IR 05000344/1989029
| ML20042D297 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 12/19/1989 |
| From: | Mendonca M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20042D296 | List: |
| References | |
| 50-344-89-29, NUDOCS 9001080241 | |
| Download: ML20042D297 (16) | |
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S.' NUCLEAR REGULATORY COMMISSION {
REGION V
. I Report No.: 50-344/89-29 ' . Docket No.: 50-344- - License'No.: NPF-1
' , ' ,, . Licensee:. Portland General Electric Company
121 S.W. Salmon Street , , Portland. 0R.97204 0 -} ; Facility Na,ne: Trojan '
Inspection at: Rainier, Oregon { ' e , Inspection conducted: : October'14, 1989 November 25, 1989 ,
. Inspectors: R. C. Barr' . ! -Senior Resident Inspector j J. F.:Melfi L Resident Inspector Approved By: %W ' O^ ' " /MW M. M. Mendonca, Chief . Date Signed ' .. ReactorProjectsSection1 Summary: . Inspection on October 14 - November 25, 1989 (Report-50-344/89-29) - Areas Inspected: Routine inspection of operational safety verification, maintenance, surveillance, event follow-up, system engineering, and open item follow-up.- Inspection procedures 30703 61726, 62703, 71707, 71710, 86700, 90712,92700,and93702wereusedas.guldanceduringtheconductofthe inspection, Results
One non-cited violation was identified (paragraph 6).
Scheduling and ' - performance weaknesses continue to exist in the-licensee surveillance pro ram.
The-licenseeappearstohave$roblemsobtainingbualifiedpartstoinstal in - the. plant (paragraphs 4 and 6. The licensee di not readily resolve one pressure transmitter failing low in a timely manner (paragraph 10).
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.' .- ,, . .. DETAILS 1.
Persons Contacted
- D. W. Cockfield, Vice President, Nuclear
- C. P. Yundt,' Plant General Manager
- T. D. Walt, General Manager, Technical Functions A. N. Roller, Manager, Nuclear Plant Engineering
- C. K. Seaman, Manager, Nuclear Quality Assurance
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- D. W. Swan, Manager, Technical Services
, M. J. Singh, Manager, Plant-Modifications J. D. Reid, Manager, Quality Support Services
- J. W. Lentsch, Manager, Personnel Protection
- J.F.Whelan,BranciMana3er, Maintenance J. Mody, Branch Manager, 1 ant S
- D. L. Nordstrom, Branch Manager,ystems EngineeringQuality Operations J. P. Fischer, PM/EA Branch Manager R. N. Prewit, Supervisor QualitySystems
- J. Reinhart, Branch Manag,er, Operations J. C. Heitzman, Acting Assistant Operations Supervisor N. A. Regoli Instrument and Control Supervisor J.A.Benjam}n, Supervisor,QualityAudits J. D. Guberski, Nuclear Safet and Regulation Department Engineer
- W. J. Williams, Compliance En ineer The inspectors also interviewed and talked with other licensee employees during the course of the inspection.
These included shift supervisors, reactor and auxiliary operators, maintenance personnel, plant technicians and engineers, and quality assurance personnel.
- Denotes those attending the exit interview.
2.
Plant Status The facility operated at 97% power throughout the reporting period.
From October 18 through October 20, 1989, the steam driven auxiliary feedwater (AFW) was out of service to correct a control malfunction identified during surveillance testing.
After repair and testing, the AFW pump was returned to service on October 27, 1989.
3.
Safety Verification (71707) Operational Safety Verification During this inspection period, the inspectors observed and examined activities to verify the operational safety of the licensee's facility.
The observations and examinations of those activities were conducted on a daily, weekly or biweekly basis.
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' p . L ' .. . ' L . . Daily the inspectors observed control room activities to verify the - licensee's adherence to limiting conditions for operation as prescribed in the facility Technical Specifications.
Logs, instrumentation, ' recorder traces, and other' operational records were examined to obtain information on plant conditions, trends and compliance with regulations.
, OnoccasionswhenashiftturnoverwasInprogress the turnover of - informationonplantstatuswasobservedtodetermlnethatpertinent
, ". . information was relayed to the oncoming shift personnel.
' , Each week the inspectors toured the accessible areas of the facility to
observe the following items: (a) General-plant and equipment conditions.
. (b) Maintenance requests and repairs.
