IR 05000029/1991004
| ML20024G904 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 04/23/1991 |
| From: | Fairtile M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20024G903 | List: |
| References | |
| 50-029-91-04, 50-29-91-4, NUDOCS 9105020069 | |
| Download: ML20024G904 (16) | |
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U.S. NUCLEAR REGULATORY COhihilSSION i
REGION I
Report No:
50 29/91-04 Docket No:
50 29
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Licensee No:
DPR-03 Licensee:
Yankee Atomic Electric Company 580 Main Street
Bolton, hiassachusetts 017401398 Facility Name:
Yankee Nuclear Power Station Inspection at:
Rowe, Massachusetts-Dates:
February 20 to April 15, 1991 Inspectors:
Thomas Koshy, Senior Resident inspector
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Mark Miller, Resident inspector
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Peter Sena, Reactor Engineer.
Approved By:
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i M'. Fairtile, Acting Chief, Reactor Projects Section 3A'
. Date Summary: Insoection Reoort On February 20 - Aoril 15.1991 Report No. 50-29/91-04
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Areas Inspected: Routine inspection on daytime and backshifts in the areas of plant operations, radiological controls,' maintenance and surveillance, security, engineering and technical support, safety assessment and quality verification.
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No viniations or deviations were identified.
i 9105020069 910424
.PDR-ADOCK 05000029 3.
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EXECUTIVE SUMMARY Plant Onerations Main condenser in leakage continued to effect plant operations, resulting in elevated steam generator chloride concentrations and three load reductions for tube plugging.
Human factors deficiencies and poor communications resulted in the inadvertent actuation of the turbine lube oil deluge system during surveillance testing.
A minor body to bonnet leak developed on the loop 3 hot leg isolation valve. A Vapor Container (VC) entry was made, the valve was inspected and cleaned and it was verl0cd that carbon steel subcomponents were not degraded by boric acid corrosion.
Radiological Controls An unresolved open item relating to potential unmonitored releases via the component cooling water system was closed. The inspector concluded that no additional risk is posed by the existing con 0guration.
Maintenance and Surveillance Observed Technical Speci0 cation surveillances were conducted in accordance with procedures and with adequate interaction with the operations staff.
A violation related to inadequate corrective actions for repeated emergency lighting failures was -
closed. The licensee performed an evaluation of potential causes of battery failure and factored their conclusions into the appropriate procedures.
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SImtity Plant security activities returned to normal levels following cessation of hostilities in the Persian Gulf.
Engineeringjmd Technical Sunnort Yankee's modi 0 cation of the pressure control valve for the turbine driven Emergency Boiler Feedwater Pump steam supply was reviewed and found to include an appropriate level of review and approval.
Safety Assessment /Ouality Verification Plant Information Reports and the one licensee event report reviewed during this period adequately characterized operational events and occurrenees.
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TAHLE OF CONTENTS EX EC UTIV E S U M M A R Y.............................
.........il TA B L E O F CO NTENTS........................................ i i i 1.
SUMM ARY OF FACILITY ACTIVITIES........................
2.
PLA NT OP ER ATION S...................................
2.1 Plant Operatione Review..............................
I 2.2 Safety System Review................................
2.3 Review of Temporary Changes, Switching, and Tagging...........
2.4 Fire Protection Sprinkler System Actuation...................
2.5 Loop Isolation Valve Leakage...........................
2.6 Steam Generator Chemistry
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3.
RADIOLOGICAL CONTROLS..............................
3.1 (Closed) Item 90 07 - 0 1.,,............................
4.
M AINTENANCE AND SURVEILLANCE,
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4.1 Maintenance Department Reorganization.....................
4.2 Technical Specification Surveillance Activities.................
4.3 (Closed) Violation 90-16-02, Station Emergency Lighting..........
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5.
SECURITY
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ENGINEERING AND TECHNICAL SUPPORT....................
6.1 Modification of Emergency Boiler Feed Pressure Control Valve
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7.
SAFETY ASSESSMENT AND QUALITY VERIFICATION
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7.1 LER 9101 Inadequate Quality Controis Result in Defective Wire Crimps on Safety Related Systems...
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7.2 Plant Information Report (PIR) 90-10, Control Rod 24 Discovered Unlatched During Core 21 Physics Testing...................
I1 7.3 Plant Information Report (PIR) 90 03, Stuck Control Rods During Drop Ti m e M ea su re m en t..................................
l 7.4 Plant Information Report (PIR) 90-07, Steam Turbine-Emergency Boiler l
Feedwater Pump (ST-EBFP) Overspeed Trip Failure.............
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8.
