ML20246N727
| ML20246N727 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/30/1989 |
| From: | Dupree D, Michael Ray, Shawn Smith TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8905190519 | |
| Download: ML20246N727 (68) | |
Text
. -___ _ _ -
d- ,
TENNESSEE VALLEY AUTHORITY NUCLEAR POWER GROUP SEQUOYAH NUCLEAR PLANT MONTHLY OPERATING REPORT TO THE NUCLEAR REGULATORY COMMISSION APRIL 1989 UNIT 1 DOCKET NUMBER 50-327 LICENSE NUMBER DPR-77 UNIT 2 DOCKET NUMBER 50-328 LICENSE NUMBER DPR-79 g/ f
==?L4 /,^/t>
Submitted by: '.T S. J., Smith, Plant Manager f(Tf b
-7 .
i
, TABLE OF CONTENTS Page i I. Operational Summary Performance Summary 1 Significant Operational Events 2-8 Fuel Performance 9 Spent Fuel Storage Capabilities 9 PORVs and Safety Valves Summary 9 Special Reports 9
, Licensee Events 10-14 i Radwaste Summary 15 Offsite Dose Calculation Manual Changes 15 II. Operating Statistics A. NRC Reports Unit 1 Statistics 16-18 Unit 2 Statistics 19-21 B. TVA Reports Nuclear Plant Operating Statistics 22 Unit 1 Outage and Availability 23 Unit 2 Outage and Availability 24 Unit 1 Reactor Histogram / Analysis 25-26 Unit 2 Reactor Histogram / Analysis 27-28 III. Maintenance Summary Maintenance 29-46 Modifications 47-50 IV. Glossary Common Abbreviations and Systems of Sequoyah Nuclear Plant 51-54 Operational Modes 55 0288f
- - - _ _ _ _ _ _ - - - _ - _ _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ - l
\ :~ .q e
)
\.
OPERATIONAL Su}uiARY
___ _ _ _ - - - - - ~ ' - ^ ^ - . _ _ _ - - --
'k
- i. t' i.
PERFORMANCE SUMHARY I
. April 1989
.i A. .
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 inciuded in this report.
Unit 1 Unit 1. operated the entire month at approximately 100 percent power.
It generated 842,260 tIWh of electrical power with a capacity factor of 99.02 percent.-
service. Not since This marks a milestone for the unit for reliable day-to-day' March 1985 has the unit accomplished a capacity of 99 percent or greater (unit 1 - 99.9 percent, March 1985, and unit 2 -
99.3 percent, April 1985). This is equivalent to generating 1.171 MWh every. hour throughout the month of April.
The unit has been in continuous operation for 77 days as of April 30, 1989.
Since restart on November 10, 1988,- a total of' 3,315,110 MWh has been produced.
Unit 2 On April 11, 1989, at 2024 (EDT), the unit was taken critical. On April 14, 1989, at 1408 (EDT), unit 2 began generating electrical power, signaling the completion of the cycle 3 refueling / modification outage.
The outage duration was approximately 87 days. A total of 17,610 MWh was
, generated for the month, comprising a capacity factor of 2.07 percent.
Three reactor trips occurred during unit restart because of erratic behavior of the feedwater controls and associated equipment. This accounts for the small amount of electrical generation and capacity. Unit 2 returned to service on April 26, 1989, at 1627 (EST).
The unit continued in service for the remainder a turbine overspeed test. of the month, except for a small portion of time dedicated for I
J
_ _ _ _ - _ - - - - - - - - - - - - - - - ^' ~ ~ a
7..
"t' STGt1IFTCANT OPERATIONAL. EVENTS Unit 1 Date Time (EST) Event
~
04/01/89 0001' 109 percent reactor power, 1,185 MWe.'
0714 RPI H-8 is inoperable.
1356 Control rod H-8 RPI'D bank operable after.ma4ntenance. !
1958 B-2 intermediate string heaters isolated on high-high.
level.
1959 Reduced turbine load because of secondary-side flow swings and decreasing S/G levels. Reactor at L
100 percent, 1,180 MW e.
2016 Began draining heater B-2 in preparation for placing back in service.
2029 Began increasing turbine load.
04/02/89 .0126 Began placing' intermediate pressure heater B string.in service.
0358 B string of intermediate pressure heaters are back in service.
0500 Reactor at 100 percent, 1,188 MWe.
04/03/89 0151 RPI B-4 is reading approximately 18 steps different
'from its group demand steps counter, declared inoperable.
1130 RPI B-4 rod is operable after maintenance and PMT were performed.
1534 Reactor at 100 percent, 1,184 MWe .
04/15/89 1444 RPI E-13 on S/D bank D inoperable.
1532 RPI E-13 operable, but being monitored.
04/24/89 0259 Isolate waterbox IB-2 for maintenance. Reactor at 99 percent, 1,175 MW .
e 1230 Waterbox 1B-2 back in service.
1528 Reactor at 100 percent, 1,180 MWe.
04/26/84 0117 Diluted RCS to increase Tavg.
I SIGNIFICANT OPERATIONM. EVENTS Unit 1 Date Time (EST) Event 04/27/89 1625 RPI E-13 inoperable, greater than 12 steps out.
j 04/28/89 1030 RPI E-13 remains inoperable. Detection circuitry -
problems.
04/29/89 0038 RPI L-13 inoperable. Indicator showing rod at the bottom of the core.
0517 Maintenance complete on RPI L-13. RPI operabic.
1336 RPI F-14 inoperable. Indicator showing rod at the .
bottom of the core.
1819- RPI F-14 operable after maintenance.
04/30/89 2400 Reactor at 100 percent, 1,372 MWe, Unit 2 04/01/89 0001 Mode 5, 178.10F at 350 lb/in2 Activities continue for start-up.
04/05/89 0415 Began heatup in preparation for mode 4 entry.
0751 Entered mode 4 0915 Mode 4, RCS at 2220F, 365 lb/in2 ,
2038 Started RCS heatup to approximately 3350F.
2225 RCS temperature at approximately 3350F.
04/06/89 1630 Entered me's 3.
1652 3560F, pressure 550 lb/in2, 2319 Terminated heatup, holding RCS temperature at 4500 F for performance of RCS RTD cross calibration verification.
04/07/89 1300 Began low-power physics testing.
04/08/89 1303 Resume temperature increase.
04/09/89 0117 Began pulling S/D bank A.
04/10/89 0708 RPI F-2 on control bank D inoperable.
I
t ,
SIGNIFICANT OPERATIONAL EVEUTS Unit 2 Date Time (EST) Event 04/10/89 1338 - RPI F-2 returned to normal after maintenance.
(cont.) '
1525 Began pulling S/D rods.
1540 Stopped pulling S/D bank A from core, RP1 0-2 indicating approximately 20 steps below other RPIs.
Maintenance initiated to repair RPI.
2039 Began pulling S/D rods in preparation for mode 2.
2043 Shutdown bank A pulled.
2052 Shutdown bank B pulled.
2058 Shutdown bank C pulled. ;
04/11/89 0019 RCS 5450F at 2,335 lb/in2, 1537 All S/D banks fully withdrawn.
1545 Entered mode 2.
1651 Began RCS dilution at various times for criticality.
2024 Unit 2 reactor critical.
04/12/89 0100 Low-power physics testing in progress. 1 I
0403 RPI B-12 inoperable.
l 0855 Low-power physics testing in progress.
l 1600 Shutdown rod bank A. RPI B-12 operable after maintenance, but RPI is being monitored.
04/13/89 0330 Low-power physics testing complete.
1024 Reactor power at 1 percent. !
j 1250 Entered mode 1, reactor at 5 percent and increasing. I 1500 Reactor at 20 percent power.
