Event Notification Report for March 1, 1999
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
02/26/1999 - 03/01/1999
** EVENT NUMBERS **
35407 35409 35410 35411 35412 35413 35415 35416 35417 35418 35419
35420 35421 35422 35423 35424
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35407 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: WATERFORD REGION: 4 |NOTIFICATION DATE: 02/25/1999|
| UNIT: [3] [] [] STATE: LA |NOTIFICATION TIME: 18:41[EST]|
| RXTYPE: [3] CE |EVENT DATE: 02/25/1999|
+------------------------------------------------+EVENT TIME: 14:18[CST]|
| NRC NOTIFIED BY: BILL MCKINNEY |LAST UPDATE DATE: 02/27/1999|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |ELMO COLLINS R4 |
|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 Refueling |0 Refueling |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| PRESSURIZER NOZZLE LEAKAGE DISCOVERED DURING REFUELING OUTAGE |
| |
| During a visual inspection, evidence of reactor coolant system leakage was |
| found on two inconel instrument nozzles located on the top head of the |
| pressurizer. The leakage was in the annulus area where the nozzle |
| penetrates the pressurizer head. The nozzles are welded on the inner |
| diameter of the pressurizer and are joined to instrument valves RC-310 and |
| RC-311. |
| |
| The NRC resident inspector has been informed of this notification by the |
| licensee. |
| |
| * * * UPDATE AT 2251 ON 02/27/99 FROM DAVID LITOLFF TAKEN BY STRANSKY * * * |
| |
| "On 02/25/99 a 4-hour report to the NRC was made per 10CFR50 72(b)(2)(i) for |
| evidence of Reactor Coolant System Leakage on two pressurizer instrument |
| nozzles. The purpose of this report is to update the 02/25/99 report for |
| additional Reactor Coolant System instrument nozzles which have been |
| identified as having evidence of RCS leakage. On 02/27/99, evidence of |
| boric acid leakage was found on one Hot Leg 1 Inconel Alloy 600 instrument |
| nozzle. Potential leakage was also found for one steam generator instrument |
| nozzle and the pressurizer side shell nozzle. Any further evidence of |
| leakage found in subsequent inspections will be included in the 30-day |
| Licensee Event Report." |
| |
| The NRC resident inspector will be informed of this report by the licensee. |
| The NRC Operations Officer notified the R4DO (Chuck Cain). |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35409 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: SEABROOK REGION: 1 |NOTIFICATION DATE: 02/26/1999|
| UNIT: [1] [] [] STATE: NH |NOTIFICATION TIME: 09:28[EST]|
| RXTYPE: [1] W-4-LP |EVENT DATE: 02/26/1999|
+------------------------------------------------+EVENT TIME: 01:22[EST]|
| NRC NOTIFIED BY: STEVE MORRISSEY |LAST UPDATE DATE: 02/26/1999|
| HQ OPS OFFICER: LEIGH TROCINE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |RICHARD BARKLEY R1 |
|10 CFR SECTION: | |
|NLTR LICENSEE 24 HR REPORT | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| DISCOVERY OF A DEAD SEAL IN THE CIRCULATING WATER FOREBAY |
| |
| The following text is a portion of a facsimile received from the licensee: |
| |
| "A dead seal was observed in the Seabrook Station's CW (circulating) water |
| forebay on February 26, 1999, at about 0122. It is not known whether the |
| seal was alive or dead upon entering the offshore intake structure. This |
| 24-hour notification is [being] made in accordance with Section 4.1 of the |
| Environmental Protection Plan, Appendix 'B' of the Operating License." |
| |
| The licensee plans to notify the NRC resident inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 35410 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PADUCAH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 02/26/1999|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 14:09[EST]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 02/26/1999|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 08:45[CST]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 02/26/1999|
| CITY: PADUCAH REGION: 3 +-----------------------------+
| COUNTY: McCRACKEN STATE: KY |PERSON ORGANIZATION |
|LICENSE#: GDP-1 AGREEMENT: Y |BRUCE JORGENSEN R3 |
| DOCKET: 0707001 |JOHN HICKEY NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: M. UNDERWOOD | |
| HQ OPS OFFICER: BOB STRANSKY | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|OCBA 76.120(c)(2)(i) ACCID MT EQUIP FAILS | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| PROCESS GAS LEAK DETECTION SYSTEM INOPERABLE (24-HOUR REPORT) |
