Event Notification Report for May 14, 1999
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
05/13/1999 - 05/14/1999
** EVENT NUMBERS **
35560 35686 35690 35695 35710 35711 35712 35713
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35560 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: SEABROOK REGION: 1 |NOTIFICATION DATE: 04/08/1999|
| UNIT: [1] [] [] STATE: NH |NOTIFICATION TIME: 15:41[EDT]|
| RXTYPE: [1] W-4-LP |EVENT DATE: 02/17/1999|
+------------------------------------------------+EVENT TIME: [EDT]|
| NRC NOTIFIED BY: KILBY |LAST UPDATE DATE: 05/13/1999|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |FRANK COSTELLO R1 |
|10 CFR SECTION: | |
|DDDD 73.71 UNSPECIFIED PARAGRAPH | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N N 0 Refueling |0 Refueling |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| UNESCORTED ACCESS GRANTED INAPPROPRIATELY. IMMEDIATE COMPENSATORY MEASURES |
| TAKEN UPON DISCOVERY. CONTACT HOO FOR ADDITIONAL DETAILS. |
| |
| ******************** UPDATE AT 1131 ON 05/13/99 FROM MIKE DAVID TO TROCINE |
| ******************** |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "North Atlantic [Energy Service Corporation] is providing this update |
| because the NRC Daily Report (Event Number: 35560) described this event as |
| 'Unescorted Access Granted Inappropriately.' North Atlantic believes that |
| the Unescorted Access was granted in accordance with the appropriate |
| procedures and that it did not violate any NRC regulations; hence, it was |
| appropriate. Furthermore, the appropriate action was taken when the FBI |
| information was received. North Atlantic believes that the event would have |
| been more accurately described as 'Unescorted Access Revoked.'" |
| |
| Contact the NRC operations officer for additional details. |
| |
| The licensee notified the NRC resident inspector. The NRC operations |
| officer notified the R1DO (Caruso). |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35686 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: MILLSTONE REGION: 1 |NOTIFICATION DATE: 05/06/1999|
| UNIT: [] [] [3] STATE: CT |NOTIFICATION TIME: 18:44[EDT]|
| RXTYPE: [1] GE-3,[2] CE,[3] W-4-LP |EVENT DATE: 05/06/1999|
+------------------------------------------------+EVENT TIME: 18:00[EDT]|
| NRC NOTIFIED BY: MARTIN |LAST UPDATE DATE: 05/13/1999|
| HQ OPS OFFICER: CHAUNCEY GOULD +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |JAMES LINVILLE R1 |
|10 CFR SECTION: | |
|ADAS 50.72(b)(2)(i) DEG/UNANALYZED COND | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
| | |
|3 N N 0 Cold Shutdown |0 Cold Shutdown |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| VALVE FAILED LOCAL LEAK RATE TEST (LLRT). |
| |
| VALVE 3QSS*V4 IS A CHECK VALVE IN THE QUENCH SPRAY SYSTEM, AND THE CAUSE OF |
| THE LLRT FAILURE IS UNKNOWN AND BEING INVESTIGATED, BUT MOST LIKELY MAY BE |
| DUE TO VALVE DEGRADATION. WHEN THE RESULTS OF ITS LLRT WERE ADDED TO THE |
| OTHER CATEGORY "C" VALVES' RESULTS, THE TOTAL LEAKAGE EXCEEDED TECHNICAL |
| SPECIFICATION ALLOWABLE LIMITS OF 43 SCFH (TOTAL MEASURED CATEGORY "C" WAS |
| 335 SCFH); HOWEVER, THE TOTAL LEAKAGE OF EVERYTHING STILL DID NOT EXCEED THE |
| 0.6 La VALUE. CORRECTIVE ACTION WILL BE TO REPAIR THE VALVE PRIOR TO |
| STARTUP. |
| |
| THE NRC RESIDENT INSPECTOR WAS NOTIFIED ALONG WITH STATE, LOCAL AND OTHER |
| GOVERNMENT AGENCIES. |
| |
| * * * UPDATE 1134 5/13/99 FROM STEVE LAWHEAD TAKEN BY STRANSKY * * * |
| |
| Local leak rate testing of valve SIL*V6 (RHS Loop 1 Cold Leg Check Valve) |
| found the leakage to be excessive. This leakage, when combined with the |
| other known leakage, caused TS LCO 3.6.1.2.b to be exceeded. The TS |
| requires a combined leakage rate of less than 0.6 La for all penetrations |
| and valves subject to Type B and C tests, when pressurized to Pa. The NRC |
| resident inspector has been informed of this update. Notified R1DO |
| (Cowgill). |
+------------------------------------------------------------------------------+
!!!!!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!!!
