IR 05000029/1986019

Insp Rept 50-029/86-19 on 861007-870126.No Violations Noted. Two Inadequacies Involving Failure to Fully Review Emergency Plan & Insufficient Implementation Procedures for Offsite Dose Calculation Manual Noted

ML20204H092
Person / Time
Site:	Yankee Rowe
Issue date:	03/19/1987
From:	Elsasser T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20204H041	List: IR 05000029/1986019 ML20204H037
References
REF-GTECI-A-46, REF-GTECI-SC, TASK-03-06, TASK-3-6, TASK-A-46, TASK-OR, TASK-RR 50-029-86-19, 50-29-86-19, IEB-79-01, IEB-79-1, IEB-79-24, IEB-86-003, IEB-86-3, NUDOCS 8703260546
Download: ML20204H092 (21)

Text

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /86-19 Docket N Licensee N DPR-3 Licensee: Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701 Facility Name: Yankee Nuclear Power Station Inspection at: Rowe, Massachusetts Inspection Conducted: October 7, 1986 - January 26, 1987 Inspector: H. Eic holz, Senio esident Inspector ApprovedBy:(OU IM 4%

T. Elsasser, Chief, Reactor Projects Section 3C

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~Date Inspection Summary: Inspection on October 7,1986 - January 26, 1987 (Report

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No. 50-29/86-19)

Areas Inspected: Routine onsite regular and backshift inspection by the resident inspector (136 hours0.00157 days <br />0.0378 hours <br />2.248677e-4 weeks <br />5.1748e-5 months <br />). Areas inspected included review of licensee action on previous findings, operational safety verification reviews, review of radio-logical controls, reviews of events requiring telephone notification to the NRC, review of plant events, maintenance observations, surveillance obser-vations, Plant Operations Review Committee activities, Cold Weather Prepar-attons, followup of Plant Information Reports (PIRs), licensee response to IE Dulletins, review of Licensee Event Reports, inspection of GE HGA Relays, and a review of changes to the licensee's organizational structur Results: No violations were identified by the inspector; however, two inadequacies involving failure to fully review the Emergency Plan and insufficient implementa-tion procedures for the Offsite Dose Calculation Manual were classified as licensee identified violations (Section 14). The licensee's administrative controls for temporary changes (Section 3 and 4), modifications being implemented to the access control point for the plant's radiation control area (Section 5) and licensed operator attention to detail regarding off-normal plant instrumentation indications and resulting timely investigation and corrective measures (Section 7) were viewed as areas exhibiting positive trends or licensee strengths. Areas needing increased licensee attention included security practices (Section 4) and documentation prac-tices associated with maintenance activities (Section 8).

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DETAILS 1

. Persons Contacted t

! Yankee Nuclear Power Station B. Drawbridge, Assistant Plant Superintendent T. Henderson, Technical Director i N. St. Laurent, Plant Superintendent l The inspector also interviewed other licensee employees during the inspection,

including members of the Operations, Radiation Protection, Chemistry, Instru-

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ment and Control, Maintenance, Reactor Engineering, Security, Training, Tech-l- nical Services, and General Office Staffs.

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l Summary of Facility' Activities At the completion of the last resident inspection period on October 6,1986,

the plant was at 75% of rated power returning to full power. A plant trip t had occurred on October 5, 1986 which resulted from low steam generator level.

! The plant was at 100% of rated power on October 7, 1986. On October 31, 1986, I a manual shutdown to hot standby was initiated by the licensee in response !

s to a failed pressurizer wide range level channel and in conformance with the

, requirements of Technical Specification (TS) 3.3.1. A reactor startup com-

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menced on November 1, 1986 following repairs to the instrument channel. The i plant was at 100% of rated power un Novernber 4,1986, and was maintained at j this power level until December 14, 1986. On this date a planned load reduc-l tion to 50% of rated power occurred to allow turbine throttle valve and main

! steam non-return valve testing. In addition, condenser tube cleaning and leak checks were conducted. The plant was at 100% of rated power on December 15,

1986, and was maintained assentially at this power level through the end of

the inspection period on January 26, 1987.

I Licensee Action on Previous Inspection Findings l (Closed) Inspector Follow Item 50-29/78-SP-08: Environmental qualification i (EQ) of stem-mounted limit switches inside reactor containment. As part of i the EQ inspection, 50-29/86-18, the inspection team reviewed the basis for i the 10 CFR 50.49 Master List and checked listed items for validity. The team

) identified no limit switches in the plant that were within the scope of 10 CFR 50.49. This item is closed.

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(Closed) Inspector Follow Item 50-29/79-BU-01: IE Bulletin 79-01, EQ of elec-I . trical equipment. The licensee's EQ Program was inspected b.v the NRC during i inspection 50-29/86-18. This inspection determined that the licensee's pro-

gram to meet the requirements of 10 CFR 50.49 was satisfacto y. This item l 1s closed.

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(Closed) Inspector Follow Item 50-29/85-05-06: Environmental qualification of inaccessible valves in the post accident sampling system. The valves in-cluded in this category are: HV-50V-1&2, VD-MOV-505 through 509, SA-MOV-51 Valves HV-SOV-1&2 are V'alcor solenoid valves Model V526-5820-9, located inside the vapor containment above elevation 1105'0". The environmental qualifica-tion of these valves was documented in EQ file YAEC-QDR-5435-104-0474. The required environmental conditions for these valves were stated on EQ work sheet SOL- Based on the Yankee Rowe Specification for Valve Operators, sheets 2 & 3, VD-M0V-505 through 508 are limitorque type SMA-00 valves, V0-MOV-509 and SA-MOV-511 are limitorque type SMA-000 valves. The qualification data for these valves are contained in the licensee's work package No. YRP-304/8 The inspector reviewed the appropriate documents in these EQ files and had no additional concern. This item is close (Closed) Inspector Follow Item 50-29/85-21-01: Statistical evaluation, upon completion of the analyses of water samples by the licensee and Brookhaven National Laboratory, was to be made. The analyses were completed and an evaluation was performed. During Inspection Number 50-29/85-21, the steam generator water and spent fuel pool water were sampled for analysis. Duplicate samples were sent to the Brookhaven National Laboratory (BNL) for independent verification of analysi Yankee Rowe Split Samples BNL YR Iron (ppb) 84 --

