ML20154D973

Monthly Operating Repts for Apr 1988

ML20154D973
Person / Time
Site:	Sequoyah
Issue date:	04/30/1988
From:	Shawn Smith TENNESSEE VALLEY AUTHORITY
To:	NRC
References
NUDOCS 8805200044
Download: ML20154D973 (51)

Text

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TENNESSEE VALLEY AUTHORITY 0FFICE OF NUCLEAR POWER SEQUOYAH NUCLEAR PLANT MONTHLY OPERATING REPORT TO THE NUCLEAR REGULATORY COMMISSION APRIL 1988 UNIT 1 DOCKET NUMBER 50-327 LICENSE NUMBER DPR-77 I

UNIT 2 DOCKET NUMBER 50-328 LICENSE NUMBER DPR-79 Submitted by:

4 S. J. Scii t h, Pla n t M a n ar,e r

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8805200044 880430 PDR ADOCK 05000327 R

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r TABLE OF CONTENTS Page 1.

Operational Summary l

Performance Summary 1

Significant Operational Events 1-3 j

Fuel Performance and Spent Fuel Storage Capabilities 4

PORVs and Safety Valves Summary 4

Special Reports 4-5 Licensee Events 5-15 Offsite Dose Calculation Manual Changes 16 l

Radwaste Summary 17 II. Operating Statistics A.

NRC Reports Unit One Statistics 18-20 i

Unit Two Statistics 21-23 l

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TVA Reports "1

3 Nuclear Plant Operating Statistics 24 g

Unit Outage and Availability 25-26

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Reactor Histogram 27 l

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111. Maintenance Summary Instrument Maintenance 28-29 Mechanical Maintenance 30-31

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Modifications 32-39 1

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I OPERATIONAL

SUMMARY

PERFORMANCE

SUMMARY

April 1988 The following summary describes the significant operational activities for the month of April.

In support of this summary, a chronological log of significant events is included in this report.

Unit I remained in an administrative shutdown the entire month because of design control review, configuration control updating, and resolution of significant employee concerns. Outage-related maintenance and modifications are being performed. Unit 1 has been off line 982 days. Preparations continue for unit 2 restart and power operation. Unit 2 has been off line 983 days.

SIGNIFICANT OPERATIONAL EVENTS Unit 1

, Date Time Event 04/01/88 0001E The reactor was in mode 5.

The administrative shutdown due to design control review, configuration control updating, and resolution of significant employee concerns continues.

04/30/88 2400E The reactor was in mode 5.

The administrative shutdown due to design control review, configuration control updating, and resolution of significant employee concerns continues.

Unit 2

, Date Time Event 04/01/88 0030E Mode 3 RCS at 375 degrees, 900 psi. Activities continue for power operation.

2300E Maintenance on the pressurizer safety valves is complete.

04/02/88 0325E Began heatup for pressurizer safety valve testing.

1616E RCS at 384 degrees, 1765 psi.

1723E Testing begins on pressurizer valve 68-563.

Valve inoperable.

2300E Begin cooldown, valve failed acceptance criteria.

2320E Terminate cooldown, valve meets acceptance criteria.

2325E Began heatup.

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Unit 2 Date Time Event 04/03/88 1635E RCS at 545 degrees, 2235 psi. Valve activities continue.

1748E Testing begins on pressurizer valve 68-564.

1752E Valve incperable.

2030E Begin cooldown, valve failed acceptance criteria.

2301E Terminate cooldown, valve meets acceptance criteria.

RCS at 421 degrees, 1750 psi.

04/04/88 1344E Testing begins on pressurizer valve 68-565. Valve inoperable.

1457E Testing terminated, valve meets acceptance triteria.

1526E Begin heatup.

04/05/88 0129E RCS at 547 degrees, 2235 psi. Activities continue for power operation.

04/06/88 1745E Primary to secondary leakage based tritium analysis, 435 gallons / day. Within technical specifications.

1846E Loop No. 4 isolated for leakage.

2148E Loop No. 4 in service.

2211E Initiate leak testing on loop No. 3.

04/07/88 0815E Tube leak on No. 3 3/G 234 gal / day.

1740E Initiated cooldown to repair No. 3 S/d tube leak.

2323E Entered mode 4.

04/08/88 1010E Entered mode 5.

1615E RCS at 120 degrees 250 psi.

2115E Draining No. 3 S/G for tube plugging.

04/10/88 1035E No. 3 S/G drained.

1659E RCS at 122 degrees, zero pressure.

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Unit 2 Date Time Event 04/14/88 0030E Tube identified in No. 3 S/G. Leak found in row No.1.

1800E The remaining S/Gs will be tested. Row No. 1 of all S/Gs will be plugged.

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04/20/88 0600E Nanways off all S/Gs.

04/21/88 0700E S/G tube plugging in progress.

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04/2$/88 1448E All S/G's manways are on and torqued. Maintenance I

complete.

04/30/88 2400E RCS at 124 cegrees, zero pressure. Activities continue for restart and power operation.

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FUEL PERFORMANCE Unit 1 The core average fuel exposure accumulated during April was 0 MWD /MTU with the total accumulated core average fuel exposure of 0 MWD /MTU.

Unit 2 The core average fuel exposure accumulated during April was 0 MWD /MTU with the total accumulated core average fuel exposure of 8097.51 MVD/MTU.

i SPENT FUEL PIT STORAGE CAPABILITIES The total storage capability in the SFP is 1,386.

However, there are five cell locations which are incapable of storing spent fuel.

Four locations (A10 All, A24, and A25) are unavailable due to a suction strainer conflict, and one location (A16) is unavailable due to an instrumentation conflict.

Presently, there is a total of 348 spent fuel bundles stored in the SFP.

Thus, the remaining storage capacity is 1.033.

l PORVs AND SAFETY VALVES

SUMMARY

No PORVs or scfety valves were challenged in Apell 1988.

SPECIAL REPORTS The following special reports were submitted to NRC in Apell 1988.

1-88-07 On March 20, 1988, at 1200E, all four fire pumps were technically i

inoperable in accordance with TSs and LCO 3.7.11.1 was entered.

Fire pumps 1A-A and IB-B were removed from service for routine testing of fire pump 1A-A while fire pumps 2A-A and 28-B were 3

inoperable because of a design deficiency described in LER 7

SQRO-50-327/87042 Revision 1.

The design deficiency identified that during a LOCA, the fire pumps could start and potentially degrade the auxiliary power system and overload the standby emergency power source. Even though the fire pumps were technically inopsrable, all four pumps were available utilizing i

operator action.

Surveillance testing of fire pump 1A-A was successfully completed on March 20, 1988, and LCO 3.7.11.1 was I

exited at 1707E with fire pumps IA-A and 1B-B operable.

On March 23, 1988, at 1403E, all four fire pumps were again inoperable to perform surveillance testing on fire pump 18-B.

Upon successful completion of surveillance testing, LCO 3.7.11.1 was exited at 2034E with fire pumps IA-A and IB-B operable.

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SPECIAL REPORTS (continued) 2-88-08 Fire door C-14 (to auxiliary instrument room, unit 1. control building elevction 685) was opened and was not reluened to opecable status within the 7-day limit allowed by TS LCO 3.7.12.

It was opened due to a broken door knob. The door was e.spablo of being e.losed and latched, but could not be opened easily witheat the knob.

The door was opened March 1, 1988, and returned to functional status on March 11, 1988.

LICENSEE EVENT REPORT (S)

The following licensee event reports (LERs) were transmitted to the Nuclear Regulatory Commission in April 1988.

Description of Event LER 1-88013 Oa March 14, 1988, at 1200E, it was discorered that four S/G blowdown system RMs had not been adequately functionally tested. They have not been tested to ensure they can provide a close signal to the FCVs and that the FCVs will close upon receipt of a high radiation condition in the S/G blowdown effluent to the cooling tower blowdown line or as a result of a loss of instrument power. The functional tests are required to meet a quarterly TS SR 4.3.3.9.

Subsequent to the discovery of this condition, manual valves in series with the FCVs were locked closed and tagged to preclude an inadvertent release via this pathway.

