IR 05000344/1986007

Insp Rept 50-344/86-07 on 860211-0317.No Violation or Deviation Noted.Major Areas Inspected:Operational Safety Verification,Corrective Action,Maint,Surveillance & Preparations for Refueling

ML20202H848
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	03/28/1986
From:	Dodds R, Kellund G, Richards S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20202H723	List: IR 05000344/1986007 ML20202H718
References
50-344-86-07, 50-344-86-7, NUDOCS 8604150469
Download: ML20202H848 (11)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION V

Report No. 50-344/86-07 Docket No. 50-344 License No. NPF-1 Licensee: Portland General Electric Company 121 S. W. Salmon Street Portland, Oregon 97204 Facility Name: Trojan laspection at: Rainier, Oregon Inspection conducted: February 11 - March 17, 1986 Inspectors: 5 3/27/96 S. A. Richards Date Signed Senior Resident Inspector Db1O rc G. C. Kellund 247/sc Date S'igned Resident Inspector I Approved By: r 3 /T R. 7.Dodds, Chief D4te Signed Reactor Projects Section 1 Summary:

Inspection on February 11 - March 17, 1986 (Report 50-344/86-07)

Areas Inspected: Routine inspection of operational safety verification,

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l corrective action, maintenance, surveillance, preparations for refueling, the inservice testing program, review of various aspects of plant operation, and

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followup on previously identified items. Inspection procedures 30702, 30703, l 40700, 60705, 61700, 61726, 62702, 62703, 71707, 71710, 92701, 93702, and 94702 were used as guidance during the conduct of the inspectio Results: No violations or deviations were identifie PDR ADOCK 05000344 G PDR

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DETAILS

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I Persons Contacted

• W.S. Orser, Plant General Manager

• R.P. Schmitt, Manager, Operations and Maintenance D.R. Keuter, Manager, Technical Services J.K. Aldersebaes, Manager, Nuclear Maint. and Construction J.D. Reid, Manager, Plant Services R.E. Susee, Operatione Supervisor D.W. Swan, Maintenance Supervisor A.S. Cohlmeyer, Engineering Supervisor G.L. Rich, Chemistry Supervisor T.O. Meek, Radiation Protection Supervisor S.B. Nichols, Training Supervisor D.L. Bennett, Control and Electrical Supervisor C.H. Brown, Quality Assurance Operations Branch Manager R.W. Ritschard, Security Supervisor H.E. Rosenbach, Material Control Supervisor The inspectors also interviewed and talked with other licensee employees during the course of the inspection. These included shift supervisors, reactor and auxiliary operators, maintenance personnel, plant technicians and engineers, and quality assurance personne * Denotes those attending the exit intervie . Operational Safety Verification During this inspection period, the inspectors observed and examined activities to verify the operational safety of the licensee's facilit The observations and examinations of those activities were conducted on a daily, weekly, or biweekly basi On a daily basis, the inspectors observed control room activities to verify the licensee's adherence to limiting conditions for operation as prescribed in the facility technical specifications. Logs, instrumentation, recorder traces, and other operational records were examined to obtain information on plant conditions, trends, and compliance with regulations. On occasions when a shift turnover was in progress, the turnover of information on plant status was observed to determine that all pertinent information was relayed to the oncoming shif During each week, the inspectors toured the accessible areas of the facility to observe the following items: General plant and equipment condition Maintenance requests and repair Fire hazards and fire fighting equipmen Ignition sources and flammable material contro l

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2 i Conduct of activities in accordance with the licensee's administrative controls and approved procedure Interiors of electrical and control panel Implementation of the licensee's physical security pla Radiation protection control Plant housekeeping and cleanlines Radioactive waste system Proper storage of compressed gas bottle 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 evolution During 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 aligned with the work activities involve The inspectors examined the licensee's nonconformance reports (NCR) to confirm that deficiencies were identified and tracked by the syste Identified nonconformances were being tracked and followed to the completion of corrective action. NCRs reviewed during this inspection period included 85-088,86-004, and 86-007. Specific comments concerning NCR control are contained in paragraph 12 of this repor Logs of jumpers, bypasses, caution, and test tags were examined ,by the inspector Implementation of radiati6n. protection controls was verified by observing portions of area surv,eys being performed, when possible, and by examining radiation work permits currently in,effect to see that prescribed clothing and instrumentation were available and u' sed.-

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Radiation protection instruments were also examined to verifytoperability and calibration statu ' #-

The inspectors verified the operabilityfof" selected engineered safety

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

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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 safety injection system, the containment spray system, and the auxiliary feedwater syste No violations or deviations were identifie . Corrective Action 4 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 E

