IR 05000341/1993018

Insp Rept 50-341/93-18 on 930909-1105.Violation Noted. Major Areas Inspected:Operational Safety Verifications, Compression Tube Fitting,Onsite Event follow-up,current Material Condition,Radiological Controls & Security

ML20059M570
Person / Time
Site:	Fermi
Issue date:	11/10/1993
From:	Philips M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20059M544	List: IR 05000341/1993018 ML20059M542 ML20059M558
References
50-341-93-18, NUDOCS 9311190167
Download: ML20059M570 (15)

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U.S. fiUCLEAR REGULATORY COMMISS10f4 REGI0f4 III Report flo. 50-341/93018 (DRP)

Docket rio. 50-341 License fios. f4PF-43 Licensee: Detroit Edison Company 2000 Second Avenue Detroit, MI 48226 Facility flame: Fermi 2  !

Inspection At: Fermi Site, flewport, Michigan l

Inspection Conducted: September 9,1993, through flovember 5,1993 Inspectors: W. J. Kropp I K. Riemer T. Vegel l Approved By: 7MV JfC M. P. Ph%ips} Chief Date '

/ Reactor Projects Section 2B Inspection Summarv Inspection from September 9.1993, throuch flovember 5.1993 (Report tio. 50-341/93018(DRP)

Areas Inspected: Routine, unannounced safety inspection by the resident inspectors of action on previous inspection findings; operational safety verification; compression tube fitting; onsite event follow-up; current material condition; housekeeping and plant cleanliness; radiological controls; security; LER followup; maintenance activities; surveillance activities; engineering and technical support; and report revie Results: Of the thirteen areas inspected, one violation was identified concerning gland steam pressure not restored by operators (paragraph 3.a).

l Four unresolved items were identified that pertained to an inadvertent i

isolation of B31-F016A due to a blown fuse (paragraph 4), a problem associated l with newly installed Valve G3300F120 (paragraph 5.a), MSIVLCS Division 11 not .

t included in the IST program (paragraph 6), and Valve B2103F600 operability li (paragraph 6). Two inspection followup items were also identified that i pertained to single element operation (paragraph 3.a) and compression tube i fitting components (paragraph 3.a).

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'l The following is a summary of the licensee's performance during this i inspection period:

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! Plant Operations

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The licensee's performance in this area was adequate. The operator's response  ;

to the overvoltage trip of the High Pressure Coolant Injection inverter was excellent. Shift turnovers and housekeeping continued to be excellen However, the pre-authorization of work which led to maintenance on a heater ,

drain valve while the system was pressurized resulted in a manual scram of the *

reactor and injury to maintenance workers. Due to recurring problems with  ;

balance of plant and safety related components, the material condition was  ;

. categorized as adequate.

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Maintenance and Surveillance i

The licensee's performance in this area was adequate. The maintenance activities on a heater drain valve while the system was pressurized were caused, in part, by poor communications between maintenance groups. Also, the  ;

licensee identified a Main Steam Isolation Valve Leakage Control System Valve *

that was not in the Inservice Testing Progra i Engineering and Technical Support The licensee's performance in this area was adequate. The inspectors were '

concerned with the cancellation in February 1993 of a Potential Design Change initiated by maintenance in January 1993 that requested engineering assistance in determining the root cause of auxiliary contact failures for motor operated v al ve s . The cancellation of the PDC was not communicated to the maintenance

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organization. Also, the initial determination for operability for a valve

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that was not in the Inservice Testing program was less than adequat !

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DETA1.L_S Persons Contacted j

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Detroit Edison Company ,

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• R. Eberhardt, Superintendent, Radiation Protection f

• P. Fessler, Director, Technical Manager

• J. Malaric, Supervisor, Modifications l

• R. McKean, Plant Manager, Nuclear Production  !