The cause of this condition is attributed to an oversight during the implementation of the SI Review Program. This program was implemented, in part, to ensure all SRs are implemented within the Sequoyah SI network. As corrective actions, the unit 2 SI used to functionally test RMs that provide automatic functions was revised to include these functional tests. The tests were satisfactorily completed following the revision, and the unit 2 manual control valve was reoponed. The appilcable unit 1 SI will also be revised in a similar manner, and the functional tests will be satisfactorily completed before the unit 1 restart.

1-58014 On March 14, 1988, with unit 1 in mode 5, a train "A" CVI occurred at approximately 1110E while maintenance was being performed.

Instrument Maintenance (IM) personnel improperly actuated the local start switch for the sample flow pump on containment purge exhaust 1-EM-90-130.

The switch actuation caused a spurious high radiation spike that was of sufficient magnitude and duration to trip the associated RM circuitry and initiate a unit 1 CVI. Operations personnel verified that the CVI was not caused by an actual high radiation condition and then reset the CVI.

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1-88014 The immediate cause of this event was an EMI-induced high radiation (cont.) spike. During subsequent investigation, it was determined that i

actuation of the pump switch induced EMI to the RN sample pump status (on/off) circuitry and resulted in the CVI. The root cause i

of this event was the failure of IN personnel to comply with l

q applicable configuration control requirements. Following the completion of a postmodification test, the subject switch was not returned to its original configuration. As soon as IN personnel discovered this, they returned the switch to the proper position without notifying Operations and requesting the RN's output signals to be blocked. For immediate corrective action, the CVI was reset and a memorandum was sont to all IN personnel stating that i

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Operations personnel shall be contacted and the RN, trip signal blocked before performing any work on RMs capable of actuating ISF l

equipment. Also, the RN samplo pump motor switch will be replaced before unit 1 entry to mode 4.

SQN is continuing the investigation a

i of the event and will submit a supplemental report by May 31, 1988, describing actions taken to prevent recurrence of this event.

1-88015 On March 17, 1988, at 1500E, it was discovered that TS SR 4.11.2.1.2 was not completely incorporated into the implementing procedure. TS SR 4.11.2.1.2 is implemented, in part, by SI-410.2, i

"Containment (Upper, Lower) Purge," which only requires that a sample be taken and analyzed of the compartment to be purged since the containment purge system can be aligned for purging the upper or lower containment compartments. TS SR 4.11.2.1.2 requires an

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t analysis before all containment purge operations. The cause of SI-410.2 incompletely implementing TS SR 4.11.2.1.2 is attributed to an oversight during the SQN initial CI-1, "Surveillance l

Program " Appendiz F review of SI-410.2.

This SI is used to review sis in order to ensure TS SR are properly implemented. The oversight made during the review was the failure to require that both containment compartments be samplot and analyzed in SI-410.2.

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SI-1. Appendir F reviews were conducted on all sis during initial program implementation in 1986, l

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purge was not in progress at the time of the discovery. As corrective actions to properly implement TS SR 4.11.2.1.2, SI-410.2 1

was revised on March 18, 1988, to require that samples be taken from the upper and lower containment compartments before conducting

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containment purge operations, 1

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a Description of Event LER 1-88016 On March 24, 1988, at approximately 1021E, a main steam line isolation signal occurred when an inadvertent actuation of a high steam flow blatable (1-FS-1-10B) occurred. At the time of this event, another high steam flow bistable (1-FS-1-21A) was already in the tripped condition because of ongoing maintenance work.

Also, the lo-lo Tayg (RCS temperature average below 540 degrees F) and low steam line pressure (below 600 psi signals were present because of the plant being in mode 5. g)herefore, all the required T

logic was completed (high steam flow in two out of four loops coincident with lo-lo T or low steam line pressure in two out avg of four loops) to give the ESF actuation signal. The SI signal also genereted from this logic was blocked as allowed by TS 3.3.2.1 below permissive P-12 (T below 540 degrees F).

Therefore, ay sincetheMSIVswerealreahyclosedformode5andtheautomaticSI circuitry was blocked, no equipment was actuated.

An immediate investigation was initiated into the root cause of the inadvertent bistable actuation (1-FS-1-108), but no root cause has been determined. An investig.tlon is continuing into the root cause of this event.

A supplemental report will be submitted by May 31, 1988, to provide details of this investigation and any necessary recurrence control actions.

1-88017 On March 31, 1933, at 1946E, a train "A" CVI occurred on unit 1.

While performing routine auxiliary building rounds, an AUO noticed that the abnormal flow alarm light for the containment purge exhaust RM (1-RM-90-130) was illuminated on the local panel.

In order to clear the abnormal flow indicater, the AUO jogged the RM sample pump switch. Upon depressing the pump switch stop button, the AUO received indication of an increased radiation count rate that exceeded the monitor's trip values, resulting in a trip alarm.

The AUO then notified the unit 1 00 of his actions and observations.

Upon receiving the AUO's report, the UO checked the status of the containment ventilation systems by review of the MCR associated isolation valves' indicator lights and found that a train "A" CVI had occurred.

The UO verified that a high containment radiation condition did not exist by reviewing the 1-RM-90-130 trace recording in the MCR, then proceeded to return the containment ventilation system to normal in accordance with SOI-30.2B, "Containment Ventilation System Isolation."

The root cause of this event is attributed to improperly controlling the operation of the sample pump local switch after it had been identified as an immediate cause of a previous CVI.

When the RM pump switch was found to have worn contacts during the investigation of a previous CVI (reference LER 50-327/88014), the pump switch should have been tagged to ensure another switch actuation would not occur before replacement of the switch.

For interim corrective actions, an HO was placed on the local sample pump switch to prevent operation. As permanent corrective actions to prevent recurrence, the local sample pump switch will be replaced befora unit 1 enters mode 4.

Description of Eveilt

_ LER 1-86022 This revision updates the corrective action section to accurately Rev. 1 reflect the recurrence control.

1-87078 This revision provides information on long-term correctiva action.

Rev. 2 1-88010 This esvision updates the corrective action.

Rev. 1 1-88011 This cotision pt9 tides additional information regarding the Rev. I corrective actions taken to prevent the recarrence of this event.

1-86S02 On March 28, 1988, at approximately 1300E, two plant laborers who were cleaning overhead pipe hangers in the unit 1 MOV board room (elevation 749 of the auxiliary building--a vital area) discovered a paper wrapper enclosed in a cellophane bag suspected to contain a small quantity of a controlled substance. The laborers were using a vacuum to clear out the pipe hangers, which were hollow hangers approximately eight feet off of the floor, when the substance was found. Upon this discovery, NSS was immediately notified. The NSS removed the substance from the vital area for examination and to conduct a field test. The substance had probably been placed in the hanger some time during the construction phase of the plant, as the paper in the bag was very brittle and had turned to a yellow color.

At approximately 1420E, the field test confirmed that the substance found was marijuana. At this time, a one-hour phone call was made to NRC in accordance with 10 CFR 73.71 requirements. A followup search was conducted of the remaining pipe h ngers in the area with negative results. A narcotics dog was also used to check the area. This i

search also did not produce any additional findings.

4 The cause of this event is unknown. The marijuana found in tite pipe hanger had obviously been in the hanger for quite a long time, possibly since construction timeframes. Thus, it would be almost impossible to determir.e who brought the substance into the plant.

NOTE: Beginning with this LER, the numbering system for safeguard LERs is i

being changed to be consistent with guidance in NUREG 1304 issued in i

February 1988. The numbering system will change from LERs with "500"

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numbers (i.e.. 1-88-501) to numbers with an "S00" (i.e., 1-88501),

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Description of Event

_ LER 2-88011 At 2030E on March 6, 1988, with unit 2 in mode 3 (1600 psig.

450 degrees F), the unit 2 cold leg accumulator 3 was declared inoperable due to the failure to perform SR 4.5.1.1.1.b. which requires verification of the boron concentration within six hours after each solution volume increase for greater than or squal to one 1

percent of the tank's volume. The RCS inventory at 1600 psig was i

leaking into the cold leg accumulator 3 at 400 pelgt therefore, the accumulator had to be drained periodically. After draining the accumulator, it was refilled with RCS inventory. The Operations shift crews did not consider the increase due to the fact that the RCS leak wss a filling operation; hence, the Radiochemistry Laboratory was not notified to verify the boron concentration in cold leg accumulator 3.

