IR 05000344/1990002

Insp Rept 50-344/90-02 on 900101-0210.Violations Noted.Major Areas Inspected:Control Room Operations,Esf Status,Maint Program,Surveillance Program,Special Insp Topic Followup & Review of Periodic Repts

ML20012C711
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	03/05/1990
From:	Mendonca M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20012C706	List: IR 05000344/1990002 ML20012C705 ML20012C710
References
50-344-90-02, 50-344-90-2, NUDOCS 9003230097
Download: ML20012C711 (28)

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REGION V

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Report No.:

50-344/90-02

' Docket No.:

50-344 License No.:

NPF-1 Licensee:'

Portland General Electric Company 121 S.W. Salmon Street Portland, OR 97204

. Facility Name:,Trcjan

' Inspection at:' Rainier,- Oregon

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' Inspection conducted: ~ January l'- February 10, 1990

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Inspectors:

R. C. Barr.

Senior Resident-Inspector J. F. Melfi

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Resident Inspector Approved By:

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M. M.:Mendonca, Chief.

Date Signed

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JReactor Projects Section 1 Summary:

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Ihspection o'n January'l - February 10, 1990 (Reporv. 50-344/90-02)

Areas-Inspected:- Routine inspection by the resident inspectors of control

' room operations,_ engineered. safety feature (ESF) status, maintenance-program,

. surveillance program, special. inspection topic follow-up, review of periodic:

reports,:and licensee actions on-previous inspection findings and on reports of defects.

During this inspection, Inspection Procedures-25027, 30702, 1-30703,-40500, 61726,'62703, 71707, 90712, 92700, 92701,:92702, and 93702.

'Safity~IssuesManagementSystem(SIMS) Items:

Closed item BL-87-02.

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General Conclusions and Specific Findings Management _and' supervisory involvement were identified as a concern in

_the implementation of compensatory actions' required for ventilating the

. control room, the administration and control of temporary modifications, and 'the implementation of'the reformatted work control program.

Significant Safety Matters:

None.

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Summary of Violations and Deviations:

Four: cited and one non-cited violations were identified.

l One of these violations resulted from the resolution of entry into

. Technical. Specification 3.0.3 for operational convenience described in inspection report 50-344/88-40.

One violation identified inadequate implementation of compensatory measures to ventilate the control room with inoperable chlorine detectors.

One violation, that has three subparts, identified procedural noncompliances in the administration of-Temporary Modifications.

One violation identified the incomplete update of the Final' Safety Analysis Report after having implemented design changes.

A'non-cited violation was' identified for the failure to perform seismic calculations as described in the Technical Specifications.

'Open Items Summary:

Seven LERs, four unresolved items, seven enforcement items, one 10 CFR 21-

" eport and-Temporary Instrc: tion (TI) 2500/27 were closed.

One open item

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and one unresolved item were identified in addition to the four j

.previously discussed violations.

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DETAILS

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Persons Contacted

• . D. W.' Cockfield, Vice President, Nuclear

'*C, P. Yundt,~ Plant General Manager

• T. D. Walt, General Manager, Technical Functions

• C..K. Seaman, General Manager Nuclear Quality Assurance.

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-D.W. Swan, Manager,TechnicalServices M. J.Singh, Manager, Plant Modifications J. D. Reid, Manager, Quality Support Services J. W. Lentsch, Manager, Personnel Protection

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D. R. Swanson, Manager, Nuclear Safety Branch J. F. Whelan, Branch Manager, Maintenance

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J. Mody, Branch Manager, 21 ant Systems Engineering D. L. Nordstrom, Branch Manager, Quality Operations A. N. Roller, Outage Manager J. P. Fischer, PM/ A Branch Manager G. L. Rich, Branch Manager, Radiation Protection

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" R. A. Magnuson, Branch Manager, Security E. F. " Petersen, Brancil Manager, Maintenance J. A. Reir,aart, Branch Manager, Operations R. N. Prewit, Supervisor, Quality Systems R. L. Russell, Assistant Operations Supervisor R. A. Reinhart, Instrument and Control Supervisor J.-A. Benjamin, Supervisor, Quality Audits

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J. D. Guberski, Nuclear Safety and Re

• W. J. Williams, Compliance Engineer. gulation Department Engineer C. H. Brown,-I&C Procedure Upgrade Manager 5.<

The inspectors also interviewed and talked with other licensee employees

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during the course of the inspection.,These included. shift supervisors, reactor and auxiliary operators, maintenance personnel, plant technicians and engineers, and quality' assurance personnel..

• Denotes those attending t'ie exit interview on February 26, 1990.

2.

Plant Status

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The facility operated at 97% until January 12, 1990, when, after resolution,of instrument uncertainty concerns, reactor power was increased to 100% on January 14, 1990.

The facility continued to operate at 100% throughout the remainder of the inspection period.

3.

Safety Verification (71707)

Operational Safety Verification During this inspection period, the inspectors observed and examined

. activities to verify t.he operational safety of the licensee's facility.

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The observations and examinations of those activities were conducted on a

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daily, weekly or biweekly basis.

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Daily _the inspectcrs observed control room activities to verify the

. licensee's adherence to limiting conditions for operation 6s prescribed in the facility Technical Specifications.

Logs, instrumentation, recorder traces, and other operational records were examined to obtain

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.information on plant conditions, trends, and compliance with regulations.

On occasions-when a shift turnover was in progress the turnover of informationonplantstatuswasobservedtodetermlnethatpertinent

information was relayed to the oncoming shift personnel'.

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Each week'the inspectors toured the accessible areas of the facility to observe the following items:

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(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.

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(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.

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(g)

Implementation of the licensee's physical security plan.

(h) Radiation protection-controls.

(i) Plant housekeeping and cleanliness.

(j) Radioactive waste systems.

-(k) Proper storage of compressed gas bottles.

Weekly, the inspectors examined the licensee'.s equipme'nt clearance control with respect to removal of.cquipment from service to determine that the licensee complied with technical specification limiting conditions for operation.- Active clearances were spot-checked to ensure

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- that.their issuance was consistent with plant status and maintenance

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evolutions.

Logs of jumpers, bypasses, caution and test tags were i

examined-by the inspectors.

L Each week the inspectors conversed with operators in the control room, and'with other plant personnel.

The discussions centered on pertinent-topics relating to general plant conditions, procedures, security,

training and other topics related to in progress work activities.

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'The' inspectors examined tne licensee's nonconformance reports (NCRs) to l

confirm that deficiencies were identified and tracked by the system.

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L-Identified nonconformances were being tracked and followed to the

completion of corrective action.

Routine inspections of the licensee's physical security program were performed in the areas of access control, organization and staffing, and l

detection and assessment systems.

The insaectors observed the access

~ control' measures used at the entrance to t1e protected area, verified the-

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integrity of portions of the protected area barrier and vital area barriers, and observed in several instances the implementation of compensatory measures upon breach of vital area barriers.

Portions of the isolation zone were verified to be free of obstructions.

Functioning of central and secondary alarm stations (including the use of CCTV l

monitors) was observed.

On a sampling basis, the inspectors verified

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thd A tequired minimum number of armed guards and individuals-

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auis,orized to direct security activities were on site.

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The inspectors conducted routine inspections of selected activities of the licensee's radiological protection program.

A sampling of radiation work permits'(RWP) was reviewed for completeness and adequacy of.

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information.

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 RWP s, and observed the-posting of radiation areas and contaminated areas.

Posted radiation

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levels at locations within the fuel and auxiliary buildings were verified

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using both NRC and licensee portable survey meters.

The involvement of health physics supervisors and engineers and their awareness-of

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significant plant activities was assessed through conversations and review of RWP sign-in records.

The inspectors verified the operability of selected engineered safety i

. features.

This was done by visual verification of the correct position j

of valves,l condition of equipment, as applicable. availability of power, cooling w and genera i

Differences Identified between the Plant and Final Safety Analysis Report (F5AR)

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During these walkdowns while comparing actual installa' tion of instrumentation and controls, the inspector identified the following

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differences from descriptions contained in the FSAR.

Specifically:

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1)

FSAR section 6.3.5.1 states that:

" Boron injection tark pressure is

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N indicated in the control room.

A high pressure alarm'is provided."

