IR 05000341/1998015
| ML20207E849 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 02/22/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20207E650 | List: |
| References | |
| 50-341-98-15, NUDOCS 9903110093 | |
| Download: ML20207E849 (6) | |
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EXECUTIVE SUMMARY Enrico Fermi, Unit 2 Fermi inspection Report 50-341/98015(DRP)
This inspection included aspects of licensee operations, maintenance, engineering and plant support. The report covers a 7-week period of resident inspection.
Operations e
The licensee conservatively entered into an Emergency Plan Alert Action Level when smoke was observed coming from an emergency diesel generator control panel.
Station management and personnel responded to the event as expected. Damage was limited to a single component in the emergency diesel generator control panel. The inspectors concurred with the licensee's conclusion that the cause of the smoke was an isolated incident of equipment failure. (Section 01.1)
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A 480 Volt engineered safety feature bus was inadvertently de-energized when the e
alternate feed breaker was opened without power available to the normal feed. The operator assigned to the task did not ensure that Bus 72F was energized as required by a prerequisite in the operations procedure due to personnel error. All appropriate actuations and isolations occurred as expected. This was a non-cited violation.
(Section 04.1)
Maintenance l
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The inspectors noted that the station Error Free Days Program clock was reset 11 times during the o/ age due to personnel errors. The inspectors observed all or portions of
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18 maintentnce/ surveillance test activities with no significant problems noted. The activities obse rved were performed in accordance with the applicable procedures, which i
provided the requisite information necessary to perform the work. (Section M1.1)
e The licensee allowed the primary containment radiation monitoring system (PCRMS) to become inoperable because a required surveillance test was not performed. Licensee managemer't failed to ensure the surveillance test was scheduled prior to the entry into Mode 4 when it was required to be operable by the Offsite Dose Calculation Manual.
The fact that the PCRMS was inoperable went unnoticed and samples of the containment atmosphere were not performed and analyzed as required by the Offsite l
Dose Calculation Manual with the PCRMS inoperable prior to containment venting. The inspectors verified that because of other operable radiation monitoring equipment, the venting operation did not result in an unmonitored release path. The surveillance test was sLbsequently performed and passed. This was a non-cited violation.
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(Section M1.2)
e A personnel error resulted in the use of a prohibited pipe thread sealant on main steam isolation valve pneumatic manifold solenoid valves. The work was re-performed and the
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solenoid valves exposed to the pipe thread sealant were determined not to have been affected by the short term exposure. The inspectors concluded the licensee's corrective actions were acceptable. (Section M1.3)
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PDR ADOCK 05000341 O
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The licensee's maintenance procedure did not provide adequate torque values or
torquing instructions to ensure the high pressure control valves were properly assembled during the previous refueling outage. Although nonsafety-related, the looseness of the internals in the Numbers 3 and 4 High Pressure Control Valves
resulted in significant reactor power fluctuations. The licensee's outage inspection efforts were thorough and comprehensive, resulting in appropriate repairs of the degraded valves. No power fluctuations were observed during unit power ascension.
(Section M3.1)
Enaineerina i
The licensee's engineering efforts to find and eliminate problems with the Reactor
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Recirculation System B Loop Speed Control Unit were ineffective. The Reactor
Recirculation System B Scoop Tube was in a locked condition for several months prior to the refueling outage. The unit was returned to service with the scoop tube operating normally. However, the scoop tube had to be returned to the locked position shortly after plant startup. (Section E2.1)
Plant Sucoort
Scheduling of a Zebra Mussel biocide treatment (Clamtro) of the general service water system during less than optimal water temperature conditions lowered the mortality rate of the Zebra Mussels. The reduced mortality rate of the treatment may require more frequent cleaning of the reactor building closed cooling water heat exchangers.
(Section R1.1)
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. 01.1 Alert Declared Due to Electrical Overheatina of Emeroency Diesel Generator (EDG) 12 Control Panel Circuitry -
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Insoection Scope (71707)
The inspectors reviewed Emergency Action Level classifications,50.72 notifications, and performed a walkdown of the EDG control panels.
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Observations and Findinas On October 8, at 2:30 a.m., during Refueling Outage 6, an Alert was declared by the nuclear shift supervisor (NSS). l The Alert was based on Emergency Plan Classification HA-2, " Fire or Explosion Affecting Operability of Plant Safety Systems
- Required to Establish or Maintain Safe Shutdown." The inspectors concurred with the
' licensee's emergency action classification. During a 24-hour run surveillance test on EDG 12, an operator observed smoke coming from behind the EDG control panel. The j
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operator informed the control room of his observation and tripped the EDG, opened the i
output breaker, and exited the residual heat removal (RHR) building. Simultaneously, a fire alarm was received in the main control room. The fire brigade responded as expected. As a precaution, the control panel was de-energized.
The inspectors noted that damage waa isolated to the EDG control panel and that a i
single component (later determined to be a current buffering transformer or linear reactor) appeared to be slightly damaged. There was no evidence of fire or smoke damage to other components. The inspectors verified that the RHR shutdown cooling function was not affected nor were other safety systems located in the EDG control room.
