ML20137S038

Insp Rept 50-344/85-39 on 851117-860103.No Violation or Deviation Noted.Major Areas Inspected:Operational Safety Verification,Corrective Action,Maint,Surveillance & Followup on Potential Sabotage

ML20137S038
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	01/24/1986
From:	Dodds R, Kellund G, Richards S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20137S025	List: IR 05000344/1985039 ML20137S020 ML20137S038
References
50-344-85-39, NUDOCS 8602130559
Download: ML20137S038 (10)

See also: IR 05000344/1985039

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U.S.NUCLEARREGULATORYCbMMISSION

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-REGION-V

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. Report No'. 05' -34'4/85-39

E Docket No.'50-344' License No. NPF-1

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Licensee: , Portland General Electric Company

E c121 S. W. Salmon Street

Portland, Oregcn 97204

Facility Name: .Troj an

Inspection at: Rainier, Oregon

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^ Inspection conduct  : ove ber , 19 -

January 3, 1986-

Inspectors: _ # , >> M/ //M/f[

. A. Ridtard5 - '

D/te' Signed

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9 . Kellund ' -

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' Approved By: fdd / [

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T. Dddds{ Chief ~ Dateisiigned

eactor Projects Section 1 .;

, Summary:

Inspection _on November 17, 1985 - January 3, 1986 (Report 50-344/85-39)

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Areas Inspected: Routine inspection of operational safety verification,

corrective action,-maintenance, surveillance, followup on a reported potential

act of sabotage, review of modification testing, and inspection of various

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aspects.of. plant operation. The inspection involved 212 inspector-hours by

r the NRC Resident Inspectors. 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> of'insp'ection were during back shift

hours. Inspection procedures 30703, 40700, 61726, 62703, 71707, 71710, 72701,

p 93702 and 94703 were used as guidance during the conduct of the inspection.

c . Results: No violations or deviations were identified.

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-DETAILS

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, J 1. . ' Persons' Contacted - .

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" " - ,*W .S~ . 0rser,IPlantU General: Manager ~ -

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~ 1*R.P..Schmitt',9 Manager, Operations and Maintenance

i. % 1*D.R.^Keuter,1 Manager, Technical Services ,

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/N "J.D. Reid, Manager,.. Plant. Services.

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, R.E. Susee, Operations. Supervisor: .

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'1D.W? Swan ~(Maintenance Supervisor

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i A.~S( Cohlmeyer, Engineering. Supervisor -

JG.L; Rich, Chemistry. Supervisor. .

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,- 7T.0. Meek, Radiation' Protection Supervisor

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g'1 . 1 ?S.B. Nichols,; Training Supervisor ,

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'l  :%D.L'.-Bennett,ControlandElectricalSupervisor

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lM.R.~ Snook,fActing Quality Assurance Supervisor ~ .

,- -R.W.:-Ritschard, Se.curity: Supervisor.. ,

_ . .H.E.'Ro~senbach,. Material Control. Supervisor ~ ..

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J.K.~ Aldersebaes, Manager, Nuclear Maint. and Construction

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.The. inspectors also' interviewed :and talked with other licensee employees-

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during~the' course of the inspection. }These included-shift supervisors,

,y reactor and auxiliary operators,Jaaintenance personnel,' plant; technicians

~, and engineers,'and quality assurance personnel'.

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~ R> 2 ? Operational Safety Verification. .

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.During this inspection. period,tthe' inspectors-bbseived and examined-

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cactivities.to verify the operational; safety.of theylicens'e'sffacility. e

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_. - ~(The observations'and examination's of those, activities were conducted on a. a

daily, weekly,' or biweekly basis.

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N On'a daily basis, the inspectors: observed control room activities to

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. verify the' licensee's adherence to? limiting' conditions for operations as. ^ ' '

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' prescribed in'the. facility technical' specifications.+ Logs, '

l ' instrumentation,' recorder traces, and other operational' records were--

. examined to obtain-informatiion on plant, conditions, trends, and -

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compliance with regulations. On occasions when asshift turnover was.in

progress,-theiturnover of information on plant status was1 observed to

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qdetermine tha't all; pertinent information,was relayed to the_onroming

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shift.

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.During'each week, the inspectors toured the, accessible' areas of the

. facility to observe the.following items:

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.7 a.- General' plant and.equipmentcconditions.

. , b ." ' Maintenance reqiests and repairs'.

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4.;9 ; . c/ Fire hazards and fire fighting equipment. .

ii d I 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.

f. Interiors of electrical and control panels.

.g. Implementation of the licensee's physical security plan.

h. Radiation protection controls.

i. Plant housekeeping and cleanliness.