• R. Newkirk, Supervisor, Licensing

• J. Nolloth, Superintendent, Maintenance

• D. Ockerman, Director, Nuclear Training

• D. Pierce, Nuclear Assistant Shift Supervisor  :

• G. Pierce, Work Control i

• J. Plona, Superintendent, Operations

• R. Szkotnicki, Supervisor, Inspection & Surveillance

• J. Tibai, Compliance, Licensing  :

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• G. Trahey, Principal Engineer i *W. Tucker, Superintendent, Technical Engineering

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• J. Walker, General Director, Plant Engineering

• Denotes those attending the exit interview conducted on November 5, 199 '

The inspectors also had discussions with other licensee employees, I including members of the technical and engineering staffs, reactor and '

auxiliary operators, shift supervisors, and electrical, mechanical and  ;

instrument maintenance personnel, and security personne i

, \ Action on Previous Inspection Findings (92701)

(Closed) Unresolved Item (341/93016-02(DRP)): The licensed operators

! did not respond to Annunciator alarms during a reactor trip on August l 13, 1993, which resulted in the closing of the Main Steam Isolation Valves (MSIV) and the use of the reactor safety valves to control reactor pressure. Also, the operators did not receive training on the simulator for a situation where the gland seal system is in the manual j mode. This item will be addressed in the response to the violation  ;

i which is discussed in paragraph 3.a of this repor '

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Plant Operations I

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Fermi 2 operated at power levels up to 93 percent from August 31, 1993, until September 17, 1993, when the licensee commenced a reactor shutdown due to problems with Heater Drain Level Control Valve, N22F415A. The j

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problem with Valve N222F415A was identified on September 8, 1993, when '

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reactor power was being raised from 65 percent after placing the 1 South

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(IS) and 2 South (25) Feedwater Heaters to service. Reactor power was

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maintained at approximately 88 percent from September 8 to September 17, 1993. During the shutdown on September 17, 1993, a manual reactor trip was initiated at 17 percent power by the operators in response to a j 3

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reported steam and water leak in the feedwater system. This event is further described in paragraph 3.b of this repor During startup on September 20, 1993, the main turbine turning gear circuit breaker failed. The reactor was shut down to minimize differential heating of the turbine shaft during the time that the turning gear was out of service. The licensee replaced the turning gear circuit breaker and the.

unit was returned to service on September 21, 199 The unit has since :

operated at power levels up to 93 percen I Operational Safety Verification (71707) i The inspectors verified that the facility was being operated in conformance with the license and regulatory requirements, and that the licensee's management control system was effective in ensuring safe operation of the plan On a sampling basis, the inspectors verified proper control room staffing and coordination of plant activities; verified operator adherence with procedures and technical specifications; monitored control room indications for abnormalities; verified that electrical power was available; and observed the frequency of plant and control room visits by station management. The inspectors reviewed applicable logs and conducted discussions with control room operators throughout the inspection period. The inspectors observed a number of control room shift turnover The turnovers were conducted in a professional manner and included log reviews, panel walkdowns, discussions of maintenance and surveillance activities in progress or planned, and associated LCO j time restraints, as applicabl ;

On August 13, 1993, a reactor trip occurred due to a spurious reactor water Level 8 signal. The event was described in paragraph 3.c of Inspection Report 50-341/9301 During the event, the licensed operators failed to respond to the annunciator alarms for a low gland seal pressure (4D40) and a high condenser pressure (4D108). As a result, the Main Steam Isolation Valves (MSIV) closed which resulted in the loss of the main condenser as a heat sink for the reactor's decay hea Annunciator Response Procedure (ARP) 4D40 " Gland Steam Pressure High/ Low," in the initial response section of the procedure, required the operators for low pressure conditions to throttle open N30F603, the Gland Steam Supply Regula*or Bypass Valve from the 52 inch manifold, to maintain gland steaa pressure at 2 to 3 psi The failure of the operators to properly respond to ARP 4D40 resulted in the loss of j the main condenser as a heat sink and the need to use safety

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relief valves (SRVs) to maintain reactor pressure and the torus as a heat sink for reactor decay heat. This failure to follow ARP 4D40 also resulted in the operators manually opening four of the five Automatic Depressurization System (ADS) SRVs (E, J, P, and R)

to maintain reactor pressur The failure to follow ARP 4D4 is j considered a violation of 10 CFR 50, Appendix B, Criterion v '

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l On October 1,1993, the licensee identified that a valve in the Main Steam Isolation Valve Leakage Control system (MSIVLCS),

B2103F600, had not been included in the Inservice Inspection and  ;

Testing program. Deviation Event Report (DER)93-055 was issued to address this deficiency and is further discussed in paragraph 6 of this report.

