IR 05000309/1986020

Insp Rept 50-309/86-20 on 861229-870202.No Violations Identified.Major Areas Inspected:Control Room,Accessible Parts of Plant Structures,Plant Operations,Radiation Protection,Physical Security & Fire Protection

ML20211F360
Person / Time
Site:	Maine Yankee
Issue date:	02/12/1987
From:	Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20211F349	List: IR 05000309/1986020 ML20211F345
References
50-309-86-20, NUDOCS 8702250062
Download: ML20211F360 (7)

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U.S. NUCLEAR REGULATORY COMMISSION Region I Docket / Report: 50-309/86-20 License: OPR-36 Licensee: Maine Yankee Atomic Power Inspection At: Wiscasset, Maine Dates: December 29 - February 2, 1987 Inspector: C rneli F. Holden, Senior Resident Inspector Approved: // b Asio E. E. Trppp, Chief, Reactor Projects Section 3A M/

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Summary: Inspection on December 29, 1986 - February 2, 1987 (Report No. 50-309/86-02)

l Areas Inspected: Routine resident inspection (108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br />) of the control room, ac-cessible parts of plant structures, plant operations, radiation protection, physi-

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cal security, fire protection, plant operating records, maintenance and surveil-lanc Results: No violations were identified. An unplanned release of 420 curies of noble gas occurred on December 30, 198 One allegation was received and closed out as unsubstantiated.

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DETAILS Persons Contacted Within this report period, interviews and discussions were conducted with various licensee personnel, including plant operators, maintenance technicians and the licensee's management staf . Summary of Facility Activities Since Last Inspection Report On December 29, 1986 the plant was at 100 percent power. On December 30, while adding hydrogen to the Volume Control Tank (VCT), an auxilary operator bumped the isolation valve for the nitrogen system which led to the release of 420 curies of primarily noble gases from the VCT via a blank flange on the nitro-gen lin Power was reduced on January 3,1987 to 75 percent for monthly surveillance testing of the turbine governor and stop valve Power was re-turned to 100 percent on January The plant began cold leg temperature coastdown operations on January 17, 198 At the end of this report period plant power level was 95 percent powe . Followup on Previous Inspection Findings (Closed) Unresolved Item (309/82-08-09). Offsite Committee to formalize methods of meeting ANSI requirements. The inspector reviewed changes to the Nuclear Safety Audit and Review Committee (NSARC) Charter which formalized the methods by which the NSARC meets ANSI requirements, (Closed) Violation (309/82-08-10). Failure to review three changes in

accordance with 10 CFR 50.59. The Nuclear Safety Audit and Review Com-mittee (NSARC) reviewed the three design changes as noted. Additionally, plant procedures have been changed so that all design change safety evaluations are now reviewed by NSAR (Closed) Followup Item (309/82-26-03), Licensee Verification of Safety i Analysis Assumptions; (Closed) Unresolved (309/83-01-02), Implementation of Measures to Prevent Operation Outside of the Assumptions of Plant Safety Analysis; and (Closed) Followup Item (309/82-08-08), Methods to Ensure Maintenance of Commitments. The licensee implemented an Inputs and Assumptions Source Documer.t (IASD) in July 1986. This document pro- .

vides a means for directly comparing plant operations and safety analysis assumptions to ensure safe operations as well as providing a reference for procedural development, engineering design changes and safety analy-i sis inputs. Procedures are inplace to utilize and update the IASD as required. Additionally, Procedure 0-05-01, Procedure Preparation, Clas-sification, Format, Use and Adherence, Rev. 1, requires that individual steps of procedures that fulfill a commitment, be referenced. This en-sures that future changes will maintain those commitments.

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3 (Closed) Unresolved Item (309/84-11-04). Non-Fire-Rated gap between Appendix R areas. The inspector observed portions of the seismic gap between adjacent Appendix R areas that are now sealed with rated fire protection foa (Closed) Followup Item (309/85-06-01). Instrument air lines are con-taminated with dirt. The licensee determinea that the probable cause for dirt in the air lines was faulty air dryers which overheated the afterfilters and contaminated the air lines. The licensee has replaced the air dryers and conducted a program of filter replacement (approxi-mately 450 filters exist in the instrument air system). (Closed) Followup Item (309/85-01-01), Air Lock Surveillance Test. The licensee reviewed the possibility of changing the containment personnel assess hatch air test in order to more closely simulate the manner in which accident pressure would affect the door seals. The licensee de-termined that other methods of testing were not feasible. The licensee has enhanced the routine testing of the personnel air lock to ensure meaningful results are obtained and corrective action, where appropriate, is effective. Management is involved when failures occur during routine surveillance tests of the personnel air lock and corrective maintenance is given high priority. When the personnel hatch air test failed on January 26, 1987, a special Plant Operations Review Committee (PORC)