.j. Radioactive waste systems.

The licensee's equipment clearance control was examined weekly by the

inspectors to determine that the licensee complied with technical

specification limiting conditions for operation with respect to removal

of equipment from service. Active clearances were spot-checked to ensure

that their issuance was. consistent with plant status and maintenance

evolutions.

During each week, the inspectors conversed with operators in the control

room, and with other plant personnel. The discussions cent'ered on

pertinent topics . relating to general plant conditions, procedures,

security, training, and other topics aligned with the work activities

involved.

The inspectors examined the-licensee's nonconformance reports (NCR) to

confirm that deficiencies were identified and tracked by the system.

Identified nonconformances were being. tracked and followed to the

completion of corrective action. NCRs reviewed during this inspection

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period included P85-40, P85-47, and'P85-63.

Logs of jumpers, bypasses, caution, and test tags were examined by the

inspectors. Implementation of radiation protection cot trols was verified

by observing portions of area surveys being performed, when possible, and-

by examining radiation work permits currently in effect to see that

prescribed clothing and instrumentation were available and used.

Radiation protection instruments were also examined to. verify operability

and calibration status.

The inspectors verified.the operability of selected engineered safety

features. This was done by direct visual verification of the correct

position of valves, availability of power, cooling' water supply, system

integrity and general condition of equipment, as applicable. ESF systems

verified operable during this inspection period included the spent fuel

pool cooling system, diesel fuel oil system, containment spray system,

the auxiliary feedwater system, and the safety injection system.

No violations or deviations were identified.

3. Corrective Action

The inspectors performed a general review of the licensee's problem

identification systems to verify that licensee identified quality related

deficiencies are being tracked and reported to cognizant management for

resolution. Types of records examined by the inspectors included

Requests for Evaluation, Event Reports, Plant Review Board meeting

minutes, and Quality Assurance Program Nonconformance Reports. The

inspectors concluded that the licensee's systems were being utilized to

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correct identified deficiencies. Plant Review Board meetings were

attended by the inspectors on December 11 and January 2. The inspectors

verified that the appropriate committee members were present at the

meeting and that the meeting was conducted in accordance with the

requirements of section 6.5.1 of the facility technical specifications.

No violations or deviations were identified.

4. Maintenance

A maintenance activity observed during this inspection period was the

calibration of the internal drawer modules of the N41 channel of the

power range nuclear instrumentation on December 19. During this

activity, the inspectors verified that the personnel performing the

activity were qualified, that the appropriate procedure was followed, and

that the equipment was removed from and restored to service in a manner

allowed by the technical specifications. The inspector also verified

that the appropriate administrative procedures were followed in

conducting the calibration, The inspectors also observed that the test

equipment was indicated to be in calibration and data were properly

recorded by the technicians when required by procedure.

No violations or deviations were identified.

5. Surveillance

The surveillance testing of safety-related systems was witnessed by the

inspectors. Observations by the inspectors included verification that

proper procedures were used, test instrumentation was calibrated and that

the system or component being tested was properly removed from service if

required by the test procedure. Following completion of the surveillance

tests, the inspectors verified that the test results met the acceptance

criteria of the technical specifications and were reviewed by cognizant

licensee personnel. No corrective action was required due to the test

results. The systems were returned an operable status consistent with

the technical specification requirements following the completion of the

test. . Surveillance tests witnessed during the inspection period were

associated with a full' core flux map on December 5, safety injection pump

inservice testing on December 18, and incore/excore nuclear

instrumentation calibration on January 3.

L No violations or. deviations were identified.

6. -Modification Testing

The inspectors reviewed the documentation of testing performed for four-

modifications which were implemented during- the 1985 outage. The

modifications, designated by the licensee as Requests for Design Change

(RDC), are as follows:

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RDC 83-042, which implemented shunt trip attachments on the reactor

trip breakers.

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RDC 83-051, which modified the function of the boron injection tank

such that heat tracing and recirculation of the tank contents is no

longer required.

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-RDC 83-054, which installed environmentally qualified electrical

connections on the reactor head vent valves and the hydrogen.

analyzer containment isolation valves.

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RDC 83-059, which replaced the limit switches on two chilled water

containment isolation valves with environmentally qualified

switches.

The inspectors verified that the testing conducted checked the

modification for proper operation and that the completed test results

were properly reviewed and filed for document retention. The inspectors

concluded that the testing for the four modifications reviewed was

adequate. The inspectors did observe that testing requirements for

modifications are not always clearly stated in the modification packages.

This concern was previously noted by the licensee's q:ality assurance

organization and is being acted on by the licensee.

No violations or deviations were identified.

