IR 05000395/1992023
| ML20128A551 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 01/21/1993 |
| From: | Cantrell F, Haag R, Keller L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20128A542 | List: |
| References | |
| 50-395-92-23, NUDOCS 9302020245 | |
| Download: ML20128A551 (14) | |
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UNITED STATES Y
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NUCLEAR REGULATORY cOMMISslOM fk
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101 MARIETTA STRE ET,N,W.-
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ATLANTA.oEORGtA 30323
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Report No.:
50-395/92-23_
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Licensee:
South Carolina Electric & Gas Company Columbia, SC 29218
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Docket No.:
50 395 License No.:
HPF-12 facility Name:
Virgil C. Summer Nuclear Station u
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Inspection Conducted:
December 1, 1992 through January 8, 1993 Inspectors:
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R. C. Haag, Senior R6sident inspector Dat'e Signed-6*' 42 M LQJ1-ihi/n L. A. GTler, Residegj. Inspector Date Sfgned Approved by:
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//Z//93 Floyd S. CantreT1, Chier/
Dite Sitned Reactor Projects Secti6n IB Division of Reactor Projects SUMMARY Scope:
This routine inspection was conducted by the resident inspectors onsite in the areas of monthly surveillance observations, monthly maintenance observations, operational safety verification, ESF system walkdown, installation and. testing.
of modifications.
Selected tours were conducted on backshift or weekends.
These tours were conducted on six occasions.
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__Resu ts:
No violations or deviations were identified.
A concern was identified involving the potential for overpressurization of the reactor building air lock from a single valve failure with the. condition being undetected (paragraph 4). The use of eyebolts in lieu of bolts as load bearing parts in Limitorque valve actuators was identified (paragraph'5.b).
Two separate failures of reactor trip. bypass breakers to close occurred..No-
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effect on the. opening ability of the-breakers was noted (paragraph 5.c)..
- Review of the licensee's basis for deviating from the vendor's recommended 10
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year replacement interval.for Agastat relays is_ an IFI (paragraph 5.d).
The adequacy of the low speed testing-for the turbine driven emergency feedwater pump was identified as an IFI (paragraph 5.f).
9302020245 930121 PDR -AJOCK 05000395 G
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REPORT DETAILS.
1.
Persons Contacted
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Licensee Employees
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W. Baehr, Manager, Chemistry and Health Physics
- C. Bowman, Manager, Maintenance _ Services-
- M. Browne, Manager, Design Engineering
- B. Christiansen, Manager, Technical Services
- M. Fowlkes, Manager, Nuclear-Licensing & Operating Experience-S. Furstenberg, Associate Manager, Operations-W. Higgins, Supervisor, Regulatory Compliance
- S. Hunt, Manager, Quality. Systems
- A. Koon, Nuclear Operations Project Coordinator
- D. Lavigne, Generai Manager, Nuclear Safety K. Nettles, General Manager, Station Support H.-0'Quinn, Manager, Nuclear Protection Services
- M. Quinton, General Manager, Engineering Services
- J. Skolds, Vice President, Nuclear Operations
' *G. Taylor, General Manager, Nuclear Plant Operations
- R. Waselus, Manager, Systems Engineering
- R. White, Nuclear Coordinator, South Carolina Public Service Authority
- B. Wi'lliams, Manager, Operations
'Other licensee employees contacted included engineers, technicians, operators, mechanics, security force members, and office personnel.
- f.ttended exit interview Acronyms and initialisms used throughout this report are listed in the last paragraph.
2.
Plant Status The p_lant operated at or near 100 percent. power throughout the inspection period.
Other. inspections or meetings:
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t During the week'of-December 1, 1992,-a regional' inspection in the area of health physics was performed.
(NRC Inspection Report No. 395/92-22).
3.
Monthly Surveillance Observation (61726)
The_ inspectors observed surveillance activities of safety-related
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systems and components listed below to ascertain that these activities-L were conducted in accordance with license requirements. The inspectors-l verified that required administrative approvals were obtained prior to
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initiating the test,. testing was accomplished bv qualified personnel in-accordance with an approved test procedure,-tes, instrumentation was calibrated, and limiting conditions for operation were met.