' c) Fire hazards and fire fighting equipment.
l d) Ignition sources and flammable material control, e) Conduct of activities in accordance with the licensee's i i administrative controls and approved procedures.
Interiors of-electrical and control panels.
> Implementation of the licensee's physical security plan.
!- Radiation protection controls.
' i) Plant housekeeping and cleanliness.
(j) Radioactive waste systems.
' (k) Proper storage of compressed gas bottles.
Weekly, the inspectors examined the licensee's equipment clearance t control with respect to removal of equipment from service to determine ! that the licensee complied with technical specification limiting - conditions for operation.
Active clearances were spot-checked to ensure , that their issuance was consistent with plant status and maintenance evolutions.
Logsofjumpers, bypasses',cautionandtesttagswere examined by the inspectors.
' Each week the inspectors conversed with operators in the control room, ' and with other plant personnel.
The discussions centered on pertinent topics relating to general plant conditions, procedures, security,
training and other topics related to in progress work activities.
The inspectors examined the licensee's nonconformance reports (NCRs) to confirm that deficiencies were identified and tracked by the system.
l Identified nonconformances were being tracked and followed to the completion of corrective action.
- Routine inspections of the licensee's physical security program were performed in the areas of access control, organization and staffing, and detection and assessment systems.
The insaectors observed the access L control measures used at the entrance to tie protected area, verified the l integrity of portions of the protected area barrier and vital area l barriers, and observed in several instances the implementation of compensatory measures upon breach of vital area barriers.
Portions of the isolation zone were verified to be free of obstructions.
Functioning of central and secondary alarm stations (including the use of CCTV monitors) was observed.
On a sampling basis, the inspectors verified . .. - -
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that the required minimum number of armed guards and individuals . authorized to direct security activities were on site.
, The inspectors' conducted routine inspections of selected activities of the licensee's: radiological protection program.
A sampling c radiation ~ work permits (RWP) was reviewed for completeness and adequacy of information.
During-the course of inspection ectivities and periodic ' tours of plant areas, the inspectors verified proper use of personnel c monitoring equipment,igning out on appropriate RWP's, and observed observed individuals leaving the radiation ^ controlled area and s posting of radiation areas and contaminated areas.
Posted radiation levels at locations within the fuel and auxiliary buildings were verified ' using both NRC and licensee portable survey meters.
The involvement of u health physics supervisors and engineers and their awareness of ' significant plant activities was assessed through conversations and review of RWP sign-in records.
The inspectors verified the operability of selected engineered safety features.
This was done by direct visual verification of the correct ' " position of valves, availability of power, cooling water supply, system integrity and general condition of equipment, as applicable.
No violations or deviations were identified, y 4.
Maintenance (62703) . Reactor Coolant Pump Seal Leakoff Failure m On November 30, 1989, at 2:44 am, the flaw indication from the Number 1 seal on the B Reactor Coolant Pump (RCP) drop)ed to zero.
Annunciators . came in noting the condition.
The operators aegan monitoring the pum) - per procedures and also began logging various other instruments in tie control room (e,.g., containment pressure, containment humidity, containment sump level, and reactor coolant drain tank level).
The filure of seal flow to the Number 1 seal can affect RCP operation and/or Reactor Coolant System leakage.
The licensee wrote an urgent Maintenance
Request (MR) to investigate this situation.
Instrumentation and Control - personnel were called * at 3:30 m and the Recident inspector was b informed at approximately 4:20 am.
, The resident observed the licensee troubleshooting the flow transmitter
electronics.
A licensee quality inspector also observed the maintenance.
-The licensee's initial investigation thought that the 15 Volt D.C. power supply had failed to that transmitter.
Subsequently, the licensee "_ personnel determined that the failure was not due to a failed power supply but due to a loose fuse clip on a circuit board.
Craftsman discove, red this by inserting the board and bending it back and forth.
The technicians determined that there was one fuse clip that was bent, allowing a poor connection.
The technician bent the fuse clip back and _" inserted the board.
The flow indicator began to read the leakoff flow.
The transmitter was then returned to service.
To make sure that no other transmitters had similar problems, the . licensee wrote another MR to investigate the other fuses.
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. , . < . , Service Water B'ooster Pump (SWBP) Switch Replacement The C Service Water Booster Pump (SWBP) switch in the control room would not go from auto to start easily.
