IR 05000344/1988029
| ML20154F645 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 08/31/1988 |
| From: | Rebecca Barr, Mendonca M, Suh G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20154F630 | List: |
| References | |
| 50-344-88-29, NUDOCS 8809200076 | |
| Download: ML20154F645 (21) | |
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U.S. NUCLEAR REGULATORY CO MISSION
REGION V
Report No.
50-344/88-29 Docket No.
50-344 License No.
NDF-1 i
Licensee:
Portland General Electric Company 121 S.W. Salmon Street Portland, OR 97204 Facility Name: Trojan Inspection at:
Rainier, Oregon Inspection conducted:
June 19 - July 30, 1988 Inspectors:
% N-AA
- />/ // r R. C. Barr
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Date Signed Senior Resident Inspector
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4;l.o/I'l G. Y. Suh Date Lyned
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Resident Inspector r
Approved Fy:
% %h M e Agn CPL /N /#f M. M. Mendonca, Chief F
Date Signed Reactor Projects Section 1 i
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Sunmary:
Inspection on June 19 - July 30, 1988 (Report 50-344/88-2s)
Areas Inspected: Routine inspection of operational safety verification, maintenance, surveillance, event follow up, and open item follow up.
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Inspection procedures 30703, 61701, 61726, 62703, 71707, 71709, 71710, 71711, 71881, 92700, 92701, 92702, 92703, and 93702 were used as guidance during the
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conduct of the inspection.
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Results:
The adequacy of root cause evaluation on several licensee event reports (LER)
was considered lacking. LER submittals by PGE are not consistently clear and specific so that readers can understand the complete event and corrective
actions to prevent recurrence.
This topic has been the subject of previous
j disc'Jssion between the utility and Regional managerent.
Further, this problem is particularly botherscoe in that LERs undergo substantial r:anagement review
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and are indicative of a managen:ent team's self-critical approach to the resolution of plant problems.
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Additionally, this inspection fcund that licensee evaluation and trending to evaluate and prevent similar events has not been totally effective. Repeat events, such as control room emergency ventilation system deficiencies, fire doors made inoperable and missed fire watches, seem to recur frequently.
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DETAILS 1.
persons Contacted
- D.W. Cockfield, Vice President, Nuclear
- C.A. Olmstead, Plant General Manager
- L.W. Erickson, Manager, Nuclear Quality Assurance R.P. Schmitt, Manager, Operations and Maintenance
- D.W. Swan, Manager Technical Services M.J. Singh, Manager, Plant Modifications
- J.D. Reid, Manager, Plant Services
- J.W. Lentsch, Manager, Personnel Protection J.M. Anderson, Manager, Material Services
- R.E. Susee, Manager, on assignment
- D.F. Levin, Supervisor, Plant Modifications R.L. Russell, Operations Supervisor R.H. Budaeck, Assistant Operations Supervisor
- D.L. Bennett, Mairtenance Supervisor R.A. Reinart, Instrument and Control Supervisor T.0, Meek, Radiation Prottetion Supervisor R.W. Ritschard Security Supervisor C.H. Brown, Operations Branch Manager. Quality Assurance
- D.L. Nordstrom Nuclear Engineer, Nuclear Safety and Regulation The inspectors also interviewed and talked with other licensee employees during the course of the inspection. These included shift supervisors, reactor and cuxiliary operators, maintenance personnel, plant technicians ar.d engineers, and quality essurance personnel.
- Denotes those attending the exit interview.
2.
Plant Status At the start of the inspection period, the plant was in Mode 5 at 90 degrees F in the fifth week of the 1988 3nnual refueling outage. Major activities in progress were prescuriZer surge line repairs, remote shutdown station relocation, large bore pipe hanger refurbishment and control room ventilation system upgrades. Outage work was completed and heatup begun on July 3,1986. With the plant in Mode 3, an inadvertent safety injection occurred on icw pressurizer pressure en July 4.
On July 5, 1988 with the plant in Mode 3. ECCS check valve leakage, exceeding technical specification requirerents, was identified during surveillance testing. The plant was returned to Mode 5 on July 6, 1988, to correct the valve leakage. The facility returned to Mode 3 on July 7 and on July 6 acceptance testing of the remote shutdown station was f orfonned satisfactorily to verify the capability to conduct a controlled cooldown of the reactor from a location other than the control room. On July 10, 1908, the reactor achieved criticality and, after completing physics testir.g. achieved 100% power en July 22, 1988 On July 27, 1988 power was reduced to 50% and returned to 100% power to replace a grounded trip solenoid on the "A" feedwater regulating valve. Additionally, throughout the remainder of the inspection period short-tenn power reductions were
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necessary to maintain turbine backpressure within specifications during hot weather conditions.
3.
Operational Safety Verification (71707. 71709, 71710 and 71881)
During this inspection period, the inspectors observed and examinco
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activities to verify the operational safety of the licensee's facility.
The observations and examinations of those activities wirre conducted on a daily, weekly or biweekly basis.
Daily the inspectors observed control room activities to verify the
licensee's adherence to limiting conditions for operation as prescribed
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in the facility Technical Specifications. Logs, instrumentation,
recorder traces, and other operational records were examined to obtain
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information en plant conditions, trends, and compliance with regulations.
On occasions when a shift turnover was in progress, the turnover of information cn plant status was observed to determine that pertinent information was relayed to the oncoming shift personnel.
Each week the inspectors teured the accessible areas of the facility to
cbserve the following items:
a General plant and equipment conditions.
b Maintenance requests and repairs.
c Fire hazards and fire fighting equipment.
d Ignition sources and flammable material control, e
Conduct of activities in accordance with the licensee's administrative controls and approved procedures.
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f Interiors of electrical and control panels.
Implementation of the licensee's physical security plan.
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Rar11ation protection controls.
Plant housekeeping and cleanliness.
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Radioactive waste systeus, j
s Proper storage of compressed gas bottles.
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Veekly, the inspectors examined the licensee's equipment clearance
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control with respect to removal of equipment from service to determine i
j that the licensee complied with technical specification limiting
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l conditions for operation.
ictive clearances were spot-checked to ensure
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that their issuance was cct.istent with plant status and maintenance
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evolutions. Legs of jumpers, bypasses, caution and test tags were examined by the inspectors.