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completion of the test, the inspectors verified that test.results conformed with technical specifications and procedure _ requirements', any -
deficiencies identified during the testing were properiy-reviewed and
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resolved and the systems were properly returned to service.
Specifically, the inspectors witnessed / reviewed portions of the following test activities:
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Quarterly surveillance test for the "A" train battery (STP 501.002).
- The inspector verified that all individual-cell voltages-and specific gravities were within the acceptance criteria.
The inspector noted that there was a fine layer of sediment at the bottom of all the cells.
The licensee, after consulting with the battery vendor, stated that the small amount of grey sediment was the result of normal chemical reactions between the battery plates and the electrolyte, and therefore was not indicative of'any problems.
The licensee subsequently performed a visual inspection of the battery cells.
They noted that cell number 13, on the "A" train battery, had some light pink, discoloration on-the supports within the cell which might indicate minor copper. contamination.
The inspectors verified that the cell readings for voltage and specific gravity have been within specifications with no adverse trends. The licensee has included cell 13 in their daily battery readings, along with the pilot-cells, in_ order to trend' the cell performance. Additionally, a vendor representative is scheduled to inspect the cell.
If copper contamination is-confirmed, the licensee plans to replace the cell during the upcoming outage (RF7).
The inspectors will followup on the licensee's monitoring of cell performance and the long term corrective-action.
Reactor building escape airlock test (STP 215.001B).
Test was
performed as a result of maintenance performed within the airlock.-
Test results were satisfactory.
Moderator temperature coefficient determination (STP 210.001). lhe
purpose of this surveillance was to perform the "at pnwer"'
measurement of moderator temperature coefficient (MTC) in accordhnte with TS 4.1,1.3.b.
The acceptance criteria required for the MTC ranges from 0 to -41 pcm per degree fahrenheit. -The measured MTC was -27.8 pcm per degree fahrenheit. This surveillance was well planned and executed.
Primary coolant sample for boron concentration (CP 903). The sample
was taken in accordance with procedures. -No discrep'ancies were noted.
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3 All tests observed were performed in accordance with the applicable procedures and demonstrated ~ acceptable results,
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4.
Monthly Maintenance Observation (62703)
Station maintenance _activitie's for the safety-related systems and components listed below were observed to ascertain that they were conducted in accordance with approved procedures, regulatory guides,. and industry codes or standards and in conformance with TS.
The following items were considered during this review:
that limiting conditions for operation were met while components or systems were removed from service, approvals were obtained prior to initiating the work, activities were accomplished ssing approved procedures and were-inspected as applicable, functional testing and/or calibrations were performed prior to returning components or systems to service, activ; ties were accomplished by qualified personnel, parts and materials used were -
properly certified, and radiological and fire prevention controls were-implemented. Work requests were reviewed to determine the status of outstanding jobs and to ensure that priority was assigned to safety-related equipment-maintenance that may affect system performance. The-following maintenance activities were absorved:-
Preventative maintenance-on the "C" chargingfpump-(EMP 295.004). The-
inspector witnessed various portions of the preventive maintenance,-
incit. ding motcr lube oil sampling ar.d bridging /megg'ering of the charging pump motor.
Corrective actions taken to reduce "C" charging pump vibrations (MWR-
92N3029).
During November 30 through December 4, 1992, the-
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inspectors witnessed corrective maintenance on the "C" charging pump which was accomplished using MMP 320.012, Charging / Safety Injection Pump Overhaul, Revision 8.
Portions of the procedure were used in an attempt 4 to determine the cause of the increased pump vibration.-
The inspectors witnessed the clearance check of the pum) bearings._
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The clearances were within the allowable tolerances.
Tae inspectors also witnessed an alignment check of the pump and speed reducer
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using aidial indicator.
The-licensee tested the charging pump and the resulting vibration measurements were reduced to levels below the alert range. However, the licensee does not believe-the root cause of the vibration was corrected and they are continuing their investigation of this problem.
Troubleshooting on -the "B waste gas ' compressor (MWR 9204309). The
compressor had been exhibiting a drop in pressure at the-inlet to~
.the recombiner.