The switch was also noted to be hard to get out of the Pull-To-Lock (PTL) position.- The SWBP is an Engineered Safety Feature (ESF) system and takes water from-the Service Water system.
and transfers it to enol certain safety-related loads (e.g. Emergency Diesel Generator).
The licensee initiated a P-1 Maintenance Request (MR 89-10679) to replace the switch.
, The inspector observed licensee electricians examine the switch after ., ' receiving permission from the shift supervisor. -The electricians ~then m s " prepared to pull the switch by making a hand sketch to show lifted leads.
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The work stopped without lifting leads when it was determined a spare replacement switch was not available.
As of December 5, 1989, work-had ' u not' progressed.to replace the switch.
' While in back of the cabinet, the inspector noted leads with two lugs < T.. each were hanging loose with untaped ends where Nuclear Instrument t Recorder (NR-46) had been removed.
The inspector determined that these , leads were still connected to the Nuclear Instruments (via isolators) and ' E still had low voltage (0-5 Volt D.C.) ap?1ied.- The inspector mentioned ' i this'to the assistant shift supervisor w1o taped the ends.
The licensee determined a significant personnel shock hazard did.not exist.
In ' discussions with licensee personnel, it was determined that this was not an acceptable practice, and the licensee discussed this with the I&C technician responsible.
No violations or deviations were identified.
5.
Surveillance (61726) . Auxiliary Feedwater (AFW) Pump Testing ! On October 17, 1989, during the performance of a Post-Maintenance Test (PMT) following a clam flush, the Auxiliary Feedwater (AFW) Terr a)peared to overspeed, and the trip and throttle valve (M0-3071)y Turbine tripped slut.
Just prior to the event the Terry Turbine was being controlled in manual from the control room with the Pressure Differential Controller (PDC) l 3083A in manual.
As the operator was reducing turbine throttle speed to l a minimum the personnel in the pump room noticed the turbine initially j l decreaseInspeed,andthennoticedtheturbinespeedincrease.
The l Terry Turbine then tripped on overspeed (later determined to be from the electrical overspeed controller).
The PDC-3083A output meter was noted not to have changed output, but the process indication went to full scale.
The licensee wrote Event Report (ER) 89-223 that developed
several potential causes for the overspeed event, but did not determine the cause of the overspeed.
The mechanical overspeed trip limit switch (LS-3) was found to be mechanically defective due to loose internal parts.
The limit switch was removed, however, electrical maintenance personnel had damaged the ! . . . - - -
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The licensee agreed that this was a problem, and was going to issue guidance on quarantining parts after an event.
Quarantiningof componentshasbeenapreviousissueatTrojan, s-During event evaluation the licensee determined the switch had 125 VDC t ! and 24 VDC entering it.
By design the 125 VDC system is not grounded, but the 24 VDC system was grotnded.
The ungrounded system is common to' the controller on the AFW pump.
The licensee did note that LS 3 could have grounded the common power ' supply (0-10) to the governor signal.
A zero signal to the governor j control signal would cause a 11aximum speed signal to go to the turbine.
, Due to the low amount of rotational inertia in the system, the Terry Turbine has'a fast response tc changes in speed from the governor controller.
The limit switch did cause a ground on the 0-10 ground detector recorder after the overspeed, but the ground detector recorder did not indicate a ground on it during the event.
To investigate the anomalous overspeed situation the licensee wrote Temporary Plant Test (TPT)-321 to install extra Instrumentation and monitor the pump's performance.
The test was reviewed and approved by appropriate levels of management.
The inspector reviewed the test procedure and observed the test, noting that permission was obtained from the control room and the testing was done by qualified personnel.
The AFW pumps are required to be operable in Mode.= 1-3 by Technical S]ecification (TS) 3/4.7.1.2, and the Limiting Condition for Operation (.00) was complied with when TPT-321 was performed.
The results from observing TPT-321 was that no abnormal degradations or operations were noted during the test on the Terry Turbine.
After one week of tests that operated the Terry Turbine 14 times without any problems the licensee decided to discontinue the special test _ program.
The licensee disconnected the instrumentation and ran Periodic Operating Test (P0T) 5-1.to verify continued AFW turbine operability, status.
Emergency Diesel Generator Surveillance ! On November 21, 1989, the inspector observed the licensee conduct the monthly surveillance on the Emergency Diesel Generators (EDG;s).