1646 Reactor at 20 percent power. Iloiding power for turbine balancing.
1847 Holding status because of excessive RCS leakage.
l E_________________-------_-------------------- - - - - - - - - - - - - - - - - - - - - - - - - - - --
s a SIGNIFICANT OPERATIONAL EVENTS Unit 2 Date Time (EST) Event 04/13/89 2121 St,arted boration for controlled power reduction to (cont.) B percent to investigate RCS leakage, power level currently 18 percent.
04/14/89 0029 Reactor at 8 percent power.
0210 RCS leakage found and resolved.
0345 Started dilution of RCS for power increase.
0515 Completed dilution of RCS.
0943 Started power increase to 20 percent.
1213 Reactor at 23 percent.
1227 Rolling main turbine.
1408 Unit 2 main generator online.
1620 Began power increase to 30 percent.
1847 Reactor at 30 percent, 284 MWo .
2211 Started turbine load decrease. Maintaining reactor power at 30 percent for turbine overspeed test.
2344 Turbine taken offline for overspeed test.
04/15/89 0009 Reactor trip on S/C low-low level, loop 4.
0011 Started emergency boration, T avg less than 5400F.
0749 Mode 3, RCS at 5470F, 2,230 lb/in2 1140 Began RCS dilution for criticality.
1457 Completed dilution.
1635 All S/D banks fully withdrawn.
2159 Unit entered mode 2, began control rod withdrawal.
2240 RP1 H-12 out greater than 12 steps, declared inoperable. Initiated maintenance for repair.
_ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - - - - - - - - - - - - i
SICMfFICANT OPERATTOMAL EVENTS Unit 2 Date Time (EST) Event 1 l
04/15/89 2301 Unit 2 reactor is critical. '
(cont.) ~
2321 Reactor at 1 percent power.
04/16/89 0040 Entered mode 1, 5 percent power.
0048 Reactor trip, began emergency boration. Tavg 1888 than 5400F, 1
i 0057 Terminate emergency boration. 1 0126 T avg increasing to 5470F. l 0713 Maintenance on various RPIs begins.
1941 initiated RCS cooldown to 4000F.
2307 Terminated cooldown at 4500F. l 04/17/89 0817 Mode 3, 4520F.
1030 Began heatup.
i 1736 RCS at 4980F, 2,085 lb/in2, 04/18/89 0109 Began pulling S/D banks.
0148 Entered mode 2.
0341 Reactor critical.
04/19/89 0058 Reactor power increase started.
0145 Entered mode 1, 5 percent power.
0444 Turbine trip because of high-high icvel in S/G 3.
Reactor at IS percent power.
0447 Reactor trip, low-low S/Gs 1 and 2.
0451 Began emergency boration. Tavg is 5400F.
0456 Terminated emergency boration.
0659 Began pulling S/D banks.
0714 Shutdown bank withdrawal completed.
i o .
SIGNYFTCANT OPERATIONAL EVENTS Unit 2 Date Time (EST) Event i.l 04/20/89 0020 RCS temperature-5470F, pressurizer 2,230 lb/in2, 04/24/89 2214 Began dilution of RCS.
2228 Completed dilution 04/25/89 0750 Mode 3, 5470F.
2323 Verified all S/D rods are fully withdrawn.
2325 Entered mode 2.
04/26/89 0009 Unit 2 reactor critical.
0755 Mode 2, 547.40F. Maintenance on secondary equipment in progress.
1934 Reactor at 2 percent power. Maintenance continues on i feedwater system.
04/27/89 0116 Started reactor power increase.
0423 Entered mode 1, 5 percent power.
0800 Began diluting RCS at various intervals for power ascension. Reactor at 10 percent power.
1446 Rolling main turbine.
1627 online.
1900 Reactor at 15 percent power, 105 MWe.
04/28/89 0130 Began power increase to 20 percent.
0431 Reactor power at 10 percent, 164 MWe .
0550 Reactor power at 23 percent, and holding for preconditioning, 184 MWe .
1905 RPI K-14 is inoperable, out greater than 12 steps.
2009
.. Removing main generator from service for overspeed test.
2324 Offline.
i
~7-
8 SICMIFICANT OPERATIONAL EVENTS Unit 2 Date Tit'.ie (EST) Evmit 04/29/89 0056 Test comalete, generator online.
0220 RPI K-14 operable after maintenance, but RPI is being monitored.
0757 Mode 1, 23 percent power, 190 MW e . llolding power for fuel conditioning and thermal power verification.
04/30/89 1700 Began power ascension.
2400 Reactor at 37 percent power. 338 MWo . Power ascension in progress.
l FUEL PERFORMANCE Unit 1 The core avera6e fuel exposure accumulated during April was 1,144 mwd /MTU, with a total accumulated core average fuel exposure of 4,644 mwd /MTU.
Unit 2 -
The core average fuel exposure accumulated during April was 46 mwd /MTU, with a total accumulated core average fuel exposure of 46 mwd /MTU.
SPENT FUEL PIT STORAGE CAPABILITIES The total storage capability in the SFP is 1,386 bundles. However, there are six cell locations that are incapable of storing spent fuel. Four locations (A10, All, A24, and A25) are unavailable because of a suction-strainer conflict, and two locations (A16 and A21) are unavailable because of an instrumentation conflict. Presently, there is a total of 428 spent-fuel bundles stored in the SFP. The remaining storage capacity is 952 bundles.
PORVs AND SAFETY VALVES
SUMMARY
No PORVs or safety valves were challenged during the month of April.
SPECIAL REPORTS There were no special reports for the month of April.
i l
l _
LERs
'The following LERs'were transmitted to the Nuclear Regulatory Commission in April 1989.
~
Description of Event LER 1-89007 On March 19, 1989, with unit 1 in mode 1 (100 percent power) and unit 2 in mode 5, an A-train CRI occurred. At approximately 1125 (EST), an A-train CRI signal, as indicated on MCR panel 0-M-27B (window 20), was received in the control room. The unit 1 ASOS responded to the CRI alarm. The ASOS had knowledge that maintenance was being performed on Control Building fresh air intake duct smoke detectors 0-XS-31A-3 and 0-XS-31A-4, and could have initiated the CRI. Subsequently, the ASOS suspended the maintenance activity on the smoke detectors. The ASOS verified that no other conditions were present, such as high radiation or a safety injection signal, and initiated realignment of the control room ventilation system to normal operation in accordance with S0I-30.18 " Control Building and Control Room Heating, Air Conditioning and Ventilation System." The Control Building ventilation system was returned to normal operation by 1310 (EST). Further investigation of the event revealed that before maintenance on the smoke detectors was authorized, the smoke detector circuit was doenergized by opening breaker No. 7 on 120-V ac preferred power rack 1-M-7, located in the control room. However, the 120-V ac power supply to an electrical interlock on the smoke-detector circuit was not deenergized.
Subsequently, while reterminating wires on the smoke detector (0-XS-31A-3, A-train), terminals 7 and 8 were accidently shorted.
This energized relay SDA3 and completed the required logic for the CRI. To prevent recurrence, thi.s event will be reviewed with Electcical and Instrument Maintenance engineers, WR planners, appropriate craft personnel, and Work Control Croup personnel to emphasize the need to doenergire olcetrical circuits while performing maintenance on electrical equipment.
1-89008 on March 20, 1989, at 1450 (EST), with unit 1 in mode 1 (100 percent power) and unit 2 in mode 5, both trains of CREVS were declared inoperable; and it was discovered that LCO 3.0.3 had been inadvertently entered earlier in the day at 0820 (EST), as a result of tornado dampers that isolate fresh air intake to the MCR having been closed. The dampers were closed to support replacement of smoke detectors. This condition could result in the loss of suction-flow path from the outside atmosphere to the Control Building emergency pressurizing fans if required upon receipt of an accident signal.