There is no boron injection tank pressure indicator or annunciator in the control room.

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FSAR section 9.2.2.2.2.3 states, in part, regarding the solenoid valve automatic controls used to control the_ Component Cooling Water

(CCW) system surge tank pressure that:

" Manual push-button controls.

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for the solenoid valves are installed on Panel C18'in the control

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room-near the surge tank level and pressure indicators to allow operators to manually adjust tank pressure to bring levels back into the normal control band during operation of the makeup pumps,!as required." The controls for these valves are not push-button, but switches.

3)

FSAR section 7.3.1.1.1 stated that certain channels are available for monitoring in Containment conditions in the post Loss of Coolant Accident (LOCA) period.

Specifically, the FSAR states that:

"The Containment air coolers cooling water exit temperatures are indicated and recorded in the control room." These temperatures are indicated but not recorded in the control room.

Items 1) and 2) appeared to result from design changes not being properly

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reflected in the FSAR.

Item 1) was removed via Request for Design Change (RDC)83-051, Design Change Package (DCP) 3 in 1985.

Item 2) was changed

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via RDC 76-068,- DCP 2, Revision 1 in 1983.

The licensee is required to update the FSAR at least annually to reflect these changes, in accordance with 10-CFR 50.71 (e). This is an apparent violation (50-344/90-02-01).

Item 3) differs from the actual plant configuration.

The licensee has submitted a change to the FSAR to correct t11s item, therefore, enforcement is not appropriate in accordance with the regulations.

The inspector also noted instances where the FSAR description implied

something different than what was actually installed.

Specifically section 6.3.5.2.2 stated that:

" Safety in pressureisindicatedinthecontrolroom.jectionpumpdischargeheader Further, section 6.3.5.3.2 states that:

" Flow through the safety injection pump header is indicated in the control room." The' text from these two statements in the FSAR implies a single flow and pressure indicator off of the common cold leg injectionline.

The pressure or flow through the single discharge header into containment are not indicated in the control room.

Flow and pressure off of each pump are indicated in the control room.

Another. inconsistency between the plant and the FSAR was also noted in

'section 7.3.2 of the FSAR ' Analysis' for the Engineered Safety Features ActuationSystem(ESFAS).

Section 7.3.2.1 is the FSAR evaluation of compliance to IEEE 279, " Criteria for Protection Systems at Nuclear Power Plants." Section 7.3.2.1 is mostly concerned with the ESF protection racks.

Subsection 7.3.2.1.5 is entitled, "Caaability for Sensor Checks and Equipment Tests and Calibration." Sub-Su3section 7.3.2.1.5.6 of the FSAR is titled, " Periodic Maintenance Inspections." This section states that typical periodic maintenance procedures (for-ESFAS) include the-

.following:

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Check cleanliness of all exterior and interior surfaces.

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Check all fuses for corrosion.

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Inspect for loose or broken control knobs and burned out indicator lamps.

4)

Inspect for rust, moisture and condition of cables and wiring.

5)

Mechanically check all connectors and terminal boards for looseness, poor connection or corrosion.

6)

Inspect the components of each assembly for signs of overheating or component deterioration.

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Perform complete system operating check.

In discussions with the I&C supervisor, procedures require that only item

~7) be performed for the protection systems.

The I&C supervisor stated that Maintenance Procedures (MP) 2-5, " Electrical Analog Instrumentation," and 2-6, " Miscellaneous Analog Instrumentation," have some of the same directions included in them, but that there were no explicit instructions to look at protection systems this way.

The I&C supervisor stated that these two procedures might be included as a

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reference in their normally scheduled PICTs.

The I&C supervisor was also not aware this section of-the FSAR required the other six items.

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As noted in-inspection report 89-24, the licensee had various problems

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with the tightness of connections in the ESF system, which may have caused plant transients.

The licensee was apparently not performing the typical maintenance procedures as separate procedures as described in the FSAR.

The inspectors will followup under routine inspection activities.

One violation and no deviations were identified.

4.

Engineered Safety Features (ESF) System Walkdown (71710)

The' inspector walked down the Emergency Diesel and various support systems for the diesel (Diesel Fuel Oil system, Diesel Air Start system, Diesel Cooling Water system, Diesel Lube Oil system).

The. inspector also walked down the control room indication for these systems.

-The diesel was found to be clean and maintained.

The diesel and diesel comments / observations:. place.The inspector had the following support systems:were in 1)

Diesel monitoring. instrumentation wires associated with panels K106Al, K106A2, K106B1 and K106B2 were not gathered into wrapped bundles or attached to the fixed surfaces.

This presented a greater likelihood of having the wiring accidentally damaged. ' The inspector shared this observation with the system engineer who acknowledged the observation.

2)

LS4905A'on the diesel fuel oil tank had a leaky root valve. When informed,-the licensee generated MR 90-1080 to fix the leak.'

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Design documentation that addressed the worst case design condition for net positive suction head (NPSH) for the diesel fuel injection

pumps was not readily available.

Specifically, as identified in the FSAR, the NPSH for the diesel fuel injection pumps is the gravity head of the fuel in the day tank minus any head loss due to piping configuration.

The day tank outlet is approximately one foot off the floor.

From a review of Calculation T89-01, the entire tank is 99 inches tall, and a tank level of ~83.5 inches is used to meet the TS limit of 1370 gallons.

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volume from the tank outlet (zero % level)-to a height of 83.5 inches.

Due to the length, size, valves, connections, and strainers in the run of pipetothefarthestfuel? injectors,andsinceoneofthefarthestfuel injectors is approximately f1ve feet off the floor, the inspector was concerned about Net Positive-Suction Head (NPSH) requirements for this pump and informed the licensee of his concerns.

The licensee researched initial plant startup records and could not determine if the volume of the Day Tank had'been tested down to the bottom (zero level).

The startup test records indicated that the tank was sized sufficiently to run the diesel for four hours, but did not specify the-starting and ending points of the diesel fuel oil level in l

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r the tank.

The' licensee then performed an analytical engineering calculation to'see if the NPSH required was less than the actual'NPSH at

rated flow through the pump.

The inspector reviewed the licensee's calculation and assumptions and found them to be acceptable. :It was determined by the licensee that the pump could go down to approximately 10.2 psia.

The calculation showed in

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t the worst case, the pump had ap3roximately 11.7 psia available.

Due to

.not having historical data for low much one of the filters could be

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clogged, the licensee assumed a number that appeared reasonable due to the amount of times the filter had been changed (annually) and the run time that actually occurred on the diesel.

No violationc or deviations were identified.

5.

Maintenance (62703)

Turbine Discharge Valve Gasket Replacement

~0n January 4, 1990 the inspector observed portions of the replacement of.

the packing on Auxiliary Feedwater (AFW) discharge valve FW-120.

The-work was performed under Maintenance Request (MR) 89-7345,. revision 1.

'The work was performed to the instructions on the MR, and the system was tagged out.

The replacement was satisfactorily performed by trained craftsmen.

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Steam-Generator Blowdown Valve Thermal Overload Replacement

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On January 9, 1990, the inspector observed portions of the replacement of the the thermal overloads on Steam Generator Blowdown valves, M0-6716, e.

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L-6717, -6718, and -6719.; The work on these valves was performed under Maintenance Requests (MRs)89-462, 89-850,89-851, and 89-8542

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respectively.

These MRs'were written to address concerns' raised in

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Non-Conformance Report (NCR)-89-109, Revision ~2.

The blowdown system was isolated and tagged out in accordance with facility Administrative Orders (A0s).

The Motor Control Centers (MCCs)

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L to theseLvalves were de-energized.

Prior to: replacing the~ overloads on-

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l these valves, the valves upstream of the valves worked were stroked, and i

electrical: current traces were taken for engineering evaluation and

trending.

The tracer and digital voltage meters used were noted to be in f

their current calibration cycle.

The licensee replaced the installed Thermal Overload (TOL) T-45'(14.7 l

amp), with the lower amperage TOL T-40 (9.84 amp).

The new TOL had the material certification with the work package.

l On observing the post maintenance testing on the first TOLs installed, the first time the licensee performed the test, the TOL did not tri).

The-testing involved feeding a predetermined current (38.25 amps) tirough the TOL.