| |
| The following text is a portion of a facsimile received from Paducah: |
| |
| "On 02/26/99 at 0845 [CST], while performing TSR surveillances on C-333 unit |
| 4 cell 10 Process Gas Leak Detection (PGLD), it was discovered that the |
| detector heads would not test fire. In the process of evaluating and |
| troubleshooting, the PGLD system was placed in a condition (override mode) |
| which would have detected a release, but was then placed back in a condition |
| (normal mode) in which the PGLD system was inoperable. After approximately |
| 90 minutes, the system was returned to an operable condition (override |
| mode). The PGLD system is required to be operable when operating above |
| atmospheric pressure. C-333 unit 4 cell 10 was operating above atmospheric |
| pressure at the time of the failure." |
| |
| The NRC Resident Inspector has been informed of this notification by Paducah |
| personnel. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35411 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: OCONEE REGION: 2 |NOTIFICATION DATE: 02/26/1999|
| UNIT: [1] [2] [3] STATE: SC |NOTIFICATION TIME: 15:20[EST]|
| RXTYPE: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-L|EVENT DATE: 02/26/1999|
+------------------------------------------------+EVENT TIME: 13:00[EST]|
| NRC NOTIFIED BY: LARRY NICHOLSON |LAST UPDATE DATE: 02/26/1999|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |ROBERT HAAG 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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
|2 N N 0 Hot Shutdown |0 Hot Shutdown |
|3 N Y 100 Power Operation |100 Power Operation |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| EFW SYSTEM DECLARED OUTSIDE OF DESIGN BASIS |
| |
| "On February 8, 1999, Duke Energy Corporation (Duke) met with the NRC staff |
| at NRC Headquarters to discuss a concern involving the differences in the |
| Oconee Emergency Feedwater (EFW) system design and post-TMI licensing basis |
| associated with the mitigation of certain Main Feedwater break scenarios. |
| The plant design utilizes the availability of EFW from any unit should the |
| affected unit's EFW system be lost during a Main Feedwater line break. In |
| response, an NRC letter, dated February 24, 1999, agreed that the issue did |
| not constitute a significant safety concern and provided an NRC |
| interpretation that the reliance of alternate EFW sources, except for |
| certain approved exceptions, was not consistent with the current licensing |
| basis. |
| |
| "The specific concern involves the failure to close of the upper surge tank |
| to hotwell makeup valve (C-187) following a main feedwater line rupture, |
| resulting in the depletion of the upper surge tank and subsequent loss of |
| EFW on the affected unit. Should this occur, operators would restore |
| feedwater by either cross-connecting EFW to one of the other units or |
| starting the Standby Shutdown Facility Auxiliary Service Water pump. These |
| alternate sources are designed and capable of supplying feedwater to the |
| affected unit. Operators are trained and procedures are established to |
| accomplish these tasks. |
| |
| "On February 26, 1999, following review of the NRC letter, it was determined |
| that the differences in the Oconee Emergency Feedwater (EFW) system design |
| and post-TMI licensing basis regarding mitigation of certain Main Feedwater |
| break scenarios, concurrent with a single active failure, constituted a |
| condition outside the licensing basis of the plant. This condition does not |
| constitute a safety concern due to the availability of multiple, diverse |
| sources of feedwater. The EFW system is considered operable but in |
| non-conformance with the licensing basis as stated in the UFSAR. Duke is |
| evaluating options to resolve the subject UFSAR discrepancy." |
| |
| The NRC resident inspector has been informed of this notification. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Hospital |Event Number: 35412 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: MAYO CLINIC |NOTIFICATION DATE: 02/26/1999|
|LICENSEE: MAYO FOUNDATION |NOTIFICATION TIME: 16:22[EST]|
| CITY: ROCHESTER REGION: 3 |EVENT DATE: 02/18/1999|
| COUNTY: STATE: MN |EVENT TIME: 12:00[CST]|
|LICENSE#: 22-00519-03 AGREEMENT: N |LAST UPDATE DATE: 02/26/1999|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |BRUCE JORGENSEN R3 |