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Each week the inspectors cenversed with operators in the control room, and with c:her plant personnel. The discussions centered on pertinent
i topics relating to general plant conditions, procedures, security.
l training and other topics related to in-progress work activities, t
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The inspectors examined the licensee's nonconformance reports (NCRs) to confim that deficiencies were identified and tracked by the system.
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Identified nonconfonnances were being tracked and followed to the
ccmpletion of corrective action, i
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Routine inspections of the licensee's physical security program were
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perfonned in the areas of access control, organization and staffing, and
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detection and assessment systems. The inspectors observed the access control measures used at the ertrance to the protected area, verified the r
integrity of portions of the protected area barrier and vital area
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barriers, and observed in several instances the implementation of
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compensatory measures upon breach of vital area barriers.
Portions of (
the isolation zone were verified to be free of obstructions.
Functioning l
l of central and secondary alarm stations (including the use of CCTV
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nonitors) was observed. On a sampling basis, the inspectors verified
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that the required minimum number of arred guards and individuals authorized to direct security activities were on site.
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l The inspectors conducted routine inspections of selected activities of l
the licensee's radiological protection program. A sampling of radiation
.i work permits (RWP) was reviewed for completeness and adequacy of information.
During the course of inspection activities and periodic
tours of plant areas, the inspectors verified proper use of personnel l
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monitoring equipment, observed individuals leaving the radiation
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i controlled area and signing out on approptiate RWP's, and observed the posting of radiation areas and contaninated areas.
Posted radiation levels at locations within the Juel and auxiliary buildings were verified
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by the inspectors using both NRC and licensee portable survey neters. The
involvement of health physics supervisors and engineers and their i
j awareness of significant plant activitiys was assessed through l
conversations and review of RWP sign-in r% ords.
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l The inspectors verified the r,perability of selected engineered safety i
features.
This was done by direct visoal verification of the correct
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position of valves, availability of power, cooling water supply, system i
j integrity and general condition of equipment, as applicable. The 120
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j Volt AC Instrument System was verified operable during this inspection
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period.
j No violations or deviations were identified.
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j 4.
Maintenance (62703)
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A containment isolation valve on the "B" steam generator blowdoan line, t
i M0-2813, failed in midqosition curing plant operation in Mode 3. Hot j
Standby. M0-2613, a motor operated, two-inch Y-pattern gicbe valve, had
been tostalled during the 1986 refueling cutage as part of major
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modifications performed on the plant steam generator blowdewn system.
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The four blowdown lines each have two containrent isolation valves, all of the same manufacturer and model, and al's installed during the outage.
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l The inspectors verified that appropriate technical specification action
i statements in response to the valve failure were completed in a timely j
nanner; reviewed clearance 88/1499, which had been processed for the
maintenance work on N0-2813, for adequacy end verified that all danger j
tags had been hung; and reviewed the essociated 56fety related outage l
sheet.
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l The valve plug was found to be galled in the valve bore, and the plug a9d i
valve stem were renoved with the assistance of a hydraulic piston. Metal
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transfer between the valve plug and valve internal components was evident. The inspectors examined the valve plug, valve stem and the valve body after the honing of the bore. Galling and metal transfer was evident on the plug. Wear was also observed where the valve stem and plug made contact.
Honing of the valve bore and replacen.ent of the valve plug and stem was accorplished with Maintenance Request MR 88-5821. The inspectors observed a dye check of the disc to the seat, cleaning out of metal shavings from the valve internal components, cleanliness material control and cleanliness closeout inspection by a QC representative, and portions of the valve reassembly.
The licensee initiated Nonconformance Repcrt NCR 88-299 in response to the failure of M0-2813. Given other failures of this valve rodel during the outage, an engineering investigation with the assistance of the valve vendor was iaitiated to detennine the root cause and recommend corrective actions. Discussions with operations management indicated that the
licensee considered the steam generator blowdown containment isolati M
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valves operable. The licensee has initiated and implemented a detailed evaluation in accordance with an action plan.
The inspectors will continue to evaluate licensee actions in routine follow-up, t
i Nu violations or deviations were identified.
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Surveillance (61726,61701,71711J, i
Remote Shutdown Station Test
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During the 1988 refueling outage, major rodifications were made to Appendix R - related remote shutdown features. A major test of the
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system was performed under Temporary Plant Test TPT-249, titled "Control i
Room Inaccessibility Test." The objectives of the test as stated in TPT-249 were to demonstrate that the pl6nt could be maintained in Mode 3 hot standby, conditions for greater than 30 minutes from the remote shutdown station and to demonstrate RCS cooldown of at least 50 degrees from the remote shutdown station.
During the test, control was dercupied i
from the control room for various components but control roon indication
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was maintained. The inspectors discussed the test procedure with
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operations and engineering personnel. Detailed review of the procedure
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for consistency with regulatory requirements was outside the scope of
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this inspection. Operations ranagement involvement was evident in
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controlling the effect of the test on plant conditions and to assure the
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return of control to the control room if specified test termination i
criteria were ret prior to test completion.
l The inspectors observed the performance of TPT-E49 from both the control roon and remote shutdown station. On shift personnel not involved in the i
test were verffied to meet technical specification requirements for
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on-site coverage. The inspectors observed that TPT-249 was conducted by a minimum shift crew necessary for safe shutdown of five operators with
the assistance of the shift technical advisor. During the first attempt i
to maintain hot standby conditions, the test was aborted due to loss of
auxiliary feedwater flow indication to two steam generators at the remote l
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shutdown station. Return of control to the control room was performed according to procedure and without incident. A temporary mudification
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was approved and installed to defeat voltage protection circuitry which had led to the loss of flow indication. The test was completed satisfactorily on the second attenpt.
The inspectors verified that test acceptance criteria were met for RCS pressure and temperature, steam generator level and pressurizer level during the course of the test. The inspectors noted that test prerequisites were met or documented to have been met; that test personnel had copies of the test procedure available during the conduct of the test; and that systems were restored to nonnal conditions per the test procedure.
