IR 05000423/1986018

Insp Rept 50-423/86-18 on 860520-0623.Satisfactory Performance Identified in All Areas.Major Areas Inspected: Plant Operations,Radiation Protection,Physical Security,Fire Protection,Ie Bulletins,Surveillance & Maint

ML20206R989
Person / Time
Site:	Millstone
Issue date:	06/25/1986
From:	Mccabe F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20206R970	List: IR 05000423/1986018 ML20206R966
References
50-423-86-18, IEB-86-001, IEB-86-1, NUDOCS 8607070273
Download: ML20206R989 (11)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /86-18 Docket N License N NPF-49 Licensee: Northeast Nuclear Energy Company P.O. Box 270 H_artford, CT 06101-0270 Facility Name: Millstone Nuclear Power Station, Unit 3 Inspection At: Waterford, Connecticut Inspection Conducted: May 20, 1986-June 23, 1986 Inspectors: J. T. Shedicsky, Senior Resident Inspector F. A. Casella, Resideht Inspector Approved by: OA. D 6/2s/86 E. C. McCabe, Chief, Reactor Projects Section 3B Date Inspection Summary:

Areas Inspectedt Routine on-site resident inspection (153 hours0.00177 days <br />0.0425 hours <br />2.529762e-4 weeks <br />5.82165e-5 months <br />) of plant opera-tions, radiation protection, physical security, fire protection, IE Bulletins, surveillance and maintenanc Results: This inspection identified satisfactory performance in all area I

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8607070273 060627 E'

PDR ADOCK 05000423 19

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TABLE OF CONTENTS P,a2e 1. S umma ry o f Faci l i ty Acti vi ti e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2. Review of Plant Operations........................................... 1 3. Review of Activities Occurring During the Inspection Period.......... 2 Service Water System Discharge Valve Wastage.................... 2 Potentially Radioactive Spill........... ....................... 2 Volume Control Tank Temperature Reduction In-Service Test....... 3 Potter-Brumfield Type MOR Relay Application..................... 4 "A" Emergency Diesel Generator Output Breaker Failure to Shut... 5 Pressurizer Safety and PORV 1eakage............................. 5 Interactions Between Seismic Class I Components................. 6 4. Observation of Maintenance........................................... 6 5. Observation of Surveillance Testing.................................. 6 6. Review of Licensee Event Reports (LERs).............................. 7 7. Steam Binding of Auxiliary Feedwater Pumps (TI 2515/69).............. 7 8. IE Bulletin 86-01 Review for Westinghouse PWRs....................... 8 9. Management Meetings.................................................. 9 d

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DETAILS 1. Summary of Facility Activities The plant was at 100% power for all but about 2.5 days during this report perio Power was reduced for about 2 days for weld repairs to a feed heater shell relief valve pipe connection. Other power reductions were required for thermal backwash of main circulating water bays and testing of turbine stop, control and intercept valves. With one pressurizer power operated relief valve (PORV) already isolated because of leakage, the other PORV was isolated during this period due to leakage. Pressurizer safety valve "A" continued to leak at about 0.4 gallons per minute. Reducing the temperature of Reactor Coolant Pump seal water rcduced seal leakage from 6-7 gpm to 3-4 gpm. The licensee made progress in getting CO2 fire suppression systems operational and clearing up problems with the radiation monitoring system. Plant clean-liness improved as major maintenance and Betterment Engineering projects were reduced in number. At the end of this report period, the plant had been on line for 42 consecutive day . Review of Plant Operations The inspector observed plant operations during regular and back shift tours of the following:

Control Room Fence Line (Protected Area)

Auxiliary Building Yard Areas Diesel Generator Building Turbine Building Intake Structure Vital Switchgear Areas Main Steam Valve Building Electrical Tunnels Waste Disposal Building The control room tours included observation of parameters related to Technical Specification requirements. Alarm conditions in effect and alarms received at the control room were reviewed and discussed with the operators, who were cognizant of board conditions. Shift manning was compared with Technical Specification Plant housekeeping controls were observed, including control of flammable materials. Posting and control of radiation and contaminated areas were inspected. Also, during plant tours, the various logs in the Con-trol Room, Chemistry department, and Health Physics deoartment were reviewe Posting and attentiveness of fire watches were checked. In addition, the in-spector observed selected actions concerning site security including personnel monitoring, access control, placement of physical barriers, alarm station operations, and compensatory measures. No deficiencies were identifie The inspector attended Plant Operations Review Committee (PORC) meetings on May 20 and 2 Technical Specification requirements for attendance were me The meeting agendas included actions to correct interaction between two Class I seismic components. The meetings were characterized by frank discussions and questioning of cause and corrective action In particular, PORC atten-