A vendor. representative was present-and participated in the overhaul. The inspecters observed' portions'of the rotor adjustment. The only noticeable item was daterioration of
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the elastomer in a water seal return valve. The valve was replaced
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and the compressor was reassembled. After running the compressor, discharge pressure dropped as previously seen. At the_end of the inspection period, the licensee was continuing with the troubleshooting effort on the "B" waste gas compressor.
o Troubleshooting the'cause of reactor building escape airlock
pressurization (MWR 92M3157).
The licensee generated this MWR on July 17, 1992, in response to indications that the airlock was slowly becoming pressurized.
A similar condition occurred in February, *.991, due to leakage past the emergency air supply valve, which is located inside the airlock, in October, 1991, this valve was replaced which corrected the problem until July,1992.
On September 11, 1992, the pressure inside the airlock reached 10 psig, prompting the licensee to make an airlock entry, it was determined that the inleakage was again from the emergency air supply valve.
Service air, at a pressure of 90 psi, is normally supplied to the emergency air supply valve to provide an_ emergency-sourco of breathing air in the airlock. Due to the leakage past this valve the upstream isolation valves, which are normally_ locked open, were closed and caution tagged until the leak could be repaired. The valve was replaced on-December 10, 1992.
Subsequent monitor Wg indicating that the leakage was corrected.
In reviewing the maintenance history and design of the airlock, the inspector was concerned that the failure of a single commercial grade valve could result in pressurizing the airlock to 90 psi.
This could challenge containment integrity, ~since the airlock is normally tested at a pressure not less than 47.1 psig. Of further concern was the fact that there were no alarm indications or periodic log readings associated with airlock pressure.
This could have resulted in an overpressurization of the airlock going undetected for a prolonged period of i.ime.
The inspector also noted that the same design existed on the personnel hatch for the reactor building.
In response to these concerns the licensee has included _
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the escape airlock and-personnel-airlock pressures in the daily -
operator logs.
The licensee is also reviewing the design configuration of emergency air supply for the airlock to determine if improvements are warranted.
The daily pressure reading will be
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continued unless the current design is modified.
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Troubleshooting an annunciator associated with the reactor-vessel
level indicating system (RVLIS) (MWR 9204402). The~ annunciator which was in-the alarm condition was " Reactor Vessel Level flydraulic
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l Isolator Trouble".
This alarm was associated with the "B" train l
hydraulic isolator, ILS 1321X.
The RVLIS uses a hydraulic isolator-l-
to separate the system-fluid in the reactor. building from the fluid-that actuates the differential pressure transmitter. The alarm indicated that the isolator bellows had moved. The actual L
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indicating needle on the isolator read approximately 0.37 cubic inches of displacement.
Engineering evaluateo the conditioa as acceptable piovided the displacement does not exceed 0.5 cubic inches.
The inspector reviewed the technical manuals for the hydraulic isolators and the RVLIS, and discussed the condition with engineering.
Based on these reviews and discussions, the inspector agreed that the limit of 0.5
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cubic inches would ensure the operability of RVLIS. A review of RVLIS history identified that isolator ILS 1321X has had larger displacements than the other channels. During the upcoming outege the licensee plans to calibrate and fill the system such that future
"at power" displacement readings will be zero.
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Inspecting and cleaning the 7.2 kV breaker for service water pump
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"A" (PMTS P0162591).
Maintena. ice activities observed were satisfactory.
The inspector noted what appeared to be commercial grade Agastat relays inside the breaker cubicles for the "A" and "C" service water pumps.
See paragraph 5d for the discussion of Agastat relays in safety-related applicctions.
Replacement of the test level potentiometer for power range nuclear
instrumentation NI-44, detector B (MWR 9213339).
While performing a surveillance test on NI-44, no output was observed from the test potentiometer.
The NI drawer was removed from the cabinet and taken to the shop for replacement of the test potentiometer.
The fine o
gain potentiometer was also replaced.
The reliability of the gain potentiometer was previously questioned when N1-44 output spiked high while using the potentiometer to adjust NI-44 output. However, g
the problem could not be repeated during subsequent troubleshooting, therefore, the licensee had planned to replace the gain potentiometer.
