Event Notification Report for August 26, 2002
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
08/23/2002 - 08/26/2002
** EVENT NUMBERS **
39141 39144 39148 39149 39150 39151 39152
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|General Information or Other |Event Number: 39141 |
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| REP ORG: WA DIVISION OF RADIATION PROTECTION |NOTIFICATION DATE: 08/21/2002|
|LICENSEE: PROVIDENCE EVERETT MEDICAL CENTER |NOTIFICATION TIME: 15:02[EDT]|
| CITY: EVERETT REGION: 4 |EVENT DATE: 08/19/2002|
| COUNTY: STATE: WA |EVENT TIME: [PDT]|
|LICENSE#: WN-M0135-1 AGREEMENT: Y |LAST UPDATE DATE: 08/21/2002|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |WILLIAM JOHNSON R4 |
| |DOUG BROADDUS NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: TERRY C. FRAZEE (e-mail) | |
| HQ OPS OFFICER: MIKE NORRIS | |
+------------------------------------------------+ |
|EMERGENCY CLASS: NON EMERGENCY | |
|10 CFR SECTION: | |
|NAGR AGREEMENT STATE | |
| | |
| | |
| | |
| | |
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EVENT TEXT
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| AGREEMENT STATE REPORT INVOLVING MEDICAL MISADMINISTRATION |
| |
| "The licensee reported that a patient received 2640 cGy (rad) during a |
| cardiac intravascular brachytherapy treatment instead of the intended 2000 |
| cGy (rad), a 32% overexposure. The patient was being treated with the |
| Guidant Corporation Galileo intravascular brachytherapy high dose rate |
| remote afterloader device (serial #27958502) with a model GDT-P32-2 source |
| wire (serial #020807016) containing 4.44 GBq (119.9 [millicuries]) of P-32 |
| at time of treatment. The patient's vessel size was larger than the |
| automatically calculated maximum diameter treatment. A manual calculation |
| of dwell time was required, based on the dose rate tables available in the |
| Guidant Manual (section 6.13 table 5). However, the dose rate for a 4.6 mm |
| diameter (3.30 mm treatment depth) was inadvertently used instead of 4.05 mm |
| diameter (3.03 mm treatment depth). This resulted in a delivered dose of |
| 2640 cGy (rad) at 3.03 mm. The cause of the event is human error. The |
| licensee's corrective action is to have a second independent calculation |
| performed by Physics and Dosimetry staff prior to treatment whenever a |
| manual calculation using the dose rate tables is necessary. No adverse |
| consequences are expected. The referring physician and the patient were |
| notified of the overexposure." |
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|Power Reactor |Event Number: 39144 |
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| FACILITY: MILLSTONE REGION: 1 |NOTIFICATION DATE: 08/22/2002|
| UNIT: [] [] [3] STATE: CT |NOTIFICATION TIME: 17:06[EDT]|
| RXTYPE: [1] GE-3,[2] CE,[3] W-4-LP |EVENT DATE: 08/22/2002|
+------------------------------------------------+EVENT TIME: 16:33[EDT]|
| NRC NOTIFIED BY: MICHAEL MARTELL |LAST UPDATE DATE: 08/23/2002|
| HQ OPS OFFICER: FANGIE JONES +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |CLIFFORD ANDERSON R1 |
|10 CFR SECTION: | |
|AUNA 50.72(b)(3)(ii)(B) UNANALYZED CONDITION | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
| | |
|3 N Y 95 Power Operation |95 Power Operation |
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EVENT TEXT
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| UNANALYZED CONDITION CONCERNING STEAM GENERATOR ATMOSPHERIC RELIEF VALVE |
| BYPASS VALVES |
| |
| Historical analysis deficiencies associated with the steam generator |
| atmospheric dump bypass valves, a condition that during a fire could cause |
| seriously degrade the safety of the plant. |
| |
| "The system affected is main steam, there are no actuation signals. The |
| cause is historical analysis deficiencies. There are no affects on the |
| plant. There are no actions taken or planned at this time and there is no |
| additional information. The NRC Resident Inspector was notified. The State |
| and Local Authorities have been notified." |
| |
| * * * UPDATE 1605EDT ON 8/23/02 FROM MICHAEL MARTELL TO S. SANDIN * * * |
| |
| The following information from Millstone Condition Report CR-02-08666 was |
| provided by the licensee as an update: |
| |
| "On August 22, 2002, with Millstone Unit 3 in Mode 1, and as a result of |
| transient analysis of the fire shutdown scenarios, the potential for |
| spurious operation associated with the Main Steam atmospheric dump valve |
| bypass motor operated valves (3MS*MOV74A thru D) from a hot short was found |
| to result in non-compliance with the BTP 9.5-1 performance criteria. The |
| Millstone Unit 3 licensing basis regarding BTP 9.5-1 performance criteria |
| provides that a fire in an alternate shutdown area shall pose a transient no |
| more severe than a reactor trip due to loss of normal power and that |