This surveillance is performed to meet the Technical Specification (TS) Surveillance 4.8.1.1.2.
L Tc meet this TS, the licensee performed parts of the approved Periodic l Operating Test (POT) 12-1, " Monthly Idle-Start and Loading of Emergency Diesel Generators.' The licensee obtained the proper tagouts for P01
12-1, and properly connected the Diesel Generator to the grid.
The licensee also installed, via Maintenance Request 89-10614, certain test gauges on the EDG to measure coolant temperature. When POT 12-1 was started, the Instrumentation and Control (I&C) technician had left the EDG room without connecting the gauge.
The operator informed the I&C technicien of the situation and he completed the installation.
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.,, - > . . . H The inspector obssrved the majority of the test and the return of the EDG ! to service; th9 instrumentation was within its calibration cycle; the testing was'cerformed by trained operators; and the test data met the acceptance criteria.
The licensee resolved minor discrepancies and the , I data was reviewed by the shift supervisor.
The licensee concluded test results met tho TS requirements.
, ~In conjunction with this test, the licensee ran POT 13-1 on the diesel, fuel system.
This POT was not completed due to problems with the level . switches (LS) 49048, 4905B and 4911B on the fuel oil day tank.
The licensee wrote MR 89-10669, -10670, -10671, and -10672 to calibrate these switches..These MRs were started on December 5, 1989, to enable the i reperformance of POT 13-1 to be successfully completed.
No violations or deviations were identified.
, 6.- Event Follow-up-(71707, 92700, 93702) A.
Control Building Thru-Wall Bolts The licensee had their Technical Specifications modified in 1980 to address a concern over the structural adequacy of two Control Building walls.
To stiffen the wall the licensee modified these two walls by adding bolts through the wall connected to bearing plates on both sides.
The license was modified with the inclusion of-Troian Technical Specification ModificationConnectionBolts."(TTS)3/4.7.11,"ControlBuilding This technical specification requires certain surveillances to assure that the bolts are carrying ' the required tension.
It was recognized at that time that there would.be concrete grout shrinkage and possible bolt relaxation, and these effects would diminish with time.
Therefore, TTS 3/4.7.11 included surveillance requirements to trend the lessening of bolt tension with time.
Further, a special sample of ten bolts was to be , E evaluated to determint the tension reduction with respect to time on a quarterly surveillance schedule for at. least the first 3 years.
The licensee performed the surveillances per Plant flodifications Surveillance.") 3 " Control Building Modification Through Bolt Procedcre (PMP i In July, 1989, a Quality Assurance audit of this technical soecification revealed that a trend analysis had not been performed for the third year inspection (1984) or the fifth year inspection in 1986.
The licensee generated Event Report (ER) 89-094 and Non-Conforming Activity Reports (NCARs) 89-277 and 89-292 on this ' evente Concerns generated by QA included: Trend Analysis was not performed as required by TS 4.7.11.1.9.
O Data was not put in permanent record storage for the completed fifth year inspection.
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P y . . .. , L . . .. . O The licensee's practice was to measure the initial tension and then retension the bolt.
This interrupted the tension decay Curve.
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The resident inspector reviewed the audit report with NRR and the licensee.
The resident inspector's review generated the following additional concerns:
The s of bolts a surve$ecialsamplegroukthinterval$pearedtohavemissedone ' 11ance interval ( by more aan 25L . The retensioning done on bolts may have adversely affected e nearby non-retensioned bolts.
The statistics done by the licensee on the bolts were not clear for the trend analysis.
In discussions with the resident inspector the licensee also concluded that one quarterly surveillance Interval had been missed.
Because the missed surveillance occurred in 1983 and the licensee has actions in progress to evaluate the surveillance program, this i violation was identified as a non-cited Severity Level V violation.
The licensee based on calculation concluded the effects of tensioningoneboltwerenotseenonadjacentbolts.
The statistics and analysis were reperformed and explained to the inspector. ' The inspector was informed that the licensee was going to formalize the calculation to get consistent results.
The licensee and the inspector concluded no safety issues existed.
B.
Pressurizer Safety Valve Disc Crack On October 27 - November 2, 1989, the licensee had Westinghouse perform testing on the rebuilt B Pressurizer Safety Valve (PSV)-80102 at an offsite facility.