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35690 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: NINE MILE POINT REGION: 1 |NOTIFICATION DATE: 05/07/1999|
| UNIT: [] [2] [] STATE: NY |NOTIFICATION TIME: 12:41[EDT]|
| RXTYPE: [1] GE-2,[2] GE-5 |EVENT DATE: 05/07/1999|
+------------------------------------------------+EVENT TIME: 09:50[EDT]|
| NRC NOTIFIED BY: LANGE |LAST UPDATE DATE: 05/13/1999|
| HQ OPS OFFICER: CHAUNCEY GOULD +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |JAMES LINVILLE R1 |
|10 CFR SECTION: | |
|AIND 50.72(b)(2)(iii)(D) ACCIDENT MITIGATION | |
|NLCO TECH SPEC LCO A/S | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 N Y 59 Power Operation |59 Power Operation |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| THE LICENSEE DECLARED THE HIGH PRESSURE CORE SPRAY (HPCS) SYSTEM INOPERABLE |
| AND ENTERED A 14-DAY LCO ACTION STATEMENT. |
| |
| THE HPCS SYSTEM FAILED TO MEET THE CHANNEL CHECK CRITERIA AND WAS DECLARED |
| INOPERABLE, BUT FUNCTIONAL. TWO TRIP UNITS WERE FOUND NOT TO BE WITHIN THE |
| 13-INCH LIMIT FOR LEVEL INITIATION INSTRUMENTATION. (THEY MEASURED 15 |
| INCHES.) A WORK ORDER HAS BEEN GENERATED TO CORRECT THE PROBLEM BEFORE THE |
| TIME CLOCK EXPIRES FOR THE LCO ACTION STATEMENT. THE REACTOR IS CURRENTLY |
| IN SINGLE LOOP OPERATION AT 59% POWER. |
| |
| THE LICENSEE INFORMED THE NRC RESIDENT INSPECTOR. |
| |
| ***RETRACTION ON 04/13/99 AT 1651 EDT FROM ROY GREEN TAKEN BY MACKINNON*** |
| |
| IT HAS BEEN DETERMINED THAT THE CHANNEL CHECK CRITERIA WAS MET AND THAT THE |
| HIGH PRESSURE CORE SPRAY SYSTEM WAS, THEREFORE, NOT INOPERABLE. PROCEDURAL |
| GUIDANCE THAT LED TO THE INOPERABLE DETERMINATION HAS BEEN CHANGED AFTER |
| ENGINEERING PROVIDED THE NECESSARY JUSTIFICATION. |
| |
| THE LICENSEE INFORMED THE NRC RESIDENT INSPECTOR. THE R1DO (JOHN CARUSO) |
| WAS NOTIFIED BY THE NRC OPERATIONS OFFICER. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 35695 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PORTSMOUTH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 05/09/1999|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 20:22[EDT]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 05/09/1999|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 16:45[EDT]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 05/13/1999|
| CITY: PIKETON REGION: 3 +-----------------------------+
| COUNTY: PIKE STATE: OH |PERSON ORGANIZATION |
|LICENSE#: GDP-2 AGREEMENT: N |MONTE PHILLIPS R3 |
| DOCKET: 0707002 |JOHN GREEVES NMSS |
+------------------------------------------------+FRANK CONGEL IRO |
| NRC NOTIFIED BY: LARSON | |
| HQ OPS OFFICER: CHAUNCEY GOULD | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NBNL RESPONSE-BULLETIN | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| NRC BULLETIN 91-01, 4-HOUR REPORT |
| |
| AT 1645 ON 05/09/99, DURING A PRESSURE DECAY TEST OF THE #1 AUTOCLAVE IN THE |
| X-342 BUILDING, A LIGHT DUSTING OF UO2F2 WAS OBSERVED INSIDE THE AUTOCLAVE, |
| AND A HEAVIER ACCUMULATION OF UO2F2 WAS OBSERVED ON THE PRESSURE DECAY |
| FILTER USED TO PREVENT URANIUM FROM ENTERING THE AUTOCLAVE DURING THIS TEST. |
| THE AUTOCLAVE WAS PLACED IN A CONTAINMENT MODE ISOLATING IT FROM ALL SUPPORT |
| SYSTEMS. HEALTH PHYSICS TECHNICIANS PERFORMED CONTAMINATION AND AIRBORNE |
| SAMPLING, AND ALL SAMPLES WERE LESS THAN DETECTABLE LIMITS. HAVING URANIUM |
| INSIDE THE SHELL OF THE AUTOCLAVE IS A LOSS OF ONE CONTROL OF THE DOUBLE |
