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Event Notification Report for October 6, 2006

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
10/05/2006 - 10/06/2006

** EVENT NUMBERS **


42877 42878 42880

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Power Reactor Event Number: 42877
Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [ ] [2] [ ]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: DONALD C. SMITH
HQ OPS Officer: PETE SNYDER
Notification Date: 10/05/2006
Notification Time: 14:18 [ET]
Event Date: 08/18/2006
Event Time: 19:58 [CDT]
Last Update Date: 10/05/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
Person (Organization):
THOMAS DECKER (R2)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
2 N Y 100 Power Operation 100 Power Operation

Event Text

INVALID PRIMARY CONTAINMENT ISOLATION SYSTEM (PCIS) ACTUATION

"This 60-day telephone notification is being made under reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of general containment isolation signals affecting more than one system.

"On August 18, 2006, at 1958 hours CDT, with Unit 2 operating at 100% thermal power, the electrical power to reactor protection system (RPS) bus 2B was interrupted during the performance of surveillance testing on RPS circuit protectors 2B1 and 2B2. The RPS buses 2A and 2B are normally powered from motor-generator (MG) sets 2A and 2B, respectively. During testing or maintenance intervals affecting either the MG set or the normal supply circuit protectors, the affected RPS bus is powered from a transformer supply through alternate power circuit protectors 2C1 and 2C2. Power to RPS bus 2B, which was being powered through the alternate power circuit protectors via a temporary transformer, was interrupted when circuit protector 2C1 actuated on a sensed under voltage condition. The Primary Containment Isolation System (PCIS) logic circuits powered from RPS bus 2B were de-energized, and PCIS logic Groups 2, 3, 6, and 8 were actuated. None of the plant conditions which require PCIS Groups 2, 3, 6, or 8 actuation (e.g., low reactor water level, high drywell pressure, abnormal area radiation levels, high area temperature, etc.) existed; therefore, these actuations are considered invalid.

"The following actuations/isolations occurred:
"Group 2: Isolation of the Pressure Suppression Chamber head tank pumps and Drywell Floor and Equipment Drains Isolation;
"Group 3: Isolation of the reactor water clean-up system;
"Group 6: Initiation of the Standby Gas Treatment System, Initiation of Control Room Emergency Ventilation, and Isolation of the reactor zone and refuel zone normal ventilation systems;
"Group 8: This logic isolates the Traversing In-core Probes (TIP) if they are inserted. The TIPs were not inserted at the time of this event.

"All equipment responded in accordance with the plant design. At the time of the loss of power from the alternate source, the surveillance testing had already been completed and activities were in progress to transfer the RPS bus back to its normal supply. These actions were completed and RPS Bus 2B was re-energized from RPS MG set 2B. The affected logic was reset, and equipment was realigned as appropriate.

"There were no safety consequences or impacts on the health and safety of the public. The event was entered into TVA's corrective action program for evaluation and resolution.

"The NRC Senior Resident Inspector has been notified of this report.

"Reference corrective action document PER 109090."

Also see similar NRC event number 42837.

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Power Reactor Event Number: 42878
Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [ ] [ ] [3]
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: DONALD C. SMITH
HQ OPS Officer: PETE SNYDER
Notification Date: 10/05/2006
Notification Time: 14:18 [ET]
Event Date: 08/19/2006
Event Time: 20:02 [CDT]
Last Update Date: 10/05/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
Person (Organization):
THOMAS DECKER (R2)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
3 N N 0 Hot Shutdown 0 Hot Shutdown

Event Text

INVALID PRIMARY CONTAINMENT ISOLATION SYSTEM ACTUATION

"This 60-day telephone notification is being made under reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of multiple main steam isolation valves.

"On August 19, 2006, at 2002 hours CDT, with Unit 3 shutdown in Mode 3 following an earlier manual scram of Unit 3 (reference Event Number 42787), the EHC System along with the turbine bypass valves was being used to cool down/depressurize the reactor. The EHC system is used to supervise the turbine control valves and turbine bypass valves to control reactor pressure. The EHC system was subsequently removed from service for the repair of a previously identified system fluid leak.