| |JOHN HICKEY NMSS |
+------------------------------------------------+KEVIN RAMSEY (fax) NMSS |
| NRC NOTIFIED BY: RICHARD VETTER |RICHARD BARKLEY R1 |
| HQ OPS OFFICER: BOB STRANSKY | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|LADM 35.33(a) MED MISADMINISTRATION | |
|CCCC 21.21 UNSPECIFIED PARAGRAPH | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| 10 CFR PART 21 REPORT - TREATMENT SOFTWARE ERROR CAUSED MEDICAL |
| MISADMINISTRATION |
| |
| The licensee reported that a medical misadministration occurred on 2/18/1999 |
| due to a problem with treatment planning software (TCP Version 1.20 upgrade) |
| provided by Nucletron of Columbia, Maryland. The misadministration was |
| identified on 2/25/1999 and verified on 2/26/1999. Specifically, a patient |
| was prescribed a dose of 4500 rads by external beam; however, due to a |
| problem with the treatment software, the patient was also given a dose of |
| 200 rads to the area from a high dose rate brachytherapy unit. The patient |
| and the referring physician have both been informed of the |
| misadministration. |
| |
| The licensee submitted the following information in accordance with 10 CFR |
| Part 21: |
| |
| "Identification of the facility, the activity, or the basic component: |
| High dose rate afterloader treatment (Ir-192 microselectron HDR V2) |
| Device software (TCS Version 1.20 upgrade) |
| |
| "Identification of the firm supplying the basic component which failed to |
| comply: |
| Nucletron, Columbia, MD |
| |
| "Nature of the failure and safety hazard that could be created: |
| The software allows more than one active cell on a treatment planning |
| sheet. |
| |
| "This allows parameters within another cell to be modified while not working |
| in that cell. In this case, dwell time and step size were simultaneously |
| active. While purposely intending to change dwell time, step size can change |
| without alerting the user. This could result in a possible therapy |
| misadministration under 10CFR35. |
| |
| "Date on which information of defect was obtained: |
| February 25, 1999 |
| |
| "Number and location of all such components: |
| Mayo Foundation has only one such device. It is located in the Charlton |
| Building, Room CHS-209. Nucletron can supply information regarding other |
| facilities using the device. |
| |
| "Corrective action: |
| Mayo Foundation modified its procedures to require a pretreatment check that |
| includes step size; this action has been completed. All individuals who |
| manually enter treatment data will be made aware of the defect and told to |
| visually confirm their entries prior to printing the pretreatment report; |
| this action will be completed by Tuesday, March 2, 1999. |
| |
| "On Friday, February 26, Mayo Foundation notified Nucletron of the anomaly |
| suggesting they correct their computer software." |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 35413 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PADUCAH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 02/26/1999|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 16:38[EST]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 02/25/1999|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 16:15[CST]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 02/26/1999|
| CITY: PADUCAH REGION: 3 +-----------------------------+
| COUNTY: McCRACKEN STATE: KY |PERSON ORGANIZATION |
|LICENSE#: GDP-1 AGREEMENT: Y |BRUCE JORGENSEN R3 |
| DOCKET: 0707001 |JOHN HICKEY NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: J. M. UNDERWOOD | |
| HQ OPS OFFICER: BOB STRANSKY | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NBNL RESPONSE-BULLETIN | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| 24-HOUR NRC BULLETIN 91-01 REPORT |
| |
| "Potentially fissile trap media was discovered in an approximately 30 gallon |
| trash can in violation of NCSA GEN-15. NCSA GEN-15 requires that |
| fissile/potentially fissile waste be accumulated in a maximum 5.5-gallon |
| waste drum. The only exception is if the waste is exempted from NCS controls |
| in accordance with requirement 2 of NCSA GEN-15. However, the trap media was |
| not exempted prior to disposal. |
| |
| "The waste was generated prior to implementation of NCSA GEN-1 5 and is |
| therefore a legacy issue; however, NCSA GEN-15 is the currently approved |