The' electricians, who performed the test, realized that.they

~should-have pulled the fuse in the circuit since the fuse provided another path for current to flow.

Directions to pull fuses were not included in any of the MR work instructions or in the referenced

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Maintenance Procedure (MP) 1-7, "480 Volt Motor Control Centers and

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Molded Case Circuit Breakers." Additionally the electricians verified thatthetripshadreset;thisalsowasnot.IntheMRinstructionsorMP

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The inspector discussed with the Maintenance Manager the subjects of MR-work instruction adequacy, and craftsman continuing work with inadequate work instructions.

As a result, the Maintenance Manager generated-Non-conforming Activity Report P90-060M, discussed this event with his

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staff, initiated a change to procedure MP 1-7, and reviewed expectations with craft' personnel.

The licensee's actions to correct these concerns

appeared acceptable.

The licensee waited for the TOL to cool down and reset, pulled the fuse,

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and reperformed e w test.

The TOL tripped within its acceptance band

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(13-90 seconds).

The TOL trip settings were:

Valve Number Phase A Phase B Phase C MOV-6716 29.14 27.97 30.01 M0V-6717 29.11 29.17 26.18 MOV-6718 33.46 36.02 29.83 MOV-6719 43.81 37.61 25.27'

The inspector also assessed the calculation used in si, zing the thermal overloads.

This calculation, TE-132, used guidance provided by the NRC

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in. Regulatory Guide 1.106 and in NUREG 1296.

This guidance states that the temperature of the overload and the motor should be considered.

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~ assumptions were conservative in that the ambient temperature-assumed at the' breaker was'104 degrees F.

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w Trojan Work Control System On 0ctober 24, 1989, Trojanmanagementestablishedaplantomodifythe

.Trojanworkcontrol-process.

Control of plant work included the I

. identification, planning, scheduling, maintenance (preventative or corrective), testing, and return to service of plant equipment.. Prior to this modification, Administrative Order (AO) 3-9, " Maintenance Requests,"

was the top tier instruction that. described how maintenance was conducted at Trojan.

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Action Plan (AP)89-052 was developed to identify-and track discrete actions required to implement the new work control process. The plan includes visits to facilities noted to have successful maintenance control programs, and resource needs identification and implementation.

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A subelement of AP 89-052 was a change in structure of the daily planning meeting.

Previously, the daily planning meeting was conducted at 8:30

• 7 a.m. each morning!and was attended by a variety of plant managers.

The meeting generally discussed events that occurred the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and= major. activities that were planned for the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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. format includes two meetings each-work day.

The 9:00 a.m. ' Planning Meeting' reviews and plans plant work tasks over a three day period.

Participants at this meeting are generally first and second line

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supervisors.

The 3:30 p.m. ' Management Meeting' ensures management maintains an overview of plant activities to assure plant problems are

resolved in a timely manner.

Participants at the ' Management Meeting'

are generally Branch managers and above.

The change in format to the planning meeting occurred in January.

On February 12, 1990, the inspector attended both the Planning Meeting and the Management Meeting.

Additionally, after the meetings the inspector discussed tM format and content change of the meetings with a

,e sampling of the attendees. The inspector made the following observations:

A specific format for the meetings did not exist.

Maintenance

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tasks discussed were not' discussed in the formatted order as

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presented in the Plan-of-the-Day.

As a result, participants

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were sometimes confused as to the tasks being discussed.

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A list of expected attendees for the meetings did not exist.

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L representative from Quality Assurance (QA) was not present to address questions in the QA area.

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Sequenced maintenance tasks that involved coordination between

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j two or more disciplines were not statused.

In one instance,

ll this led to a late, unanticipated change in planned work.

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.Two attendees at the Planning Meeting were not aware of a

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planning change regarding which safety train would be involved J

in an outage for the following day.

As a result, shops had to reschedule planned preventative maintenance and surveillance l

activities for the following day.

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gained-the impression that the change in format of the meetings was seen-

.as an improvement over the previous format.

The supervisors felt they

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to manage their resources.

They also indicated that improvements such as l.

a standardized format, a current plan-of-the-day and a standardized list l

of attendees were needed.

The ins?ector also learned that tranagement.

involvement in and. assessment of t1e effectiveness of the ' Planning-Meeting'. had been very limited.. Since the beginning of the two-meeting

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format,-the responsible manager had attended only.several of the planning

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meetings.

l The inspector also reviewed AP 89-052 for timeliness of implementation.

The plan subelement due dates were being achieved near to or on schedule.

The inspector will continue.to followup during routine inspection.

No violations or deviations were identified.

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Surveillance (61726)

Containment Pressure Transmitters The inspector observed-the licensee conduct Channel-Functional Tests on the containment pressure instruments. These Channel Functional Tests were performed to meet, in part, the requirements of Trojan Technical Specification 3/4.3.2, " Engineered Safety Features Actuation System Instrumentation." The licensee used Periodic Instrumentation and Control Tests (PICTs) 9-1 through 9-4 to check the four Containment Pressure Channels and PICT 9-5 to calibrate the pressure transmitters.

Three channels provide a High Containment Pressure signal (Phase A) and all four provide a high-high Containment Pressure signal (Phase B).

The breakdown of what the PICTs check for is as follows:

Prot. Set PICT Phase A Phase B I

9-1 No Yes II 9-2 Yes Yes III 9-3 Yes Yes IV 9-4 Yes Yes The inspector observed the licensee conduct PICT 9-1 through 9-5.

The surveillance testing was performed by qualified craftsmen using calibrated instruments.

The systems were appropriately tagged out of service.

The testing met the Limiting Condition for Operation (LCO) of the system and the testing met the TS requirements.

During the conduct of the test, there were certain annunciators that were activated signalling that the indicated pressure on that channel was high. -These annunciators were generally acknowledged by the control room

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operator.

After the test was over, the annunciator was reset.

The inspector, on two occasions, observed the Instrument and Control (I&C)-

technicians resetting with concurrence of the licensed operator, the annunciators.

The inspector questioned the appropriateness of the I&C technicians resetting the annunciator with plant management.

The inspector determined that this was not a' control manipulation, and it was

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a past practice at the plant.

Plant management, determined that this was not a good practice, and it will be discontinued.

Escalation to 100% Power

.The inspector observed portions of the licensee's escalation to 100%

power.

The licensee had been at 97% power due to concerns about the instrument accuracies of the Reactor Coolant System (RCS) average Temperature and Pressurizer-Pressure instruments.

These concerns are noted in Licensee Event Reports (LERs) 89-20 and 89-24.

The concerns stem from T.S. 3.2.5, that places a limit on the core of 589 degrees F.

for Departure from Nucleate Boiling (DNB) concerns.

It had been previously believed that the 589 degrees F. value included instrument inaccuracies, but it was determined that the 589 degrees F. value was an actual value to be in the core.

The instrument loop inaccuracies accounted for approximately 3.9 degrees F.

The reactor average temperature (Tavg) can be automatically controlled by the rod control

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system within a 1.5 degree F. deadband.

The Reference Temperature (Tref)

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used for a 100% power value'was 584.7 degrees F.

With the deadband and instrument inaccuracies, the maximum value of 589 degrees F. could be

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exceeded (584.7 + 1.5 + 3.9 = 590.1).

Due to these concerns, power had

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been limited to'97% power to assure a Tave below 589 degrees F.

The inspector observed portions of the escalation, which was performed under Temporary Plant Test (TPT)-322, " Escalation to 100% Power."

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TPT-322 had received management review and was followed. 'The procedure was performed by licensed operators.

The plant achieved 100% power on

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January-14, 1990.

No violations or deviations were identified.

7.

Event Follow-up (92700, 92701, and 93702)

Cognitive Operator Error Resulting'in the Inoperability of Both containment Hydrogen vent Systems

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The containment hydrogen vent system provides the capability to monitor gas releases from containment while reducing containment pressure.

Two trains (train A and train B) of the containment hydrogen vent are provided for redundancy.

In 1989, the licensee identified that effective monitoring of.th9. effluent gas was not occurring, due to obtaining a nonrepresentative sample due to the location of the sample point.

As corrective action, the licensee declared the A train inoperable and tagged it out of service.

The licensee implemented a temporary t

modification for the B-train sample point so that a representative sample would be obtained thereby assuring accurate monitoring.