Pressurizer Surge Line Movement
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During the 1988 refueling outage, unexpected pressurizer surge line
movement was observed which resulted in contact with a pipe whip
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restraint and apparent plastic defomation of the surge line.
Inspection Report 50-344/E0-25 docunented follcw-up inspection of licensee actions.
Inspection of the subsequent replacement of the 90-degree elbow directly i
below the pressurizer was documented in Inspection Report 50-344/88-24.
The licensee initiated a detailed examination of the replaced elbow and
adjoining piping to detemine the nature and cause of the relevant indications observed in ultrasonic examination of the two welds associated with the elbew. NRR has requested a portion of the elbow and piping for metallurgical examination.
In addition to replacewent of the surge line elbcw, the licensee
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perfomed modifications of supports and whip restraints associated with the pressurizer surge line.
During this inspection period, the
inspectors observed the licensee's program to monitor the temperature and deflection of the surge line during pressurizer heat-up, pressurizer bubble formation, ahi plant heat up to nomal operating temperature.
The program was controlled by Temporary Plant Test TPT-261, titled
"Pressurizer Surge Line Monitoring." The objectives of the program as
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to monitor and evaluate surge line movement for i
possible contact with pipe whip restraints; monitor the degree of themal i
stratification in the surge line: and to confim the validity of the i
licensee's piping stress analysis and fatigue analysis.
The inspectors
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reviewed the procedure and observed portions of the data acquisition
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during plant startup from the refueling cutage. The inspectors noted that significant events were evaluated by the cognizant engineer a.d supervisor as documented on "unusual condition evaluation" forms. TPT-261
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provided for further data acquisition during puwer operation and plant cooldown.
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i The inspectors perforr.ed a walkdown prior to pressurizer heat-up of the pressurizer surge line instrun,entation installed per Temporary
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Hodification TH 88-071. Temperature was monitored by resistance temperature detectors located along the surge line to detect suspected
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temperature stratification.
Surge line movement was reasured with linear
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potentiometers. At the time of the inspection, ell linear potentiometers j
were in place except for those to be located between whip restraint 1.5 and the pressurizer where modifications to pipe supports were still in progress. TF1-261 provided for linear potentiemeters located to monitor
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pressurizer movements being reflected on the surge line. At the time of inspection, initial data evaluation indicated that acceptance criteria for surge line terrperature and deflection were being met with significant conditions noted for further review on "unusual condition evaluation" forms. The licensee stated in a May 31, 1988 meeting that it would provide a sumary of the test results to NRR for review.
Plant Start-up from Refueling The licensee's procedures to control startup from the refueling outage included Administrative Order A0-3-25, titled "Ready for Startup "
General Operating Instruction G01-1, titled "Plant Startup frem Cold Shutdown to Hot Standby," and General Operating Instruction G01-2, titled
"Plant Startup from Hot Standby to Power Operation." The intent of these procedures was to provide the administrative controls for returning to an operable status those safety related systems which were disturbed during the refueling outage and to assure adherence to licensee technical specification requirements.
The inspectors observed portions of the licensee's ready for startup program and use of GOI-1 and G01-2 during plant startup. The inspectors performed walkthroughs of portions of the main feedwater lir.es which had been replaced during the outage and of the pressurizer surge line which had an elbow and adjoining piping replaced.
Nondestructive examinations and hydrostatic pressure tests were performed prior to returning these comporients to service.
The inspectors witnessed portions of pressurizer heatup, pressurizer bubble formation, RCS heatup to Mode 3, and low power physics testing in Mode 2.
The inspectors observed measurement of hot zero power control bank differential worth for bank 0, measurement of the n:oderator tt.mperature coefficient at hot zero power with all rods out and with bank 0 inserted, and measuretrent of the critical boron concentration with bank 0 inserted at hot zero power.
Calibration of instruments associated with the reactivity ccirputer were verified to be current.
Physics testing was controlled by Periodic Engineering Test PET-13-1, titled "Reload Cycle 11 Startup Low Power Physics Tests," and Periodic Engineering Test PET-13-2, titled "Reload Cycle 11 No Load and at Power Tests." The inspectors will review the test results upon completion of physics testing and data evaluation.
No violations or deviations were identified.
6.
Follow-up On L1 ensee Event Reports (LER) (92700, 92701)
LER 86-02, Rev. 1, (Closed) "Control Room Habitability Decraded Due To Structural Equipment, Operational and Design Deficiencies." Associated with this event, the licensee was cited with a severity level 2 violation and assessed an $80,000 civil penalty. The licensee responded to the hotice of Violation via letter, dated May 22, 1986. Additionally, via letters dated May 19, 1986 and June 13, 1986, PGE provided an action plan and updated action plan, respectively. PGE delayed some of the proposed actions past the intended implementation dates with NRR knowledge and concurrence. The licensee continues to track Control Building Mcdifications through two action plans (Control Room Emergency Ventilation system (CB-1) and Spent Fuel Pool Exhaust System (AB-4))
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Eased on PGE's response to the Notice of Violation and the updated action plans this LER is closed.
However, LER 88-18, Rev 0 indicates that one of the actior s. leak identification by smoke detection, was not capable of measuring leak rates required to be met in the original design assumptions.
Therefore, the Control Room ventilaticn system may have operated outside the assumed 10 cfm unfiltered in-leakage since 1986.
Further discussion of excessive unfiltered air in-leakage is discussed in LER 88-18, Rev. O.
LER 86-03, Rev.1, (Closed) "Residual Heat Rcn' eval Systen Inoperable Due
_to Hisunderstanding of Design Bases." Associated with this event, the licensee was cited with a severity level 3 violation and assesser.i a
$50,000 civil penalb. The licensee responded to the violation via letter, dated Novernber 14, 1986. This LER revision, as follow up to the event, further outlined proposed actions to resolve additional concerns identified during licensee followup of this event. The inspectors performed an in-office review of the event report and further verified that the corrective actions comitted to in the LER were completed or in process. The renaining corrective action to be completed is the re-evaluation of the environmental qualification of all ECCS valves to assure the proper level of qualification. Based on the action stated in the responsn to the Notice of Violation and those stated in this revision of the LER, this LER is closed.