7. Technical Review Meeting

As discussed in inspection report 85-21, the number of engineering

discrepancies noted during the past 18 months had caused the inspectors

to question the adequacy of licensee's reviews of technical work. On

December 13, 1985, the licensee met with several representatives of the

NRC, at the licensee's corporate office, to discuss actions underway to

improve their performance in this area. These actions are summarized as

follows:

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Additional resources are being provided to the nuclear division.

Twenty-nine new positions have been approved for 1986 and 42

temporary or contract positions will be made permanent.

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Design change and calculational procedures have been or are being

revised to eliminate the source of past errors.

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Programs are underway to update and reverify vendor manuals and

electrical vendor drawings.

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Specific problem areas, such as safety related tank volumes being in

error, are receiving detailed engineering reviews.

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The licensee's management has initiated several actions to more

closely control and monitor activities in this area.

The NRC representatives discussed the importance of ensuring that strong

. independent technical reviews are being performed by personnel in the

nuclear division. They also stressed-the need for management to

encourage a questioning attitude in their personnel and to reinforce that

atmosphere by frequent personal contact with workers at the site. The

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NRC' representatives concluded that-the actions taken by the licensee are

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a positive step towards minimizing errors in technical work.

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No violations or deviations were identified.

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Shift Crew Manning

. The inspectors. reviewed the information associated with the Possible-

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Reportable Occurrence (PRO) report dated June 25, 1985. This PRO

. concerns potentially inadequate shift crew manning on the day shift of

June 25, 1985. On this date, the plant was in Mode 5 and in a solid

plant condition. Technical Specification 6.2.2 and Administrative Orders

11-4, 3-1 and 3-8 require a minimum shift crew of six operators in Mode 5.

In addition, Administrative Order 3-8 requires that during solid plant

operations, one operator will monitor RCS' parameters and have no other

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' concurrent duties.

The inspectors discussed this event with the initiator of'the, PRO and

with the Operations Planner / Scheduler. Based on these discussions and

review of the associated ~ records, the inspectors determined that th'e

' shift was adequately manned. (The inspectors did, however, question the

lack of firm criteria for determining the -availability of fire. brigade

members for response-to a fire. -In this, instance, one of.the fire

. brigade members was inside the- containment building for a portion of the

shift, and his ability to respond to a fire in the uncontrolled areas of

the plantL in a timely manner 1was~ in ' question. The. Operations

. Planner / Scheduler agreed to investigate this_ issue to determine if

additional guidance on7 fire brigade member availability is necessary._

This issue will be followed up in a future inspection (344/85-39-01).

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9. Potential Sabotage Event

On December 9, 1985, while in the process of performing a semi-annual

preventative maintenance inspection on the 'A' emergency diesel generator

(EDG), a licensee mechanic discovered an 8 ounce ball peen hammer under a

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rocker-arm cover on the east unit of the EDG. The engine was operating

at_the time and the worker immediately removed the hammer from the

engine. The hammer had not caused'any-damage to the engine. Its

location was such that the probability for damage to occur should the 3

hammer have shifted its position due - to engine vibration appeared very i

Iow. The licensee initially thought that the hammer had been

inadvertently left in the engine by a maintenance worker, however, a

review of maintenance records indicated that during the time frame in

question, no work had been performed on the engine which~could account

.for the hammer. Because the licensee was unable to determine how the

hammer ~came to be placed in the engine, tLe licensee reported.the event

to.the NRC and the FBI as a potential act of sabotage.

A special agent from the Portland office of the FBI commenced an

investigation into the circumstances surrounding this event. The.

licensee initiated action to survey the plant for other evidence of

tampering. These actions included detailed visual examinations of

electrical panels, rotating equipment, and other selected vital

' equipment; sampling of oil from selected safety equipment; a -visual

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examination of the 'B' EDG rocker arm assemblies; extensive plant wide

tours by operations personnel looking for out-of-normal conditions; and 1

verification of the locked valves in the EDG and auxiliary feedwater

systems. No other evidence of tampering was found.

The inspectors reviewed the licensee's actions with plant management and

clcsely .followed the licensee's efforts to survey the plant. The

inspectors also independently reviewed maintenance records associated -

with the EDG. The Plant Review Board (PRB) met and discussed the event

and the actions'being taken. Because no further evidence of tampering

was found, the PRB recommended to plant management that no further action

be taken pending the completion of the FBI investigation. At the

conclusion of the inspection period, the FBI investigation remained open.

The licensee's security organization is also reviewing this event. The

inspectors concluded that the licensee response to this event, to date,

has been appropriate. Licensee management stated that the resident

inspectors will be kept informed of any further developments.

