IR 05000271/1994013

Insp Rept 50-271/94-13 on 940515-0625.Violations Noted.Major Areas Inspected:Operations,Maint,Engineering,Plant Support & Safety Assessment & Quality Verification

ML20149F101
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	08/04/1994
From:	Conte R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20149F094	List: IR 05000271/1994013 ML20149F089 ML20149F099
References
50-271-94-13, NUDOCS 9408100004
Download: ML20149F101 (29)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No.

94-13

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Docket No.

50-271 Licensee No.

DPR-28 Licensee:

Vermont Yankee Nuclear Power Corporation RD 5, Box 169 Ferry Road Brattleboro, VT 05301 Facility:

Vermont Yankee Nuclear Power Station Vernon, Vermont

Inspection Period:

May 15 - June 25,1994 Inspectors:

Ilarold Eichenholz, Senior Resident Inspector Paul W. IIarris, Resident Inspector l

Approved by:

M#d dr IN4 Richard J. Conte, Chief C/

/D/te ~

Reactor Projects Section 3A Scos:

Station activities inspected by the resident staff this period included Operations, Maintenance, Engineering, Plant Support, and Safety Assessment and Quality Verification. Backshift activities amounting to 11.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> were performed on May 18,19, 31 and June 1. Interviews and discussions were conducted with members of Vermont Yankee management and staff as necessary to support this

inspection.

Findines:

An overall assessment of performance during this period is summarized in the

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Executive Summary. A violation of NRC requirements was identified regarding inadequate safety assessments performed by the Plant Operations Review

Committee in their review of: (1) reactor coolant system leakage into the "A" core spray subsystem low pressure piping, and (2) cross-connecting the service water and fire water systems (Sections 2.5 and 4.2). Vermont Yankee review of the level of quality assurance applied to reactor coolant system pressure boundary components classified as Safety Class-0 was identified as an unresolved item (Section 4.2).

I 9408100004 940so4 PDR ADOCK 05000271 G

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EXECUTIVE SUMMARY Vermont Yankee Inspection Report 94-13

Operations

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A short duration main condenser vacuum transient caused by the inadvertent isolation of the steam jet air ejector system was promptly identified and responded to by the control room staff.

A power reduction and rod pattern exchange were conducted safely with appropriate management oversight.

The Vermont Yankee Plant Performance Summary report was considered a good initiative. Appropriate operator actions were observed in response to the loss

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of the fire suppression water system, however, the Plant Operations Review Committee (PORC)

failed to conduct reviews to detect potential safety hazards associated with the use of the cross

connection of the service water and fire water systems.

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afaintenance Timely corrective maintenance was performed to restore the "B" hydrogen / oxygen monitor.

Good performance occurred during the performance of residual heat removal system valve maintenance. Vermont Yankee's recovely from a fire in reactor water clean-up (RWCU) system control panel was performed in a technically competent and safe manner. A very good safety assessment of this event was performed by the PORC.

Engineering

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An Excellent initiative involving operating experience reviews by the Reactor and Computer Engineering Department resulted in a timely assessment that the cause of slow control rod scram times experienced at other nuclear facilities was not an operational concern a Vermont Yankee.

The technical evaluation performed for core spray injection valve CS-12A leakage was accurate

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and generally comprehensive. However, potential safety hazards associated with operational safety considerations involved with the closing of the core spray injection valve CS-11 A were

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not reviewed by the PORC.

l Plant Support Good performance of Vermont Yankee's Emergency Preparedness staff resulted in the timely maintenance of the FTS2000 emergency communication equipment, proposed enhancements to in-plant emergency communication equipment, and improved reliability of the emergency response pager system.

A positive initiative by the Security Department involved their sponsoring a Pennsylvania State Police briefing at the plant regarding the TMI unauthorized intrusion event. Additionally, the Security Department's investigation of a plant event that could have involved tampering, was well controlled and professionally conducted. Plant-wide clean ii

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(EXECUTIVE SUMMARY CONTINUED)

up activities in support of Vermont Yankee's housekeeping " vision" resulted in improved housekeeping, reductions in fire loading, and included active participation from all levels of the organization, including senior corporate managers.

Safety Assessment and Qualify Verification Inspections of the activities performed by the PORC detennined that their reviews of plant operations were not demonstrating a consistent high level of performance in the conduct of safety assessments necessary to detect potential safety hazards.

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SUMMARY OF FACILITY ACTIVITIES Vermont Yankee Nuclear Power Station (VY) continued full power operations this inspection period. On May 11, reactor power was reduced to approximately 75 percent rated power to perform a control rod pattern exchange and main steam isolation valve full closure testing. An increase in reactor coolant system (RCS) leakage past the "A" core spray (CS) system isolation

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valves was observed on May 17 resulting in closure of a second, normally open CS system isolation valve to prevent pressurization of low pressure piping.

On June 8, the members of the International Brotherhood of Electrical Workers, which represents all VY union personnel on-site, ratified a new 3-year labor contract. The existing and new contract end dates are June 20,1994 and June 20,1997.

2.0 OPERATIONS (71707, 62703, 93702, 37551, 92903, 90713, 92700)

2.1 Operational Safety Verification Daily, the inspectors verified adequate staffing, adherence *, procedures and Technical Specification (TS) limiting conditions for operation (LCO), operability of protective systems,

status of control room annunciators, and availability of emergency core cooling systems. Plant tours confirmed that control panel indications accurately represented safety system line-ups.

Safety tagouts properly isolated equipment for maintenance.

On June 6, control rod 22-15 was identified to be inserting slowly during recent exercising. The Operations Planning Group (OPG) Coordinator involved the Reactor and Computer Engineering (R&CE) Department in assessing this performance. Performance indicators for this control rod indicated sporadic elevated temperatures but, no concerns involving scram times were identified.

Further investigation by the Operations Department id':ntified that the sluggish control rod response was due to a directional control valve that was leaking by. Repairs were completed

by June 8, following verification by Yankee Nuclear Services Division (YNSD) that adequate Shut Down Margin was available to remove the control rod from service while it was at position

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"48".

Leaving the control at this position was necessary because of control rod configuration and power peaking concerns.

On June 7, control rod OC-15 experienced a failure of the position "31" switch in its rod position indication probe. This caused the control rod display to indicate position "98" in lieu of the actual position "48", and also resulted in the initiation of a rod drift alarm. Until the position indication probe can be replaced during a plant shutdown that includes a drywell entry,

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a Temporary Modification (TM) No.94-014 was installed to remove the masking rod drift alarm.

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The inspector observed appropriate operating staff performance in dealing with these deficient

conditions or problems, and determined that strong management support existed to correct these conditions.

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2.2 Main Condenser Vacuum Transient On May 17, an engineer performing a system walkdown to verify a design change installation and test procedure inadvertently dislodged an electrical connection associated with the steam jet air ejector (SJAE) system. The loss of power caused normally open SJAE isolation valve, OG-512B, to shut, isolating the main condenser from the SJAEs. The resulting reduction in offgas flow annunciated in the control room and control room operators (CROs) cbserved decreasing main condenser vacuum. Main condenser vacuum decreased from a steady state value of 1.37 inches-mercury absolute ("Ilg) to approximately 2.37 "lIg in ten minutes. The low condenser vacuum alarm occurs at 5 "Ilg with a subsequent turbine trip at 7.5 "IIg The CROs initiated the actions in accordance with the alarm response sheet and procedure OP 3120, Loss of Condenser Vacuum, and restored condenser vacuum within 10-minutes. System walkdowns are routinely performed in accordance with design change packages to verify as-built drawings and wiring diagrams, and routinely performed prior to maintenance or the installation of temporary (

modifications.