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tion was given to confirming plant conditions related to several Technical Specifications LCOs and Action Statements. Individual members' opinions were encouraged. No deficiencies in PORC performance were observe . Review of Activities Occurring During the Inspection Period Service Water Pump Discharge Valve Wastage In early April 1986, the licensee discovered the rubber coating on the interior of the service water pump discharge butterfly valves was de-teriorating and the carbon steel valve bodies were corroding. At that time, the licensee elected to replace the valves. Since these are long lead time items, a program for monitoring wall thickness reduction using ultrasonic thickness gaging was instituted. The inspector has been monitoring the results of the measurement program. A majority of the measured points show thickness reduction between 0 and 0.05 inches, with an instrument accuracy of +/- 0.05 inches. The worst case is a 0.15 inch loss at a nominal 4.00 inch point, reducing wall thickness to under the 3.875 inch minimum wall thickness specified on vendor drawing However, licensee engineering has evaluated the valve design for its current ser-vice and determined that the minimum acceptable wall thickness is 3 inches. The new valves are being expedited and should arrive in time for the planned mid-cycle outage this autumn. The inspector will con-tinue to monitor licensee's activities in this are Potentially Radioactive Spill On May 22, Low Level Waste Tank LLWT-A was overflowed to the waste build-ing sump. At the time, the sump pumps were lined up to that LLWT due to a temporary filtration system line-up, and the sump overflowed through the floor drains. About 40 feet of the lower level waste building and 3 feet of the lower level auxiliary building passageways were covered with approximately 30 gallons of potentially contaminated water before the problem was discovered. The spill was stopped and the area was roped off. Wet swipes showed activity levels less than background. No per-sonnel contamination occurre No release occurre Cleanup was com-pleted on May 23. A final swipe survey confirmed that no contamination was presen The cause of the overflow was a lack of Operator attentiveness during the evolution. The plant equipment operator (PEO) was.not in the imme-diate vicinity when the drains began backing up. A contributing factor was the fact that the radwaste PE0 watch had been turned over twice dur-ing that shift due to illnes It appears that complete turnover infor-mation was not provide The temporary 10 micron filtration system had been installed to filter sump effluents before LLWT entry in order to reduce the frequency of the 25 micron discharge filter changeouts. Due to residual fines and dirt from construction, an accumulation of sludge in the LLWTs was necessi-

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tating changing the discharge filter once or twice per day. Excessive low level liquid accumulation from Pressurizer Relief Tank (PRT) cool-downs (due to safety and relief valve leakage), pump seal water flow rates and various other leakage, coupled with the frequent changeouts of the discharge filter, were taxing the plant's ability to process low level wast Following the overflow, the licensee made sure that PE0s standing the radwaste watch were fully cognizant of this temporary lineup. In addi-tion, a management individual is now assigned to reducing the volume of radwaste and cleaning up the zones that drain to the sumps, with the goal of eliminating the temporary filter ri Low level liquid volumes are reducing and the discharge filter now requires changeout about once every two week The inspector will continue to follow licensee activities in this are c. Volume Control Tank Temperature Reduction In-Service Test ilumber two Reactor Coolant Pump (RCP-2) number one seal leakoff had been increasing steadily and reached 6.7 gallons per minute. RCP-4 seal leakoff was at 5 gpm. Leakoffs from RCP-1 and RCP-3 were between these values. The vendor technical manual states that the vendor will evaluate the conditions and make recommendations at 5 gpm leakoff. The vendor recommendation in this case was to allow a leakoff of 7.5 gpm based on the cooling capacity of the thermal barrier heat exchanger If seal injection flow is lost, this heat exchanger will be able to cool 8 gpm of primary water sufficiently to allow proper seal functionin On May 15, 1986, the licensee implemented an In-Service Test (IST-3-86-010) to study the effects of lowered injection flow temperatures on RCP seal leakoff. The suspected cause of increased leakoff flow is an elec-trophoretic induced corrosion product buildup on the seal surfaces. High velocity flow resulting from a 2200 psi pressure drop through a tight restriction causes an uneven static charge to build on seal surfaces, attracting corrosion particles whose presence pertubates the flow. That increases the seal to runner clearances and results in increased flow through the sea The inservice test lowered the temperature of the volume control tank (VCT) and implemented data collection to dc. ermine the effect on seal leak-off. The temperature reduction was recommended by the vendor after similar experience at other plants; the procedure underwent a safety evaluation and was approved by the Plant Operations Review Committee (PORC) in meeting 3-86-142. The actual results have been favorablei RCP-2 leakoff flow is down from 6.7 gpm to 2.6-3.0 gpm with a correspond-ing reduction in RCP-1 and RCP-3. RCP-4, the pump with the least origi-nal leakoff flow, had the least reduction in flow, from 5 gpm to 4.1 gp I l