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. 3 minutes, and Quality Assurance Program Nonconformance Reports. The

' inspectors concluded.that the licensee's systems were being utilized to correct identified deficiencies. Plant Review Board meetings were-attended by'the inspectors on February 12 and February 26. The inspectors verified that the appropriate committee members were present at the meetings and that the meetings were conducted in accordance with the requirements of section 6.5.1 of the facility technical specification No violations or deviations were identifie . Maint'enance During this inspection period, the inspectors witnessed corrective maintenance performed on March '5 on the 'B' train auxiliary feedwater (AFW) system flow throttle valves. This maintenance was conducted due to the discovery of an incorrectly coupled flow throttle valve in the AFW system 'A' train. Details of this event are contained in paragraph 10 of this report. .The specific goal of this maintenance activity was to ascertain whether the couplings between the valve motor operators and valve stems for the 'B' train valves were securely made up and to correct any deficiencies noted. The inspectors made the following observations:

- The maintenance mechanic performing the work possessed a maintenance request which properly described the work to be performe The 'B' train of AFW was declared inoperable prior to performing the wor The maintenance request required an independent quality control inspection of each valve af ter work on the valve was complete. The-inspector witnessed the quality control inspectio In addition to the NRC inspector and th'e quality control . inspector, the work was witnessed by the mechanical maintenance supervisor and the

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assistant shift supervisori ,

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No violations or deviations' werciobserved by the Inspectors with regard to this maintenance activity. -

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The surveillance testing of safety-rela,ted systems'was witnessed by the inspectors. Observations by proper procedures were used,1,the -inspectors ' included test' instrumentation verificationand was calibrated thatthat 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 appropriate acceptacce criteria. One test witnessed was a_special test conducted on March 4, to ensure that proper 'A' train AFW flows would be achieve during an automatic initiation of the system. This test resulted in adjustments being made to the limit switch positions for several-of the

--flow throttle valves. The inspectors witnessed portions of the-adjustments being made. The other two tests witnessed required-no i

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. 4 corrective action due to the results. These tests were associated with verifying the proper operation of the 'A' train emergency diesel generator (EDG) on February 14, and verifying proper operation of the EDG fuel makeup system on March 14, following a system modificatio No violations or deviations were identifie . Followup On Previously Identified Items Iollowup Item 84-08-02 (Cloced): Preferred instrument bus electrical lineup requirements. The licensee has revised Periodic Operating Test (POT) 21-2, "ESF and Offsite Power Availability," to clearly require that a preferred instrument bus that is powered off of either bus Y01 or YO2, be declared inoperabl Followup Item 85-16-04 (Open): Local leak rate testing of the fuel transfer tube. The licensee has reviewed the testing requirements associated with the fuel transfer tube and has concluded that the type A testing normally performed by the licensee is adequate. The inspectors have concluded that 10 CFR 50, Appendix J, requires the fuel transfer tube to be type B tested because the tube is a piping penetration fitted with expansion joints. The inspectors have requested a review of the requirements by the.NRC Office of Nuclear Reactor Regulation (NRR). This item will be further acted upon, pending the. completion of NRR's review.

' Followup Item 85-39-02-(Closed): Failed mechanical and. hydrauli snubbers. The licensee has submitted Licensee Event Report (LER) 85-13, which provides a detailed description of'the-surveillance testing of the snubbers. This item will be reviewed via the normal followup of LER LER 84-13 (Closed): Pressurizer. level transinitter calibration error The inspectors reviewed the calculations performed to properly scale the pressurizer level instruments." No deficiencies were identified. With regard to the programatic control of instrumentation scaling information, the licensee has initiated several:' actions:

- The licensee's corporate-engineering department is reviewing the scaling of all safety-related instrumentation. The effort is scheduled for completion by approximately July,-198 The corporate engineering department is now responsible for providing scaling information for all new instrumentatio The M-500 drawings will be updated, such that these drawings will be the controlling document for instrumentation dat The licensee has conducted. training on proper _ scaling of instrumentation data. An outside contractor was retained to provide this instructio .Through review'of the above, the inspectors concluded that the licensee

.has taken adequate' corrective action to ensure that the scaling of safety-related instrumentation is correc . . ._

. 5 LER 84-16, Revision 1 (Closed): ' Reactor trip and safety injection. This revision to the original LER provided additional information concerning the cause of the emergency diesel generator trip, which occurred following the safety injection. The licensee has' installed protective cover plates to prevent the crankcase pressure trips from being activated by oil sling within the engine. The other details of this event have been previously addressed and closed by the. inspector LER 85-07 (Closed): Nuclear instrument rate > trip setpoints improperly se The licensee has reset the rate trip'setpoints in accordance with