The inspector concluded that the loss of condenser vacuum was an isolated occurrence and of low safety significance. The control room staff identified the cause quickly and restored vacuum. The conduct of maintenance, surveillance, and equipment walkdowns have not resulted i

in repetitive occurrences of equipment failure or challenges to safe reactor power operations.

Vermont Yankee temporarily suspended field activities by the onsite maintenance / modification contractor and plant maintenance departments pending the completion of lessons learned and correcti"e actions.

2.3 Rod Pattern Exchange On Mcy 17, VY reduced reactor power to approximately 75 percent rated power to conduct a rod patiern exchange. Reactor operation information was provided to the control room operators describ ng the rate of power changes, conduct of surveillance testing, and the rod pattern exchange. The information regarding the rate of reactor power ascension was based on General Electric (GE) and YNSD analyses of core thermal performance for control rod withdrawal

sequences.

The R&CE Department personnel were observed to effectively use the process computer to i

predict estimated core thermal performance based on anticipated rod movements. Coordination l

with the control room staff was professional. Operators appropriately questioned recommended

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rod movements and verified that the moves were in accordance with the controlled copy move sheet. Plant management discussed the power reduction and associated activities during Daily Plant Manager's (PM's) meetings. The inspector reviewed the R&CE logs and identified no

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concerns.

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2.4 Plant Performance Summary The Operations Department (OD) evaluates and trends plant performance information to inform plant management of abnormal trends indicative of degrading plant or equipment performance.

The inspector reviewed the Monthly Summary of Plant Performance report for April 1994, to assess its effectiveness in informing plant management of significant trends important for safe plant operations.

This report, as described in procedure DP 4153, Plant Performance Monitoring, evaluated and trended steam cycle thermal performance, emergency core cooling system equipment performance and other operating parameters. Assessments were performed for emergency diesel generator failures, core spray leakage (Section 4.2), and safety / relief valve RV-2-71C bellows leakage (NRC Inspection Report 93-09).

The inspector concluded that this report represented a good initiative for the compilation of plant performance data for management review. The performance indicators assessed and trended focused on significant issues. According to the OD's Senior Operations Engineer, VY has identified the need to improve their assessment of trends and the documentation of corrective actions for identified degradations. For example, although the report accurately documented in-service testing (IST) results for safety system pumps, no assessment was performed and no corrective actions were documented regarding observed performance. Some of these pumps have been in the IST Alert range since October 1993, for degraded flow, differential pressure, or vibration. Further, the use of discrete performance data for plant thermal performance did not facilitate a trend assessment. Actions have been initiated to improve the quality of the report.

2.5 Cross-Connection of the Service Water and Fire Systems Backcround An inoperability of both fire water system pumps occurred on February 9,1994. The equipment problems that resulted in this event were reviewed in Inspection Report 94-01. In response to this event, plant operators cross-connected the station service water (SW) system with the vital fire suppression water supply (FP) system. On March 30, VY determined that this action placed the SW system in a configuration that was outside of the plant's design basis and made the required 1-hour Emergency Notification System report. This event was reported as Licensee Event Report (LER) 94-05 on April 29.

As described in Section 10.6.1 of the Final Safety Analysis Report (FSAR), cross ties are provided between the FP and SW systems. The FSAR states that the 12-inch cross tie valve (SW-8) is norme"y closed. Regarding the design aspects of the subject systems, the FSAR states that those portions of the SW system supplying safeguard equipment and those portions of the system which are not capable of being isolated from the SW supply are of class I seismic design. Vermont Yankee has traditionally viewed this part of their design basis to include the

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assumption that significant amounts of time were available to isolate small, non-safety class, non-seismic service water lines. The FP system is not seismically designed except for small portions which are considered to be part of the structural boundary for the SW system.

Procedure OP 2181, Rev. 29, Service Water / Alternate Cooling Operating Procedure, provides in Section K.1 instructions for cross-connecting the fire and service water systems. Specifically, it is stated that if the fire system does not have adequate flow open the 12-inch cross tie valve SW-8, run all available SW pumps, check SW and fire system pressure, and check component temperatures to ensure that there is adequate flow for components in service. In addition, VY has opened the SW-8 cross tie as part of procedures that implement fire system surveillances.

Event Details At 9:05 p.m. on February 9 the second fire water pump became inoperable causing an l

inoperability of the plant's FP system. Within ten minutes the shift supervisor (SS) notified the

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Operations Manager (OM) of the plant condition. Together, they concluded that TS 3.13.B.3.a,

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which requires a backup fire suppression water system to be established within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, could be established by opening the SW-8 valve. However, they elected to hold off implementing the

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opening of the valve utilizing the 24-hour action statement until additional reviews could be conducted on the following day shift. The SS indicated that he did not open the SW-8 valve at this time because it was recognized that the opening of this valve could potentially lead to a non-seismic SW branch line being unisolatable during a safe shutdown earthquake. The SW-8 valve

was opened by the next shift at 00:25 a.m. on February 10 and then closed at 3:58 p.m. later in the day following the return to service of one of the fire water pumps. The SS that directed that the valve be opened was concerned that the FP system was inoperable and a number of automatic fire suppression systems were relying on the immediate availability of fire suppression

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water to perform their fire protection funcHon.

At the Daily PM's Meeting on February 10, the Engineering Department (ED) was directed to

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review the acceptability of having the cross tie valve open. Engineering personnel from YNSD involved with the ongoing SW self-assessment were at the plant that day and were asked to review the condition. Their review identified a concern that for the design basis loss of coolant

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accident, ancurrent with a seismic event, the non-seismic fire suppression water piping (outside the intake structure) would be assumed to fail and that there was inadequate time for operators

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to close the SW-8 valve. With a single failure of one of the emergency diesel generators assumed to also occur, SW flow to the remaining diesel would have been inadequate and would

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potentially result in its shutdown. The YNSD engineers proceeded to review alternative actions

to address what they considered to be an unacceptable condition however, it appears that the

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immediate concerns were ameliorated by the return to service of one of the fire water pumps.

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The identified concern was communicated to the plant staff and resulted in the Operations Department issuing Night Orders on February 11 that specified that the Technical Programs

Supervisor was to be contacted to evaluate the method of establishing a backup system if the FP i

system becomes inoperable.

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Potential Reportable Occurrence (PRO) Report No. 94-25 was issued on February 9 to document and determine the reportability of the event involving both fire pumps being out of service. The PRO review conducted by the Technical Superintendent on February 15 documented: (1) that if the engineering determination is that the SW-8 valve cannot be opened, then the issue of

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reportability will have to be re-visited; and (2) that the OD is to ensure that the SW-8 valve is not to be opened until the engineering assessment is completed. Switching and Tagging Order No. 94-0228 was issued to maintain the valve shut until required evaluations are completed. On

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March 25, the YNSD engineers documented their conclusions on this issue. This document was reviewed by the ED and resulted in the issuance of a PRO on March 30, which identified that the plant's SW system had operated in a manner that was outside of its design basis by the opening of the SW-8 valve. As documented in LER 95-05, SW System Configuration Outside of Design Basis, evaluations are ongoing 'o identify alternative backup fire suppression water supplies that will not impact SW system operability. These evaluations are expected to l'e completed by August 1,1994.

Inspection Findines Regarding the role of the Plant Operations Review Committee (PORC), the inspector noted that the first time that the Committee reviewed the events of February 9-10 was on February 16 at

Meeting 94-13. This was the first meeting of the PORC since the event in question but, only discussed the issues associated with the failure of the electric fire pump to start. The inspector determined that the PORC had not reviewed the issue of cross-connection of the systems until April 29, where at Meeting No. 94-34 the Committee discussed the subject event as part of its review of LER 94-05. The root cause of the event submitted for PORC's consideration was i

inadequate procedure. The PORC reconunended approval of the LER contingent upon the root cause being reevaluated to determine if this was the proper root cause. The LER submitted to the NRC by VY designated the root cause of the event as inadequate design of the cioss-

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connection portion of the SW system, and that inadequate procedures were generated due to the inadequate design. In the review of LER 95-05, the PORC demonstrated a weak depth of review surrounding the circumstances of this LER.

