evaluation that demonstrated that at no time was the design basis water level for service water exceeded. The report of the Tech Spec required shutdown is retracted, but the report of the RPS activation stands as previously updated.
The licensee notified the NRC Resident Inspector.
The R3DO (Monte Phillips) has been notified.
"UNIT 1 WAS MANUALLY TRIPPED FROM 40% POWER AFTER AN ANTICIPATORY POWER REDUCTION AT 0239 CST DUE TO DECREASING LEVEL IN THE CIRC. WATER INLET FOREBAY. DECREASING LEVEL WAS APPARENTLY DUE TO ICING OF THE INTAKE STRUCTURE. UNIT 1 CIRC. WATER SYSTEM HAS BEEN SECURED AND FOREBAY LEVEL HAS INCREASED. THE UNIT 1 CIRC. WATER SYSTEM WAS SECURED TO MAINTAIN OPERABILITY OF THE SERVICE WATER SYSTEM. UNIT 2 REMAINS AT 100% POWER AND IS STABLE."
FOREBAY LEVEL HAD DECREASED TO -11 FT FROM THE NORMAL OPERATING -8 FT. THE -11 FT CORRESPONDS TO THE MINIMUM DESIGN LIMIT FOR THE SERVICE WATER SYSTEM. CURRENTLY, LEVEL IS -6 FT 4 INCHES. STEAM GENERATORS USING THE ATMOSPHERIC DUMPS ARE PROVIDING THE HEAT SINK FOR DECAY HEAT REMOVAL SINCE THE CONDENSER IS NOT IN SERVICE. ALL CONTROL RODS FULLY INSERTED FOLLOWING THE MANUAL TRIP.
THE LICENSEE WILL INFORM THE NRC RESIDENT INSPECTOR.
* * * UPDATE ON 01/21/00 @ 1339 EST BY KRAUSE TO GOULD * * *
THE LICENSEE UPDATED THIS EVENT BY ALSO CLASSIFYING IT AS A RPS ACTUATION DUE TO THE MANUAL REACTOR TRIP.
THE NRC RESIDENT INSPECTOR WILL BE NOTIFIED.
THE R3DO (HILLS) WAS NOTIFIED.
* * * UPDATE ON 01/24/00 @ 1201 EST BY KRAUSE TO JONES * * *
The licensee performed an engineering evaluation that demonstrated that at no time was the design basis water level for service water exceeded. The report of the Tech Spec required shutdown is retracted, but the report of the RPS activation stands as previously updated.
The licensee notified the NRC Resident Inspector.
The R3DO (Monte Phillips) has been notified.
Power Reactor
Event Number: 36612
Facility: OYSTER CREEK
Region: 1 State: NJ
Unit: [1] [] []
RX Type: [1] GE-2
NRC Notified By: GLENN HUTTON
HQ OPS Officer: FANGIE JONES
Region: 1 State: NJ
Unit: [1] [] []
RX Type: [1] GE-2
NRC Notified By: GLENN HUTTON
HQ OPS Officer: FANGIE JONES
Notification Date: 01/21/2000
Notification Time: 11:17 [ET]
Event Date: 01/21/2000
Event Time: 10:48 [EST]
Last Update Date: 01/24/2000
Notification Time: 11:17 [ET]
Event Date: 01/21/2000
Event Time: 10:48 [EST]
Last Update Date: 01/24/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(ii) - RPS ACTUATION
10 CFR Section:
50.72(b)(2)(ii) - RPS ACTUATION
Person (Organization):
FELICIA HINSON (R1)
FELICIA HINSON (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | M/R | Y | 65 | Power Operation | 0 | Hot Shutdown |
MANUAL REACTOR SCRAM PER PROCEDURE FOLLOWING TRIP OF ALL RECIRCULATION PUMPS
During the performance of a surveillance test, with a half scram signal inserted, a spurious trip signal from the other system initiated a trip of all the reactor coolant system recirculation pumps. The operators manually scrammed the reactor per procedure, all rods fully inserted. Heat removal is via the main condenser with main feedwater feeding the reactor. All systems functioned as designed and the plant is proceeding to cold shutdown. A trip review will be performed by the licensee.
The licensee notified the NRC Resident Inspector.
* * * UPDATE AT 1943 EST ON 01/24/00 BY FRANK CIGANIK TO FANGIE JONES * * *
On 01/21/00, the licensee inserted a manual scram in accordance with station procedures in response to a trip of all the reactor recirculation pumps. The trip of all the pumps was due to an error in the performance of the ongoing surveillance test.