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Repair of intermediate building fire service deluge valve XVM 6935 e
(MWR 92G0040). When attempting to re:et the weighted block, which provides the mechanical force to trip the valve open, the Hock would not reset. The valve was disassembled and no unusual conditions were noted.
Af ter the valve was reassembled, the block was reset satisfactorily. The inspector was informed that resetting the weighted block on deluge valve is a sensitive task and the licensee has experienced previous difficulties in resetting the
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blocks.
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All the maintenance activities observed were performed per the applicable procedures / instructions. While the final resolution to several plant g
problems, i.e.,
"C" charging pump vibration,
"B" waste gas compressor discharge pressure and RVLIS annunciator, was not fully completed, the licensee has taken appropriate actions to resolve these problems.
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5.
Operational Safety Verification (71707)
a.
Plant Tour and Observations The inspectors conducted daily inspections-in the following areas:-
l control room staffing, access, and operator behavior; operator adherence to approved procedures, TS, and limiting conditions for
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operations; and review of control room operator logs, operating orders, plant deviation reports, tagout logs, and tags on components to, verify compliance with approved procedures.
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The inspectors conducted inspections for the operability verification of selected ESF systems by valve alignment, breaker -
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positions, condition of equipment or component (s), and operability of instrumentation and support items essential to system actuation or performance. The emergency feedwater system and the high head safety -injection system were included in these inspections.
Plant tours included obscrvation of general plant / equipment
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conditions, fire protection and preventative measures, controi-of activities in progress, radiation protection controls, physical security contro(s, plant housekeeping conditions / cleanliness, and missile hazards. Reactor coolant system leakirates were reviewed to ensure that detected or suspected leakage from the system was recorded, investigated, and evaluated; and that_ appropriate actions were taken if required.
Selected tours were conducted on backshifts or weekends.
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On January 5,1993, during a tour of the "B" chill water chiller room the inspector noted a small service water (SW) leak at the
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chiller SW return header relief valve XVR 3144B.
The leak appeared to be located at the socket weld for the pipe stub
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attached to the valve. The amount of leakage was approximately 1-l to 2 drops per ininute.
The inspector informed the licensee ~of the
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leak and NCN 4571 was subsequently written.to address this problem.
The licensee plans to cut the-socket weld to_ allow
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replacement of the pipe stub. This. work is currently. planned.for.
l-the week of January 25-29, 1993. The. inspector reviewed the engineering evaluation which accepted this condition until the.
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repair is completed. The evaluation provides adequate.
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justification to accept this condition on a temporary basis.
In addition, operations is monitoring the leak every shift to detect any increase in leakage.
L b.
Questionable Housing Cover Bolts on Certain Motor Operated j
Valve (MOV) Actuators During plant tours the inspectors noted several safety-related MOV-Limitorque actuators with eyebolts located in the taped holes.for
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the housing cover bolts, for the actuators observed, two of the four taped holes had eyebolts on SMB 000 actuators, and two of the six for SMB 00 actuators.
The eyebolts had apparently been installed for use as lif ting lugs to facilitate installation of the actuators.
The inspectors were concerned with the suitability of these eyebolts for application as actuator housing cover fasteners.
Subsequent conversations between the inspectors and the vendor (Limitorque) revealed that housing cover bolts are considered critical components for use on SMB 000 and SMB 00 a>plications.
The housing cover and bolts receive the resulting tarust load during the closing stroke of the vahe. The normal housing cover bolts are specified by Limitorque as SAE Grade 5 material.
The material grade of the eyebolts could not be determined by the vendor.
For the MOV actuator that was provided, as an example, by the inspectors (XVG 06517 VU), the vendor verified that eyebolts were included on the bill of material.
However, there was no documentation regarding whether the eyebolts were meant to reinain installed af ter initial MOV installation in the plant.
As a result of the concerns raised after those conversations with the vendor, the inspectors requested that the licensee address the use of eyebolts as actuator housing cover fasteners on MOV's considered important to safety.
In response, the licensee provided an engineering evaluation which assumes that only the " qualified" bolts accept the entire load developed from the maximum thrust rating of the actuator.
