IR 05000344/1993003
| ML20035E921 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 04/02/1993 |
| From: | Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20035E920 | List: |
| References | |
| 50-344-93-03, 50-344-93-3, NUDOCS 9304200072 | |
| Download: ML20035E921 (14) | |
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i U. S. NUCLEAR REGULATORY COMMISSION
REGION V
i Report No.
50-344/93-03 Docket No.
50-344 i
License No.
NPF-1 Licensee:
Portland General Electric Company 121 S. W. Salmon Street Portland, Oregon 97204 Facility Name:
Trojan Nuclear Plant Inspection At:
Rainier, Oregon Inspection Conducted:
January 21 - March 8, 1993 Inspectors:
K. E. Johnston, Senior Resident Inspector J. F. Melfi, Resident Inspector Approved By:
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P. H./ ohnson, Chief Date Signed React Projects Section 1 Summary:
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Inspection on January 21 - March 8. 1993 (Inspection Report No. 50-344/93-03)
Areas Inspected:
Routine inspection of operational safety verification,
maintenance, surveillance, and followup of previously identified items.
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Inspection procedures 61701, 61726, 62703, 71707, 71710, 92700, 92701, and 92702 were used as guidance during the conduct of the inspection.
Results:
General Conclusions and Specific Findinas:
The licensee maintained required systems and components operable. A safety valve in the spent fuel pool system had to be reset (paragraph 7.a) and a need for improvement in the SFP emergency procedure was identified (paragraph 7.b).
Sionificant Safety Matters:
None 9304200072 930402 PDR ADOCK 05000344 G
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i Summary of Violations and Deviations:
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Open Items Summary:
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i One Information Notice (IN) (Paragraph 9), one enforcement item (Para-l graph 10), and 17 LERs (Paragraph 8) were closed. One Part 21 item i
(Paragraph 11) was examined but remains open.
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DETAILS l
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Persons Contacted a.
Portland General Electric Company i
J. E. Cross, Vice President and Chief Nuclear Officer l
W. R. Robinson, Vice President Nuclear i
- R. D. Machon, Plant General Manager
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G. D. Hicks, General Manager, Plant Support l
C. K. Seaman, General Manager, Nuclear Plant Engineering l
D. L. Nordstrom, General Manager, Nuclear Oversight t
- T. D. Walt, General Manager, Technical Functions
C. P. Yundt, Project Manager, Special Projects A. R. Ankrum, Manager, Nuclear Training J. A. Benjamin, Manager, Quality Control l
L. K. Houghtby, Manager, Nuclear Security l
H. K. Chernoff, Manager, Licensing
M. B. Lackey, Manager, Planning and Control t
- J. M. Mihelich, Manager, Nuclear Plant Engineering
S. B. Nichols, Outage Manager l
- W. O. Nicholson, Manager, Operations l
J. W. Patterson, Manager, Maintenance
S. M. Quennoz, Manager, Technical Services M. Singh, Manager, Plant Modifications
W. J. Williams, Manager, Nuclear Compliance
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G. P. Enterline, Branch Manager, Operations M. G. Cooksey, Supervisor, Maintenance
- D. A. Desmarais, Supervisor, Nuclear Plant Engineering
C. M. Dieterle, Supervisor, Individual Plant Examination i
- J. M. Pedro, Compliance Specialist
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E. W. Ford, Compliance Specialist b.
Oreoon Department of Eneroy i
A. Bless, Resident Safety Manager j
V. Sarte, Resident Inspector
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- Denotes those attending the exit interview.
The inspectors also interviewed and talked with other licensee employees during the inspection. These included shift supervisors, reactor and
auxiliary operators, maintenance personnel, plant technicians and engineers, and quality assurance personnel.
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Plant Status
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At the beginning of the reporting period the licensee was making preparations to defuel the reactor (Mode 6). The licensee finished defueling the reactor at 6:07 p.m. on January 27, 1993.
The reactor remained defueled for the rest of the reporting period.