| CONTINGENCY PRINCIPLE CONTAINED IN NCSA_342A002. THE ACCUMULATION OF UO2F2 |
| ON THE OUTSIDE OF THE PRESSURE DECAY FILTER IS NOT EXPECTED. ADDITIONAL |
| EVALUATIONS ARE BEING CONDUCTED AT THIS TIME. |
| |
| THERE IS NO LOSS OF HAZARDOUS/RADIOACTIVE MATERIAL OR |
| RADIOACTIVE/RADIOLOGICAL CONTAMINATION EXPOSURE AS A RESULT OF THIS EVENT. |
| THIS EVENT DOES NOT PRESENT ANY IMMINENT DANGER OF A CRITICALITY. |
| |
| THE AUTOCLAVE HAS BEEN ISOLATED FROM ALL POTENTIAL MODERATOR SOURCES, SO |
| CRITICALITY IS NOT POSSIBLE. MODERATION, ENRICHMENT, AND GEOMETRY |
| PARAMETERS WERE MAINTAINED THROUGHOUT THIS EVENT. THE UNLIKELIHOOD OF |
| URANIUM INSIDE THE AUTOCLAVE SHELL AND OUTSIDE OF THE UF6 CYLINDER WAS A |
| LOST PARAMETER. THE AUTOCLAVE WAS TAGGED TO PREVENT ANY OPERATION UNTIL |
| RECOVERY PLANS ARE DEVELOPED. |
| |
| THE NRC RESIDENT INSPECTOR WAS NOTIFIED BY PORTSMOUTH PERSONNEL. |
| |
| ***UPDATE ON 05/13/99 AT 2248 EDT FROM J. CASTLE TAKEN BY MACKINNON*** |
| |
| REVISION TO THE NCS ANOMALOUS CONDITIONS INCIDENT REPORT (NSI-99-02603 REV. |
| 1) IDENTIFIES THAT THE MAXIMUM AMOUNT OF U-235 THAT COULD HAVE BEEN RELEASED |
| IN THE AUTOCLAVE WAS 755.2 GRAMS. THIS AMOUNT OF U-235 IS LESS THAN THE |
| ALWAYS SAFE MASS OF 800 GRAMS FOR 5% ENRICHED MATERIAL AND IS WELL BELOW THE |
| SUBCRITICAL LIMIT OF 1640 GRAMS AT 5% ENRICHMENT. THIS INCIDENT REPRESENTS |
| A DEGRADATION IN THE MARGIN OF SAFETY FOR THE MASS CONTROL PARAMETER BUT DID |
| NOT RESULT IN A LOSS OF DOUBLE CONTINGENCY BASED ON THE MAXIMUM AMOUNT OF |
| MATERIAL POTENTIALLY LOST. |
| |
| SAFETY SIGNIFICANCE OF EVENTS: THE EVENT RESULTED IN A DUSTING OF URANIUM |
| INSIDE THE AUTOCLAVE (A/C) SHELL AS WELL AS URANIUM ON THE PRESSURE DECAY |
| FILTER. THE MAXIMUM CREDIBLE AMOUNT OF U-235 RELEASED INSIDE THE A/C WAS |
| CALCULATED BY ENGINEERING TO BE 755.2 GRAMS OF U-235 BASED ON ASSUMING THE |
| ENTIRE AVAILABLE UF6 PIPING VOLUME AND PRESSURE DECAY FILTER VOLUME WERE |
| FILLED WITH URANIUM AT MAXIMUM DENSITY PRIOR TO THE EVENT. THIS IS LESS |
| THAN THE SAFE MASS OF 800 GRAMS U-235 AT 5% ENRICHMENT AND MUCH LESS THAN |
| THE SUBCRITICAL MASS LIMIT OF 1,640 GRAMS OF U-235. THEREFORE, A |
| CRITICALITY COULD NOT HAVE OCCURRED WITHOUT THE ADDITION OF MORE MASS AND |
| MODERATION TO THE AUTOCLAVE. BOTH ADDITIONS ARE UNLIKELY AS THE AUTOCLAVE |
| IS SHUTDOWN AND ISOLATED, PENDING CLEANUP. THE DOUBLE CONTINGENCY PRINCIPLE |
| WAS, THEREFORE, STILL MET FOLLOWING THIS EVENT. |
| |
| POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO OF HOW CRITICALITY |
| COULD OCCUR): A CRITICALITY WOULD ONLY BE POSSIBLE IF MORE THAN 300 lbs OF |
| WATER IS ADDED TO THE AUTOCLAVE. THE AUTOCLAVE HAS BEEN ISOLATED FROM ALL |
| POTENTIAL MODERATOR SOURCES. AT LEAST 885 GRAMS OF U-235 WOULD NEED TO BE |
| ADDED TO THE AUTOCLAVE IN ADDITION TO AT LEAST 300 POUNDS OF WATER IN ORDER |
| FOR A CRITICALITY TO BE POSSIBLE IN THE AUTOCLAVE GEOMETRY. THE AUTOCLAVE |
| HAS BEEN ISOLATED FROM ALL POTENTIAL SOURCES OF MODERATION AND URANIUM. |
| MODERATION AND ENRICHMENT CONTROL WERE MAINTAINED, AND ALSO, GEOMETRY OF THE |
| PRESSURE DECAY FILTER WAS MAINTAINED. |
| |
| NUCLEAR CRITICALITY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE |
| FAILURES OR DEFICIENCIES: THE PRESENCE OF UO2F2 INSIDE THE AUTOCLAVE SHELL |
| IN AN AMOUNT LESS THAN A SAFE MASS REPRESENTS A DEGRADATION IN ONE OF THE |