The cold leg accumulators are designed to ensure a sufficient volume of borated water will be forced into the core in the event of a large line break. At the time of this event, the RCS boron concentration was above 2000 ppm; thus, the inleakage into the accumulator would not have caused a dilution of the accumulator below TS limits. The missed SR did not have an adverse affect on the safety of the plant.

l The causes and corrective actisn sections of this report will be revised by May 15, 1988, to detail the results of the investigation of this event.

2-88012 On March 5, 1988, at approximately 2230E, with unit 2 in mode 3, it was discovered that 2-LCV 's-175, which supplies AFW from the turbine-driven AFW pump to number 4 S/G, could not be opened from the MCR. However, the Operations shift crew did not immediately enter LCO 3.7.1.2 because it was believed that a previous TS interpretation involving LCOs 3.0.5 and 3.7.1.2 allowed this condition to exist without entering the LCO. Following a subsequent evaluation of this event, SQN determined that action statement (b) to LCO 3.7.1.2 had been applicable during this event and a late entry was made into the subject LCO. A similar event occurred on March 6, 1988, at approximately 2120E.

In this case, Instrument Maintenance persannel were backfilling an instrument sensing line to 1

correct a deviation in S/G 4 level indicator LI-3-107.

Since this 1evel indicator sharea a connon sensing line with level P.ransmitter LT-3-175, which controls 2-LCV-3-175 position, the backfilling i

procedure required the subject LCV to be placed in manual. At this time, the Operations shift crew did not declare the valve inoperable because it was known that an automatic AFV signal would override the manual control mode. However, because the sensing line for 2-LCV-3-175 was inoperable, it could not be assured that the subject valve could properly control the level. Thus, the valve should have 1

been declared inoperable, and LCO 3.7.1.2 should have been entered.

The events were caused by an inadequate and improperly utiliaid TS interpretation. In addition, the procedure that was used to backfill the common sensing line of the subject level instrumentation did not adequately address the applicability of LCO 3.7.1.2.

In order to prevent the recurrence of this event, plant procedures have been revised and operator training increased, and a formal review of all TS interpretations currently in use was performed.

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Description of Event LER 2-88013 This summary described three events in which a manual reactor trip was initiated in accord 6nce with the action statement of TS LCO 3.1.3.3.

On March 16 and 17, 1988, with unit 2 in, node 3, rod control system testing was being performed in preparation for entry into mode 2.

At two different times during the performance of this testing, it was determined that the rod group demand position indication was not within i 2 steps of the actual domand position as required by TS LCO 3.1.3.3.

As a result, plant operators compiled with the action statement of the subject LCO and opened the reactor trip breakers.

On March 19, 1988, unit 2 was maintaining mode 3 conditions while additional testing of the rod control system was performed. During this testing, rod group movement for shutdown bank "B" was out of sequence by three steps. Although the demand step counters accurately indicated rod position, the operator conservatively assumed that LCO 3.1.3.3 was applicable and opened the reactor trip breakers.

The first event was caused by a failure of the demand step counter circuitry, and the second event was caused by internal binding of the step counter itself.

In both of these events, the rod groups moved as designed. The third event was caused by an open switch to the rod lift coils that precluded movement of one group of rods. As a result, a shutdown bank was withdrawn out of sequence, and the operator conservatively tripped the reactor, even though the rod demand position indication was accurate. To prevent recurrence of these events, various components have been replaced in the rod control system, and procedures will be impler.ented to clean and test the rod control circuitry on a periodic basis. Alsc, LCO 3.1.3.3 will be evaluated to determine if a change to this LCO is appropriate.

2-88014 On March 20, 1988, at 0702E, with unit 2 in mode 3 (547 degrees F.

2235 psig), an ESF signal was generated while placing the "A" string intermediate pressure heaters in service for filling and deserating portions of the condensate system. A pressure anomaly in the condensate system occurred while opening 2A intermediate heater isolation valve from the MCR. The condensate piping downstream of the intermediate heaters was at low pressure due to vacuum in the main condenser and the 2B main MF#P racirculation valve to the condenser being partially open. The pressure change in the condensate system caused MFWP gland seal injection water pressure to drop below the MFVP trip setpoint. Since only one gland seal injection pump was in service, the presaure drop could not be compensated for and subsequently tripped both MFWPs. This condition generated ca AFV pump start signal as designed. Since both motor-driven AFW pumps were in service at the time of the event (as part of normal plant operation during mode 3), only the turbine-driven feedwater pump started (ESF actuation). The immediate cause of the event was concluded to be the balance of plant system perturbation. The root cause of the event was that the MFVP trip bus was not deenergized during filling and deserating the condensate !

Description of Event

__LER 2-88014 system. At 0703E, the seal injection water pressure returned to (cont.) normal, the MCR operators stopped the turbine-driven AFV pump, removed the auto start signal, and stabilized the plant condition.

As a corrective action, appropriate plant procedures will be revleeu to doenergize MFVp ttip bus while filling and deserating the condensate system and reenergize the trip bus before entry into mode 2.

2-88015 On March 24, 1988, while unit 2 was in mode 3. Electrical Maintenance (EM) personnel were performing maintenance on train "A" reactor trip bypass bresker in accordance with an approved maintenance instruction (MI). At approximately 0130E, during the performance of the bypass breaker cell switch inspection, an inadvertent train "A" MFI occurred. This was the result of a reactor trip breaker "open" signal concurrent with a low RCS average temperature. A subsequent evaluation of this event revealed that the trip breaker "open" signal was the result of EM personnel operating the cell switch in the train "A" bypass breaker compartment as required by MI-10.9.2.

All maintenance activities on the reactor trip bypass breakers were halted and actions were taken to stabilize the plant.

A review of the procedure revealed that it did not have the necessary steps to preclude an MFI when more than one reactor trip breaker was "racked out" or completely removed from its compartment. In addition, an evaluation that had been performed to assess the impact of MI-10.9.2 on plant operation was inadequate. To prevent recurrence of this event, instruction change forms were issued to permanently change the mis for the reactor trip breakers as well as the bypass breakers. Also, a review of the content of this LER will be performed with the personnel responsible for reviewing the plant impact evaluation associated with the performance of the mis.

In addition. TVA will review this event with individuals involved in the fleid validation of the subject mis.

2-88016 During in-service pressurizer coda safety valve lift pressure setpoint testing conducted on April 2-4, 1988, with unit 2 in hot standby, all three relief valves excGeded the TS lift setting. Upon this discovery, each valve was declared inoperable, the appropriate action statement of TS was entered, setpoint* adjustments were made, and subsequent testing was conducted to ensure compliance with TS.

The in-service testing performed on April 2-4, 1988, determined the valve lift settings through calculations based on measured parameters

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using a lift assist assembly (Trevitest). The previous testing and adjustments of these lift settings were made during a bench test at a contract laboratory using hot nitrogen pressure as the lifting medium.

On April 8, 1988, unit 2 was returned to cold shutdown in order to locate and repair S/G tube leaks. At this time, the relief valves were removed and sent to the same contract laboratory for additional bench testing. The bench testing was conducted using water as the lifting nodium. Test pressure was controlled by regulating the nitrogen pressure which was loaded on the water system. During this Description of Event

_ LER 2-88016 testing, the valve lift settings did not meet the TS required (cont.) setpoint until adjustments were made. The lif t-settings determined by this bench test were below the TS limit. Sequoyah chose to recognize these bench test results for the lift settings. With this position, the valves were rendered inoperable during the performance of the in-service testing. The cause of this event was attributed to several factors: 1) different methodology used to determine the valves' lift settings between the in-service and bench tests.

2) the replacement of valve parts made before the most recent bench test was conducted, and 3) the different environmental conditions between the in-service and bench testing. As corrective actions to prevent recurrence, a consistent method will be used to determine the pressurizer code safety valve lift settings.

2-88017 On April 1, 1988, an inadvertent reactor trip signal was generated on unit 2 when Operations personnel attempted to secure a control power fuse on the source range (Neutron Flux Channel N-31).

The fuse locks into position by a metal connector that has two metal tabs that lock into the fuse holder as the connector is inserted and rotated.

