Event Notification Report for May 5, 1999
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
05/04/1999 - 05/05/1999
** EVENT NUMBERS **
35668 35669 35670 35671 35672 35673 35674 35675
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35668 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: GINNA REGION: 1 |NOTIFICATION DATE: 05/04/1999|
| UNIT: [1] [] [] STATE: NY |NOTIFICATION TIME: 00:09[EDT]|
| RXTYPE: [1] W-2-LP |EVENT DATE: 05/03/1999|
+------------------------------------------------+EVENT TIME: 23:30[EDT]|
| NRC NOTIFIED BY: DOUGLAS J. GOMEZ |LAST UPDATE DATE: 05/04/1999|
| HQ OPS OFFICER: JOHN MacKINNON +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |JAMES LINVILLE R1 |
|10 CFR SECTION: |CECIL THOMAS NRR |
|ASHU 50.72(b)(1)(i)(A) PLANT S/D REQD BY TS |CHARLES MILLER IRO |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 90 Power Operation |85 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| TECHNICAL SPECIFICATION SHUTDOWN DUE TO THREE OUT OF FOUR OVERTEMPERATURE |
| DELTA TEMPERATURE AND OVERPOWER DELTA POWER CHANNELS BEING DECLARED |
| INOPERABLE |
| |
| During calibrations of Protection Channels, it was determined that three out |
| of four Overtemperature and Overpower Delta Channels were inoperable |
| requiring entry into Technical Specification 3.0.3 requiring Mode 3 (Hot |
| Standby) within 6 hours (0530 EDT on 05/04/99). |
| |
| The summer coming out of the Delta Temperature Channels is superimposing an |
| AC ripple on top of the DC output, and depending on whether it is feeding |
| out through an NUS bistable or Foxboro bistable, the Foxboro bistable |
| apparently allows the AC ripple to continue through and cause the Delta |
| Temperature Setpoints to be non-conservative. The licensee did not know |
| when a surveillance test had last been performed on the channels. |
| |
| All Emergency Core Cooling Systems and the Emergency Diesel Generators are |
| fully operable. The electrical grid is stable. |
| |
| The NRC Resident Inspector will be notified by the licensee. |
| |
| **** Update on 05/04/99 at 0442 EDT from Dan Berry taken by MacKinnon **** |
| |
| The licensee exited Technical Specification 3.0.3 at 0437 EDT when three |
| Channels were declared operable and the fourth channel was defeated |
| (bistables were tripped). Reactor power level was reduced to 21% before |
| Technical Specification 3.0.3 was exited. The licensee plans to increase |
| reactor power level to between 30 to 35%, at which point, the licensee plans |
| to stabilize reactor power level and repair a steam leak on MSR "2B." |
| |
| The NRC Resident Inspector will be notified by the licensee. The R1DO (Jim |
| Linville) was notified by the NRC Operations Officer. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35669 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: MCGUIRE REGION: 2 |NOTIFICATION DATE: 05/04/1999|
| UNIT: [] [2] [] STATE: NC |NOTIFICATION TIME: 10:17[EDT]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 05/04/1999|
+------------------------------------------------+EVENT TIME: 09:30[EDT]|
| NRC NOTIFIED BY: GRADY PICKLER |LAST UPDATE DATE: 05/04/1999|
| HQ OPS OFFICER: JOHN MacKINNON +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |AL BELISLE R2 |
|10 CFR SECTION: | |
|AUNA 50.72(b)(1)(ii)(A) UNANALYZED COND OP | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 N Y 100 Power Operation |100 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| MECHANICAL PROBLEMS ASSOCIATED WITH MANUAL STEAM GENERATOR PORV OPERATION |
| WERE DISCOVERED AND CORRECTED. |
| |
| On March 14, 1999, McGuire Nuclear Station identified mechanical problems |
| with one Unit 2 Steam Generator Power-Operated Relief Valve (PORV). A |
| similar problem was identified with another Unit 2 Steam Generator PORV on |
| March 20, 1999. On May 3, 1999, it was determined that, as a result of the |
| observed mechanical problems, both of the affected Unit 2 Steam Generator |
| PORVs were inoperable for a period of time greater than allowed by McGuire |
| Technical Specification 3.7.4. This condition affected the ability of the |
| PORVs to operate in manual. Manual operation of the PORVs is credited in |
| mitigation of design basis accidents. Consequently, on May 4, 1999, at 0930 |
| hours it was determined that this issue was reportable as a 1-hour |
| Unanalyzed Condition event. |
| |
| The mechanical problem discussed above is that a pin is used to connect the |
| handwheel to the shaft that operates a Steam Generator PORV valve. It was |
| found that the pin for operation of two of the Steam Generator PORV valves |
| was inserted into the handwheel, but the pin was not inserted far enough to |
| go into the shaft. Therefore, the handwheel would have turned, but the |