meeting was held to discuss operability of the air lock and corrective maintenanc The hatch was repaired and tested on January 27, 1987. . Routine Periodic Inspections Daily Inspection During routine facility tours, the following were checked: manning, ac-cess control, adherence to procedures and LCO's, instrumentation, recor-der traces, protective systems, control rod positions, containment pres-sure, control room annunciators, radiation :aonitors, emergency power source operability, control room logs, shift supervisor logs, and operat-ing order System Alignment Inspection Operating confirmation was made of selected piping system trains. Ac-cessible valve positions and status were examined. Power supply and breaker alignment was checked. Visual inspection of major components was performed. Operability of instruments essential to system perform-ance was assessed. The Auxilary Feedwater System was reviewed. No dis-crepancies were identifie .. - - - . . . ... -

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. Biweekly Inspections During plant tours, the inspector observed shift turnovers, chemistry sample results and the use of radiation work permits and Health Physics procedures. Area radiation and air monitor use and operational status was reviewe Plant Housekeeping and cleanliness were evaluate , Plant Maintenance The inspector observed and reviewed maintenance and problem investigation

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activities to verify compliance with regulations, administrative and maintenance procedures, codes and standards, proper QA/QC involvement, safety tag use, equipment alignment, jumper use, radiological controls for worker protection, fire protection, retest requirements, and report-ability per Technical Specifications. The repair of MS-P-168, steam pressure control valve for auxilary feedwater pump P-258 was reviewe No discrepancies were identifie Surveillance Testing The inspector observed parts of tests to assess performance in accordance with approved procedures and LC0's, test results, removal and restorttion of equipment, and deficiency review and resolution. Surveillance tests witnessed included diesel generator overspeed test and containment hatch pressurization tes Control Room Environment The inspector reviewed the control room environment and determined the status of the work atmosphere in which the operators are required to function. Recent improvements in control room access has limited access and provided a work area for processing of work orders / tag outs. Busi-ness is conducted in an orderly manner with few distractions. Alarm

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conditions are kept to a minimum and are aggressively pursued to correc-i tion. The professional appearance of the control room and operators dis-plays the sense of pride they take in the performance of their dutie ' Emergency Diesel Generator The emergency diesel generator (EDG) IB was tagged out of service on January 13, 1987 for preventive maintenance. Scheduled maintenance in-cluded lube oil and fuel oil filter check, brush check and an overspeed trip test upon testing of the diesel after maintenance. The inspector witnessed the overspeed test which was performed satisfactorily. At the conclusion of the test a small amount of smoke was noticed coming from the EDG control cabinet. The diesel generator was again tagged out and troubleshooting was begun to determine the cause of the smoke.

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A burned relay was identified as an "SFA" relay, Start Fail Auxilar The apparent cause of the failure was metal filings in the relay which prevented full travel of the relay armature causing arcing of the con-tacts and overheating. This is considered an isolated incident. The relay was replaced. The probable source of the metal filings was a re-cent installation of a NVR relay which required drilling in the vicinity of the SFA relay for seismic mounting of the NRV relay. Other relays were inspected in the area for metal filings and none were foun Following completion of maintenance the tags were cleared on EDG 1 When the tag was cleared from the " Start 1" breaker and the breaker was closed, the EDG 1A auto started. The diesel was again tagged out of service. Three sets of contacts were identified as probable causes for the activation of the start circuitry. These contacts are located in a different control panel in the EDG room from the SFA relay. All three sets of contacts were manually repositioned to check for mechanical binding. No binding was identified. All three sets of contacts were replaced. These contacts are designated Start Failure Backup (SFB)

1 & 2, Pinion Failure (PFD) 1 & 2 and Start Failure Delay (SFD) 1 & During preparations for retesting EDG 1A, a portion of the mechanical linkage on the fuel rack was broken. This part was replaced. EDG 1A was returned to service following a two hour post maintenance run at 3:00 p.m. on January 16, 1987. During the time frame when EDG 1A was out of service, the licensee conducted a daily test of EDG 18 as required by Technical Specification 3.1 No significant concerns were identified during inspection coverage of the above activitie . Gaseous Release On December 30, 1986 at approximately 3:50 a.m. the primary auxilary operator was directed to add hydrogen to the Volume Control Tank (VCT). This requires the manipulation of two manual valves in the Primary Auxilary Building (PAB).