REVIEW OF PERIODIC REPORTS........................,
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9.
M A N A G EM ENT M EETI N G S......................,........ 13 9,1 Preliminary inspections Findings.........................
9.2 N R C A ctivi ties..................................... 13 ili
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DETAILS
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1.
SUhth1ARY OF FACILITY ACTIVITIES On hiarch 2,1991, power was reduced to approximately 50% for one week for condenser tube sleeving.
I On hiarch 12, 1991, power was reduced to the intermediate range to allow inspection of loop
3 hot leg isolation valve leakage, and inspection and plugging of main condenser tubes Power
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was returned to 100% on hiarch 14, 1991, j
On April 2,1991, a reorganization of the YAEC maintenance department was announced. The new organization involves the creation of 2 assistants to the hiaintenance Director and the division of the maintenance group into a hicchanical hiaintenance Department and an Electrical hiaintenance Department, each with a supervisor.
On April 5,1991, power was reduced for condenser tube plugging. Full power operation was restored on April 6,1991.
2.
PLANT OPERATIONS (IP 71707, 71710)
2.1 Plant Operations Review-The inspector observed plant operations during regular and backshift tours of the following areas:
Control Room Safe Shutdown System Building Primary Auxillary Building _
Fence Line (Protected Area)
Diesel Generator Rooms
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Intake Structure Vital Switchgear Room i
Turbine Building Spent Fuel Pit (SFP) Building Safety injection Building The following items were checked during daily routine facility tours: shift staffing, access-control, adherence to procedures and limiting conditions of operation (LCOs), instrumentation, recorder traces, protective systems, control room annunciators, area radiation and process
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monitors, emergency power source operability, operability of the Safety Parameter Display System (SPDS), control room logs, shift supervisor logs, and operating orders. On a weekly
- basis, selected Engineered Safety Feature (BSF) trains were verified to be operable. The condition of plant equipment, radiological controls, security and safety were assessed. On a r
I biweekly frequency, the inspector reviewed safety related tagouts, chemistry sample results, shift turnovers, portions of the containment isolation valve _ lineup and the posting of notices.to =
workers. Plant housekeeping and fire protection were also evaluated.
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Inspections of the control room were performed on weekends and ba;kshifts as follows: February 20, 28; hiarch 5, 6,13,15,18,19, 20, 24, 29; April 1, 8, 9,10,11,12 and Deep Back Shift inspections were performed on hfarch 2 - 8:00 a.m. to 1:00 p.m.; hiarch 16 - 6:45 a.m. to 11:15 a.m.; hf arch 17 - 8:00 a.m. to 12:30 p.m.; hf arch 24 - 2:15 p.m. to 6:15 p.m.; April 4 -
8:00 a.m. to 12:00 p.m.; April 13 - 7:00 a.m. to 12:00 p.m.; April 14 - 9:00 a.m. to 3:45 p.m.
Operators and shift supervisors were alert, attentive and responded appropriately to annunciators and plant conditions.
Cognizant shift personnel were knowledgeable of plant conditions and ongoing maintenance and surveillance activities. Shift turnovers were conducted professionally with effective control exercised over control room access. Shift documentation adequately characterized operating history and the observed off normal conditions. Equipment problems were resolved in a timely manner.
2.2 Safety System Review The emergency diesel generators (EDG), EDG fuel oil, station batteries, and safety injection systems were reviewed to verify proper alignment and operational status. The review included verification that: (i) accessible major flow path valves were correctly positioned, (ii) power supplies were energized, (iii) lubrication and component ecoling were proper, and (iv)
components were operable based on a visual inspection of equipment for leakage and general conditions. The risk-based inspection guidelines were factored into the safety system review.
High Pressure Safety injection System (HPSI), Low Pressure Safety Injection System (LPSI) and the Accumulator System were included in the routine physical inspections to assess material conditions and operational readiness. Selected accessible valves were verified to be in the correct position and locked when required by plant procedures.
2.3 Review of Temporary Changes, Switching, and Tagging Temporary change requests (TCRs), which were approved in support ofimplementing lifted leads and jumper requests and mechanical bypasses, were reviewed to verify that: controls established by AP-0018, " Temporary Change Control," were met; no conflicts with the Technical
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Specifications were created; the requests were properly approved prior to installation; and a safety evaluation in accordance with 10 CFR 50.59 was prepared if required. Implementation of the requests was reviewed on a sampling basis.
The switching and tagging log was reviewed and tagging activities were inspected to verify plant equipment was controlled in accordance with the requirements of AP 0017, " Switching and Tagging of Plant Equipment."
l Licensee administrative control of off normal system configurations by the use of TCRs and
switching and tagging procedures, as reviewed above, was in compliance with procedural instructions and was consistent with plant safety. No unacceptable conditions were identified.