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On October 14, 1993, the licensee switched from three element  !

reactor vessel level control to single element contro Prior to ,

that time, the licensee had experienced several random ,

perturbations in steam flow which caused minor variations in actual reactor vessel water level. Subsequent to the switch to  !

single element control, a transient response scenario was run on '

the simulator to determine plant response to a reactor scram while

! in single element control. The simulator response indicated that ,

reactor feed pumps would trip after the level shrink caused by the  ;

scram. An engineering department analysis contradicted this scenario, and the analysis showed that the reactor feed pumps had '

adequate margin to prevent a low suction pressure trip following a 1

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scram while in single element control. Engineering personnel re-

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. analyzed the event and the subsequent analysis validated the l simulator response; the results at the second analysis showed that the reactor feed pumps would trip on low suction pressure after a '

reactor scram, similar to the simulator response. The results of .

the analysis and simulator response were discussed in the l Operations Department Night Orders and discussed at shift turnover. This is considered an Inspector Followup Item (341/93018-03(DRP)) pending NRC review of licensee actions taken to resolve the root cause of the reactor vessel level control issu b. Comoression Tube Fittinos While conducting routine tours, the inspectors noted several

examples where components on compression tube fittings were interchanged with those of another manufacturer. Specifically,

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the tube fitting body and associated nuts were manufactured by different vendors. These questionably configured compression fittings were identified on various instrument manifolds and components includin ;

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Emergency Diesel Generator 11 Standby air compressor and air  !

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receiver instruments R30-N562A and R30-N560A

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Panel H21 P448 - EECW Pressure and Level Equipment

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Panel H21 P505 - DW Drain Sump Division 1 (

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Instrument rack for LXP N22-N400C, N401B, N400B As discussed in NRC Information Notice 92-15, " Failure of Primary System Compression Fittings," the interchanging of hardware from  ;

different manufacturers was one of the problems that might impact the effectiveness of the compression fit. In addition, maintenance procedure 35. CON.017, " Fabrication, Installation,

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Repair, and Removal of Q and Non-Q Instrument Tubing," Section i 2.0, Precautions and Limitations, states, "Do not interchange ;

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compression fitting parts (bodies, ferrules, nuts, etc.) between manufacturers." The inspectors discussed the observation of mixed compression fittings with the license Due to previous incidents involving compression fitting problems, reactor water cleanup instrument line failures in May and July 1988, the licensee was cognizant of the potential problems that could result from ;

improper compression fitting installation. Though no mixed compression fittings were identified on high energy lines, further '

licensee attention is required. The observation of mixed compression fittings in the field may be indicative of lax maintenance practices, inattention to detail of a procedural requirement, in ensuring that proper components are being used while installing compression fittings. This matter is considered Inspection Followup Item pending further review by the NRC (341/93018-02(DRP)).

c. Onsite Event Follow-up (93702)

During the inspection period, the licensee experienced several events, some of which required prompt notification of the NRC pursuant to 10 CFR 50.72. The inspectors pursued the events onsite with licensee and/or other NRC officials. In each case, the inspectors verified that any required notification was correct and timely. The inspectors also verified that the licensee initiated prompt and appropriate action The specific events were as follows:

e On September 17, 1993, the operators manually tripped the reactor at 17 percent power when a , team and water leak was reported in the feedwater system. At the time of the trip, the licensee was performing a cor trolled shutdown of the unit to repair a Heater Drain Leo el Control Valve, N22F415A.