During plant operations when the boron concentration of the primary coolant is being diluted, hydrogen levels in the VCT are depleted and must be re-plenished. The operator opens two manual valves which unisolates a regulator and increases the hydrogen overpressure in the VCT. Connected to the hydrogen line downstream is a nitrogen line. One valve isolates a regulator and a blank flange. A flange is aed to prevent inadvertant Nitrogen addition to the VCT since this type of puile is used primarily during plant outage Several minutes after the manipilation of the hydrogen valves was performed a high primary vent stack gaseous monitor alarmed. Plant operators took action in accordance with Abnormal Operating Procedure (A0P) 2-25, High Radiation Levels, and began to survey the PAB for sources of a leak. At 4:35 a.m. plant operators located a leak on the blank flange in the nitrogen system and tightened the flange to isolate the leak (the leak was only apparent after the operators identified that nitrogen valve N-7 may have been bumped and thereby pressurized the blind flange). Radiation levels in the PAB returned to norma . _

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i The licensee calculated the highest release rate and the amount of activity (420 curies) released. Both of these values were extremely small compared to the Technical Specifications limits for gaseous releases (less than 0.004%

of quarterly limits). Additionally,.the licensee conducted a Human Perform-ance Evaluation of the incident to determine the root cause of the leak. The results of this evaluation indicated three areas for action. These included requesting maintenance check the operation of check valve (N-6), a recommenda-tion for removing the handwheel from valve N-7 and requiring maintenance to install and remove the spool piece / blank flange in the nitrogen line. The inspector questioned whether other release paths had been reviewed to deter-mine if the potential exists for similar problems. The licensee indicate 1

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that they had not identified similar potential release pathu; the inspector will continue to review this area in future inspection (s). Observations of Physical Security Checks were made to determine whether security conditions met regulatory re-quirements, the physical security plan, and approved procedures. Those checks

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included security staffing, protected and vital area barriers, vehicle searches and personnel identification, access control, badging, and compensatory meas-ures when required.

' Radiological Controls l Radiological controls were observed on a routine basis during the reporting period. Standard industry radiological work practices, conformance to radio-logical control procedures and 10 CFR Part 20 requirements were observe . Relay Inspection The inspector reviewed the application of General Electric (GE) HGA relays at Maine Yankee. The objective of this inspection was to determine how GE HGA relays are utilized at Maine Yankee. Some HGA relays were determined to chatter during seismic testing for another nuclear power plant. Based on this information the inspector determined the affect of contact chatter on the

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specific applicatio HGA relays are employed in a variety of safety related functions at Maine Yanke HGA relays provide loss of control power alarms for various safety related motors. The HGA relay is in parallel with the closing mechanism for the assigned breaker. On a loss of control power the relay would deenergize and cause an alarm in the control room. Contact chatter during a seismic event would not affect the operability of the breaker since the relay only provides for a control room alar HGA relays are also used to provide overcurrent alarms to the control room for certain safety related motors. Once again, these relays only serve to give an indication of an overcurrent condition and do not affect the operability of the pump .

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HGA relays also provide for automatic starting of the two electric driven emergency feedwater (EFW) pumps on low level in the steam generator. This is an energize to start function. Contact chatter could cause the pump to start prematurel HGA relays are used to prevent a low suction pressure trip of the high pres-sure safety injection (HPSI) pumps. During normal operations the charging pumps are provided with a low suction pressure trip to prevent damage of the pump due to cavitation. During a safety injection sequence, the charging pumps become the HPSI pumps and are protected against low suction pressure trips by an HGA relay. The HGA relay must energize to prevent the low suction pressure trip. If a seismic event were to cause contact chatter it it possible the HPSI pump would trip with a concurrent low suction pressure condition, however, the HPSI pump would immediately restart upon the return of the HGA relay to its required positio In the case of the EFW and HPSI application of the HGA relay, both are in an energized state to produce the pump start or prevent the low suction pressure tri The EFW and HPSI pumps are controlled from the main control board in the control room. All pumps are instrumented to show breaker status, running amperage, flowrates, etc. Contact chatter appears to be less likely while the HGA relay is energized since the coil would tend to maintain the u 7 tacts in the required position versus spring pressure when the relay is deenergize Several different events must take place in sequence to arrive at a situation where a seismic event could affect the operability of the pum The inspector presented these findings to the licensee and indicated that this area was currently receiving generic review by NR . Allegation No. RI-86-A-102 The inspector received an anonymous allegation regarding the qualifications of the computer section head. The allegation stated that the individual re-cently promoted to computer section head did not meet the requirements of Regulatory Guide 1.8 or American National Standards Institute (ANSI) N18.1,

Selection and Training of Nuclear Power Plant Personnel.

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Due to a recent temporary reassignment of the Radiation Protection Manager, the plant made several personnel changes. As a result the D mputer Section Head was assigned as the Assistant to the Technical Support Department Hea The inspector determined that this assignment is in addition to the normal duties he maintains as the Computer Section Head. The inspector verified the qualifications of the Computer Section Head. Based on this review, the in-spector concluded the Computer Section Head meets the requirements of Regula-tory Guide 1.8 and ANSI N18.1. Allegation number RI-86-A-0102 is closed.

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10. Exit Interview Meetings were periodically held with senior facility management to discuss the inspection scope and findings. A summary of findings for the report period was also discussed at the conclusion of the inspection.

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