Event Notification Report for March 28, 2005

UNUSUAL EVENT DECLARED DUE TO HYDROGEN LEAK

The licensee faxed the following information:

"Following the Unit 2 trip, (due to a malfunction of the generator protection circuitry), a hydrogen leak was identified on the Unit 2 main generator. The leakage was sufficient enough to cause a flammable gas release affecting normal plant operation. The Unusual Event was declared under HU6 - Hazards and Other Conditions. The State and local authorities were notified at 14:09 CST. There is NO fire, it is a hydrogen release only."

The licensee said the hydrogen totalizer, which indicated a flow rate of 100 cfm, may be a probable area of leakage. The hydrogen leaked directly into the turbine building. Air samples indicated personnel breathing apparatus was not required. The licensee expects to exit from the UE once the hydrogen system was purged with CO2. Estimated timeframe is 3-6 hours.

The licensee notified the NRC Resident Inspector.

*** UPDATE FROM D. BRAGLIA TO J. KNOKE AT 17:40 EST ON 3/28/05 ***

The licensee terminated their Unusual Event at 16:40 CST, and the plant status is 0% power / Mode 3. The hydrogen leak was determined to be from a bushing on the main generator. The hydrogen leakage is believed to have lasted only 5 minutes.

The licensee notified the State and NRC Resident Inspector.

Notified R3DO (Riemer), IRD Manager (Crlenjak), NRR EO (Berkow), FEMA and DHS.

*** UPDATE FROM J. GERRITY TO J. KNOKE AT 14:30 EST ON 3/29/05 ***

Received Event Summary Report from Braidwood station. Corrected Unusual Event termination time from 16:40 to 16:23 CST. Notified R3DO (Riemer) and NRR EO (Brenner)

REACTOR TRIP DUE TO GENERATOR PROTECTION CIRCUITRY

The licensee faxed the following:

"Unit 2 reactor trip due to generator protection circuitry. Auxiliary feedwater actuated as expected. There were no additional malfunctions or unexpected plant response. The cause of the generator protection circuitry induced trip is still under investigation.

"This is a 4 hour notification of an RPS actuation per 10CFR 50.72(b)(2)(iv)(B). The 8 hour notification of an auxiliary feedwater system actuation per 10CFR 50.72(b)(3)(iv)(A) is being made under the same telephone call."

See Event 41534.

The licensee notified the NRC Resident Inspector.

RCS THROUGH WALL LEAKAGE IDENTIFIED AT WELDED PIPING JUNCTION

"At 1035 hours on the morning of 3/28/05, investigation into a previously reported leak (Event #41530) identified an approximate 3 inch through wall flaw at the welded junction of the 4 inch diameter decon port and the 28 inch diameter 'B' Reactor Recirculation Systems pump suction piping. The identification of the indication required the removal of insulation in order to complete the inspection. The unidentified leakage rate is below Tech Spec limits and the leak location is within the isolable boundary of the 'B' Reactor Recirculation Pump. This failure constitutes a welding or material defect in the primary coolant system that is not acceptable under ASME Section XI for ASME Code Class 1 piping.

"This report is being made under Hope Creek Event Classification Guide RAL# 11.2.1 in accordance with 10CFR50.72(b)(3)(ii). At the time of this notification, the Hope Creek Generating Station is in OPCON 4 at 150 degrees reactor coolant temperature to support piping repairs."

The licensee informed the Lower Alloways Creek Township and NRC Resident Inspector.

NOTIFICATION OF A FORCED OUTAGE EXPECTED TO LAST LONGER THAN THREE DAYS.

The licensee faxed the following information:

"At 1248, 03/28/05, Callaway Plant notified the Missouri Public Service Commission (MPSC) of a forced outage expecting to last longer than three days. This notification was performed in accordance with 4CSR2403.190(3)(B).

"The information was supplied to the MPSC:
* the plant shutdown occurred at 0520, 03/26/05 due to a minor leak in the Essential Service Water system,
and that additional repairs were going, to be performed while shutdown.
* the scope of the work addition had not been finalized and a formal restart date had not been established.

"This ENS notification is being made in accordance with 10CFR50.72(b)(2)(xi) for notification of another government agency."

The licensee notified other governmental agencies and the NRC Resident Inspector.

LOSS OF EMERGENCY SIREN CAPABILITY DUE TO POWER OUTAGE

"The purpose of this 8 Hour notification is to inform the Nuclear Regulatory Commission of a Loss of Emergency Preparedness Capabilities per NUREG 1022 and 10CFR50.72(b)(3)(xiii). Specifically, at 1455 central standard time on 3/28/05, Point Beach Nuclear Plant was notified that Emergency Plan Sirens, P012, P013, and P014 were out of service. These three sirens constitute 59.32% loss of population coverage.