Loss of suction to the pressurization fans could preclude the system from performing its design function to maintain the MCR habitability area at greater than or equal to 0.125-inch positive static pressure during accident periods.
The cause of this event is attributed to an incomplete evaluation of the effect of closing the tornado dampers by a licensed operator. Upon discovery of this condition, immediate corrective actions were to open the tornado dampers and exit LCO 3.0.3. Both trains of CREVS were returned to operable status
Description of Event l LER l 1-89008 at 1451 (EST), on March 20, 1989. In addition to the disciplinary (cont.) action taken, long-term corrective actions include reviewing this event with Operations personnel; emphasizing the need to do a thorough review of all available drawings and information before deciding to operate equipment; replacing an ineffective placard on containment / auxiliary vent board 1A1 with one that more adequately i details the consequences of closing the tornado dampers; and revising SOI-30.7, "On Site Electrical power Systems Board Rooms lleating, l Venting, Cooling," and SOI-30.1, " Control Building and Control Room Heating, Air Conditioning and Ventilation Systems," to include j warnings about effects on the CREVS when tornado dampers are closed. '
1-89009 On March 19, 1989, at approximately 0600 (EST), with unit 1 in mode 1 (100 percent power) and unit 2 in mode 5, it was discovered that the switch on the local control panel for the CO 2 fire-suppression system that protects the computer room was in the OFF position, thus making the system inoperable. The Fire Operations Shift Supervisor was immediately notified, and a fire operator was dispatched to investigate the condition. A check to determine if the system should be inoperable revealed that no log entries were made'to remove the CO2 system for the computer room from service, and the system was immediately returned to operable status. Subsequent to returning the system to normal, an investigation ensued to determine how the switch was left in the.0FF position. It was revealed by discussion with one of the fire operators that he had isolated the CO2 system for the unit 2 auxiliary instrument room on March 18, 1989, to facilitate work activities. Upon arriving at the work area, the fire operator recalled that he had opened the door for the control panel for the CO2 system that protects the unit 2 auxiliary instrument room, as well as the adjacent control panel for the computer room, because of a lack of identification external to the panels. The control panels for the CO2 systems are controlled to allow limited access.
Because of the controls in place to limit access, and the fact that the fire operator recalled opening the CO2 control panel for the computer room, it is concluded that the fire operator inadvertently placed, and left, the switch in the OFF position. A contributing cause of this event is attributed to the lack of identification tags on the control panels that apparently confused the fire operator as to which control panel corresponded to the unit 2 auxiliary instrument room. To prevent recurrence, this event will be reviewed with the fire operators to provide a lesson Icarned on how inattention to detail can lead to rendering a system inoperable.
Additionally, identification tags will be placed on the outside of the CO2 control panels.
1-89G10 on March 29, 1989, at 1830 (EST), with unit 1 in mode 1 (100 percent I power) and unit 2 in mode 5, a review of SI packages determined that l SI-307.2, " Degraded Voltage Relay Response Time Test and Timor l Calibration," was out of its 18-month TS required frequency. This instruction satisfies SR 4.3.2.3 by performing a response-time test of the 6900-V shutdown board 2A-A and 2B-B degraded voltage timers i
4 6 Description of Event LER , ,
1-89010 and relays. Two auxiliary relays (DS and DV) associated with (cont.) degraded voltage logic on 6900-V shutdown board 2B-B had not been tested within frequency. The scheduled performance date of SI-307.2 had been, based on an incomplete package performed on November 10, 1987, by the EM Group. The November package had been incorrectly designated as a basis for scheduling the next performance of the SI and led to the auxiliary relays not being tested within the required frequency.
Unit 1 entered the action statements of TS LCOs 3.3.2.1 and 3.8.2.1 at 1830 (EST). Unit 2 was in mode 5; therefore, these LCOs were not applicable. At 2130 (EST), SI-307.2 was initiated to test the 6900-v shutdown board 2B-B degraded voltage relays and was successfully completed with no deficiencies at 0030 (EST), on March 30, 1989. As a corrective measure, the EM supervisor discussed this event with personnel involved in reviewing the SI package. As an enhancement, SI-1, " Surveillance Program," will be revised to updato all SI data package coversheets to include the scheduling questions in the text.
The person who signs for review and approval of the test package will verify the scheduling questions were correctly answered.
1-88033 This revision provided additional information concerning the (Rev. 2) corrective action taken by TVA on the generation of an unplanned reactor trip signal because of an instrument failure during the performance of SI-94.22, " Channel Calibration of delta T/Tavg-Channel II, Rack 6 (T-68-25)," on October 4, 1988. Immediate corrective action was to initiate a WR to diagnose the erratic behavior of 1-TI-68-2A, Troubleshooting concluded 1-TM-68-2E was unreliable; and a similar instrument was procured, calibrated, and installed to allow a railure analysis on the removed module.
Based on the inconclusiveness of the Foxboro failure analysis and TVA's inhouse investigation, there is no further action associated with this event recommended at this time.
1-89004 This revision provided NRC with an update of the causes and corrective (Rev. 1) actions taken as a result of this event.
2-89002 On March 25, 1989, at 1059 (EST) ard 1549 (EST), with unit 2 in modt $, two separato unplanned reactor-trip signals were generated from the safety-injection logic during the performance of IMI-99 RT-601A, " Response Time Test For Turbine Trip." Before these events, on March 23, 1989, the A-train SSPS had been removed from service j
and the input error inhibit switch located on the SSPS logic test panel was placed in the INHIBIT position, disabling inputs fecm the relay contacts to the logic; however, this condition does not block the permissives.
Also, the modo selector switch was placed in TEST, which unblocked several of the at-power trips and ESF functions. On l
I March 25, 1989, performance of IMI-99 RT-601A was commenced, and the input error inhibit switch was returned to NORMAL. The blocks for the at-power trips and ESF actuations had not been reinstated; and, 4
I 1
i s .
Description of Event I LER {
1 i
2-89002 since one of the ESF signals results from low-pressurizer pressure, (cont.) an SI signal was generated, resulting in an unplanned reactor trip i
at 1059 (EST), A second reactor trip was generated at 1549 (EST),
by-an SI signal from high-steam flow coincident with low-low T avg-Investigation rev,ealed that one flow transmitter had been removed from service for maintenance and a second flow transmitter drifted high. With the RCS temperature well below the setpoint for low-low T avg, the coincident logic necessary for an SI signal was complete.
Neither event resulted in an SI because the SI system had been adequately removed from service. The cause of the first reactor trip was a deficient procedure, because the blocks for the reactor trips and ESF functions were not reinstated. The cause of the second event was not promptly addressing the generic ramification of the condition that caused the first event. To prevent recurrence, IMI-99 RTI-601A will be revised to reinstate blocks before returning the input error inhibit switch to NORMAL. Additionally, other response time verification tests associated with SSPS will be reviewed to ensure that this condition does not exist.
2-89003 On April 1, 1989, at 0150 (EST), Operations personnel noted an unusual odor adjacent to the 120-V ac Class IE vital invertor 2-IV. Initial investigation found a smoking capacitor, and the unit 1 SRO was immediately notified of t ae situation. The SRO was directed to transfer the power supply for 120-V ac vital instrument power board 2-IV from the inverter (normal supply) to the alternate / maintenance power supply in accordance with SOI-250, " Low Voltage AC/DC Electrical Systems." The alternate power supply for board 2-IV is 120-V ac instrument power distribution panel 2B. When attempting the transfer, the SRO discovered the inverter output frequency would not synchronize with the distribution board 2B output frequency by the absence of a sync light on vital board 2-IV.