| shaft going to the PORV would not have turned. The licensee has corrected |
| this problem, and an investigation of Unit 1 did not find this type of |
| problem. |
| |
| The NRC Resident Inspector was notified of this event by the licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35670 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: CATAWBA REGION: 2 |NOTIFICATION DATE: 05/04/1999|
| UNIT: [] [2] [] STATE: SC |NOTIFICATION TIME: 13:53[EDT]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 05/04/1999|
+------------------------------------------------+EVENT TIME: 12:55[EDT]|
| NRC NOTIFIED BY: BILL RUDY |LAST UPDATE DATE: 05/04/1999|
| HQ OPS OFFICER: FANGIE JONES +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |CHRIS CHRISTENSEN R2 |
|10 CFR SECTION: | |
|AOUT 50.72(b)(1)(ii)(B) OUTSIDE DESIGN BASIS | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 N Y 8 Power Operation |8 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| AUXILIARY FEEDWATER DECLARED INOPERABLE |
| |
| All Auxiliary Feedwater (AFW) Pumps have been declared inoperable, including |
| the #2A and #2B motor-driven and the #2 turbine-driven AFW Pumps. |
| |
| "The suction piping to the [AFW] System from the Nuclear Service Water |
| System was discovered to have excessive fouling that resulted in the AFW |
| system being outside its design basis. Engineering analysis determined that |
| under certain accident scenarios involving AFW system runout flow, there is |
| inadequate suction pressure to assure AFW pump operability. The |
| turbine-driven AFW pump has been isolated as an interim |
| compensatory measure. This action reduces the system runout flow which |
| should increase the suction pressure to the two remaining motor-driven AFW |
| pumps sufficiently to assure pump operability. Engineering analysis is in |
| progress to verify the adequacy of this action. The affected piping will be |
| cleaned which will ultimately correct the problem." |
| |
| Unit 2 will remain at its present power level until the engineering analysis |
| is complete and corrective actions are determined to be adequate. |
| |
| The licensee notified the NRC Resident Inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|General Information or Other |Event Number: 35671 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: FLORIDA BUREAU OF RADIATION CONTROL |NOTIFICATION DATE: 05/04/1999|
|LICENSEE: UNIVERSITY COMMUNITY HOSPITAL |NOTIFICATION TIME: 17:14[EDT]|
| CITY: TAMPA REGION: 2 |EVENT DATE: 05/03/1999|
| COUNTY: STATE: FL |EVENT TIME: 12:00[EDT]|
|LICENSE#: 0549-1 AGREEMENT: Y |LAST UPDATE DATE: 05/04/1999|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |CHRIS CHRISTENSEN R2 |
| |JOHN GREEVES NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: CHARLES ADAMS | |
| HQ OPS OFFICER: FANGIE JONES | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NAGR AGREEMENT STATE | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| AGREEMENT STATE REPORT - RADIATION THERAPY SEEDS BURNED |
| |
| "475 Pd-103 seeds arrived at the licensee's receiving department on Friday, |
| 30 April 99. The package was taken to the radiation therapy section, and |
| the technician says he advised the physicist that the package was in the |
| hall outside the door. The physicist says that he doesn't remember being so |
| advised. Later that day, a janitor stated that she picked up what she |
| assumed was a empty box and placed it in the trash. The package was taken |
| to the county waste facility and incinerated. The package was consumed in |
| the 2,000�F heat, but since palladium does not vaporize until 2,800�F, it |
| should have remained as ash. That small amount of ash diluted in huge |
| bunches of ash was not detected by the waste stream monitor. A state survey |
| of the waste energy facility did not detect any activity. The package was |
| noted missing on 3 May 99, and this office [State of Florida Bureau of |
| Radiation Control] was so advised. Further action is referred to Material |
| Licensing." |
| |
| The seeds constituted 666 mCi as of 1200 on 4 May 99. |
| |
| (Call the NRC operations officer for a contact telephone number.) |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35672 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PILGRIM REGION: 1 |NOTIFICATION DATE: 05/04/1999|
| UNIT: [1] [] [] STATE: MA |NOTIFICATION TIME: 17:35[EDT]|
| RXTYPE: [1] GE-3 |EVENT DATE: 05/04/1999|
+------------------------------------------------+EVENT TIME: 16:45[EDT]|
| NRC NOTIFIED BY: ERIC OLSON |LAST UPDATE DATE: 05/04/1999|
| HQ OPS OFFICER: FANGIE JONES +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |JAMES LINVILLE R1 |
|10 CFR SECTION: | |