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As corrective action in the near term, the licensee intends to replace the seal injection filters with charged, high efficiency 1-micron glass

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filters to reduce the input of. corrosion products. In the longer term, j the seals will be replaced by new seals of a harder material. The.ori-ginal reason for replacement was that the harder seals are better able

, to withstand contact with the seal runner during low flow startup period But, as part of the work that has been done with electrophoresis, it was

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j found that the new seal material has much better electrical conductivity l than does the old material'and distributes the static charge buildup more

! evenly over its surfac An even distribution of charge and therefore j of corrosion particle buildup reduces the hydraulic disturbances and s should improve seal leakoff.

The Millstone 3 Probablistic Safety Study has shown that failure of the i

i RCP seals after a loss of seal injection flow resulting from a station blackout is a contributor to the sixth ranked dominant accident' sequenc , The inspector will continue to follow licensee activity in improving RCP-j seal effectiveness, with emphasis on correlation of measured' leakage to

worst case leakage.

Potter-Brumfield Type MDR Relay Application I

i Seabrook Station filed a 10CFR50.55e report on misapplication of Potter-l Brumfield type MDR relays in the Westinghouse Solid State Protection 1 System (SSPS). Contacts from these relays were used_in DC solenoid valve I

circuits to interrupt current beyond the contacts' rating. In addition,'

i high voltage arcing across the contact resulting-from 1-1.5 amp rated-solenoid inductive kick could have caused burning of the contacts. The

failure of these safety-related relays could prevent safety equipment from performing its intended safety function.

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occurred at Millstone 3. During start-up testing of Engineered Safety Features in test INT-2003, the licensee found that the.MDR contacts were

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l significantly underrated (0.8 amps) for operating Main _ Steam Isolation i Valve (MSIV) pilot valve solenoids-(30 amps). Interposing relays were' '

installed under E&DCR TC-07056 to eliminate the' current rating' mismatch.

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On the Deficiency Form that documented the original' problem (UNSAT 5336),

! a note stated that the deficiency was not generic. The inspector has l found no documentation that a review was conducted to' determine if other L MDR' relays were-used to drive DC solenoids in safety systems. If the.

{ MDR relays had been used to control similar direct acting solenoids, they i very probably would have failed _during ESF testing. The inspector will~

continue to review this subjec ,

"A" Emergency Diesel Ganerator Output Breaker Failure to Shut i

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The "A" Emergency Diesel Generator (EDG) output breaker failed to operate i

during a post-maintenance test of the diesel generator on June 4. .The

unit had been removed from stand-by service for preventive maintenance at 0630 that day. The licensee conducted an immediate. investigation ~

