IR 05000454/1986002

Safety Insp Rept 50-454/86-02 on 860101-31.Violations Noted: Failure to Perform Surveillances within Required Time Interval,Follow Procedures & Effectively Implement Independent Verification Program

ML20153F114
Person / Time
Site:	Byron
Issue date:	02/18/1986
From:	Forney W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20153F085	List: IR 05000454/1986002 ML20153F082 ML20153F098
References
50-454-86-02, 50-454-86-2, NUDOCS 8602250315
Download: ML20153F114 (12)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION III

Report N /86002(DRP)

Docket N License N NPF-37 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Byron Station, Unit 1 Inspection at: Byron Station, Byron, IL Inspection Conducted: January 1 - 31, 1986 Inspectors: J. M. Hinds, Jr.

, P. G. Brochman J.,A.,Malloy Approved By: G &c L. forney, w Lbifef% W Wrs

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Reactor Projects Secti Date Inspection Summary

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l Inspection on January 1 - 31, 1986 (Report No. 50-454/86002(DRP))

l Areas Inspected: Routine, unannounced safety inspection by the resident inspectors of 10 CFR Part 21 reports; operations summary; LERs; design changes and modifications; surveillance; maintenance; operational safety; event l followup and other activities. The inspection consisted of 141 inspector-hours onsite by 3 NRC inspectors including 24 inspector-hours during off-shift Results: Of the eight areas inspected, no violations or deviations were identified in seven areas; four violations were identified in the remaining area: (failure to perform surveillances within their required time interval -

Paragraph 4.b; failure to follow procedures and to effectively implement an independent verification program - Paragraph 4.c; failure to take timely and effective corrective actions for an significant condition which was adverse to safety - Paragraph 4.d; failure to perform a surveillance prior to returning a component to service following maintenance - Paragraph 4.e). Violations 1 and 4 were of minor safety significance, when the surveillances were performed the components were verified to be operable; therefore, the public health and l

safety were not affected. Violations 2 and 3 were of more than minor safety i

significance with the potential to affect the public health and safety.

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B602250315 860220 PDR ADOCK 05000454 0 PDR

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DETAILS 1. Persons Contacted Commonwealth Edison Company

• R. Querio, Station Manager

• R. Pleniewicz, Production Superintendent

• R. Ward, Services Superintendent

• L. Sues, Assistant Superintendent Operations

• T. Joyce, Assistant Superintendent Technical Services

• F. Hornbeak, Technical Staff Supervisor

• A. Chernick, Compliance Supervisor

• Flahive, Assistant Technical Staff Supervisor

• Yates, Onsite Nuclear Safety

• Britton, Quality Assurance Inspector

• J. Langan, Compliance Staff R. Steder, Technical Staff J. Ventimiglia, Technical Staff D. Prisby, Technical Staff The inspector also contacted and interviewed other licensee and contractor personnel during the course of this inspectio * Denotes those present during the exit interview on January 31, 198 . 10 CFR Part 21 Report Followup (92716)

(Closed) 10 CFR 21 Report (454/85003-PP): Diesel Generator tachometer erroneous readings. On September 6, 1995, the licensee identified a problem with the Cooper-Bessecer Emergency Diesel Generators (DGs). This problem involved spurious noise in the speed sensing tachometer which caused an RPM reading to be produced when the engine was secured. At greater than 280 RPM, the starting air solenoid valves become de-energized and due to design of the starting circuit the valves would not energize and open on a starting signal. Consequently, had a valid signal requiring the DGs to start occurred, the DGs could have failed to star The licensee initially ioa tified the problem as coming from a DC to DC regulator in the tachometer circuit; the regulator was replaced and the noise was no longer present. Subsequent investigation by licensee corporate and station personnel determined that a DC to DC converter should be installed to provide better signal isolation. Modification M6-1-85-0638 was issued to accomplish this work and wa:; completed by December 12, 1985. The inspector reviewed the modification package and verified that the DC to DC convertors had been installed. Since the installation of the converters no additional spurious tachometer signals have occurre . Summary of Operations The unit operated at power levels up to 98% until 1103 on January 13, 1986 when a manual turbine runback to 55% power occurred following the trip of the 18 Circulating Water (CW) Pump (see Paragraph 9.b). The CW