According to licensee representatives, allowing the use of the SW-8 valve to cross-connect the systems has existed since initial operations. It appears that reviews conducted as a result of the required periodic bi-annual reviews of procedure OP 2181 (or for that mattei some of the surveillance procedures that specify the opening of the valve) did not questioned the issue of the

cross-connection, its relationship to either the TSs (i.e., as an envisioned backup supply for the FP system) or the FSAR provision that the SW-8 valve is normally closed.

As specified in Section 13.8.2 of the FSAR, the PORC reviews operating procedures and originates recommendations for any changes needed to optimize operation, improve safety, and assure compliance. They are also described as promptly investigating any abncrmal conditions which might have safety implications. Technical Specifications 6.2.A.6.e specifies that it is the responsibility of the PORC to review plant operations to detect any potential safety hazards.

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The failure of the PORC to conduct a review of plant operations to detect potential safety hazards associated with the need for, and the results of, the implementation of a cross-connection of the SW to FP systems is an example of a violation of the requirements of TS 6.2.A.6.e (VIO 94-13-01)

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The inspector also noted no concerns with the response of the plant operators in addressing the off-normal equipment conditions that existed on February 9-10. Each shift demonstrated a good

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level of safety concern and proper judgement, especially when challenged by a situation that involved the unavailability of detailed procedural guidance, in attempting to minimize the impact that inoperable FP equipment had on the operating plant.

3.0 51AINTENANCE (62703, 61726, 93702)

3.1 51aintenance The inspectors observed selected maintenance on safety-related equipment to determine whether these activities were effectively conducted in accordance with VY TS, and administrative i

controls (Procedure AP-0021 and AP-4000) using approved procedures, safe tagout practices and appropriate industry codes and standards.

Interviews were conducted with the cognizant engineers and maintenance personnel and vendor equipment manuals were reviewed.

3.1.1 IIydrogen\\ Oxygen Monitor Failure During this inspection period, VY experienced spurious high pressure alarms in the "B" hydrogen \\ oxygen (1I202) monitoring system and declared the system inoperable to support troubleshooting and corrective maintenance. The 11202 system is used to sample the drywell

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atmosphere and analyzes hydrogen and oxygen concentrations for abnormal increases, potentially

indicative of post-accident fuel failure or combustible gas conditions. Two 11202 systems are normally available and a 30-day TS LCO is applied when one system is out of service.

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The inspectors verified that the OPG and the Instrument and Controls (I&C) Department assigned the appropriate high priority for the restoration of this TS equipment. Coordination of maintenance was discussed at the Daily PM's Meetit r The system was restored to operation following corrective maintenance and post maintenane: testing. The maintenance was well I

controlled, the tagout assured worker safety, and no inspector concerns were identified.

3.1.2 Atmospheric Monitoring Piping Configuration In NRC Inspection Report 94-09, the inspector documented that rubber hoses were installed in the process lines for the drywell and reactor building ventilation radiation monitoring systems.

This observation was discussed with ED and Maintenance Department representatives who acknowledged the inspector's concern that hoses were not specified on the controlled drawings

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or vendor manual, and that the use of rubber in lieu of piping could represent a reduction in long-tenn system integrity. Failure of a hose could result in a dilution of sampled air resulting in a non-conservative radiation indication.

Currently, VY has detennined that the functional integrity of these systems is adequate. The hoses are structurally sound and show no evidence of cracking. Although hoses are installed on the sampling pump discharges which are subject to elevated temperatures, the cognizant ED engineer has determined that there is reasonable assurance that the hoses will not thermally degrade and cause a system failure. A commitment item was initiated by the Maintenance Department to resolve and track this issue to closure. Preliminary corrective actions included drawing changes and an assessment regarding the implementation of preventive maintenance to assure the long-term ;ategrity of the hoses.

Based upon the above information, the inspector determined that there are no immediate system operability or reliability concerns. The installation of the hoses, which appears to have existed for a long period of time, was performed without maintenance or engineering documentation.

Adequate administrative controls are currently in place, as required by procedure AP 0021, Work Orders, to preclude recurrence.

3.1.3 Residual Heat Removal Valve Maintenance During the period of June 9-15, corrective maintenance occuned on components of the "B" subsystem of the residual heat removal (RHR) system. On June 9, the minimum flow valve, RHR-16B, lost control power while the valve was being closed from the control room.

Troubleshooting and repairs were controlled by Work Order (WO) No. 94-05242. A blown fuse was found in the control circuit. Maintenance personnel identified a loose wire in the valve's control circuit, although no conclusive cause for the blown fuse was identified. There was an intensive investigation due to this condition having also occurred on November 6,1993, as documented in WO No. 93-09789. The valve was returned to service on June 9.

On June 10, following the securing of the "B" RHR subsystem from the suppression pool (SP)

cooling mode of operation, operators determined that the subsystem would not pressurize and that the level in the SP was increasing. Backflow from the keep fill system was causing the level increase. The "B" RHR pump was isolated from the pool and declared inoperable. The pump's discharge check valve, RHR-48B, was determined to be leaking. The Maintenance Department issued WO No. 94-05361 to investigate, repair or replace components, and conduct post maintenance testing. The investigation determined that the valve disc was not properly seating on the valve seat at the top of the disc by approximately 0.090 inches. This was caused by a combination of: (1) the valve body casting was not symmetrical, (2) minor corrosion buildup was noted on the disc stud and arm stud bore, and (3) there was little to no clearance between the arm stud bore and the disc stud outside diameter. Vennont Yankee contacted the manufacturer of the valve and confirmed that the as-found dimensions and clearances were within manufacturing tolerances and expected normal wear.

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Repairs included increasing the size of the hanger to disc attachment hole to allow the disc to float relative to the swing hanger, since this free float is necessary to allow the disc to properly seat; and the insertion of a shim in the hanger suppon spacing to prevent binding of the disc on the valve body side wall. The ED reviewed the repair techniques, determined that they were similar to those the manufacturer would perform, and that the work was properly performed in accordance with plant procedures. Regarding previous maintenance on this valve, the inspector noted that VY performed an inspection in August 1990 and the disc, disc washer, disc nut, cotter pin and hanger arm were replaced. A Significant Corrective Action Report (CAR) was assigned to maintenance to determine the root cause and ensure that appropriate corrective actions will be accomplished.

Although repairs to the RHR-48B valve were completed on June 13, the post maintenance testing could not be performed due to a problem with lining up the subsystem to the SP. Specifically, the "B" RHR pump's SP suction valve, RIIR-13B, would not open properly. The Maintenance Department issued WO No. 94-06537 and determined that the problem was due to a defective set of contacts on the valve's control room located key-lock control switch. On June 14 the RHR-13B valve was repaired and returned to service, the post maintenance testing was successfully accomplished for the repairs to the RHR-48B valve and the "B" RHR subsystem was declared operable.

The inspector observed that the above enumerated repairs received a good level of management interest at the Daily OPG meeting and PM's meeting. Plant and department managers were involved in the reviews of the conduct of investigations and repairs. A good level of ED involvement occurred. An appropriate level of documentation was found in completed WO packages.

The inspector also noted that maintenance personnel were knowledgeable and accomplished work in accordance with the WO instmetions and applicable maintenance procedures. Proper housekeeping controls, for an open system were implemented.

3.1.4 Fire in the Reactor Water Clean-up Panel On May 30 a fire in the demineralizer control panel for the reactor water clean-up (RWCU)

system occurred. This panel is located on the 318 foot elevation of the reactor building (RB).