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revealing that an internal micro-switch had not reset, preventing con-tinuity from being restored to the breaker closing circuit. This was evident because the white "breaner ready" lamp on the front of the switchgear was not illuminate The micro-switch in question (52SM/LS) resets when the breaker closing spring stored energy mechanism is fully charged. In this case, the spring was charged but the monitoring switch had not reset. The licensee replaced the switch, GE type CR2940V301, and successfully tested the diesel generator at 1324, June 5. The "A" EDG was declared operable following the completion of survellance testing at 143 The licensee performed failure analysis of the switch in questio The device appeared to operate properly when examined. In addition, identi-cal switches were examined and their contact resistance measured. Both new switches and those in use were tested. Several in-service switches were found to have 50 to 100 ohms contact resistance, which was reduced to between 0.08 and 0.10 ohms after rinsing the switch with a contact cleaning solution. Based on these findings, the licensee is considering the addition of contact resistance measurements as part of the normal routine maintenance of these breakers. Based on their total operating experience with this type of breaker, the licensee considers this to be an isolated event at this tim The inspector confirmed that all 4160 Volt Switchgear Breaker Ready lamps were illuminated, with the exception of breakers racked out for personnel safety during maintenance and those for de energized swing component The inspector verified compliance with applicable Technical Specifica-tions concerning those system There were no unacceptable conditions identifie f. Pressurizer Safety and PORV Leakage The second Power Operated Relief Valve (PORV) Block Valve was shut on June 5 to reduce leakage into the Pressurizer Relief Tank (PRT). Prior to this, the PRT required almost daily feed and bleed cooldowns that consumed about 20,000 gallons of water each and resulted in significant number of low level radwaste discharges. Both PORVs are now blocke One safety continues to lea The PRT requires puinping down about once every 3 days. In-service test 3-86-103 was written and performed to measure leak rate with the PORVs both shut; the result was a measured 0.4 gp ,

The PORV isolations are in accordance with Technical Specification I There is nonetheless a concern that a pressure transient may challenge I an already leaking safety valve and compound leakage problems. Moreover, the leaking safety valve is causing steam to flow into all 3 safety valve I tail pieces, cooling the heated junction thermocouple indicators that serve as positive indication of safety valve opening. As a result, all three safeties have continuous flow indication on Main Board ]

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The licensee is taking positive steps to correct thesa deficiencie Alternatives are under review. A design change to modify the setpoints of the heated junction thermocouples is in progress. Two new PORVs are on order. Rebuild kits will be available if the new PORVs are not available for the mid-cycle outage. The licensee has 3 spare safety valves on han The inspector will continue to follow licensee correc-tive action y Interactions Between Seismic Class I Components The licensee discovered on May 20 that the Number 3 Battery Charger, as-sociated with 125 volt Battery 301A-2, had insufficient seismic shake space between it and 120 volt a.c. Vital Bus Number VIAC- Although both components remained in service, the Battery Charger was considered to be inoperable under Technical Specification Limiting Condition for Operation.(LCO) 3.8.2.1.b. The LC0 concerning the 120 volt a.c. Vital Bus Number VIAC-3 was met as the bus was energized as required. Perman-ent corrective action will establish an adequate shake space between these devices. The licensee temporarily corrected the deficiency by banding the two cabinets together, causing them to act as a single seis-mic structure.- This corrective action had been endorsed by licensee engineering analysis. The problem, inadequate shake space, had existed during construction. Corrective action has been implemented for the other three Battery Chargers by moving the static inverters. The Number 3 inverter had not yet had this design change mad The inspector verified that this condition did not exist on the other three battery systems. There were no unacceptable condition . Plant Maintenance The inspector observed and reviewed preventive and ' corrective maintenance to verify compliance with regulations, use of administrative and maintenance procedures, compliance with codes and standards, proper QA/QC involvement,

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use of bypass jumpers and safety tags, personnel protection, and equipment alignment and retest. The following activities were included:

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"A" Emergency Diesel Generator monthly preventive maintenanc Weld repair to Feed Heater 1 No deficiencies were note . Surveillance Testing The inspector observed parts of tests to assess performance in accordance with approved procedures and Limiting Conditions for Operation, removal and re-storation of equipment, and deficiency review and resolution. The following 'l tests were reviewed- '

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In-service testing of "B" Containment Spray and "B" Safety Injection Pump In-service testing of the turbine driven auxiliary feedwater pum Solid State Protection System Surveillances SP3443E01 and D2 The inspector had no questions on the conducted test . Review of Licensee Event Reports (LERs)

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LERs submitted during this report period were reviewed. The inspector as-sessed LER accuracy, whether further information was required, if there were

generic implications, adequacy of corrective actions, and compliance with the reporting requirements of 10CFR 50.73 and Administrative Control Procedure ACP-QA-10.09. Selected corrective actions were checked for thoroughness and implementatio Those LERs reviewed were

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86-031-00, CBI Signal from Chlorine Detector B Trai , Reactor Trip on Low Steam Generator Water Leve , Violation of Technical Specification 3.3. This violation of NRC requirements was reviewed during the previous inspection perio The problem was identified by the licensee and communicated to the in-spectors. The Plant Incident Report circuit documented the occurrence, and corrective action was quickly implemente , Missed Surveillance on ESF Radiation Monitor Sample Rate Flow.