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pump was restarted and the unit returned to 98% power the same day. At 0449 on January 16, 1986, the reactor tripped when reactor trip breaker A was inadvertently opened during a surveillance (see Paragraph 9.c). The unit was taken critical at 1549 and tied to grid at 1845 on the same da During power escalation problems with the main condenser were identified and the unit was taken off the grid and placed in Mode 2. During the restart an additional problem was identified with the 1A Main Steam Isolation Valves (MSIV). The MSIV was repaired and the unit was tied to the grid at 1839 on January 18, 1986, and returned to 98% power. At 0006 on January 29, 1986 the reactor tripped on Low-Low Steam Generator water level when the 1A MSIV went shut inadvertently (see Paragraph 9.d).

Following repairs the unit was taken critical at 2147 on the same day and tied to the grid at 0142 on January 30, 1986, and returned to 98% powe . Licensee Event Report (LER) Followup (90712 & 92700) (Closed) LERs (454/85007-02-LL; 454/85080-01-LL; 454/85098-LL; 454/85099-LL; 454/85100-LL): An in-office review was conducted for the following LERs to determine that the reportability requirements were fulfilled, immediate corrective action was accomplished and corrective action to prevent recurrence had been accomplished in accordance with Technical Specification LER N Title 454/85007-02 Control Room Ventilation Isolation Due to a Voltage Transient Induced by a Circulating Water Pump Start 454/85080-01 Incorrect Calorimetric Calculation Resulting in Exceeding 100% RTP 454/85098 Actuation of the Main Control Room Ventilation System Due to De-Energization of the Associated Radiation Monitors 454/85099 Actuation of the Main Control Room Ventilation System Due to the Spiking of the Iodine Channel of the OPR31J Rad Monitor 454/85100 Construction Equipment Blocking Fire Door in Open Position No violations or deviations were identified, (Closed) LER (454/85096-LL): This LER described events from January 10 - December 7, 1905, while in Modes 1, 2, or 3. These events involved the failure to perform Technical Specification l Surveillances on Post Accident Monitoring (PAM) Containment Pressure 1 Channel ;

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Personnel at the Braidwood Nuclear Power Station were conducting a review of procedures, as the licensee intended to use the same procedures at the Byron and Braidwood station. These personnel identified an error in Byron Operating Surveillance, IBOS 3.3.6-1,

" Accident Monitoring Instrumentation Monthly Surveillance", in that the procedure checked the Containment Spray pressure channels 934, 935, 936, and 937 instead of PAM containment pressure channels PC004 and PC00 Technical Specification 4.0.2 states, in part: "Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval..." Technical Specification 4.3.3.6 states:

"Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7." Table 4.3-7, Instrument 1, " Containment Pressure" requires that a CHANNEL CHECK be performed at least once every month and that a CHANNEL CALIBRATION be performed at least once every 18 month Byron Operating Surveillance 180S 3.3.6-1, " Accident Monitoring Instrumentation Monthly Surveillance" performs a CHANNEL CHECK on the Containment Pressure Post-Accident Monitoring Instrumentation channel; however, the B0S was performed on Containment Spray Containment Pressure channels 934, 935, 936, and 937, instead of {

channels PC004 and PC005. Consequently, from January to November J 1985 channels PC004 and PC005 did not have a CHANNEL CHECK performed on them and were therefore considered inoperable. The failure to perform a monthly CHANNEL CHECK on PC004 and PC005 within the required time interval is a violation of Technical Specification 4.3.3.6 (454/86002-01a(DRP)).

l A CHANNEL CALIBRATION was performed on PC004 and PC005 on August 22 and 23, 1983; however, these channels were not recognized as being Technical Specification related and were placed on a 48 month {

calibration schedule. On July 7 and 8, 1985 the 18 month time intervals plus 25% extension were exceeded for PC004 and PC005, respectively. The failure to perform an 18 month CHANNEL CALIBRATION on PC004 and PC005 within the required time interval is a violation of Technical Specification 4.3.3.6 (454/86002-01b(DRP)).