Copper (ppb) 246 230 Nickel (ppb) <70 --

Chloride (ppb) 275 160 Ammonia (ppb) 394 775 Hydrazine (ppb) <4 870 Boron (ppm) 1040 1067 The chloride, ammonia and hydrazine differences resulted apparently because the sample sent to BNL was not stabilized by acid / alkaline additions, as have been found necessary during similar NRC inspections at other facilitie Future comparisons of analyses will include the use of NRC provided spiked samples to minimize these kinds of analysis errors. This item is close (Closed) Unresolved Item 50-29/84-20-01: Revise procedure AP-0018 in accord-ance with 10 CFR 50.59 and Technical Specification (TS) requirement 6.5. for Plant Operations Review Committee (PORC) review of safety evaluation This item concerned the lack of a PORC review requirement in procedure AP-0018 for safety evaluations that resulted from lifted lead and jumper activitie Procedure AP-0018, Revision 12, Temporary Change Control, provides the neces-

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sary controls to insure that lifted lead and jumper activity that changes the facility as described in the FSAR will result in the preparation of a safety evaluation and receive a PORC review prior to installation. Also, the licen-see's procedure now provides extensive management controls for all forms of temporary alterations that result in plant equipment being different than design and licensing documents. The licensee actions in response to this item were determined by the inspector to be fully responsive to NRC concerns. This item is close . Operational Safety Verification Reviews

, Daily Inspection During routine facility tours, the inspector checked the following items:

shift manning, access control, adherence to procedures and limiting con-ditions for operations (LCOs), instrumentation, recorder traces, protec-tive systems, control rod positions, containment temperature and pressure, control room annunciators, radiation monitors, radiation monitoring, emergency power source operability, control room and shift supervisor log, tagout log, and operating orders. No inadequacies were identified except as noted belo Throughout the inspection period, the inspector noted that the power range neutron monitoring system's channel No. 6 was exhibiting periodic spiking of the channel output. This spiking would result in both a high flux alarm and channel trip conditio The reactor protection system is actuated when any 2 of the 6 channels of power range and intermediate power range monitors indicate that a high flux condition exists. The licensee issued Maintenance Request 86-1352 to investigate and correct the condition. The initial troubleshooting efforts of the I&C department indicates that the problem may be in system cabling or the detector itsel The lic-ensee is waiting for a plant shutdown of sufficient duration to facilitate further corrective actions. This item will be reviewed by the inspector during routine inspection activitie On November 13, 1986, the operations department issued Special Order 86-115, which specified that a water treatment demineralizer waste overboard resin indicator and collector was installed in the de-mineralizer waste outlet line. Instructions were provided to the plant operators for action to be taken, should resins appear in the sightglass during a rinse cycle. The inspector reviewed OP-2425, Rev. 9, Operation of the Water Treatment Plant, and noted that, as of the end of the inspection period, no licensee action was taken to revise the procedure and address the licensee's concerns for pumping resins overboard. This item reflects continuing concern that the inspector has expressed about the licensee's use of Special Orders in lieu of approved procedures. Further licensee management attention is warranted to address this concer . _ _ . . . _ _ _ - _ _ _ - .

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On December 31, 1986, the high voltage power supply switch for -

source range neutron monitoring channel No. 2 was placed in the off positio The high voltage for the source range's BF3 detector is controlled by. intermediate range neutron monitoring channel No. 4; i however, an observed drifting low of the channel 4 output resulted in licensee concern. This concern involved the potential for the downward drifting indication to automatically reconnect the high voltage to the source range detector during power operation, which would result in damage to the affected source range detector. Al-though intermediate range channel No. 4 was considered inoperable in accordance with TS 3.3.1, Table 3.3-1, continued power operation was permitte In response to this condition, the licensee initiated Temporary Change Request (TCR) No. 8C-72 on December 31, 1986. This TCR pro-vided the written bases and necessary approvals to insure that the change did not constitute an unreviewed safety question as described in 10 CFR 50.59. The TCR was reviewed by the Plant Operations Re-view Committee prior to implementation at it's Meeting 86-78 on December 31, 1986. The inspector verified that the TCR was devel-oped, reviewed, and implemented in accordance with the requirements of AP-0018, Rev. 12, Temporary Change Control. Additionally, the licensee identified within the TCR the necessary procedure changes that will instruct the operator to manually energize the Channel 2 detector high voltage at the appropriate time. Special Order 86-134, issued December 31, 1986 provided appropriate guidance, and the revised instructions, to plant operators for their review and us The procedures changed by the licensee were OP-3000, Emergency Shutdown From Power and Op-2103, Reactor Startup and Shutdown. No inadequacies were identified as a result of reviewing the licensee's actions in this are The inspector also determined that the licensee's development and implementation of the TCR process, as administratively controlled by AP-0018, was fully responsive to past NRC concerns in the area of lifted leads and jumper control This matter is discussed fur-ther in Section 3 of this report as part of the closure of Unre-solved Item 50-29/84-20-0 b. System Alignment Inspection Operating confirmation was made of selected piping system trains. Ac-cessible valve positions and status were examined. Power supply and breaker alignments were checked. Visual inspections of major components were performed. Operability of instruments essential to system perform-ance was assessed. The following systems were checked during plant tours and control room panel status observations:

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Emergency diesel generator (EDG) unit