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86-035-00, Reactor Trip Resulting from Turbine Trip Due to Low Condenser Vacuum There were no inadequacies identifie . Steam Binding of Auxiliary Feed Pumps (TI 2515/69)

The inspector verified that the licensee's written reply to IE Bulletin 85-01 was timely (within 120 days of issuance of the operating license) and in ac-cordance with the Bulleti Change Number 9 to Operating Procedure (0P) 3322 ,

" Auxiliary Feedwater System" was approved by the Plant Operations Review Com- l mittee and implemented on March 19, 1986. This change fully instituted the '

action requested by the Bulletin. It was noted that the procedure was imple-mented 21 days later than the stipulated 90 days following receipt of an operating license. (At the date of procedure issue, the plant was still in power ascension testing and had not yet exceeded 50% power.) 'l

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The inspector checked that current changes were incorporated in the working copies of OP3322, the Auxiliary feed System procedure. He also discussed the procedure with some operators to ascertain their understating of the require-ments and bases. There were no discrepancies note Motor Driven Auxiliary Feed Pump "A" is currently experiencing back leakage that was detected by elevated casing temperatures. The pump is being run periodically in accordance with OP3322 to reduce the temperature and reseat the check valves. The resident inspector will continue to follow this trend as a matter of routine inspectio No unacceptable conditions were identified during this inspectio . IE Bulletin 86-01 Revi s for Westinghouse PWRs Although this Bulletin is addressed to BWRs only, a 10 CFR 21 report from a Westinghouse plant showed that similar problems may exist in PWRs. The in-spector therefore reviewed Millstone 3 Emergency Core Cooling Systems for common recirculation lines for which isolation could disable their respective high, intermediate or low head safety injection pump Each of the Millstone 3 low head Residual Heat Removal (RHR) pumps have motor-operated minimum flow valves that independently recirculate to their respective suction The valves are controlled by independent circuits with inputs from independent flow transmitters. No possibility of a single recir-culation valve failure causing both RHR pumps to fail was foun Intermediate head Safety Injection (SI) pumps have normally open motor-opera-ted minimum flow recirculation valves in lines that merge to a common line which penetrates the Refueling Water Storage Tank (RWST). There is another normally open motor-operated valve in that common line. Operator acticn is required to shut these valves while lining up for containment recirculation; there are no automatic closure signals. The valves are interlocked with the SI pump containment sump suction valves so that the latter cannot be opened without all three minimum flow valves being shut. In this manner, contamin-ated sump water is prevented from entering the atmospheric vented RWS Shut-ting any one of these three recirculation valves will annunciate two alarms on Main Board 2. It was concluded that, while closure of the common recircu-lation valve could result in loss of botn SI pumps, the Millstone 3 configur-ation, normally open fail-as-is valves with no automatic closure signals, provides a reasonable safeguard against such closur Charging pump (high head SI) minimum flow recirculation is isolated by a motor-operated valve for each pump as well as by a common motor-operated valve upon receipt of a safety injection signa Simultaneously, discharge relief valves are unisolated by the same signal. These reliefs are set at 2200 psig and are capable of passing the full 60gpm required to prevent pump damag These recirculation valves have position indication on Main Board 2 and will actuate an annunciator if in an off-normal positio ~

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No concern that a loss of power to these common recirculation line motor-operated valves will disable the SI or Charging Pumps was identified: the valves will fail as is. Further, all of the SI and Charging pump recircula-tion valves have position indication and off-normal annunciators on Main Boards 2 and 3, making their closure readily apparent. The most probable cause of closure of these valves would be failure to fully recover from a maintenance lineup. Since these are ECCS valves, they require dual verifica-tion of position during system restoration from a tagout. It was concluded that there are valid equipment and administrative measures to assure proper lineu In the Millstone 3 Probablistic Risk Assessment, disabling the charging and safety injection pumps due to closure of common minimum flow recirculation valves did not surface as being significant: these are undeveloped events on the fault tree model No unacceptable conditions were identifie . Management Meetings During this inspection, periodic meeting were held with senior plant manage-ment to discuss the scope and findings. No proprietary information was iden-tified as being in the inspection coverage. No written material was provided to the licensee by the inspector.