The required calibration wss subsequently performed on December 9, 198 Technical Specificaticn 4.0.3 states, in part: " Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation...." Technical Specification 3. and Table 3.3-10 require a minimum of one Containment Pressure channel to be OPERABLE, while in Modes 1, 2, 3, and 4. Technical Specification 3.3.3.6.b states, in part: "With the number of OPERABLE accident monitoring instrumentation channels...less than the Minimum Channels OPERABLE requirements of Table 3.3-10, restore theinoperablechannel(s)toOPERABLEstatuswithin48hnurs; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />." With both channels PC004 and PC005 inoperable for greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> the licensee

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failed to place the unit in at'least HOT STANDBY within the next 6-3- hours and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The failure *

' to take these actions within the-required time is a violation of y Technical Specification 3.3.3.6(454/86002-01c(DRP)). 4

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As corrective action the licensee has written a new procedure 1 BIS 3.3.6-214, " Surveillance Calibration of Containment Pressure Loops"

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to perform the 18 month CHANNEL CALIBRATION and is preparing a

, revision to 1805 3.3.6-1 to specify channels PC004 and PC005 for

containment pressure. This revision is being tracked by Action Item

! Record AIR 6-85-414. Based on these corrective actions the

! inspector has no further concerns regarding this matter and this j item is considered clcsed; consequently, no reply is required.

j i (Closed)LER(454/85097-LL): This LER described an event on 1985, while in Mode 5. This event involved an

] December inadvertent 12, Safety Injection (SI) actuation occurrenc At 0220 a licensed operator was performing Byron Operating I Surveillance 180S 3.1.1-20, " Train A Solid State Protection System (SSPS) Bi-Monthly Surveillance". Paragraph F.45 states: "At the

logic test panel, PLACE the INPUT ~ ERROR INHIBIT switch to the-I INHIBIT position.", and requires " INDEPENDENT VERIFICATION" of this

} step. Instead, the operator placed the MULTIPLEXER TEST switch in

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the INHIBIT position and when Paragraph F.46 was performed next, a SI occurred. The MULTIPLEXER TEST switch is located' adjacent to the~ i IhPUT ERROR INHIBIT switch on the logic test panel of SSPS. The SSPS

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i i memory circuits must be reset by activating the blocking switches at

the main control board prior to placing both the INPUT ERROR INHIBIT j and the MODE SELECTOR switches in the OPERATE position. Since this '

j was not done the valid Low Steamline Pressure and Low Pressurizer '

i Pressure SI signals were not blocked and a SI actuation occurre An independent verification of the position of the INPUT ERROR

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i INHIBIT switch was not performed prior to the performance of Paragraph F.46. 10 CFR 50, Appendix B, Criterion V states, in part: " Activities i

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affecting quality shall be prescribed by documented instructions, l

procedures, or drawings, of a type appropriate to the circumstances ;

i and shall be accomplished in accordance with these instructions, j

procedures, or drawings...." The failure to place the INPUT ERROR 4 INHIBIT switch in the INHIBIT position as required by Paragraph F.45 is a violation of 10 CFR 50, Appendix B, Criterion V (454/86002-02a(DRP)).

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Byron Administrative Procedure BAP 100-13. " Guidelines for Performance of Independent Verification of Proper Equipment

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i Alignment", Paragraph C.1 states, in part: "All Components tha i

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provide a safety function should be independently verified when alignment changes have been made."' BAP 100-13 Paragraph states: " Independent verification may also be satisfied by a visua i

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verification, apart in time, and documentation of equipment alignment by a second qualified person." The operator performing }

the independent verification in the surveillance would verify the i switch positions for several steps at once, rather than after each ,

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step which required independent verification. This was done to meet '