Seven relays (General Electric Co. type CR120A) and associated wiring were damaged by the fire. The fire, which was self extinguished by a lack of oxygen, caused an electrical short that resulted in the opening of the "A" precoat inlet valve (SP-12-4-40A) and caused the pressurization of a section of RWCU 150 psig rated piping to 1000 psig. This resulted in lifting the 150 psig relief valve (SR-12-4-36) on the precoat inlet line for the "A" RWCU demineralizer and the discharge of approximately 300 gal. of reactor coolant to the radwaste system via the RB floor and equipment drain system. There was a minor spill in the vicinity of the fuel pool cooling pumps at the 303 foot elevation of the RB due to a backup in the floor drain system.

The actuated relief valve caused increased flow in the RWCU system and an approximate 2-inch reactor water level decrease from the steady state water level of 160 inches. The decrease in reactor water level was automatically mitigated by an increase in feedwater flow causing a subsequent short-duration reactor power transient to 1614 MWt from an initial power level of l

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1594 MWt (100 percent rated power). When the RWCU system isolated on high inlet water temperature, the release pathway was isolated and the reactor water level transient was terminated.

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Based upon radiation monitoring system alarms, the shift supervisor initiated conservative actions

to secure access to the RB pending the conduct of radiation and contamination surveys.

Contamination of the RB occurred, however, affected areas were decontaminated in a timely manner. An evaluation and a walkdown of low pressure system piping was performed by ED l

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personnel. They determined that the short term over-pressure condition should not have caused

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any short or long term damage. Repair efforts conducted by the I&C Department, with assistance from the ED and two representatives from the manufacturer of the RWCU system, were successful in returning the system to service on June 5. During the time that the RWCU system was out of service, the Chemistry Department was monitoring reactor coolant quality closely and reporting its status to plant management daily. No TS or VY administrative limits

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for reactor coolant conductivity were exceeded as a result of this event.

The I&C Department was directed to develop a Significant CAR. A VY corporate engineer was

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assigned to lead the development of the CAR. As developed, the CAR attributed the root cause I

of the event to be ineffective corrective action. This was because the 1991 establishment of a 15-year service life for normally energized CR120A relays to address reliability concerns (i.e.,

fires or relay failures caused by coil overheating) was not applied to non-nuclear safety (NNS)

relay applications, as is the case for the RWCU system controls.

Reliability problems with

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older CR120A relays that used a 115 volt coil design was a known issue, in that, an operating

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experience (OE) report from another facility was received and dispositioned by VY in 1989.

Problems with this relay were also documented in GE Service Information letter (SIL) No. 229 in 1977. This SIL was followed by NRC Bulletin 78-01, Flammable Contact-Arm Retainers in G.E. CR120A Relays. Here too, the corrective action was limited to safety related system

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applications and VY's current investigation concluded that the contact arm retainers were not replaced in the RWCU system. This condition added to the severity of the fire event. Vermont Yankee had experienced a failure with a safety related CR120A relay in the primary containment

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isolation system, with corrective actions dispositioned by LER 91-10. Other failures of the

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subject relay had also occurred but, in NNS applications. The inspector noted that upgraded l

relays were used in replacing fire damaged units, and that thermography was used to detect and replace if warranted in safety and NNS systems relays that indicated abnormally high coil

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temperatures.

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As a result of the re-review of OE, including the lessons learned from this event, VY developed the following corrective actions associated with the use of CR120A relays: (1) all CR120A relays that are normally energized, or have a high duty cycle of energization, will be identified and have their relay coils replaced with the upgraded 120 volt coil if that has not previously occurred; (2) the NNS normally energized CR120A relays will have a service life applied to them and be included in a periodic replacement program; and (3) the priority of te RWCU demineralizer control panel design change (Plant Design Change Request 91-06) be reviewed.

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I The latter item is related to an August 1989 event involving the backwash and precoat of the

"A" RWCU system's precoat tank. A sticking switch contact in the control panel resulted in a 12 foot high column of water being discharged from the tank and the contamination of a large l

area in the RB. As a result of this event, it was recommended by the plant organization that the old style electromechanical timing system used in the demineralizer's control panel be replaced I

with a new programmable controller. This proposed modification has been budgeted for each l

year since 1989 and resulted in the development of PDCR 91-06 in 1991. Implementation cf this l

modification would have resulted in the elimination of all CR120A relays used in this system.

l According to the CAR, it was stated that it was unfortunate that this project has had a lower l

priority and for a variety of reasons has been canceled each year.

l This event and the CAR were reviewed at PORC Meetings 94-44 and 94-47 on June 3 and 7 respectively. A number of additional corrective actions were recommended by the Committee, including: (1) review past G.E. SILs and NRC Information Notices for issues similar to those identified in this CAR (i.e., aging, partial implementation); (2) the implementation of thermography to identify degraded equipment in other areas; and (3) ensure that funding is approved for PDCR 91-06. This latter item was assigned by the Plant Manager (PM) to be reviewed by the Integrated Planning Committee (IPC). The inspector learned that the IPC is a new group of senior managers formed in May 1994 for the purpose of ensuring that VY budget goals are met while also ensuring that projects are properly prioritized so as to ensure safe and steady plant operation. This Committee; which consists of the Operations and Technical Superintendents, the Engineering Director, the Controller, the Assistant to the PM, and the Corporate Engineering Support Department Manager, will oversee the allotment and re-allotment of funds to meet these goals.

The performance of the I&C Department in restoring the damaged system to an operable status reDected a high level of technical competence of its personnel and management. A cautious approach to system restoration was observed.

All recovery efforts were preplanned.

Coordination and communications between responsible individuals and departments were outstanding, as evidenced by the uneventful return to operations of the system. A very good level of management involvement and oversight was observed. Special attention was evident in the manner in which the OPG Coordinator established test controls to assure that further overpressurization of low pressure components of the RWCU system would not occur during post maintenance testing. A thorough and probing review, as evidenced by the additional recommendations made to the CAR's recommendations, was provided by the PORC. A good OE focus was evident. The IPC and process was considered to be a positive 3.2 Surveillance The inspector reviewed procedures, witnessed testing in-progress, and reviewed completed surveillance record packages. The surveillances which follow were reviewed and were found effective with respect to meeting the safety objectives of the surveillance program. The

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inspector observed that all tests were performed by qualified and knowledgeable personnel, and in accordance with VY TS, and administrative controls (Procedure AP-4000), using TS approved procedures.

• OP 4126 Rev. 33 Diesel Generator Surveillance

OP 4111, Rev. 26 Control Rod Drive System Surveillance

OP 4113, Rev.18 Main and Auxiliary Steam System Surveillance

OP 4179, Rev.1 Standby Fuel Pool Cooling Surveillance

OP 4110, Rev. 21 Reactor Recirculation System Surveillance

OP 4152, Rev. 23 Equipment and Floor Drain Sump and Totalizer Surveillance

OP 4115. Rev. 32 Primary Containment Surveillance 4.0 ENGINEERING (71707, 37551, 93702)

4.1 Events Assessment Feedback - Slow Scram Times The occurrence of slow scrams times recently experienced at the Washington Public Power Supply System (WPPSS), Unit 2, and the Boston Edison Company's Pilgrim Nuclear Power Station, were evaluated by VY to determine whether similar conditions had the potential to occur at their facility. Both stations identified hardening and cracking of the buna-n diaphragms that are integral to the scram solenoid pilot valves (SSPVs). During their investigations, these licensees found no definitive correlation between individual control rod scram times and the functional integrity of its respective buna-n diaphragm. They also found that the diaphragms were in-use only three to four years which indicated that the failure mechanism was age and/or material related. Neither station found any indication of thread scalant contamination, high external temperatures, or unreasonably high solenoid coil voltages. At both stations, all SSPV diaphragms were replaced regardless of the associated control rod scram time. Root cause determinations continue.