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Low pressure and high pressure injection systems

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Spent fuel cooling system

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Charging system (control board status observations)

No unacceptable conditions were observe c. Biweekly and Other Inspections (1) During plant tours, the inspector observed shift turnovers, compared boric acid tank sample analyses and tank levels to Technical Speci-fications requirements, and reviewed the use of radiation work per-mits and radiation protection procedures. Area radiation levels and air monitor use and operational status were reviewed. Veri fi-cation of tagouts indicated the action was properly conducte On November 21, 1986 the plant operators added boric acid and de-mineralized water to the boric acid mix tank at 9:40 a.m. The con-centration of the tank's contents was sampled at 2:10 p.m. and de-termined to fall within the TS 3.1.2.11.a.2 requirement of 12 to 12.5% by weight boric acid solution. The licensee's actions were determined to be in accordance with procedure OP-2167, Rev. 10, Boric Acid Mix Tank Makeup. No deficiencies were observed by the inspecto (2) Observations of Physical Security

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On October 21, 1986, the inspector questioned the on-duty security personnel about one of the CCTV monitors that had a picture out of focus. Following questioning by the inspector, and informing the on-shift Security Shift Supervisor of the concerns about the poor equipment performance, a maintenance request was submitted. Although the licensee is making pro-gress in the area of identifying and correcting equipment de-ficiencies, the threshold upon which requests are made to cor-rect equipment problems or deteriorated performance needs to be lowere The inspector noted that prior NRC concerns associated with snow removal activities at the site (i.e., maintaining the integrity of the isolation zone by not allowing snow to pile up within the zone) received attention by the Security Super-visor prior to the onset of the snow season. A memo from the Security Supervisor to the Maintenance Manager, issued on November 13, 1986, raised the level of awareness about the issue and requested actions by the maintenance department to preclude repetition of prior events. The inspector acknowl-edged the licensee's effort to be responsive to NRC initiatives in this are .

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During the early part of this inspection period the inspector noted that an excessive number of security alarms associated with the control room doors was continuing to occur. This condition had been recurring since the beginning of the year, and resulted in the need for the on-shift security force to respond to many alarms and assess each situation. Also, the condition was compounding existing security concerns because 1) the security force was below the authorized personnel com-plement, and 2) the security force was providing compensatory measures in response to the Regulatory Effectiveness Review conducted between July 28 and August 1,198 The inspector discussed this issue with the assistant plant superintendent and recommended expeditious remedial actio The inspector noted that the conditions were occurring because of 1) door hardware problems, 2) a differential pressure con-dition across the doors caused by control room air handling requirements, and 3) uncooperative plant personnel that would not insure that the doors fully close behind them on entrance or exiting the control roo On November 10, 1986, Special Order 86-113 was issued to dis-cuss the problem and to remind operations personnel to aid in preventing unnecessary alarms as a result of using these door The assistant plant superintendent issued a memorandum on November 12, 1986 to all department heads to obtain the co-operation of all plant personnel to take the necessary actions to preclude generating door alarm Additionally, signs were placed outside the control room doors to remind personnel for the need to insure door closure. The necessary long term hardware fixes have been identified by the license This item will be reviewed during routine facility inspection . Fire Protection and Housekeeping No inadequacies were noted regarding licensee housekeeping or fire protection practices. A strong commitment to proper housekeeping conditions and practices by the plant staff is routinely observed by the inspecto . Review of Radiological Controls Radiological controls were observed on a routine basis during the reporting period. Standard industry radiological work practices and conformance to radiological control procedures and 10 CFR Part 20 requirements were observe Independent surveys of radiological boundaries and random surveys of ronradio-logical areas throughout the facility were taken b the inspecto _, - . _ - . - -

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No deficiencies were identified in this area as a result of the inspector's review. Additionally, during this inspection period, the licensee has initi-ated significant facility modifications involving the access control point to the radiation contro.1 area. These modifications represent licensee self-identified improvements in efficiency.and control of radiological activities, involve substantial resource commitment, and represent a high level of man-agement commitment to further improve the Radiological Protection Program at

the plan . Review of Events Requiring Telephone Notification to the NRC

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The circumstances surrounding the following events, which required NRC noti-fication via the dedicated ENS-line, were reviewed. A summary of the inspec-tor's review findings follows or is documented elsewhere as noted below:

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At 11:50 a.m. on October 31, 1986, the NRC was notified in accordance with 50.72 (b)(1)(A) that a plant shutdown to hot standby (Mode III) was initiated at 11:05 p.m. This resulted from declaring the pressurizer wide

range level indicator channel PR-L-8 inoperable. This event is discussed in Section 7.b of this repor At 5

55 p.m. on January 6,1987, the NRC was notified in accordance with 50.72 (b)(2)(iii)(B) that the safety injection building's ventilation fans PR-V-1 and PR-V-2 were declared inoperable. This event is discussed in Section 7.d of this repor . Review of Plant Events At the start of the current inspection period, the plant was at 75% of rated power and returning to full power operations from the October 4,

, 1986 reactor scram that was the result of a partial loss of control air event. The event was caused by 1) the failure of an air pressure switch

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on the main (No. 3) control air compressor, 2) an inability of the back-up air compressors to rapidly restore the air pressure, and 3) recovery l procedure guidance inadequacies for resetting of the feedwater control valves upon lockup. This event was last discussed in Inspection Report

, 50-29/86-08, Section 7.

l Subsequently, on October 8, 1986 the operations department issued Special Order 86-95 that specified interim guidance to the operators that was needed as a result of equipment and personnel response to the even The licensee has issued alarm response procedure OP-3750 Rev. 7, Bailey

Valve Position Lock, that incorporates the guidance from the special l order. The licensee had conducted load tests on the control air compres-

! sors on October 17, 1986. This testing identified a deficient load /

unloader solenoid operated valve on the No. 1 control air compressor.