| pressurizer level remain in the indicating range. Spurious operation of |
| plant components due to credible hot shorts must be postulated and have |
| methods for mitigation that do not require the use of any other component |
| potentially affected by the fire. The current BTP 9.5-1 compliance report |
| and shutdown methods address the spurious opening of the Main Steam |
| atmospheric dump valve bypass motor operated valves. Fire safe shutdown |
| procedures provide methods for mitigating this spurious actuation. |
| |
| "On-going effort to validate assumptions used in the fire safe shutdown |
| analysis concluded that insufficient operator response time was available to |
| support mitigation of this transient. It has been determined that spurious |
| opening of one of the dump valves could cause pressurizer level to go below |
| the indicating range in a time frame shorter than originally assumed. This |
| condition is being reported as an unanalyzed condition that significantly |
| degrades plant safety, pursuant to 50.72 (b)(3)(ii)(B). |
| |
| "It should be noted that, consistent with the Licensing Basis, this analysis |
| assumes a fire of sufficient magnitude to result in an instantaneous hot |
| short. Such fires are viewed to be very low probability events. The fire |
| areas of concern are the control room and cable spreading area. The control |
| room is continuously manned, and currently a continuous compensatory fire |
| watch is required in the cable spreading room for other conditions. No |
| additional compensatory actions are judged to be necessary at this time. |
| Corrective actions are under evaluation in accordance with the Millstone |
| corrective action program." |
| |
| The licensee informed the NRC Resident Inspector. Notified R1DO(Noggle). |
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|General Information or Other |Event Number: 39148 |
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| REP ORG: GENERAL ELECTRIC COMPANY |NOTIFICATION DATE: 08/23/2002|
|LICENSEE: GE NUCLEAR ENERGY |NOTIFICATION TIME: 15:45[EDT]|
| CITY: SAN JOSE REGION: 4 |EVENT DATE: 08/23/2002|
| COUNTY: STATE: CA |EVENT TIME: [PDT]|
|LICENSE#: AGREEMENT: Y |LAST UPDATE DATE: 08/23/2002|
| DOCKET: |+----------------------------+
| |PERSON ORGANIZATION |
| |WILLIAM JOHNSON R4 |
| |KEN BARR R2 |
+------------------------------------------------+VERN HODGE - FAX NRR |
| NRC NOTIFIED BY: JASON POST | |
| HQ OPS OFFICER: FANGIE JONES | |
+------------------------------------------------+ |
|EMERGENCY CLASS: NON EMERGENCY | |
|10 CFR SECTION: | |
|CCCC 21.21 UNSPECIFIED PARAGRAPH | |
| | |
| | |
| | |
| | |
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EVENT TEXT
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| 10 CFR 21 REPORT: MAIN STEAM LINE OUT-OF-SERVICE |
| |
| The following is taken from a facsimile report: |
| |
| "This letter provides notification of a Reportable Condition under 10CFR |
| 21.21(d) and as an interim report per �21.21.(a)(2) for other plants that |
| may be determined to be affected. The basis for this conclusion is that a |
| 1988 GE Nuclear Energy (GE) analysis for Brunswick Units 1 and 2 full power |
| operation with one Main Steamline Isolation Valve (MSIV) Out of Service |
| (OOS) provided to Carolina Power and Light (CP&L) did not adequately address |
| the increased flow induced vibratory loads on the MSIVs to assure they would |
| be able to perform their required safety function which could result in |
| potential offsite exposures in excess of those in 10CFR100.11. |
| |
| "The GE MSIV OOS analysis evaluated plant operation at 100% power with three |
| active steamlines and one set of MSIVs closed (OOS). The GE analysis did not |
| address the increased steam flow hardware effect of potential long-term flow |
| induced vibration degradation on the MSIVs, including the effect on the MSIV |
| air operated controls. During three steamline operation the steam flow in |
| each line would increase to approximately 133% of normal flow. No vibration |
| measurements (empirical or experimental data) exist for either Brunswick |
| Units during operation up to this increased steam flow level. |
| |
| "If it is postulated that the plant operated for an extended period in the |
| MSIV OOS condition and then a main steam line break is postulated to occur |
| in one of the three operational steam lines, then there is the potential |
| that neither MSIV would close to terminate the release from a steamline |
| break. Because GE has no analytical or experience basis (no available |
| empirical or experimental data) to support higher main steam line flow rates |
| greater than previously tested, it could be postulated that a common mode |
| failure of both MSIVs, in the broken line, could occur. Alternatively, |
| failure of one MSIV due to the high flow induced vibration and the other |