This valve was rebuilt after it had been removed from the pressurizer due to steam leakage.
ihe - valve was totally rebuilt, and a new solid stellite disc put in the
valve.
On the 13th lift test, the new disc cracked, catastrophically failed and blew down the test rig.
Examination of the disc revealed a visible fracture extending across and through the entire disc.
The disc was held together due to a raised portion on the back of the disc which connects the disc to the valve stem.
A similar failure
of the disc installed ir. the plant could lead to a Small Break Loss l ofCoolantAccident(SBLOCA).
The licensee contacted the Nuclear Steam System Sup)1ier (Westinghouse) and the manufacturer of the valve (Crosby) w1o stated that the history of stellite disc failure. y were not aware of any other ! During the testing performed by Westinghouse on the valve, excessive chattering was noticed on the valve (on the lith lift).
This chattering was significant enough to trip the boiler off the line during one lift test.
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V-a . - . The licensee initiated Event Report (ER) 89-233 and Non-Conformance , ' Report (NCR) 89-548 to track this problem. ~The licensee also
searched procurement records for discs of the same-lot number to see if any were installed in the )lant.
There were 3.similar discs: . the one that failed, one in t1e warehouse, and the other with p Westinghouse.
The licensee performed a metallurgical investigation of the failed disc. -The investigation revealed no material problems and found that the gross properties of the material were within limits.
The licensee machined off the raised pcrtion of the disc and the disc broke apart in two pieces.
To date, the licensee has not determined the cause of the failure but concluded that there were two likely causes: L 1) A material flaw not readily apparent in the stellite disc existed.
2) The testing method used by Westinghouse resulted in the failure.
The licensee sent one half of the disc to Westinghouse for their ' ' analysis.
The analysis by Westinghouse resulted in Westinghouse L, stating that it probably was a materials problem (e.g. carbon segregation banding).
, .The licensee sent the other half.of the disc for independent s analysis.
At the end-of the inspection period, the licensee was , . still pursuing the cause of the failure.
The cause of the failure - will be followed up as part of the routine resident and engineering i section inspection.
C.
NEMA-4 Electrical Cabinets During the course of this inspection period, the licensee's Quality j L . Inspection'(QI) group identified a concern with NEMA 4 electrical cabinets.
These cabinets are' designed to exclude water from the E electrical components within the cabinet, and these cabinets may ! have safety-related connections in them.
A QI inspector noted that ' the hinge covers for 52 cabinets were not properly secured.
TheQI ' inspector then wrote a Non-Conforming Report (NCR) documenting the problem.
The Engineering department signed the NCR, and took it to the control room for. Shift Supervisor approval.
The Shift Supervisor and the Assistant Shift Supervisor did not approve the NCR, since it would place them in an unknown condition with respect to operability and it did not provide enough information to determine operability.
The engineer who signed the NCR then determined which cabinets were safety-related.
Concurrent with the correlation, the licensee tightened the NEMA 4 cabinets without a Maintenance Request; however, the as found-as left conditions of the cabinets were documented.
The plant personnel were informed by the Trojan Update (11/17/89) to properly secure cabinets after accessing them.
Additionally, facility . . .
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, managers were directed to reemphasize to their personnel to make ' sure these cabinets were tight.
After the licensee tightened the electrical cabinets, the inspector saot checked the cabinets to evaluate licensee corrective actions.
T1e inspector identified several of these cabinets as still having-loose covers.
Later, the inspector took a compliance engineer and the maintenance Branch Manager to show which cabinets he had identified.
At that time, the inspector branch manager, and complianceengineernotedachemistopenIngthedoortoa Containment Radiation Monitor.
The person had not fully closed this , ' door.
When asked why not he stated that he.had to be in that cabinet often.
Insidethiscabinettherealsowasagreasetube,a screwdriver and two ' hot' tools (crescent wrench and a box end - wrench).
Due to these findings, the licensee again told personnel.about these cabinets.
The licensee plans on writing a lower level procedure to inspect these cabinets.
The licensee's actions will be followed by routine resident inspection.
, D.
Auxiliary Feedwater Flow Transmitter 3043-During the course of routine inspection, the inspector noticed a Maintenance Request (MR) tag, dated July 24, 1989, on Flow Indicator (FI)-3043 in the Control Room and Remote Shutdown Station.
This is the indicator for Auxiliary Feedwater (AFW) to the.B Steam Generator.