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By design one of the-B train L

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containmenthydrogenventcontainmentisolationvalvesispoweredfrom l

the B emergency bus and the other from the A bus.

On January 24, 1990 the B emergency diesel generator (EDG) was removed.

fromserviceformalntenance;t1usinanaccidentscenario,theB

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L emergency bus would not be energized rendering the B and the A

containment hydrogen vent system inoperable.

The operators, therefore, unknowingly entered a 30-day action.

Additionally, during the event

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critique of January 29, 1990, the licensee considered that technical specification 3.0.3 may have been entered unknowingly due to having the

,

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redundant train inoperable and having no emer ency power source.

This L

would have required the reactor be shutdrwn w thin one hour.

At the

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conclusion of this evaluation period, the licensee was continuing to evaluate this event as event report 90-018.

The inspectors reviewed the critique minutes and discussed the event with

. plant management and members of the Nuclear Safety and Regulation Department (NSRD).

The inspectors also noted that an ' outage related L

work sheet' had been generated and was filed for the A train containment

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hydrogen vent system.

A previous event associated with RHR inoperability occurred due to insufficient evaluation of an outage related work sheet.

The inspectors also noted the maintenance work request had been processed according to licensee procedures.

Because the final evaluation of this L

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11-j

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event (ER 90-018) is not complete and neither the cause of the event nor

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its safety significance has been determined, this is an unresolved item (50-344/90-02-02).

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Reactor Coolant Sample Line Leak-

,

Maintenance Request (MR) 89-1870 was generated to repair reactor coolant

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sampling check valve SS-2009 that was leaking by its seat.

As part of preparations to perform the repairs per MR 89-1870, a clearance was issued that, in addition to isolating the sampling system from the primary system to perform the check valve repair, had operators open vent

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isolation valves SS-5017, S5-5018 and a pipe cap to assure the system was depressurized.

The valve was repaired on January 24, 1990, the clearance removed, and valves restored to their normal configuration by plant operators.

During restoration of the valve positions, the operator recognized that S5-5017 and S5-5018 were hard to close; however, not harder than some other valves in the facility and, therefore, did not reise further questions to plant supervision.

The operator replaced the

. pipe cap, but did not tighten the pi e cap since he did not have a wrench.

The operator intended to ti hten the cap later, but was given other tasks and did not return to ti hten the pipe cap.

An independent verification by another operator checked that 55-5017 and SS-5018 had been shut and the pipe cap was in place.

At approximately 1:00 a.m., January 15, 1990, plant chemists obtained a reactor coolant system (RCS) sample.

Sampling had bee'n completed by approximately 1:30 a.m.; however, per procedure, the sample line remained pressurized.

l At approximately 2:56 a.m., January 25, 1990, a roving operator heard an alarming personnel radiation monitor (frisker) and found the alarm b.

,f originating from a frisker being used by a security guard as he was exiting the electrical penetration area.

The operator ascertained where-i the guard had been and proceeded,-after notifying health physics but without donning protective clothing, into the electrical penetration

.

area.

The operator noted steam and water had accumulated in the sample l

valve area.

Upon leaving and frisking out of the area, the operator also

!

found himself contaminated.

The control operator secured the sample line by shutting the containment isolation valve,_thereby terminating the

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leak.

Later that morning, area decontamination began when dayshift radiation protection personnel were called in early.

After decontamination was completed, an operator, as part of follow-up, was able to further close l

isolation valves SS-5017 and S5-5018 by an eighth to a quarter of a turn.

He also tightened the pipe cap which he found loose.

Later that day, the licensee, to gather facts for an event evaluation, l

conducted a critique.

The critique raised questions whether a procedure or guidance was in place for installing pipe caps, for re-torquing valves after a system has come to normal operating temperature and pressure, and for inspecting a system after pressurization upon removing a clearance.

Additionally, the critique noted that the contaminated water had potentially accumulated because the floor drains in the electrical

p--

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r

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penetration room had been covered with a metal plate to prevent drainage of rain water into the Dirty Waste Drain-Tank (DWDT).

The critique concluded with the assignment of the following immediate

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corrective actions:. emphasize the use of pickup items (maintenance

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requests to remove boric acid accumulations), establish a policy for visual inspection for leakage upon releasing a clearance, determine if

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Operations Management expectations are bein met, evaluate the management

' issues generated by the event, investigate ustification for plugged drains in the electrical penetration area, ydrostatically test the sample line to assess the leak tightness of 55-5017 and 55-5018, determine a criteria for torquing pipe caps and walkdown other valves in the area ins)ecting for boric acid accumulations.

The licensee is continuing t1eir evaluation via event report 90-21.

t The inspectors evaluated this event by walking down'the electrical penetration area following its decontamination noting the valves in the area were properly positioned, attending the event critique, and

. discussing the event with licensee management.

The inspectors also reviewed MR 89-1870. They noted that neither the work instructions nor

'

the post maintenance test required the system to be verified leak tight upon return to service.

The work package and the clearance were processed in accordance with administrative procedures.

Because the final licensee evaluation of this event is not completed, this event will be followed as open item (50-344/90-02-03).

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Inadequate Implementation of Compensatory Measures for Ventilating the Control Room

,

As described in NRC Inspection Report 50-344/89-33 and Licensee Event Report (LER) 89-30, on December 28, 1989, tho licensee was issued an w-emergency change to Trojan Technical Specifications (T.T.S.) that permitted ventilation of the control room with both trains of the control room chlorine detection system inope~rable for periods of up to one hour provided that compensatory measures were implemented.

One of the compensatory measures was to station at the south-east vehicle security access point, a security guard, who was to have continuous communications

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with'the control room, to notify plant operators of approaching trains that may have a chlorine shipment on board so that operators could secure control room ventilation prior to the train passing the facility.

On January 24, 1990, at approximately 1:05 p.m., the control room called l

the security guard at the south-east vehicle access gate and notified him L

that control room ventilation was going to start per procedure OI-T-64, Revision 0, and to notify the control room operator if a train was sighted.

01-T-64 step 3.2.7 states " Setup a communication between the guard and the control room." While communications were established, it was not continuously maintained per the emergency Technical Specification change.

At approximately 1:35 p.m., the guard, who was in the process of issuing a vehicle pass to a visitor, looked up and noticed a train I

rapidly approaching the facility.

The guard immediately returned to the l

guard shack and attempted to contact the control room on the ROLM phone; l

however, the phone rang approximately four times without being answered.

The train passed the facility as the guard was redialing the control room

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p

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on the Bell phone.

The security guard, shortly after the train had

?assed, reached the control operator and notified him that a train, that ind only an empty gondola car, had passed by the facility.

This is an apparent violation (50-344/90-02-04).

Licensee management met afterwards and discussed the event.

They concluded this event had limited safety significance since the train did

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not have a shipment of chlorine nor had the train derailed.

As

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corrective action, management required the security guard to maintain continuous communication, as required by the emergency technical

specification change, by hand-held side band radio.

Inspector followup identified that during or subsequent to implementation

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of the waiver, it had not been explained to the security guards that continuous communications were required.

The security guards were only aware of the need to notify the control room of an oncoming train.- Also identified was that normally two security guards were posted at the south-east gate; however, on the day of this event only one guard was

. attending the gate while the other guard was in training.

Finally, the inspector noted that the Nuclear Safety and Regulation Department had not verified proper implementation of the compensatory measures nor had the Quality Assurance Department surveilled proper implementation.

Subsequent to sharing these findings with the licensee, additional clarification of responsibility was provided to the security force.

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One violation and no deviations were identified.

8.

Follow-up of Licensee Event Reports [LERs] (90712, 92700)

i LER 89-12, Revision 1, (Closed), " Control Room Ventilation Hangar Design

Requirement Not Met."

This revised licensee event report described the causes and corrective actions for an event in which control room-ventilation duct supports were found not to have the safety margin described in the final safety analysis report. The licensee concluded the causes of the event'were inadequate construction controls for duct

supports by the design / installation contractor and inadequate licensee L>

'as-built' inspections. Additionally, the scope of the licensee design s

l basis' document (DBD) program did not include as-building or examination of supports.

