IR 05000344/1987027

Insp Rept 50-344/87-27 on 870622-26.No Violations or Deviations Noted.Major Areas Inspected:Usi A-26 & Followup & Closure of Previous Insp Findings

ML20235Y298
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	07/09/1987
From:	Mendonca M, Pereira D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20235Y285	List: IR 05000344/1987027 ML20235Y283
References
REF-GTECI-A-26, REF-GTECI-RV, TASK-A-26, TASK-OR 50-344-87-27, NUDOCS 8707250191
Download: ML20235Y298 (7)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION V

Report No: 50-344/87-27 Docket N License No. NPF-1 Licensee: Portland General Electric Company 121 S. W. Salmon Street Portland, Oregon 97204 Facility Name: Trojan Nuclear Plant Inspection at: Rainier, Oregon Inspection cond ct d: 26, 1987 Inspector: . ayl Mt #  ?? Y D7 B. Pereira, Reactor Inspector Date Signed Approved by:

M. M. Mendonca, Chief, Date Signed Reactor Project Section 1 Summary:

Inspection During the Period of June 22-26, 1987 (Report 50-344/87-27)

Areas Inspected: This routine, unannounced inspection by the Project Inspector involved the areas of Temporary Instruction 2500/19 (Unresolved Safety Issue A-26: Reactor Vessel pressure transient protection for pressurized water reactors), and follow-up and closure of previous inspection finding During this inspection, inspection modules 30703, 25019, and 92700 were use Results: No violations or deviations were identifie I l

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DETAILS Persons Contacted Licensee Personnel

. C. A.'Olmstead,' Plant Manager

• R. P. Schmitt,. Manager, Operation; and Maintenance C.~ Jarman, Manager,, Quality Assurance Departmen >

J. D."Reid, Manager,' Plant Services C. H. Brown, Operations Branch Manager, Quality Assurance 14; Ankrum,-Supervisor, Nuclear Regulation' Branch G. A. Zimmerman,' Manager, Nuclear. Regulation Branch

• D. Nordstrom, Engineer, Nuclear Safety and-Regulation Department

• W. L. Kershul,. Engineer, Nuclear Safety and Regulatio .

Department D. L. Bennett, Supervisor, Control and Electrical R. Reinart, Supervisor, Instrument and Control

• D. W. Swan, Manager, Technical Services R. Russell, Assistance Operations Supervisor, Quality Assurance U. S. Nuclear Regulatory Commission G. Suh '

.c.' Oregon Department of Energy

• H. Moomey, Oregon Resident Inspector

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• Attended the Exit Meeting on June 26,.198 . Temporary Instruction 2500/19 Unresolved Safety' Issue A-25: Reactor Vessel Pressure Transient Protection for Pressurized Water Reactor The purpose of this inspection was to verify that the licensee has -

developed and implemented an effective mitigation system for low-temperature ~ overpressure transient conditions in accordance with their commitments concerning Unresolved Safety Issue (USI) A-2 The background of USI A-26 is that a technical issue was identified concerning the safety margin-to-failure for pressurized water reactors (PWR) should they be subject to severe pressure transients while at a relatively low temperatur The majority of the transients that occurred were during startup and shutdown operations when the reactor coolant system (RCS) was in a water-solid condition (i.e., no steam bubble present in the pressurizer to act as a surge volume). During such conditions, the RCS is susceptible to a rapid increase in system pressure through thermal expansion of the RCS water or through injection of water into the systems without adequate relief capacity or discharge flow path to control the pressure increas _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ - _ _

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Plants receiving an operating license before March 14, 1978, committed to design reviews, procedure changes, equipment modifications, operator training, and surveillance using a combination of operator personnel and automatic equipmen The Trojan Overpressure Mitigation System (OMS) consists of two separate trains, each containing a power-operated relief valve (PORV), an