| MSIV as the design basis single failure, would result in an un-terminated |
| release, which would exceed the existing 10 CFR 100 radiation release |
| limits. |
| |
| "GE has verbally communicated to CP&L the need for the 75% power limitation |
| when exercising the MSIV OOS flexibility and will follow-up with a written |
| communication. |
| |
| "GE is reviewing all other MSIV OOS analyses performed by GE for other BWRs |
| and will communicate to any similarly affected utilities, similar corrective |
| actions. GE will notify all affected utilities that GE recommends operation |
| at the 75% power level when operating with one MSIV 005, unless there is |
| sufficient test data to support operation at a higher power level. This |
| effort will be completed by September 30, 2002." |
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|Power Reactor |Event Number: 39149 |
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| FACILITY: ROBINSON REGION: 2 |NOTIFICATION DATE: 08/24/2002|
| UNIT: [2] [] [] STATE: SC |NOTIFICATION TIME: 08:50[EDT]|
| RXTYPE: [2] W-3-LP |EVENT DATE: 08/24/2002|
+------------------------------------------------+EVENT TIME: 07:50[EDT]|
| NRC NOTIFIED BY: DON KNIGHT |LAST UPDATE DATE: 08/24/2002|
| HQ OPS OFFICER: RICH LAURA +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |KEN BARR R2 |
|10 CFR SECTION: | |
|ACOM 50.72(b)(3)(xiii) LOSS COMM/ASMT/RESPONSE| |
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| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|2 N Y 100 Power Operation |100 Power Operation |
| | |
| | |
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EVENT TEXT
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| PLANNED OUTAGE ON ERFIS SYSTEM |
| |
| "Emergency Response Data System (ERDS) and Safety Parameter Display System |
| (SPDS) inoperable due to planned outage of the Emergency Response Facility |
| Information System (ERFIS). |
| |
| "At 07:50 Hours on August 24, 2002, the H. B. Robinson Steam Electric Plant, |
| Unit No. 2, ERFIS computer system was removed from service for a planned |
| outage to upgrade this system. The expected duration of the outage is |
| approximately 5 days. During this time, the ERDS and SPDS will be |
| unavailable. Alternate means are identified in the HBRSEP, Unit No. 2, |
| emergency preparedness procedures to collect and distribute data that would |
| normally be available by the SPDS. The HBRSEP, Unit No. 2, Emergency |
| Response Organization remains able to respond to an event during the time |
| the ERFIS is unavailable. |
| |
| "The planned ERFIS changes. when completed, are not expected to change the |
| ERDS transmission format or data point library. Therefore, additional |
| reports in accordance with 10 CFR 50, Appendix E, VI. 3. a and b are not |
| expected. |
| |
| "The NRC Resident Inspector has been notified." |
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|Power Reactor |Event Number: 39150 |
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| FACILITY: CALLAWAY REGION: 4 |NOTIFICATION DATE: 08/24/2002|
| UNIT: [1] [] [] STATE: MO |NOTIFICATION TIME: 09:33[EDT]|
| RXTYPE: [1] W-4-LP |EVENT DATE: 08/24/2002|
+------------------------------------------------+EVENT TIME: 08:26[CDT]|
| NRC NOTIFIED BY: STEVE KOCHERT |LAST UPDATE DATE: 08/24/2002|
| HQ OPS OFFICER: RICH LAURA +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |WILLIAM JOHNSON R4 |
|10 CFR SECTION: | |
|ACOM 50.72(b)(3)(xiii) LOSS COMM/ASMT/RESPONSE| |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
| | |
| | |
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EVENT TEXT
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| PLANNED LOSS OF THE EOF FOR 20 HOURS |
| |
| "During performance of planned maintenance, the Emergency Operation Facility |
| (EOF) will be without ventilation capabilities. These maintenance |
| activities, including electrical isolation and restoration are expected to |
| last approximately 20 hours. Contingency plans for emergency situations have |
| been established. |
| |
| "This event is reportable per 10 CFR50.72(b)(3)(xiii) since this constitutes |
| a loss of an emergency response facility for the duration of the evolution. |
| |
| "The NRC Resident has been notified." |
| |
| * * * UPDATE AT 1445EDT ON 8/24/02 FROM STEVE KOCHERT TO S. SANDIN * * * |
| |
| At 1335CDT on 8/24/02 the EOF was declared functional and restored to |
| operation. |
| |
| The licensee will inform the NRC Resident Inspector. Notified |
| R4DO(Johnson). |
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|Power Reactor |Event Number: 39151 |
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| FACILITY: CALVERT CLIFFS REGION: 1 |NOTIFICATION DATE: 08/24/2002|
| UNIT: [1] [2] [] STATE: MD |NOTIFICATION TIME: 10:40[EDT]|
| RXTYPE: [1] CE,[2] CE |EVENT DATE: 08/24/2002|