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-2-On January 27, the licensee submitted an application for a Possession i
Only License (POL), a letter stating their intent to permanently cease (
operation of the Trojan Nuclear Plant. At the end of January, the I
licensee gave approximately 415 employees 60 days notice of their pending
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termination.
j 3.
Doerational Safety Verification (71707)
I During this inspection period, the inspectors observed and examined plant
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activities to verify the safety of the licensee's facility. Observations i'
and examinations of those activities were conducted on a daily, weekly or biweekly basis.
I Daily the inspectors observed control room activities to verify the l
licensee's adherence to limiting conditions for operation as prescribed i'
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.
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On occasions when a shift turnover was in progress, the inspectors
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observed the exchange of information on plant status to determine that
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pertinent information was relayed to oncoming shift personnel.
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Each week the inspectors toured accessible areas of the facility to observe the following items:
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General plant and equipment conditions
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Maintenance requests and repairs L
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Fire hazards and fire fighting equipment l
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Ignition sources and flammable material control e.
Conduct of activities in accordance with the licensee's i
administrative controls and approved procedures f.
Interiors of electrical and control panels
Plant housekeeping and cleanliness h.
Radioactive waste systems 1.
Proper storage of compressed gas bottles Each week the inspectors conversed with operators in the control room, l
and with other plant personnel. The discussions centered on pertinent
topics relating to general plant conditions, procedures, security, l
training and other topics related to in-progress work activities.
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l The inspectors periodically observed radiological protection practices to
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j determine whether the licensee's program was being implemented in con-
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formance with f acility policies and procedures and in compliance with
regulatory requirements. The inspectors verified that health physics supervisors and professionals conducted frequent plant tours to observe
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activities in progress and were aware of significant plant activities, particularly those related to radiological conditions and/or challenges.
ALARA considerations were found to be an integral part of each RWP l
(Radiation Work Permit).
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Security activities were observed for conformance with regulatory i
requirements, implementation of the site security plan, and administra-tive procedures, including vehicle and personnel access screening, i
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personnel badging, site security force manning, compensatory measures, and protected and vital area integrity.
Exterior lighting was checked
during backshift inspections.
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r The inspectors conducted routine inspections of selected activities of
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the licensee's radiological protection program.
During the course of
inspection activities and periodic tours of plant areas, the inspectors
verified proper use of personnel monitoring equipment, observed
individuals leaving the radiation controlled area and signing out on
appropriate RWPs, and observed the posting of radiation areas and i
contaminated areas. The involvement of health physics supervisors and
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engineers and their awareness of significant plant activities was assessed through conversations.
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No violations or deviations were identified.
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4.
Maintenance (627031 The inspector observed work on the service water (SW) manual isolation valve (SW-158) to the B emergency diesel generator (EDG) jacket cooling
water and combustion air coolers.
SW-158 is a butterfly valve with a i
soft liner seat. SW-158 has a manual actuator using a steel worm and a
cast iron worm gear to rotate the butterfly valve.
Operators isolated SW-158 as part of a clearance boundary for a mainte-f nance outage on the B EDG. After completing this outage, operators tried
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to open SW-158 and could not. The licensee used Maintenance Request (MR)
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90-4175 to remove the actuator and see if the valve was stuck. Mainte-nance personnel removed the actuator and found that the valve would not
rotate, even with 450 foot-pounds of torque on the stem.
Maintenance personnel disassembled the actuator and found that the actuator worm gear had several missing teeth.
Since the butterfly valve
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was stuck closed and had to be removed, the licensee initiated MR 93-00186 to remove and repair the valve. The licensee believes that a i
previous overhaul on this valve was inadequate.
i The inspector and licensee found that the configuration for this valve was incorrect.
The inspector also questioned the adequacy of some of the licensee's previous corrective actions with similar actuators.
The details found during the licensee's work were:
The valve liner was lubricated to the valve body instead of glued in
place. The licensee believes that the valve was stuck due to pinching of the valve disc and valve liner.