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8, 1993 when reactor power was being raised from 65 percent after returning the 1 South (15) and 2 South (25) Feedwater Heaters to service. The licensee's investigation determined that the valve's stem and disc were no longer connecte The design of the valve consisted of the stem being screwed into the disc body with a pin inserted into the disc and i stem at approximately 45 degrees. To repair the valve, the i licensee decided to shut down the unit to hot standby since Valve N22F415A could not be isolated during power operatio l The licensee developed an outage schedule for the repair of Valve N22F415A that included the tagging out of service the heater drain (HD) system and subsequent draining of the HD system prior to working on Valve N22F415A. Prior to the draining of the HD system, the Nuclear Shift Supervisor (NSS) authorized a partial tag out of the air lines to Valve ,

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N22F415A to allow instrumentation and control (I&C)

personnel to disconnect the valve's actuator from the air line This authorization to commence work was contrary to the established outage schedule. Maintenance personnel performed work beyond the scope defined by the NSS. The mechanical maintenance personnel proceeded to remove the actuator from Valve N22F415A which resulted in the ejection of the disconnected stem from the valve since the HD system was still at approximately 800 ps The resulting opening in the HD system caused a steam and water leak in the feedwater system which resulted in 1st and 2nd degree burns to the mechanical maintenance personnel and the subsequent manual trip of the reactor by the operator The manual trip allowed the operators to isolate the HD system. This event was investigated by the NRC and the results of the investigation will be documented in Inspection Report No. 50-341/9302 e On October 29, 1993, while performing preventive maintenance on the spare 130 Vdc battery charger, 2B l-2, the charger tripped on overvcitage. The overvoltage condition on the battery charger caused an overvoltage condition at the High !

Pressure Coolant Injection (HPCI) inverte The HPCI inverter tripped on the overvoltage which caused the flow controller for HPCI to switch to the manual mode and the flow demand to drop to approximately 40 percent. The flow control demand dropping to 40 percent resulted in the HPCI system being inoperable since the required HPCI flow of 5500 gallons per minute would not be met. The licensee entered the limiting Condition for Operation (LCO) for HPCI. This condition existed for approximately two minute The HPCI inverter was restored to an operable status and the flow controller was placed in automatic with a 100 percent demand setpoin However, the licensee verified that there was no damage to the HPCI inverter before exiting the Technical ,

Specification LC0 for HPCI. The inspectors will review the '

associated LER for the results of the licensee's investigation and subsequent root cause and corrective l action )

l d. Current Material Condition (71707) l The inspectors performed general plant as well as selected system and component walkdowns to assess the general and specific material condition of the plant, to verify that work requests had been initiated for identified equipment problems, and to evaluate housekeeping. Walkdowns included an assessment of the buildings, components, and systems for proper identification and tagging, accessibility, fire and security door integrity, scaffolding, ,

radiological controls, and any unusual condition Unusual l conditions included but were not limited to water, oil, or other l

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liquids on the floor or equipment; indications of leakage through  !

ceiling, walls, or floors; loose insulation; corrosion; excessive noise; unusual temperatures; and abnormal ventilation and lightin Based on recurring equipment / component problems such as ,

i the voltage regulator problems with Module Power Units (MPU) and i valve G3300F120 noted in paragraph 5.a of this report; the i

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auxiliary contact switches (Deviation Report (DER)93-505) noted in paragraph 6 of this report; heat deterioration of wiring (DER  !

539); blown fuses on dedicated shut down panel (DER 547); and  !

other equipment / component problems noted in the licensee's ,

1 Equipment Concerns list, the material condition of the plant is  ;

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considered adequat :

i Housekeepina and Plant Cleanliness

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The inspectors monitored the status of housekeeping and plant *

cleanliness for fire protection and protection of safety-related ,

equipment from intrusion of foreign matter. The housekeeping was  !

considered excellent during this inspection perio ;

] Radioloaical Controls (71707)

r The inspectors verified that personnel were following health  !

physics procedures for dosimetry, protective clothing,' frisking, l

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posting, etc., and randomly examined radiation protection l

- instrumentation for use, operability, and calibratio The i i radiological conditions of the plant as noted by few personnel  !