The licensee inspected the accessible M0V's to verify the assumptions in the evaluation, i.e., maximum number of eyebolts installed and that the eyebolts were diametrically located in the bolting pattern.
The resulting calculated stress to the " qualified" bolts under these conditions was less than the fastener proof strength (fastener proof strength is the maximum load a bolt can withstand without being permanently deformed). Additionally, it was noted in the evaluation that there had been no instances of fastener failures during operation or MOV testing.
Therefore engineering concluded that for the conditions observed in the field, there was not an immediate operability concern. The inspectors reviewed this engineering evaluation and agreed with the calculational mathods and conclusions.
Limitorque subsequently published Maintenance Update 92-02, which stated that it was Limitorque's opinion that the eyebolts would not require immediate replacement. However, they recommended that any eyebolts used as housing cover bolts for SMB 000 and SMB 00 actuators be replaced with qualified Grade 5 bolts at the next scheduled maintenance. The licensee indicated that the subject eyebolts will be replaced with Grade 5 bolts as part of their
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routine scheduled maintenance, Since the maintenance schedule for i
some MOVs will not be completed for several years, the inspector
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questioned if there existed any inaccessible valves that would not
have maintenance completed during the upcoming outage.
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response, the licensee stated that they would inspect during the i
outage any previously inaccessible. valve that was not scheduled
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for outage maintenance to ensure the engineering evaluation i
assumptions were valid for all applicable MOVs.
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Reactor Trip Breaker Deficiencies When performing solid state protection system (SSPh) testing the
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reactor trip bypass breaker is placed in service to allow testing of the reactor trip breaker. On December 4, 1992, while
attempting to test the "B" train bypass breaker (R14), prior to placing it in service, the breaker failed to remain in the closed position.
The breaker would cycle through the closed position and immediately reopen. A spare breaker was placed in the "B" bypass position and the SSpS testing was completed satisfactorily.
Subsequent inspection of breaker RT4 by-the licensee revealed that the trip latch had become detached from the trip latch pivot pin.
The inspector observed the detached trip latch on RT4 and com9ared theconditiontoareactortriptrainingbreakerlocatedintbe shop, in order to assess the impact of this condition on breaker operation.
The training breaker was cycled through the open and closed positions to allow observation of breaker operation.
The inspector also reviewed the technical manual (TM) for the reactor trip breakers (IMS-940-1066).
The following sentence in Section-2.7.2 of the TH describes the effect that the trip latch has on closing operation, "Upon closing, the trip latch comes in contact
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with and is kept in place by the overlap of the latch surface on the breaker trip shaft, latching the breaker in the closed
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position".
Based on the review of the TM and observation of the breakers, the inspector concluded that the detached trip latch
only affected the ability of the breaker to close and did not have
the potential to impact the breaker opening function. The breaker was shipped to Westinghouse for repair and evaluation of the failure.
On December 18, 1992, while performing SSpS testing on "A" train, a problem occurred with the "A' train bypass breaker (R15)._ Prior to placing R15 in service, it was closed and tested using the shunt trip device. The breaker tripped open as' expected from the
shunt trip..The charging s) ring motor attempted to charge the
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closing springs, however,-tic springs discharged as soon as-they became charged.
Several cycles of the closing spring being charged then immediately discharging occurred.
This sequence of tripping the breaker open via the shunt trip device with the
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closing spring charging and discharging was repeated when electricians were called to investigate the problem.
The reactor trip breakers are designed such that the closing springs are automatically charged when the breaker is opened.
The licenseo verified from the cycle counter that the breaker was not cycling
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through the open and close positions when the clocing springs were being charged and discharged repeatedly.
The breaker was taken to i
the electric anop for troubleshooting.
The SSPS testing was completed by moving the breaker from the "B" train bypass cubicle into the "A" train bypass cubicle.
Visual inspection of the breaker revealed a small amount of grease on the spring charging motor cut off limit switch, which may have caused the problem observed in the-plant.
The switch was replaced and the breaker subsequently tested satisfactorily in the shop.
However, after further review the licensee questioned if_ a limit switch failure could cause the cycling problems with the charging springs.
Due to the unknown cause of the cycling problems, tin!