At 0246 (EST), the inverter ac output breaker was opened to perform the transfer and resulted in a loss of power to board 2-IV.
While attempting to deenergize the inverter, popping sounds from inside the inverter were heard, and a significant amount of smoke began emitting from the inverter. During the review of the MCR boards for loss of equipment, it was discovered that a CVI had occurred; and RHR suction valve 2-FCV-74-2 had closed, which isolated i RHR flow. Immediately, RHR train 2A pump was secured. At 0256 (ES1),
~
vital board 2-IV was energized from distribution panel 2B. This caused 2-FCV-74-2 to 1sopen, and RHR train A pump was started, reestablishing RHR flow for unit 2. The root cause of this event was the inability to manually transfer the power supply for vital instrument board 2-IV from the normal supply to the alternate / maintenance supply without momentarily losing power to the board. As immediate corrective actions, the unit 2 vital inverters were verified to synchronize with the vital instrument board alternate supply; and Operations will perform by May 31, 1989, the appropriate portions of SOI-250 to verify that the synchronization signal, necessary for inverter j transfer, is present. '
~13-L_-_-_-__------
- v. c .
l Description of Event-
-LER h
2-88041 This' revision provides information on the' completed' corrective action (Rev. 1)'of.this. event, which was a result of insufficient seismic qualification of instrument cabinets containing. class 1E devices.
On December 9, 1988, Sequoyah Site NE personnel identified that as-installed instrument cabinets supplied by the Bailey Meter Company did not meet seismic qualification requirements and issued CAQR SQP880601. The. seismic qualification of the cabinets requires that two restraint bars be installed on each row of instrument modules:
in the cabinetc.- Instead, as-installed cabinets had only one restraint bar installed on each row of instrument modules. In November 1979, the Bailey Meter Company notified TVA that two restraint. bars are necessary to effectively restrain instrument modules in the event of a seismics occurrence. <This modification was not implemented in.12 instrument-cabinets.
To ensure the seismic qualification, a temporary alteration installed a 3/64-inch diameter aircraft cable across the front face of the instrument modules. The aircraft cable provided a stable and adequate restraint <until permanent seismic-restraints were installed.
TVA'has now installed permanent seismic restraint bars in the affected instrument ' cabinets as specified by DCN M01020A for unit 1 on February. 12, 1989, and'DN M01021A for unit 2 on March 23,.1989.
Aircraft cable across'the front face'of the instrument modules, "which was installed as.immediate corrective action, was removed upon installation of the permanent seismic restraint bars.
?-
l l
t
'O A i
RADWASTE
SUMMARY
l' April 1989
- 1. Total volume of solid waste shipped offsite:
A. Dry active waste: 915.2 ft3 Activity: 8.088 curies B. Spent resins, sludges, bottoms: 158.5 ft3 Activity: 7.1774 curies Shipped: Barnwell, Inc. - April 10, 1989 (2)
April 20, 1989 (1)
- 2. Radwaste onsite and awaiting shipment:
A. Resin in storage: 744.0 ft3 ;
B. Estimate resin that will be generated: 225.0 ft3 C. Dry active waste awaiting shipment: 1035.0 ft3 1 - Dry active wasto 2 - Spent resin OFFSITE DOSE CALCULATION MANUAL CHANCES No changes were made to the Offsite Dose Calculation Manual for the month of April, 1
l.,
l-l l
l OPERATING STATISTICS (NRC REPORTS) e i
.3 CPERATIfJG DATA REPORT-
,, D O C Kli T L N O .00-327 DATE MAY 0D,1989 COMPLETED.DY D.C.DUPRhF 4
TELEPHONE (615)843-67;';!
'~
- Oi'ER ATING STATUS ,
l1 J UNIT NAME; SEGUDYAH NUCLEAR PLANT, UNIT ~ 1 NOTES: -
- 2. ^ REPORT. PER IOD: APRIL 1989
- 13. f LICENSED THERMAL POWER (MWT): 3411.0
- 14. NAMEPLATE RATING (GROSS MWE): 1220.6
- 5. . DESIGN ELECTRICAL RATING (NET MWE): 1140.0 c 6. MAXIMUM DEPENDADLE CAPACITY (CROSS MWE): 1183,0 7.. MAXIMUM DEPENDADLE CAPACITY (NET MWE): 1148.0
- 8. IF CHANGES . OCCUR IN CAPACITY RATINGS (ITEMS NUMDERS 3 THROUGH 7)SINCE LAST REPORT, GIVE REASONS-
- 9. PCWER' LEVEL TO WHICH RESTRICTED,IF ANY(NET MWE):
- 10. REASONS FOR RESTRICTIONS, IF ANY: __
THIS MONTH YR.-TO-DATE CUMULATIVE
- 11. HOURS IN REPORTING PERIOD 719.00 2879.00 68664.00-
? i 2. NbMDER OF HOURS REACTOR-WAS CRITICAL 719.00' 2830.25 27654.69
- 13. REACTOR ' RESERVE SHUTDOWN HOURS 0.00 HO. OO O.00 14.- HOURS. GENERATOR ON-LINE '
719.00 2804.66- 26868.54
- 15. UNIT PESERVE SHUTDOWN HOURS 10.00 0.00 0.00 1A GFOSS' THERMAL ENERGY GENERATED (MWH) 2445906.48 9413478.21- 86Y90415.08
- 17. : GMOSS ELECTPICAL ENERGY GEN. (MWH) 842260.00 3247870.00 >29383146.00 18.' . NET ELECTRICAL ENERGY GENERATED- (MWH) 813470.00 3135491.00. 20006Y19.00 119..UN)T SERVICE FACTOR 100.00 97.42 39.13
- 20. UNIT AVAILABILITY F ACTOR 100.00 97.42 39.13
- 21. UNIT CAPACITY FACTOR (USING MDC NET) 98.55 94.87 35.59 Ei22.' UNIT CAPACITY FACTOR (USING DER NET) 98.50 94.87 35.59 23.1 UNIT FORCED OUTAGE RATE O.00 2.58 54.93 .
- 24. 11,UTDOWNS ECHEDULED OVER NEXT 6 MONTHS (TYPE, DATE. AND DURATION OF EACH):
' 2 5. IF SHUTDOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP:
'. NOTE THAT THE YR.-TO-DATE AND CUMULATIVE VALUES HAVE DEEN UPDATED.
4 1
, , *2 s..;_--
. w;. g --$ ,k,, d ig n ,' e
'SEQUOYAH NUCLEAR PLANT AVERAGE DAILY POWER LEVEL DOCKET NO. : 50-327 UNIT : ONE DATE : MAY 08,1989 COMPLETED BY : D.C.DUPREE TELEPHONE : (615)870-6722 MONTH: APRIL 1989 AVERAGE DAILY POWER LEVEL' AVERAGE DAILY POWER LEVEL DAY (MWe Not) DAY (MWe Net) 01 1140 16 1137 02 1140 17 1137 03 1139 18 1135 04 1138 19 1131 05 1137 20 1127 06 1137 21 1135 07 1139 22 1138 08 1137 23 1132 09 1132 24 1132 l
10 1137 25 1131 11 1136 26 1120 12 1136 27 1128 i
~13 1139 28 1130 14 1137 29 1129 15 1137 30 1130 l
a ,D8 tfL) 7 y 2 o5 . )
" co R C5 e e u rE 3 u- rnoL
- qy 1 v c t of(
0 ea . 6 i n si 5 sMD( t e nt st) cor . I at r1 et r - reo6
.EEYE l
r u OMTBN r nc i G. aphe e0 p1 NAA O ooe r p t eSR -
NDDH CiR i r G T EP t A bP ytE ET TE & ct i - rnR KI EL A n r o h rt eU CN OU LE 'e e xo nvN PT s v f EfEE(
- D M u e 4 O a r s .