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the guidance provided in Westinghouse' Technical Bulletin NSID-TB-85-1 After taking this action, the plant experienced'a negative rate, reactor trip during a main turbine runback, indicating that the trip was set excessively conservative. The licensee then changed the plsnt setpoint such that less margin exists to the technical specification allowable limit, however, calculations by the licensee indicate that a negative rate trip'would still likely occur during a turbine runback or a single dropped rod event. The licensee has initiated action to encourage the nuclear steam supply system vendor to correct this discrepanc LER 85-13 (Open): High failur.e rates during snubber inservice testin This item has been previously discussed in Inspection Reports 85-39 and 86-0 During this inspection period, the licensee informed the inspectors that the steam generator column supports will be inspected during the 1986 outage to verify that excessive stresses were not placed on the reactor coolant system piping due to snubber problems. The inspectors will follow the results of the licensee's inspection effor LER 85-14 (Closed): Automatic actuation of the 'B' emergency diesel i

generator (EDG). The EDG automatically started due to a loss of power to the V-82 bus, a 230 KV switchyard bus. The licensee determined that two blown fuses caused the loss of power to the V-82 bus, however, the cause of the blown fuses was not found. The licensee plans additional action related to this event, however, because the EDG functioned per design and the redundant source of offsite power was maintained, the inspectors concluded that no additional NRC followup of this event is warrante No violations or deviations were identifie . Preparation for Refueling Prior to the unit shutdown for the annual refueling outage, the inspectors reviewed the licensee's outage schedule and examined procedures pertaining to the following areas:

FHP-1 New Fuel Transfer FHP-5-1 Refueling Organization FHP-5-2 Refueling Procedure FHP-5-3 Reactor Vessel Head Removal and Installation FHP-5-13 Fuel Shuffle and Position Verification FHP-13 Fuel Handling Emergency Procedures

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The procedures appeared adequate to control and coordinate.the refueling ~

activities. Additionally, the inspectors observed new fuel receipt activities to ensure proper assembly identification, conduEt of adequate -

radiation surveys, inspection for physical damage and proper hr.ndling activities.

No violations or deviations were id2ntifie ,

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8. Inservice-Testing Program .,

$ The' inspectors reviewed the-licensee'c inservice testing.(IST) program to i

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, determine the status.of changes made subsequent to the.NRC inspection of l ,

this area conducted in June, 1985 (Inspection Report 50-344/85-20). The ,

review included an examination of the IST program manual, review of  ;

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selected Periodic Operating Tests and Maintenance Procedures that-

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implement the IST program, examination of a sample of' pump and valve test-I data, and review of the licensee's equipment performance trending

program.

j The inspecto'rs' review of the IST program manual indicated that the

manual had not been' updated to reflect the current state of Code j exemptions granted since 1982. The inspectors discussed this matter with .

I the Plant Test Engineer (PTE) responsible for IST. He stated that the manual-is in the process of being updated and that a revised manual that incorporates all relief requests to date is complete and is awaiting i

final NRC approval of outstanding relief requests prior to issuance of i i the manua '

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During the previous inspection of this program, a concern was raised l i

regarding use of gauges for testing that did not meet the ASME Section XI or the licensee's procedural requirements for proper range. The plant 3 was not originally designed to accommodate inservice testing, and the

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gauges in place were intended for operational use rather than for test purposes and did not meet the Code specifications for range. The PTE

stated that a Request for Design Change has been initiated to' install i

gauges that meet the'ASME Section XI range specification alongside the

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j gauges currently in place.

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Several procedural problems were identified in.the-previous-inspectio The IST procedures for testing the containment spray pumps (POT-4-1),

component cooling water pumps (POT-8-1), and the residual heat removal

pumps (POT-16-1)

did not preestablish either flow or differential pressure to a reference value. The inspectors reviewed these procedures and verified that POT-4-1 has been revised to preestablish a reference

,; flow. .The inspectors, discussed the status of the:other two procedures

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with the PTE who stated that revisions to these procedures have been j written and are in the review cycle. ' Additional procedural problems identified previously- concerned safety and relief valve testin Procedures MP-5-1, " Pressurizer Safety Valve Inservice Test" and MP-12-15, Safety-Related Safety Valve Inservice Testing" appeared to;

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allow averaging of lift test values to determine acceptability. The

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inspectors reviewed the subject procedures and verified that MP-5-1 was

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revised to preclude use of average lift valuesifor acceptability. ~ In - _

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$ addition, the PTE stated that MP-12-15 was also being revised to address

this concern.