LER 87-06 Rev. 2, (Closed) "Deficiencies in Flood Protection Design Provisions."
This licensee event report was submittea to update the status of corrective actions for inadequately sized turbine tuilding flood louvers and other flood protection design feature deficiencies.
This LER is closed based on in-office review and installation verification of the newly designed flood louvers.
LER 88-03, Rev. 0/Rev. 1, (Closed) "Single Failure Mechanism Discovered Which Could Overpressurize Containment Electrical Penetration Seal." As
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a result of a system engineer's review of the electrical penetration sealing system, a sinole failure mechantsn that could have rendered the penetrations inoperable was icentified.
Inadequate pressure relieving capability downstream of the nitrogen supply regulator could have i
resulted in electrical penetrations seeing 200 psi vice the design 100 psi if the nitregen supply regulator would have failed.
Upon recognition of the problen the licensee isolated the nitrogen supply, which is acceptable since the nitrogen purge is not required for the penetration to be considered operable.
Subsequently, the licensee is, on an every shift basis, maintaining the electrical peretrations purged by nanus 11y adjusting nitregen pressure.
The licensee is awaiting additional infernation from the tranufacturer of the electrical renetrations prior to implementing final corrective actions. This LER is closed based on the future cwaitted actions of submitting a revised LER addressing root cause and the finel corrective actions.
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LER 88-04 Rev 0/Rev. 1, (Closed) "Containment Penetrations Not Verified Closed as Required by Technical Specifications." On March 28, 1988, i
feedwater drain valves FW-079, 082, 083 and 086 were found open. The f
Final Safety Analysis Report assumed these valves to be locked closed.
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i Technical Specifications 3.6.3.1 identifies these valves as manual j
containment isolation valves and must, therefore, be closed to be i
considered operable. The licensee determined the root cause of this event to be the inadequate perfonnance of the 10 CFR 50.59 review included in the associated safety evaluation.
Based on the proposed
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corrective actions in these licensee event reports and the licensee's Notice of Violation response dated July 15, 1988, these LER's are closed.
Continued follcwup of license corrective action will be under the violation open item 88-13-01.
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L_ER 68-05, Rev. O. (Closed) "Surveillance Interval for Chilled Water Return Valves Exceeded." Due to personnel error and insufficient procedural detail, Chilled Water Return Yalves CV-10014 end CV-10015 were not surveilled at minimum required intervels. Subsequent valve testing
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of CV-10014 and CV-10015 necessitated further reducing valve cycling
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j surveillance frequency to weekly. Also, as a result of this event, the licensee reviewed in-service testing records and identified additional
valves that had not been surveilled at correct intervals.
To ensure surveillance inspection intervals are met in the future, the
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licensee has revised FET 9-4, "Documentation of Inservice Testing Data for Pumps and Valves," to clarify when testing at increased frequency is t
required. Also, to improve Inservice Testing Program implementation.
l organization changes have been made, i
LER 88-06 Rev. O. (Closed) " Charging Pum) Seal Leakage d
l Greater than F5AR Assumed Limits". With t1e reactor at 100% power, the j
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"B Centrifugal Charging pump (CCP) mechanical seal experienced leakage l
of approximately 3000 cc/hr on April 5,1988. Since this leakage
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contributed to total ECCS leakage external to containment and exceeded
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the 1580 cc/hr leakage limit, the licensee declared both trains of (
control room ventilation inoperable which required a reactor shutdown in
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one hour if the leakage could not be isolated. The "B" CCP was isolated
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and the seal ledage stopped within thirty-five minutes.
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The licensee determined the seal failure to be due to uneven mechanical
seal rear caused by boric acid crystal buildup on the seal. To prevent a
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similar CCP mechanical seal failure, CCP operating practice has been
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revised to provide for more frequent pump operation which is expected to i
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prevent boric acid crystal buildup on the seal.
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LER 88-14. Rev. O, (Closed) "Containment Ventilation Isolation On High Containment Radioactivity Signal." On May 17, 1988, the contairrer.t i
a purge supply and exhaust systems isolated on hioh radioactivity as sensed
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by the contaircent low level noble gas monitor (PRM-1C). The radioactive l
gas originated from the primary side of the 'D' Steam Generator (S/G)
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j which had been opened for maintenance and was being exhausted to the j
containment atmosphere.
Planning for the evolution of exhausting the 'D'
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S/G primary side had correctly identified the need to filter the (
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exhausted gas for particulates but, by oversight, did not acccunt for the i
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presence of noble gases. Licensee corrective action included a planned revision to MP-5-3 "Steam Generator Primary Manway Removal and
Installation" to add a cautionary statenent that a noble gas release
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would be probable and to evaluate outage evolutions, on a case by case l
basis that could result in radioactive releases. The licensee also noted a previous event (as reported on LER 88-10) had recently occurred. Based
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on licensee planned actions this event is closed.
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LER 88-15, Rev. O, (Closed) "Containment Ventilation isolation on High
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Containment Radioactivity While Testing Process and Effluent Radiation
Monitor (PRM-lC)." On May 19, 1968, the containment purge supply and exhaust systems isolated due to a technician error while perfoming corrective maintenance on the radiation monitor.
The technician released
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the alarm reset button prior to the monitor radiatien level decaying
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below the isolation level subsequent to performing a source check.
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Licensee corrective action included counselling the technician and l
training all I&C technicians on this event. The event report did not
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address whether technician training en radiation nonitors had included I
operation of the check source, alam and reset levels. The licensee via
LER 88-23 Rev. O committed to performing a trend analysis on all
yl containment ventilation isolations from January 1,1987 to present.
Based on licensee actions described, this item is closed.