"During the performance of an instrument logic surveillance test on the isolation condenser actuation logic which also inserts trip signals into the recirculation pump trip logic, the reactor recirculation pumps tripped. After testing the first instrument, the trip signal associated with the isolation condensers were reset; however, the recirc pump trip logic was not properly reset before proceeding to the second instrument. When a test signal was introduced on this instrument, the reactor recirculation pumps tripped as designed. In response to the pump trip, the operators manually scrammed the reactor. The plant is currently in cold shutdown. A trip review has been performed. Oyster Creek plans to take advantage of this shutdown to perform some other maintenance activities."
The original report should have read "half trip" rather than "half scram" referring to the trip logic of the reactor recirculation pumps; this was entered wrong the first time.
The licensee notified the NRC Resident Inspector. The R1DO (John Rogge) has been notified.
During the performance of a surveillance test, with a half scram signal inserted, a spurious trip signal from the other system initiated a trip of all the reactor coolant system recirculation pumps. The operators manually scrammed the reactor per procedure, all rods fully inserted. Heat removal is via the main condenser with main feedwater feeding the reactor. All systems functioned as designed and the plant is proceeding to cold shutdown. A trip review will be performed by the licensee.
The licensee notified the NRC Resident Inspector.
* * * UPDATE AT 1943 EST ON 01/24/00 BY FRANK CIGANIK TO FANGIE JONES * * *
On 01/21/00, the licensee inserted a manual scram in accordance with station procedures in response to a trip of all the reactor recirculation pumps. The trip of all the pumps was due to an error in the performance of the ongoing surveillance test.
"During the performance of an instrument logic surveillance test on the isolation condenser actuation logic which also inserts trip signals into the recirculation pump trip logic, the reactor recirculation pumps tripped. After testing the first instrument, the trip signal associated with the isolation condensers were reset; however, the recirc pump trip logic was not properly reset before proceeding to the second instrument. When a test signal was introduced on this instrument, the reactor recirculation pumps tripped as designed. In response to the pump trip, the operators manually scrammed the reactor. The plant is currently in cold shutdown. A trip review has been performed. Oyster Creek plans to take advantage of this shutdown to perform some other maintenance activities."
The original report should have read "half trip" rather than "half scram" referring to the trip logic of the reactor recirculation pumps; this was entered wrong the first time.
The licensee notified the NRC Resident Inspector. The R1DO (John Rogge) has been notified.
Power Reactor
Event Number: 36613
Facility: OYSTER CREEK
Region: 1 State: NJ
Unit: [1] [] []
RX Type: [1] GE-2
NRC Notified By: BROWNE
HQ OPS Officer: CHAUNCEY GOULD
Region: 1 State: NJ
Unit: [1] [] []
RX Type: [1] GE-2
NRC Notified By: BROWNE
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 01/21/2000
Notification Time: 14:26 [ET]
Event Date: 01/21/2000
Event Time: 12:00 [EST]
Last Update Date: 01/21/2000
Notification Time: 14:26 [ET]
Event Date: 01/21/2000
Event Time: 12:00 [EST]
Last Update Date: 01/21/2000
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):
FELICIA HINSON (R1)
FELICIA HINSON (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Hot Shutdown | 0 | Hot Shutdown |
THE LICENSEE NOTIFIED THE NEW JERSEY DEPARTMENT OF ENVIRONMENTAL PROTECTION OF A FISH KILL.
THERE WAS A FISH KILL IN THE DISCHARGE CANAL OF MORE THAN 100 FISH (SEVERAL SPECIES) DUE TO COLD STRESS. THIS OCCURRED FOLLOWING THE REACTOR SCRAM WHEN COOLER WATER WAS DISCHARGED TO THE CANAL DECREASING ITS TEMPERATURE.
THE RESIDENT INSPECTOR WILL BE NOTIFIED.
THERE WAS A FISH KILL IN THE DISCHARGE CANAL OF MORE THAN 100 FISH (SEVERAL SPECIES) DUE TO COLD STRESS. THIS OCCURRED FOLLOWING THE REACTOR SCRAM WHEN COOLER WATER WAS DISCHARGED TO THE CANAL DECREASING ITS TEMPERATURE.
THE RESIDENT INSPECTOR WILL BE NOTIFIED.