The GE Co. issued RICSIL No. 69, Revision 1, " Scram Solenoid Pilot Valve Diaphragm Degradation," describing the above events, results of their studies, and interim recommendations to assure proper operation of SSPVs. In addition, GE stated that the conditions observed at WPPSS and Pilgrim differed from the conditions reported by SIL 575, "CRD Slow Start of Motion." In SIL 575, anomalies in scram times were found to be a combination of scram air header pressure, scram time measurement technique, age / temperature degradation, and diaphragm hole diameter. This SIL was issued, in part, due to the lessons learned from the VY slow scram insertion time issue that occurred in April 199?. (NRC Inspection Report 93-09).

Vermont Yankee evaluated the conditions observed at the WPPSS and rus im stations and c

concluded that they had reasonable assurance that similar conditions were not occurring at their plant. This justification was verified by the inspector based on an evaluation of recent scram time data and the replacement of all SSPV diapnragms in April 1993. Results from technical studies involving material characteristics, inspection, and data evaluation will be incorporated into the VY scram timing program. Daily management briefings were conducted to discuss the

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status of VY's evaluation of diaphragm functional integrity. Frequent communications with the BWR Owners Group representative and YNSD provided good technical expertise for VY's evaluation of these events. Recommendations developed by GE are planned for implementation during the March 1995 refueling outage. The inspector considered the R&CE Department's

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efforts to closely follow these current OE issues to ne excellent.

4.2 Leakage of "A" Core Spray Subsystem Valves

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Component Description and Backcround The functions of the three core spray (CS) pump discharge valves are described in Sections 5.2.3.5 and 6.4.3 of the FSAR. These valves are shown in Attachment 1. The check valve, CS-13A, prevents loss of reactor coolant outside containment. The motor-operated, flexible

wedge, 8 inch gate valves (CS-12A and 11 A) isolate the system from reactor coolant system (RCS) pressure and prevent over-pressurization of the 450 psig low pressure CS piping; these valves are leak rate and stroke time tested, however, all three valves are not TS primary containment isolation valves. The CS-12A valve is normally closed and the CS-IIA valve is

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normally open during reactor power operations to minimize the equipment needed to operate

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during an accident.

I During the Operational Safety Team Inspection Follow-up Inspection No. 94-08, the inspectors observed that the pressure within the low pressure piping of the "A" CS subsystem had increased higher than expected. In normal standby readiness, this section of CS piping (i.e. keepfill portion) remains at constant pressure at approximately 80 psig as controlled by a keepfill system.

Ilowever, since plant startup on October 25, 1993, the low pressure piping of the "A" CS subsystem continuously pressurized at a rate of appror.imately 10-15 psig/ hour. Operators were bleeding off system pressure 3-7 times a week, which maintained a margin between the 300 psig

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high pressure alarm setpoint of pressure switch PS-14-47A and as-found piping pressure

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(upwards of 225 psig).

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A Basis for Maintaining Operability (BMO) No. 94-05 dated April 6, was developed to document VY's just fication that the plant may continue to operate safely with this degraded i

condition, as long as the leak rates remained low and no adverse trend was occurring. The pressure increase was attributed to RCS leakage past CS-13A and 12A valves. This leakage was considered to be caused by degradation of the CS-12A seating surface. Inservice testing results and leak rate testing data indicated acceptable CS-12A valve performance and did not reveal any degradation or material concerns. The maintenance history for the CS isolation valves indicated that valves CS-13A and II A were previously inspected, found acceptable, and restored without valve lapping or machining: CS-12A was never inspected. Additionally, the BMO specified that if the 300 psig, "A" CS valve leakage high, annunciator alarmed, then Alarm Response Sheet (ARS) 3-D-4 was expected to be followed. Actions required by ARS 3-D-4 included cycling C-12A and closing CS-11 A if pressure continues to increase. The ARS references PORC Meeting 90-08 and a Safety Evaluation (SE, October 10, 1990) written for an RCS

leakage situation that occurred that resulted in plant operation with both "B" CS subsystem

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valves CS-12B and 11B closed. The NRC's review of this issue is contained in Inspection Reports 90-08 and 90-80, and resulted in the identification of violations of 10 CFR 50.59 requirements.

On April 6, the PORC at Meeting No. 94-29 reviewed BMO 94-05, and assigned actions involving monitoring and trending, operator instructions, review of similar industry events, and calibration oflow pressure alarm instrumentation (PS-14-47A&B). Since the BMO required that plant management be informed if the pressurization rate exceeded 220 psig in an 8-hour period, revised operator log sheets were implemented requiring notification of the Operations Manager (OM) should this pressurization rate occur or ifit became necessary to bleed pressure more than once per shift to maintain pressure below 220 psig. Corrective actions to perform internal inspections of valves CS-11 A and 12A and to review the adequacy of the CS valve preventive maintenance prior to the March 1995 refueling outage were assigned. NRC Generic Letter 87-06, " Leak Tight Integrity of Pressure Isolation Valves," was reviewed and PORC documented a concern regarding the potential for pressure locking of the valve bonnet assembly for CS-12A and 11 A gate valves however, no action item was assigned to address this concern. The PORC concluded that continued plant operation was safe, that the 'A" CS subsystem was operable, and that the leakage (0.03 gph calculated) and rate of pressurization (80-100 psig/8-hour shift) was acceptable.

The inspector's review of BMO 94-05 determined that the VY operability assessment was acceptable (UNR Item 94-08-01, Closed).

Pressurization Rate Increase On May 8, the pressurization rate in the CS system increased (180-200 psig/8-hour shift). On May 12, at 4:45 p.m. and 9:40 p.m., it became necessary to bleed off the pressure, and as required the OM was informed. Subsequently, the inspector discussed what appeared to be..n escalating condition with the OM, and assessed potential VY response actions, including the actions of the ARS. The inspector questioned the acceptability of the pressure switch PS-14-47A to be classified as Safety Class 0, given that full implementation of the ARS could result in closure of the CS-11 A. Potentially, this component could be subjected to full RCS pressure.

The OM indicated that he could not address the quality and pressure retaining capabilities of the component but would investigate the matter.

During the Daily PM's Meeting on May 13, the higher pressurization rate was discussed. It was concluded that the actions in BMO 94-05 were still applicable, however, the Engineering Depanment was tasked to re-evaluate the BMO in light of the current pressurization rate. The OM raised the concerns about the pressure switch, which resulted in assigning the isste to the Engineering Department for review and disposition.

On May 17 at 1:30 a.m., the "A" CS valve leakage high pressure alarm actuated and the local pressure gauge at the pump discharge indicated 350 psig. The OM was notified and the pressere was i rimediately bled off to approximately 80 psig. At 4:40 a.m., the pressure increased to 265 psig a id operators cycled the CS-12A valve in accordance with ARS 3-D-4. At 5:45 a.m., the OM was again notified that the "A" CS subsystem pressurization rate was approximately 80

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psig/ hour, indicating that the cycling of the CS-12A valve was not effective in reducing the leakage. At this time, the OM and SS reviewed the ARS and discussed closing the CS-11A valve. It was decided to hold off closing the valve until this condition cou!d be reviewed at the PM's Meeting. The "A" CS subsystem was still considered operable by the Operations Department. Subsequent analysis indicted that the valve seat leakage increased approximately 20 percent to 0.036 gph.

During the Daily PM's meeting on May 17, the PM tasked the ED and OD to review the BMO, the October 1990 SE, and the ARS and make recommendations regarding closure of the CS-11 A valve. By 3:00 p.m., this review was completed and continued plant operation with both the CS-11 A and CS-12A valves closed was considered acceptable.