, In addition, the tests determined the need to adjust the lead backup l

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control air compressor starting pressure upwards from 85 to 90 psig be-cause of pressure switch setpoint drift. The licensee is in the process of revising procedure OP-4228, Rev. 8, Compressor Emergency Cooling Water

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Supply Test and Auto Start Test, which will more accurately verify proper backup compressor operation, provide for additional data recording, and s

add as an acceptance criterion that the backup compressors carry the loa '

Full details of the licensee's investigation, and appropriate near term and long term corrective actions, are contained in LER 50-29/86-012 issued on November 3, 1986. The inspector identified no violations and had no further questions of the licensee as a result of reviewing this event _and the corrective action b. On October 31, 1986, with the plant at 100% power, the licensee deter-

.._ mined that the main coolant system pressurizer wide range level channel PR-L-8 was not respor. ding accurately to changes in pressurizer leve This chann'el'provides the single high pressurizer water level trip input

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into the reactor protection system. Accordingly, the plant operators declared the channel inoperable and entered TS 3.3.1, Table 3.3-1, Action Statement No. 8 at 9:15 a.m.. This provides for the plant to be in hot standby wf thin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if the single channel cannot be returned to ser-vice. At 9:58 a.m. licensee personnel made a containment entry to troubleshoot the problem. The licensee determined that the channel's level transmitter was faulty. At 11:05 a.m. plant operators commenced an emergency load reduction to hot standby. The NRC was informed of the licensee's action at 11:50 a.m. via the ENS line. By 2:23 p.m., the plant was placed in hot standby. Proper plant operator attention to detail, with regard to off-normal plant instrument operation, and timely licensee investigation and corrective measures were strengths demon-strated during this even To correct the equipment deficiency, the licensee issued Maintenance Request 86-1493 on October 30, 1986. Maintenance procedure OP-6219, Rev. 2, Maintenance of Rosemount 1153B Transmitters, Section B, was utilized to effect the transmitter's replacement, and provided the necessary controls to insure that EQ program requirements were maintaine A temporary procedure change was made to this procedure to allow an operational leak check upon valving the transmitter and process tubing into service, which was an acceptable alternative to the hydro leak test that would otherwise be conducted in accordance with OP-fl05,

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I&C Department Pressure Leak Test. "Following post-maintenance testing l of the new' transmitter, the channel was declared operational at 10:20 The reactor was critical at 5:15 a.m. on November 1, 1986, with the plant at 100% of rated power on, November 3, 198 ~s .

The licensee issued LER 50-29/86-014 on November 28, 1986 in response ( to this event. The licensee is eva7aating the transmitter's failed sensor module for 10 CFR Part 21 reportability. No deficiencies were l

i identified by the' inspector during the review of this event.

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c. At 12:01 a.m. on December 14, 1986 the licensee commenced a controlled l load reduction to 50% power to conduct turbine valve testing in accord-l s ance with procedure OP-4225, Rev. 16 and main steam non-return valve

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operability testing in accordance with Procedure OP-4261. During the performance of OP-4225, plant operators observed load swings that were apparently due to oscillations of the No. 3 control valv With the plant at approximately 75% and 65% of full rated power, respec-tively, the licensee removed the No. 3 Boiler Feedwater Pump from service for a lubricating oil change and the No. 2 Heater Drain pump for repack-ing. Subsequent to the valve testing, the licensee conducted main con-denser tube leakage checks. A total of five leaking condenser tubes were discovered in the west water box and one leaking tube was located in the east water box. Following plugging of the defective tubes a plant load increase was initiated at 3:20 p.m. on December 14, 1986. The inspector reviewed Special Order No.86-127 dated December 12, 1986, which provided pre planned instructions for the scheduled load reductio Included with these instructions were appropriate precautions and reference to proce-dure AP-7104, Core Operational Limits. The plant returned to full load at 4:15 p.m. on December 15, 198 The inspector noted that the licensee consistently provides prior plan-ning and appropriate assignment of priorities for these scheduled load reductions. Ensuing activities are controlled with well written proce-dures and instruction No violations were identifie d. On January 6,1987 at 5:55 p.m. the NRC was notified via the ENS line that the exhaust fans, PRV-1 and PRV-2, for the safety injection building were declared inoperable. As a result of performing calibrations on the temperature switches that initiate the building ventilation system on high internal temperature, the licensee discovered that the thermal over-load relays for the fan's motor starters had tripped. Immediate correc-  :

tive measures included starting the fan following overload reset, and

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to assure continued operability of the fans, they were manually started once per hour. Subsequently, on January 12, 1986 the surveillance in-terval was extended to once per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift. Special Order 87-03 and 87-07 were issued on January 6 and 12,1987, respectively, to instruct

, the operators on the increased testing program for the fans, and required

, data gathering on fan performanc These fans support the operability of the ECCS system when the outside

ambient temperature is above 40 degrees fahrenheit. The outside tempera-j ture on January 6, 1987 was between the range of 9 to 33 degrees fahren-I heft.