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2.4 Fire Protection Sprinkler System Actuation The nre protection sprinkler system was inadvertently actuated (Event Number 033765) on hlarch 20,1991 by operations department personnel during the performance of surveillance OP-4242 " Fire Protection Sprinkler System Operability Tests," Revision 9. The deluge system for the turbine tube oil reservoir was manually initiated when valve FS-V-763 was mispositioned.
This deluge valve is a normally shut " knife-switch" valve and is in the closed position when "in-line." Valve FS-V 763 is located within an enclosed housing in the switchgear room fan room.
The valve was mispositioned due to informal communications between the operators performing the surveillance and a lack of understanding of valve operation. Valve FS V-763 was initially in the "in line" position and thus closed; however the operator assumed this was the open position. The error occurred when the operator manipulating the valves asked the question "Do you want this closed?" referring to FS V-763. The operator with the procedure in hand answered simply "yes," referring to the cabinet door. "'he valve operator then repositioned FS-V-763 to the open position and thus initiated the deluge system. Approximately 14 ga!!ons of a water / foam solution sprayed onto the turbine lube oil reservoir. No damage was inflicted on the oil system. The resulting Gush and refill of the Orc system resulted in the inoperability of the turbine lube oil reservoir foam system per Technical Specincation 3.7.10.6. The inspector veri 0ed that the licensee complied with the associated action statement by establishing a continuous fire watch with backup fire suppression equipment. The licensee has generated a Plant Information Report (PIR) to address the root causes, generic implications and corrective actions.
2.5 Loop Isolation Valve Leakage On March 12,1991, inspections performed in the VC revealed significant deposits of boron on MC-MOV-309, the loop 3 hot leg loop isolation valve. The deposits were viewed from the top of the Loop 3 cubicle due to high levels of radiation which exist in the cubicle when the reactor is at power. The assessment resulting from this inspection was that a packing leak existed, resulting in large deposits of boron accumulating on the valve and its motor operator.
The inspector attended PORC meeting 91-15, conducted later the same day. At the meeting, PORC considered the scope of the leak and the possible adverse effects of boric acid on carbon
.; teel subcomponents of MC-MOV-309. As a result of these discussions, a decision was made i
to bring the plant to Mode 2, zero power conditions, to allow personnel entry into the loop 3 cubicle for inspections, valve cleanup, and packing adjustment.
l On March 13, 1991, the inspector accompanied members of the plant operations, maintenance, l
and radiation protection (RP) organizations on a VC entry to inspect the subject valve. The l
inspection was well coordinated, with a pre-entry brienng on individual activities conducted by RP personnel. Once in the VC, RP personnel were effective in controlling activities and performed briefings explaining survey results,
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The inspection of MC MOV-309 revealed that a packing leak did not exist. The area of highest boron conecatration was in the region of the stem leakoff line in the vicinity of the body-to-bonnet interface. When the valve was cleaned, a minor body-bonnet leak was identified.
YAEC's actions included verifying that carbon steel subcomponents were not degraded, thoroughly cleaning the affected areas, and removing insulation from the around the valve.
Insulation removal was intended to both prevent pockets of boric acid from building up between the valve and insulation and to allow cooling and subsequent contraction to improve the body-to-bonnet seal.
YAEC considered the possibility of lapping the body-to bonnet seating surfaces, but concluded that, given the minor nature of the leak, this action could be deferred. The llevnsee concluded that, if followup inspection showed a worsening trend, a maintenance outage could be arranged.
YAEC management did not wish to attempt a lapping procedure until an auton ated lapping tool arrived on site. The need for an automated lapping tool, and the potential for man-rem reduction associated with its use, was discussed in NRC Inspection Report 50 29/90-16.
Surveillance of the valve continued through the end of the inspection period as a function of planned VC entries. Personnel were viewing the valve, through binoculars, from the top of the cubicle.
On April 6,1991, in conjunction with a power reduction for condenser tube plugging, an entry was made into the loop 3 cubicle for an inspection of MC MOV 309. The inspection was conducted by the Maintenance Support Supervisor, who noted that insulation removal was effective in improving the body to-bonnet seal and thus reducing the leak rate. Minor boron buildup was noted around a body-to-bonnet joint stud onto which the leakage impinged. The management attention to this event was prompt and effective.