The upper bushing and two 0-rings were not installed.
- The coupling nut (connecting the actuator to the valve) should have
been rotated to a different position to prohibit the stops from hitting the gear tooth.
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Worm gear teeth in the actuator were missing due to a travel stop
setscrew misadjustment. The end of the setscrew touched the worm gear tooth and not the worm gear body.
Since the worm gear is made of cast iron and is brittle, the teeth snapped off due to excessive force.
The handwheel had open arrows in the counterclockwise and clockwise
directions, although the valve should only be opened in one direction.
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The inspector reviewed previous maintenance work instructions and the vendor's work instructions and found that the instructions were not specific enough to ensure that this valve would be installed properly.
The vendor directions were generic, allowing for four different actuator
installation configurations and for lubricating or gluing the valve liner in place.
Following the vendor's guidance and the licensee's maintenance practices appeared to be the appropriate course to repair the valve and
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put it in its as-found configuration.
If the valve actuator configura-tion was not correct it could lead to similar problems, which can be seen in the MR history.
The licensee verified proper setup of SW-158 in the shop before it was installed in the field. This included verifying the correct positions for the travel stop setscrews and the coupling nut.
The licensee also modified the handwheel indication to have the valve open in only one direction.
In reviewing MR history for SW-158, the inspector found that the actuator had failed three times in the last six years by the gear teeth breaking off. Since this valve is rarely operated, this is a high failure rate.
Further review showed that the licensee had generated previous corrective action documents (i.e., Event Reports, Nonconformance Reports, Corrective Action Requests) documenting similar problems with this type of valve and actuator.
The licensee's review indicated that there were only three valves with this model actuator, and two of these valves had had problems with the teeth breaking off. The other valve was SW-140, the service water booster pump (SWBP) outlet isolation, which had warning tags.
Previous corrective actions for this model actuator were to install a different model actuator. While this would solve the problem, the licensee has not l
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installed new actuators for SW-140 or SW-158.
Due to problems with the newer model actuators, the licensee now intends to use the old actuators.
The licensee contended that these actuators are now configured properly.
The inspector noted that better maintenance instructions regarding assembly of this actuator could be appropriate, along with additional operating instructions specifying the valve configuration and cautioning operators about the potential for valve damage.
The licensee committed to provide better work instructions for these two valves to prohibit their being installed incorrectly, and expected that this would avoid their susceptibility to damage during normal operation.
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i Both SW-140 and SW-158 are manual isolation valves used to isolate parts of the service water system for maintenance. Other automatic isolation valves could have been used to isolate these portions of the SW system.
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Due to the licensee's decision to close the plant, the actions appear
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appropriate.
I No violations or deviations were identified.
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Surveillance (67101. 61726)
The inspector observed portions of Temporary Plant Test (TPT) 421, " Post Shutdown Heat-up Rate Test of Spent Fuel Pool." The licensee performed
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this test to compare the actual heat-up rate to calculated heat-up rates.
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The instrumentation used was in calibration and the data were evaluated by plant system engineers and nuclear plant engineers.
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The licensee maintained the spent fuel pool (SFP) pumps running, but cut off cooling flow to the SFP heat exchangers. The test results showed
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that the pool heated up at a rate of about 2.5 degrees F/ hour.
The inspector reviewed two engineering calculations, TC-719, " Spent Fuel Decay Heat Generation," and TC-720, " Trojan SPnt Fuel Pool Heatup and Boil Down Rates." The licensee had calculated a heatup rate of 4.1 degrees F/ hour. The test was performed on February 16, 99 days following the November 1992 shutdown. The differences between the calculated and observed heatup rates were due to underestimating heat sink and evapora-
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tion losses. Using these conservative calculations, the water makeup rate was approximately 16 gallons / minute and the time to boil in the SFP was 27.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The water makeup rates will continue to decline and the time to boil will continue to increase, since the fuel heat generation
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rate will continue to decline. The inspector determined that the test was performed acceptably.