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i An extensive investigation was conducted by the licensee to determine

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the cause of failure and appropriate corrective actions which included 1) disassembly, cleaning, inspecting, installing new bearings; 2) flush-  ;

ing and lubrication of the fan bearings with EQ lubrication that has an

! excellent operating temperature range for the expected operating condi-tion for the fans; and 3) review of system design and operating condi-tions. This latter item determined that the observed operability problem had not been detected earlier because the fans were usually started in

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l mild weather, and that a change to the motor starter heater selection was warranted. Further, the licensee is performing an evaluation to determine the adequacy of the sizing of the existing fans. A surveil-

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lance procedure is.in the process of being developed to check the safety injection building ventilation system operability once per seven day The inspector will continue to review the licensee's remedial actions during subsequent routine inspection As part of Special Order 87-07, the Operations Department provided guid-ance to the plant operators on the actions they should take when both PR-V-1 and PR-V-2 fans are inoperable, since they are a subsystem of the safety injection system. Additionally, the special order raised the issue about single failures but, other than indicating that Yankee Nuc-lear Services Division was looking into it, did not provide guidance to the operators on what actions to take should only a single fan become inoperable. The licensee has determined that active ventilation is not required until the ambient temperature is greater than 40 degrees fah-renheit. The inspector requested that the licensee resolve this issu The inspector's concerns were based upon past modifications implemented by the licensee in the ventilation system that were responsive to the NRC's Systematic Evaluation Program concerns for the potential of single-active-failure vulnerability of the ventilation syste . Maintenance Observations The inspector observed and reviewed maintenance and problem investigation activities to verify compliance with regulations, administrative and mainten-

ance procedures, codes and standards, proper QA/QC involvement, safety tag use, equipment alignment, jumper use, personnel qualification, radiological controls for worker protection, fire protection, retest requirements and re-portability per Technical Specifications. The following activities were in-cluded:

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Maintenance Request (MR) 86-1454, BAMT (Boric Acid Mix Tank) Level Transmitter SI-LT-221 Will Not Calibrate Properly

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MR 86-1462, SI-FT-6 HPSI (High Pressure Safety Injection) Header Flow Transmitter EQ Cable Terminations 4 --

MR 86-1493, PR-L-8 Pressurizer Wide Range Level Trending Downward

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MR 86-1480, Pinhole Leak in Sensing Line For No. 3 MCP (Main Coolant Pump) Flow Indicator

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MR 86-1506, MC-TR-2 T/C-D-8 Main Coolant Temperature Green Pen Does Not Operate

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MR 86-1552, Replace ETLs (Electro Thermal Links) on CREACS (Control Room

, Emergency Air Cleaning System) Fire Damper Following the Performance of

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Fire Detection System Testing

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MR 86-1565, Transformer Oil Cooler Deluge System Switch Module Failure

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MR 86-1582, Low Battery Voltage on Safety Injection Building Fire Panel

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MR 86-1711, P-6-2 Service Water Pump Has Excessive Motor Noise

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MR 86-1732, SIAS (Safety Injection Actuation System) High Containment i Pressure Switch SI-PS-239 Replacement

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MR 87-0047 & 0049, Emergency Diesel Generator Nos 1, 2, & 3 Starter Motor Contact Replacement / Inspection Based upon a review of licensee activities in this area the inspector noted the following:

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Regarding MR 86-1711, the licensee initiated maintenance activities to inspect the motor, air circuit breaker (ACB), pump, and check valve

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associated with the No. 2 service water pump. The licensee utilized the following procedures to perform the maintenance activities: OP-5650, Re , Maintenance of Service Water Pump No. P6-2; OP-5755, Rev. 8, Inspec-tion and Maintenance of ACB and/or Contactor; OP-5756, Rev. 6, Inspection _

, and Maintenance of Electric Motors. This MR was issued on December 17,

! 1986 and the work was initiated and completed on January 5,1987 and

! January 13, 1987, respectively. On December 23, 1986, the operations department issued Special Order 86-31 to the plant operators to utilize this service water pump only if absolutely needed because of the main-tenance department's preliminary investigation that indicated the pump

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had bearing problem On January 12, 1987, the inspector reviewed the re-assembly of the pump at it's installed location, which is in the intake structure. The in-spector verified that OP-5650 was being utilized by the maintenance per-sonnel at the worksite, and that the precautions and prerequisites were being adhered to. The inspector observed that the maintenance personnel were torquing the pump's discharge flange. According to the Pump In-l spection Data Sheet, OPF-5650.1, torque wrench No. 2347 was used and the only torque value written in the " Miscellaneous" section of the form was for the flange bolting torque of 160 ft-lb. The inspector reviewed the I calibrated equipment signout card for torque wrench No. 2347 (a 600 lb range) and noted that it indicated that for this MR it was used on January 6,1987 for the service water pump strainer. The calibrated equipment signout card for the 1000 ft-lb. torque wrench No. 3302 indi-cated that it was used on January 8 and 12,1987 to torque the service water pump check valve and discharge flange, respectively. The licen-see's procedure, OP-5650, contained neither instructions for torquing of the various components that were part of the service water pump main-tenance nor specified the proper values to be used.

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Following discussions with maintenance department personnel on their torquing practices, which apparently utilizes an informal process to arrive at desired bolt torquing for non-Class 1 applications, the in-spector requested that they consider 1) a review of their torquing prac-tices, 2) insure that for safety related maintenance that proper controls are in place such that the selected torque value is consistent with the bolting in use, and 3) consider the development of a maintenance proce-dure that will specify fermal guidelines to be considered when bolting is to be torqued. The acting maintenance manager acknowledged the in-spector's comments and concerns, and indicated that the aforementioned items would be taken under advisement. In the last SALP report, 50-29/

85-98, that resulted from a December 4, 1986 SALP board meeting, a con-cern was raised that reflected the fact that proper documenting of main-tenance was not being accomplished. The inconsistencies and lack of

proper documentation of the above described torquing activities suggest the need for more aggressive management attention to this area. The inspector had no further questions on this ite The inspector reviewed the licensee's repair activities associated with MR 86-1493, which covered the replacement on an EQ transmitter in the Pressurizer wide range level instrumentation channel. The inspector's review is documented in Section 7 of this repor . Surveillance Observations The inspector observed tests or parts of tests to assess performance in ac-cordance with approved procedures and LCOs, test results (if completed), re-moval and restoration of equipment, and deficiency review and resolution.