2.6 Steam Generator Chemistry l
YAEC has experienced elevated levels of chloride concentration in steam generator secondary l
water inventories since November,1990. The cause has been failures in condenser tubes and tubesheets. The increased chloride levels factored into three power reductions during the inspection period. On March 1,1991, power was reduced to approximately 50% to allow condenser tube plugging and sleeving. This evolution took approximately one week. While l
power was reduced, chloride levels rose from approximately 36 ppb (average) to a peak value of 341 ppb, then dropped to approximately 74 ppb by March 5,1991.
The inspector attended a PORC meeting on this subject on March 5,1991. In response to the
l elevated chloride level, the Plant Superintendent directed that power be further reduced to approximately 25 30% and that steam generator blowdown rates be maximized. Power was reduced that day and, by March 9,1991, chloride levels averaged 15 ppb. Following condenser tube sleeving, chloride levels steadily increased; however, dropping in response to additional tube plugging on March 13,1991 (coinciding with the power reduction for MC-MOV 309 inspection)
and sawdust injection on March 21,1991..
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Oa April 5,1991, power was again reduced to allow for tube plugging. During this power reduction, blowdown was again maximized and a number of tubes in both waterboxes were plugged. This resulted in a decrease in steam generator chloride levels to approximately 12 ppb, averaged over the 4 steam generators.
At the end of the inspection period, chloride concentrations were trending upward and Yankee was evaluating the impact of frequent load reductions to accommodate tube pbigging. YAEC management displayed a conservative approach toward chloride levels throughout the inspection period, initiating condenser maintenance whenever concentrations approached 20 ppb. However, recent experience has indicated that cycling the condenser for maintenance tends to create new leaks. Yankee is currently pursuing a strategy ofincreasing steam generator blowdown rates and injecting sawdust into condenser tubes to slow the upward trend, and plugging condenser tubes when chloride levels approach the YAEC administrative unit. Yankee's administrative limit for steam generator chlorides is 30 ppb, based upon known corrosion mechanism rates of the 304 stainless steel steam generator tubes. The inspectors concluded that Yankee ~s actions in response to this issue were thorough and conservative, given the constraints imposed by the nature of the problem. The licensee has long term corrective actions under consideration.
3.
RADIOLOGICAL CONTROLS (IP 71707)
Radiological controls were reviewed on a routine basis relative to industry radiological standards, administrative and radiological control procedures, and regulatory requirements. Selected work evolutions were observed to determine the adequacy of program implementation commensurate with the ladiological hazards and importance to safety.
3.1 (Closed) Item 90-07 01 Component Cooling Water System Review Unresolved Item 90-07-01 resulted from inspector review of the component cooling water (CCW)
l system. The inspector noted that the CCW system operated at a higher pressure than the service water (SW) system which cools it. This pressure differential, combined with a CCW heat
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exchanger tube leak, could result in CCW being released to the environment via an unmonitored pathway. CCW is chromated for corrosion control and, as it serves to cool various systems containing primary coolant, could potentially become radiologically contaminated. It was noted that the CCW system is provided with one radiation monitor and that monitor operability was not covered in Technical Specifications (TS).
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Yankee addressed these concerns in an internal memorandum, a copy of which was provided to the inspector.
In the memorandum, the topics of chromate and radioactive release were discussed. Yankee stated that the total amount of potassium dichromate in the CCW system, assuming concentrations at the upper limit of the current band (500 ppm), was 31.7 pounds, which is less than the 50 pound / day limit requiring reporting under 310 CMR 40.900. Yankee further stated that the YNPS NPDES permit requires reporting any non routine discharge of any toxic pollutant which is not limited in the permit that will exceed 500 mg/ liter. Based on the flowrate at the circulating water discharge to the ultimate heat sink, Yankee stated that a CCW
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leakrate of 130 GPM must occur to exceed the 500 mg/ liter limit. A leak of this size would result in a CCW surge tank low level alarm in approximately 11 minutes, allowing operators to respond.
Yankee pointed out that the issue of intersystem leak detection was addressed by NUREG 0825, the YNPS Safety Evaluation Program. In NUREG 0825, the NRC staff concluded that neither the addition of radiation monitoring equipment in the SW system nor the addition of the CCW radiation monitor to TS would signincantly effect the risk associated with YNPS. The inspector noted that the analyses presented in NUREG 0825 took no credit for the CCW radiation monitor, The inspector reviewed YAEC procedure OP 4801, Revision 17, " Functional Test and Alarm Settings of the Process Radiation Monitoring System and found that the CCW radiation monitor was functionally checked monthly. The inspector reviewed a sample of functional test records for 6 consecutive months in 1990. It was apparent from these records that the CCW radiation monitor had performed well and that the tests were performed at the prescribed frequency. The inspector also noted that AP-9001, Revision 9, " Primary Chemistry Test Frequencies and Specifications," requires that CCW is sampled weekly for pH, chromate level and gross activity.