| NCSA for the generation and handling of potentially fissile waste. |
| |
| "This event is being categorized as a 24-hour event in accordance with |
| Safety Analysis Report Table 6.9-1 Criteria A.4.a and NRC Bulletin 91-01, |
| Supplement 1 report. |
| |
| "SAFETY SIGNIFICANCE OF EVENTS: |
| |
| "This violation resulted in the loss of one leg of double contingency. |
| Although double contingency was not maintained, there was not enough |
| material present to result in a critical configuration. |
| |
| "POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO[S] OF HOW |
| CRITICALITY COULD OCCUR): |
| |
| The trash can contained approximately 15 gallons of contaminated alumina. |
| Based upon data from KY/S-208, Subcritical Dimensions For Water Reflected |
| UO2F2 and Water Systems at Two Weight Percent Enrichment, at 2.0 wt % U-235, |
| the safe volume of UO2F2 solution is 23 gallons. Additionally, KY/S-208 |
| modeled optimal concentration UO2F2 solution in a spherical geometry |
| reflected with 30 cm of water. The trash can contains trap material |
| intermixed with the UO2F2, and the material is not in the optimum |
| configuration modeled in KY/S-208, therefore, in reality it would take much |
| more than 23 gallons to achieve a critical configuration. Based upon this |
| information, a criticality is not possible. |
| |
| "In order for a criticality to be possible much more than 23 gallons of the |
| trap material would have to be present In the trash can. |
| |
| "CONTROLLED PARAMETERS (MASS, MODERATION. GEOMETRY, CONCENTRATION, ETC.): |
| |
| "Controlled parameters are geometry and spacing. |
| |
| "ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS |
| LIMIT AND % WORST CASE OF CRITICAL MASS): |
| |
| "The trash can contained approximately 15 gallons of contaminated alumina at |
| a maximum assay of 1.04 wt % U235. |
| |
| "In order for a criticality to be possible, much more than 23 gallons of the |
| trap material would have to be present in the trash can. |
| |
| "The determination that the material in the drum was fissile is based on |
| conservative sample results. Two independent smears and two independent bulk |
| samples were taken and analyzed. One of the bulk sample results indicated an |
| assay of .944% U-235. All of the remaining sample results were below 9%. A |
| .1% error is conservatively applied to lab sample results as a general rule |
| to account for uncertainties. Much lower uncertainties are routinely |
| achieved but have not been established far these samples at this time. This |
| Incident Report conservatively assumes the material in the drum is fissile |
| based on the .1% error applied to the one sample result above .9% U-235. |
| |
| "NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION |
| OF THE FAILURES OR DEFICIENCIES: |
| |
| "Loss of spacing control. Double contingency control leg was lost since |
| geometry process condition was not maintained. |
| |
| "CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEM AND WHEN EACH WAS IMPLEMENTED: |
| |
| "The trap media will be disposed of in a minimum 5.5 gallon drum in |
| accordance with plant procedure CP2-EW-WM1036. |
| |
| "A minimum 6 ft. spacing is being maintained between the cold trap and the |
| container of contaminated trap media. A minimum 2 ft. spacing will be |
| maintained between the maximum 5.5 gallon waste drum containing the trap |
| media and all other fissile/potentially fissile material." |
| |
| The NRC resident inspector has been informed of this notification. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Hospital |Event Number: 35415 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| REP ORG: SAN DIEGO MEDICAL CENTER |NOTIFICATION DATE: 02/26/1999|
|LICENSEE: VA MEDICAL SYSTEM |NOTIFICATION TIME: 17:31[EST]|
| CITY: SAN DIEGO REGION: 4 |EVENT DATE: 02/26/1999|
| COUNTY: STATE: CA |EVENT TIME: 08:10[PST]|
|LICENSE#: 04-15030-01 AGREEMENT: Y |LAST UPDATE DATE: 02/26/1999|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |ELMO COLLINS R4 |
| |JOHN HICKEY NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: MIKE ZORN | |
| HQ OPS OFFICER: BOB STRANSKY | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|BAAA 20.1906(d) SURFACE CONTAMINATION E| |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| PACKAGE RECEIVED WITH SURFACE CONTAMINATION ABOVE LIMITS |
| |
| At 0810 PST, a courier delivered a shipment of radiopharmaceuticals from the |