The following tests were reviewe OP-4656, Rev. 4, Functional Test of the NRV (Non-Return Valve) Main Steam Line Pressure Channels

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OP-4685, Rev. 1, Functional Test of the Second Level Grid Undervoltage Protection Relays

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OP-4644, Rev. 11, Functional Test of the Fire Detection Instrumentation

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OP-4500, Rev. 9, Weekly Check of the Station Batteries

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OP-4207, Rev. 22, Surveillance of the Station Power DC and AC Distribu-tion Systems and the Emergency Diesel Generators

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OP-4634, Rev. 12 Safety Injection Actuation High Containment Pressure

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Sensors (SI-PS-238 & SI-PS-239) Functional Test

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OP-4716, Rev. 6, Vapor Container Personnel Hatch and CA-V-765 Leak Test

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OP-4702, Rev. 11, Vapor Containment Type B & C Penetration Tests

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OP-4801, Rev. 14, Functional Test and Alarm Settings of the Process Radiation Monitoring Syste OP-4214, Rev. 12, Chemical Shutdown System Operability Check Based upon a review of the licensee activities in this area, the inspector noted the following: During the performance of OP-4207 on January 13, 1987, the surveillance identified that the No. 3 emergency diesel generator (EDG) would not star Plant operators implemented TS 3.8.1.1 Action Statement (a),

which requires demonstration of the operability of the remaining AC sources by performing TS Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.5. The applicable portions of procedure OP-4207 which satisfies these requirements were performed within the required time specified in the TSs. Additional response to the surveillance test failure by the licensee included the initiation of MR 87-047, which re-sulted in determining that the contacts in the starter for the EDG start-ing motor were defective. Following repair and post-maintenance testing, the No. 3 EDG was returned to service. The licensee's expeditious re-sponse to the surveillance test discrepancy resulted in a timely return to service, in that only four hours of the 72-hour limiting condition for operation were utilized. As a preventive measure, the licensee issued MR 87-049 to inspect and clean the starter contacts on EDG No I and 2. This activity was completed in January 14, 1987. The failure of the plant's EDGs, either during surveillance testing or when called upon to function in response to an event is a rare occurrence at this facility. The inspector had no further questions of the licensee on this matte On December 23, 1986, the licensee performed OP-4634 and found that the containment pressure switch SI-PS-239 failed to operate. The I&C Department notified the plant operators of the problem, who in turn declared the instrument channel inoperable per TS Table 3.3-2, Action Statement 10. The instrument and controls (I&C) technician who per-formed the monthly functional test tapped the switch and observed that it subsequently functioned within allowable limits. The switch con-tinued to operate properly following repeated testing and within approximately ten minutes the channel was declared operable. As a precaution, however, the I&C Department replaced the switch with a new unit from stock. MR 86-1732 was issued to control and document the safety-related maintenance, with procedure AP-6007 Rev. 5, I&C Department Corrective Maintenance used to provide the method for performing the corrective activitie The new switch was functionally tested satisfactorily using OP-4634. The licensee re entered the TS Action Statement to facilitate the switch replacement. From the time the switch was determined to have failed the surveillance until it was replaced with a new unit and retested, a period of three and one-half hours had elapsed. The TS Actien Statement allows a six-hour period to effect repairs or be in Mode _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

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Based upon discussions with the I&C Supervisor, the inspector learned that the licensee considered two options. The first reflected the switch replacement, and the second involved retaining the original switch and increasing the surveillance frequency. Plant management opted for the first option. The switch that was removed from service was not returned to stock; however, the I&C department did not disassemble and visually inspect the switch internals. As a result of inspector concerns relative to the licensee not fully performing root cause determinations of this equipment failure, the I&C Supervisor agreed to improve on their root cause analysis techniques in the future. The inspector reviewed the I&C department's performance trending charts for the four identical pressure switches used to detect high containment pressure, which included pres-sure switch SI-PS-239, and noted that this switch was last replaced on August 19, 1985. The inspector has determined that the licensee closely tracks the per_formance of these switches to detect deterioration in per-formanc No violations were identifie . Onsite Review Committee Activities On December 10, 1986 the inspector observed the meeting of the Yankee NPS on-site review committee to ascertain that the provisions of TS 6.5.1 were me During this meeting, the committee reviewed OP-3305 Onsite Medical Emergency and OP-3330, Emergency Radiation Exposure Control. The inspector noted a minor discrepancy between these procedures that involved the requirement for management approval to allow individual exposures to exceed 5 Rem in emergency conditions. This condition was brought to the attention of the licensee sub-sequent to the completion of the meeting. The licensee agreed to review the inspector's comments and take appropriate action. The inspector had no fur-ther questions on this matte As a result of reviewing the licensee's activities in this area, no violations were identifie . Cold Weather Preparations The inspector reviewed implementation of the licensee's program on extreme cold weather protective measures to determine whether the licensee has 1) in-spected systems susceptible to freezing to ensure the presence and operability of heat tracing, space heaters, and/or insulation; 2) set the thermostats properly; and 3) energized the heat tracing and space heating circuit In preparing the plant for cold weather operation, the operations department implements OP-2115, Rev. 13, Warm or Cold Weather Operations. The inspector determined that this procedure was completed by the licensee on October 31, 1986. To ensure the operability of the plant's heat tracing system, the main-tenance department implements OP-5751, Rev. 9, Heat Tracing Inspections, which was verified by the inspector to be completed on October 31, 198 .

Throughout the cold weather period, the auxiliary operators for the primary and secondary sides of the plant perform routine shift checks on the status of the heat trace and heating of systems and structures. The inspector re-viewed the PA0 and SA0 Log Sheets regularly and verified these routine acti-vities were being accomplished. The licensee's actions associated with cold weather preparations were determined to be performed in accordance with ap-proved plant procedures and consistent with commitments made in its response to IE Bulletin No. 79-24, Frozen Lines (WYR 79-123, October 10,1979).