The inspector concluded that Yankee's response to the open item was technically sound and that adequate provisions are in place to protect the public and the environment from significant releases of chwmates and radioactivity via a CCW system failure. This item is closed.
4.
h1AINTENANCE AND SURVEILLANCE (IP 71710,61726,62703,92700,93702)
The inspector observed and reviewed maintenance and surveillance activities relative to industry standards, administrative controls, and regulatory requirements. Selected work evolutions and surveillance tests were observed to verify safety requirements and compliance. Speci0c areas examined were licensee use of station procedures, codes and standards, QA/QC involvement, management oversight, safety tag use, jumper use, equipment alignment and post-maintenance testing (PMT). In addition, the inspector evaluated radiological controls for worker protection, Gre protection, limiting conditions for operation (LCOs), deficiency review, resolution and reporting per Technical Specifications (TS).
4.1 Mnintenance Department Reorganization On April 2,1991, a reorganization of the YAEC maintenance department was announced. The current organization consisted of a Maintenance Director (reporting to the Plant Superintendent)
directing the activities of the Maintenance Support Department (MSD), the Instrumentation and Controls Department (I&C), and the Maintenance Department (covering mechanical and electrical work). The new organization involves the creation of 2 assistants to the Maintenance Director and the division of the maintenance group into a Mechanical Maintenance Department and an Electrical Maintenance Department, each with a supervisor.
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The current hiaintenance Supervisor has been named as one of the assistants to the hiaintenance Director. His area of responsibility is to be reactor vessel disassembly for the 1992 outage inspection. The current hiaintenance Support Supervisor has been named as another assistant to the hiaintenance Director and his area of responsibility will include coordination of the reactor vessel examinations. The licensee has begun a search to fill the positions of the hiaintenance Support Supervisor, hicchanical hiaintenance Supervisor and Electrical hiaintenance Supervisor.
Once these positions are filled, the reorganization will be fully staffed.
4.2 Technlent Specification Surveillance Activities
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The inspector observed performance of OP 4665A "Comsip model K ll! Post Accident H2 hionitor (HWGA 2) Calibration," Revision 15. This surveillance is conducted on a quarterly basis and is required per Technical Specincation 3.6.3.1/4.6.3.1 to demonstrate operability of the hydrogen monitor. The technician performing the surveillance verined the prerequisites after obtaining the Shift Supervisor's clearance to commence work. However, prior to ecmmencing the procedure, the NRC inspector informed the technician that he was utilizing an outdated procedure. The correct procedure for the calibration was Revision 15, dated January 1991, while the technician was performing the surveillance using Revision 14. Work was immediately j
secured and the matter was further investigated as appropriate levels of Yankee management were notified, Upon further review, it was discovered that the changes incorporated into Revision 15 were non-technicalin nature and did not effect the scope of the work " Document Change Notice" number il16, dated January 22, 1991 was properly issued for updating procedure OP-4665. Per procedure APF-0223.1, " Document Control," Revision 9, the DCN was signed by Instrumentation and Control (l&C) department personnel documenting that the procedure was
" entered in the applicable department manuals and/or files and all obsolete copies have been discarded or identified as obsolete. I&C department manual, master copy number 29, was found to be updated however the working files still contained the incorrect revision. Yankee initiated a nonconformance report (NCR) and performed a 100% review of all 1&C and hiaintenance Department surveillance procedures and general plant procedures for revision status.
Additionally, the inspector performed an independent review of selected 1&C surveillance procedures. The inspector also verined that the previous completion of OP-4665 was performed to the correct revision. No additional discrepancies were identified by the inspector or the licensee. The inspector concluded that this oversight was an isolated event and that no programmatic deficiencies existed in Yankee's document control system.
Procedure OP-4665A was recommenced with the correct revision. Control room personnel were properly briefed and cognizant of expected alarms. The technician was properly qualified and found to be knowledgeable of the equipment, job and procedure. The inspector concluded that the surveillance was properly performed with appropriate precautions, prerequisites and verifications.
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The inspector winessed selected portions of surveillance procedure OP-4656 " Functional Test of tne NRV Main Steam Line Pressure Utnnel/ Switches," Revision 11.