The licensee's failure analysis of the linear reactor revealed the iron core reactor over heated due to internal turn to turn shorts in the coil layers close to the core. Internal arcing, which welded the turns together, was the most likely cause for the shorts. The internal arcing was caused either by insulation failure between the 17th layer and the
. core, or by arcing between the ends of a broken turn. The linear reactor was replaced
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and tested successfully. Inspections of other components revealed no damage.
The licensee inspected the other EDG control panels and found no similar problems.
Additional research indicated that this failure of the linear reactor was an isolated event.
The EDG control panel was repaired and the EDG was retumed to service on October 10,1998.
c.
Conclusions The licensee conservatively entered into an Emergency Plan Alert Action Level when smoke was observed coming from an EDG control panel. Station management and personnel responded to the event as expected. Damage was limited to a single component in the EDG control panel. The inspectors concurred with the licensee's conclusion that the cause of the smoke was an isolated incident of equipment failure.
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M3.1 BAactor Power Fluctuations inspection Scooe137551)
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The inspectors reviewed CARDS 98-14280, 98-16075, 98-14692, 98-16353, and i
98-15947. The inspectors also reviewed Maintenance Procedure 35.109.002,
Temporary Modification 980016, and General Electric Transient Analysis Recording
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traces. The inspectors observed main turbine control valve and stop valve
maintenance, and conducted interviews with mechanical, maintenance, and engineering i
personnel.'
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Observations and Findinas t
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On September 4,1998, during the power reduction in preparation for RFO6, the plant a
experienced reactor power fluctuations approaching 10 percent peak to peak. Similar -
power. fluctuations had occurred on July 19,1998, during the conduct of a control rod j
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sequence exchange. In the July 19,1998, power fluctuation, the root cause was
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attributed to the identification of looseness in the No. 4 Turbine Control Valve actuator
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linkage.. For the September 4,1998 power fluctuation event, loose linkage was also
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identified in the actuator associated with the No. 3 Turbine Control Valve. In response i
to both the July and September power fluctuations, control room operators placed the l
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reactor mode cwitch in " Shutdown" and manually scrammed the plant. This was discussed in inspection Report 50-341/98013.
During the RFO6 outage, the licensee conducted inspections of the No. 3 and
j No?4 High Pressure Control Valves (HPCVs). The valves were stroke tested and the I
resulting readings compared to previous results. The data showed that the effective
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stroke lengths for HPCV Nos. 3 and 4 had increased, indicating looseness in the tongue j
and spindle connections.
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Given the looseness of the HPCV tongue and spindle connections, the licensee evaluated the torque that had been applied. According to station procedure, a torque value of 150 ft-lbs should have been adequate. The licensee's review of this torque -
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specification with the vender determined that the actual torque should have been
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1000-1200 ft-lbs. When the valves were reassembled during RFO5, the torque applied
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was approximately 150 ft-lbs. This work was performed by both contractors and licensee pers_onnel. When the valves were reassembled during RF06, the as left torque
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(The licensee also revie'Ned the method of applying torque and determined it to be
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inadequate. Specifically, the procedure did not contain adequate instructions to guide craft personnel to ensure the proper application of torque. During power ascension, the licensee maneuvered the plant several times through the power transition point where
ithe fluctuations were typically observed. No power fluctuations'were evident.
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C'onclusions
' The licensee's maintenance procedure did not provide adequate torque values or
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torquing instructions to ensure the HPCV valves were properly assembled during the previous refueling outage. Although nonsafety-related, the looseness in the Nos. 3 and 4 HPCVs resulted in significant reactor power fluctuations. The licensee's outage inspection efforts were thorough and comprehensive, resulting in appropriate repairs of the degraded valves. No power fluctuations were observed during unit power
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ascension.
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E2.1 Reactor Recirculation System (RRS) B Looo Control
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Insoection Secoe (37ff2.L
.The inspectors reviewed Work Request 00Z982413 and conducted interviews with engineering and work control management personnel, b.
Observations and Findinas
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Tha licensee operated with the "B" Reactor Rec!rculation Scoop Tube in the locked position for several months. The licensee placed the scoop tube in the locked position
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due to unpredictable changes in the output of the reactor recirculation pump speed controller, The licensee was unsuccessfulin finding and eliminating the problem during i
the outage. Although the unit was retumed to service with the scoop tube operating
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normally, the scoop tube had to be retumed to the locked position shortly after startup.
At the end of the inspection period, a ground was identified that could potentially affect speed control circuitry performance.
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Conclusions i
The licensee's engineering efforts to find and eliminate problems with the RRS B Loop Speed Control Unit were ineffective. The RRS B Scoop Tube was in a locked condition for several months prior to the t ofueling outage. The unit was retumed to service with the scoop tube' operating normally, However, the scoop tube had to be retumed to the locked position shortly after plant startup.-
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