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isolation valve and associated circuitry. When in the low pressure mode the system provides low pressure setpoints of 440 psig for valve PCV-455A and 490 psig for valve PCV-456 for the respective trains. When the-system is enabled, it will terminate all analyzed pressure transients below the Appendix G limit by automatically opening the PORV A manual switch is used to enable and disable the low pressure setpoint of each relief valv An enabling alarm which monitors system pressure is provided to alert the control room operator to enable the OMS when system pressure drops to a predetermined point (375 psig). In addition, an alarm is provided in the control room to indicate when an overpressure transient is occurrin Supplementing the protection provided by the OMS is the low temperature overpressure protection provided by the residual heat removal (RHR)

system. The RHR system has four safety relief valves. One has a setpoint of 450 psig with a relief capacity of 900 gpm, and the other three have setpoints of 600 psig with each one having a relief capacity of 20 gp The RHR system also has an independent letdown flow path from the reactor coolant system (RCS) to the volume control tank or one of th holdup tank A review of the design of Trojan's OMS and single failure criteria indicated that the Trojan's OMS meets this criteria without requiring '

operator action for all cases reviewed except where the initiating event is a loss of power from one 125V DC bus during heatup or cooldown when a charging pump is running and when' the RHR system is not in operatio This event a.t this time would result in isolation of the letdown line and one PORV failing to open upon demand. When a single failure is postulated for the other PORV there is no overpressure protectio In the analysis of this scenario, the licensee indicated that during all of the normally encountered plant cooldown and heatup conditions that there would either be a vapor space in the pressurizer or the RHR system would be in service. The vapor space.in the pressurizer would provide a buffer against overpressurization of the RCS, which would allow the operators time to take corrective action to prevent exceeding the Appendix G limits. The amount of time provided to the operator is dependent on the water level in the pressurize At the normal no-load level of 30% there would be 1260 ft of vapor space, which would result in an 18% pressure increase 10 minutes after the operator was alerted to the transient by various alarms. This would not exceed Appendix C limit During cooldown, the RHR system is normally placed in service prior to collapsing the pressurizer bubble and is not normally removed from service, during heatup, until after a steam bubble has been established in the pressurize The RHR system provides protection from this scenario in two ways; first by providing a second letdown flow path and

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i c second by providing relief through the'900 gpm safety valv The capacity is sufficient to mitigate the pressure transient resulting from-

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._ isolation of letdown at Troja i The limited time in Mode _4 operation (approximately a 2-hour period during heatup and cooldown) coupled with the likelihood of the failures

.(loss of DC bus when a charging pump is running with PORV failure) makes ,

this'a low' probability even j

~The inspector reviewed the administrative controls and procedures for th OMS and determined the following items: The licensee's procedures allow the plant to be operatedonly'with a steam bubble in the pressurizer at all times except for ' hydrostatic tests.- This effectively minimizes.the time in a water-solid condition. General. Operating Instruction (G01)-1, entitled: " Plant Startup from Cold Shutdown to Hot Standby," Revision 31, in ste .4, establishes a steam bubble.in the pressurizer prior to plant

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heatup and prior to' RHR removal from service. GOI-4, entitled:

" Plant Shutdown from Hot Standby to Cold Shutdown," Revision 20, in 'I step 3.15.3, places the OMS in service prior to' collapsing the steam-bubble in' the pressurizer in step 3.1 I The licensee's procedures restrict the number of-high pressure safety injection (SI) charging pumps to no more than one when the

' reactor coolant system is in'the low-temperature overpressure condition. GOI-4 in step 3.15.4 disables one SI pamp by closing the pump discharge valves, placing the pump control switch in pull-to-lock, and racking out the' pump breake The licensee's operators are alerted since an alarm at 375 psig occurs in the control room if the OMS is not enabled.- License' Amendment #78 to the Technical Specification provides-justification that the plant installed system is in accordance with the plant license, The power-operated relief valves (PORVs) are spring-loaded-closed, air-required-to-open valves, which are supplied by a co~ntrol air source. To assure operability of the valves upon loss of control air, a barkup air supply is provide The backup air supply consists of a seismically qualified passive air accumulator for each PORV. 'Each accumulator contains enough air to ensure. that it will provide 32 cycles of the PORV which will be sufficient to mitigate ;

an overpressure event for the 10 minutes during which no credit can be taken for operator actio The inspector reviewed the training and equipment modification concerning OMS and determined the following:

All operators as of the time of this inspection had received training concerning low-temperature overpressure (LTOP) event causes, the operation and maintenance of the OMS that mitigates the events, and the consequences of inadvertent actuatio The

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4 inspector interviewed the instructors, examined their briefing papers, and interviewed operators. This training was conducted during session #3 between October 27, 1986, to December 5, 198 No problems were discovered in the training are Permanent modifications and procedural changes have been made that result in a system that provides mitigation for reactor coolant system LTOP events. A permanent second setpoint of 440 psig has been installed on PORV PCV-455A, and a second setpoint of 490 psig has been installed on PORV PCV-456. Procedural changes have been-added to various operating instructions to tag out one of two SI pumps, shut their respective isolation valves, and rack out their breakers prior to an RCS cold leg temperature is < 290 F per Technical Specification 3.4. An enabling alarm which monitors system pressure. is provided to alert the control room operator to enable (UNBLOCK) the OMS when system pressure drops to 375 psig. In addition, an alarm is provided in the control room at 400 psig when an overpressure transient is occurrin The inspector reviewed the surveillance requirements and activities associated witn OMS and determined that PICT-17-1, Revision 9, performs the functional test of the Solid Reactor Coolant System OMS (Channel 403). This test is performed within 31 days prior to having an RCS cold leg temperature of 290 F or less, and at least once per 31 days thereafter while at such temperature. PICT-17-2 performs the functional test on channel 405. Channel 403 activates PORV PCV-45SA, and channel 405 activates PORV PCV-45 The inspector determined that the PORV electronics and setpoints are verified periodically and are performed to ensure operability of the system electronics before each cold shutdow Based on this inspector's review of the OMS, it appears that Trojan has an effective mitigation system for LTOP transient conditions in accordance with their commitments concerning USI A-26. TI 2500/19 is considered close No violations or deviations were identified in this inspectio . Follow-up on Previous Inspection Findings (Closed) Licensee Event Report No. 85-13, Revision 2, High Failure l Rates Snubber Inservice Testing i Licensee Event Report No. 85-13, Revision 2, describes that during-the 1985 refueling outage, snubber inservice testing was performed on both mechanical and hydraulic snubbers, which consisted of visual inspections and functional tests of the snubbers. As a result of this testing, a high percentage of PSA -1/4 and PSA-1/2 mechanical snubbers manufactured by Pacific Scientific were determined to be inoperable. Additionally, hydraulic snubbers manufactured by both Anker-Holth and Bergen-Paterson were found significantly degraded, primarily due to deteriorated seal _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ - _

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g This snubber inservice testing was conducted in accordance with Technical Specification 4.7.10 (issued as Amendment 102 to the

' Trojan Technical Specifications on February 6,1985), which requires visual inspections of all of the hydraulic snubbers and a minimum of 82 mechanical snubber In addition, 10 percent of each type of snubber is required to be functionally teste As a result of the functional testing, a number of snubbers were declared inoperable and were either rebuilt, retested, and I reinstalled or replaced with operable snubbers. In 1985, one hydraulic snubber was declared inoperable as a result of the visual inspections. Three mechanical snubbers failed the visual inspection in 1986 but were subsequently determined to be operable. For each snubber which failed a functional test, additional snubbers of that I type were teste The results of the testing and effects of the j failed snubbers are discussed adequately in the LE As a result of the 1987 refueling outage snubber testing, 100 percent of the hydraulic snubbers were examined and were determined to be satisfactor In addition, the functionally tested snubbers were satisfactory.