This monthly surveillance provides a detailed method of testing the three main steam line non-return valve (NRV) pressure channels and switches. A two-out-of three low main steam line pressure signal on any of the four main steam lines will close the NRVs and will scram the reactor. Yankee considers this a "high risk" evolution, as the procedure places the three pressure channels in a one out of two trip logic. The operations personnel were properly briefed by the technicians performing the surveillance and were aware of expected alarms, in addition, the operators were found to be knowledgeable in the required Technical Speci0 cation action statements if a train or channel was found to be inoperable. The technicians performed the surveillance according to procedure and properly signed off each step as it was accomplished, Effective communication was present between the I&C technicians and operations personnel regarding the status of the surveillance. The inspector also noted that proper independent verincation of required sections of the procedure was accomplished as each channel was returned to service.
The inspector concluded that the surveillance was performed satisfactorily in accordance with appropriate requirements.
4.3 (Closed) Violation 90-16 02, Station Emergency Lighting This item pertains to the inadequate corrective action in response to four consecutive failures of the emergency lights during surveillance test. The corrective action did not address potential common mode failures in spite of repeated failures, in response to the above concern, Yankee conducted a comprehensive evaluation of the causes of the emergency lighting failure. They identified two potential causes: (1) batteries for the lighting units had reached their end of life (three to four years), and (2) the battery 00at voltage was out of speciDed range, either high or low, wluch would shorten the battery life. Both of these causes were addressed by Yankee through appropriate procedure revision to monitor Goat-voltage and to confirm that a battery in service is within its service life. This event was reviewed by the Maintenance and Maintenance Support supervisors with emphasis placed on the importance of ensuring that defective equipment is identified and the corrective actions preclude repetition of the deficiency. Yankee's corrective actions in this matter were deemed adequate.
This item is closed.
5.
SECURITY (IP 71707, 92700)
Selected aspects of plant physical security were reviewed during regular and backshift hours to verify that controls were in accordance with the security plan and approved procedures. This review included the following security measures: guard staffing, vital and protected area barrier integrity, maintenance of isolation zones, and implementation of access controls including authorization, badging, escorting, and searches. No inadequacies were identified.
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The inspectors reviewed loggable events for the period of February to March,1991 and concluded that events were classified appropriately, and that corrective actions and compensatory measures (where applicable) were adequate, i
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Following the cessation of hostilities in the Persian Gulf, security measures were returned to pre-Desert Storm levels.
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6.
ENGINEERING AND TECIINICAL SLJPPORT (IP 37828,92701)
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6.1 Modification of Emergency Holler Feed Pressure Control Valve
Following the rebuilding of the steam turbine emergency boiler feed pump (EBFP) turbine
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governor (see PIR 9407, Section 7.4), the EBFP steam supply pressure control valve, AS PCV-451, demonstrated excessive hunting during surveillance testing. The control valve was declared
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inoperable and subsequent surveillance testing was performed using a manual bypass valve to l
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control steam pressure to the turbine. YAEC determined that the hunting may have been attributable to over sensitive responsiveness resulting from the governor overhaul.
AS PVC-451 was comprised of a controller, a valve positioner, and an air _ operated valve l
(AOV). The controller sensed steam pressure upstream of the EBFP governor and provided a
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linear 3 to 15 psig signal for a given setpoint and proportional band. The "setpoint" of the
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controller is the value of the process variable (in this case steam pressure) at which controller
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output is at midrange. The " proportional band" corresponds to the range of the process variable represented by the 3 to 15 psig output signal of the controller. The proportional band was
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expressed as a percentage of the controller design process range. In the case of AS PCV-451, the process range was 0-200 psig, the setpoint was 125 psig and the proportional band was 20%,
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resulting in a 3-15 psig output signal which corresponded to 125 psig +/- 20 psig. The controller was reverse acting, meaning that a 3 psig controller output (corresponding to a 0%
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l valve position) resulted from a steam pressure of 145 psig, and a 15 psig output signal
l-(corresponding to a 100% valve position) resulted from a 105 psig steam pressure. - Controller i
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output was directed to a valve positioner, which supplied 16 28 psig (corresponding to the 315 i
psig input signal) to the AOV actuator.
-Plant personnel, working in conjunction with Yankee Nuclear Services Division ('YNSD),
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turbine governor respon'siveness, (b) the controller was actually capable of delivering 0-30 psig, I
at a lower air delivery rate than the valve positioner, (c) the controller response remained linear over the 0 30 psig range, with slope and intercept determined by setpoint and proportional band,
and (d) the valve positioner could, by design, be bypassed, allowing the controller output to act (
directly on the AOV actuator.