The licensee is considering performing a SFP heat-up test without SFP pumps running.
This test will show if natural convection will occur in the pool and determine the maximum the pool would reach without forced circul ation. The purpose of this test would be to determine the continued necessity of active SFP cooling.
No violations or deviations were identified.
6.
Event Follow-up (93702. 62703. 92701)
a.
Fuel Bundle Damace
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During defueling, the licensee observed that fuel assembly F18R appeared to have one damaged rodlet. The licensee noted an apparent hole in the cladding of one rodlet above the sixth grid strap, and
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an apparent missing section between the seventh and eighth grid straps.
This assembly came from location K8 (near core center).
The licensee initiated Corrective Action Request (CAR) 93-0011 to
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investigate this condition. The initial observations were made in the refueling cavity using a black-and-white underwater camera with
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poor lighting. The licensee looked at this assembly with better i
lighting in the spent fuel pool and determined that the fuel clad-
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ding was not missing, but was discolored due to zirconium hydriding.
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This conclusion was also consistent with reactor coolant system (RCS) chemistry, which did not show gross defects in the cladding.
j The licensee reconstituted this assembly in 1989 from assemblies that experienced the core baffle jet impingement phenomenon in 1984.
t Although scme of the fuel rodlets in the assemblies which had
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experienced " baffle-jetting" had cladding failure, others were
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determined to be acceptable for re-use. All the rodlets used in i
this reconstituted fuel reassembly were ultrasonically tested before insertion into this reconstituted assembly. Fuel assembly F18R was used in cycle 13 and the last cycle (cycle 14).
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The licensee determined that this was not an LER or a vendor
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I deficiency (Part 21) report, since the fuel clad had remained substantially intact after full use. The inspector found the licensee's review of this issue to be acceptable.
b.
Service Water Booster Pumo (SWBP) Failure On February 15, 1993, operators noticed noise coming from the A SWBP; its bearings were hot and its mechanical seal was leaking.
The licensee declared the pump inoperable and disassembled the pump.
After pump disassembly per Maintenance Request (MR) 93-0017, the
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mechanic found the impeller washer missing. Without this washer,
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the impeller slid over the lockscrew and came in contact with the
pump volute. The licensee initiated Corrective Action Request (CAR)
l 93-0022 to address this problem.
l The licensee concluded that the washer had corroded away and/or
broken into parts after being corroded.
Licensee review of mainte-
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nance history revealed that this pump was last disassembled in 1980.
The other three pumps had been overhauled since 1980, in June 1987
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(D SWBP), January 1989 (C SWBP) and August 1989 (B SWBP)
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maintenance records for these overhauls were poor and the licensee
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could not determine from their review the condition of the impeller washers or if, when and with what they were replaced. The licensee speculates that the washer for this pump was replaced with a differ-ent, non-corrosion resistant material in 1980. The original washers were of stainless steel (A410), which is corrosion resistant. The
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licensee did not have historical documentation showing replacement i
for the washers. The replacement A SWBP washer was stainless steel.
The inspector questioned why this pump was not overhauled for 13
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years. During discussions with the inspector, the Preventive i
Maintenance (PM) supervisor stated that these pumps were not in the
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PM program until January 1992. The A SWBP was scheduled to be
overhauled during the 1993 outage.
The licensee concluded that it was acceptable not to inspect the other pumps immediately, since the other pumps were repaired more recently. Therefore, any washer corrosion, if present, would still
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be significantly less than the A SWBP. The licensee also noted that
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l concurrent failures of the other SWBPs is not likely. The inspector reviewed the licensee's. technical specifications, which indicate that if all SWBPs failed, the diesels would be inoperable. The actions required by the licensee in that unlikely event would be to l
not move any spent fuel.
To verify that the other pumps do not have a similar problem, the
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l licensee intends to overhaul the D SWBP by June 1, 1993.
Based on I
the results of this inspection, the licensee will decide if they l
need to inspect the other pumps.