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Based on those results, the 1988 refueling outage snubber testing will consist of the Technical Specification

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sampling basis which will require inspecting all the hydraulic '

snubbers and a minimum of 82 mechanical snubber Based on this inspector's review of the LER and with discussions with licensee personnel, it is determined that LER No. 85-13, Revision 2, is close b, (Closed) Licensee Event Report 87-01 Technician Inadvertently Shorted Feedwater Pump Control Power-Feedwater Pump Trip / Reactor Trip Licensee Event Report 87-01 describes the circumstances that occurred on January 6, 1987, when the plant was operating at 100 percent power at normal operating temperature and pressure. At 1129, the "A" main feedwater pump tripped while maintenance was being performed on the pump control circuitry. This resulted in a reactor trip because the feedwater flow / steam flow misma'r.1 reactor trip bistables were in a partial trip condition due to routine instrumant surveillance testin !

The cause of the event was personnel error. While troubleshooting the "A" main feedwater pump control oil pressure indicator, an I&C

~ Technician shorted the pump control power supply. This caused the  ;

power supply fuse to blow which also provides power to the pump 1 overspeed control circui This resulted in actuation of the pump ]

overspeed trip and tripping of the pum )

The licensee's permanent corrective action consisted of the following:

(1) Evaluate rescheduling of PICTs performed concurrently with maintenance activities that create a high risk of inadvertent

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plant tri In letter DLB-0 M-87, dated June 12, 1987, Bennett recommends to R. SchmtT,t that PICTs with the highest risk performed during plant operation could be started before normal work hours, during which time scheduled electrical and I&C maintenance will not be performed. Typically, this will start one to.two hours before normal day shift. It would significantly reduce the risks associated with. concurrent instrument surveillance and on-line maintenanc (2) Evaluate establishing additional management controls of maintenance activities that are performed on inservice equipment which has a potential for tripping the plan In letter DLB-083-87, PICTs will continue to be scheduled well in advance and published on the Plan of the Day. Advanced prior to performance discussion between Operations and Maintenance supervision will occer before start of surveillanc In addition, each Maintenance Request will continue to be reviewed by the work group supervision as to the scope of the work and the risks involved. Additional emphasis will continued to be placed on deenergizing the equipment where that is possibl (3) This event was discussed with I&C Technicians and the importance of carefully planning instrument maintenance was emphasize (4) Evaluate improvements to the control room indication of automatic turbine runback. The turbine automatic runback circuitry was checked following the reactor trip. This was accomplished by resetting the circuitry and manually tripping the main feedwater pumps. The runback contacts in the local feedwater pump cabinets were verified to actuate properly. A Plant Configuration Change (PCC) will modify the annunciation in the control room for a turbine automatic runbac Based on the upon corrective actions, the inspector determined that Licensee Event Report 87-01 is closed, c. (0 pen) Licensee Event Report 87-07 Pressurizer Safety Valve Lifted Out-of-Tolerance During Surveillance Testing -

Licensee Event Report 87-07 describes that during a performance of pressurizer safety valve (PSV) testing on April 1, 1987, PSV-8010C lifted at 2596 psi This exceeded the Technical Specification allowed tolerance of 2485 psig +/-1% (i.e., 2460 to 2510 psig). The valve lift setpoint was adjusted and the valve retested satisfactoril The exact cause of this event is unknown. The licensee felt that inherent differences between bench testing and in place testing, in combination with setpoint drift, may have been the caus The immediate corrective action was to adjust the PSV setpoint and test the valve in place with acceptable result Test methods are j being reviewed to ensure that the correlation between bench testing l

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and in place testing is adequate. In place testing on PSVs 8010A and B will be performed prior to startup from the 1987 refueling outage. The last setpoint testing performed on these PSVs was via bench testing. This licensee event report will remain open pending the outcome of the in place testin . Exit Interview The inspector met with the licensee representatives denoted in paragraph 1 on June 26, 1987, and summarized the scope and findings of the inspection activitie ,

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