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YAEC found that, by changing the controller setpoint to 160 psig and the proportional band to 25%, and by bypassing the valve positioner, the controller would deliver 16-28 psig linearly to the AOV actuator (corresponding to 0-100% valve position) with steam pressure ranging from 105115 psig. Because of the slower air delivery rate of the controller, there was a slower overall response of the AOV, allowing the valve / governor to reach stable conditions. Tests
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verined that stability was achieved for EBFP discharge pressures up to 1000 psig.
The basis for the modification was transmitted from YNSD to YNPS by memorandum. The memorandum included a determinvlon that the modification did not create an unreviewed safety question, pursuant to 10 CFR 50.59. The modification was implemented through a maintenance request (MR) referencing the memorandum. The memorandum and changes to the applicable surveillance test procedure were reviewed by the Plant Operations Review Committee (PORC)
prior to initiation of the modification.
While the engineering involved in this modification showed a very resourceful approach to resolving the hunting problem, the inspector questioned the practice of effecting modi 6 cations by memorandum. Yankee stated that this modification did not constitute a design modification, as there was no change in the form, fit or function of AS-PCV-451 and no drawing changes were required. The inspector reviewed the instrument loop folder for the control valve and found that, while a copy of the memorandum was included in the folder and the appropriate material history forms were ulxiated to reflect the change, form DPF 6001.4, " Instrument Equipment Data," had not been updated to reflect the new setpoint and proportional band as required by YAEC
procedure DP-6001, Revision 11, "I&C Department Equipment History Card / Loop Folder System." The inspector pointed this fact out to 1&C personnel and the appropriate change was made the same day, e
7.
SAFETY ASSESSMENT AND QUALITY VERIFICATION (IP 40500)
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l The inspector reviewed selected portions of Yankee's self assessment program to verify
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implementation and determine if those programs contribute to the prevention of problems through
monitoring and evaluating plant performance, providing assessments and findings, and communicating and following up on corrective action recommendations.
l-l 7.1 LER 91-01 Inadequate Quality Controls Result in Defective Wire Crimps on Safety Related Systents
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LER 91-01 documented fmdings and corrective actions resulting from the discovery of numerous nonconforming wire connectors on three emergency diesel generators (EDGs) and in other plant systems. The nonconformances were the result of inadequate quality-controls applied to an unqualified (under the YAEC QA Plan) electrical contractor during the 1990 refueling outage.
The details of this event are discussed in NRC inspection report 50-29/90-03. The LER was reviewed and found to be generally consistent with the fmdings of IR 90-03, l
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I1 Corrective actions described in the LER, including dates for completion, were identical to those committed to in a YAEC letter (UTR 91-042) to the NRC dated h1 arch 20,1991. The corrective actions cited were deemed to be sufncient to verify the quality of existing contractor installed electrical connections.
The adequacy of quality control-related corrective actions will be reviewed when such actions are completed.
The LER was submitted on hiarch 28, 1991, 30 days after the discovery of numerous nonconformances of EDG 1, under 10 CFR 50.73(a)(2)(v). The inspectors were informed that the time between discovery and notincation was due to a weakness in YAEC's procedure covering reportability. The amplifying information contained in 10 CPR 50.73(a)(2)(vi) was not cal'ed out in Yankee's procedures, resulting in an initial failure to recognize the discovery of fabiication inadequacies as an " event" covered in 10 CFR 50.73 (a)(2)(v) YAEC stated that their procedure would be revised to include this additional information.
7.2 Plant Informntion Report (PIR) 9010, Control Rod 24 Discovered Unlatched During Core 21 Physics Testing On October 30,1991, during Cycle 21 zero power physics testing, YAEC suspected that control rod 24 was not latched to its drive shaft. The plant entered hiode 5 (Cold Shutdown), the Control Rod Drive hiechanism (CRDht) for control rod 24 was removed, and the rod was latched from above the reactor head. The details of the event were discussed in NRC inspection Report 50-29/90-20.
In PIR 90-10. YAEC concluded that, because of the length of time which passed between rod latching and discovery of the unlatched condition (from July 21 to October 30,1990), the cause could not be determined with certainty. However, the PIR discusses three possible scenarios that could have led to the unlatching of control rod 24. The corrective actions delineated in the PIR include additional procedural guidance and the requirement of additional visual inspections to ensure that control rods are adequately latched. The inspector noted that the corrective actions were comprehensive in scope and that they were sufficient to detect an unlatched control rod under any of the possible scenarios identified in the PIR.
l 7.3 Plant Information Report (PIR) 90-03, Stuck Control Rods During Drop Time hicasurement
PIR 90-03 addresses the failure of two control rods to fully insert due to excessive bowing. This event occurred at the end of Cycle 20 with the plant in hiode 2. OP-7000.39, " Simultaneous and Individual Rod Group Drop Time hicasurement" was being performed to determine if control rod drop times would increase when all rods were dropped simultaneously rather than by group. When the scram breakers were opened, rods 17 and 18 failed to fully enter the core, sticking at 39 and 36 inches, respectively.