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LER 88-18, Rev. 0 (Closed) "Leakage into the Control Room Emergency
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Ventilation System." Licensee Event Report 88-18 dealt with the
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r licensee's detemination that CB-1, the control room emergency L
i ventilation system, exhibited greater inleakage than that assumed in FSAR l
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safety analyses. The condition was discovered with the plant in cold i
shutdown in the 1988 refueling outage. Region V review of LER 88-10 Rev. O, Indicated that the report inadequately described the event with
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regard to the testing requirements and testing history of inleakage for j
the system, considering the relationship of this event with the l
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identified problems and issues for CB-1 which resulted in NRC escalated
enforcement actions in 1986.
In addition, the LER did not fully
describe the root cause for the observed inleakage, nor why inleakage was j
j observed in 1988 testing and not identified in 1986 testing.
t in a conference call with licensee representatives, regional management
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i and the inspectors discussed the adequacy of the LER and obtained the
l following information.
Following identification of the problems in 1986,
the licensee developed an action plan which in part included smoke testing to determine CB-1 integrity in 1086 and replacenent of filter j
housing access doors in 1988. The licensee detennined that the latter l
modification required a subsequent inleakage test per ANSI standards
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referenced in NRC Regulatory Guide 1.52.
The licensee was committed,
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with exceptions, to R.G. 1.52.
In 1988, inleakage testing was performed l
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for the filter housing.
Based on engineering recomendations, the supply
ductwork located between the filter housing and the CB-1 fans was also
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tested.
In response to questions, the licensee stated that (1) there
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j were no technical specification surveillerce requirenents nor licensee
j comitments, particularly in response to 1986 NRC enforcement actions, to
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j perfom CB-1 inleakage testing following the 1986 outage except in the
case of major n.odifications; (2) the epoxy paint which was used in part
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effective over the current operating cycle; and (3) a review had not been
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performed to determine the details of possible inleakage tests perforired prior to 1986.
The inspectors questioned whether inleakage was measured for CB-1 ductwork on the return lines from the control room envelope to the filter housing. Measurenent had not been perfomed, and a formal evaluation of the potential effect of inleakage on the return lines on control room habitability had not been performed.
in response, the licensee performed pressure tests on CB-1 and completed an engineering calculation to derronstrate that the system had enough margin to accomodate the potential contribution to control room calculated doses from return line inleakage. The inspectors reviewed the calculation and discussed the following observations with licensee engineering representatives:
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the calculation made the assumption that the return lines had integrity equivalent to the suppl performed in 1988; (2) y lines for which actual inleakage measuretrent was the reduction in s documented for future reference; and (3) ystem margin should be the calculation and supporting pressure testing appeared to be reasonable and conservative.
As a result of the atove, the licensee contaitted to revise LER 88-18 by August 5, 1988, with subsequent revisions as needed.
The licensee was ccntinuing to evaluate the event through its event report process. The licensee also committed to perform evaluations to detemine the need for CB-1 inleakage testing during the 1989 refueling cutage for the return line cuctwork.
Current plans included major CB-1 modifications in 1989 which may obviate the need for inleakage testing over various portions of the system, aside from testing required by R.G.1.52 for rajor modificaticns.
The licensee cormitted to reperform the inleakage testing of 1988 in the 1989 outage in the event the planned CB-1 modifications are not perforfred in 1989.
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The inspectors concluoed that Rev. O to LER 88-18 did not fully describe the event, including generic implications and the root cause of leaking ducting.
In addition, the licensee failed to formally evaluate the potentially significant effect of inleakage in the CB-1 return lines, given the low rargin to control room habitability limits. At the July 29 exit meeting, licensee manager %nt stated that the root cause of the events described in LER 88-18 had been determined and that corrective actions had been implemented or scheduled for completion.
Based on the above inspection, LER 88-18. Rev. 0, is considered closed.
Revision I to LER 88-10 is currently under review.
Preliminary review of this revision found probims. Specifically, 1) the LER does not clearly explain PGE justifications for smoke testing, and 2) the LER does not clearly define what is meant by system integrity and hcw to deterrrine
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systen integrity. The inspectors understand that another revision is planned for this LER.
Fire Protection Associated Licensee Event Reports
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The followirg five licensee event reports describe events associated with r(portable cccurrences related to the area of fire protectier. The i
inspectors assessed the thoroughr.ess of the licensee's evaluation of each
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event, the effectiveness of the proposed corrective actions and trended the events. The inspectors noted two areas of improvement.
First, the licensee has lowered the threshold for identifying and reporting fire protection related events. Secondiy, in one instance, a quality inspector observed a deficiency in a required fire patrol while conducting an unrelated cuality activity. The inspectors noted three areas requiring improvement.
First, the licensee does not appear to effectively trend events in the fire protection area.
Each event appears to be treated as an isolated case, therefore deeper prograruratic problems took longer to identify and frequently resulted in repeat events (Specific examples are documented below). Secondly, licensee event evaluations that are attributed to personnel error do not always consider the adequacy of the inplementing procedures or the training given to irplement the procedures. Thirdly, all corrective actions conriitted to in licensee event reports are not always effectively tracked and implemented. The concerns are particularly bothersome in the area of fire protection where the licensee has been placing additional emphasis and was evaluated as a '3' in the previous SALP.
These specific weaknesses will be followed as part of the Appendix R team inspection in August 1988. Further, general weaknesses in these areas will be followed under routine LER inspection (this closed item 87-40-03).
LER 87-29 Rev. O, (Closed) "Fire Watch Not Established Due to Personnel Error." On July 6, 1967, as a compensatory measure for maintenance on fire suppression systems, a fire watch was to have been posted for the remote shutdown panel room and the electrical penetration areas. However, a change in work schedule caused the fire systens to be placed out-of-service earlier than expected. Through persor.ael error because the time to initiate the compensatory fire watch was not coincident with placing fire systems out of service, the system was inoperable without having a compensatory watch. The licensee, as corrective measures, verified Administrative Order 10-2 "Fire Protection' was adequn e, counseled the personrel involved and determined the event to be an isolated incident.
The licensee's evaluation did not assess the adequacy of training.
LFR 87-35 Rev. O. (Closed) "Fire Doors Made Inoperable Due to Personnel Error." On November 18, 1967, the integrity of a Corponent Cooling Water ILU I fire door was compromised because cables and hoses being used for maintenance were rcuted through the fire decr.