At that time the licensee had written a Priority 1 (P-1)MRtofixthisindIcator,whichwasreadingabout100gpmwhen the pump was not running.
A P-1 MR is one required to be started in one day.
On November 23, 1989, the inspector researched the reason for the l delay in repairing FI-3043.
The inspector was informed by the licensee that the manufacturer was going out of business.
The licensee provided the insaector with a chronology of the procurement of the items necessary.
- rom-the discussion with the licensee, it
, was determined that procurement on these items started on October i 12, 1989.
The priorities assigned to Maintenance' Requests (MRs) are stated in l Administrative Order (AO) 3-9, Maintenance Requests.
The classification assigned to the MR, the reasons for the classification, and the person responsible for authorizing the MR is shown below.
Priority Classification Reason Authorizing Manager , Urgent-MRs requiring immediate action Shift Supervisor or ,to correct deficiencies that Assistant Shift may cause: Supervisor Immediate Reduction in Plant - e, .
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. . , . - . Output - _ . Degraded Reactor Safety - Imminent Personnel Danger - P-1 'MRs requiring to_be started in Shift Supervisor or 1 day to correct deficiencies Assistant Shift which: Supervisor Affect' equipment safety - Create a personnel Safety - Hazard Restrict Return to Operation - at Full Power , P-2 MRs recuiring to be evaluated or Shift Supervisor or startec within 7 days Assistant Shift Supervisor Affect equipment safety - Create a personnel Safety - Hazard Restrict Return to Operation - at Full Power Routine MRs that may be handled as scheduled to correct deficiencies.
The inspe-tor was informed that the licensee had numerous P-1 MRs that had been open since 1988.
There were also several MRs that were awaiting parts to start work that were several months old.
The licensee's actions to monitor the status of P-1 MRs and the backlog of open work items will be followed up as a course of routine resident inspector activity.
One non-cited violation was idehtified.
7.
Follow-up of Licensee Event Reports [LERs) (90712, 92700) LER 88-32, Pevision'1, (Closed), " Control Room Emergency Ventilation System Inoperable Due to Inability to Maintain Sufficient Positive Pressure in the Control Room." This revised event report provided additional.information as to the cause of the event and corrective acticns' taken or proposed.
The licensee concluded that on September 24, .1989, during the conduct of Periodic Operating Test (P0T) 20-1, " Control Room Emergency Ventilation Performance," Control Room Emergency Ventilation System (CB-1A) could not maintain a positive pressure of greater than 0.125 inches water gauge because balancing damper BD-104, which is in the return air duct, was mispositioned.
Initially, the licensee concluded the cause of the event was due to having auxiliary j building doors, which are normally closed, propped open.
l l ! ' Inspection report 50-344/88-43 closed Revision 0 of this licensee event report and, due to the incompleteness of the LER, generated open item 50-344/88-43-01.
Inspection report 50-344/88-50 documented additional
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l inspection concerning the event and closed the open item based on j ' determining the cause of the event to be a mispositioned damper (BD-104).
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, i This revised LER addressed the concerns raised in previous inspections and established corrective actions that when implemented should prevent ! recurrence of the event.
The inspectors verified procedural changes
committed to.in the-LER had been implemented.
The inspectors examined
the present locking device on damper BD-104 and concluded the device would prevent accidental mispositioning of the damper.
The inspectors, to ensure procedure charges were understood by shift management ' discussed with a number of the shift supervisors the actions req,uired if.
control room ventilations systems could not achieve the positive pressure criteria.
LER 89-07, Revision 1, (Closed), " Control Room Isolation Damper Closure ' Time exceeds Required Maximum." This revised event report provided additional information as to the root cause of the event.
While the cause of the event could not be firmly established, the licensee concluded the most probable causes of the event were pressure switches > that control ventilation fans (consequently damper closure time) were ! -incorrectly set.
Other contributing causes were a poorly written
surveillance procedure and a wiring error.
As corrective actions, the licensee changed the setpoints of the pressure switches, verified correct wirin{i of the damper indicating lights, revised the surveillance procedure and reperformed the surveillance.
The inspectors reviewed the pressure switches' setpoint change and the respective safety evaluation, the revised surveillance procedure and.the completed surveillance data sheet for accuracy and completeness.
No deficiencies were identified.