C C P e
i t
i
_ v i
t 'm.i c at a .rs mci
- S o aS x N N r E O cc .
I Sif T - t o C ll a U 9 : a a m D 8 d uuot a u E
R 9
1 ggu onnt nt dh h a a uoue tMMACORO R l Mgqgu e - - - -
.. E i H1 234 W r _ a -
O p P A .
D 3,'7 u N
A H T
ggg S N N O W M O e #
D T e T R st t U O nnr H P eeo S E c vp )
R i E e n T -
L R i I
N a
l U p n x o n$ygM fn -
E - i ytgc.m ( t s
t c
uo yc e= ee .i g
- I rT u t r
I ! 1%
l r s
- ,g gc c io e e a R Ft e
c - y ggo v nan no g r enit osi gt 32ac mel at
. : pt el a na 1 e
. ri nuurir ouif gemeh s qa eeplip aE! ! RROA
_ gy e -
RABCDEFG 2 '
e t d a e D l d u ed ce rh oc FS o
N : :
FS I
. ,! i!
4 4 OPERATING DATA REPORT DOCKET NO 00-328 DATE MAY OD,1989 COMPLETED DY D.C.DUPREE TELEPHONE (615)843-6722 GFERATING STMUS -
- 1. UNIT NAME: SEGUOYAH NUCLEAR PLANT, UNIT 2 NOTES:
- 2. Rt70RT PERIOD: APRIL 1989
- 3. LICENSED THERMAL POWER (MWT): 3411.O
- 4. NAMEPLATE RATING (GROSS MWE): 1220.6 '
- 5. DESIGN ELECTRICAL RATING (NET MWE): 1148.O 6,., MAX I MUM DEPENDABLE C AP AC I TY ,,( GR,OSS MWE ) :
- 7. M'AXIMUM DEPENDABLE CAPACITY (NET' MWE):
, 1183 1 14 8. 0) ,O _,, , , ,,, ,
O. IF CHANCES OCCUR IN CAPACITY RATINGS (ITEMS NUMBERS 3 THROUGH 7)SINCE LAST REPORT, GIVE REASONS: __
- 9. FGWER LEVEL TO WHICH RESTRICTED,IF ANY(NET MWE):
- 10. REASONS FOR RESTRICTIONS, IF ANY: _ _ _
=
THIS MONTH YR,-TO-DATE CUMULATIVE
- 11. HOURS IN REPORTING PERIOD 719.00 2879.00 60624.00
- 12. NUMDER OF HOURS REACTOR WAS CRITICAL 222.47 651.72 27838.36
- 13. REACTOR RESERVE SHUTDOWN HOURS 0.00 0.00 0.00
- 14. HOURS GENER ATOR ON-LINE 88.00 516.97 27108.94 *
- 15. UNIT RESERVE SHUTDOWN HOURS O.00 O.00 O.00
- 16. GROSS THERMAL ENERGY GENERATED (MWH) 78400.23 1112587.67 82436685.73
- 17. GROSS ELECTRICAL ENERGY GEN. (MWH) 17610.00 309889.00 20046C00.00
- 18. NET ELECTRICAL ENERGY GENERATED (MWH) -1750.00 314101.00 26705197.00
- 19. UNIT SERVICE FACTOR 12.24 1 7 . 9 6- 44.72
- 20. UNIT AVAILABILITY FACTOR 12.24 17.96 44.72
- 21. UNIT CAPACITY rACTOR(USING MDC NET) O.00 9.50 38.37
- 22. UNIT CAPACITY FACTOR (USING DER NET) 0.00 9.50 38.37 1
- 23. UNIT FORCED ' OUTAGE R ATE '77. 51 36.98 49.48
- 24. SHUTDOWNS SCHEDULED OVER NE) T 6 MON'lHS (TYPE, DATE; AND DURATION OF EACH):
i
- 25. [F HUTDOWN AT END OF REPORT PERIODI ESTIMATED DATE OF STAR NOTE THAT THE YR.-TO-DATE AND CUMULATIVE VALUES HAVE DEEN UPDATED.
I q
'a h> 4.; t,y % :.
p ; nqw p. ; . . . E
.w ;c
'SEQUOYAH NUCLEAR PLANT AVERAGE DAILY' POWER LEVEL DOCKET NO. : 50-328 UNIT : TWO DATE : MAY 08,1989 COMPLETED BY : D.C.DUPREE TELEPHONE : (615)843-6722 l-MONTH: APRIL 1989 AVERAGE DAILY POWER LEVEL AVERAGE DAILY POWER LEVEL DAY (MWe Net) DAY (MWe Net) 01 -21 16 -28 02 -12 17 -29 03 -21 18 -33 04 -14 19 -28 05 -27 20 -30 06 -29 21 -32 07 -29 22 -28 00 -30 23 -29 9
09 -26 24 -30 10 -26 25 -32 11 -30 26 -29 12 -30 27 6 13 -31 28 110 14 69 29 154 15 -29 30 253
h1 u4 ,
a r oa oD cF a ,D8 t r or e . i i 8 y ' s rt f r t t fL) 2 o5 . ) e ea s c o R 3 u C5 e e .
t l d e e u rE
- q y 1 v c e ueh t r noL 0 e a . 6 i n d h gvT d
t of(
5 sMD( t e e t e o e si cor e r m nt st) et r p f e . e I at r1 EEYE r s p - r e o6 ot I TBN u .
or rt s G a e p1 NAA O r nc e g
r e s r ph e0 NDDI oo e e et ge e
J CiR a v s a nt v t eSR -
T EP t t o uwi i r G ET TE & ct u _
ad ed . o bP ytE KI EL A n o e c eb e r i rnR CN LE e e nc e e e e e h rt eU OU PT s v g i u bf s p w n xo nvN D 1 1
u e i
i b o a s o i EfEE(
0 l r p d ewrp b 4 C
a r r C P l eh e e
u utn pt evt u t t p non n f r e if i e r e e o c ro b mc o c g r f r t r rr f r .n e guoe e mi 3 e p rn t f p e p at n oi r n .rs e i 0 t gee 0 i 3 mci S l l 3 cn l p3 aS x N c f ai i l f
2 r E O y f t ew h ot f cc I C o a t R s wt a O a Sif T t o C ll a U 9 : a a m D 8 d uu ot a E 9 R 1 n,gu o nnt ntd ht a a uo u oCggg u l
f
! ACO
. R l E i !e 1-234 W r 3 O p P A D
N ggu A !
l gu g T
S N N 0 W !