Power Reactor
Event Number: 36614
Facility: BEAVER VALLEY
Region: 1 State: PA
Unit: [1] [2] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: THOMAS
HQ OPS Officer: CHAUNCEY GOULD
Region: 1 State: PA
Unit: [1] [2] []
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: THOMAS
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 01/21/2000
Notification Time: 20:05 [ET]
Event Date: 01/21/2000
Event Time: 19:45 [EST]
Last Update Date: 01/21/2000
Notification Time: 20:05 [ET]
Event Date: 01/21/2000
Event Time: 19:45 [EST]
Last Update Date: 01/21/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(1)(ii)(A) - UNANALYZED COND OP
10 CFR Section:
50.72(b)(1)(ii)(A) - UNANALYZED COND OP
Person (Organization):
CHRIS CAHILL (R1)
CHRIS CAHILL (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 |
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
INSUFFICIENT GUIDANCE FOR ENSURING THAT A SAFETY RELATED SYSTEM WAS OPERATING DURING A LOCA
During the evaluation of a separate Condition Report concerning a Technical Specification 3.0.3 entry involving the Beaver Valley Power Station (BVPS) Unit 1 Supplemental Leak Collection & Release System (SLCRS), it was discovered that the operators did not have sufficient guidance for ensuring that this safety related system was operating following an accident.
BVPS Units 1 and 2 Technical Specifications 3/4.7.8 Bases states "System operation was assumed in that portion of the design basis accident LOCA analysis which addressed ESF leakage following the LOCA Based on the results of the analyses, the SLCRS must be OPERABLE to ensure that ESF leakage following the postulated DBA LOCA will not exceed 10CFR 100 limits." ESF leakage is a concern during a LOCA following a Safety Injection Transfer to Recirculation.
The BVPS Unit 1 and 2 operators did not have sufficient guidance to ensure that a SLCRS fan is operating during a LOCA after Transfer to Recirculation occurs. Thus, the SLCRS may not operate as assumed and credited in LOCA safety analyses. SLCRS also provides cooling air flow to the emergency core cooling pumps motors. The need for SLCRS for maintaining the qualification of emergency core cooling pumps motors is proportional to the ambient outdoor air temperatures. Current guidance does not ensure SLCRS would be available in all conditions where SLCRS flow is required to maintain the qualification of the emergency core cooling pump motors.
This condition is an unanalyzed condition and is reportable pursuant to 10CFR50.72(b)(1)(ii).
Verbal guidance has been provided to the BVPS Unit 1 and 2 operating crews to ensure a SLCRS fan is running in the event of a reactor trip or safety injection. This guidance is currently being formalized.
The NRC Resident Inspector was notified.
During the evaluation of a separate Condition Report concerning a Technical Specification 3.0.3 entry involving the Beaver Valley Power Station (BVPS) Unit 1 Supplemental Leak Collection & Release System (SLCRS), it was discovered that the operators did not have sufficient guidance for ensuring that this safety related system was operating following an accident.
BVPS Units 1 and 2 Technical Specifications 3/4.7.8 Bases states "System operation was assumed in that portion of the design basis accident LOCA analysis which addressed ESF leakage following the LOCA Based on the results of the analyses, the SLCRS must be OPERABLE to ensure that ESF leakage following the postulated DBA LOCA will not exceed 10CFR 100 limits." ESF leakage is a concern during a LOCA following a Safety Injection Transfer to Recirculation.
The BVPS Unit 1 and 2 operators did not have sufficient guidance to ensure that a SLCRS fan is operating during a LOCA after Transfer to Recirculation occurs. Thus, the SLCRS may not operate as assumed and credited in LOCA safety analyses. SLCRS also provides cooling air flow to the emergency core cooling pumps motors. The need for SLCRS for maintaining the qualification of emergency core cooling pumps motors is proportional to the ambient outdoor air temperatures. Current guidance does not ensure SLCRS would be available in all conditions where SLCRS flow is required to maintain the qualification of the emergency core cooling pump motors.
This condition is an unanalyzed condition and is reportable pursuant to 10CFR50.72(b)(1)(ii).
Verbal guidance has been provided to the BVPS Unit 1 and 2 operating crews to ensure a SLCRS fan is running in the event of a reactor trip or safety injection. This guidance is currently being formalized.
The NRC Resident Inspector was notified.