| SYNCOR pharmacy in San Diego, CA. Upon receipt, the licensee performed a |
| routine wipe sample of the external surfaces of the outer container (ammo |
| box), and discovered contamination in excess of the reporting requirements |
| of 10 CFR 20.1906. Initial wipes indicated up to 30,000 CPM of gross |
| activity for a swab that had been run over all surfaces of the container. |
| The package contained two vials of radiopharmaceuticals; 10 mCi of Tc-99m |
| META solution, and 10 mCi of Ga-67 (not NRC regulated). The licensee did |
| not report any damage to the vials, and they were administered to patients. |
| No contamination occurred at the medical center as a result of this |
| shipment. |
| |
| A more detailed survey of the container revealed up to 50,000 cpm/300cm2. |
| The licensee did not determine the isotope of the contaminant. A |
| representative of SYNCOR visited the medical center, and took several wipe |
| samples for isotopic identification. The licensee plans to investigate this |
| event with SYNCOR. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35416 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: COOK REGION: 3 |NOTIFICATION DATE: 02/27/1999|
| UNIT: [] [2] [] STATE: MI |NOTIFICATION TIME: 00:31[EST]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 02/26/1999|
+------------------------------------------------+EVENT TIME: 21:30[EST]|
| NRC NOTIFIED BY: BRIAN MUTZ |LAST UPDATE DATE: 02/27/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |BRUCE JORGENSEN R3 |
|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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 N N 0 Cold Shutdown |0 Cold Shutdown |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - Potentially excessive thermal stress on two containment penetrations due |
| to blocked cooling flow - |
| |
| At 2130 on 02/26/99, with Unit 2 in cold shutdown mode in a refueling |
| outage, the Licensee determined that on 08/02/96, with Unit 2 at 100% power, |
| a temporary plant modification was implemented on Unit 2 which allowed for |
| continued power operation with component cooling water (CCW) to containment |
| penetrations #CPN-3 and #CPN-4 isolated. These penetrations contain the |
| steam generator #2 and #3 main steam headers. This condition is reportable |
| under 10CFR50.72(b)(2)(i) as an event found while the reactor is shutdown, |
| that, had it been found while the reactor was in operation, would have |
| resulted in the nuclear power plant, including its principal safety |
| barriers, being in a seriously degraded condition that significantly |
| compromised plant safety. The component cooling water return header |
| upstream of the containment isolation valve, #2-CCR-441, (containment |
| penetrations #CPN-3 and #CPN-4 inner cooling coils CCW outlet containment |
| isolation valve) was discovered to contain blockage during a post |
| maintenance activity associated with the repair of valve #2-CCR-441. The |
| blocked line eliminated cooling flow to the penetration inner coolers, which |
| is designed to assure integrity of the penetration sleeve. The result of |
| operating with the CCW isolated to penetrations #CPN-3 and #CPN-4 was the |
| creation of potentially excessive thermal stress on the penetration sleeves. |
| Design basis information indicates that the penetration sleeve may be |
| exposed to temperatures of as high as 150�F without experiencing |
| degradation. It is estimated that the penetration sleeves on penetrations |
| #CPN-3 and #CPN-4 were operated at a temperature approximating main steam |
| temperature of approximately 600�F. |
| |
| This condition was identified during an expanded system readiness review. |
| No immediate corrective action is planned since the main steam system is |
| currently out of service and containment integrity is not required in the |
| current operational mode. Further analysis and corrective action will be |
| considered during ongoing investigation under the corrective action |
| program. |
| |
| The Licensee notified the NRC Resident Inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35417 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: CRYSTAL RIVER REGION: 2 |NOTIFICATION DATE: 02/27/1999|
| UNIT: [3] [] [] STATE: FL |NOTIFICATION TIME: 09:07[EST]|
| RXTYPE: [3] B&W-L-LP |EVENT DATE: 02/27/1999|
+------------------------------------------------+EVENT TIME: 08:10[EST]|