The licensee determined the cause of the event to be procedural inadequacy, since the CCW fire door was not labeled as a fire door. The licensee's evaluation did not assess the adequacy of fire protection training. As corrective actions the licensee established a coepensatory watch, added signs to identify the door as a fire door, changed procedures to require periodic checks of fire doors and cermitted to having rollup fire doors closed as a romal position.
Through personnel eversight, the corrective action of establishing rollup fire doors as normally closeo was not implenented and was a contributing cause to a similar reportable event (LER 88-12) on the sane rollup fire door.
LER 88-09, Rev. O, (Closed) "Hcurly Fire Patrols for 1roperable Fire Dcors Missed." On April 12, 1986, es a compensatory measure for opened fire doors during battery replacerent, a roving fire watch was
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established on the fire doors of the east and west battery rooms. On i
April 21,15c8, a quality control inspector, assigned to inspect the i
battery replacement, alertly noted the fire watch was not being conducted
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and notified security. The licensee's evaluation concluded the fire watch had determined that it was not necessary that all of the areas (
associated with the battery replacement needed to be observed by a roving fire watch since other workers were in those areas. The fire watch, however annotated the fire watch log as though the areas had been surveilled. Licensee's corrective actions included innediate posting of a roving fire watch on the missed fire areas, reerrphasizing the importance of performing fire patrols to all those charged with that responsibility and
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discontinuing employrent of the individual involved. The licensee's
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evaluation did not assess the adequacy of fire watch training.
LER 88-12 Rev.0, (Closed) "Fire Door Made Inoperable Due To Personnel Error." On Xpril 30,19E8, the integrity of a Component Cooling Water (CCW) rollup l
fire door was cerpromised because rigging gear, associated with
in-progress maintenance, was routed through the fire door. The licensee
determined the event was caused by personnel error and noted the event was a repeat occurrence. Additionally, the licensee noted a corrective
action to a previous similar event had not been perforced and another
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corractive action had not been effective. As corrective actions to this
event the licensee irrediately removed the rigging and reestablished the integrity of the fire door.
The door was also more effectively labeled t
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j and additional labeling was added to all fire doors. The licensee's
evaluation did not assess the adequacy of fire protection training, j
LER 88-17, Rev. O, (Closed) "Hourly Fire Patrols For Inoperable Fire Barriers
Missed." On June 5,1988, as a compensatory measure for maintenance and l
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constructicn activities, a fire watch was to have been implemented for
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fire barriers in the Electrical Penetration Area and the Service Water
Pump Room. However, due to personnel error, as determined by the licensee's assessments, the fire watch was not posted.
The licensee's
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corrective actions included innediately posting a fire watch, revising i
direction to the security force to clearly define specific duties of each l
assignment and implementing mandatory hourly verification that scheduled
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fire patrols are performed. The licensee's evaluation did not assess the
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adequacy of fire patrol training.
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In Office Review j
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Based on in office review for accuracy of the event report and adequacy of correcthe actions and plans as stated in the LERs, the following
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LERs were closed out:
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LER 86-12
"Control Roori Boundary Degradation Resulting in Inability to f
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Maintain Fositive Pressure in Control Room"
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LER 37-25
"Containment Ventilatico isolation Due to Spuricus Actuation
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LER 87-32
"PCS Wide pange Pressure Transmitters Out-of-Calibration Due
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to Instrument Drif t"
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l LER 87-33
"Main Steam Pressure Transmitters Out-of-Calibration -
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LER 87-36
"Seismic Monitoring Instrumentation Surveillance Missed -
Personnel Error" LER 87-37
"EHC Switch Failure Caused Load Rejection Reactor Tripped Marually" LER 88-08
"Pressurizer Safety Valve Setpoint Found Out-of-Tolerance During Surveillance Testing" LER 88-13
"Cerponent Cooling Water Valve Positions Not Verified as Required by Technical Specifications Surveillance Requirements" LER 88-16
"Steam Generator Water Level Instrument High-High Trip Outside Technical Specification Allowable Values" LER 88-20
"Pressure Mitigation System Actuated Following Inadvertent Letdown Isolation" LER 88-21
"Operational Hode Change made Without Technical Specification Required Surveillance Having Been performed following Maintenance" LER 88-22
"Train A Safety Injection Initiation During Plant Heatup" LER 88-23
"Containment Ventilation Isolation Due to Spurious Spike on PRM-1C" 7.
Follow-up On Previous inspection Findings (92701, 92702, 92703)
Violation (Closed) (50-344/87-24-01) Failure To Pake Timely 50.72 Report."
On June 29, 1987, while conducting a sirulated loss of power test with the plant in Mode 5 and the reactor coolant system at 110 F, a reactor protective system (RPS) actuation (reactor trip breakers opened)
occurred. The licensee did not report the event within the four hour reportability requirements of 10 CFR 50.72.
The licensee respended to the notice of violation on September 11, 1987, acknowledging the violation and stating that operators, because of plant conditions, did not recognize the RPS actuation as reportable.
As corrective actions, the licensee clarified RPS actuation reporting to the operations staff by discussing the event and providing additional written guidance to the staff.
The inspectors verified the adequacy of the written guidance and have, since the violation was issued, evaluated licensee performance in the area o' reporting RFS actuations as adequate.
Follows) Item (Closed) (50-344/87-37-02) ("Emergency Core Coolino System TRTTJ External Leekage Greater lhan Design Allowable."
On septec.ber 27, 1987, while at 65% reactor pcwer, the licensee identified that ECCS external leakage exceeded the design basis value of
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1580 cc/hr via a leak from the boron injection tank (BIT) outlet relief valve. This event raised the following concerns:
(a) management recognition and response to the safety significance of ECCS external leakage (based on the leakage having existed for greater than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following the leak identification);
(b)'implementationofNUREG-0737III.D.1.1(basedonplantwalkdowns that indicated other ECCS external system leakage existed but was not documented or tracked against the 1580 cc/hr criteria); and, (c) the disparity between the FSAR and the as-found design (based on the PSV-8852 relief discharge line routing to the auxiliary building drain system vice the pressurizer relief tank).