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licensee intends on shi) ping R15 to Westinghouse for a complete
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evaluation and refurbis ament upon return of RT4 from Westinghouse.
The next scheduled SSPS testing will be in January, 1993, for "B" train.
The inspectors will closely monitor reactor trip breaker
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and reactor trip bypass breaker performance during upcoming tests, and will also review the Westinghouse evaluations for the two breakers.
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d.
Agastat Relays in Class IE Switchgear Exceed Manufacturers
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Qualified Life While ob:erving maintenance on the breaker for the "A" service water pump the inspector noted an Agastat 7000 series relay inside the breaker cubical.
The model and serial number for the relay indicated that it was manufactured prior to 1977, as:a commercial grade relay. Agastat 7000 series relays which meet Class IE requirements have the letter E preceding the model number (i.e.
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E7012). They have a qualified life of.10 years unless they have-undergone additional qualifications.
The original relay identified by the inspector functioned in an alarm circuit only and therefore did not perform a critical function, llowever, subsequent inspections revealed similar relays, in terms of model series and ago, that were located in Class lE switchgear throughout the plant.
There are approximately 100 safety related Agastat relays in the plant.
The inspector was only able to inspect 63 of these relays due to limiting plant conditions.
Of the. 63 relays inspected, there were 12 relays-that were non "E" series,_that appeared to be-in potentially critical applications. All 12 of-these relays were well past-their qualified life based on date of manufacture.
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These questionable relays were in the switchgear for the "A".E.00,
When the licensee set up their program for Agastat relay replacement in June, 1983, their EQ files were used as the basis for which relays were to be included in the replacement program.
The inspector questioned why the 12 relays mentioned above were not included in the replacement program.
Af ter reviewing the application of these relays the licensee stated that the relays were in alarm circuits associated with the
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components and therefore they did not perform critical functions.
During this review the inspector questioned the basis nf the licensee's PM program which _ allowed deviating from the vendor's recommended 10 year replacement interval for "E" series relays.
As stated in the EQ program, the actual replacement schedule will be developed from surveillances and trending of componcnt failures by engineering.
For approximately 19 "E" series relays which are greater than 10 years old the licensee has deferred replacement of the relays based on the above program.
Based on the initial reviews, the inspectors questioned if the amount of surveillance testing and trending data collection was adequate to justify deferring the relay replacements.
The licensee is continuing to review the Agastat relay replacement issue.
Additional inspector followup of the current relay replacement PM progam and any_ future program changes is identified as Ifl 50 395/92-23 01.
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Control Room Evacuation Panel (CREP) and Technical Sup) ort Center (TSC) Indication for Steam Generator "A" Level Failed ow During a routine plant tour on December 28, 1992, the inspector I
noted that the level indication for steam generator "A" on the CREP had failed low.
Subsecuent troubleshooting by the licensee revealed that the associatec level transmitter (ILT 00477A) had failed.
This also affected indications for the TSC.
Loss of this level indication on the CREP panel placed the plant in the LC0 for remote shutdown indication (TS 3.3.3.5), which has a 7. day action statement. A reactor building entry was made the following day to-replace the failed transmitter, and _the LCO was exited shortly:
thereafter.
The ins)ector noted that-the required channel check for this instrument lad been satisfactorily performed on December 6, 1992. Actions taken by the licensee to correct this nonconformance were prompt and effective.
f.
Turbine Driven Emergency feedwater Pump (TDEFP) Maintenance and Testing Activities The inspector reviewed the scheduled maintenance for the TDEFP during the u) coming _ refueling outage.
Specific areas reviewed dealt with tie turbine low lube oil (LO) trip feature. 'The TDEFP is protected from low speed operation by the corresponding low LO
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trip feature. The trip setpoint is approximately five psig LO piessure which corresponds to a turbine / pump speed of approximately 1,000 RPMs.
The licenseo stated the EfW pump can be damaged during prolonged operation at low saced (approximately 700 RPMs).. Steps 7.5.5, 7.5.6 and 7.5.7 of Meclanical Maintenance-Procedure, MMP 300.015, Turbine Maintenance, Emergency feedwater Pump, cleans, ins)ects and repairs, as necessary, the oil cylinder which initiates tie low LO trip.