LER 89-10, Revision 0, (Closed), " Failure to Follow Procedure Actuates - Auxiliary Feedwater System." This event report described an actuation of ' the engineered safety' feature (ESF) auxiliary feedwater pumps due to noncompliance with procedures. As corrective action, the individual involved received positive discipline.
The inspectors verified the auxiliary feedwater pumps started and operated per design, and steam generator water levels were maintained.
LER 89-16, Revision 1, (Closed), " Inadequate Management Oversight ' Procedure Deficiencies, and Personnel Errors Rasult in Power Operation with an Inoperable Containment Recirculation Sump." This revised licensee event report provided clarifying information on the sequence of events and additional corrective actions taken or proposed to correct deficiencies.
Special inspection report 50-344/89-19 described NRC inspection associated with this event.
LER 89-19, Revision 0, (Closed), " Pressurizer Safety Valve Setpoint Found Out-of-Tolerance During Testing." This event report described the licensee finding a pressurizer safety valve, PSV-80108, lift set of specification high (2552 psig vice between 2435 to 2535 psig) point out during the conduct of surveillance testing.
The licensee concluded the cause of 'PSV-8010B being out of specification was a change of the testing method.
Specifically, previously the safety valve had been tested with the loop
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Industry experience indicated testing without the loop seal intact may result in setpoint errors.
Therefore, because the valve had previously been tested without a loop seal and was subsequently tested with a loop seal, the setpoint had changed as industry experience indicated.
Thelicenseeadjustedtheliftsetpointtowithintechnical specification requirements and revised the safety valve test procedure to test with full loop seals.
The inspectors observed the lift tests conducted per Maintenance Procedure 5-1, " Pressurizer Safety Valve Inservice Test".
The inspector's observations are documented in inspection report 50-344/89-24.
LER 89-20, Revision 0, (Closed), " Instrument Loop Inaccuracies not Accounted for in Performing Surveillance on Reactor Coolant System Average Temperature Could Allow Operatici. Outside Analysis." This event ' report described that since 1976, the reactor may have been operated with an average temperature (Tave) higher than the average temperature that was assumed in the safety analysis.
This was due to the licensee not recognizing that technical specification 3.2.5 values were actual values vice values that included instrument error.
Thelicensee,inconjunction with the vendor, concluded since additional conservatism existed in the > -Departure from Nucleate Boiling (DNB) parameters and the Rod Control System and the Pressure Control System were in automatic with the appropriate program that safety was not impacted.
As corrective action, i , ' procedures were changed to account for instrument uncertainties.
- Additionally, the licensee will by March 1992, complete and review all of ' the other Technical Specification values to determine if instrument , uncertainties are properly applied.
The inspectors reviewed the vendor .
. provided documentation that concluded additional safety margin existed in the DNS parameters.
< 8.
Spent cuel Pool Inspection (86700) . The Spent Fuel Pool (SFP) is a Seismic Category I reinforced concrete, stainless steel lined structure which provides storage space for irradiated fuel from the reactor.- The fuel stora the Spent Fuel Pool (SFP), Cask Loading Pit (CLP)ge structure consists of and the Fuel Transfer
Canal (FTC).
The inside surfaces of the SFP, CLP and FTC are lined with . 1/4 inch thick welded stainless steel.
The Fuel Storage and handling systems are described in section 9.1 of the Final Safety Analysis Report (FSAR).
There were no fuel handling activities onsite during the inspection period, but the SFP design, leakage, and system were compared to technical specifications 3.9.11 and 3.9.12 and the FSAR.
There is also a SFP leak detection system which consists of a network of 2" by 2" monitoring trenches located in the reinforced concrete under the liner plates.
These leak detection trenches will drain through normally open valves into manifolds located on the 45 foot level of the fuel building and then into the Dirty Radioactive Waste System.
The approximate location of the leak may be determined by which leak detection valve exhibits a discharge and the rate of the discharge.
There is one valve going into the manifold that shows evidence of leakage.
The inspector determined that this leakage had probably been happening since initial startup.
The leak is so slow that the concentrated boric acid had crystallized before reaching the drain.
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the FTC only.
The leakrate is low and has not been quantified, j The inspector verified that the SFP water level was higher than the . minimum level specified in the technical specifications.
The spent fuel i ventilation system was operable at the time, and the SFP tem)erature was within limits. There was fine dust / debris floating on the S P surface ! due to the SFP skimmer pump being repaired.