Licensee corrective actions included:

immediately declaring both trains of control room emergency ventilation inoperable, establishing administrative controls to prevent plant mode changes until the seismic qualification of the system was determined, repairing l

i deficient hangers and supports, reviewing the adequacy of all safety related system ventilation supports, and reviewing design guidance for

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structural supports.

The licensee concluded that this event had 'ninimal safety significance because the as-found conditions, even though not

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L installed as design would have provided sufficient structural support duringadesignbasisseismicevent.

.

Initial NRC inspection on this event was documented in NRC inspection report 50-344/89-09.

Followup on this event included a review of the licensee response to the associated Notice of Violation (PGE letter of October 12, 1989, " Reply to Notice of Violation"), verification of

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selected hanger modifications, and discussions with licensee civil engineers on analysis techniques and improvements.

LER 89-18, Revision 1, (Closed), "Both Trains of Residual Heat Removal (RHR) System Not Known to be 0)erable as a Result of Maintenance Work on a Flow Indicating Switch for tie B Train Pump." The revised LER provided additional information as to the cause of the event, the safety significance of the event and corrective actions that are planned to prevent event recurrence. -The licensee concluded that the health and

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safety of the public was not affected since one-train of RHR had been

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available throughout the duration of the event.

The licensee additionally committed tre evaluate the shift turnover process for improvements.

NRC special inspection 50-344/89-27 was conducted to r

evaluate this event.

Two violations of NRC requirements were identified in 50-344/89-27 and are discussed in section 9 of this report,

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50-344/90-02.

LER 89-23, Revision 1, (Closed), " Containment Integrity Violated During Local Leak Rate Testing." This revised licensee event report provided a chronology of the licensee's experiences with Bunker-Ramo containment electrical penetrations,.and concluded the cause of violating containment integrity, during testing of these penetrations while in Modes 1 through 4, was due to cognitive personnel error since it was not recognized that the inner seals of the electrical penetration modules were not a containment boundary.

Additionally, the licensee concluded a contributing factor of not understanding the design of the electrical i

penetration was an absence of design documentation and the inability to obtain design information because the vendor was no longer in business.

The licensee had identified the need to obtain additional design

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information on Bunker-Ramo electrical penetrations during.the design basis document reconstruction effort.

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The licensee concluded, since the inner self-energized seal provides a seal for leak testing purposes only, that during testing while in Mode 1 through 4, a potential leak path existed from inside containment to outside containment via a purge plug through the penetration inner seal and out a nitrogen test line.

To prevent further containment integrity

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violations during testing, the surveillance procedure was revised.

Additionally, the licensee is evaluating the replacement of the t-electrical penetration modules.

The licensee concluded the event had no effect on the health and safety of the public since no overpressurization of containment had occurred.

The inspectors met periodically with licensee engineers and management L

throughout the evaluation of the event to understand the design of the j

penetrations and the significance of the event.

Additionally, the inspectors reviewed the revised surveillance procedure to ensure that testing was performed without dolating containment integrity.

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LER 89-27, Revision 1, (Closed), "Cortainment Ventilation Isolation Due i

to Spurious Electronics Noise Spike." This revised LER described the i

conclusions of the initial event investigation.

On November 8, 1989, a

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hydrogen vent exhaust isolation occurred during containment pressure I

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r.:

reduction.

Licensee evaluation of this event found a radiation monitor-with a sensing device that did not have a normal plateau curve and t

several high voltage connectors that were dirty.

Additionally, the licensee cleaned PRM-ID circuit board connectors and the mode selector switch, and checked both high and low voltage signals for noise.

The licensee concluded that alone these items did not cause the observed PRM-ID response (25 cpm to 160 cpm over a six minute time period) and

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eventual hydrogen vent system isolation.

After cleaning contacts and

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replacing and calibrating the detector, the radiation monitor functioned properly.

Licensee long term corrective actions included evaluating a design change-for the radiation monitoring system (C41), which had previously been recognized as needing increased ventilation (cooling) and had a restricted area for conducting maintenance; continuing to monitor and trend spikes on process radiation monitor channels; and developing an action plan to investigate and resolve issues associated with process

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radiation monitors.

The inspectors evaluated the trend charts of PRM-1D and the other PRM-1 radiation monitor channels and concluded, as did the licensee, that no release above limits occurred.

The inspectors also verified preventive maintenance and surveillances associated with PRM-1 had been performed as required.

The inspectors concluded the licensee actions and response-with respect to this event were acceptable.

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LER 89-29, Revision 1, (Closed), " Fire Dampers' and Penetrations'

Surveillances Not Performed Within the Required Time Frames Due to Personnel Errors." This revised LER provided additional information concerning the findings of a task group reviewing the performance of surveillances in the area of fire protection.

The tas( group identified

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four additional surveillances for sprinkler systems as being missed.

The licensee is examining their Technical Specifications' surveillance program to ensure future requirements are met.

LER 89-30, Revision 0, (Closed), " Chlorine Detector Response Time Greater L

Than Used in Toxic Gas Analysis Due to Not Specifying a Response Time in the Desic n Specification for the Detector." This licensee event report describec initial licensee actions upon discovering the response time of control room ventilation chlorine detectors, whose purpose is to shutdown the normal control room ventilation system and initiate emergency ventilation when high chlorine levels are detected, was greater than assumed in the safety analysis.

Licensee immediate corrective actions included securing the normal control room ventilation system (CB-2) and starting the emergency control room ventilation system (CB-1).

Follow-up

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licensee actions included obtaining an emergency waiver to the Trojan Technical Specifications to permit ventilating of the control room for periods of up to one hour when compensatory measures have been established, and procuring replacement chlorine detectors.

The licensee committed to providing a revised fu:2 discussion LER by January 26, 1990, and has subsequently notified the NRu that the revised LER will be delayed until mid-Februar n-

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'Theinspectorsattendedthelicenseeeventcritiqueonthissubject, reviewed the emergency' technical specification waiver, and verified licensee compliance with the waiver. : Documentation of this inspection is i

contained in NRC inspection report 50-344/89-33.

The inspector also evaluated an event associated with inadequate implementation of the waiver's compensatory actions (section 7 of this report).

Based on licensee corrective actions taken to date, the commitment to provide a revised LER and the commitment to install chlorine detectors that meet

time respons,e requirement, this LER is closed.

LER 89-32, Revision 0,-(Closed), " Incomplete Surveillance of Power Operated Relief Valves (PORV) Due to an Inadequate Procedure Review Upon Issuance of a License Amendment." On December 5, 1989, the licensee determined that surveillance testing did not fully test the auxiliary relays interfacing between the Control and Protection Circuitry, of the pressurizer PORVs.

As immediate corrective action, the licensee shut the PORV block valves in accordance with Technical Specifications.

Subsequentiv, the relays were satisfactorily tested and the system restored.

'ihe licensee concluded the cause of the event was an

~inadeguate review of procedures to identify changes needed to implement a

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Technical Specification Amendment.

The licensee noted that this weakness had been recognized in 1987, and tiiat Nuclear Division Procedure (NDP)

700-2, " Control of the Trojan Operating License and Licensing Documents,"

had been revised in August of 1987 to require that both the cognizant discipline supervisor and the Plant Review Board engineer identify changes to licensee procedures associated with an ameridment change.

However, the licensee had not retroactisely, applied the review of procedures to previous technical specification amendments.

Additional inspector discussion with the cognizant manager on this event resulted in the added commitment to review procedures affected by amendments previous to August 1987 for adequacy.

w Inspection on this event was documented in NRC report 50-344/89-33.

Unresolved item 50-344/89-33-01 was generated pending evaluation of licensee corrective actions.

The corrective action review of all surveillance procedures impacted by technical specification amendment changes, for technical adequacy, is considered adequate.

No violations or deviations were identified.

9.

Followup on Open Items, Corrective Actions for Violations and Unresolved Items (92701, 92702)

Unresolved Item 50-344/88-40-02, (Closed), " Entry into Technical Specification 3.0.3. for Operational Convenience while Testing Reactor Coolant System (RCS) Check Valves in Mode 3."

As explained in Lhe Bases of the Technical Specification, T.S. 3.0.3 delineates the ACTION to be taken for circumstances not directly provided for in the ACTION statements, the occurrence of which would violate the intent of the specifications.