j During review of pump test data, the inspectors observed that reference values were reestablished quite frequently for many pumps. In discussing

this issue with the PTE, he stated that this was required in the past in

- order to bring the acceptance criteria for the pump tests in conformance

! with the Code guidelines by creating uniform-test conditions. In

} addition, recent system flow alterations performed.on certain systems

, such as the service water system have_ necessitated changes in pump reference values. The-inspectors did note, however, that changes in '

i reference values are_ currently being made less frequently than in the

. past. The inspectors expressed a concern that changes in pump reference values makes trending of equipment performance difficult. The licensee's

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current method for trending equipment performance consists of a binder maintained by the PTE. He enters test data from each test of.'a component and uses this information to monitor the component's performance. The I-PTE did indicate that the licensee is considering utilizing personal

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computers to more effectively trend equipment performance.

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The inspectors concluded that the=IST program has improved over the past year. The licensee's plans for additional personnel support for the PTE

, should produce further improvements in the implementation of the program, and additional training of personnel involved >with the' program would be beneficial as wel >

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No violations or deviations were identified.

t Surveillance of Containment Isolation' Valves '

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Duringinspectionofcontainmentisolationvalv'es,Nheinspectorsraised f

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the concern that test vent and drain valvesilocate'd(on containment-penetrations between the containment wall and.the outer isolation valve (or. blind flange) may be required ,to be verifie'd a'siclosed 7 once'every 31 days, on penetrations required torbe c1'osed during accident conditions, per Technical Specification'4.6.1;1.a.1.' The inspectors determined that

the licensee was not including ~these jalves on-Periodic;0perating> Test (POT) 3-3, " Containment. Penetration' Valve Inservice Test," which implements the surveillance requirements of Technicalt Specification .

4.6.1.1.a.1. This was discussed _with'lic'ensee representatives;who stated

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that the licensee's position was*that' penetrations capable of being

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closed by automatic containment' isolation ialves an'd required"to be

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closed during accident conditions are not. governed by Technicalk '

Specification 4.6.1.1.a.1 and therefore,. test vent and drain valves on, these penetrations are excluded from monthly position' checks. The

-inspectors felt that these valves should be included in the monthly

surveillance to verify integrity of the subject' containment penetration The licensee committed to provide the NRC Region.V office with a letter i outlining the formal company position on this issue'.- The Region V Office

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will contact the Office of Nuclear Reactor _ Regulation (NRR) to obtain an interpretation of the technical specification requirements concerning

these test vent and drain valves, and any Region V action will await this interpretation. 'In the interim, the licensee has verified that these valves are closed and has committed'to include them in the monthly

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position checks performed by FOT-3-3 until the issue is resolved. This is an unresolved item (86-07-01).

No violations or deviations were identifie . Auxiliary Feedwater Valve Coupling Problems On March 3, 1985, a licensee maintenance mechanic commenced action to perform routine preventative maintenance on auxiliary feedwater (AFW)

flow control valve CV3004D1. This valve is a motor operated valve which is normally positioned in a throttled position to control 'A' train AFW flow to the 'D' steam generator, should the AFW system receive an automatic start signal. There is a similar valve associated with the ' A'

train AFW flow path to each of the four steam generators. There are also four valves performing the same function for the 'B' train of AFW, making a total of eight AFW flow control valves. These valves have only one automatic function, which is to close if a high flow rate to their associated steam generator is sensed by installed instrumentatio The licensee mechanic took manual control of the valve operator to fully backseat the valve prior to commencing his work. While bringing the valve to the backseat, the valve stem came out of the coupling which connects the valve stem to the motor operator shaft. The problem was brought to the attention of the shift supervisor and the maintenance supe rvisor. The other three valves in the 'A' train were then visually inspected. The valve to the ' A' steam generator appeared properly made up, however, the other two valves appeared to have couplings that were not properly assembled. This information became available during the evening of March 3, with the plant operating at 100 percent power. Plant management directed personnel to commence action to rework the 'A' train valve couplings as necessary to properly mate the valve operators to the valve stems and then to test the system to determine the proper throttle position for the valves. The 'A' train of AFWehad been declared inoperable due to the original maintenance wor In accordance with the

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technical specifications, one train may be removed from service without affecting plant operation.for up to 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> With regard to the 'B' train valves, plant management decided not to visually inspect these valves until the 'A' train was returned to service, apparently based on the assumption that access to the valve l couplings was prevented by seismic supports that'couldlnot'be removed without rendering the 'B' train inoperable' . Further, the valves had successfully passed routine surveillance tests'and had functioned normally during relatively recent plant startups and trip Later, the licensee determined that access to' the 'B' train valves was not restricted by supports andsthat visual inspections of the couplings were