"Following the shutdown of EHC system, reactor pressure and temperature were allowed to slowly increase. At 2240 hours CDT upon completion of the leak repair, operations personnel placed an EHC pump back in service in accordance with the system operating instruction to support post-maintenance testing activities. However, when the EHC system was returned to service, the pressure control set point was lower than the reactor pressure. The significance of the delta between the EHC setpoint and the actual reactor pressure was not recognized by operations. Because the actual reactor pressure was higher than the existing control set point, the EHC system responded by opening turbine bypass valves to lower the reactor pressure. Operations personnel observed the bypass valve response, identified the cause, and raised the pressure control set point. This action caused the bypass valves to rapidly close. The abrupt cessation of steam flow caused by the rapid closure of the bypass valves initiated a reactor pressure transient that affected the reactor water level instrumentation. The affected level instruments' output signals exhibited a ringing effect of a magnitude sufficient to reach the low level set point for primary containment isolation system (PCIS) Group 1 actuation. The Group 1 isolation logic actuated in accordance with its design, and the main steam isolation valves and the main steam line drain valves automatically closed.

"Designed time delays in other logic circuits affected by these water level signals prevented additional equipment actuation during this event. Actual reactor water level did not change, remaining within the normal level band; therefore, the isolation signal is considered invalid.

"All equipment responded in accordance with the plant design. Upon verification that no actual water level anomaly existed and that the transient instrumentation response had stabilized, the affected PCIS logic was reset, and equipment was realigned as appropriate.

"There were no safety consequences or impacts on the health and safety of the public. The event was entered into TVA's corrective action program for evaluation and resolution.

"The NRC senior resident inspector has been notified of this report.

"Reference corrective action document PER 109118."

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General Information or Other Event Number: 42880
Rep Org: SOUTHERN TESTING SERVICES
Licensee: OTEK CORPORATION
Region: 1
City: KNOXVILLE State: TN
County:
License #:
Agreement: Y
Docket:
NRC Notified By: WILLIAM R. WILLIS
HQ OPS Officer: PETE SNYDER
Notification Date: 10/05/2006
Notification Time: 19:20 [ET]
Event Date: 10/05/2006
Event Time: [EDT]
Last Update Date: 10/05/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
Person (Organization):
PAUL KROHN (R1)
THOMAS DECKER (R2)
VERN HODGE (Email) ()

Event Text

POTENTIAL DEFECT IN OTEK PANEL METERS

Southern Testing Services (STS) provided information on a potential defect in OTEK Panel Meters. The information applies to 90-265 Volt AC / 10- 32 Volt DC, OTEK Part Numbers: HI-Q114, HI-Q116, HI-Q117, HI-Q118, and HI-Q119.

"STS has identified a potential failure cause for the HI-Q class meters that would result in a frozen indicator with no indication that the failure has occurred. The specific failure mode is a failure of the main processor (only) that freezes the display processor. The frozen display inhibits the detection of the main processor failure except by cycling power to the Panel Meter. This failure mode was previously addressed in the Failure Mode Effects Analysis (FMEA) conducted by STS in Test Report S4000-RP-03; however, no credible causes for the failure beyond infant mortality were postulated or had been experienced at the time. Infant mortality failures are minimized by the dedication process that verifies operation of the units prior to shipment to the client.

"STS has just learned from OTEK that of one (1) OTEK Panel Meter had experienced a main processor failure which resulted in a frozen display indicator. This Panel Meter was reportedly in a non-nuclear application. OTEK determined that the likely cause of this failure was a high frequency spike on the power lines caused by running the meter off of a DC generator.

"The STS EMI/RFI qualification program for the OTEK meters qualified them with an anomaly, limiting the surge protection to 500 [Volts peak to peak] for DC powered units. Additionally, high frequency susceptibility testing was successfully completed with a continuous signal of 3.5 [Volts RMS] on the power lines (both AC and DC powered units) from 10 kHz to 200 MHz. Additional testing was done on AC powered units at 7 [Volts RMS]. Testing was conducted in accordance with EPRI TR-102323 to the methods specified in EN 61000-46. The units passed the susceptibility testing at the limits specified, as reported in Test Report S4000-RP-03.

"Based on the successful surge and high frequency susceptibility testing conducted by STS, it is concluded that the single known failure noted by OTEK was related to operating outside of the parameters tested during qualification of these OTEK meters. The likelihood of such a failure in a nuclear safety related application is considered to be remote."

STS determined that 294 units were provided to nuclear plants in potentially safety related applications where EMI/RFI requirements were imposed. The following plants were effected purchasers: Vermont Yankee, Pilgrim, St. Lucie, and Brown's Ferry.

"OTEK is currently in the process of updating the display board processor programming to detect a failure of the main processor, and provide an indication on the display that a main processor failure has occurred. STS will coordinate with the above utilities when this update becomes available."

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Friday, March 30, 2012