As further explained in NRC Generic Letter 87-09, Technical Specification 3.0.3. was not intended to be used as an

" operational convenience" which permits routine voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components

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-being inoperable.- As documented in inspection report 88-40 the

<

inspector identified instances of apparent voluntary entry Into Technical

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i S)ecification (T.S.) 3,0.3 for operational convenience.

It appeared that tie licensee had voluntarily entered T.S. 3.0.3 nineteen times for periods-from five minutes to greater than two hours when the. testing was-

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performed in Mode 3, Hot Standby.

Entry into T.S. 3.0.3 occurred since none of the ECCS safety systems as required by T.S. 3.5.2 were operable.

The following is an apparent violation (50-344/90-09-05) where testing placed the facility in T.S. 3.0.3 in July 1988 during the 1988 Refueling Outage:

Testing of the first off ECCS cold leg check valves (8948 A, B, C and'D) during the closure of M0-8835 and M0-8809A while in Mode 3.

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Testing of residual heat removal second off cold leg check valves (8818 A, B, C and D) during the separate closures of M0-8809A and M0-8809B while in Mode 3.

Testing of residual heat removal second off hot leg check valves-(8736 A and B) during the opening of M0-8802A while in Mode 3.

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Testingofsafetyinjectionsecondoffhotlegcheckvalves(8905A, B,s0 and D) during the closure of M0-8923A M0-8923B, M0-8835, M0-8821A,theopeningofM0-8802A,andpla$ingbothsafetyinjection pumps in the pull-to-lock position while in Mode'3.

Although the conduct of these tests were generally in Mode 3 after a long.

outage and the overall actual safety significance was minimal, it is important that the licensee establish a consistent implementation of T.S.

u 3.0.3.

At NRC's request, the licensee performed a review of plant procedures to identify other instances where T.S. 3.0.3 could be entered as an operational convenience.

Other systems had been made inoperable and T.S.

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3.0.3 entered.

It was determined that six procedures had the potential to place the plant in T.S. 3.0.3:

POT 2-4'on ECCS Check Valves; ONI 14 on the Component Cooling Water System; A0 3-21 on the Control Room Emergency Ventilation Boundary; Temporary Plant Test 258 on ECCS Check Valve Leakage; POT 4-2 on the Containment Spray System; and MDP 5-7 on Component Cooling Water. Surge Tank Nitrogen Regulator Adjustment. The licensee revised these procedures to prevent entry into T. S. 3.0.3 for operational convenience.

Unresolved Item 89-16-01,(Closed) " Acceptability and Review Process for Lower Tier Procedures." During a routine inspection, a concern developed over the calibration of the Power Range Nuclear Instruments by the lower tier Instrument and Control Procedure (ICP)-21-76, " Calibration of Power Range Nuclear Instruments" which received only supervisory review and approval vice Plant Review Board review and Plant Manager approval.

The lower tier procedure (ICP-21-76) was issued April 7,1988, and previous to the use of this procedure, the vendor manual was used. The lower tier procedure was signed by the supervisor of the I&C shop, but PICT-11-1 which tests the alarm functions and trip setpoints settings was signed by

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.the plant manager.

The-lower level procedure calibrated the instruments, but did not have the in-depth review that checks the alarm functions.

l The inspector was concerned about the appropriate level of review for lower level procedures as required by ANSI 18.7.-

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.The; inspectors noted that in the licensee's Nuclear Quality Assurance

L Program, the licensee is committed to Regulatory Guide (RG) 1.33, L

Revision 2 with the exceptions noted in the supplement to the manual.

L This guide endorses ANSI 18.7-1976/ ANSI 3.2, " Administrative Controls and Quality" Assurance Requirements for the Operational Phase of Nuclear Power Plants.

None of these exceptions relate to the appropriate level of review for a procedure.

Section5.3.5.q4)oftheANSIStandardstates that for maintenance procedures, these ' procedures shall receive the same level of review and approval as operating procedures." Furthermore, RG 1.33, Section H.2,-states for I&C calibration that " implementing procedures are required for each surveillance test,"insaection or

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calibration listed in the Technical Specifications.

Tie licensee has

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taken action to resolve this item.

The licensee is currently upgrading

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all technical specification procedures to an upper tier procedure.

The

' licensee is also upgrading their procedures for instruments that meet T.S. equipment requirements.

The inspector discussed the action plan for resolution of the problem with the Manager of the I&C Procedure Up rade project.

The licensee has

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hired several contractors to upgrade the r I&C procedures.

The licensee

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intends to revise the I&C procedures by the end of September 1990.

Based i

on the discussions with the licensee, the inspector concluded that the licensee's response is acceptable.

This item is closed.

Unresolved Item 50-344/89-20-02, (Closed), " Seismic Calculational Methods Used Different Than Identified in Final Safety Analysis Report." This

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unresolved item concerned the methods used by the licensee to seismically qualify Category I equipment.

During that inspection the inspector identified to the licensee that Trojan Technical 5)ecIfication 5.7.1 states that the seismic design of equipment shall 3e designed and maintained to the original provisions contained in section 3.7 of the Final Safety Anal sis Report (FSAR).

The licensee determined they had committed to seis ically qualify Category I equipment by the Absolute Sum Method (FSAR section 3.7).

Specific examples where seismic analysis was

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performed using calculational techniques different than the absolute sum method include:

the modifications to the control room emergency ventilation system and the analysis on the design for the'new battery rack. This use of other design methods than specified could yield more conservative answers than what was used in these designs.

Further, the licensee recognized that they had not always been analyzing their seismic design in accordance with the FSAR, since cases existed where other methods (methods identified in Regulatory Guide 1.92, " Combining Modal Responses and Spatial Components in Seismic Response Analysis") were employed. The licensee's In-House Position (IHP) 1.92-1-1 to RG 1.92 noted that this guide is not applicable to Trojan.

As a result of these findings, the licensee generated Non-Conforming Activity Report (NCAR) P89-380M, documenting that IHP 1.92-1-1 states that they will use chapter 3.7 of the FSAR for seismic design.

One of

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the recommended actions noted on NCAR P89-380M was to determine which

method to use in performing seismic design analysis in the future-(FSAR 3.7 or RG 1.92) followed by making necessary revisions to the FSAR.

The inspector reviewed the licensee's actions with~ respect to this NCAR.

In talking with licensee civil representatives, they wanted to be able to use the other methods for seismic analysis, since they would be easier to implement (e.g. computer codes available), more accurate and not take as long to determine an answer.

The licensee has taken actions to commit

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(via their IHP) to use RG 1.92.

The licensee has also-reviewed the

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Trojan Technical Specifications and plans to submit a license change request to delete all reference to the original design in the' technical specifications.

The licensee plans to modify their FSAR to reflect these

changes.

The technical specification currently states that the seismic design is to the original methods in chapter 3 of the FSAR.

Due to the actions is a non-citeo violation.d in accordance with Part 2 of 10 CFR 50, this taken by the licensee an This item is closed.

Unresolved Item 50-344/89-33-01, (Closed), " Incomplete Reactor Coolant System Power Operated Relief Valve (PORV) Channel Functional Calibration."

Refer to section 7, LER 89-32, Revision 0.

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Enforcement Item 50-344/89-09-04, (Closed), " Failure to Meet Required Factor of Safety for CB-1 Intake Pipe Support Anchor Bolts." The

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licensee responded to the violation via letter dated October 12, 1989.

The licensee concluded the factor of safety for the CB-1 intake pipe s.

support anchor bolts was approximately six which was greater than the required factor of safety of four.

The factor of safety of six was derived by using a realistic wind drag coefficient around the two-intake ducts for 300 mile per hour winds.

The inspector reviewed these-calculations and found them acceptable, t

The licensee concluded that a violation of. Nuclear Division Procedures existed since Calculation TC-508, that calculated Anchor Bolt Factor of Safety, did not have sufficient detail as to purpose, methods, assumptions and design input.

Because the assumptions and design inputs of the calculation were not adequately detailed, a reviewer could not ascertain the actual loading placed on the intake structure support anchor bolts.

As corrective actions, the licensee reperformed the calculations properly documenting the purpose, methods, assumptions and design input, and conducted training sessions to reinforce the requirement for documentation of design information in calculations.