The licensee formed an independent review group to address these concerns.
Based on the findings of the independent review group, the subsequent a: tion plan generated by the plant staff and the actions proposed in LER 87-27, this item is closed.
Unresolved Items 86-23-04 and 86-23-05 (Closed) "Battery Testing and Specific Gravities Not Corrected for Level Per Standard Industry Practice." Trojan Technical Specification 4.8.2.3.2 requires pilot cell specific gravities to be corrected for temperature only. Standard Westinghouse Technical Specification 4.8.2.1. requires pilot cell specific gravities to be corrected for tempecture and level.
IEEE Standard 450, Appendix A2, states, "The readings (specific gravities)
must be corrected for the actual electrolyte temperature and level". PGE had not been correcting specific gravity for level. After review, PGE detemined that to correct specific gravities for level a baseline would have been needed to have been established during initial battery installation and was not. PGE detemined without the baseline specific gravity correction would be inappropriate. Additionally, PGE replaced station batteries during the 1988 Refueling Outage and following the replacement have been correcting specific gravity for level per Maintenance Procedure PP-1-14. "125 Volt Station Batteries" which was revised to require battery specific gravity to be corrected for level.
The new batteries also provide additional capacity to address previous testing deficiencies.
Finally, PGE creriitted to suteit a license change application request, which will include level correction of specific gravity, when the proposed change to Standard Technical Specifications (now urder consideration) is adopted by NRC.
Violation (Closed) (50-344/08-13-06) "Inadequate Lecking of Feedwater Drain Valves." On March 25, 1988, feedwater drain valves FW-079, 082, 083 ano 086 were found open. The Final Safety Analysis Report assured these valves to be locked closed. Technical Specification 3.6.3.1.
icentifies these valves as manual containment isolation valycs and as such must be locked closed to be censidered operable. The licensee determired the cause of this event to be personnel error to reinstall locking devices. Based on the proposed ccrrective actions in Licensee Event Report 88-04, Rev. I and licensee's Notice of Violation response, this item is close '
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Followup Item 50-344/87-37-04 (Closed) "Subcooled Margin Menitor (SMM)
Operability." The licensee in the October 22,19'7 exit conrnitted to further examine an event in which confusion existed over the operability of the SMB. The licensee's evaluation concluded the training given the operators and shift technical advisors on the Fluke data logger and the SFM could have been more detailed. The training on the SMM and the Fluke data logger has been subsequently improved. Operations has assigned a duty general manager to improve management involvement with operability concerns.
This item is closed based on licensee corrective action.
Temaorary Instruction 2515/93 (Closed) "Inspection for Verification of Quality Assurance Request Regardino Diesel Generator Fuel Oil Multi-Plant Action Item A-15."
The inspectors verified diesel generator fuel oil had been added to the areas governed by the licensee's quality assurance program. Diesel generator fuel oil is on the licensee's Q-list and fuel oil quality is verified whan received.
Open item inoffice Review The following open items were reviewed. These open items were reviewed considering safety significance and licensee responses.
Further, these items were verified with the licensee to be acceptably completed or tracked by the licensee.
Followup items 87-13-04, 05, 06, 07, 08, 09, and 10 on
"Environmental Qualification Program":
87-13-04 - This open item dealt with clarification of QA inspection hold points in the equipment installation process.
In addition to the licensee's tracking and verification, general inspection of such processes indicated that QA holdpoints have been generally understood and verified.
Specific licensee guidance (Installation Guidance E-10 and Revision tc QCP-2-1) had been provided for QA inspection and licensee has cocpleted their action on this item.
87-13-05 - The required modification to EQ files for equipment design life were verified by the licensee to be completed.
j 87-13-06 - The actual work on the associated corrective maintenance and modifications on Limitorque operators was verified to be complete.
87-13-07 - This open item dealt with QA coverage of work on equipment.
Similar to one of the above items general inspection has
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verified that this activity is generally acceptable in addition to
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the licensee's verification. Additional licensee guidance and directions were referenced and the licensee closed their followup of this item.
87-13-08 - Improved forvalization of site EQ training was verified as in progress with the license _ _ _ _ - - _
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e 87-13-09 - The licensee's ccanpletion of improvements to feedwater control and CCW make pumps in accordance with RDC 87-28 were
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verified with the licensee.
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87-13-10 - The inspector verified that the records related to this
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f open item were retrieved and found complete _ by the licensee.
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Followup item 87-18-01 on "Motor Operated Yalve Repair" was verified to l
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be complete.,
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Followup item 87-18-03 on "Feedwater Pipe Restraint Damage" has been
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addressed in licensee action plans, NRR reviews, and routine inspection
activities and the licensee has completed their followup of this item.
Followup items 87-31-02, 07, 09, 10, 11, 14, 16, 17, 18, 19, 20, 21, 22
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and 23 dealing with maintenance will be grouped and selectively followed
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as item 88-38-01 during the upcoming maintenance team inspection.
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f Followup item 87-44-02 "Lifted Leads Control " was verified to have en
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action plan, which included visits to other RV plants and incorporation
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I of additional independent. verification requirements. The licensee has completed their followup.
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l Unresolved item 88-13-05 on "Stop Watches in Calibration Program." The i
Ifcensee committed to evaluate the need for stop watches in the
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calibration program. This item was entered and tracked on the licensee's comitment tracking If st.
The following violations were reviewed and the licensee's responses were considered acceptable to close.
l Violation 87-31-15 on"InserviceTesting(IST)ofDieselGenerator Relief Valves" Violation 88-01-01 on "!ST Data Review" l
Ylo1ations 88-13-02, 03 and 06 on "!ST and Administrative Controls" Violation 88-24-02 on "Work Hour Limits"
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Event Follow-up (93702. 92701)
Inadvertent Safety injection With the plant in Mode 3 during startup from the refueling outage, an inadvertent safety injection occurred on low pressurizer pressure on July 4, 1988.