"The time that the emergency sirens went out of service was 1343 CST, 3/28/05. These sirens are fed from Two Rivers Water/Light (WIPPI). Point Beach Nuclear Plant has contacted Manitowoc County Dispatch at 1517 of the loss of population coverage, and subsequent power restoration. Two Rivers Municipal Utilities has restored power to the Emergency Plan Sirens at 1533 on 3/28/05."

The licensee informed the NRC Resident Inspector.

*** UPDATE FROM J.FOUSE TO J. KNOKE AT 19:02 EST ON 03/28/05 ***

The licensee faxed the following update:
"The purpose of this update is to provide additional clarification to EN#41538. At 1343 CST, Emergency Plan Siren P012 went out of service due to a power outage. The loss of P012 represents a 3.38% loss in population coverage in and of itself. At 1502 CST, Emergency Plan Sirens P013, and P014 went out of service due to a power outage. When all three sirens are out, as they were from 1502 to 1533 CST, the loss of coverage to the population is 59.32%.

"Power was restored to Emergency Plan Sirens P012, P013, and P014, and they were returned to service at 1533 CST. The remainder of EN#41538 remains accurate as written."

The licensee informed the NRC Resident Inspector.
Notified the R3DO( Riemer).

HISTORICAL REPORT - AFTER-THE-FACT OPERABILITY ASSESSMENT

The licensee provided the following information:

"This event notification reports a previously unrecognized reportable event that occurred in 2002. This past event is being reported under the criterion of 10CFR50.72(b)(3)(ii)(B), an event or condition that results in the plant being in an unanalyzed condition.

"On 1/23/2002, at 1815 hours, while the plant was operating at full power, the plant entered a 72 hour Technical Specifications (TS) limiting conditions for operation (LCO) for one train of the Residual Heat Removal (RHR) system. The RHR system LCO was entered when the plant discovered a potential for reaching pump runout under specific post-accident operating conditions for single pump operation of the Component Cooling Water (CCW) system. The CCW system provides cooling support to the RHR system for post-accident long term recirculation core and containment cooling operation. The RHR train was returned to service at 2002 hours on 01/25/2002, within the TS allowed 72 hours, after a flow limiting device was installed on a non-critical CCW system flaw control air operated valve.

"An after-the-fact past operability assessment of the condition found that, compared to corrective actions taken to remove the potential for runout, the condition potentially existed since original plant design. As a minimum, it existed for a period longer than the TS allowed LCO for the RHR system. Therefore, an unanalyzed condition did exist and as such this is a late report."

The licensee notified the NRC Resident Inspector.

HIGH PRESSURE COOLANT INJECTION TEMPORARILY INOPERABLE

The following information was obtained from the licensee via facsimile (licensee text in quotes):

"Unit 2 High Pressure Coolant Injection (HPCI) [was] declared inoperable due to a loose control power fuse clip associated with the HPCI Pump Suction from Suppression Pool Valve HV-055-212041. The loose clip resulted in loss of control power to this DC motor operated valve and therefore rendered the HPCI system inoperable. The fuse clip has been replaced and the HPCI system returned to an operable status. This report is being made pursuant to 10CFR 50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of the safety function of a system that is needed to mitigate the consequences of an accident."

The licensee will notify the NRC Resident Inspector.

PLANNED OUTAGE OF SPDS DUE TO SYSTEM UPGRADE

The licensee faxed the following information:

"On Tuesday, March 29, 2005, Surry Power Station will remove a portion of the Emergency Response Facilities Computer System (ERFCS) in the Main Control Room (MCR) for planned upgrades to a new system for increased reliability and improved human interface design. For approximately 1 week, there will not be a qualified Safety Parameter Display System (SPDS) available in the MCR.

"During the time period that SPDS is unavailable in the MCR while the upgrade is occurring, the replacement SPDS system is available in the MCR on the Plant Computer System (PCS), but final testing will not qualify the software until completion of this work. There will also remain one ERFCS terminal operable in both the Technical Support Center (TSC) and Local Emergency Operations Facility (LEOF) with operable SPDS throughout the duration of this work. The Emergency Response Data System (ERDS) will remain available from those terminals for the duration of this planned work and normal data transmission capability will remain in the event of an emergency.

"Since SPDS will be out of service for greater than 1 hour, this report is being submitted in accordance with the guidance in 10CFR50.72(b)(3)(xiii) and NUREG-1022."

The licensee has notified the NRC Resident Inspector.

* * * UPDATE PROVIDED TO NRC (HUFFMAN) BY DEANE ON 1827 EDT ON 4/7/05 * * *

The following information was provided by the licensee via facsimile:

"This report is an update to previous report EN #41533, dated 03/28/05, regarding the availability of the Emergency Response Facilities Computer System (ERFCS) and Safety Parameter Display System (SPDS) in the main control room for Surry Power Station.

"The ERFCS and SPDS modifications and testing were completed at 1630 hours on 04/07/05. These functions have been transferred to the Surry Plant Computer System.

"Senior NRC resident, Norm Garrett was notified via alphanumeric page. Notification was made to NRC duty officer Bill Huffman."

R2DO (Rogers) notified.