| NRC NOTIFIED BY: LARRY MOFFATT |LAST UPDATE DATE: 02/27/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |ROBERT HAAG R2 |
|10 CFR SECTION: | |
|APRE 50.72(b)(2)(vi) OFFSITE NOTIFICATION | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|3 N Y 100 Power Operation |100 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - STATE NOTIFIED OF A KEMP'S RIDLEY SEA TURTLE RETRIEVED FROM THE PLANT |
| INTAKE WATER - |
| |
| At 2232 on 02/26/99, a young Kemp's Ridley sea turtle was taken from the |
| water at the intake of Crystal River Unit 3. The turtle was found pinned |
| against the bar rack and was retrieved by site personnel in accordance with |
| the Florida Power Corporation Turtle Protection Guidelines. Crystal River |
| Mariculture Center personnel took custody of the sea turtle and will return |
| it to the Gulf of Mexico In the afternoon of 02/27/99. At 0810 on 02/27/99, |
| the Florida Department of Environmental Protection was notified of the |
| retrieval of the sea turtle. |
| |
| The licensee notified the NRC Resident Inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35418 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: GINNA REGION: 1 |NOTIFICATION DATE: 02/27/1999|
| UNIT: [1] [] [] STATE: NY |NOTIFICATION TIME: 13:37[EST]|
| RXTYPE: [1] W-2-LP |EVENT DATE: 02/27/1999|
+------------------------------------------------+EVENT TIME: 11:39[EST]|
| NRC NOTIFIED BY: DOUGLAS GOMEZ |LAST UPDATE DATE: 02/27/1999|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |RICHARD BARKLEY 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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 70 Power Operation |70 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| CONTAINMENT VENTILATION ISOLATION DURING I&C WORK ON RADIATION MONITOR |
| |
| "During Instrument and Control (I/C) activities on radiation channel R-12, |
| an unexpected Containment Ventilation Isolation (CVI) occurred. The planning |
| activities for this maintenance recognized the potential of generating a CVI |
| signal and directed the technicians to install a jumper to prevent the |
| actuation. Even with the jumper installed an unexpected CVI occurred when |
| the R-12 drawer was deenergized. The CVI was therefore due to maintenance |
| activities and was not the result of an actual high radiation condition. |
| |
| "No plant system other than the containment ventilation monitoring system |
| was affected by this event. The plant is stable at approximately 70% power |
| with a plant coastdown in progress. |
| |
| "This event is reportable under lOCFR50.72(b)(2)(ii), 'Any condition that |
| results in a manual or automatic actuation an Engineered Safety Feature.'" |
| |
| The NRC resident inspector has been informed of this notification. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35419 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: FITZPATRICK REGION: 1 |NOTIFICATION DATE: 02/27/1999|
| UNIT: [1] [] [] STATE: NY |NOTIFICATION TIME: 22:39[EST]|
| RXTYPE: [1] GE-4 |EVENT DATE: 02/27/1999|
+------------------------------------------------+EVENT TIME: 21:56[EST]|
| NRC NOTIFIED BY: STEVE CAROLIN |LAST UPDATE DATE: 02/27/1999|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |RICHARD BARKLEY R1 |
|10 CFR SECTION: | |
|AINT 50.72(b)(1)(vi) INTERNAL THREAT | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |65 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| FIRE ONSITE LASTING LESS THAN 10 MINUTES |
| |
| At 2156, the 'A' circulating water pump tripped, and the control room |
| received indication of a fire in the pump motor. The onsite fire brigade |
| responded, and the fire was extinguished at 2204. The licensee reported |
| that the pump motor does not appear to be extensively damaged, and that no |
| other equipment was involved in the fire. Reactor power was reduced to 65% |
| of rated due to the unavailability of the circulating water pump. No |
| personnel injuries were reported. |
| |
| The licensee will inform the NRC resident inspector of this event. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35420 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: SALEM REGION: 1 |NOTIFICATION DATE: 02/28/1999|
| UNIT: [1] [] [] STATE: NJ |NOTIFICATION TIME: 02:55[EST]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 02/28/1999|
+------------------------------------------------+EVENT TIME: 01:38[EST]|
| NRC NOTIFIED BY: JACK GRANT |LAST UPDATE DATE: 02/28/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |RICHARD BARKLEY R1 |