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the operator's understanding of the " apart in time" requirement of BAP 100-1 Other departments at Byron, such as the Instrument Maintenance department, did not subscribe to this understanding, but would perform an independent verification after each step in which it was required. The failure of licensee management to ensure that the Independent Verification program was effectively implemented during the performance of 180S 3.1.1-20 is a violation of 10 CFR 50, Appendix B, Criterion V (454/86002-02b(DRP)). The safety significance of this event is of more than minor significance due to management involvement and the potential for this event to have occurred in Modes 1, 2, and d. (Closed) LER (454/85101-LL): This LER described an event on December 27, 1985, while in Mode 1. This event involves a Reactor Trip due to an incorrectly positioned control switch in the IB Steam Generator Feedwater Regulating Valve (FRV) control circui CFR 50, Appendix B, Criterion XVI states, in part: " Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrcctive action taken to preclude repetition..."

ANSI N18.7-1976/ANS-3.2, Section 5.2.11 states: "

...In the case of significant conditions adverse to safety, the measures shall assure that the cause of the condition is determined and corrective action taken..." '

At 1230 on December 27, 1985 an instrument mechanic (IM) was performing maintenance on Steam Flow Channel 522 for the IB Steam Generator (SG). The control circuit for the 1B SG Feedwater Regulating Valve (FRV) has three inputs of two channels e'ach: steam flow, feedwater flow, and SG level. The control switch for feed flow was selected to channel 520. The licensed operator who gave permission to the IM to work on channel 522 verified that channel 522 was not selected to control the FRV, as this was the only work to be done on the FRV control circuit. However, Byron Instrument Surveillance 1 BIS 3.2.1-200, " Surveillance Calibration of the Steam Generator Steam Flow / Feed Flow Mismatch Protection Set I" which performs the calibratior, of channel 522, directs the IM to place channels 520 and 522 and 524 (SG level) in test. When this was done the FRV control circuit perceived a large mismatch between steam flow and feedwater flow ar.d caused the FRV to go full open. Even though the operator took manual control of the FRV, SG level rose to the High-High icvel trip setpoint and a turbine trip and subsequent reactor trip occurred. On June 26, 1985 an identical event occurred, except that the operator was able to take manual control in time to prevent the reactor trip. This event was described in Deviation Report DVR 6-1-85-194. The corrective actions taken in response to this DVR was to affix a cover sheet to surveillance packages to specify the correct switch position; however, this was not done for maintenance packages, even though they used the same BI The failure of licensee management to take timely and effective corrective action for a significant condition which was

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adverse to safety is,a violation of 10 CFR 50, Appendix B, Criterion XVI and ANSI N18.7-1976/ANS-3.2, Section 5.2.11 (454/86002-03(DRP)).

e. (Closed) LER (454/85102-LL): This LER described an event during the period of December 1Y to 31, 1985, while in Modes 1, 2, 3 and This event involves the failure to perform a stroke time test on containment isolation valve 1FW039D, "1D Feedwater Preheater Bypass Valve" prior to its return to service following maintenance. This event was identified by the licensee on December 31, 1985, when the System Engineer was reviewing the work package for 1FWO39D prior to performing a stroke time test on the valve. The valve was then satisfactorily tested on Cecember 31, 198 Technical Specification 4.6.3.1 requires that containment isolation valve 1FWO39D shall be demonstrated OPERABLE prior to returning the valve to service after maintenance work is performed on the valve by performance'of a cycling test and verification of isolation tim The failure to perform an isolation time surveillance on 1FWO390 prior to returning it to service is a violation of Technical Specification 4.6.3.1(454/86002-04a(DRP)).

Technical Specification 3.6.3 requires that with containment isolation valve 1FWO39D not OPERABLE action be taken within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore the valve to OPERABLE status or secure the valve in the isolation position; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SilVTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

4 With the containment isolation valve 1FWO39D inoperable for greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the failure to place the unit in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />

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is a violation of Technical Specification 3. (454/86002-04b(DRP)).