O e #
D T e T R st t U
I O n n r l P e e o S E vp T
R -hE t R e )
n I i N a U l p n lou g:c g6n x o ygu ein 4 9 3 E
( t i
t 9 s c 44o ygc:
6 ee i rT u
r ti go ge C B A B l r i o a R s
e F e c y gyo v 3 2 2 3
t ngr c tgso 1
4 9
. 5 na nor e nit o 5
2 4 1 :
mel at 0 pt el a n 3 3 ninuuri o uif geme s qae e pd
,e y S S F 5
aE! RROAO e - - - -
2 RABCDEFG 8 4 5 e 1 1 1 8 t 1 4 4 2 a 0 0 0 4 d D 9 9 9 0 e 8 8 8 9 l 8 d u ed c e rh oc 2
FS o 1 3
_ 4
_ N : :
_ FS I
,ll ,'
,w 7
- - r -w-------_--- -_-_-_____ _____
- ?, e
~
G OPERATIllG STATISTICS (TVA REPORTS) il I
i
aa---_x--_---.____---- - - - - - - - - - - - ------- - -----.____----____----w s--- --
TvA T382A toNP 4 88) NUCLEAR PLANT OPERATING STATISTICS 1 . ,
i Sequoyah Nuc1 car Plant a
719 Month April 19 89 Period Hours f tem No . Unit No. - UNIT ONE UNIT TWO PLANT 1 Average Hously Gross Load. kW 1,171,433 200,114 1,065,514 2 Maumum Hour Net Generation, MWh 1,156 300 1,432 3 core Ther mal Energy Gen, GWD (t)2 101.9128 4.1000 106.0128 4 steam cen. Thermai Eneigy Gen., GWD (t)2 102.2991 4.1168 106.4159 8 5 Gross Electrical Gen., MWh 842,260 17,610 859,870 j 6 Station Use, MWh 28,790 19,360 48,150 E 7 Not Elect scal Gen., MWh 813,470_ -
1,750 811,720
$ 8 station Use, Percent 3.42 109.94 5.60 9 Accum. Core Avg. Exposure, MWD / Ton 1 4,644 46 4,689 10 CTEG This Month,106 BTU 8,347,879 335,840 8,683,719 11 SGTEG This Month,106 BTU 8,379,523 337,213 8,716,736 1?
13 Hours Reactor Was critical 719.0 222.47 941.47 14 Unit Use, Hours Min. 719:00 88:00 807:00 15 capacity Factor, Percent 99.02 2.07 50.55 16 Turbine Avail. Factor, Perr ent 100.0 54.38 77.19 g
17 Generator Asail. Factor. Percent 100.0 55.37 77.69 ~ ;
e O 18 T ur boqan Avad Factor, Percent 100.0 54.38 77.19 l
$ In peactoi Avam ractor. Percent 100.0 57.39 78.69 0 PO Unit Avo A Factor. Percent 100.0 12.24 56.12 T urbine Startups 0 3 3 21 0 1 1 22 Reactor Cold Startups
??
24 Gross Heat Rate, Blu/kWh 9,910 19,070 10,100 3 10,700 j ?5 Net Heat Rate, BtulkWh 10,260 N/A 3 26 Gross heat Rate, Btu /kWh (w/ oil) 10,100 0 27 Net Heat Rate, Btu /kWh (w/ oil) 10,700 0 28 Throttle pressure, psig 84Y.1 940.1 859.1 s 2 29 Throttte Temperature, F 526.8 538.8 528.1 j 30 Exhaust Pr essure, inHg Abs. 2.2 5.1 2.5 E 31 Intake Water Temo., F 56.8 61.7 57.3 f- 3 "' ' ~
3? Main Feedwater, M lb/hr 14.9 3.1 13.6
$ 34
[ 35 36 37 Full Power Capacity. ErPD 404.86 411.60 816.46 38 Accum. Cycle Full Power Days, EFPD 121.29 1.20 122.49 j 39 Oil fired for Generation, Gallons 1,056 138,000 3 40 Oil Heatinq Value. Bto/ Gal. ~
16 41 Qgsel Gener ation. MWh 4?
Max. Hour Net Gen. Max, Day Nc' Gen- Load MWh Time Date MWh Date f actor, %
a 43 1,432 2400 4-30-89 32,496 4-30-89 47.04 3 Remarks: AFor BF NP this value #5 MWD /STU and for SONP and WBNP this value is MWU/MTU.
5 2(t) indicates Thermal Energy.
G
<#' O Daie suemitted FAY 12 'I989 Date nevised d. .Ou et
, [ Tiant Manages
,l !i Il i ,1 ,
lI 8
I i
~ -
IUl IL-f g
Vr 4Lo
.Bg #~
9 I P,i 8 V 0 ;,
I '
9 l 1 g 1
l4 y l
i dL, g f r
p 9 dF l
A 1
7 O TA ~ E L -
s .
e a
y rn SG L - '
y i t
'TUN G t
n u I A _
rt ei r
e ' A RT f T UU "SDO P
X t
G R N NO I WT T
TOC UD A H E S R i
t E
S U
C la n A C
P E
V r a G T e A I
L lc T u
I U
v B
A L
I A
N O
V ~__
A e n ~
D O R ,
A n.
E G
h a
o y
o i
e i n 'u me s
'! a e . ; * ! . i ' i * ' i a ' i !; i i !' i i i ' ' i i I X
A u q
N u T t. s U e S
t
+
n e
h t
X O U n.
T et u u _ .
t
- m. O I . .
!Xt g ,~;
! iI ! I!' i e t' !; i i i 8 ' . . t!le
?a 1 U .
X. _
.n t
u u e * ; s 8 '
j.
ls i.i e!i8 '
W . i i t8 ' ; i i6 !; iIi l t.
i _
r n.
te s
t c
o a
u,i a i 'ts
- ! i !i8 li' l!;' il;
's.
- t. s !i ' : * ; !i!i 4, te
- - i _
A e .
to
- H .
ni n u, e.i e s ; e * - .i ' i' ; e' G
i ' i'
's.
w i!i!i8 li * ! . e :I t
a W e i
M n c
In ;8 , e , '
e . t !i $ ; * ;
u s e ' i !i!;i !i* ; * ! i :i l T i.
) 3 e ht 8 f e
( 1 .
W w 1 O u M to tl g e ' s i!l # i ;i! .; i!; i* i i ! :i ; e8 e s e !
1 M g 9t a
N n e
1 4 W l.
3 f
- n. a w n. n 0 0 ta A 3_ 5 R n u O0 0 Q .. !!l!i '
. o r
t.
!!l!t::t!!!tll " o 0 e
w 0 la ce s!] e ; s i' Ii e ' ' i' !' . i i3 ii I !i i!i!i i b o
2 2 r u e 43 4
- 6) P 1 r i
c e
s 22 ~
2 Q..!! l!!!!!! !!l:' .::l!
4 2
9 1
7 r g f.
r .i.
7 3
o .
fr: 1 a 05 0
P t
N E t
u Q..!!l* 0 0
,!!!!!!::::i xe sa l!
',s 40G s '0 0 0 P( t o e i f s * ; e ' ' .; i l ' '
C R i n e
T 3 4
. : ' ' ; . iI ' ; .I 0 d w.
G r e s 22 2 Q.. !! !
!!!!! !!!!!!!:!!:l :
v 4
2 1 9 _
% e t. 7 6 s n 4. p y 2 3 4 5 6 7 8 9 0 1 2 111345s7 I
A e a is 8 9 0 1 2 3 4 5 6 7 8 9 0 1 ts.
v *t 1 1 1 i 1 1 2 2 2 2 2 2 2 2 2 2 wfb 1
l 3 3 t T L i n
, o ty
,ll!
.m.n- ~ ..
3 4
6 :
- OU l - l, w
. 4 J > s. t :
4 u C_ , ,, i f g
e ag- e
- > ~ d
..O y'- 6' a t.
V ! I 3 e n" a,,.
l ,h.
b n .e %d* %
, l.
' s fs.
4 N %< 3
, I D *- , t b
, f.
- l E I
& I
> -4 w)Ee Oc u i k "
-8 E_ 'A'* -4
{ t ._ p % w g ,
W 4 W b & $
MOO f x z
> g.\
l I
o et L. O. g l 6 m C
O U2O
%e 3* D D" u w w
_e - q O
,3 y w w O r
_@ . . ,r M pr. I h
I f,
W w w I M .
m i
- D e e 1 c w w m <
y g
g e e w e W .u o e h Q C* b*
t s < "
- s. . g a .
i c Z 0 0 0
4 -
e :
.a *
- it o
A .C. .C.e t A A f
< o 3
w m n Q v e H
.c *n w g *
- o o S
[3 .s~ .... . . ..;.