| TWO CONTROLS FOR DOUBLE CONTINGENCY, BUT IT DOES NOT REPRESENT A COMPLETE |
| LOSS OF THAT CONTROL. TWO UNLIKELY, INDEPENDENT AND CONCURRENT EVENTS WOULD |
| STILL NEED TO OCCUR IN ORDER FOR A CRITICALITY TO BE POSSIBLE IN THE |
| AUTOCLAVE. |
| |
| CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEM AND WHEN EACH WAS IMPLEMENTED: |
| AT 1805 HOURS ON 05/09/99, THE AUTOCLAVE WAS DANGER TAGGED TO PREVENT ANY |
| OPERATION UNTIL RECOVERY PLANS ARE DEVELOPED. |
| |
| THE NRC RESIDENT INSPECTOR WAS NOTIFIED OF THIS EVENT BY THE CERTIFICATE |
| HOLDER. THE NMSS EO (JOE HOLONICH) AND R3DO (ROGER LANKSBURY) WERE NOTIFIED |
| BY THE NRC OPERATIONS OFFICER. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35710 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: BYRON REGION: 3 |NOTIFICATION DATE: 05/13/1999|
| UNIT: [1] [] [] STATE: IL |NOTIFICATION TIME: 09:47[EDT]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 05/13/1999|
+------------------------------------------------+EVENT TIME: 08:10[CDT]|
| NRC NOTIFIED BY: DAVE FLOWERS |LAST UPDATE DATE: 05/13/1999|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |ROGER LANKSBURY R3 |
|10 CFR SECTION: | |
|ARPS 50.72(b)(2)(ii) RPS ACTUATION | |
|AESF 50.72(b)(2)(ii) ESF ACTUATION | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 A/R Y 100 Power Operation |0 Hot Standby |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| AUTOMATIC REACTOR TRIP DUE TO AN INADVERTENT POWER RANGE HIGH FLUX SIGNAL |
| DURING THE PERFORMANCE OF SURVEILLANCE TESTING |
| |
| At 0810 CDT, Unit 1 experienced an automatic reactor trip from 100% power |
| due to an inadvertent power range high flux trip signal during the |
| performance of instrumentation and control surveillance testing. All rods |
| fully inserted. Following the trip, feedwater isolated as expected, and the |
| auxiliary feedwater pumps automatically started on low 2 steam generator |
| level (18% narrow range) as designed. (Steam generator level decreased to |
| 0% narrow range, and normal steam generator level is 63%.) In addition, the |
| 'K' steam dump failed to reclose, and this caused reactor coolant system |
| temperature to decrease slightly below the P-12 setpoint of 550�F for a |
| short period of time. (Normal temperature is 557�F.) When temperature |
| reached 550�F, the steam dumps automatically isolated. The 'K' steam dump |
| was manually isolated. The steam dumps were re-opened when temperature went |
| above 550�F. All other post-trip actuations occurred as expected. None of |
| the primary safety valves or power-operated relief valves lifted. However, |
| some of the secondary feedwater heater relief valves lifted. The licensee |
| stated that it is not unusual for these valves to lift following a reactor |
| trip, and the licensee plans to isolate them. The licensee also stated that |
| these valves are not large enough to cause a problem with cooldown. There |
| were no emergency core cooling system actuations, and none were required. |
| An investigation to determine the exact cause of the reactor trip is in |
| progress. |
| |
| The unit is currently stable in Mode 3. The reactor coolant pumps, normal |
| charging and letdown, and pressurizer heaters and sprays are being utilized |
| for primary system transport, level, and pressure control. Containment |
| parameters are normal. Secondary steam is being dumped to the condenser, |
| and water is being supplied to the steam generators via the auxiliary |