The licensee's investigation identified a deficiency in the outage out of service program. During an outage from October 25 to December 13, 1985, the entire feedwater system was taken out of i

service under a " blanket" out of' service, i.e. every component in the system was declared out of service. With the whole system out of service, work could be done freely without additional out of services being processed. During the outage only work on the c$mputerized outage list was to be performed under the blanket out of service. The computerized system was to track work to be performed and testing required prior to unit start-up. This valve, 1FWO390, was not on the outage list. There was no program in place to explicitly track Technical Specification / Mode change required work if it did not appear on the outage list but was under the blanket out of servic The licensee's corrective actions-included revising the computerized outage tracking program to include all work performed during an outage. Additionally, Byron Administration Procedure BAP 300-18,

! " Removing and Returning Equipment Out of Service" will be revised to

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ensure that no work can be performed on a piece of safety related equipment unless it has been specifically taken out of servic This revision is.being tracked by Action Item Reccrd AIR 6-86-00 t

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5. Design Changes and Modifications Review (37700)

The inspector evaluated the below listed Design Changes / Modifications to verify that they had been reviewed and approved in accordance with Technical Specification and Quality Assurance requirement Incluhd in this evaluation were: 1) a review of post-modification test records, results evaluation, deviations and deviation resolution; 2) a verification that operating and surveillance procedure changes required by the design change / modification were made and approved in accordance with Technical Specifications; 3) a verification that training programs were revised to reflect the design change / modification; 4) a verification that as-built drawings were revised and that control room drawings were updated prior to system startup; and 5) a review of the 10 CFR 50.59(b)

annual report to the NRC to verify that it listed those design changes /

modifications which had been performed without prior NRC approva Modification M6-1-84-113 M6-1-84-181 M6-1-84-185 M6-1-85-106 The inspector reviewed the outstanding safety-related Facility Change Requests (FCR) to determine that an excessive backlog of FCRs did not exis The inspector reviewed the licensee's program for temporary modifications, lifted leads and jumpers to verify that: 1) tempcary modifications were reviewed and approved in accordance with Technical Specifications and 10 CFR 50.59; 2) formal records maintained the status of temporary modifications, lifted leads, jumpers and temporary trip setpoints; 3) independent verification and functional testing following the installation or removal of temporary modifications was accomplished; and 4) periodic reviews of the lifted lead and jumper log were conducte No violations or deviations were identifie . Monthly Surveillance Observation (61726)

The inspector observed Technical Specifications required surveillance testing on the Axial Flux Deviation and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that . limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test results conformed with Technical Specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne .

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The inspector also witnessed portions of the following test activities:

SPP 86-06 Split Feed Flow No violations or deviations were identifie . Monthly Maintenance Observation (62703)

Station maintenance activities of safety related systems and components listed below were observed / 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 considered during this review: the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality

, control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and fire prevention controls were

! implemented. Work requests were reviewed.to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performanc The following maintenance activiti- .<ere observed / reviewed:

SI Accumulator Hi/ Low Level Alarm 1MS001A Main Steam Isolation Valve Repair Following completion of maintenance on the SI Accumulator Alarm and 1MS001A the inspectors verified that these systems had been returned to service properly.

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No violations or deviations were identifie . Operational Safety Verification (71707)

The inspectors observed control room operation, reviewed applicable logs and conducted discussions with control room operators during the month o January. During these discussions and observations, the inspectors ascertained that the operators were alert, cognizant of plant conditions, attentive to changes in those conditions, and took prompt action when appropriate. The inspectors verified the operability of selected ,

emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of the auxiliary, turbine and rad-waste buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks.and excessive vibration and to verify that maintenance requests had been initiated for equipment in need of maintenanc .

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The inspectors by observation and direct interviews verified that the physical security plan was being implemented in accordance with the station security pla The inspectors observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection controls. During the month of January, the inspectors walked down the accessible portions of the Containment Spray and Combustible Gas Control systems to verify operability. The inspectors also witnessed portions of the radioactive waste system controls associated with radwaste shipments and barrelin These reviews and observations were conducted to verify that facility operations were in accordance with the requirements established under Technical Specifications, 10 CFR and administrative procedure No violations or deviations were identifie . Onsite Followup of Events at Operating Reactors (93702) General The inspector performed onsite followup activities for events which occurred during January 1986. This followup included reviews of operating logs, procedures, Deviation Reports, Licensee Event Reports (where available) and interviews with licensee personne For each event, the inspector developed a chronology, reviewed the functioning of safety systems required by plant conditions, reviewed licensee actions to verify consistency with procedures, license conditions and the nature of the event. Additionally the inspector verified that licensee investigation had identified root causes of equipment malfunctions and/or personnel error and had taken appropriate corrective actions prior to plant restart. Details of the events and licensee corrective actions developed through inspector followup are provided in Paragraphs b through e belo Manual Turbine Runback on January 13, 1986 -