|AOUT 50.72(b)(1)(ii)(B) OUTSIDE DESIGN BASIS | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 80 Power Operation |80 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| HPCI AND RCIC DIFFERENCES FOUND BETWEEN DESIGN BASIS AND TECHNICAL |
| SPECIFICATIONS. |
| |
| "During license design-basis reconstitution efforts it was discovered that |
| the specific values in the PNPS TS for HPCI and RCIC operability testing are |
| not in accordance with the plant design. In accordance with the TS, the |
| HPCI System is tested to ensure the HPCI pump can deliver at least 4,250 gpm |
| for a system head corresponding to a reactor pressure of 1,000 to 150 psig. |
| The RCIC test requirement is that RCIC shall deliver at least 400 gpm for a |
| system head corresponding to a reactor pressure of 1,000 to 150 psig. The |
| applicable Tech Spec Sections are 3.5.C for HPCI and 3.5.D for RCIC. |
| |
| "The design requirement of HPCI and RCIC is to achieve 4,250 gpm and 400 |
| gpm, respectively, for a reactor pressure corresponding to the Safety Relief |
| Valve (SRV) setpoint. Currently, the SRV upper setpoint limit is 1,115 |
| +/-11 psig. Therefore, the corresponding discharge pressure of the pumps |
| shall be that required to achieve the required flow rate at the given |
| reactor vessel pressure (1,126 psig) taking into account system head loss, |
| elevation changes, lowering level in the CST or suction taken from the torus |
| and instrument setpoint error. For both HPCI and RCIC, this pressure should |
| be approximately 1,243 psig (slightly less for RCIC due to lower head loss |
| from the lower flow criteria). |
| |
| "Both the HPCI and RCIC systems are considered operable per Operability |
| Evaluation #99-024 and #99-025. This is a verbal evaluation based on |
| engineering judgement that there is sufficient horsepower available to |
| achieve the design parameters of both systems and that, [during] past |
| testing, the actual values have, in fact, been achieved. However, the |
| systems have not been tested to the design values on a periodic (quarterly) |
| test basis. During a normal surveillance test of HPCI on 03/13/98, the HPCI |
| discharge pressure was 1,260 psig. During startup testing, the RCIC system |
| was tested to 1,280 psig discharge pressure." |
| |
| The licensee plans to notify the NRC Resident Inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35673 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: COOK REGION: 3 |NOTIFICATION DATE: 05/04/1999|
| UNIT: [1] [2] [] STATE: MI |NOTIFICATION TIME: 18:19[EDT]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 05/04/1999|
+------------------------------------------------+EVENT TIME: 15:00[EDT]|
| NRC NOTIFIED BY: DONALD KOSLOFF |LAST UPDATE DATE: 05/04/1999|
| HQ OPS OFFICER: FANGIE JONES +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |MONTE PHILLIPS R3 |
|10 CFR SECTION: | |
|ADAS 50.72(b)(2)(i) DEG/UNANALYZED COND | |
|AIND 50.72(b)(2)(iii)(D) ACCIDENT MITIGATION | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N N 0 Cold Shutdown |0 Cold Shutdown |
|2 N N 0 Cold Shutdown |0 Cold Shutdown |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| SI THROTTLE VALVES CAVITATION DURING LOCA COULD LEAD TO FAILURE OF SI |
| PUMPS. |
| |
| "On March 27, 1999, engineering personnel investigating NRC Information |
| Notice 97-76 concluded that a preliminary flow analysis indicated that six |
| Unit 1 safety injection (SI) throttle valves could experience cavitation |
| during a LOCA. As a result of that conclusion, SI throttle valve #1-SI-121S |
| was radiographed to determine its position. On April 8, 1999, a review of |
| the radiograph indicated that the valve was about 43 percent open. The |
| radiograph also showed indications of possible erosion of the valve that |
| could have been caused by cavitation. Valve cavitation during a LOCA could |
| cause the valves to allow excessive flow, leading to SI pump runout and |
| subsequent failure of the SI pumps. Valve #1-SI-141L1 was also radiographed |
| and determined to be 27 percent open. |
| |
| "A fax from the valve vendor indicated that SI throttle valves that were |
| less than 32 percent open may not be capable of allowing passage of sump |
| debris of the expected maximum size, 0.25 inch diameter. Several of the |
| valves may be less than 32 percent open. This condition could restrict SI |
| flow to the reactor coolant system during a LOCA. |
| |
| "On May 4, 1999, during continuing evaluation of the above conditions, plant |
| personnel determined that the conditions were reportable. Both conditions |
| may also exist in Unit 2. Both Units are currently in Mode 5. Evaluation |
| of these conditions, including determination of the need for physical |
| modification, is ongoing, and the conditions will be resolved prior to |