| feedwater pumps and startup feedwater pump. The licensee plans to secure |
| auxiliary feedwater. Offsite power is available, and the emergency diesel |
| generators are operable and available if needed. The licensee is in the |
| process of re-closing the ring bus located in the switchyard to increase |
| offsite power reliability. |
| |
| The licensee notified the NRC resident inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Other Nuclear Material |Event Number: 35711 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: GREEN BAY PACKAGING |NOTIFICATION DATE: 05/13/1999|
|LICENSEE: GREEN BAY PACKAGING |NOTIFICATION TIME: 14:12[EDT]|
| CITY: WINCHESTER REGION: 2 |EVENT DATE: 05/21/1998|
| COUNTY: FREDERICK STATE: VA |EVENT TIME: 12:00[EDT]|
|LICENSE#: 45-25268-01 AGREEMENT: N |LAST UPDATE DATE: 05/13/1999|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |RUDOLPH BERNHARD R2 |
| |SCOTT MOORE NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: M. DICKEY | |
| HQ OPS OFFICER: JOHN MacKINNON | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|IBBF 30.50(b)(2)(ii) EQUIP DISABLED/FAILS | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| INTERMITTENT STUCK SHUTTER (MECHANICAL PROBLEM) ON OHMART SCANNING GAUGES |
| REPORTED AFTER PERFORMANCE OF AN NRC INSPECTION. |
| |
| The licensee reported two past incidents which occurred at their facility |
| after an NRC inspection determined that the two past incidences should have |
| been reported. |
| |
| On 05/19/98 or 05/20/98, the shutter of an Ohmart scanning gauge, used to |
| measure the thickness of paper, was found to intermittently stick open. As |
| of late 1993, the Ohmart scanning gauge (Model # OD120) contained 300 |
| millicuries of Kr-85. Kr-85 is a beta-gamma emitter with a half life of |
| 10.76 years. On 05/21/98, Alternative Process Service of Alabama repaired |
| the intermittently sticking shutter on the gauge. There was no exposure to |
| personnel in the area when the shutter was stuck open because the gauge is |
| hard to get to (0.5" gap). It was determined that broken screws had caused |
| the shutter to intermittently stick open. |
| |
| On 05/03/99, a shutter on a different Ohmart scanning gauge was found to be |
| stuck open. The licensee lightly tapped the source case of the Ohmart |
| scanning gauge (Model # OD120) which caused the shutter to go closed. The |
| licensee called Honeywell Measurex and informed them that the shutter had |
| stuck open on the scanning gauge. Honeywell Measurex will send someone to |
| Green Bay Packaging during the week of 05/17/99 to repair the scanning |
| gauge. This gauge also contained 300 millicuries of Kr-85 as of late 1993. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35712 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: CLINTON REGION: 3 |NOTIFICATION DATE: 05/14/1999|
| UNIT: [1] [] [] STATE: IL |NOTIFICATION TIME: 00:30[EDT]|
| RXTYPE: [1] GE-6 |EVENT DATE: 05/13/1999|
+------------------------------------------------+EVENT TIME: 23:00[CDT]|
| NRC NOTIFIED BY: RICH CHEAR |LAST UPDATE DATE: 05/14/1999|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |ROGER LANKSBURY R3 |
|10 CFR SECTION: | |
|AOUT 50.72(b)(1)(ii)(B) OUTSIDE DESIGN BASIS | |
|NLCO TECH SPEC LCO A/S | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 4 Startup |4 Startup |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| DISCOVERY THAT THE LEAK RATE OF A MAIN STEAM ISOLATION VALVE IS GREATER THAN |