While in Mode 1, with reactor power at 96% a manual turbine runback to 55% power was initiated when the 1B Circulating Water (CW) pump tripped. The 1B CW pump tripped on loss of excitation to its synchronous motor. The excitation was lost when the supply feeder breaker to motor control center MCC 134U1 opened on a fault. A temporary modification was powered from a breaker at MCC 134U1; however, when the modification was installed the insulation on one of the cables was damaged, resulting in a phase-to-ground faul This fault was not identified prior to energizing the circuit and when the circuit was energized the MCC circuit breaker and the MCC feeder breaker both opened. The licensee is still investigating this and the deviation investigation report will be reviewed when it is issue l Reactor Trip on January 16, 1985  ;

While in Mode 1 with Reactor Power at 98%, the reactor tripped when l the A Reactor Trip Breaker was inadvertently opened during the - 1 performance of Byron Operating Surveillance 1BOS 3.1.1-11, j

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"Bi-Monthly, Staggered Basis, Reactor Trip Breaker Short and Independence Test - Train B". At 0449 a Temporary Undervoltage Procedure thange Trip (85-1-925) to 180S 3.1.1-11 was being perforned for the first time. When paragraph F.10 of the procedure was performed the Train A Test push button was depressed instead of Train B push button which caused the A Reactor Trip breaker to open. Reactor Trip Bypass Breaker A was open at this time consequently a reactor trip occurred. The test push buttons are located in two different cabinets at switchgear 1RD05E; however, this was not indicated in the Temporary Change; additionally, the test buttons were not labeled to indicate that one was Train A and the other was Train B.

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The inadequate procedure and inadequate labeling of plant test equipment were the causes of this reactor trip. All systems responded normally during the reactor trip. As corrective action the licensee has revised the procedure to indicate in which cabinet of 1RD05E Train A and Train B are located and has installed labels on switchgear 1RD05E indicating which test button is Train A and which is Train b. During the restart banging noises were heard in the main condenser at approximately 35% power and the unit was shutdown to Mode 2 to inspect the main condenser. No problems or damage could be identified. The Main Steam Isolation Valves (MSIVs)

were shut to provide isolation for the condenser inspection; when the operators attempted to open the 1A MSIV it failed to do so. The valve was repaired and when functionally tested failed to close within the required time interval; consequently, the valve was declared inoperable. Additional repairs were performed and the valve tested satisfactorily. The unit was returned to service on January 18, 1985. This event will be reviewed in a subsequent report when the LER is issue Reactor Trip on Low-Low Steam Generator (SG) Level on January 29, 1986 While in Mode 1, with reactor power at 98%, the reactor tripped on Low-Low Level in the 1A SG when the 1A MSIV inadvertently closed. A ground was present on the negative side of a standby test closure relay, which is powered frcm 125 VDC Bus 112. This relay is used to perform a fast closure surveillance test, approximately two seconds to close, on the MSIVs. At 0006 the trip coil for ventilation fan OVA 03CE shorted out, and caused a positive ground on Bus 112. As a result of the positive and negative grounds on Bus 112 a current path was producca in the test relay causing it to energize. When the MSIV closed rapidly SG level shrank to the Low-Low Reactor Trip setpoint. The licensee replaced the test relay and the unit was returned to service on January 30, 1986. This event will be reviewed in a subsequent report when the LER is issue No violations or deviations were identifie r:. a

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I 1 Exit Interview (30703)

The. inspectors met with' licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on January 31, 1986. The inspectors summarized the purpose and scope.of the inspection and the findings. The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspection.- The licensee did not identify any such .

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