|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 60 Power Operation |0 Hot Standby |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| - AUTO Rx TRIP FROM 60% DUE TO MAIN TURBINE TRIP DUE TO LOW AUTO STOP OIL |
| PRESSURE - |
| |
| AT 0130 ON 02/28/99, THE UNIT 1 REACTOR AUTO TRIPPED FROM 60% POWER DUE TO A |
| MAIN TURBINE TRIP (WITH REACTOR POWER ABOVE THE P-9 SETPOINT OF 50% POWER) |
| DUE TO LOW AUTO STOP OIL PRESSURE. ALL CONTROL RODS INSERTED COMPLETELY. |
| THE AUXILIARY FEEDWATER SYSTEM AUTO STARTED TO MAINTAIN STEAM GENERATORS AT |
| NORMAL WATER LEVELS. NO SAFETY OR RELIEF VALVES LIFTED AND STEAM IS BEING |
| DUMPED TO THE MAIN CONDENSER. UNIT 1 IS STABLE IN MODE 3 (HOT STANDBY). |
| THE LICENSEE IS INVESTIGATING THE CAUSE OF THE LOW AUTO STOP OIL PRESSURE. |
| |
| THE LICENSEE PLANS TO INFORM THE NRC RESIDENT INSPECTOR. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35421 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: VOGTLE REGION: 2 |NOTIFICATION DATE: 02/28/1999|
| UNIT: [1] [] [] STATE: GA |NOTIFICATION TIME: 03:14[EST]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 02/27/1999|
+------------------------------------------------+EVENT TIME: 23:40[EST]|
| NRC NOTIFIED BY: CHUCK MEYER |LAST UPDATE DATE: 02/28/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |ROBERT HAAG R2 |
|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 M/R Y 18 Power Operation |0 Hot Standby |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| -MAN Rx TRIP FROM 18% DUE TO NUCLEAR INSTRUMENTS ANOMALY DURING PLANT |
| SHUTDOWN- |
| |
| WHILE SHUTTING UNIT 1 DOWN FOR A PLANNED REFUELING OUTAGE, CONTROL ROOM |
| OPERATORS MANUALLY TRIPPED UNIT 1 FROM 18% POWER DUE TO A CONCERN THAT THE |
| NUCLEAR INSTRUMENTS INTERMEDIATE RANGE FLUX TRIP WOULD NOT RESET BEFORE THE |
| POWER RANGE (P-10) AUTO UNBLOCK OCCURRED. ALL CONTROL RODS INSERTED |
| COMPLETELY. NO SAFETY OR RELIEF VALVES LIFTED. CONTROL ROOM OPERATORS |
| MANUALLY ACTUATED THE AUXILIARY FEEDWATER SYSTEM TO MAINTAIN STEAM |
| GENERATORS AT THEIR NORMAL WATER LEVELS. UNIT 1 IS STABLE IN MODE 3 (HOT |
| STANDBY). THE LICENSEE PLANS TO INVESTIGATE THE CAUSE OF THE NUCLEAR |
| INSTRUMENTS ANOMALY AND PROCEED WITH THE PLANNED REFUELING OUTAGE. |
| |
| THE LICENSEE INFORMED THE NRC RESIDENT INSPECTOR. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35422 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: KEWAUNEE REGION: 3 |NOTIFICATION DATE: 02/28/1999|
| UNIT: [1] [] [] STATE: WI |NOTIFICATION TIME: 18:41[EST]|
| RXTYPE: [1] W-2-LP |EVENT DATE: 02/28/1999|
+------------------------------------------------+EVENT TIME: 16:50[CST]|
| NRC NOTIFIED BY: CRAIG BYALL |LAST UPDATE DATE: 02/28/1999|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |BRUCE JORGENSEN R3 |
|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 |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 97 Power Operation |97 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| UNIT PLACED IN UNANALYZED CONDITION DUE TO CLOSURE OF CONTAINMENT ISOLATION |
| VALVE |
| |
| At 0400 CST on 02/28/99, the plant was incorrectly placed in an unanalyzed |
| condition when a manual valve between the reactor coolant drain tank and the |
| chemical volume control holdup tank was closed. This manual valve was |
| located downstream of two containment isolation valves that had failed |
| timing tests, and the manual valve was being relied upon to maintain |
| containment integrity in accordance with NRC Generic Letter 96-06. However, |
| when the manual valve was closed, overpressure protection for that line was |
| lost. The valve subsequently was reopened, restoring the penetration at |
| 0800 CST on 02/28/99. The reportability of this condition was identified at |
| 1650 CST. The NRC resident inspector has been informed of this |
| notification. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35423 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: SUSQUEHANNA REGION: 1 |NOTIFICATION DATE: 02/28/1999|
| UNIT: [1] [] [] STATE: PA |NOTIFICATION TIME: 22:54[EST]|
| RXTYPE: [1] GE-4,[2] GE-4 |EVENT DATE: 02/28/1999|
+------------------------------------------------+EVENT TIME: 22:00[EST]|
| NRC NOTIFIED BY: DAVID WALSH |LAST UPDATE DATE: 02/28/1999|
| HQ OPS OFFICER: BOB STRANSKY +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |RICHARD BARKLEY R1 |
|10 CFR SECTION: | |