Attached to the outer portion of the fuse connector is a plastic cover that is used to faellitate removal / installation of the fuse.

The plastic cover has two flat surfaces that are normally in the horizontal position when the fuse connector is locked into position correctly. Before this event, the plastic cover was observed to be rotated slightly such that the flat surfaces were not horizontal.

Thus, the fuse appeared to be in a condition such that the potential existed for the fuse to disconnect at any time and generate a reactor trip signal. When Operations personnel attempted to lock the fuse in place, control power was interrupted momentarily and, subsequently, the reactor trip signal generated. No reactor trip occurred because the reactor trip breakers were open and the rods inserted. The fuse was left in place with control power restored and the trip signal reset; but the flat surfaces on the plastic cover still did not line up. The fuse assembly was subsequently replaced on April 13, 1988, and the only deviation found, as compared with other similar fuse connectors, was that the outer plastic cover was rotated slightly.

This is a cosmetic condition only and has no effect on the function of the fuse.

The cause of this event is attributed to the lack of a work document being initiated before manipulating the fuse and a misunderstanding

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of the fuse connector construction. To prevent recurrence, this i

event will be reviewed with Operations shift personnel by May 31, 1988, to emphasize the need for proper work package t

preparation before work and to provide an awareness of the construction of this type fuse assembly. -

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Description of Event LER 2-88018 On March 30, 1988, at 1440E while unit 2 was in hot standby, it was discovered that a threaded cap was installed on a tube fitting tee in the sense line to a local p-essure gauge (2-PI-63-74).

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installed between the inboard and outboard containment isolation valves in a test system line used to measure RCS boundary valve seat

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1eakage. Due to the location of the cap. It serves as s 5oundary

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required to satisfy containment integrity requirements. When Operations personnel were made aware of the pipe cap finding, the action statement of TS LCO 3.6.1.1 was entered for appropriate compensatory measures. This event was caused by an inadequate review by SQN of the penetration which had a threaded cap following receipt of the NRC position. This review has since been verified to be complete. No additional penetrations were found to have inadequate containment integrity devices.

2-88019 At 1215E on Apell 7, 1988, with unit 2 in mode 3. it was determined that TS LCO 3.0.3 had been in effect since approximately 1156E that morning. This LCO was applicable because the handswitches for the "B" train RHR pump and the "A" train CCp had both been placed in the "pull-to-lock" position, thereby rendering both pumps inoperable. As a result, two independent trains of ECCS subsystems were not 1

available, ar.d unit 2 was not in compliance with either the LCO or the action requirements of TS 3.5.2.

At 1226E, CCP 2A-A was returned to service and LCO 3.0.3 was exited.

i The immediate cause of this event was a personnel error that resulted from the high level of activity in the MCR. The root cause of this event was inadequate work control. The handswitches for both pumps l

were placed in the "pull-to-lock" position in accordance with approved plant procedures; however, these procedures should not have i

been allowed to be performed concurrently. To ensure that the activity in the MCR is maintained at a reasonable level during the restart of unit 2, the shift operations advisor will assist the shift supervisor in monitoring the activity level in the MCR and recommend that activities be slowed or stopped when, in his opinion, too many activities are in progress simultaneously. In addition, SQN will revise the plant procedure for work control to ensure that activities controlled by Operations are adequately considered during the wort impact evaluation, t

i 2-88003 This revision updates the cause and corrective action sections of 1

Rev. I this report, i

2-88005 This revision provides an update of completed corrective a:tions and Rev. I a restated event analysis.

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AB3REVIATIONS Page 1 of 2 Auxiliary Building Gas Treatment System 1.

ABGTS Auxiliary Building Secondary Containment Enclosure 2.

ABSCE Auxiliary Building Isolation 3.

ABI 4

AFW

- Auxiliary Feodwater Abnormal Operating Instruction l

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AOI Assistant Unit Operator 6.

AUO Boric Acid Storage Tank 7.

BAT 8.

BIT

- Boron Injection Tank Condition Adverse To Quality Report 9.

CAQR 10.

CCP

- Centrifugal Charging Pump 11.

CCW

- Component Cooling Water 12.

CRI Control Room Isolation 13.

CREVS

- Control Room Emergency Ventilation System 14.

CSS (CS) - Containment Spray System 15.

CVI

- Containment Ventilation Isolation Diesel Generator (s) 16.

D/G(s) 17.

DCR

- Desi$r. Change Request 18.

DNE

- Division of Nuclear Engineering Emergency Core Cooling System 19.

ECCS 20.

ECN

- Engineering Change Notice Emergency Gas Treatment System 21.

EGTS Electromagnetic Interference 22.

EMI Environmentally Qualified / Environmental Qualification 23 EQ 24.

ERCW

- Essential Raw Cooling Water Engineered Safety Feature (Actuation) 25.

ESF(A) 26.

FCV

- Flow Control Valve Final Safety Analysis Report 27.

FSAR 28.

FVI

- Feedwater Isolation 29.

GOI

- General Operating Instruction 30.

GPM

- Gallons Per Mintte 31.

H0 Hold Order 32.

IMI

- Instrument Maintenance Instruction 33.

LCV Level Control Valve Limiting Condition for Operation l

34.

LCO 35.

LOCA

- Loss Of Coolant Accident Maximum Allowable Stroke Time 36.

MAST 37.

MFI Main Feedwater Isolation Main Feedwater Pump 38.

MFP 39.

MOV Motor Operated Valve 40.

MSI

- Main Steam Isolation I

Main Steam Isolation Valve 41.

MSIV 42.

MCR

- Main Control Room 43.

NSS

- Nuclear Security Service 44.

NSSS

- Nuclear Steam Supply Systen 45.

PORC

- Plant Operation Review Committee 46.

PRO

- Potential Reportable Occurrence 47.

RCS Reactor Coolant System 1

48.

RHR Residual Heat Removal 49.

RM Radiation Monitor (RAD Monitor / RAD MON) 50.

RVST Refueling Water Storage Tank

._=.

.. ~.

i t

ABBREVIATIONS i

t

)

Page 2 of 2 I

51.

SCR Significant Condition Report Spent Fuel Pit j

52.

SFP Steam Generator (s) i 53.

S/G(s)

Surveillance Instruction /or Safety Injection 54.

SI i

55.

SMI Special Maintenance Instruction System Operating Instruction 56.

SOI Sequoyah Nuclear Plent 57.

SQN Surveillance Requirement 58.

SR Solid State Protection System 59.

SSPS Temporary Alteration Control Form 60.

TACF Technical Instruction 61.

TI Technical Specification (s) 62.

TS(s)

Unit Operator 63.

U0 64.

WP Workplan Work Request 65.

WR 1

OFFSITE DOSE CALCULATION MANUAL CHANGES No chant.cs were made to the SQN Offsite Dose Calculation Manual (ODCM) in Apell 1988.

1 RADWASTE

SUMMARY

I i

April 1988 i

I 1.

Total volume of solid waste shipped offsite:

A.

Dry active waste:

936 ft.3, Activity: 3.9067 curies B.

Spent resins, sludges, bottoms: 94 ft.3, Activity:

2.5595 curies A

1 j

Shipped: April 22, 1988 Chem-Nuclear. Inc.

i 2.

Radwaste on site and awaiting shipac7t:

A.

Rosin in storage: 109 ft,3

]

B.

Estimate r9 sin that will be generated: 148 ft.3 C.

Dry active waste: Number of drums - 32 j

Number of boxes -

6 i

i il 1

a 1

I i

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I i

b i

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i I

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)

i 4

17 k

I e

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l.

(

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l r

l j

I t

i OPERATING STATISTICS t

(NRC PCPORTS) i I

1 a

l 1

5 a

t

{

5 L

r i

d j

i d

i i

a l

OPERATING DATA REPORT DOCKET NO. 50-327 DATE MAY 05. 1988 COMPLETED BY D.C.DUPREE i

TELEPHONE (615)B70-672?

CFERATING STATUS 1.

LNI T NAME: SEGUOYAH NUCLEAR PLANT. UNIT 1

NOTES:

2.

Pt'POR T PERIOD: APRIL 1988 3.