Power Reactor
Event Number: 36615
Facility: MILLSTONE
Region: 1 State: CT
Unit: [] [] [3]
RX Type: [1] GE-3,[2] CE,[3] W-4-LP
NRC Notified By: JAMES V. GROGAN
HQ OPS Officer: JOHN MacKINNON
Region: 1 State: CT
Unit: [] [] [3]
RX Type: [1] GE-3,[2] CE,[3] W-4-LP
NRC Notified By: JAMES V. GROGAN
HQ OPS Officer: JOHN MacKINNON
Notification Date: 01/21/2000
Notification Time: 23:33 [ET]
Event Date: 01/21/2000
Event Time: 23:13 [EST]
Last Update Date: 01/21/2000
Notification Time: 23:33 [ET]
Event Date: 01/21/2000
Event Time: 23:13 [EST]
Last Update Date: 01/21/2000
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):
CHRIS CAHILL (R1)
CHRIS CAHILL (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | N | Y | 100 | Power Operation | 100 | Power Operation |
ULTIMATE HEAT SINK (UHS) BELOW MINIMUM DESIGN TEMPERATURE LIMIT OF 33 DEGREES F (CONCERN IS SERVICE WATER TEMPERATURE).
Below is the Basis for Initial Reasonable Expectation of Continued Operability performed by Engineering:
CCI is Closed Cooling for Safety Injection, CCE is Closed Cooling for Charging, SW is Service Water, and Hx is Heat Exchanger
"The concerns with low UHS temperature are pipe stress calculations and possible freezing of stagnant loops such as CCI and CCE. Other portions of the SW system are not vulnerable to small changes in UHS temp due to constant flows and large heat sinks. Additionally, the 33 degree F limit is the design basis of the plant per the FSAR and therefore is reportable if exceeded.
"CCL and CCE freezing:
"A review of the thermodynamic characteristics of the CCI and CCE heat exchangers, assuming UHS of 31 degrees F, shows that the temperature gradient of the process fluid, due to the high conductivity of the copper nickel pipe and room temperatures, results in CCI and CCE bulk fluid temperatures remaining above freezing for an extended period of time, although the bulk temperature of the shell side of the CCI heat exchangers is only 32.5 degrees F. To assure operability of the SI pumps, the CCI system should be started at a SWP temperature of 33 degrees F decreasing to move the water through the system. The historical UHS temperature dips below 33 degrees F have been for a short duration of hours followed by longer warming periods. Design calculations show that CCI/SIH operate properly with the entire bulk temperature at 32 degrees F. This is conservative as the majority of the CCI system will remain at ambient room temperature in the standby mode. Once operating, there is additional heat input from the SIH pump. Therefore, there is no limit to UHS temperature which would render CCI/SIH inoperable even over several hours.
"A similar argument applies for the CCE/CHS system. The temperature gradient and bulk temperature characteristics are identical. Additionally, the CCE system utilizes a temperature control valve to maintain proper temperatures. The failed position of this valve (LOP) is to maximum flow through the SW Hx and has a position limiter designed for proper Charging lube oil temperature with a 33 degrees F UHS. The supporting calculation demonstrates that a 41.9 degrees F CCE inlet temperature at the lube oil cooler results in a 55 degrees F lube oil temperature (minimum required temperature for charging pump operability). If a service water temperature of 30 degrees F is used in the LMTD equation for the CCE HX, and the delta-T on both the SW & CCE sides are assumed to remain unchanged, the calculated CCE inlet temperature at the lube oil cooler is 42.2 degrees F, which is > 41.9 degrees F needed for the 55 degrees F lube oil temperature.
"To prevent potential freezing of the shell side fluid in the CCI heal exchangers and assure the Operability of the Safety Injection pumps, the CCL pumps should be started at a SWP temperature of 33 degrees F decreasing.
"Piping Stress:
"The Service Water System piping, supports and equipment have been analyzed to meet Code stress limits at a minimum temperature of 33 degrees F as described in OD MP3-213-96. Lower temperatures will have no affect on the qualification status of the piping analysis since thermal expansion induced pipe stress is secondary, and is not evaluated under GL 91-18 Operability rules.
"Increases in loads for pipe supports, equipment nozzles and equipment anchorage for temperatures below 33 degrees F are judged to be negligible and also have no affect on the operability of these components. ERC MP3-DE-96-059 revision 1 transmitted Technical Report TR-89014-1, revision 2, which evaluated low temperature piping at Millstone Unit 3. This report concludes that systems where the maximum operating temperature does not exceed 200 degrees F, and the temperature range does not exceed 200 degrees F, satisfy ASME Code requirements for thermal expansion loads. The Service Water maximum temperature range is well below 200 degrees F.