! contaminations and low contaminated square footage were considered  !

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i Security (71707)  ;

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Each week during routine activities or tours, the inspectors '

monitored the licensee's security program to ensure that observed

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actions were being implemented according to the approved security l pl an . The inspectors noted that persons within the protected area )

displayed proper photo-identification badges, and those l

individuals requiring escorts were properly escorte Additionally, the inspectors also observed that personnel and packages entering the protected area were searched by appropriate ,

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l One violation was identified.

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Licensee Event Report (LER) Follow-up (92700)

Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to 1 determine that reportability requirements were fulfilled, that immediate corrective action was accomplished, and that corrective action to prevent recurrence had been or would be accomplished in accordance with Technical Specifications (TS):

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• (Closed) LER (341/93008): Recorder T50-R802A, " Division 1 l Drywell and Torus Pressure" was discovered by the licensee I to be powered from an outlet other than the designated power source. The inspectors review of the licensee's correction actions and root cause was documented in Inspection Report ,

No. 50-341/93012 and subsequent Enforcement Conference !

documented in Inspection Report No. 50-341/9301 e (Closed) LER (341/93010): Reactor scram caused by a spurious reactor water Level 8 signal when a Nuclear Power Plant Operator (non-licensed operator) cleaned a tape off an instrument valve. The inspectors reviewed the licensee's root cause and corrective actions. Based on the licensee's analysis of this event, this LER is considered close * 1 Closed) LER (341/930111: Inadvertent isolation of B31-F016A due to blown fus The licensee determined that B31-F016A closed due to a blown fuse in the control logic; the most probable cause of the blown fuse was the grounding of a jumper installed by a nuclear supervising operator (NS0).

The root cause of the event was a design deficiency which did not allow secure installation of the jumper Two prior LERs occurred at the station due to inadvertently grounding jumpers during removal, one of which also involved operations department personnel. Pending further NRC review of the event and of the licensee's past corrective actions to address problems associated with jumper removal, this item is considered an Unresolved item (341/93018-04(DRP)).

e (Closed) LER (341/93012): High Pressure Coolant Injection (HPCI) suction re-aligned from the condensate storage tank to the torus. The inspectors reviewed the licensee's root cause and corrective actions. Based on the licensee's analysis of this event, the inspectors consider this LER close e (Closed) LER (341/93013): Manual reactor scram from 17 percent power when an inadvertent breach of the reactor feedwater system occurred during maintenance activities.

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The description of this event is further discussed in paragraph 3.b of this report and will be reviewed during a special inspection documented in Inspection Report No. 50-341/9302 In addition to the LERs, the inspector reviewed licensee's Deviation Event Reports (DERs) during the inspection period. This was done in an effort to monitor the conditions related to plant or personnel performance, potential trends, etc. DERs were also reviewed to ensure that they were generated appropriately and dispositioned in a manner consistent with the applicable procedures. The inspectors identified a concern with the j

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operability justification in DER 93-0555 that pertained to inadequate testing on a Main Steam Isolation Valve Leak Control System (MSIVLCS) valve and DER 93-0507 that pertained to auxiliary contacts for motor operated valves. These DERs are further discussed in paragraph 6 of this repor No violations or deviations were identifie . Maintenance / Surveillance (62703 & 61726) Maintenance Activities (62703)