L ICENSED THERMAL POWER (MWT):

3411.0 4.

NAMEPLATE RATINC (GROSS MWE):

1220.6 5.

1)ESIGN ELECTRICAL RATING (NET MWE):

1148.0 6.

MArjMUM DEPENDABLE CAPACITY (GROSS MWE):

1183.0 7.

MAXIMUM DEPENDABLE CAPACITY (NET MWE):

1140.0 0.

I' CHANGES OCCUR IN CAPACITY RATINGS (ITEMS NUMBERS

- THROUGH 7)SINCE LAST REPORT. GIVE REASONS:

9.

IOWER LEVEL TO WHICH RESTRICTED, IF ANY(NET MWE):

10. REASONS FOR RESTRICTIONS. IF ANY:

THIS MONTH YR.-TO-DATE CUMULATIVE

11. t3UP5 IN REPORTING PERIOD 719.00 2903.00 59904.00 1c. TENDER OF HOURS REACTOR WAS CRITIC AL O.00 0.00 24444.91

13. IE AC T OR RESERVE SHUTDOWN HOURS 0.00 0.00 0.00 14 K UR S GENERATOR ON-LINE O.00 0.00 23781.13

15. UNIT RESERVE SHUTDOWN HOURS 0.00 0.00 0.00

16. GROSS THERMAL ENERGY GENERATED (MWH) 0.00 0.00 77060971.91

17. GPOSS ELECTRICAL ENERGY GEN. (MWH) 0.00 0.00 25976386.00

10. NET ELECTRICAL ENERGY GENERATED (MWH)

-4239.00

-23140.00 24831183.00

19. UNIT SERVICE FACTOR O.00 0.00 39.70

20. t 'NI T AVAILABILITY FACTOR O.00 0.00 39.70

21. UNIT CAPACITY FACTOR (USING MDC NET) 0.00 0.00 36.11

22. UNIT CAPACITY FACTOR (USING DER NET) 0.00 0.00 36.11

23. UNIT FORCED OUTAGE RATE 100.00 100.00 53.S'

24. ?HJTI;OWM SCHEDULED OVER NEXT 6 MONTHS (TYPE. DATE. AND DURATION OF E ACi-25.

l' SnUTDOWN AT END OF REPORT PERIOD. ESTIMATED DA*.E OF STARTUP:

TARTUP IS DETERMINED AT THIS TIME PENDING DESIwN CONTROL REVIEW, C ONF I GUR AT ION CONTROL UPDAT ING. AND RESOLUTION OF SIGNIFICANT EPPLOYEE CONCERNS.

NOTE. HAT lHE THE YR.-TO-DATE AND CUM 01ATIVE VALUES HAVE DEEN UPDATED.

J j

)

SEQUOYAH NUCLEAR PLANT l

AVERAGE DAILY POWER LEVEL t

.i DOCKET NO. : 50-327 UNIT : ONE DATE : MAY 02,1988 i

COMPLETED BY : D.C.DUPREE TELEPHONE : (615)870-6722 a

]

MONTH: APRIL 1988 a

AVERAGE DAILY POWER LEVEL AVERAGE DAILY POWER LEVEL DAY (MWe Net)

DAY (MWe Net) 01 0

16 0

l 02 0

17 0

i 03 0

18 0

j 04 0

19 0

j

.i 05 0

20 0

i 06 0

21 0

l 5

07 0

22 0

+

08 0

23 0

09 0

24 0

4 1

10 0

25 0

4 11 0

26 0

i 12 0

27 0

13 0

28 0

14 0

29 0

J 2

15 0

30 0

i 4

i

4

UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO.

50-327 UNIT NAME s,mmvah one DATE

%v 7.

1988 COMPLETED BY n_ r_ Dunree TELEPHONE (615) 870-6722 REPORT MONTH Anri1 19RR

-o c

cg oyg Licensee go, va, Cause & Corrective c*

No.

Date g

g 5; g

ygg Event gg gg Action to y

3o g

f g a:

Report #

gu go Prevent Recurrence e.c u

c m h v

oc o

n

~

1 880101 F

719 F

4 Design Control, Configuration Updating, and Employee Concerns.

j l

9 e

3 Method:

• Exhibit C-Instructions 2 Reason:

3F: Forced S: Scheduled A-Equipment Failure (Explain) 1-Manual for Preparation of Data B-!!aintenance or Test 2-Manus 1 Scram.

Entry Sheets for Licensec l

C-Refueling 3-Automatic Scram.

Event Report (LER) File l

D-Regulatory Restriction 4-Cont. of Existing (NUREG-0161)

E-Operator Training T. License Examination Outoge l

F-Administrative 5-Reduction 5 Exhibit 1-Same Scurce C-Operational Error (Explain) 9-Other l

Il-Other (Explain)

1 i

OPERATING DATA REPORT

[

1 DOCKET NO. 50-328 i

DnTE MAY 05, 1998

{

j COMPLETED DY D.C.DUPREE i

TELEPHONE (615)S70-6722

}

1 OTERAT149 STATUS h

2 j

1.

_u!T NAME SEOVOYAH NUCLEAR PLANT, UNIT 2

NOTES:

[

i 2.

REPORT PERIOD. APRIL 19G8 3.

ICLN3ED THERMAL POWER (MWT )

3411.0

{

}

4 NAMEPLATE RATING (GROSS MWE)-

1220.6 j

5 DFSIGN ELECTRICAL RATING (NET MWE):

1140.0 6.

r:A < IMUM DEPENDADLE CAPACITY (GROSS MWE):

1183.0

?.

MnXIMUM DEPENDADLE CAPACITY (NET MWE):

1148.0 i

O.

IF CHANGES OCCUR IN CAPACITY RATINGS (ITEMS NUMBERS r

7 THROUGH 7)SINCE (,AST REPORT. O!VE REASONS' i

i 9

'CUER LEVEL TO WHICH RESTRICTED. !F ANY(NET MWE):

{

10. REASONS FOR RESTRICTIONS, IF ANY:

i i

1 l

THIS MONTH YR.-TO-DATE CUMULATIVE i

11. 64tHS IN REPORTING PERIOD 719.00 2903.00 51864.00 l

12. fANDER OF HOURS REACTOR WAS CRITICAL O.00 0.00 21994.54 j

i

13. NEACTOR RECERVE SHUTDOWN HOURS 0.00 0.00 0.00 14 HOURS GENERATOR ON-LINE O.00 0.00 21494.22

13. i.N!i RESERVE SHUTDOWN HOURS 0.00 0.00 0.00

{

j 16.

Gh0SE lHERMAL ENERGY GENERATED (MWH) 0.00 0.00 69127977.22 l

J t 7.

Gh0SS ELECTRICAL ENERGY GEN. (NWH) 0.00 0.00 23536780.00 t

}

10.

M' ELECTRICAL ENERGY GENERATED (MWH)

-9153.00

-51751.00 22456394.60 17 eNIT SERVICE FACTOR O.00 0.00 41.44 i

i

20. UNIT AVAILABILITY FACTOR O.00 0.00 41.44

21. UNIT CAFACITY FACTOR (USING MDC NET) 0.00 0.00 37.72 i

22. UNIT CAPACITY FACTOR (USING DER NET) 0.00 0.00 37.72 l

23. UNIT FORCED OUTAGE RATE 100.00 100.00 54.23 1

]

24 GHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE. DATE. AND DURATION OF E 1

i 25.

r SHUTDOWN AT END OF PEPORT PERIOD, ESTIMATED DATE I.

STARTUP:

i TF E R EST ART OF UNIT-2 IS MAY 1988.

I I

l NOTE THAT THE THE YR.-TO-DATE AND i

CUMut oTIVE VALUES HAVE DEEN UPDATED.

I 1

1

'I f

j ;

1

._.y...