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12 PIR 90-03 documents the sequence of events comprising this event, the root cause, an analysis of safety significance, and corrective actions.
YAEC committed to: (a) revise OP-2103,
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" Reactor Startup and Shutdown," to include instructions to perform OP-4703, " Control Rod Drop Time Measurement," whenever the reactor is in Mode 3 and drop time measurements have
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not been performed within 60 days, (b) conduct an independent peer review of the YAEC control
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rod bowing model by Core 21 mid cycle, and (c) evaluate additional data concerning neutron i
induced zirconium growth.
The inspector reviewed YAEC's progress to date with respect to the proposed corrective actions.
OP 2101, " Plant Start-up From Mode 2 or Mode 3" was revised in February,1991, to include i
instructions to perform OP-4703 for the conditions describea above. This addition was in lieu
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of the revision of OP 2103 committed to in the PIR. However, the inspector noted that the
l addition was made at a point in OP-2101 just prior to entry into OP-2103. The inspector found l
that making the revision in this way satisfied the intent of the corrective action. The inspector i
rioted that this additional information did not contain a reference to either PIR 90 03 or l
Inspection Report 9012 (which documented the event and commitments) as required for l
commitment tracking by AP-001, Revision 19, " Plant Procedures." The inspector notified
YAEC management of this fact and was informed that a reference would be added. At the end of the inspection period, the balance of the corrective actions had not been completed.
l 7.4 Plant Information Report (PIR) 90-07, Steam Turbine Emergency Boller Feedwater
Pump (ST-EHFP) Overspeed Trip Failure PIR 90-07 discusses an event which occurred with the plant in Mode 5 (cold shutdown) during j
capacity testing of the ST-EBFP. During preparations for the test, the Shift Supervisor adjusted l
the turbir.e governor valve due to its failure to trip properly in previous tests. The adjustment was not performed in accordance with plant procedures. When the ST-EBFP was brought to t
speed and hand tripped by the operator performing the test, the turbine slowed but did not trip.
i The operator adjusted the governor valve stroke and the turbine stopped. The turbine was then restarted and successfully hand tripped.
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The test conthued to the point of the overspeed trip test. The turbine failed to trip on overspeed i
j and had to be manually tripped. The step was repeated, at which tin e the turbine accelerated i
to "well above rated speed" and sparks were seen coming from the governor control area. The operator then secured steam to the turbine and increased pump discharge pressure to quickly slow-j the turbine to a stop. The turbine was subsequently overhauled.
i Yankee identified the apparent cause of the occurrence as a failure of the turbine overspeed trip _
mechanism to close the governor valve. Corrective actions included repair of the governor valve
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and Revisions to OP-4211, " Emergency Feedwater Operability Test," to include guidance on the sensitivity of the trip mechanisms and to require that the maintenance department make
adjustments to the governor or governor valve, when needec. The inspector reviewed OP-4211
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Revision 3, and found that the corrective action related changes had been incorporated, i
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8.
REVIEW OF PERIODIC REPORTS (IP 90712,90713)
Upon receipt, the inspector reviewed periodic reports submitted pursuant to Technical
Specifications and other internal licensee reports. This review verified, as applicable: (1) that the reported information was valid and included the NRC-required data; (2) that test results and supporting information were consistent with design predictions and performance spccincation; and (3) that planned corrective actions were adequate for resolution of the problem. The inspector also ascertained whether any reported information should be classined as an abnormal occurrence. The following reports were reviewed:
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Monthly Statistical Report for plant operations for December, January, and February
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i 1991 Certified Financial Report for 1990.
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9.
MANAGEMENT MEETINGS (30702)
9.1 Preliminary Inspections Findings At periodic intervals during this inspection, meetings were held with senior plant management
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to discuss the Dndings. A summary of Dndings for the report period was also discussed at the conclusion of the inspection and prior to report issuance.
9.2 NRC Activities
On March 21,1991, Mr. Thomas T. Martin, NRC Regional Administrator for Region I, met with YAEC management and conducted a plant tour.
A special inspection (50-29/91-04) on electrical splice problems was conducted by the resident inspectors. On March 25, 1991, John F. Rogge, NRC Region I Project Section Chief with responsibility for Yankee Nuclear Power Station, attended the exit meeting for this inspection.
The findings from this inspection will be discussed in an enforcement conference scheduled for April 29,1991.
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