Enforcement Item 50-344/89-09-08, (Closed), " Failure To Test The Control Room Emergency Ventilation System Per Technical Specification 4.7.6.1."

The licensee responded to the violation via letter dated October 12, 1989. The licensee concluded the cause of the event was the failure to communicate and correlate a design change with the operational objectives it should have fulfilled.

As corrective action, the licensee proposed l

upgrading the supplemental control room cooling system to safety related

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e ~,1

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status based on a probabilistic. risk assessment (PRA).

In the interim, the licensee has been' testing CB-1 with CB-16 in operation.

If CB-16 becomes inoperable, the licensee will test CB-1 accordingly.

This item is closed based on the submittal on December 21, 1989 of the PRA to the

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NRC for review.

Enforcement Item 50-344/89-20-01, (Closed), " Rod Position Indication Surveillance Violation." The licensee responded to the violation via letter dated October 30, 1989.

ThelicenseeconcludedTrojanTechnical 5)ecification 4.1.3.2, Reactivity Control Systems-Position Indication C1annels, was violated due to personnel error in that following a computer outage, a computer variable was not restored.

Contributing

)

causes for the violation were that no administrative controls existed to periodically verify operability of the Rod Deviation Monitor and the surveillance data sheet did not have space for documenting shutdown rod

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bank position during inoperability of the Rod Deviation Monitor.

Licensee corrective actions included:

issuing a Training Information Bulletin (TIB-65) that provided training and guidance on the Rod

. Deviation Monitor and the requirement for increased surveillance when the l

plant computer is inoperable, changing the plant computer software so that the position variable would be automatically restored following a computer restart, and changing the surveillance procedure to periodically check that the Rod Supervision Computer is active.

The inspectors attended the critique of this event, evaluated the event report (ER 89-222), reviewed the training information' bulletin, sampled licensed operators knowledge on the purpose and operation of the Rod

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Deviation Monitor, and reviewed the surveillance procedure change.

The

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inspectors will evaluate the effectiveness of these changes during routine inspection.

This item is closed.

>.,

Enforcement Item 50-344/89-24-01, (Closed), "Use of Uncalibrated Dial Indicators on Safety Related Equipment." The licensee res)onded to the Notice of Violation via letter dated December 18, 1989.

T1e licensee concluded the reason for the violation was a procedural deficiency in that conflicting instructions existed regarding the calibration of dial l

i indicators used as measuring and test equipment.

As corrective actions, L

procedure MDP-1-12, " Dial Indicators," was revised to provide specific instructions for calibration; mechanical maintenance planners were instructed to include the requirement in maintenance work instructions for dial indicators to be calibrated; and mechanical maintenance personnel were trained to the specific requirements.

The inspectors verified that dial indicators were now included in the maintenance and test equipment calibration program, MDP-1-12 had been revised, and the training was conducted.

Enforcement Item 50-344/89-24-02, (Closed), "Non-Initiation of Out-of-Calibration Investigation for Out-of-Specification Detector

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Voltage." The licensee responded to the Notice of Violation via letter i

dated December 18, 1989.

The licensee concluded the cause of the l

violation was personnel error in that the instrument technician supervisor and the reviewer failed to recognize the out-of-tolerance condition.

The instrument technician did not consider the reading l

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(n,

,e

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out-of-tolerance because it was not a bistable.

As corrective action, the power supply was adjusted to within tolerance, the affect of the

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out-of-tolerance indication was evaluated and found to have no adverse affect on detector operation, training was' conducted with the Maintenance Department, and unique tag numbers will be provided for instrument drawer power supplies.

The inspectors verified the power supply had been adjusted to within required calibration voltages, that training had been conducted and that

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the impact on instrument operability was negligible.

Enforcement Item 50-344/89-27-01, (Closed), " Noncompliance with Administrative Order ( AO) 3-9, ' Maintenance Requests'." The licensee responded to the violation via letter dated December 18, 1989.

The licensee concluded the violation was due to personnel error in that the shift supervisor believed that identifying the Limiting Condition for

Operation (LCO) and not the applicable LC0 subpart was adequate.

Also, a contributing cause for the violation was that the shift supervisor, due

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.to not having the work instructions for the maintenance to be performed, could not accurately determine which subpart of the LCO would be applicable.

As immediate corrective action, the licensee issued a memorandum to all Shift Supervisors emphasizing the requirement to identify the applicable technical specification and its subpart when

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completing maintenance requests.

As long term corrective action, the licensee plans to revise the work control process so t, hat the Work Control Group licensed senior operator has the responsibility for operability determinations.

Based on the above near and long term corrective actions, this item is closed.

Enforcement Item 50-344/89-27-02, (Closed), " Noncom)11ance with N

Administrative Order ( AO) 3-14, ' Control of Safety Related Equipment Outages'." The licensee responded to the violation via letter dated December-18, 1989.

The licensee concluded the cause of the event was personnel error in that the shif t supervisor did not recognize the maintenance being performed on a plant component would render a safety pump, and therefore, a safety train inoperable.

Contributing causes to the violation were that the maintenance work instructions were not sufficiently detailed to clearly call out what the impact on safety equipment was, and A0-3-14 did not have specific direction as to the dispositioning of the safety related equipment outage worksheet if the shift supervisor disagreed with the planners decision that the sheet was required.

Licensee corrective actions included:

incorporating this event's lessons learned in operator requalification training, disciplinary action with involved personnel, scheduling of train related work on the Plan of the Day, assigning a licensed senior reactor operator to provide an independent review of proposed work for safety-impact, and revising A0-3-14 to include instructions for the disposition of safety related outage worksheets.

The inspector attended the licensee critique for this event, reviewed the internal event report, attended PGE management meetings that focused on revising the Trojan work control program, and interviewed the personnel involved with the event.

The inspector verified the above mentioned corrective actions.

The effectiveness of the corrective actions and the

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implementation of the new work control program will continue to be evaluated as part of routine inspection.

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Two violations and no deviations were identified.

10._ Followup on 10 CFR 21-Reports 89-29-P, (Closed), " Dresser Pump Division-Potentially Defective 2 inch Charging Pump Barrel Sleeve." A license phone conversation with Pacific Pump identified that Portland General Electric did not purchase any of-the defective sleeves.

No violations or deviations were identified.

l 11. Temporary Instruction-(TI) 2500/27 " Inspection Requirements for NRC Compliance Bulletin 87-02."

Tnis TI was issued to evaluate the licensee's root cause determination and the implementation of corrective actions in response to bulletin

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87-02. NRC Bulletin 87-02 requested licensees to test safety-related (SR) and nonsafety-related (NSR) fasteners. Supplements 1 and-2 to the 1-L Bulletin requested licensees to provide a list of the suppliers and/or manufacturers from whom the fasteners may have been purchased. Temporary Instruction 2500/26 required an NRC inspector to participate in the i

E licensee's selection of fasteners to be tested to assure that they were L

representative of installed fasteners. The inspection of 2500/26 was documented in inspection report 50-344/88-03. The licensee's testing identified samples with deficiencies. These samples, the Nonconformance Report (NCR) generated, and the material type are noted in the table.

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Sample NCR Number Material Type TNP-22 88-004 ASTM A193 GRB7 5/8-11x2 Stud TNP-21 88-005 ASTE SA193 GRB 16 7/8-9x5 Stud TNP-7R 88-006'

ASTM A194 GR8M 3/4 Nut l

l TNP-24 88-007 ASTM A193 GRB8M 5/16-18x1 1/2 Bolt 1-TNP-25 88-008 ASTM A193 GRB8M 1/2-20x4 Bolt TNP-26 88-009 ASTM A193 GRB711/4-8x71/4 Bolt TNP-33 88-010 ASTM A193 GRB7 3/4-10x71/2 Stud

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The licensee also generated other NCRs on fasteners that were not part of the bulletin, which included NCRs87-417, and 87-430.

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l-Temporary Instruction 2500/27 was specifically concerned with Sample L

Numbers TNP-7R, TNP-21, TNP-24, TNP-25, and TNP-33. TNP-33 was installed in a safety-related application while the other bolts identified were installed in non-safety-related applications.