,._~_.-~__m.._,,

, _, _,,.. _ ~ ~..,,, _ _,

\\

l l

1 SEQUOYAH NUCLEAR PLANT j

AVERAGE DAILY POWER LEVEL i

)

i DOCKET NO. t 50-328 i

UNIT : TWO DATE : MAY 02,1988 COMPLETED BY : D.C.DUPREE i

TELEPHONE (615)870-6722

{

MONTH: APRIL 1988 AVERAGE DAILY POWER LEVEL AVERAGE DAILY POWER LE'!EL DAY (MWe Net)

DAY (MWe Net) 01 0

16 0

02 0

17 0

03 0

18 0

r 04 0

19 0

05 0

20 0

06 0

21 0

07 0

22 0

I 08 0

23 0

09 0

24 0

40 0

25 0

11 0

26 0

12 0

27 0

l I

13 0

28 0

14 0

29 0

15 0

30 0

~

UNIT !>HUID0VNS AND P0k'ER REDUCTIONS DOCKET NO.

50-328 UNIT NAME seminvah Two DATE May 2, 1988 COMPLETED BY D. c. Dupree REPORT 110 NTH April 1988 TELEPHONE (615) 870-6722 l

"s.

a Y cc 0 Licensee ro en Cause & Corrective ee ce oa c

co oe ce 36; oga Event gg gg Action to l

No.

Date e

o y

5 u cr:

Report #

gu o o n

ei u Prevent Recurrence i

=w cc e.c c u

z v. g a

l 1

880101 F

185 F

4 Design Control, Configuration Updating, and Employee Concerns.

2 880407 F

513.2 F

9 Repair No. 3 S/C tube leak.

U, 3

880430 F

20.8 F

9 Maintenance of the pressurizer safety valves.

3tlethod:

4 Exhibit G-Instructions 2 Reason:

8 F: Forced S: Scheduled A-Equipment Failure (Explain) 1-flanua l f or Preparation of Data B-!!aintenance or Test 2-Manus! Scras.

Entry Sheets for Licensee C-Refueling 3-Automatic Scram.

Event Report (LER) File D-Regulatory Restriction 4-Cont. of Existing (NUREG-0161)

E-Operator Training 1. Licessse Examination Outage F-A.iministrative 5-ReJuction G-Operational Error (Explain) 9-Other 5 Exhibit I-Same Source II-Other (Explain)

l OPERATING STATISTICS (TVA REPORTS)

b TV A 7302 A tcW*P4-all NUCLEAR PLANT OPERATING STATISTICS j

i 1

SE0t'0_YAH Nt'Ct. EAR Plant l

l i

719 ANL 88 P w od Hours Month 39 l

Item No.

Unit No.

UNIT 1 UNIT 2 PLANT i

l 1

Ave < ace Hourly Gross Load. kW 0

0 0

k 3

Core Thermal Energy Gen. GWD (t)2 0

0 ~

0 2

Ma.imum Hour Net Generation MWh 0

0 0

4 Steam Gen. Thermal Energy Gen, GWD (t)2_

0 0

0 j

j 5

Gross t lectricai nen. MWh 0

0 0

6 station use, MWh 4,239 9.153 13.392 l

j

[

7 Net Electrical Gen, MWh

-1 239

-9,153

-13.392 U

8 station Use, Percent N/A N/A N/A -

i l

9 Accum. Core Avg E aposure, MWD / Ton!

0 0

0 l

10 CTE G This Month,106 BTU 0

0 0

0 11 SGTE G This Month,106aTU 0

0 12 13 Hours Reactor was critical 0.0 0.0 0.0 14 Un.t use, Hours Myn 0:00 0:00 0:00 15 capacity F actor, Percent 0.0 0.0 0.0 16 Turbine Avail Factor Percent 0.0 0.0 0.0 I

17 Gspnator Agt. F actor. Percent 0.0 0.0 0.0 g

18, A teen Avait Factor. Percent 0.0 0.0 0.0 t

19 Reactaduj;L11tter. Pucent 0.0 0,0 0.0 0

20 Umt Ava.t rector. Percent 0.0 0.0 0.0

_2L.Aeme starturi 0

0 0

.l.3Jctor Cold Startues 0

0 0

?3 3

24 Gross Heat Rate, Blu,%Wh N/A N/A N/A

[

25 Net Heat Rate, atu %Wh N/A N/A N/A i

1 26 C

27 g

28 TNottle Pressure, ps+g N/A N/A N/A g 2 Threttle Temperature.

• F N/A N/A N/A j

30 E n haust Pr essure, inHg Abs.

N/A N/A N/A q

[

31 Intake Water Temp.

• F N/A N/A N/A 1

32 31 Main F eedwater, M ie/hr N/A N/A N/A i

34

[

35 36 37 Full Power Capeityi EFPO 404.86 363.65 768.51 l

38 Accum Cycle Full Power Days, F FPO 0.0 210.8416 210.8416 l

s 39 Oil Fned for Generation, Gallons 1,$18 I

5 40._QEutma V2.we. Bis / cat 138,000 l

_.il._Dtte.LQenerat ren. M W h 23 A?

Ma s. Hour Net Gen.

Man. Day Net Gen-Load I

MWh Time Date M Wh Date Factor, %

l 43 N/A N/A N/A N/A N/A N/A

/

j R ema r k s - A For BF NP this value is MWD /5TU and for SQNP and WBNP this value is MWD /MTU.

}

l 2(t) indicates T hermal Energy.

1, 5

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SEQUOYAH ONE REACTOR HISTOGRAM 110 100 - 90 - 80 - I { 70 - U 60 - i 50 - 40 - g h 30 - r 20 - 10 - 0--c -10 i i i i i i i i i i i i i i i i o 4 8 12 16 20 24 28 32 APRL 1988 SEQUOYAH TWO REACTOR HISTOGRAM 110 100 - 90 - 60 - l 3 70 - U 60 - i 50 - 0-s h 50 - r 20 - 10 - 0--c 0 -10 - e i i i i i i e i i i i i i i o 4 8 12 16 20 24 28 32 APRL 1988 -

a .) l a 6 1 1 i ( t i k i I I 9 I i .l i 1 i i

SUMMARY

OF MAINTENANCE ACTIVITIES l i i a i i / I I h 1 i \\ i t i l I i i 1 1 ) i t i I I I

~ MAINTENANCE

SUMMARY

(INSTRUMENTATION) l i l

4 i e f INSTRUMEMT MAIMTENANCE MONTHLY REPORT FOR APRIL 1988 COMMON 1. Completed corrective action for CAQR SQP880159. This involved training for the Instrument Maintenance employees on the proper use of "information only" drawings. 2. Completed remaining documentation requirements for Workplans (WP) 7343-02 (unit 1) and WP-7344-02 (units 0 and 2) that added seal-in rolsyr to various radiation monitors to reduce the effects of electromagnetic interference. This was needed to satisfy NRC commitments identified in CCTS-NC08602229003, NC0870362003, and NC0870362004. 3. Completed and issued the trending evaluations for component failure reports identified by the NPRDS, EQIS, and 10 CFR 50.49 databases. Initiated proposed corrective actions to resolve potential equipment reliability problems. UNIT 1 1. Completed a revision to SI-281 (functional test of RMs with automatic actuations) to incorporate quarterly functional testing which verifies the automatic isolation actuations of radiation monitors, 1-RM-90-120 AND 121. This was an NRC cosnitment item identifed in CCTS-NC0880080001. UNIT 2 1. Installed pulsation dampeners in the sense lines for the turbine-driven AFW pump pressure switches, 2-PS-3-121A, -1218, and -121D. This was needed to help prevent system pressure pulsations from continuosly actuating the switches and causing spurious alarms and damage to associated electrical relays. This work was authorised by TACF 88-2008 and will be made a permanent modification when ECN 7118 is implemented. l I l 2. Installed temporary RCS level instrumentation for use by the operators during draindown of the RCS for steam generator repair work. This involved installation of a transmitter to a sense line connected to the '1 RCS loop 1 hotleg. The level signal was routed to an esisting recorder in the main control room as a diverse means of ald-loop level monitoring. i l 3. Replaced defective RCS wide-range temperature detector, 2-TE-68-65, as a result of data collected during performance of SI-488 RCS RfD 1 cross-calibration. 4. Repaired nuclear instrumentation system sourca range detecor, N31, as needed to clear electrical grounds (referenco CAQR SQP880265). 1 I j I i -28