Instrurentation and control technicians were working to remove the automatic block of the safety injection accumulator outlet valves, associated with permissive P-11, to allow time cycling of the valves and in preparation for reactor coolant system (RCS) leakage tests. One pressurizer pressure channel, associated with PT-456, had a dumy signal to simulate RCS pressure above P-11.
The technicians then proceeded to the pressurizer pressure channel associated with PT-455. The licensee's troubleshooting indicated that an apparent voltage spike in the PT-455
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test instrument loop cleared P-ll, and with actual RCS pressure at approximately 1550 psig, an "A" train safety ihjection actuation occurred. The "B" train did not actuate. Safety injection from the centrifugal charging pumps raised RCS pressure to a maximum of approximately 1830 psig. Operators responded per plant emergency operating procedures and secured safety injection.
The licensee's investigation of the event identified the apparent cause of the "B" train failure to actuate to be response time differences in the solid state protection system input relays for the "A" and "B" trains given the short duration of the voltage spike. Voltage spikes on the test instrument loop for PT-455 were not consistently reproducible and were not found on the other pressurizer pressure channels. The licensee was continuing to investigate the cause of the voltage spika The licensee confimed the operability of associated safety in, t 'n circuitry by completing channel functional tests for the instrument loops and solid state protection system to the output relays for both the "A" and "B" trains.
The Ifcensee has revised the applicable procedure to reduce the likelihood of an inadvertent safety injection when attempting the removal of the automatic block of the safety injection accumulator outlet valves.
The inspectors discuned the event with the operating crew, reviewed relevant chart recorder data, reviewed the alam printer listing and sequence of events record, and reviewed other shift records. The licensee's Administrative Order A0-3-7, titled "Post-Trip Review and Pemission for Reactor Trip Recovery and Pode Changes " required that the Post-Trip Review fom be completed by the shift supervisor and shif t technical advisor following a safety injection actuation to assess plant
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conditions and to identify problems, anomalies, and potential failures of
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equipment to start. As a result of the 1983 Salem ATWS events. NRC Generic Letter 83-28 dealt in part with the need to complete post trip review foms to ensure that the cause and progression of a reactor trip was understood, to improve the capability to detemine the root cause of the reactor trip, and to make available sufficient infomation abcut the reactor trip event so that a decision on the acceptability of a reactor restart can be made. The inspectcrs' review of the completed Post-Trip Review fem identified two concerns.
First, the completed fom was taken from Revision 15 of A0-3-7.
Revision 16 to A0-3-7 was approved on March 3, 1988. Discussions with licensee personnel indicated that when Post-Trip Review forrs were needed, they would be copied from a controlled copy of the procedure or taken from a supply kept in the shift supervisor's office. The latter source had foms from Revision 15 of A0-3-7.
Although there were no significant changes between the Revision 15 and 16 foms, the inspectors consider the possible use of outdated procedures to be a potential concern.
Second, review of the completed Post-Trip Review fom for the July 4, 1988, safety injection event indicated that the form was not filled out completely and was potentially incorrect for one iten. The items which were not filled out ranged from administrative-type itens (e.g.
verification that an Energency Notification System report was ccepleted)
to items such as whether the "A" train containment spray pump, main steam isolation, and feedwater isolation actuated.
The apparent incorrect item
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dealt with the "A" train safety injection pump which was checked as having started, when in fact the pump was apparently in pull-to-lock position and did not start.
The inspectors reviewed the ccepleted Post Trip Review fom for the July 4 event about four hours after event initiation.
Review of the fom by the Operations Supervisor resulted in some improvement in the completeness of the Post Trip Review fom.
Based on discussions with the Operations Supervisor, his review probably occurred on July 5.
The inspectors reviewed Post Trip Review foms for 1987 and 1988 for completeness and use of the current revision of the procedure. A review for accuracy was not within the secte of the review. The inspectors did not observe a pattern or programatic problems. Although a trend was not identified, the inspectors consider the failure to completely ar.d accurately fill out the Post Trip Review fom to be a concern.
The inspectors discussed the concerns with the Operations Supervisor.
The Operations Supervisor stated that the outdated revision to A0-3-7 was imediately discarded upon identification and a check will be made for other procedures which may be subject to the use of outdated revisions and to impose controls as necessary for those procedures. On the completeness of the Post Trip Review, the Operations Supervisor stated he believed this to be an isolated incident; he was aware of the need to keep shift crews familiar with forms; he would discuss this event with shift management; and additional training would be considered in this area particularly when the simulator becomes available. At the exit meeting conducted on July 29, the inspectors discussed the findings and the need to take app opriate corrective actions.
The appropriateness of perfoming a surveillance, the 6bove surveillance test on mode 3 is currently under review by the inspector and will be documented in the next inspection report.
Seal Injection Ff1ter Housing Leakage With the plant in Mode 2 in low power physics testing, an 0-ring on the
"A" seal injection filter housing failed on July 10, 1988, resulting in a spill of approximately 1500 gallons.
Filter replacement, and consequent 0-ring replacement as necessary, had occurred frequently on plant startup. The "A" filter had reportedly been replaced on July 6.
Operators in the auxiliary building responded to the event, identified the source, and isolated the leakage within 27 c'inutes. Contaminated i
plant areas were restricted to three filter pits and an adjoining room I
where leakage entered through piping penetrations and was contained by a floor bem. No plant radioactivity release was attributed to this event.
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Control room operators entered off-ncmal instructions for reactor coolant leakage and for the observed malfunction in the charging line.
Shif t crew rantgement determined entry into the plant radiological emergency response plan was not r.ucessary.
The inspectors discussed the event with members of the crerating crew, reviewed relevant chart recorders and logs, and performed a walkdown of the centaminated area in 9e auxiliary building. A review of process radiation monitor data confirmed that no releases had occurred.
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inspectors discussed the applicability of the plant radiological
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emergency response plan for this event with the Region V emergency
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inspectors will continue to monitor licensee actions in routine i
follow-up.
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No violations or deviations were identified.
9.
Exit Interstew (30703)
The inspectors met with the licensee representatives denoted in paragraph 1 on July 29, 1988, and with Itcensee management throughout the inspection period.
In these meetings the inspectors sumarized the scope and findings of the inspection activities.
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