Engineering Department recommendations regarding future inservice inspection requirements, a temporary modification for monitoring and annunciation of the low pressure piping, and closing of the PS-14-47A pressure Switch's root isolation valves were transmitted to the OD for implementation. At 3:15 p.m. on May 17, the CS-11 A valve was closed.

"A" Core Sorav System Inonerability Subsequently on May 19, the ED determined that with both CS-12A and 11A valves closed inadequate margin existed to assure that both valves would open during a design basis accident due to differential pressure (dp) trapped between the valves. It was postulated that dp could occur because CS-12A leaked, pressurizing the space between CS-12A and the now closed CS-11 A. Therefore, without assurance that there would not be trapped pressure between the valves during an accident condi: ion, CS-12A would have been required to open with an estimated 650 -

1000 psid (1000 psig RCS pressure minus 0-350 psig due to the opening of the valve at the low pressure permissive setpoint, or less if complete vessel blowdown occurs) and CS-11 A with 915 psid (1000 psig RCS pressure minus 85 psig keepfill pressure) across their disks. The dp for these valves were approximately 2 to 3 times higher than the 280 psid at their thrust settings were calculated. Because VY Engineering determined that inadequate margin existed between the required thrust and available thrust of 29.800 psi and 30,000 pounds respectively, they l

concluded that an " indeterminate" state of operability existed and therefore conservatively l

recommended that the valves be considered inoperable.

On May 19 at 1:30 p.m., VY declared the "A" CS subsystem inoperable. The 7-day TS 3.5 LCO Action Statement was back dated to 3:15 p.m. on May 17, the date that the CS-11 A valve was closed. Supplement I to BMO 94-05 and a replacement 10 CFR 50.59 SE were reviewed at PORC Meeting No. 94-40 on May 20 which justified the valve lineup deviation and change to system operation that would result from opening CS-12A valve and leaving CS-11A valve closed. Prior to implementation additional short-term corrective actions involving pressure monitoring, changes to operating and ARS procedures, and pressure alarm / indication were instituted. On May 21, the CS-12A valve was placed in the open position and the "A" CS l

subsystem was considered operabl _ _ _ _ _ _

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Current Status and Safety Assessment

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Based on current leakage monitoring, VY has determined that adequate CS-11A valve seat integrity exists to prevent pressurization of the low pressure CS piping. Plant management continues to periodically review the status of leakage at the daily PM's Meetings. Temperature monitoring of the piping will be considered to independently confirm that RCS leakage into the low pressure piping is not occurring.

Inservice inspection will be performed to provide assurance that pipe welds are structurally sound to contain RCS pressure.

The RCS leakage through the CS-12A valve was oflow safety significance. The leakage rate was small and monitored by operating personnel and installed instrumentation.

Over-pressurization of the "A" CS subsystem was continuously protected by relief valve SR-14-20A and high pressure conditions were alarmed in the control room. Based on in-service test results and local leak rate testing, neither CS-12A nor CS-II A demonstrated potential for catastrophic failure. In fact, even if both motor-operated gate valves were open, in-situ testing confirmed that the seat integrity of CS-13A would have limited RCS leakage to approximately 8 gph, well within the 100 ppm relief capacity of relief valve SR-14-20A. This condition is highly unlikely because, with RCS pressure greater than 350 psig, valve interlock circuitry prevents both CS-12A and CS-II A valves from being simultaneously opened. At this leak rate, CS-13A met its design function as described in the FSAR. Conservatism in the VY thrust assessment was also

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evident, because the calculations were based on the maximum 1000 psid which assumed immediate RCS depressurization and relatively conservative stem friction and valve factors.

Inspection Findines The technical evaluation performed for CS-12A valve leakage was accurate and generally comprehensive.

Jurtifications and conclusions were supported by reasonable engineering judgements, appropriate regulatory guidance, and information gained from inservice and leak rate testing. The multiple design functions and actual performance associated with the CS-12A and CS-II A valves were reviewed and discussed in sufficient detail to provide confidence that operation with the degraded CS-12A valve seat was acceptable with respect to both design and regulatory requirements. Weaknesses in technical evaluation were observed regarding: (1) the initial comparison of CS-12A leak at RCS pressure to that observed at 44 psid during leak rate testing; and (2) crediting of availability of the relief valve SR-14-20A to mitigate the CS-12A valve leakage without being aware of its capacity.

However, these technical evaluations weaknesses did not significantly detract from the appropriateness of the engineering conclusions.

The inspector determined that VY's corrective action to review the adequacy of CS valve preventive maintenance was appropriate.

Although acceptance criteria were established to define the CS-12A valve leakage limits for which BMO 94-05 was valid, once these limits were reached, it was not clear as to what the follow-up actions would be. This was demonstrated on May 13 following the depressurization of the system twice in an 8-hour period on May 12. When this occurred, subsequent actions were unclear and the engineering staff was tasked to re-evaluate the trend because they were ill-e

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prepared to quickly articulate the significance of this occurrence. A similar situation occurred on hiay 17 when CS-12A leakage increased 20 percent. When this was subsequently discussed during the Daily PM's Meeting, it was unclear whether the BMO was still valid. As a result, the ED was tasked a second time to evaluate the significance of exceeding a pre-established acceptance criterion. This latter occurrence, contributed to an 8-hour delay in closing the CS-11 A valve as required by ARS 3-D-4.

The safety assessment performed by the PORC during the reviews of the original BMO 94-05, the ARS 3-D-4, and the October 10, 1990 Safety Evaluation failed to review all operational safety considerations involved with closing the CS-11 A valve. Specifically, the following three issues were not adequately evaluated prior to the increase of CS-12A valve leakage on May 17.

(1)

Pressure Ratine and Ouality Reauirements of Pressure Switch PS-14-47A: This pressure switch was classified as Safety Class-0 (SC-0) and is a RCS pressure retaining component. In some cases, SC-0 is considered comparable to a component that is classified as NNS. Typically, NNS components have no 10 CFR Part 50, Appendix B quality assurance requirements specified. Following implementation of BMO 94-05, and after concerns were raised by the inspector, VY initially determined that the component was rated for 800 psig and not qualified for the 1000 psig RCS pressure. Subsequently, VY determined that a " gauge saver" feature existed to protect the component from RCS pressure failure. This feature was not described in the VY Maintenance Planning end Controls system, the applicable SE, or the BMO. Similar situations exist in other safety-related systems where SC-0 components retain RCS pressure.

Vermont Yankee evaluation of the level of quality assurance applied to RCS pressure boundary I

components classified as SC-0 remains an unresolved item (URI 94-05-01).

(2)

less of Alarm Function From PS-14-47A: As described in FSAR Section 7.4, this pressure switch provides monitoring of the CS system pressure between the two pump discharge valves to permit detection of RCS leakage into the CS system outside containment. As documented in their response to NRC Generic Letter 87-06, " Periodic Verification of Leak Tight Integrity of Pressure Isolation Valves," VY stated that the design function this pressure switch was credited to provide over-pressure annunciation for control room operators. Although the Generic Ixtter 87-06 issue was reviewed by PORC at its April 6,1994 meeting, the loss of alarm function was not fully evaluated by PORC or the VY staff until after CS-12A leakage had increased on May 17 and the

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issue was raised by the inspector. Neither the SE nor the BMO addressed this issue. The alarm function contributes to safe plant operation because it alerts plant operators to a potential over-pressurization condition of low pressure CS piping.

(3)

Differential Pressure Onenine Capabilitief the CS-12A and 11 A Valves: As discussed earlier, it was not until May 20 that PORC addressed this issue curing its deliberations of RCS leakage into the CS subsystems. Ilowever, Special Test Procedure 93-009, in-Situ Differential Pressure Testing of Valves CS-11 A and CS-12A, was PORC reviewed (September 24,1993, meeting 93-78) and completed October 14,1993. It is difficult to

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understand why, as a minimum, EMO %05 failed to address this issue or be identified by PORC as an area in need of assessment.