UNIT 2 (Cont.) 5. During heatup, Operations reported loop 1 steam flow indication (loop 2-F-1-3A) reading higher than other channels. Investigation revealed tv, problems: A. A section of the S/G insulation had split apart, which caused the transmitted sense line condensate pot to overheat and stop functioning properly. B. The condensate pot elevations were two inches different than assumed in the transmitter scaling. The insulation was repaired and the transmitter rescaled and re:alibrated appropriately. 3 i l l 1.

w h 4 e 9 l t l MAINTENANCE

SUMMARY

(MECHANICAL) t } 1 i

MECHANICAL MAINTENANCE MONTHLY REPORT FOR APRIL 1988 COMMON 1. Completed SI-181.1, "Fire Hose Hydrostatic Test (Annual)," on fire protection system. 2. Completed repair on control room emergency vent dampers 31-9 and -11. 3. Completed work on auxiliary boiler. UNIT 1 1. Completed repair on IB-B boric acid transfer pump. 2. Completed repair on la-A and 1B-B speed increaser. 3. Completed repair on 1-VLV-012-0713. 4. Installed terry turbine governor valve. 5. Rebuilt IB-B CCP vent valve (1-VLV-062-0513). 6. Completed SI-102 (Diesel Generator Inspection) on diesel generator IA-A. 7. Replaced diaphragm in 1-LCV-003-0164A. 8. Completed repair on various valves during RHR outage. 9. Completed SI-648, "Test of Door Play on Containment Personnel Airlocks," on airlock doors. 10. Installed new valves at locations 1-VLV-012-0776, -0501, and -0713. 11. Completed workplan on thermal barrier booster pump. l UNIT 2 1. Completed repairs on 2-FCV-003-0081. 2. Rebuilt 2-RM-090-0100 rad monitor pump. 3. Gagged /ungagged valses 2-FSV-068-0563 and -0564. 4. Completed repairs on 2-HEI-035-0094A. 5. Completed repair on 2A-A main feed pump governor speed changer. 6. Supported tube leak repair on unit 2 steam generator No. 3. I l,

MECHANICAL MAINTENANCE MONTHLY REPORT FOR APRIL 1988 UNIT 2 (Cont.) 7. Pulled, repaired, and replaced pressure safety valves 63-563, -564, and -565. 8. Replaced bearings in 2A-A MDAFP. 9. Rebuilt 2A-A boric acid transfer pump. 10. Supported uncoupling, realignment, and recoupling of RCP No. 1. 11. Completed SI-102 MM (D/G monthly mechanical inspection) on diesel generator 2A-A. 12. Rebuilt hydrogen side of seal oil pump. 13. Repaired various hangers on condensate system. 34 Recoupled RCP No. 2. 15. Completed repair on SIS test valves 63-24, -79, -121, -158. -165, -167, and -174. 16. Replaced trim in 2-FCV-1-114. 17. Completed repair on 2-FCV-1-18. 18. Cleaned ERCW check valves 67-562A and C. 19. Repacked main steam check valve 1-624. 20. Completed SI-648, "Test of Door Play on Containment Personnel ] Airlocks," on airlock doors. J 1 9 21. Closed unit 2 steam generator manways. 22. Completed work on "A" boric acid evaporator, i 23. Completed workplan on thermal barrier booster pumps. OTHER 1. Continued closure of various CAQRs, CARS, DRs, etc. l

s.- 9 8 0 e MAINTENANCE

SUMMARY

(MODIFICATIONS)

n b l

SUMMARY

OF WORK COMPLETED HODIFICATIONS - CURRENT STATUS APRIL 1988 ) l

, e Major Capital Projects: PN7101: ECW 6388 - 500-kVA Switchyard Current Transformer Heaters Workplan (WP) 12223 is in progress pending safe access to current transformers. PN7102: ECM 5938 - Replace Feedwater Heaters 3 and 4 No work in progress at this time. PN7105: ECM 5009 - Essential Raw Cooling Water (ERCW) Piping Changeout From Carbon Steel to Stainless Steel Insulation of new stainless steel piping has been completed. No additional pipe replacement is scheduled in the near future. PN1108: ECW 6720 - Crane Consistency Proaram Unit 2 polar crane modification is complete. Postmodification testing (PNT) by Electrical Maintenance is not complete. Unit 1 polar crane work started on Ju:

• 6, 1987 and is approximately 98 percent complete.

Pa. .ing of blocks and limit switch weights remain to be completed. Auxiliary Building crane WP 12596 is in the approval cycle. Drawings have not been issued on the remaining cranes. WP 12596 is in nonwork status. Work is stopped pending resolution of budgetary problems. PN7115: ECW 6719 - Volumetric Intrusion Detection System ECM 6719 is still on hold. Security is dissatisfied with system operation. Security and DME have not concluded discussions on resolution. Design is working on making the system functional. Lighting is not finished. Workplan change for drainage is being written. Field Change Request (FCR) 6645 for DNE changes is approved. Work is stopped pending resolution of budgetary problems. PN7122: DCR 1373 Secondary Side Steam Generator Peccervation ECW 5657 - Installation of Moisture Separator Reheater (MSR) Drain Valves Work is essentially complete on all the ECNs. Some minor insulation installation activities are in progress as resources permit. _. -

Major capital Projects (cont.): ECW 5841 - Hot Shop Fire Protection / Evacuation Alarm - WP 12637 All fieldwork for evacuation alarm is complate. Awaiting Work Request (WR) B240406 to be worked to restart fans to do functional test on fire protection. PN7123: ECWs 5938, 6305. 6571 - Replace Feedwater Heaters 1 and 2. Units 1 and 2 and Eroded Pipe Modifications Group B is supporting PNT on heaters 1 and 2 on both units as requested. Insulation work on unit I continues. PN1130: DCR 1156 - Post Accident Monitoring This work is now scheduled for unit 1 by unit I cycle 4 (U1CA) and unit 2 by unit 2 cycle 4 (U2C4). PN7132: DCN 0026 - Sewage Treatment Facility and Civil Upgrade city of Soddy-Daisy will be responsible for interface work on site, lift station, and pipeline between Sequoyah Nuclear Plant and Highway 27. WP 0026-01 is in approval cycle. DCW is being revised to include the Solar Building discharge { effluent to Soddy-Daisy interface. t PN7136: ECW 6259 - MSR Tube Bundle Replacement j ECW is complete except for PNT and inservice leak test. Leak checks will be performed during system heatup. PN7161: ECW 5855 - Replacement of Doors A56 and A57 I Functional testing is in process. PN7181: DCR 1898 - ECWs 6832 and 6596 - Dry Active Waste (DAW) Building Electrical interface work is complete. Workplan closure held for Instrument Maintenance (WP 12478) and Electrical Maintenance (WP 12612) checks and update of SOIs by Operations.

. e Major Capital Projects (cont.): PN7199: Miscellaneous Activities Under $100.000 This is for various work orders prepared for work under $100,000 total site cost. This work is done as manpower resources are available that will not impact unit 2 restart effort. Sinnificant Items: The following major significant items are under Operation and Maintenance Expense (O&M) and will continue to be until restart of unit 2. 1. Fuse Program Work ECW 5880 - Work complete for class IE unit 1 and 2 and non-IE fuses associated with unit 2 restart. Remaining work is non-class IE unit 1 and non-class It unit 2. WP 5880-02 is being written for non-IE checklist for unit I and is scheduled to be in werk May 13, 1988. 2. Cable Ampacity Program Unit 1 Cable Ampacity Program is complete.