Safety-Related Fasteners

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I The sample which failed to meet specification requirements was TNP-33. A stud failed to meet both the chemical and mechanical tests for A193 grade 87. The material was a resulfurized AISI 1137

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steel.. The licensee reviewed records, and identified one possible-I safety-related activity for which the stud could be issued. The studs were verified not-installed at the facility.

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Non-Safety-Related Fasteners The Non-Safety-Related (NSR) fasteners and the material defect listed in the temporary instruction are summarized below..

Sample NCR Number Nonconforming Condition TNP-21 88-005 Actual elongation 14% vs. required min, of 18%

TNP-7R 88-006 Chromium and Nickel contents not in specification TNP-24 88-007 Hardness less than required H

TNP-25 88-008 Hardness less than required and sulfur content too high.

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The licensee p~erformed an analysis that concluded the installed

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nonsafety fasteners w'ere acceptable for the application in which they are used. The remaining nonsafety fasteners and spares have been discarded. The licensee concluded that the root cause for L

these problems was inadequate receipt inspection.

L Further, it was noted in Information Notice 86-25, Revision 1, that there was a problem of intentionally mismatched SAE Grade 8.2 bolts for SAE Grade 8 bolts.. The licensee had not purchased any SAE grade bolts; they used equivalent bolts.

h The NRC was further concerned with the licensee's commercial grade

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l procurement receipt inspection activities for pipe and issued escalated I.

enforcement on:this issue. This is documented in inspection report 50-344/88-46. The licensee has initiated corrective actions to address these concerns. The inspector reviewed a sample of the licensee's receipt. inspection for bolts. This TI is closed.

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No violations or deviations were identified.

12. Trojan Nuclear Operating Board [TN0B] (40500)

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Trojan Technical Specification 6.5.2. requires the establishment of the TNOB as an oversight committee functioning to provide independent review

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and audit of.a broad spectrum of Nuclear Division activities. The

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. Technical Specification requires the TN0B to meet at least once each six l.

months. The TNOB reports to and advises the Vice President, Nuclear on H

those areas under review by the TN0B.

The licensee's Nuclear Division Procedure (NDP) No. 500-1, " Trojan l

Nuclear Operating Board" describes the purpose and responsibility of the l

TN0B. The TN0B is one of the licensee's groups that perform a self-assessment function. Self-assessment organizations act in a I

measurement and advisory capacity; monitoring the overall performance of the plant; identifying substandard or anomalous performance and

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precursors of potential problems; reporting findings in an understandable format in a timely fashion to a level.of line management having the authority to effect corrective action; and reporting those assessment

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results to line management.

- To prepare for this inspection, the inspectors reviewed recent licensee event reports (LER), recent plant event reports, TNOB meeting minutes, and various Quality Assurance audits and surveillances. The inspectors attended the TNOB meeting of January 17 and 18, and reviewed the licensee's minutes of the January TNOB meeting.

Significant recent plant events and reports evaluated by the TNOB

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included: debris and design concerns with the containment recirculation sump fire protection program adequacy, RCS Tave and Tref operational concerns, and the Cascadia Subduction Zone Earthquake report. The TNOB review of these issues appeared to be comprehensive. The TN0B focused on the causes of the events and the adequacy of proposed or implemented corrective actions.

The inspectors concluded the licensee is meeting the requirements of the Technical Specifications with respect to the composition, duties, meeting frequency and responsibilities of the TN0B associated committees. The inspectors, in the course of routine inspection, will assess the effectiveness of TN0B initiatives to improve Trojan performance with respect to safety.

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No violations or deviations were identified.

13. Temporary Hodifications Temporary Modifications (TMs) are minor, temporary changes to plant equipment that do not conform to approved drawings or other design documentation. Temporary modifications are intended-to be installed for a short period of time for evaluative purposes. Since-these are modifications to equipment, they receive independent technical review, Quality Assurance (0A) review, and plant manager approval. Temporary-Modifications include:

a.

Lifted Leads b.

Pulled Circuit Boards c.

Disabled Annunciator Alarms d.

Mechanical Jumpers Temporary supports, enclosures, or other structures attached to e.

permanent equipment.

The licensee had 79 open TMs at the end of January. The licensee's

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program on Temporary Modifications is specified in Administrative Order (A0) 5-8, " Temporary Modifications (TMs)."

Age of Temporary Modifications The inspector found that the majority of TNs were greater than six months old, and the average age of the TMs was about one and a half years old.

Of the 79 TMs eleven related to temporary instrumentation to monitor

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system performance. The three oldest TMs (from 1985) relate to valve modifications and replacements, and spurious alarms from abandoned radweste equipment.

j Record Storage.

The inspector found that the original records were kept in the tagging office, in an unlocked and non-fire proof cabinet. The TM originals could be removed by completing a takeout card (i.e. honor system).

During a selective review of the TMs, the inspector determined that TM 89-072, " PERM 1 Sampling," had certain parts of the TM with white out

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over them.- The licensee procedure on records (A0 7-1) explicitly states that white out.is not to be used (step 4.7.1).- This is contrary to the licensee's procedure, and is an apparent violation (50-344/90-02-06).

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response to this item, the licensee initiated an NCAR. As of February 21, 1990, this NCAR had not been issued. The closure of this NCAR will be tracked by the residents as part of routine inspection.

Temporary Modifications Restoration The~ inspector identified from his review of the Temporary Modifications, a TM (87-076) which had been closed in August 1989, but which had the tags still in place. This TM was to modify the outside of panel C-41, which is located in the. control room. As of February 5, 1990, the tags were still in place. The licensee's procedure A0 5-8,. step 4.8.1.a.

-states in part that "Upon receipt of the As-Built cover letter and

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processing of the turnover check list NPE on site shall:

a.

Remove the l

TM tags and-discard."

In response to this item, the licensee initiated I

an NCAR. As of February 21, 1990, this NCAR had not been issued. The

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>;, observed condition is contrary to the licensee's procedure and is an L

apparentviolation(50-344/90-02-06).. The licensee's OA organization had previously identified (on December 15,1987) in NCAR P87-172 that TM tags l-'

had not been removed following restoration of a modification. The licensee's corrective action consisted of discussions with maintenance personnel regarding removal of the tags. The corrective action taken appears to have been ineffective.

Temporary Modifications Safety Reviews

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The licensee's procedure (A0 5-8) specifies that periodic TM reviews are

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to take place. The licensee's procedure, step 4.11.1 states the following:

"4.11.1 The PSE Branch Manager shall direct a PSE engineer to perform a periodic TM Review when one of the following conditions exists:

a.

Any TM has been installed for more than six months and no TM Review has been initiated or completed 0,. R l

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More than one year has passed since the last periodic TM Review."

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In reviewing the TMs, the inspector questioned the periodicity of the six

month and one year 1eviews. The inspector was informed that the above

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steps were not performed as stated and that it was not the intent of the

~ l procedure step. The intent of the procedure was for annual reviews only.

The engineers wanted to make sure that the TM was flagged at six months to ensure that the review was done before the first year was over.

The licensee had not been performing the reviews as stated in the procedure.

Specifically, the following data indicates the reviews for the TMs.

-TM Number Installed 1st Review Days Late 2nd Review Days Late i

28 days N/A it

89-072 7/13/89 89-058 6/19/89 53 days N/A

89-002 4/27/89 105 days N/A-

88-080 6/30/88 7/05/89 186 days N/A 88-011 2/8/88 11/28/88 122 days 12/19/89 21 days88-004 1/15/88 10/06/88 83 days 10/18/89 12-days Notreviewedasoftheendoftheinspectionperiod(2/10/90)

This is an apparent violation and will be tracked under open item number 50-344/90-02-06 with the other violations in this section.

The licensee's OA organization separately identified a concern for lack of TM reviews by the operations department. This is noted on NCAR 90-053, dated January 25, 1990, which states that operations has not been documenting their monthly review of TMs. The last record of completion of a TM review was March 14, 1989.

>., One violation with three examples and no deviations were identified.

14. Exit Interview (30703)

The inspectors met with the licensee representatives denoted in paragraph 1 on February'26, 1990, and with licensee management throughout the inspection period.

In these meetings the inspectors summarized the scope and findings of the inspection activities. The inspectors emphasired the continuing instances of procedural noncompliance, personnel error and inadequate supervision of routine and off-normal activitier.

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