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The licensee did not find the missing impeller washer. The licensee then evaluated the potential impact of the washer on the components downstream of the pump and decided that it would pass through the large bore piping and might cause flow restrictions in a heat exchanger. The licensee intends to monitor flow to the affected heat exchangers to see if any blockage exists.
Based on the shutdown of the plant and the decreased accident loads on the heat exchangers, this action appeared to be appropriate.
No violations or deviations were identified.
7.
Spent Fuel Pool (SFP) System Walkdown (71710)
The inspectors walked down accessible portions of the SFP system and evaluated procedures related to SFP operation. The walkdown was generally acceptable, with concerns noted by the inspectors regarding relief valves and procedures.
The walkdown showed that:
Hangers and supports were aligned and working properly.
- l Housekeeping was satisfactory and no flammable materials were
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Valves were properly labeled and not leaking excessively.
Further,
telltale drains underneath the SFP liner did not show any leakage.
Instrumentation for the SFP was operating and in calibration.
- The licensee performs surveillance on the emergency flow path of
service water to the SFP.
a.
SFP System Pelief Valves The SFP system has three relief valves, one on each SFP heat ex-changer (PSV-5277 and -5278) and one on the refueling cavity suction line (PSV-5279). The inspector noted the following concerns:
The SFP heat exchanger relief valves had not been tested since
plant startup.
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The setpoint for the relief valve on the refueling cavity
suction line was incorrect.
The licensee had made commitments to the NRC previously (see Inspec-tion Report 50-344/92-03) to test all relief valves in the plant on a scheduled basis. These relief valves were scheduled for surveil-lance testing by 1994. Based on plant status, the licensee may reevaluate this commitment for testing all their relief valves.
However, the licensee intends to test these valves in the future.
The design pressure rating of the refueling cavity suction line is 150 pounds per square inch (psig). The nominal setpoint for the
refueling cavity suction line relief valve (PSV-5279) was a corresponding 150 psig. The inspector found that the setpoint for i
PSV-5279 was incorrect since the static head on the valve discharge
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was not considered. This static head pressure was 18 psig, and to
keep the system piping below 150 psig, the valve setpoint should have been 130 psig.
The inspector discussed this issue with the system engineer and a maintenance engineer. The system engineer found records that demonstrated that the licensee had previously identified this setpoint difference on a Request for Evaluation (RFE) (9/11/84), and had initiated Plant Setpoint Change (PSC) 85-01 to change the setpoint to 130 psig.
To change the setpoint, the licensee initiated MR 85-0722 (dated 2/12/85). This MR was canceled on October 15, 1985, since it was believed that a design change was necessary to remove the valve.
No design change to the plant was initiated. The need to change the setpoint was not communicated effectively within the plant.
This valve subsequently developed a leak and ultimately became a radiation protection department concern. The licensee installed freeze seals on the piping to the valve and replaced it.
This last maintenance on the valve (MR 88-4228 completed 5/15/90) left the
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valve at 150 psig.
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Based on the inspector's concerns, the licensee generated RFE 93-001. This evaluation of the piping and valves affected by this relief valve showed that the design ratings for these components was not exceeded.
Further licensee review indicated that the design purpose of this relief valve was to prevent possible overpressure
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from the residual heat removal (RHR) system.
Since the RHR system i
is no longer pressurized to the valve's lift setpoint, the licensee does not intend to reset this relief valve.
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and projected status of the plant, this appeared appropriate.
b.
ONI for SFP System Trouble l
The inspector reviewed Off-Normal Instruction (ONI) 4-4, Revision 1,
" Spent Fuel Pool System Trouble /High RWST Temperature," and walked j
down the steps necessary to establish service water system (SWS)
i makeup to the SFP.
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The Final Safety Analysis Report (FSAR, Amendment 19) Section 9.1.3,
"SFP Cooling and Demineralizer System," stated that a seismic Cate-gory I makeup water supply from the SWS was available to the spent fuel pool. The normal SFP cooling system is not seismically quali-fied and does not receive power from safety related power supplies.