. possible without affecting the valve operabilit Late in the morning of March 4, with the ' A' train still' inoperable, plant management ordered reactor power reduced to 35 percent. Plant personnel had by then visually inspected the 'B' train and considered the

'B' train still operable, however, plant power was reduced as a conservative measure until a detailed inspection of the 'B' train valves could also be performed. The 'A' train valves were repaired, tested, and

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. 9-returned to service by the evening of March 4. The 'B' train valves were-then closely inspected and all four couplings were. determined to be functional. The inspectors observed the testing performed to reset the throttle ~ positions of the 'A' train valves and witnessed the maintenance inspection of the 'B'z train couplings. The licensee's corporate quality _

assurance organization reviewed the plant history of these valves an concluded that' the couplings that required rework were probably assembled incorrectly in 1975,.at the completion of plant. construction. They also examined other valves in the-plant and determined that the specific coupling used in the'AFW flow control valves were not in generic use in the plan Both during and after this event, NRC personnel expressed concern that licensee management was not_more aggressive in determining th6 status of theB' train valve couplings. Although; the plant staff originally thought. that the valves were inaccessible for' inspection, a more thorough review would ha.4 revealed that a preliminary inspection could be easily perfo rmed. NRC personnel conveyed their expectations to licensee management that when operational readiness of; safety-related equipment comes into question, action to ascertain that equipment operability should be as prompt and thorough as possible.< - ,

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The inspectors' review of this event" concluded th'at'at all times,_thelAFW system was capable of providing water to the; stea'm generato'rs for decay heat remova k ,'

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No violations or deviations were identifie .

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1 Steam Generator Nozzle Dam Safety Evaluation Review . - *

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's While reviewing preparations for the upcoming-refueling outage, the inspectors examined the use of: steam' generator nozzle dans during the 1985 refueling outage. The nozzle-dams ~ar,e# designed.to fit into'the hot

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leg and cold leg nozzles of a steam; generator and serve to isolate the channel head from the -primary coolant, loop. This allows the reactor-cavity to be flooded for refueling' operations while permitting concurrent'

steam generator work. The licensee contracted -Nuclear Energy Services (NES) to provide the nozzle dans to be used_during the.1985 outage'. . The inspectors specifically examined the safety evaluation addressing-failure of the NES design nozzle-dans and the timeliness of-the evaluation. The documentation associated with the use of the nozzle dans indicates that extensive preplanning of the job occurred, however, the associated formal safety _ evaluation does not appear to have been approved in a timely manner. The'dans in the 'B' steam generator were installed on May 13 and -

the reactor cavity was partially flooded on May 15 when the dans were discovered to be leaking and.were.subacquently removed. .The' Plant Review Board, however, did not review the pro:edure for using the dans'and the associated safety' evaluation until May 15, the same day they were put into use. This last minute action indicates that the communication and coordination surrounding the installation of the dams.was not well-

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The inspectors did note that the Trojan Nuclear Operations Board requested a staff review of the nozzle dam installation attempt, and this

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. 10 review identified similar concerns with the timeliness of completing required actions. The licensee also stated that the dams wculd not be

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used during the 1986. outage. Nonetheless,.the inspectors concluded that '

both the' licensee's and the; inspectors' findings' indicate an increased need for licensee management to ensure that formal safety evaluations and procedures are completed and reviewed well in advance of implementation of plant changes where circumstances allo '

No violations or deviations were identifie .

12. Miscellaneous Observations While reviewing the licensee's active nonco formance ' reports (NCR), the inspectors observed that two NCRs were approximately one month old and had not yet received an initial engineering evaluation. The inspectors contacted the action engineer and determined that one of :the NCRs had the initial evaluation completed, however NCR 86-004 was still being considered. The inspectors further determined that the procedure for processing an NCR does not provide guidance concerning the timeliness of the initial engineering evaluation. .The inspectors discussed their concern with the plant manager and the operations quality assurance-branch manager, that NCRs should receive a high priority during their

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initial review and that delays in p'erforming the initial review should be brought to the attention of management. The operations quality assurance branch manager stated that a revision in process to the NCR procedure may address this concern, however, the revised procedure was not available during the reporting period. This item will be further reviewed during a

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future inspection period (344/86-07-02).

No violations or deviations were identifie . Unresolved Item

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Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of '

noncompliance, or deviations. An unresolved item disclosed during the inspection is discussed in paragraph . Exit Interview The inspectors met with the plant general manager and the manager of operations and maintenance at the conclusion of the inspection perio During this meeting, the inspectors summarized.the scope and findings of

the inspection.

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