| startup of the units." |
| |
| The licensee notified the NRC Resident Inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 35674 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PORTSMOUTH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 05/04/1999|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 23:52[EDT]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 05/04/1999|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 13:50[EDT]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 05/04/1999|
| CITY: PIKETON REGION: 3 +-----------------------------+
| COUNTY: PIKE STATE: OH |PERSON ORGANIZATION |
|LICENSE#: GDP-2 AGREEMENT: N |MONTE PHILLIPS R3 |
| DOCKET: 0707002 |JOHN GREEVES NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: KURT SISLER | |
| HQ OPS OFFICER: LEIGH TROCINE | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NCFR NON CFR REPORT REQMNT | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| VALID SAFETY SYSTEM ACTUATION OF AN OVERHEAD CRANE HOIST BRAKE RESULTING IN |
| A FULL, 10-TON, UF6 CYLINDER BEING SUSPENDED ABOVE THE AUTOCLAVE ROLLERS IN |
| THE X-343 BUILDING |
| |
| The following text is a portion of a facsimile received from Portsmouth: |
| |
| "On May 04, 1999, at 1350 hours, during removal of a full, 10-ton, UF6 |
| cylinder from autoclave #5, the hoist on the North overhead crane stopped |
| due to an actuation of the hoist brake caused by a power failure. The hoist |
| brake did perform its design function upon this loss of power. This power |
| failure resulted in the cylinder being suspended approximately 1 foot above |
| the autoclave rollers. Procedure XP2-TE-TE5030 steps were implemented, and |
| after the power breaker was reset, the cylinder was lowered within the |
| confines of the autoclave, and the North crane was declared inoperable." |
| |
| "This event is being categorized and reported as a valid actuation of a 'Q' |
| Safety System in accordance with Safety Analysis Report, Section 8.9 |
| (24-hour report)." |
| |
| "There was no loss of hazardous/radioactive material or |
| radioactive/radiological contamination exposure as a result of this event." |
| |
| Portsmouth personnel notified the NRC resident inspector and the Department |
| of Energy site representative. (Call the NRC operations officer for a site |
| contact telephone number.) |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35675 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: LIMERICK REGION: 1 |NOTIFICATION DATE: 05/05/1999|
| UNIT: [] [2] [] STATE: PA |NOTIFICATION TIME: 04:38[EDT]|
| RXTYPE: [1] GE-4,[2] GE-4 |EVENT DATE: 05/05/1999|
+------------------------------------------------+EVENT TIME: 02:00[EDT]|
| NRC NOTIFIED BY: GREG SOSSON |LAST UPDATE DATE: 05/05/1999|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |JAMES LINVILLE R1 |
|10 CFR SECTION: | |
|AESF 50.72(b)(2)(ii) ESF ACTUATION | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 N N 0 Refueling |0 Refueling |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| ISOLATION OF THE 'B' LOOP DRYWELL CHILLED WATER INBOARD SUPPLY AND RETURN |
| VALVES DURING PERFORMANCE OF A SPECIAL PROCEDURE TO DEENERGIZE A SAFEGUARDS |
| BUS |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "On 05/05/99 at 0200 hours, it was discovered that an [engineered safety |
| feature] actuation occurred on the Unit 2 Drywell Chilled Water (DWCW) |
| system. This isolation occurred at approximately 2300 on 05/02/99 during |
| performance of a special procedure to deenergize the D22 safeguards bus." |
| |
| "With Unit 2 in OPCON 5, the 'B' loop DWCW inboard supply and return valves |
| HV--087-222 and HV-087-223 isolated when the power supply to an interposing |
| relay was deenergized per a special procedure on 05/02/99. This special |
| procedure did not properly address the valve closure. During this special |
| procedure, power was later removed from both valves. This removed control |
| room indication of their position, and their closure was not immediately |
| detected. Later, station personnel observed increasing drywell |
| temperatures. The follow-up investigation revealed the isolation valves were |
| closed by local verification." |
| |
| "Additional investigation found several containment atmosphere sample valves |
| and primary containment instrument gas [primary containment isolation |
| valves] that also closed during the loss of power. These conditions were |
| also expected but not properly documented in the special procedure." |
| |
| The licensee notified the NRC resident inspector. |
+------------------------------------------------------------------------------+
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Page Last Reviewed/Updated Thursday, March 25, 2021