No violations or deviations were identified.

10. Miscellaneous Observations

During a routine control room tour, the inspectors noted that component

ceoitng enter (CCW) flow had been secured to the B-2 and B-3 containment

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. air coolers (CAC) in an effort to increase CCW flow to the excess letdown

heat exchanger, which was then in service. The inspectors questioned

whether the operability of the CACs was affected by this condition,

however, the control operator indicated that the CACs could still be

considered operable.with CCW flow secured. After a review of technical

specification requirements, the personnel on shift agreed that CCW flow

to the CACs was required to consider them operable. A technical

specification violation did not occur due to the short period;of time

that the system was in this condition. Based on discussions with other

operations personnel, the inspectors concluded that this was an isolated

weakness in'the individual operator's knowledge. The inspectors

discussed this occurrence with the plant general manager.

Because of the recent removal of the boron injection tank from the

service for which it was originally designed, valves MO 8803 A/B have

been placed in the open position with power to the valve operators

removed. This also deactivated the valve position indication in the

control room. -These valves are in the direct flow path of the high

pressure injection pumps and as presently aligned are basically manual

valves. The inspectors questioned whether they should be designated as

locked valves. The licensee is considering this concern.

-The licensee has continued to experience an increasing primary to

secondary leak in the 'C' steam generator. At the close of the

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inspection period, the leak rate was approximately 140 gallons per day. '

The licensee has taken action to increase health physics monitoring of

secondary plant systems. The inspectors will follow the licensee's

actions to monitor the leak closely.

.No. violations or deviations were identified.

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11. Inservice Testing of Snubbers

The.results of the licensee's inservice testing of snubbers in

conformance to technical specification requirements during the 1985

refueling outage was examined by a review of test and maintenance data

and discussions with responsible engineers. The testing was performed

pursuant to procedure number 0816N.0185, Snubber Inservice Test Program.

All changes to test acceptance criteria and/or deviations had been

properly reviewed and approved by the Plant Review Board and the General

Manager.

The licensee determined that a high percentage of PSA-1/4 and PSA-1/2

mechanical snubbers manufactured by Pacific Scientific were inoperable.

The cause of these failures has not yet been determined. Additionally,

hydraulic snubbers manufactured by both Anker-Holth and Bergen-Paterson

were found significantly degraded, primarily due to deteriorated seals.

The inoperable mechanical snubbers were replaced with operable mechanical

snubbers, and the hydraulic snubbers were rebuilt and retested

satisfactorily.

All of the mechanical pipe snubbers at Trojan were manufactured by

Pacific Scientific. An initial sample of 10 percent of each type of

mechanical snubber ranging from PSA-1/4 to PSA-35 was tested. Each

snubber was evaluated for operability based on predetermined acceptance

criteria established for Trojan on the basis of manufacturer's acceptance

criteria and generalized stress analyses. Each snubber which failed the

predetermined acceptance criteria was declared inoperable and an

additional 10 percent of that type of snubber was tested. As a result of

high failure rates, 100 percent of the PSA-1/4s and PSA-1/2s were

functionally tested. The following table displays for each type of

mechanical snubber, the number tested, the percent of the total of that

type which were tested, and the number of failures. Several snubbers

which did not meet the predetermined acceptance criteria were later

declared operable when a specific stress analysis for the particular

installation was performed. The table represents the final failure

total.

Snubber Type No. Tested Percent of Total No. of Failures

PSA-1/4 50 100 13

PSA-1/2 74 100 12

PSA-1 8 20 0

PSA-3 11 10 0

PSA-10 10 10 0

PSA-35 2 10 0

The cause of some of the failures (22) was attributed to exceeding the

5 percent drag force criterion. The failure mechanisms for the snubbers

was still being evaluated by the licensee and the manufacturer.

There are four 900-kip, Anker-Holth hydraulic snubbers installed on each

of Trojan's four steam generators. -During the 1985 refueling outage, all

16 snubbers were visually inspected with no significant discrepancies

noted. Paul-Munroe Incorporated, was contracted to perform the

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functional testing of these-snubbers. The first two snubbers tested

(from.the D steam generator) would not respond under a 100-kip load. The.

snubbers appeared to be locked in a cold position (i.e., fully

compressed). As a result of these test failures, and in light of the

uncertainty.regarding the time required to rebuild the snubbers, a

' decision was made to assume all the steam generator snubbers were

inoperable, and not perform any further testing. The snubbers were then

removed and overhauled by Paul-Munroe. During the overhaul of the

snubbers, marks were found on the cylinder walls indicating the snubbers

had been moving. The snubber seals were found to be degraded and the

hydraulic fluid was heavily contaminated with seal material and rust.

