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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 2,000F heat, but since palladium does not vaporize until 2,800F, 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.                            |
+------------------------------------------------------------------------------+