The FSAR recognized that in the event of a loss of SFP cooling, the SFP could reach boiling. Operators would manually align service water (SW) and SFP cooling valves to provide water to compensate for evaporation losses. ONI 4-4 provided these instructions. The inspector determined that SW could be aligned to the SFP.
However, the inspector noted the following weaknesses during his review:
The only SFP level indication in the control room was provided
by a high/ low level alarm. Actual level can only be determined at the SFP. ONI 4-4, which covered actions for a low SFP level, did not direct operators to the SFP to determine actual level and the rate of decrease.
ONI 4-4 did not address actions to be taken if it was
determined that the SFP was losing inventory during normal operations. Actions sech as isolating the SFP cooling system or assessing the SFP leakoff lines would seem appropriate.
I ONI 4-4 did not provide the location of the manual valves which
j would need to be operated to establish SWS makeup to the SFP.
ONI 4-4 did not provide guidance regarding the approximate
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l makeup which would be needed if the SFP reached boiling and how to control SFP level over the long term.
A fire water hose station and a temporary demineralized water
l manifold were located near enough to the pool to be used as i
alternate makeup water sources. These sources were not included in ONI 4-4.
ONI 4-4 did not provide SFP level information, such as the high
and low alarm setpoints and the level of the siphon breaks.
The inspector discussed these findings with the operations manager.
The operations manager noted that he was generally aware that ONI 4-4 lacked specificity. He stated that he had already requested operations personnel to review the procedure for improvements.
He recognized that in the defueled condition, ONI 4-4 would be the relevant emergency procedure.
He stated that the procedure review would be coordinated with the defueled safety evaluation which was underway. The operations manager stated that the inspector's findings would be assessed during their review and that he expected to have ONI 4-4 revised by April 30, 1993.
The inspector will review this procedure during the course of routine inspection.
No violations or deviations were identified.
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8.
Follow-up of Licensee Event Reports (92700)
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The inspector closed the following LERs based on in-office review and the projected phase-out of the plant. This review determined that the licensee had adequately described the event, determined the root cause, and implemented or identified appropriate corrective actions.
LER Number Title LER 90-01, Revision 2
" Failure to Test Containment Personnel Air
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Locks Equalizing Valve and a Valve Configuration Error Compromised Containment Integrity."
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LER 90-06, Revision 2
"Both Trains of Emergency Core Cooling System Were Inoperable During Mode 3 Surveillance Testing Due to Procedural Inadequacy."
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LER 92-02, Revision 1
" Inadequate Disc Hut Locking Design Results
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l in Residual Heat Removal Pump Discharge Check Valve Failure."
LER 92-14, Revision 0
" Reactor Trip Caused by Failure of a Manual Pushbutton on a Steam Generator Feedwater Regulating Valve Controller."
i LER 92-19, Revision 0
" Inappropriate Changes to the Trojan
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l Scheduling System Data Base Led to l
Staggered Surveillance Activities Not
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Performed as Required."
LER 92-21, Revisions 0 and 1
" Incorrect Installation of the Main Steam Isolation Valves' Air Supply Vent Valves
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Caused Vent Valves To Be Inoperable for Isolating Main Steam."
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j LER 92-22, Revisions 0 and 1
" Personnel Error in the Development of a
Surveillance Procedure Following a 1985 Design Change led to Inadequate Performance of a Surveillance."
LER 92-23, Revision 0
" Incorrectly Reporting that the Under-voltage Surveillances of The 4.16kV Busses Were Complete led to failure to Perform the i
Surveillance Within the Required Interval."
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LER 92-25, Revision 0
" Error in Implementation of a Technical
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Specification Change Results in a Failure To Perform a Surveillance at the Required Frequency."
LER 92-26, Revs 0 thru 4
" Fire Barrier Deficiencies Identified by Ongoing fire Barrier Improvements."