Paul-Munroe was of the opinion that the foreign material in the hydraulic

fluid would not have affected the normal operation of the snubbers

because of the relatively large channels through which the fluid would

normally flow. In the case of a seismic or other severe dynamic event,

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it was determined the snubbers would have locked up but that the foreign

material could have blocked the bleed orifice, thereby preventing the

snubber from unlocking. There have been no seismic or other severe

dynamic events at Trojan which would have caused the snubbers to lock up.

This was further evidenced by the fact the seals showed no signs (e.g. ,

extrusion) of having been under a large load.

Following overhaul, the snubbers were retested using the criteria in

Section 5.4.12.1.7 of the Trojan Updated' Final Safety Analysis Report;

-namely the snubbers maximum drag force is 1,000 lbs. at a minimum

displacement rate of 25 mil / min. The snubbers could not satisfy these

i criteria. Each time the snubber. velocity approached 25 mil / min., the

snubber locked up.

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the acceptance criteria were revised to a minimum displacement rate of

7.6 mil / min with a maximum drag force of 5,000 lbs. The snubbers tested

satisfactorily with these criteria.

There are four hydraulic pipe. snubbers installed on safety-related piping

p at Trojan. These snubbers were manufactured by Bergen-Paterson, and are

installed'on the four main steam lines inside Containment. The snubbers

installed on the A and B main steam lines are rated at 130 kip, and those

installed on the C and D main steam lines are rated.at 70 kip. Each of

these snubbers was visually inspected and functionally tested.

Due to the problems encountered during the functional testing of the

first two steam generator hydraulic snubbers, all four main steam line

snubbers were declared inoperable and sent offsite to be rebuilt before

being tested. Two were found to have physical defects which would have

presented them from performing their intended function. One snubber had

a damaged reservoir, which was found by visual inspection. The second

snubber had a compression side poppet and spring installed backwards and

would only have been able to carry load in the tension direction. The

seals in all of the snubbers were found degraded. This degradation alone

would not have caused the snubbers to restrict thermal growth, but would

have affected the capability of the snubbers under severe dynamic or

seismic events. Following overhaul, the snubbers tested satisfactorily.

At the time of the inspection the licensee was still evaluating the

effects of the failed snubbers on system components. The analysis had-

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apparently only been underway for about a month and appeared to have been

prompted by an inquiry from plant engineering to downtown engineering.

The need to~ perform this analysis required by the technical

specifications had not been identified as an open item on any of-the

licensee's tracking systems. The results of the licensee's investigation

will be furnished to the NRC as a special report, according to the

licensee.

Preliminary analysis indicated that the piping had not been affected by

the failed mechanical snubbers.

As a separate issue, the licensee has been meritoring unusual pipe

movements of the pressurizer surge line since 1982. A walkdown of this

line at the beginning of the 1985 refueling outage revealed additional

. movement had occurred. A consultant was hired to evaluate and analyze

the movements of the. pressurizer surge.line. The consultant analyzed

various potential causes of the observed movement. It was determined

that none of the potential causes, either alone or combined, could have

produced the forces required to ' result in the observed movement.' The

consultant was advised that some problems.had been encountered in testing

the steam generator snubbers. The licensee directed the consultant to

analyze the surge line movements using the. worst-case assumption that the

snubbers may have been locked. The preliminary analyses, which were

completed in November 1985, revealed that locked-up snubbers could have

produced the movement necessary to displace the surge line as observed.

Based on this finding, further worst-case analyses of reactor coolant

loop thermal expansion with locked-up snubbers was to be performed to

' demonstrate the structural integrity of the Reactor Coolant System (RCS),

and its associated supports. The analyses was to be performed under the

rules of Subsection NB-3600 of'Section III of the ASME Boiler and

Pressure-Vessel Code. The licensee subsequently stated that the

worst-case analysis was on the B reactor coolant loop and revealed the

stress at the elbow where the B RCS hot leg enters the B steam generator

would be in excess of the yield stress for the material. Subsequently, a

plastic analysis was performed for this elbow in accordance with

Subsection NB-3228 of Section III, with strain acceptance criteria as

specified in Appendix T of Code Case N47. This' analysis revealed the

strain in the elbow due to thermal expansion loads would be less than the

one percent limit specified in Appendix T of Code Case N47. The fatigue

usage factor was determined to be less than 0.1 based on 30

heatup/cooldown cycles.

This work will be followed up as open item 85-39-02.

11. Exit Interview

The inspectors met with the plant general manager and members of his

l staff at the conclusion of the inspection period. During this meeting,

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the inspectors summarized the scope and findings of the inspection.