No violations or deviations were identified, f: 10.
Engineered Safety Feature Walkdown (71710) , b A.
Containment Spray System While walking down the containment spray system at the beginning of the inspection period, /86the inspector noted ceveral gauge calibration stickers (FI-2067:11/13 PI-2061:12/3/85) that appeared overdue beyondthe2yearcalibratlonfrequency.
The inspector then reviewed the calibration records for these indicators and other indicators in the Containment Spray (CS) system.
The indicators reviewed and calibration dates found were: ' Indicator Description Calibration Dates PI-2061 A CS Pump Suction Press.
12/14/83, 12/3/85, 12/3/87 PI-2070A A CS Pump Discharge Press, 1/17/83, 9/25/85, 10/11/87 FI-2066 A CS Pump Flow 9/23/85, 9/21/87 PI-2063 8 CS Pump Suction Press.
2/17/83,3/15/85,3/19/87, 3/9/89 PI-2070B B CS Pump Discharge Press.
11/5/84, 11/7/86, 11/13/86, 6/1/88 FI-2067 8 CS Pump Flow 11/5/84, 11/13/86, 12/21/88 ' The inspector noted problems with PI-2070A and PI-2063.
The data indicate that the 2 year frequency was exceeded in 1983 for PI-2070A.
On followup, the inspector determined that the frequency for calibration of PI-2070A was changed in 1984 to 3 years.
The data review for PI-2063 found that it was always failing low.
The inspector discussed the reeson for PI-2063 always failing low with licensee personnel.
Due to the questions asked by the inspector, , the instrumentation and control group wrote a Request for Evaluation ' (RFE) by the Plant System Engineering (PSE) department.
In April 1989 while reviewing the Out-of-Calibration (000) investigation form a licensee engineer noticed that this indicator had a history of l failing low.
He documented in the comments section of the 000 i investigation sheet that an RFE should be evaluated; however, he did i not initiate an RFE.
The inspector discussed PI-2063 with the ' system engineer.
The engineer developed several proposed actions to investigate this situation, as noted in letter MRM 002-89.
The i licensee's plan was to initially calibrate the transmitter during the next surveillance, and if it was 000, to calibrate it every time ' the surveillance needed to be run.
If the indicator was not out of
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s ' The inspector discussed this approach with the manager of Plant ' ~ , , . System Engineering, who did not have immediate concerns with the .
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. approach. The inspector was concerned because this approach may not i determine when or how fast the transmitter was going 000.
Subsequently after discussions with the inspector, the licensee
changed their approach to check if the transmitter is 000, and if ( not, then not recalibrate. The recorded values.of the transmitter calibration should give an indication of when the' transmitter would " , ," go out of calibration.
! The licensee had-not effectively trended the failure of this
indicator.
The inspector had previously identified other problems with the licensee's 000 program (see 50-344/89-24). The' licensee's solution to this indicator failing low will be followed as a course of continuing inspection.
B.
Service Water System While performing a walkdown of the Service Water system, the.
inspector noticed that lubricating water cooling valves' (SWO17, ' . SW-019 and SW-018, SW-020) positions to the Swing Pump (1080) were a reversed. The inspector noted this and informed the licensee about L this finding. The Piping and Instrument Diagram (P&lD) and isometric drawings both showed the valves in another configuration.
Subsequently, the inspector noted that the tags on these valves were switched back. The licensee had rehung the tags after determining ' '
that the tags were wrong, 'These valves were on the locked valve list. During the performance of valve position verification, the licensee did not recognize that ' these valves were not properly labeled.
This suggests personnel who position plant valves do not always pay attention to detail. The inspector discussed his concerns with the licensee on identifying valves properly on the lock valve list. The licensee agreed to do some spot checking of the valve lineups and emphasized the importance of the locked valve list to the operators. This will be-followed up in routine inspection activities.
No violations or deviations were identified.
' _11. _ Severity Level V Violations As_ stated in Section V.A of 10 CFR Part 2, Appendix C, " General Statement of Policy and Procedure for NRC Enforcement Actions," 53 Fed. Reg. 40019 (October 13,1988), a Notice of Violation will not formally be issued for isolated' Severity Level V violations provided that the licensee has ' initiated appropriate corrective actions before the inspection ends. One ! apparent Severity Level V violation for which a Notice of Violation was not issued is discussed in paragraph 6.A of this report.
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