Additionally, the inspector reviewed plant reporting records, determined that events involving frozen systems or components are infrequent and, when they do occur, result in corrective action to preclude recurrenc . Plant Information Reports Plant information reports (PIRs) prepared by the licensee per AP-0004 were reviewed. The inspector determined whether the conditions were reportable

,

as defined in the TS and whether the licensee's system of problem identifica-tion and corrective action is being effectively utilized. The following PIRs were reviewed:

PIR N Occurrence Date Report Date Subject 85-11 11/11/85 01/22/86 Nos. 2 & 3 Steam Generator Feed-ring Cracks

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85-13 11/10/85 02/13/86 Series 20 Electro Switch Instal-lation Problem 85-15 12/29/85 02/19/86 No. 3 Boiler Feed Pump and Motor Failure Except for the following comments, the inspector had no further question PIR 85-11: This event was described in Inspection Report 50-29/85-18, Section Inspector concerns pertaining to the need for the licensee to strengthen procedural controls associated with maintenance, surveillance and operations activities, that are used to ensure containment integrity is being maintained when required, will be reviewed during the followup to Inspector Follow Item 50-29/85-18-0 PIR 85-15: This event was described in Inspection Report 50-29/85-24, Section The inspector identified no violations regarding the licensee's actions as-sociated with these events, and noted that the licensee determined the cause of the occurrence and specified appropriate short term and long term correc-tive action _

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13. Licensee Response to IE Bulletins The licensee's response to the following IE Bulletin (IEB) was reviewed. This review included: adequacy of the response to IEB requirements, timeliness of the response, completion of identified corrective actions and timeliness of completio IE Bulletin No. 86-03: Potential Failure of Multiple Emergency Core Cooling System (ECCS) Pumps due to Single Failure of Air-Operated Valve in Minimum Flow Recirculation line, dated October 8, 1986. This IEB requested that lic-ensees determine if they could be affected by this problem, and take appro-priate corrective actions if requirea. The licensee was to provide a written report to the NRC within 30 days from receipt of the bulleti The inspector reviewed licensee letter FYR 86-110 to NRC:RI dated November 7, 1986. In this letter the licensee stipulated that the problem does not exist at the Yankee Nuclear Power Station (YNPS). In the attachment to their letter, the licensee described the plant's ECCS and demonstrated that the single failure proof design of this system precludes the bulletin's concern from occurring in the minimum flow recirculation lin This bulletin is close . Review of Licensee Event Reports Licensee Event Reports (LERs) submitted to NRC:R1 were reviewed to verify that the details were clearly reported, including accuracy of the description of cause and adequacy of corrective action. The inspector determined whether further information was required from the licensee, whether generic implica-tions were indicated, and whether the event warranted onsite followup. The following LERs were reviewe Report Report LER N Date Date Subject 50-29/86-01 3/11/86 4/10/86 TS Violation Concerning the Yankee Emer-gency Plan 50-29/86-02 3/11/86 4/10/86 Insufficient implementation Procedures for the Offsite Dose Calculation Manual 50-29/86-03 4/16/86 5/16/86 Failure to comply with a TS Action Statement 50-29/86-04 6/1/86 7/1/86 Reactor Scram - Loss of Heater Orain Pumps 50-29/86-05 6/2/86 7/2/86 Dose Equivalent I-131 Greater than Microcuries per Gram 50-29/86-06 6/19/86 7/18/86 Inoperable No. 3 Steam Generator Blowdown Monitor

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Report Report LER N Date Date Subject 50-29/86-12 10/4/86 11/3/86 Plant Trip on Low SG Level Due to Loss of Control Air 50-29/96-14 10/31/86 11/28/86 Plant Shutdown Due to Loss of Pressurizer Wide Range Level Channel LER 50-29/86-01, TS Violation Concerning the Yankee Emergency Plan (E-Plan). This event involved a change to the E-Plan without a plant opera-tions review committee (PORC) review, which resulted from a lack of cor-porate level administrative controls. The event was identified during the performance of a quality assurance (QA) audit, with a Nonconformance Report (NCR) No. 86-14 being issued. The inspector considers this event to be a licensee-identified violation, and in accordance with the provi-sions of 10 CFR 2, Appendix C, a Notice of Violation will not be issue LER 50-29/86-02, Insufficient Implementation Procedures for the Offsite Dose Calculation Manual (00CM). As above, '.his event was discovered during the conduct of a QA audit. The licensee attributed this event to personnel error at the corporate headquarter's radiological engi-neering group. Appropriate corrective measures were implemented by the licensee to correct the immediate condition, with actions specified to preclude similar events from occurring in the future. The licensee issued NCR No. 86-12 as a result of this event, which is also con-sidered to involve a licensee-identified violation for which a Notice of Violation will not be issue LER 50-29/86-03, Failure to Comply with a TS Action Statemen Documen-tation of the licensee's corrective actions and inspection findings is contained in Inspection Report 50-29/86-05, Section LER 50-29/86-04, Reactor Scram - Loss of Heater Drain Pumps. Documenta-tion of the licensee's corrective actions and inspection findings is contained in Inspection Report 50-29/86-05, Section LER 50-29/86-05, Dose Equivalent I-131 Greater than 1.0 Microcuries per Gram. This event resulted from a plant scram in June 1, 1986 and in-volved cladding defects in one or more fuel rods from second cycle Core XVIII Exxon Nuclear Fuel Company fuel assemblies. Documentation of the event is contained in Inspection Report 86-05, Section This LER was submitted to meet the reporting requirements of TS 3. LER 50-29/86-06, Inoperable No. 3 Steam Generator Blowdown Monito Documentation of the licensee's corrective actions and inspection fhd-ings is contained in Inspection Report 50-29/86-08, Section .