Notwithstanding the above, as an operational consideration, differential opening capability of valves should be a routine consideration and therefore addressed in terms of the current knowledge of the performance capabilities of the equipment.

As a result of the inability of PORC to self-identify for review and disposition the above enumerated issues, the inspector determined that this condition constituted a failure of the PORC to review plant operations to detect potential safety hazards as required by TS 6.2.A.6 e. This failure to meet the requirements of TS 6.2.A.6.e is considered an example of a violation (VIO 94-13-02).

4.3 Core Spray Pressure Alarm Temporary Modification

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Vennont Yankee modified the alarm function for pressure switch PS-14-47A to allow its

continued use after the CS-11A valve was shut. The inspectors discussed the loss of alarm function with members of the OD and ED on May 17, prior to the closure of CS-11 A. At this time, VY had not fully evaluated the Icas of the alarm function, however, VY representatives acknowledged the inspector's comments and concerns regarding the lack of a SE describing the

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loss of this FSAR described component.

l Temporary Modification (TM) No. 94-13 was installed prior to the closure of valve CS-11 A.

Using the existing wiring and control room annunciator, VY installed a temporary pressure switch near the "A" CS pump discharge and electrically connected it to the as-built PS-14-47A alarm circuitry. The root isolations for PS-14-47A were shut to prevent spurious actuation,

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significant changes to operator logs were not required, and the annunciation of a high pressure condition remained unchanged.

The inspector observed the implementation of the initial phases of the TM. The cognizant engineer accurately described the TM to the SS and referenced applicable system drawings. In parallel with this, the SS tasked the Shift Engineer to independently assess the TM from an operational standpoint. As a result, consequences resulting from a change in the alarm setpoint were identified. Because the alarm setpoint was lower than the discharge pressure of the "A" CS pump, the Shift Engineer found that the alarm function would be lost during quarterly pump surveillance and nuisance alarms could be generated. This excellent observation was brought to the attention of the PORC and revisions to applicable procedures were implemented. The normal alarm setpoint of 300 psig was lowered to 180 psig by PORC to conservatively inaease the margin between an alarm condition and the lifting of relief valve, SR-14-20A, at 375 psig.

The inspector reviewed the TM and identined no concerns. The applicable control room drawings were updated to reflect the TM and the operating staff was informed of the ci anges l

to "A" CS subsystem operation.

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5.0 PLANT SUPPORT (71707, 71750, 93702)

5.1 Radiological Controls Inspectors routinely observed and reviewed radiological controls and practices during plant tours.

The inspectors observed that posting of contaminated, high airborne radiation, radiation and high radiation areas were in accordance with administrative controls (AP-0500 series procedures) and plant instructions. Iligh radiation doors were properly maintained and equipment and personnel were properly surveyed prior to exit from the radiation control area (RCA). Plant workers were observed to be cognizant of posting requirements and maintained good housekeeping.

Appropriate considerations to reduce personnel exposures were discussed and implemented prior to the conduct of Geld activities to assess the integrity of the "A" CS subsystem injection valves.

The RP staff demonstrated effective control of radioactive materials generated during the plant-wide cleanup activities (Section 5.4). Improvements in the quality of radiological survey maps

had resulted in improved worker information.

5.2 Emergency Preparedness 5.2.1 Emergency Communications Circuit Maintenance During this inspection period VY effectively managed offsite maintenance conducted to enhance emergency communications.

The offsite maintenance on the FTS2000 communications equipment was conducted by a vendor and affected the Emergency Response Facility located at the VY corporate office. Appropriate discussions occurred to inform plant management and control room operators of the short-duration communications disruption.

Alternate communication circuits were verided operational prior to the performance of the maintenance.

5.2.2 Emergency Communication Circuit Enhancements As identified in NRC Inspection Report 94-09,the plant communications system within the emergency diesel generator (EDG) rooms is difficult to hear during diesel operations. This observation was confirmed with a number of plant operators who routinely conduct EDG surveillances.

Vermont Yankee's Emergency Preparedness (EP) staff accepted responsibility for this issue because of their concern that other areas within the plant may also exhibit difficulty in emergency announcements and alarms being readily heard. A task force was chartered to review and disposition this issue. Corporate management is currently reviewing the recommendations developed to resolve concerns in this area.

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i 5.2.3 Pager System Malfun. tion During this inspection period, the emergency communications pager system malfunctioned preventing its use as the primary means :o notify VY emergency response personnel of emergency events, as described in the VY EP Plan.

'this off-site system is operated and maintained by a contractor.

The inspector observed that the pager system was quickly restored to operation.

Communications between the EP staff and the contractor resulted in the implementation of a redundant and independent paging system for use by system maintenance personnel, so as to notify them of system problems requiring repair. The loss of the paging system was of minor safety significance and VY response to a plant emergency was not precluded because a manual call-in system using telephone lines is proceduralized and would be implemented in parallel with

pager system actuation. The failure of the pager system was identified during VY weekly

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surveillance testing. A VY Quality Services Group audit of the contractor was performed to assess the adequacy of contractor controls to assure system reliability.

5.3 Security The inspector verified that security conditions met regulatory requirements and the VY Physical Security Plan. Physical security was inspected during regular and backshift hours to verify that controls were in accordance with the security plan and approved procedures.

5.3.1 Security Event Review

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On May 11, the Pennsylvania State Police (PSP) conducted a briefing at VY regarding the unauthorized intrusion at the Three Mile Island Nuclear Station (TMI) on Febmary 7,1993.

The PSP brief was attended by the inspector,34 individuals representing five nuclear power facilities, and six local law enforcement agencies. The purpose of the brief was to review the lessons learned from the event and to discuss coocdination activities. The brief was held at the request of the VY Security Department.

lessons learned from this event were reviewed in detail. Experiences and insights gained regarding the challenge of accurately disseminating information to the media were reviewed.

Command, control and communications between the PSP, TMI representatives and emergency j

management organizations were also reviewed. Lessons learned generated from difficulties encountered during response activities, staging of command posts, and activation of communication circuits were reviewed.

Pctsonnel performance and programmatic issues involving command clarity, handling of the media, and implementation of search tactics highlighted the discussions mentioned above, i

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F The planning and conduct of the brief was a positive initiative. Based on the materials reviewed and discussed, important lessons learned regarding security response strategy, command and control, policies and procedures were presented to the VY Security staff. The Security Manager

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indicated that the security and emergency response procedures will be revised to incorporate the lessons-learned.

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5.3.2 Diesel Room Exhaust Fan Breaker Found Open

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On June 2 at 5:55 a.m., a contrc,1 room operator identified that the indicator lights for the "B" EDG room exhaust fan (TEF-3) were extinguished indicating that the control power circuit j

breaker was open. An Auxiliary Operator (AO) verified that the breaker located in a motor

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control center (MCC) in the "B" EDG room was in the "open" position verses the " tripped open" position, and shut the breaker in accordance with instructions from the control room.

Power was restored to TEF-3 at 6:00 a.m. and the breaker properly operated without excessive

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friction or movement. The AO verified that no similar condition existed in the "A" EDG room.

The power supply breakers for the EDG room exhaust fans are normally shut enabling the fans to auto-start based on EDG room temperature and diesel operation. Once operating, the fans automatically cycle on and off to maintain EDG room ambient temperatures between 60 and 80 degrees. As described in the FSAR, loss of EDG room ventilation could result in a loss of the associated EDG during engine operation if room temperatures became excessive. The exhaust fans (one fan for each EDG room) are powered from their respective EDG supplied busses.

Vermont Yankee aggressively investigated this event and determined that the breaker was inadvertently opened during floor surfacing refurbishment activities conducted the previous day.