| THE INBOARD LEAKAGE CONTROL SYSTEM PROVEN TREATMENT FLOW RATE |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "Clinton Power Station has identified that the Inboard Main Steam Isolation |
| Valve (MSIV) Leakage Control System (LCS) is not calibrated to process the |
| design MSIV leakage of 28 standard cubic feet per hour (scfh) per one main |
| steam line. One MSIV has been identified as having a leakage rate of 22.5 |
| scfh which is greater than the Inboard LCS proven treatment flow rate of |
| 21.3 scfh. Currently, the engineering team is still investigating." |
| |
| The licensee stated that the unit was in a 30-day technical specification |
| limiting condition for operation as a result of this issue. |
| |
| The licensee notified the NRC resident inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35713 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: SAN ONOFRE REGION: 4 |NOTIFICATION DATE: 05/14/1999|
| UNIT: [] [] [3] STATE: CA |NOTIFICATION TIME: 01:54[EDT]|
| RXTYPE: [1] W-3-LP,[2] CE,[3] CE |EVENT DATE: 05/13/1999|
+------------------------------------------------+EVENT TIME: 21:44[PDT]|
| NRC NOTIFIED BY: MATT THURBURN |LAST UPDATE DATE: 05/14/1999|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |CHARLES CAIN R4 |
|10 CFR SECTION: | |
|ARPS 50.72(b)(2)(ii) RPS ACTUATION | |
|AESF 50.72(b)(2)(ii) ESF ACTUATION | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
| | |
|3 M/R Y 97 Power Operation |0 Hot Standby |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| MANUAL REACTOR TRIP AND MANUAL INITIATION OF EMERGENCY FEEDWATER DUE TO THE |
| LOSS OF BOTH MAIN FEEDWATER PUMPS DURING THE PERFORMANCE OF MAINTENANCE ON |
| THE FEEDWATER CONTROL SYSTEM |
| |
| At 2144 PDT on 05/13/99, both main feedwater pumps tripped for unknown |
| reasons. Reactor operators recognized this loss, manually tripped the |
| reactor from 97% power, and manually initiated emergency feedwater. All |
| rods fully inserted, and all systems functioned as required. There were no |
| engineered safety feature actuations other than the manual initiation of |
| emergency feedwater. The main steam isolation valves remained open, none of |
| the relief valves opened, and there was no loss of offsite power. |
| |
| The cause of the loss of both feedwater pumps is under investigation. The |
| licensee stated that maintenance personnel were performing voltage checks in |
| the feedwater control system at the time of the event. |
| |
| The reactor is currently stable in Mode 3. The reactor coolant pumps, |
| normal charging and letdown, and pressurizer heaters and sprays are being |
| utilized for primary system transport, level, and pressure control. |
| Containment parameters are normal. Emergency feedwater is supplying water |
| to the steam generators, and secondary steam is being dumped to the main |
| condenser. Offsite power is available. The 'A' emergency diesel generator |
| is currently out of service for planned maintenance. However, the 'B' |
| emergency diesel generator and the crossties from the other unit are |
| available if needed. |
| |
| The licensee notified the NRC resident inspector and plans to issue a |
| media/press release. |
+------------------------------------------------------------------------------+
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Page Last Reviewed/Updated Thursday, March 25, 2021