i 1 r o Other Items: ) ECW 5111 - Provide Permanent Power to Manholes 42-46 Electrical conduit and wiring installation for manholes 44-46 is 1 complete according to WP 12262. Manholes 42 and 43 are being held for j information from DNE on power cable routing and terminations. FCR 4572R1 was sent to DNE and was logged in, but it was.not sent to the responsible engineer. Manholes 42-46 will be worked by FCR 4572R1. This item is not considered unit 2 restart by DNE, and they will not authorize the engineer to do the necessary work to obtain pull cards or drawings for manholes 42-46. FCR 4572R2 will be written to supersede 4572R1 for conduit and cable routing for manholes 42-46. ECN 5435 - Fire Doors No work is in progress at this time. ECN 5503 - Evacuation Alarms O&PS/ Fire Detection O&PS No work in progress at this time. WP 12664 - Work is in progress and is approximately 95 percent complete. Remainder restrained by CAQR SQP 871616 (evacuation alarm O&PS). ECW 5552 - Condensate Domineralizer Modifications and High Crud Filter Upgrade to higher range instrumentation for condensate domineralizer system neutralization and nonreclaimable waste pumps. A mode 2 restart item. WP 5552 Work is complete. WP 5552 Install flow recorder, on hold for material. ECN 5609 - Evacuation Alarm / Fire Detection Valve 26-290 WP 12387 is in work and is 90 percent complete. Workplan change is needed to allow local testing of alarm horn. ECW 5609 - Alteration to the Makeup Water Treatment Plant WP 12576 - Work is in process and approximately 95 percent complete. 1 ; j

l e ,2 e Other Items (cont.): WP 12633 - Work is in process and approximately 75 percent complete. WP is being held for material on WP 12665 (Mechanical). I WP 12731 - WP is in work. l WP 12684 - WP in work, approximately 70 percent complete. WP 12665 - WP is being worked, and is 30 percent complete. WP 12682 - WP is being worked, and is 30 percent complete. ECW 5626 - Containment Ladders, Unit 1 A workplan is being written to incorporate this ECN. Modiftsacions j needs additional design information to complete. DNE needs to issue all drawings listed on this ECN. Work has not begun because of this holdup. ECW 5726 - CAQR SQP870478 - Reroute Sense Line Piping for Train Separation to Panels 2-L-58 and 2-L-68 Fieldwork is complete. In-service leak check will be performed during heatup. ECW 5841 - Hot Shop Fire Detection WP 12360 is 98 percent complete. ECW 5935 - Correct Power Block Lighting Defielencies WP 12437 has been approved and is complete. WP 12275 is written. Modifications needs DNE to provide light mounting detail for refueling water storage tanks and estimate design date after unit 2 restart. WP 5935-01 has been written to install security grills and gratings. All I work is currently held for manpower. j i ECW 6005 - Replacement of AFW Steam Ejector, Unit 2 In-service leak check performed and reinstallation of piping insulation is complete. ECN 6057 - Cable Tray Covers This activity is complete for the unit 2 restart. The only remaining work is the unit 1 annulus and this will be completed prior to June 15, 1988. - -

Other Items (cont.): ECN 6082 - Vent for Boric Acid Tank (BAT) Work complete except final inspections and painting on duct supports and performance of SI on High Efficiency Particulate Air. Work is nearing completion. ECN 6196 - Pressurizer Hangers and Valves pMT is scheduled for unit I restart. Remaining unit 2 work is scheduled for U2C3 refueling outage. ECN 6205 - Replacement of Instrument Loop Power Supply Fuses Unit 2 and unit 0 work is complete except for spares. WP 12447 has been turned in to Document Control for partial completion. WP 12611 initiated to perform unit I work. Unit I work is progressing. ECW 6357 - Essential Raw Cooling Water (ERCW) Roof Access and Rails for Security Equipment WP 12238 is in work and is approximately 20 percent complete. ECN 6380 - Replacement of Barton Pressure Transmitters Unit 2 fieldwork is complete. Final closure awaiting instruction revision by Instrument Maintenance. ECW 6388 - Hydrogen Monitors in Switchyard l Workplaa 12223 has insufficient craft support available for installation of hydrogen analyzers at 500-kV switchyard and is i 75 percent complete. ECNs 6402 (Unit 1) and 6439 (Units 1 and 2) - Pressurizer Instrumentation Relocation Inservice tests will be performed when the system is brought up to temperature and pressure. ECN 6429 - Component Cooling Heat Exchanger B Replacement DNE to procure piping, hanger, and plate heat exchanger frame i material. DNE to issue hanger drawings. Workplans are in work to fabricate and wald piping and fabricate and install new plate heat exchanger. Estimated completion date for outage is during the month of June 1988. - _ ~ _ - _,

y \\ Other Items (cont.): ECM 6455 - Upgrade CU-3 Box Battery Packs WP 12295 has been issued. Modifications are complete for all CU-3 boxes. Site Security still needs to perform some PMT before the workplan can be closed out. ECNs 6491 and 6534,- Essential Rav Cooling Water (ERCW) Supports All restart work is complete. A workplan to remove previously installed piping was approved and will be worked as resources permit. ECW 6543 - Install Public Safety Access Portals and Modify Entrance Road Work is being held pending the release of drawings from DNE. ECW 6601 - Removal of Unit 1 Emergency Gas Treatment System (EGTS) Backdraft Dampers PMT remains to be completed by the Mechanical Test Section. Fieldwork is complete. ECW 6610 - Modify Air Return Fan Supports Unit I work is incomplete. ECW 6631 - Modify Snubbers Unit 2 work is complete. ECW 6689 - Relocation of Main Steam Power Operated Relief Valves (PORV) All work is complete for unit 2. Work on unit I has started. ECW 6698 - Repull 120-Volt Cables Unit 2 is complete. Unit 1 is 90 percent complete and held by ECW 6742. This ECN should be completed prior to unit I restart. ECN 6706 14 Support Enhancement / Lost Calculations Repairs continue on unit 1. This project has been combined with the calculation regeneration project for unit 1. Unit 2 work is complete and workplans are closed. There are 25 modifications in work with 46 having been completed. In addition, 82 maintenance items are in 1 i progress with 293 complete. ;

e 'O <A other Items (cont.): ECN 6739 - Alternate Ana'ysis All unit 1 modification work has been completed. Review and closure has begun. Work is in progress on the maintenance itams. ECH 6742 - Install Fuses in Radiation Monitor Power Supply Circuits The workplan is complete for unit 2. Unit 1 is in work. ECN 6761 - East Valve Room (EVP.) Blowout Panels Implementation of unit 1 work is progressing as resources are available. Unit 2 work is complete. ECW 6784 - Documentation to Show Pipe Class Breaks Final closure is awaiting a revision to AI-19 deleting requirements to mark shif t supervisor drawings. ECN 6815 - Installation Power Circuit Breaker Install 500-kV power circuit breaker and associated equipment for bay 1. Retire 161-kV PCB and associated equipment. A total of eight workplans will be required. Foundations and conduit installation are complete. WP 12740 for lighting, drain pipe, and surface ground mat is in work. WP 12739 for the structural steel installation is in work. WP 6815-02 has been written and is in work. WP 6815-31 is in work. ECW 6860 - Control Room Bullet Resistivity - DCR 2268 - ECW 6860 WPs 12602 and 12604 are field complete. WP 12603 is held for disposition of CAQR 880183. WPs 12605 and 6860-01 are held for material on Contract 74014A. Essentially all work is complete except for pull handles, replacement of door closures, rework of one lockset and replacement of one electric hinge. Estimated date of completion is June 30, 1988. 7 ECW 7000 - Addition of Support to Main Steam Dump Header Fieldwork is complete. Workplan 12217 is closed. l - -

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g e TENNESSEE VALLEY AUTHORITY Sequoyah Nuclear Plant P. O. Box 2000 Soddy-Daisy, Tennessee 37379 May 12, 1988 Nuclear Regulatory Commission Office of Management Information and Program Control Washington, DC 20555 Gentlemen: Enclosed is the April 1988 Monthly Operating Report to NRC for Sequoyah Nuclear Plant. Very truly yours, TENNESSEE VALLEY AUTHORITY S J. Smith Plant Manager Enclosure cc (Enclosure): Director, Region II INPO Records Center Nuclear Regulatory Conunission Suite 1500 Office of Inspection and Enforcement 1100 Circle 75 Parkway Suite 3100 Atlanta, Georgia 30339 (1 copy) 101 Marietta Street Atlanta, Georgia 30323 (1 copy) Mr. K. M. Jenison Resident NRC Inspector Director, Office of Inspection O&PS-2, Sequoyah Nuclear Plant and Enforcement Nuclear Regulatory Commission Washington, DC 20555 (12 copies) Director Office of Special Projects 4350 East West Highway EW 322 Bethesda, Maryland 10814 (10 copies) Mr. T. Marston Electric Power Research Institute P. O. Box 10412 Palo Alto, California 94304 (1 copy) i\\ An Equal Opportunity Employer _. _ - -}}