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LER 92-28, Revision 0
" Manual Reactor Trip Due to Main Feedwater Pump Trip on Low Suction Pressure Caused by Loss of Heater Drain Tank Pump Flow."
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LER 92-30, Revision 0
" Inattention to Detail Caused Missed Surveillance on Remaining AC Power Sources
i After Removing Emergency Diesel Generator
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for Maintenance."
LER 92-32, Revision 0
" Personnel Error During Procedure Develop-ment Results in Allowable Residual Heat Removal System Flow Below Value Assumed in Accident Analysis."
LER 92-33, Revision 0
" Failure to Properly Stress Relieve a Steam Generator Tube Sleeve Causes Primary to Secondary Leakage to Exceed Limits."
l LER 92-34, Revision 0
" Deficient Fire Barrier Penetration Seals Result in Technical Specification Violations."
LER 92-35, Revision 0
" Power Operated Relief Valve and Block
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Valve Position Indicator Alarms Channel
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Functional Tests Hot Performed Due to Personnel Error."
LER 93-01, Revision 0
" Personnel Error in Failing to Properly Seal Fire Barrier Penetration Leads To a
Non-Functional Three Hour Fire Barrier
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No violations or deviations were identified.
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Followuo of Information Notices (92701)
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NRC Information Notice (IN) 92-06. " Reliability of Anticipated Trio Without Scram (ATWS) Mitication System and Other NRC Reouired Eouipment i
Not Controlled By Plant Technical Specifications (TS) " (Closed)
This item was previously inspected in HRC Inspection Report 50-344/92-24.
The availability of this system then was 97%.
This item remained open pending a licensee evaluation of the equipment to determine if a 97%
availability was sufficient and whether there were any additional items that should be included in the Technical Specifications. The licensee's comparison of standard technical specifications revealed that they did not include the loose parts monitor in their technical specifications.
Based on the projected status of the plant, neither the ATWS Mitigation System Actuation Circuitry (AMSAC) nor the loose parts monitor is necessary. This IN is closed.
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Followup of Enforcement Items (92702)
Enforcement Item 92-32-03 (Closed). " Warehouse Access Not Controlled By
l Authorized Personnel."
On November 12, 1992, an Oregon Department of Energy (0 DOE) inspector entered the Trojan Issue Warehouse for a routine inspection.
He found:
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the door open, 2.
the access point unattended,
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3.
the gate to the parts storage area open.
The ODOE inspector informed the control room that the warehouse did not appear to be controlled as required by Trojan Plant Procedures (TPP)
16-12, " Material Storage and In Storage Inspection and Checks." On November 15, 1992, the ODOE inspector found the issue desk unattended.
Due to the repeat nature of these events and historical events showing that this procedure was not followed, the NRC issued a corrective action
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violation.
The inspector reviewed the licensee's January 8, 1993 response to the Notice of Violation. The licensee initiated corrective actions by
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maintaining the issue counter in a docked condition, locking the entrance door if obtaining a part would require an extended length of time, and
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training warehouse personnel on the procedural requirements to maintain access control.
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Spot checks at various warehouses by the inspector verified these corrective actions.
Based on these actions and the closure of the plant, this item is closed.
11.
Followup of Part 21 Items (92701)
Part 21 Report. "Conax Buffalo Corporation Electrical Kit Defect." (00en)
This Part 21 report concerned the validity of an environmental qualifica-tion (EQ) test done on some electrical splices.
PGE is reperforming the
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test and expects the test to be completed by April 1,1993, and the results to be evaluated by June 15, 1993. This Part 21 report remains open pending inspector review of PGE's evaluation.
12.
Exit Interview (30703)
The inspectors met with the licensee representatives denoted in para-graph 1 on March 18, 1993, and with licensee management throughout the inspection period.
During these meetings the inspectors summarized the scope and findings of the inspection activities.
The licensee did not identify as proprietary any of the information discussed with or reviewed by the inspectors during the inspection.
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