Routinely, station maintenance activities were observed and/or reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards, and in conformance with technical specification The following items were also considered during this review:

limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; and activities were accomplished by qualified personne Portions of the following maintenance Work Requests (WR) were observed or reviewed:

e 000Z934523 Repair G3300F120 actuator

• 0002934525 Repair tubing on EDG-12 air compressor

• 000Z934733 Install EDP 26009

• 000Z934739 Repair of MPU # 1

• 000Z934800 Replace actuator spring-G3300F120 1) Modular Power Unit No. 1 A failure of MPU No. I had occurred on September 6, 1993, and Technical Specification (TS) Action Statement 3.8.3.1.a.l.e was entered. Investigation by the licensee determined that the regulated supply voltage from Modular Power Unit (MPU) No. I to distribution cabinet No. 2 was 98 Vac instead of the expected 130 Vac. The repair was completed on September 7, 1993, and consisted of replacing ;

the voltage adjust module in the voltage regulato '

On October 11, 1993, the licensee received Annunciator Alarms 8D57, "Div. I H2/02 Channel Inop," and 8D63, "Div.I H2/02 Flow Low." The licensee determined that decreased voltage ,

(98 Vac) at Distribution Cabinet No. 2 alarms 8D57 and 8D63 i to annunciate. The licensee at 5:08 a.m. on October 11, 1993, declared MPU No.1 inoperable and entered TS Action Statement 3.8.3.1.a.l.e which required that MPU No.1 be i

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restored to an operable condition within 8' hours or be in )

Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold Shutdown l within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The licensee initiated work request (WR) 000Z934739 to repair MPU No.' The repair j consisted of replacing the voltage regulator for  !

Distribution Cabinet No. 2 which was completed prior to the l time required to be in Hot Standby by the TS Action !

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i Since MPU No. 1 failed twice within 35 days, the inspectors }

reviewed the Work Request (WR) history on the six MPUs !

1 installed in the plant. MPUs Nos. 1, 3, 4, 5, and 6 ,

utilized internally mounted voltage regulators while MPV N ;

2 has the regulator mounted externally. The inspectors :

determined that other voltage regulator problems had !

occurred on the MPU No. 2 in 1988, MPU No. 3 in 1990, and }

MPU No.5 in 1991, in addition to the two problems noted with ;

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i MPU No. 1 in 1993.

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On October 6, 1993, Reactor Water Clean-Up (RWCU) to l Feedwater Spring Assist Close Check Valve, G3300F120, _j failed Surveillance 24.707.001, "RWCU Valve Operability

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i j Test." The licensee initiated WR 000Z934523 to replace the j actuator. After replacing actuator Valve G3300F120, it !

still would not close properly and the actuator was removed i from the valve. To ensure valve G3300F120 would close, the i licensee's engineering organization issued a Engineering

Change Package (ECP) to install a counterweight on the ;

actuator shaft to further assist the valve to close along ,

with the spring in the actuator. While investigating why :

the new installed actuator would not close the valve, the i licensee determined that the spring in the actuator did not ,

have the correct spring constant. The correct spring was l l installed using WP, 000Z934800, along with the counterweight !

] using WR 000Z934733. In April 1993, the actuator for Valve i j G3300F120 was replaced due to intermittent movement noted j

, during Surveillance 24.707.01 performed in January and March !

i 1993. The licensee determined that the actuator installed !

4 in April 1993 also had the wrong spring installed. Based on l

the problems with Valve G3300F120 in January, March, April, i j and October 1993, the inspectors obtained the work history for Valve G3300F12 Pending further review of the maintenance and surveillance history i of the MPUs and G3300F120 by the NRC, this matter is considered an '

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Unresolved Item (341/93018-05).  ;

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l b. Surveillance Activities (61726)

During the inspection period, the inspectors observed technical specification required surveillance testing and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that results conformed with technical specifications and procedure requirements and were reviewed, and that any deficiencies identified during the testing were properly resolve The inspectors also witnessed or reviewed portions of the following surveillances:

• 24.000.05 Monthly Continuity Light and Channel Check

• 24.137.03 MSIV Leakage Control System Monthly Operability Test

• 24.137.15 MSIV Leakage Control System Functional Test

• 24.137.16 MSIV Leakage Control System Monthly Operability Test

• 24.137.18 Main Steam Line Drain and Drain Isolation Valve Operability Test

• 24.307.014 EDG-ll start and Load Test, Fast Start On October 4, the inspectors noted that torus water temperature, as indicated at the Dedicated Shutdown Panel, was approximately 79 degrees fahrenheit. Torus water temperature as indicated in the main control room was approximately 39 degrees fahrenheit. The inspectors questioned the control rcom operators about the readings and associated channel check requirements. The acceptance criteria per Procedure 24.000.05, " Monthly Continuity Light and Channel Check," is that the indicated difference is no greater than 4 degrees Fahrenheit. The licensee performed the monthly channel check surveillance procedure (already scheduled for that day), declared the alternate shutdown instrumentation inoperable, and entered the appropriate LCO. A calibration performed on the instrumentation was satisfactory. The licensee concluded that the instrumentation was accurate, and the observed temperature differences were due to stratification. The licensee later initiated torus cooling and the temperature indications converge A historical review by the licensee indicated that prior performance of the surveillance had passed because they had been performed relatively soon after torus cooling was put into operation. The licensee's investigation showed that the thermal stratification was most likely due to safety valve leakage, and the thermal stratification effects were seen due to the different elevations of the temperature sensor The licensee revised Procedure 24.000.05 to recommend that the Residual Heat Removal system be put in the torus cooling mode of operation prior to recording data to promote mixing and breakup of stratificatio The inspectors reviewed the licensee's investigation results and corrective actions and had no concern .

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Ilo violations or deviations were identifie . Enaineerina & Technical Support (37700)

During this inspection period the inspectors reviewed the following Deviation Event Reports (DER):

• DER 93-055 During Surveillance Test 24.137.18 " Main Steam Line Drain and Drain Isolation Valve Operability Test," on October 1,1993, the system engineer and Inservice Testing (IST) engineer identified that Valve, B2103F600, Main Steam Drain Isolation valve, was not included in the station's Inservice Inspection and Testing (IST)

progra The IST and system engineer had determined that Valve B2103F600 was utilized as the Main Steam Isolation Valve Leak Control System (MSIVLCS) Isolation Boundary Valve for Division 11 of MSI'1LCS and should have been required to be stroked time tested in accordance with the licensee's IST program. Therefore, the <

Technical Specification (TS) Surveillance requirement 4. l appeared not to have been met for Valve B2103F600 and DER 93-055 l was issued with a determination for operability. The j determination for operability documented in DER 93-055 on October 1,1993, was based on Surveillances 24.137.16 and 24.137.18 and l Preventive Maintenance Activity B750. Surveillance 24.137.16 performed monthly verified that Valve B2103F600 strokes (not timed) and Surveillance 24.137.18 performed every 18 months verified that the MSIVLCS system functions properly which indicated adequate leak tightness of Valve B2103F600. Preventive Maintenance (PM) B750 to inspect, lubricate, and test the motor operator that was performed in September 1989 and April 1991 showed no degradation. The DER stated the only test not completed was the stroke time test; however, the stroke time test for closing was not a factor in the valve operability since the operators had 20 minutes following a postulated loss of coolant accident (LOCA) to close Valve B2103F600 and initiate MSIVLC Therefore, the licensee's position was that MSIVLCS Division 11 had been and remains operable even though not formally included in the IST progra The inspectors concluded that the licensee did not adequately assess the operability of Valve B2103F600 on October 1, 199 Plant conditions on October 1, 1993, did not preclude the ability to stroke time Valve B2103F600 and comparing the results with the stroke times obtained during PM activity B75 performed in 1989 and 199 Stroke timing of B2103F600 was a predictive maintenance activity to identify degradation of valve performance and not directly related to the amount of time after a LOCA the Valve B2103F600 was required to operat The basis used by engineering