|AOUT 50.72(b)(1)(ii)(B) OUTSIDE DESIGN BASIS | |
|AINB 50.72(b)(2)(iii)(B) POT RHR INOP | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| UNIT OUTSIDE OF DESIGN BASIS DUE TO SHEARED VALVE STEM IN RHR SYSTEM |
| |
| "On 2/11/99 the Unit 1 'B' RHR Loop was removed from service for a scheduled |
| maintenance work window. During the system restoration at 2330 Hrs, it was |
| identified that the keepfill system did not respond as expected. An |
| investigation into the degraded keepfill condition was initiated. An |
| Operability Determination was performed and it was determined that the RHR |
| system was operable with the degraded keepfill system. |
| |
| "On 2/16/99 at 0400 hrs, the Unit 1 'A' RHR Loop was removed from service to |
| perform a scheduled maintenance work window. The 'A' RHR Loop was returned |
| to service at 2115 hrs on 2/16/99 and is currently operable. |
| |
| "On 2/26/99, after further trouble shooting of the degraded keepfill |
| condition on the 'B' RHR Loop it was determined that the most likely cause |
| was the RHR Loop 'B' Injection Flow Control Valve, HVI51F017B, being failed |
| closed. The 'B' RHR Loop was declared inoperable at 1600 Hrs on 2/26/99. The |
| valve was inspected and found to have the stem sheared from the disk. |
| Following a review of the time line of the events, it was identified that |
| both the 'A' and 'B' RHR Loops were inoperable from 0400 hrs to 2115 hrs on |
| 2/16/99 during the scheduled maintenance work windows for the 'A' RHR Loop. |
| |
| "This report is being made due to the Plant being Outside of the Design |
| Basis requiring a 1 Hr ENS Notification under 10CFR50.72(b)(1)(ii)(B) and a |
| Loss of a Safety System requiring a 4 Hr ENS notification under |
| 1OCFR50.72(b)(2)(iii)(B)." |
| |
| The NRC resident inspector has been informed of this notification. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 35424 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: OCONEE REGION: 2 |NOTIFICATION DATE: 03/01/1999|
| UNIT: [] [2] [] STATE: SC |NOTIFICATION TIME: 00:17[EST]|
| RXTYPE: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-L|EVENT DATE: 02/28/1999|
+------------------------------------------------+EVENT TIME: 20:40[EST]|
| NRC NOTIFIED BY: MIKE HILL |LAST UPDATE DATE: 03/01/1999|
| HQ OPS OFFICER: DICK JOLLIFFE +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |ROBERT HAAG R2 |
|10 CFR SECTION: | |
|ARPS 50.72(b)(2)(ii) RPS ACTUATION | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 A/R Y 98.5 Power Operation |0 Hot Shutdown |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| -AUTO Rx TRIP ON HIGH RCS PRESSURE DUE TO MAIN TURBINE CONTROL VALVES |
| FAILING CLOSED- |
| |
| At 1609 on 02/28/99, the Unit 2 electro-hydraulic control system lost |
| various power supplies. Main steam pressure increased from a normal 900 psig |
| to 942 psig and reactor power increased from 100% to 100.4%. The main |
| turbine control valves had throttled closed for unknown reasons causing the |
| main steam pressure to increase. Main feedwater was throttled to reduce |
| main steam header pressure since the turbine header pressure control station |
| had no effect. Unit 2 was stabilized at 98.5% power with the main steam |
| pressure at 938 psig and the main feedwater master control stations and the |
| reactor control station in manual. |
| |
| At 2040 on 02/28/99, Unit 2 automatically tripped from 98.5% power due to a |
| reactor protection system actuation (reactor coolant system high pressure |
| trip). All control rods inserted completely. The main steam code safety |
| valves lifted to dump steam to the atmosphere for approximately 10 minutes. |
| Plant operators verified that the valves reseated properly. Steam is being |
| dumped to the main condenser. The main feedwater system remained |
| operational throughout the event. The reactor control station was in |
| automatic at the time of the trip. Unit 2 is stable in hot shutdown mode. |
| |
| The licensee is investigating the cause of the main turbine control valves |
| failing closed and plans to make necessary repairs. |
| |
| Units 1 and 3 remain at 100% power and were unaffected by this event. |
| |
| The licensee notified the NRC Resident Inspector. |
+------------------------------------------------------------------------------+
Page Last Reviewed/Updated Thursday, March 25, 2021
Page Last Reviewed/Updated Thursday, March 25, 2021