"Therefore, Service Water temperatures below 33 degrees F (to the freezing point of sea water 27.5 degrees F) will have no effect on the operability of Service Water piping, supports, equipment nozzles, and equipment anchorage, from a stress standpoint. No compensatory action are required.
"An OD has been prepared to support continued operation of MP3 with a Service Water temperature of 31 degrees F. This OD is stored on the MP3 server at K:\DeptData\MacDeb\UHSOD. Scoping calculations to support this OD have been completed and reviewed. This OD must be reviewed and PORC approved as required by the SM, if SWP temperatures lower than 33 degrees F are recorded."
When the tide goes out the water temperature of the Ultimate Heat Sink temperature decreases and when the tide comes back in the temperature of the Ultimate Heat Sink increases. At the present time the tide is going out and the water temperature is 32.7 degrees F. In a couple of hours the tide will start coming back in and the water temperature of the Ultimate Heat sink should increase to above 33 degrees F (37 to 38 degrees F).
The licensee notified State and Local Counties of this event.
The NRC Resident Inspector was notified of this event by the licensee.
Below is the Basis for Initial Reasonable Expectation of Continued Operability performed by Engineering:
CCI is Closed Cooling for Safety Injection, CCE is Closed Cooling for Charging, SW is Service Water, and Hx is Heat Exchanger
"The concerns with low UHS temperature are pipe stress calculations and possible freezing of stagnant loops such as CCI and CCE. Other portions of the SW system are not vulnerable to small changes in UHS temp due to constant flows and large heat sinks. Additionally, the 33 degree F limit is the design basis of the plant per the FSAR and therefore is reportable if exceeded.
"CCL and CCE freezing:
"A review of the thermodynamic characteristics of the CCI and CCE heat exchangers, assuming UHS of 31 degrees F, shows that the temperature gradient of the process fluid, due to the high conductivity of the copper nickel pipe and room temperatures, results in CCI and CCE bulk fluid temperatures remaining above freezing for an extended period of time, although the bulk temperature of the shell side of the CCI heat exchangers is only 32.5 degrees F. To assure operability of the SI pumps, the CCI system should be started at a SWP temperature of 33 degrees F decreasing to move the water through the system. The historical UHS temperature dips below 33 degrees F have been for a short duration of hours followed by longer warming periods. Design calculations show that CCI/SIH operate properly with the entire bulk temperature at 32 degrees F. This is conservative as the majority of the CCI system will remain at ambient room temperature in the standby mode. Once operating, there is additional heat input from the SIH pump. Therefore, there is no limit to UHS temperature which would render CCI/SIH inoperable even over several hours.
"A similar argument applies for the CCE/CHS system. The temperature gradient and bulk temperature characteristics are identical. Additionally, the CCE system utilizes a temperature control valve to maintain proper temperatures. The failed position of this valve (LOP) is to maximum flow through the SW Hx and has a position limiter designed for proper Charging lube oil temperature with a 33 degrees F UHS. The supporting calculation demonstrates that a 41.9 degrees F CCE inlet temperature at the lube oil cooler results in a 55 degrees F lube oil temperature (minimum required temperature for charging pump operability). If a service water temperature of 30 degrees F is used in the LMTD equation for the CCE HX, and the delta-T on both the SW & CCE sides are assumed to remain unchanged, the calculated CCE inlet temperature at the lube oil cooler is 42.2 degrees F, which is > 41.9 degrees F needed for the 55 degrees F lube oil temperature.
"To prevent potential freezing of the shell side fluid in the CCI heal exchangers and assure the Operability of the Safety Injection pumps, the CCL pumps should be started at a SWP temperature of 33 degrees F decreasing.
"Piping Stress:
"The Service Water System piping, supports and equipment have been analyzed to meet Code stress limits at a minimum temperature of 33 degrees F as described in OD MP3-213-96. Lower temperatures will have no affect on the qualification status of the piping analysis since thermal expansion induced pipe stress is secondary, and is not evaluated under GL 91-18 Operability rules.
"Increases in loads for pipe supports, equipment nozzles and equipment anchorage for temperatures below 33 degrees F are judged to be negligible and also have no affect on the operability of these components. ERC MP3-DE-96-059 revision 1 transmitted Technical Report TR-89014-1, revision 2, which evaluated low temperature piping at Millstone Unit 3. This report concludes that systems where the maximum operating temperature does not exceed 200 degrees F, and the temperature range does not exceed 200 degrees F, satisfy ASME Code requirements for thermal expansion loads. The Service Water maximum temperature range is well below 200 degrees F.