. +
. _. . . .. . . g... . _. _ ....._ . ...
N e '
et s v.
- [-
.n c - 3- -, e c .
D Y E - c !\
O $ I H w.
k.
s llI .i 4 a
~
4 l \.
- a. >6 --
c
- 4' 1' 4 I I h * .'
3 -A m m f \
l l t n m i f o c o o o ko c,lo l <,l ..! . ~ .e e
.l .... .o.o.. c_c_c.... ...... .. . ...l .. . . ...._g
.s. ........
> e~ e
.c.
3 _ _ _ .... _ .,
- .,c.i,mp I < .l . .. . ...
. ~, i~ ... . . ......
.> ~ .. ~ m..... .......... . . . . .. . . ...
.~, .e e e e o :
c O. oo h 4 4'"I .* N @ m 3
3
.$ s$o g ..c. o
_ <f..g . _
4.>
.. ~_ ... .
4 m e -
.r...
~
3 E c o
- ~ ~~<r..l..,...
m o i
. . > ~o m e o o o - .e
- s o o e g
j f e c o l o 3 o o c ( .m N j
> c .n e g . .. .".s .. . . . , ...... ... ... _.......... ... .. . _ _..
,. r 4 m 4 .1 4 . o g ~ n n <. . . .... .. . . .. . ..) m o 7 g D c o o e i o- e a 2 m e j c c c( . * . . .4 .>.-
> cC O 4 - e C
. ... .. _ . '.s. =.-. = -.... .. .... . . . . ....... ._ . . . ... .
D . 4 m 4 4 m mn 4 o e ..C o
ec eS g m -
' (I..i .i.......... .)
e N .O.*C c c C N a
U %
- t e _ .... . .. _.... .. .. .. ._... . ..... . . .. _ .. ...
..... j Z
4*
n v = .
2 3 i f 2
m ~
4lo e y ,n
, o
, .f .
.. . . .. . . . . . _...e. . . . . _ _. __ ..._
m 4 C
. ,g O .
C 0 N Y 3 # l os e m mle oc C
g 4 - % ?
l C N NjN EC
- P
& -E 3 E e m m oio o S- a s a m 4 i
g -g ;3 2 .. . . . .. . . .... .. . . . .. . . _. . . .
cl,c C
< w . .
m et,.
.e 3 w i
- I 41 C
o $ hs > A eM .t A C N O @
- -C -#
- f N
-O C
- N M ** an O h T m'nO'
- ; -O -==' - N - M N N 7 N N N N N m m - -
9 i 2 i
i 5
- 2 i
V A E 1
i 2 R N GO i O T
~ I 7
N i
S 1
U i
H A
R Y i 3 OOU 1
T Q i C E AS E i 9 R
i
. 5 1
0 0 0 0 0 0 0 0 0 0 0 0 0 1 0 9 8 7 6 5 4 3 2 1 1 1 1 sZdewct g'$o0 - oosVoe-l l
UNIT 1 REACTOR HISTOGRAM ANALYSIS Unit 1
- 1. Condenser waterbox 1B-2 out of service for maintenance.
d i
1
9 9
- 2 a
2 5
- 2 A
\
- A R O G TW 1
c 2
, ~
0 T I
6 >
~
S U N e
> 7 1 3
H '+
r >
A 3 u
R Y 3 2' -
OOU 1
a 3
~
1-
~
CQ E s
AS E , 9 R
~
, 5
- 1 0 0 0 0 0 0 0 0 0 0 O 0 0 9 8 7 6 5 4 3 2 1 1 1
ywooWo_ $3Oc : t -
o0 o<W0-h 1
4 UNIT 2 REACTOR HISTOGRAM ANALYSIS 4
Unit 2
- 1. IIolding/ reduce power because of excessive RCS leakage.
- 2. Tying online.
- 3. Iloiding for turbine overspeed test.
l 4. Reactor trip, low-low level on S/G 4 (04/15/89).
1
- 5. Reactor trip, low-low level on S/G 1 (04/16/89).
- 6. Reactor trip, low-low icvel on S/G 2 (04/19/89).
- 7. Holding power at 2 percent power, maintenance on the feedwater system.
- 8. Iloiding for fuel conditioning and thermal power verification test.
- 9. Power ascension.
I
6 ,
SUMMARY
OF MAINTENANCE ACTIVITIES l
4 [
MAINTENANCE
SUMMARY
(ELECTRICAL) b'- _.- - - - - - - -___----__--.x---_- - - - --____-_---. _ ___ - - --- - - - - - - --
Electrical Maintenance Report 1
02/13/89, 2-FCV-062-0022, 03/24/89 Durire unit 2 refuelirg outage, a special inspection indicated RCP 2 seal return flow contzul valve was not operating 'Ihe solenoid coil housire and coil retainer were missing, allowing air to leak from the solenoid preventire the control valve fmn operating. Also, the solenoid coil was installed up-side-down. 'Ihe solenoid was abecembled and reassembled correctly. 'Ihe problem persisted.
Replaced solenoid, ruterminated wiring and returned valve to service. (WR B256476) .
l 03/31/89, 2-FCV-001-0183-A, 04/02/89 Durirg unit 2 refueling outage, a visual alarm indicated steam generator 3 blowdown isolation valve throttled. 'Ihis is a generic Masonellan stem rotation problem. Removed stem nut and applied thread locker. Reinstalled nut ard staked threads. Returned valve to service. (WR B256280) .
1
e e
.' =
MAINTENANCE
SUMMARY
(INSTRUMENTATION)
Instrunent Maintenance Report
'05/04/88, 2-XE-068-0366, 03/23/89 During routine observation the RCS acoustic. valve positioner accelerometer special element was not emitting a proper signal. m e element was'found to have a broken hardline cable. A new hardline cable was install 6d with Raychem, and the element was functionally tested. (WR B785232).
06/17/88, 2-Pl'068-0301,
'03/07/89' Durire unit shutdown the RCS pressurizer relief tank pressure transmitter required replacirg because of instrument drift problems.
Cause was determined to be the age of equipment ard wear and because ,
of drift. Installed new transmitter, rechecked calibration and found i in toleran . Ibrformed postmaintenance test and returned to i service. (WR B295731).
11/03/88, 1-PDI-030-0133, 1 02/25/89' During unit operation the ventilation system containment annulus delta P pressure differential iniicator was found out of tolerance while performing SI-193, " Containment Building and Auxiliary Building Ventilation Systems - Units 0,1 and 2." Root cause is unknown. It may have been because of the age of equipment.
Recalibrates to desired tolerance and returned to service. No PFO report was necessary. (SI-193) .
12/27/88, 2-PT-068-0066-E, 04/06/89 Durirg unit 2 cycle 3 outage while performing SI-484, " Periodic Calibration of Reactor Vessel level Instrumentation (RVLIS) and RCS Wide Range Pressure Channels (P-403, P-406) (Refueling Outage) 10 CFR 50.49, - Units 1 and 2," the RCS loop 4 hot leg pressure ;
transmitter was found out of operational tolerance. m e cause of j failure was unknown but may have been because of transmitter drift !
or equipment age. W e transmitter was recalibrates and returned to ;
service during the performance of SI-484. EQ maintenance was '
performed in accordance with G E requirements. j 12/27/88, 2-F1'068-0069-D, 04/06/89 During unit 2 cycle 3 outage while performiIg SI-484, the RCS ;
wide-rarge pressure loop 3 hot leg pressure transmitter was found l cut of operational tolerance. Se cause of failure was unknown but '
may have been because of drift or equipment age. me transmitter was '
recalibrates and returned to service during the performance of SI-484. D2 maintenance was performed in accordance with GE requirements, q
l I
l
Instrunent Maintenance Report 01/20/89, 2-IIP-063-0060, 03/30/89 During unit outage the SIS accunnlator tank 4 level tran0mitter was found out of tolerance while performig SI-161, " Channel Calibration of S.I.S. Amlator Tank Water Invel ard Pressure Instnmentation
- Units 1 and 2."'The root cause is unknwn. It may have been because of agirq of equipment and/or cycling. Recalibrates to desired tolerance and returned to service. PRO 2-89-56 was initiated ard submitted for failure analysis. (SI-161).