+------------------------------------------------+EVENT TIME: 07:50[EDT]|
| NRC NOTIFIED BY: LEO GETZ |LAST UPDATE DATE: 08/24/2002|
| HQ OPS OFFICER: RICH LAURA +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |JAMES NOGGLE R1 |
|10 CFR SECTION: | |
|ACOM 50.72(b)(3)(xiii) LOSS COMM/ASMT/RESPONSE| |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|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 Y 100 Power Operation |100 Power Operation |
| | |
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EVENT TEXT
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| LOSS OF CALVERT COUNTY EP SIREN ACTIVATION SYSTEM DUE TO LIGHTNING STRIKES |
| |
| "At 07:50 on 8/24/02, notified by system engineer that Calvert County sirens |
| are out of service due to multiple lightning strikes affecting the Calvert |
| control center siren activation circuit. The county is using route alerting |
| as a compensatory measure." |
| |
| The NRC resident inspector was notified. |
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|Power Reactor |Event Number: 39152 |
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| FACILITY: BEAVER VALLEY REGION: 1 |NOTIFICATION DATE: 08/25/2002|
| UNIT: [] [2] [] STATE: PA |NOTIFICATION TIME: 00:45[EDT]|
| RXTYPE: [1] W-3-LP,[2] W-3-LP |EVENT DATE: 08/24/2002|
+------------------------------------------------+EVENT TIME: 20:00[EDT]|
| NRC NOTIFIED BY: PETE SENA |LAST UPDATE DATE: 08/25/2002|
| HQ OPS OFFICER: RICH LAURA +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: NON EMERGENCY |JAMES NOGGLE R1 |
|10 CFR SECTION: | |
|AUNA 50.72(b)(3)(ii)(B) UNANALYZED CONDITION | |
|AIND 50.72(b)(3)(v)(D) ACCIDENT MITIGATION | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
| | |
|2 N Y 100 Power Operation |100 Power Operation |
| | |
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EVENT TEXT
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| GAS VOIDING IN ECCS PIPING |
| |
| "At 0425 hrs on 8/24/2002, a gas void was identified in Emergency Core |
| Cooling System (ECCS) piping at Beaver Valley Power Station (BVPS) Unit No. |
| 2 that exceeded the gas void volume limit of .872 cubic feet. A gas void |
| which exceeds .872 cubic feet could potentially disable a single High Head |
| Safety Injection (HHSI) pump if ingested. The gas void was located in the |
| 'B' train piping which would be used (only) following the |
| transfer-to-recirculation phase of a Loss of Coolant Accident (LOCA). |
| Technical Specification Action 3.5.2.a and 3.5.2.d was entered for 'B' ECCS |
| train not being operable. The piping where the void was located leads to a |
| common HHSI pump suction header which connects to both trains' HHSI pumps. |
| |
| "At 1345 hrs on 8/24/2002, an isolation valve (2SIS-MOV863B) was |
| de-energized closed. De-energizing this isolation valve prevents the gas |
| void traveling to the common HHSI suction header during |
| transfer-to-recirculation flow. This was done as a general precaution to |
| strengthen the operable 'A' HHSI train during the ongoing gas void |
| generation investigation since this gas void generation process was not yet |
| fully understood. |
| |
| "At 1638 hrs on 8/24/2002 it was calculated that the actual gas void volume |
| in the 'B' train piping was 1.3 cubic feet. It was also identified that the |
| previously established gas void volume limit of .872 cubic feet was |
| incorrect and the applicable gas void volume limit was .319 cubic feet. With |
| an evaluation of the new gas void limit, it was concluded at 2000 hrs that |
| BVPS Unit No. 2 had been vulnerable to a degradation of both trains' HHSI |
| pumps between 0425 and 1345. This would be possible since the gas void could |
| potentially have split in half (0.65 cubic feet) and migrated during |
| post-LOCA transfer-to-recirculation flow through the common HHSI suction |
| header. Each half-sized void could enter each train's HHSI pump, potentially |
| affecting both trains of HHSI pumps (.65 cubic feet would exceed the limit |
| of .319 cubic feet for each pump). This is reportable pursuant to |
| 10CFR50.72(b)(3)(ii)(B) as being in an unanalyzed condition that |
| significantly degraded plant safety. This is also reportable pursuant to |
| 10CFR50.72(b)(3)(v)(D) as a condition that at the time of discovery could |
| have prevented the fulfillment of the safety function of systems needed to |
| mitigate consequences of an accident. |
| |
| "Currently with 2SIS-M0V863B de-energized closed, the gas void can not |
| travel to the 'A' train HHSI pump. Actions are being initiated to eliminate |
| this gas void. BVPS Unit No. 2 remains in Tech Specification Action 3.5.2.a |
| and 3.5.2.d for one ECCS subsystem inoperable. The investigation of the gas |
| void generation process is continuing." |
| |
| The NRC Resident Inspector was notified. |
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