Step 7.5.8 states to perform a i
maintenance run per step 7.11 as necessary.
While reviewing step 7.11, the inspector noted that the lowest speed obtained during the maintenance run/ testing is 2100 to 2300 RPMs.
Information supplied to the inspectors by the licensen indicated that minimum S/G flow can be provided at TDEfP speeds of 1415 RPM.
The licensee was not aware of any previous tests that-had been performed to verify the correct setpoint for the L0 trip -
feature. The low LO trir. had previously occurred due to a clogged L0 strainer and a premature lifting relief valve, but the actual trip setpoint had not been captured.
Based on the upcoming maintenance which could affect the actual tripping mechanism and the planned testing which would not test the IDEfP at the lowest s3eed at which it could be operated, the inspector was concerned t1at the low LO trip setpoint was not being verified.
in response, the licensee informed the inspector that additional low speed testing requirements would be added to MMP 300.015, This additional testing would be performed at the same five year interval that the TDEfP overhaul would be performed. Review of the procedure change which will test / verify the TDEfP low L0 trip setpoint is identified as Ifl 50 395/92-23 02.
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6.
[SF System Walkdown (71710)
The inspectors verified the operability of an ESF system by performing a walkdown of the accessible portions of the service water (SW) system.
The inspectors confirmed that the licensee's system line up procedures matched plant drawings and the as-built configuration. The inspectors looked for equipment conditions and items that might degrade performance (hangers and supports were operable, housekeeping, etc.).
The inspectors verified that valves, including instrumentation isolation valves, were in proper position, power was available, and valves were locked as appropriate.
The inspectors. compared both local and remote position indications. With the exception of several administrativo deficiencies involving component labeling and poor housekeeping in the west piping penetration room at elevation 436, the
SW system was properly aligned and maintained. The licensee initiated-
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Installation and Testing of Modifications (37828)
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As part of the ongoing evaluation of plant modification activities, MRF
22365 was reviewed, included in the review was the direct observation of portions of the installation, post modification testing and review of the MRF package.
The MRF installed test connections on "A" EDG fuel oil tank vent and drains.
This will allow for connection of test equipment when measuring flow rates for the fuel oil transfer pumps.
The inspector was particularly interested in the controls for cutting and welding since the day tank remained full of fuel oil while work was being performed. The level of management attention and review this work activity received was appropriate.
The ins)ector considered that the precautions taken to ensure that the hot wor ( was not a safety hazard were adequate.
8.
Exit Interview (30703)
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The inspection scope and findings were summarized on January.8, 1993, with those persons indicated in paragraph 1.
The inspectors described the areas inspected and discussed the inspection findings.
No dissenting comments were received from the licensee. The licensee did not identify as proprietary any of the materials provided'to or reviewed by the inspectors during the inspection, 9.
Acronyms and Initialisms CP Chemistry Procedure CREP Control Room Evacuation Panel EDG Emergency Diesel Generator EFW Emergency Feedwater EMP Electrical Maintenance Procedure EQ Equipment Qualification-ESF Engineered Safety feature IFI Inspector Followup Item LCO Limiting Conditions for Operations LER Licensee Event Reports LO Lube Oil MMP Mechanical Maintenance Procedure MOV Motor Operated Valve MRF Modification Request Form MTC Moderator Temperature Coefficient MWR Maintenance Work Request NCN Nonconformance Notice NI Nuclear Instrumentation NRC Nuclear Regulatory Commission NRR Nuclear Reactor Regulation PCM Percent Millitho pH Preventive Maintenance l
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PMIS Preventive Maintenance lask Sheet PSI Pounds Per Square Inch PSIG Pounds Per Square Inch Gaugo
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i RVLIS Reactor Vessel level Indicating System RWP Radiation Work Permits SAE Society of Automotive Engineers S/G Steam Generator SPR Special Reports SSPS Solid State Protection System i
S1P Surveillance Test Procedures SW Service Water-
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TDEfP Turbine Driven Emergency feedwater Pump 1M Technical Manual TS Technical Specifications 1SC Technical Support Center
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