Reconstruction of the previous day activities included interviews with maintenance personnel,

evaluation of access control information, and a review of the maintenance conducted. It was I

determined that a cable draped above the MCC had inadvenently dropped and tripped open the breaker. Vermont Yankee demonstrated that the force required to open the breaker was small i

and that it was reasonable that the breaker opened without the maintenance personnel in the area

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being aware that it occurred.

Security and Maintenance Departments coordinated the investigation activities. Persons independent of the EDG room maintenance were assigned lead investigation responsibility.

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l The investigation activities in the EDG room were appropriately controlled so as to not adversely affect the operability of the "A" or "B" EDGs. The breaker was replaced as a conservative measure to allow bcnch testing; the breaker properly operated and no root cause was identified.

Based on the ambient air temperature in the EDG room and at the EDG mtake plenum, diesel

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operability was not affected by the loss of the fan. The inspector had no further questions on this event at this time.

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5.4 Housekeeping

The VY staff conducted plant cleanup days on May 11 and June 8 to remove unnecessary f

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materials and assure the proper storage of materials.

This effort was intended by plant

management to improve the overall material condition of the plant, advance the VY's " vision"

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of good housekeeping and provide another way for all the employees (both plant and corporate)

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to demonstrate the pride that they have in the plant. Plant wide activities were focused in the

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turbine and radioactive waste buildings, intake structure, administrative and control buildings, reactor and advanced off-gas buildings, and the outside protected areas.

The inspector conducted tours prior to, during, and following the cleanup days and verified that material conditions of the plant improved. Most notable, a significant reduction in transient materials and combustible loading was observed near the main generator stator cooling system l

and iso-phase busses. In the turbine building, temporary material storage areas were organized i

to improve access and aided in the conduct of Auxiliary Operator rounds. In the service water intake stmeture, clemliness of piping and pumps were improved enhancing the material l

condition of this safety-related system.

General clutter consisting of debris, dirt, and

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miscellaneous material from previous work activities were removed improving personnel safety.

j Appropriate management guidance was communicated to the plant staff assuring the proper

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conduct of cleanup activities without adversely affecting safe reactor operations. Management i

and supervisors were assigned plant areas to assure that activities were well controlled.

j Radiation protection personnel were assigned and equipment laydown areas identified to assure

appropriate control of potentially radioactive materials. Fire protection considerations and

contingencies were established that kept the control room staff appraised of transient fire j

loading, The inspector observed senior plant and corporate managers, including the president

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and vice-presidents, working side-by-side with VY staff in performing cleanup activities. The accomplishments of the two cleanup days reflected positively on plant organization team work.

6.0 SAFETY ASSESSMENT AND QUALITY VERIFICATION (40500,90712,92700, l

90713)

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6.1 Plant Operations Review Committee Performance A number of Committee activities were reviewed this inspection period. As documented in Sections 2.5, 3.1.4, and 4.2 of this report, the inspector determined that the PORC was not

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i demonstrating a consistent high level of performance in the conduct of safety assessments required by the plant's TSs during their review of plant operations to detect potential safety hazards. Although strong safety assessment performance was evident during the PORC's review

of the event and Significant Corrective Action Report for the fire that occurred in the RWCU i

demineralizer's control panel, their reviews for issues involving the cross-connection of service i

water and fire water systems and the leakage of reactor coolant system past isolation valves in the core spray system have historically been inadequate. These latter performance issues were l

determined to be a violation of TSs requirements.

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6.2 Open (VIO 93-21-02): Undersized Core Spray Strainers

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On Febmary 9,1994, in response to a Notice of Violation (NOV) issued on January 21,1994, j

VY committed to revise by June 16, 1994 its procedure for preparation of engineering l

calculations and analyses to provide specific instructions on obtaining input data. On June 15, i

1994 the inspector was informed by the ED's Technical Program Manager (TPM) that due to a combination of factors, including senior management direction to obtain a broader perspective t

from industry on the planced process changes, VY was requesting an extension of the I

commitment date until the end of June 1994. The inspector reviewed this request with an NRC

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Region I management representative and informed the TPM that this request was acceptable.

i This extension request would also cover the due date for corrective actions stipulated in LER 93-015. The inspector acknowledged that the need for schedular extensions for NRC related

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corrective action commitments are not frequent and usually occur for sound technical reasons.

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6.3 Review of Written Reports

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The inspectors reviewed Licensee Event Reports (LERs) submitted to the NRC to verify accuracy, description of cause, and adequacy of corrective action. The inspectors considered

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the need for further information, possible generic implications, and whether the event warranted

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further onsite followup. The LERs were also reviewed with respect to the requirements of 10

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CFR 50.73 and the guidance provided in NUREG 1022.

e LER 93-18, Sup.1 Group 4 Primary Contaimnent Isolation on Initiation of "A" Shutdown Cooling System Due to Pressure Spike

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With this occurrence, VY has now experienced pressure spiking in both the "A" and "B" shutdown cooling (SDC) systems (LERs91-006 and 93-011) immediately after residual heat removal (RIIR) pump start. The pressure spikes have resulted in pump trips and SDC isolations, however system and component failures have not occurred. Inspector conclusions regarding VY initial assessment and resolution of this issue were documented in NRC Inspection Report 93-33.

This supplement to LER 93-18 documented that the pressure spikes were caused by void collapse in the section of SDC piping between the injection valve and downstream check valve located within containment. The root cause was attributed to the lack of a high point vent due to inadequate design specifications. Vermont Yankee plans to use the condensate transfer system to flush the piping to prevent void formation pending completion of thermal stress, seismic qualification, and procedural reviews.

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e LER 94-05 Service Water System Configuration Outside of Design Basis On March 30, VY identified that the service water (SW) system operated outside its design for i

approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> when plant operators cross-connected the SW system to the fire j

suppression water supply system.

Inspector review of this event, corrective actions, and management involvement are documented in Section 2.5 of this repon.

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Periodic and Special Reports

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Vermont Yankee submitted the following periodic and special reports which were reviewed for accuracy and found to be acceptable:

Monthly Statistical Report for May,1994

Vermont Yankee Cycle 16 Operating Report This report was submitted in accordance with 10 CFR 50.59(b)(2) and 10 CFR 50.4 describing facility changes, tests, and experiments performed this period. The inspector reviewed a selection of the facility changes described in this report and concluded that the changes and safety evaluations were accurately described. Vermont Yankee also justified the facility changes based on increased margins to operating limits and improved reliability, maintenance, or operation. The inspector discussed the reportability of several design changes and special test procedures with plant management and identified

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no concerns.

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Report of Failed Fuel Status and Parameter Trends for April and May 1994 j

The inspector reviewed the reactor coolant gross activity, offgas activity and slope, and radioactive isotope release rates and identified no abnormal trends. The iodine dose equivalent reached an equilibrium value which corresponds to values observed the previous operating cycle. In the April report, an error involving the use of incorrect units for reactor coolant iodine dose equivalent was identified and subse (ly corrected

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7.0 MANAGEMENT MEETINGS (71707, 30702)

7.1 Preliminary Inspection Findings Meetings were held periodically with VY management during this inspection to discuss inspection findings. A summary of preliminary findings was also discussed at the conclusion of the inspection and prior to report issuance. No proprietary information was identified as being included in this report.

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ATTACIIMENT 1 CRP 9-3 sR-14-20A dllll '

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Safety Classification Designation Spec.A PS-14-47A l

Design Pressure: 1250 psig Design Pressure: 800 psig Design Temperature: 575 degrees Alarm SetPoint: 300 psig Gauge Saver : 10,000 psig SR-14-20A Design Pressure: 450 psig Lift Pressure: 375 psig Design Temperature: 175 degrees Design Flow: 100 gpm Not drawn to scale

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