19 LER 50-29/86-12, Plant Trip on Low Steam Generator Level Due to Loss of Control Air. Documentation of the licensee's corrective actions and in-spection findings is contained in Section 7 of this inspection report and Inspection Report 50-29/86-08, Section . LER 50-29/86-14, Plant Shutdown Due to Loss of Pressurizer Wide Range Level Channel. Documentation of the licensee's corrective actions and inspection findings is contained in Section 7 of this repor . Inspection of General Electric HGA Relays The inspector was directed by NRC: Region I Division of Reactor Projects to perform an inspection to determine if General Electric HGA relays are utilized in functions that are important to safety at the YNPS and ascertain the potential consequences of maloperation as a result of contact chatter initi-ated by seismic activity. This inspection was conducted in accordance with Region I Temporary Inspection No. RI-86-03, dated October 21, 1986. The re-sults of this inspection are contained in the tabulation below:

i RELAY RELAY FUNCTION AND POTENTIAL SYSTEM DESIGNATION MODEL N CONSEQUENCE OF MAL-0PERATION Main Coolant K4 & 2K4 12HGA111J9 Energized by low main coolant Pressure coolant pressure to initial SIA Chanel Contact chatter can initiate a Safety Injectio Process Radi- RM-K-108 12HGA De-energizes on hi radiation, ation Moni- RM-K-109 channel failure, and abnormal toring RM-K-111 switch positio Intermittent operation of solenoid operated air valves on the component cooling vent line, steam genera-tor blowdown line, and the test tank effluent line will occu Containment K21 & K22 12HGA111J2 Energized by high containment Isolation pressure. Contact chatter will cause containment isolation and shut the non return valves (NRVs).

K9, K10, KJ1, 12HGA111J2 Energized by CIAS Signa Con-K12 tact chatter can cause contain-ment hydrogen vent valves to cycl K-Cl-31 & 12HGA11A52F Energized by SIAS Actuatio K-C1-32 Contact chatter can cause con-

, tainment isolation.

l l

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RELAY RELAY FUNCTION AND POTENTIAL SYSTEM DESIGNATION MODEL N CONSEQUENCE OF MAL-0PERATION

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Non-Return K40, K408, 12HGA111J2 Energized to Close NRVs. Contact Valves K42A, K428, Contact chatter can cause NRV K44A, K448, closur K46A, K46B K3AX, K3BX, 12HGA111JL Energized for 10% Test. No K4AX, K4BX, adverse consequences due to KSAX, K5BX, contact chatte K6AX, K6BX,

K41A, K41B,

' K43A, K43B, K45A, K45B, K47A, & K478

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Reactor Trip K53A & K538 12HGA111J2 Energized on trip signal from

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NRV System or main coolant pres-sure channels. Contact chatter can actuate the trip coils of the reactor trip breakers, i

Safety R1 & R2 12HGA11J52 Energize with SIAS Actuatio Injection No adverse consequences due to contact chatte Emergency 74C-1, 74C-2, 12HGA11J52 De-energized with loss of an EDGs Diesel & 74C-3 starting circuit power. Alarm Generator function only.

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27 (6-3), 27 12HGA11J70 De-energized with loss of 480V (4-1), & 27 a-c bus. No adverse consequences (5-2) due to contact chatte Containment PS-239X 12 HGA Energized by high containment i

Pressure pressure. Contact chatter will

Channel initiate a safety injectio '

Currently, the NRC is reviewing the YNPS design as part of the Systematic Evaluation Program (SEP). As stated in Paragraph 4.11 of the Integrated Plant Safety Assessment, NUREG-0825, dated June 1983, the operability of safety-

, related electrical equipment with respect to seismic design considerations (SEP Topic III-6) will be addressed under the NRC's Unresolved Safety Issue

. A-46. The licensee has designed and installed a seismic hot shutdown system, to resolve the issues associated with SEP Topic III-6, Seismic Design Con-siderations. The inspector has determined that none of the HGA relays are utilized in this syste !

The inspector had no further questions on this item at this tim ___ . . _ _ . _ - - _ _ _ . _ _ _ _ _ . _ _ . _ __ _ _ __ ___.__ _ _._ _ _ _ _

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16. Organization and Administration During the inspection period, the inspector reviewed changes to the licensee's staff or organization structure as described below. The review included:

verification that licensee's onsite organization structure is as described in the facility TS, and verification that personnel qualification levels are in conformance with ANSI N18.1-1971, as described in TS Section 6. As a result of the retirement of the plant operations manager, the lic-ensee announced the promotion of operations department personnel for the positions of plant operations manager, assistant plant operations manager, and shift supervisor. The personnel assignments were effective December 1, 198 On December 29, 1986, the licensee announced changes in the technical director's organization structure that became effective January 1,198 These changes reflect 1) that the training department manager will report directly to the technical director and, 2) a new technical assistant director position was created that will report to the technical directo The licensee indicated, that, a proposed change to the technical speci-fications will be submitted to the NRC to reflect the organization change No deficiencies were identifie . Management Meetings During the inspection period, the following management meetings were conducted or attended by the inspector as noted below:

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The inspector attended an exit meeting held on October 24, 1986 by an NRC team at the conclusion of Inspection 50-29/86-18, review of the lic-ensee's implementation of a program for establishing and maintaining the qualification of electric equipment within the scope of 10 CFR 50.4 The inspector attended an exit meeting on November 7, 1986, by a region based licensing examiner at the conclusion of Inspection 50-29/86-20, review of the licensee's operator licensing requalification training progra An onsite management meeting was held on January 6, 1986 between Region I and Yankee Atomic Electric Company to discuss the results of the NRC Systematic Assessment of Licensee Performance review ccnducted for the period of February 1,1985 through October 6,1986. This assessment is documented in Inspection Report 50-29/85-9 At periodic intervals during the course of the inspection period, meet-ings were held with senior facility management to discuss the inspection scope and preliminary findings of the resident inspector.