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that the valve stroke time was not required based on the operators having 20 minutes following a LOCA to close the valve was not adequat During a review of the DER on October 4,1993, the Plant Manager also had a concern with the basis for operability as stated in DER 93-055 and requested a re-review of the operability determinatio On October 14, 1993, a revised operability determination was completed. The revised operability determination included a comparison of the stroke time of Valve B2103f600, which was obtained on October 5, 1993, to the stroke times obtained at the motor control center breaker for the valve in September 1989 and April 1991, during Preventive Maintenance activity B750. The ,

stroke times observed on October 13, 1993, September 18, 1989, and April 6, 1991 were approximately 28 seconds which were within the master valve list indicated stroke time of 21 to 35 seconds. The inspectors have concluded that the revised basis for operability was satisfactory. Pending further review by the NRC and the licensee, the initial determination for operability for Valve B2103F600 performed on October 1, 1993, is considered an *

Unresolved item (341/93018-06).

• DER 93-507 DER 93-507 was issued on August 30, 1993, to identify the failure of the Reactor Core Isolation Cooling (RCIC) minimum flow to open to relieve a high RCIC suction pressure. The licensee's ;

maintenance effectiveness group identified that the root cause was a stuck auxiliary "b" contact on the closed contactor. The DER identified that the maintenance effectiveness group had previously initiated PDC 13770 in January 1993 after analyzing the 1988-1992 inputs to the motor operator valve (MOV) trending program. This analysis identified at least eight other MOV failures attributed to auxiliary contacts. PDC 13770 requested engineering assistance in determining root cause of the problems with the auxiliary '

contact On February 5, 1993, system engineering sent a voided notification back to the maintenance organization stating that the concern with the auxiliary switches should be handled via communications with nuclear engineering so specific design changes can be determine While reviewing the failure of the RCIC minimum flow valve in August 1993 and other recent MOV failures, l the maintenance effectiveness group determined that PDC 13770 was voided by engineering. Because of 1993 valve failures caused by

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problems with auxiliary contacts, the maintenance effectiveness group initiated Potential Design Change (PDC) 14218 on September 22, 1993, to again request engineering assistance to address the repeated problems noted with the auxiliary contacts. The inspectors were concerned with the cancelling of PDC 13770 and the lack of adequate followup by engineering and maintenance from January 1993 to October 1993 to address a potentially common mode I

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failure of MOVs. Pending further review by the NRC, this matter j is considered an Unresolved Item (341/93018-07).  !

No violations or deviations were identifie i Report Review l

During the iir,pection period, the inspector reviewed the licensee's  !

Monthly Performance Report for September 199 The inspector confirmed ,

that the information provided met the requirements of Technical i Specification 6.9.1.6 and Regulatory Guide 1.1 l No violations or deviations were identifie ! Inspection Followuo items Inspection Followup items are matters which have been discussed with the ;

licensee, which will be reviewed by the inspector, and which involve '

some action on the part of the NRC or licensee or both. Two Inspection  !

Followup Item (s) disclosed during the inspection are discussed in .

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paragraph . Unresolved Items  ;

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Unresolved items are matters about which more information is required in ;

order to ascertain whether they are acceptable items, violations, or  ;

deviations. Four Unresolved Items disclosed during the inspection are  ;

discussed in paragraphs 4, 5.a, and . Meetinos and Other Activities Management Meetinas (30702)

On September 30, 1993, Mr. Ed Greenman, Director, Division of .

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Reactor Projects, toured the fermi plant and met with licensee management to discuss plant performance and plant material condition. On October 25 and 26, 1993, Mr. Jack Martin, Regional -

Administrator, Region III, Mr. Ed Greenman, Director, Division of Reactor Projects, and other members of the Region 11I staff toured :

the Fermi plant and met with licensee management to discuss plant i j performance and plant material conditio .

1 + Exit Interview (30703)

The inspectors met with the licensee representatives denoted in ,

paragraph I during the inspection period and at the conclusion of l the inspection on November 5, 1993. The inspectors summarized the -

scope and results of the inspection and discussed the likely

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content of this inspection report. The licensee acknowledged the information and did not indicate that any of the information disclosed during the inspection could be considered proprietary in ;

natur j 15  ;

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