"Therefore, Service Water temperatures below 33 degrees F (to the freezing point of sea water 27.5 degrees F) will have no effect on the operability of Service Water piping, supports, equipment nozzles, and equipment anchorage, from a stress standpoint. No compensatory action are required.
"An OD has been prepared to support continued operation of MP3 with a Service Water temperature of 31 degrees F. This OD is stored on the MP3 server at K:\DeptData\MacDeb\UHSOD. Scoping calculations to support this OD have been completed and reviewed. This OD must be reviewed and PORC approved as required by the SM, if SWP temperatures lower than 33 degrees F are recorded."
When the tide goes out the water temperature of the Ultimate Heat Sink temperature decreases and when the tide comes back in the temperature of the Ultimate Heat Sink increases. At the present time the tide is going out and the water temperature is 32.7 degrees F. In a couple of hours the tide will start coming back in and the water temperature of the Ultimate Heat sink should increase to above 33 degrees F (37 to 38 degrees F).
The licensee notified State and Local Counties of this event.
The NRC Resident Inspector was notified of this event by the licensee.
General Information or Other
Event Number: 36627
Rep Org: SOLIDSTATE CONTROLS, INC.
Licensee: SOLIDSTATE CONTROLS, INC.
Region: 3
City: COLUMBUS State: OH
County:
License #:
Agreement: Y
Docket:
NRC Notified By: W. G. HAMPTON
HQ OPS Officer: FANGIE JONES
Licensee: SOLIDSTATE CONTROLS, INC.
Region: 3
City: COLUMBUS State: OH
County:
License #:
Agreement: Y
Docket:
NRC Notified By: W. G. HAMPTON
HQ OPS Officer: FANGIE JONES
Notification Date: 01/26/2000
Notification Time: 16:11 [ET]
Event Date: 01/21/2000
Event Time: 00:00 [EST]
Last Update Date: 01/26/2000
Notification Time: 16:11 [ET]
Event Date: 01/21/2000
Event Time: 00:00 [EST]
Last Update Date: 01/26/2000
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
Person (Organization):
MONTE PHILLIPS (R3)
JOHN ROGGE (R1)
CAUDLE JULIAN (R2)
GAIL GOOD (R4)
VERN HODGE (fax) (NRR)
MONTE PHILLIPS (R3)
JOHN ROGGE (R1)
CAUDLE JULIAN (R2)
GAIL GOOD (R4)
VERN HODGE (fax) (NRR)
10 CFR 21 - POTENTIAL DEFECT INVOLVING STATIC SWITCH CONTROL ASSEMBLY AND REGULATED RECTIFIER CONTROL ASSEMBLY USED IN AN UNINTERRUPTIBLE POWER SYSTEM
This reports a potential for a defect in Static Switch Control assemblies and Regulated Rectifier Control assemblies used in Uninterruptible Power Systems manufactured under the Solidstate Controls, Inc., 10CFR50 Appendix B Program.
Each of the plants has been notified of this report and requested to return the affected assemblies for replacement. The following plants have affected components: Indian Point 2, Catawba, Oconee, Arkansas Nuclear One, Waterford 3, Crystal River, Diablo Canyon, Sequoyah, and Surry.
The problem is associated with a component part - Unijunction Transistor, part number 03-650007-00, that has been found to sometimes cause random, sporadic transfers of the static switch due to a higher level of sensitivity to noise spikes on the input line or cause the regulated rectifier to output a higher DC voltage. This increase in sensitivity will not cause the system to drop power to the load, if the alternate source is not available the static switch will not transfer, continuing to protect the load.
This reports a potential for a defect in Static Switch Control assemblies and Regulated Rectifier Control assemblies used in Uninterruptible Power Systems manufactured under the Solidstate Controls, Inc., 10CFR50 Appendix B Program.
Each of the plants has been notified of this report and requested to return the affected assemblies for replacement. The following plants have affected components: Indian Point 2, Catawba, Oconee, Arkansas Nuclear One, Waterford 3, Crystal River, Diablo Canyon, Sequoyah, and Surry.
The problem is associated with a component part - Unijunction Transistor, part number 03-650007-00, that has been found to sometimes cause random, sporadic transfers of the static switch due to a higher level of sensitivity to noise spikes on the input line or cause the regulated rectifier to output a higher DC voltage. This increase in sensitivity will not cause the system to drop power to the load, if the alternate source is not available the static switch will not transfer, continuing to protect the load.