01/20/89, 2-LS-063-0060B, 03/30/89 During unit outage the SIS accumulator tank 4 lw level switch was I found out of tolerance while performing SI-161. We root cause is unknwn. It may have been because of agirg and cycling of equipment.
Recalibrates to desired tolerance and teturned to service.
PRO 2-89-56 was initiated and submitted for failure analysis. i (SI-161).
01/20/89, 2-IIP-063-0082, 03/30/89 During unit outage the SIS level transmitter was fourd out of tolerance while performirg SI-161. 'Ihe root cause was unkncun. It may have been cecause of age ard/cr cycling. Recalibrates to desired tolerance ard returned to service. PRO 2-89-56 was initiated and submitted for failure analysis. (SI-161) .
01/20/89, 2-IlP-063-0089, 03/30/89 During unit outage the SIS accumulator tank 3 level transmitter was found out of tolerance while performing SI-161. 'Ihe root cause is unknown. It may have been because of age of equipnent. Recalibrates to desired tolerance and returned to service. PRO 2-89-56 was initiated and submitted for failure analysis, (SI-161).
01/20/89, 2-IlP-063-0099, 03/30/89 Durirg unit outage the SIS accumulator tank 2 level transmitter was fourd out of tolerance while performirq SI-161. 'Ihe root cause was unkncun. It may have been because of the age of equipment.
Recalibrates to desired tolerance and returned to service.
PRO 2-89-56 was initiated and submitted for failure analysis.
(SI-161).
01/20/89, 2-IS-063-0099B, 03/30/89 Durjng unit outage the SIS accumulator tank 2 level switch was fourd out of tolerance Wile performing SI-161. 'Ihe root cause is unknown.
It may have been tause of agiry of equipment. Recalibrates to desired tolerance ard returned to service. PRO 2-89-56 was initiated and submitted for failure analysis. (SI-161).
i I
l l
Instntment Maintenance Report 01/20/89, 2-IT-063-0109, 03/30/89 During unit outage the SIS level transmitter was found out of
-tolerance while performing SI-161. We root cause was unknown. It '
may have been because of equipment aging and/or cycling.
ILcalibrated to desired tolerance and returned to service.
PRO 2-89-56 was initiated and submitted for failure analysis.
(SI-161).
01/20/89, 2-LT-063-0119, .
03/30/89 During unit outage the SIS accumulator tank 1 level transmitter was found out of tolerance while performing SI-161. Se root cause is unknown. It may have been because of age of equipment. Recalibrates to desird tolerance and returned to service. PRO 2-89-56 was j initiated and submitted for failure analysis. (SI-161).
01/20/89, 2-10-063-0129, 03/30/89 During unit outage the SIS accumulator tank 1 level transmitter was found out of tolerance while performing SI-161. me root cause is unknown. It may have been because of aging of equipment.
Pamlibrated to desired tolerance and returned to service.
PRO 2-89-56 was initiated an submitted for. failure analysis.
(SI-161).
01/21/89, 2-Il -003-0038-E, 03/16/89 During unit 2 cycle 3 outage while performing SI-94.4, " Reactor Trip / Engineered Safety Feature / Accident Monitoring Instrument Steam Generator Invel Channel Calibrations (18 Months) Units 1 or 2," the min aM auxiliary feedwater system steam generator 1 level transmitter was found out of operational desired values. W e transmitter was out of tolerance because of transmitter drift. .Wis is a generic problem to Barton 764 transmitters and has been identified. Trending data is being evaluated me transmitter was recalibrates and returned to service during the performance of SI-94.4. Applied lubricant and installed inner and outer o-rings and torqued cover screws. (SI-94.4).
01/21/89, 2-LT-003-0039-F, 03/16/89 During unit 2 cycle 3 outage while performing SI-94.4, the main and auxiliary fecdwater system steam generator 1 level transmitter was found out of operational desired values. W e transmitter was out of tolerance because of transmitter drift. h is is a generic problem to Barton 764 transmitters and has been identified. Trending data is being evaluated. We transmitter was recalibrates and returned to service during the performnoe of SI-94.4. PRO 2-89-35 was initiated, but after evaluation the problem was determined not reportable Applied lubricant and installed inner and outer o-rings and torqucd cover screws. (SI-94.4) .
1 l
)
f Instrument Maintenance Report L
01/21/89, 2-LT-003-0097-G, 03/16/89 During unit 2 cycle 3 outage while performing SI-94.4, the main arxi auxiliary feedwater system steam generator 3 level transmitter was found out of-operational desired values. S e transmitter was out of tolerance because of. kcusdtter drift. mis is a generic problem to Barton 764 transmitters and has ban identified. Trending data is being evaluated. W e transmitter was recalibrates and returned to service during the performance of SI-94.4. Applied lubricant and installed inner and outer o-rirgs and torqued cover screws.
(SI-94.4).
01/21/89, 2-LT-003-0106-E, 03/16/89 Durire unit 2 cycle 3 outage while performing SI-94.4, the main and -
auxiliary feedwater system steam generator 4 level transmitter was found out of operational desired values. W e transmitter was out of tolerance because of transmitter drift. mis is a generic problem to Barton 764 transmitters and has been identified. Trending data is being evaluated. he transmitter was recalibrates and mturned to .
service during the performance of SI-94.4. Applied lubricant und installed inner and outer o-rings and torqued cover screws.
(SI-94.4).
01/25/89, 2-TS-001-0018B-B, 02/25/89 During unit. 2 cycle 3 outage while performirg SI-580, " Periodic Calibration of Main Steam System (Refueling Cycle), - Units 1 and 2," the main steam system flow to AWP turbine isolation high temperature switch was found out of operational setpoint tolerance.
W e cause of failure was unknown but may have been because of setpoint drift or equipment age. W e switch was recalibrates and returned to service durirg SI-580. EQ maintenance was performed in accordance with GE requirements. (SI-580) .
01/27/89, 2-PT-063-0086, 02/10/89 During unit operation the SIS accumulator tank 3 pressure transmitter output reading was 94mA with no pressure applied.
Defective amplifier probably because of age ard/or wear. Replaced amplifier and powered up transmitter. ' Calibrated and returned to service. (WR B757658).
01/27/89, 2-PT-068-0068-F, 03/06/89 During unit 2 cycle 3 outage while performing SI-199, " Periodic Calibration of Reactor Coolant System Instrumentation (Refuelirg Cycle), - Units 1 and 2," the RCS loop 4 hot leg pressure transmitter was found out of operational tolerance. W e cause of failure was unknown but may have been because of transmitter drift.
We transmitter was recalibrates and returned to service durirg the performance of SI-199. D2 maintenance was performed in accordance with G4DS requirements. (SI-199) .
l l
L E________________________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
Instrunent thintenance Report 01/28/89, 2-LT-068-0325C, 03/02/89 Durirq unit outage the RCS pressurizer level transmitter was found out of calibration while perfoming SI-88.2, " Remote Shutdown Monitorirg Instrumentation - Pressurizer Level Channel Calibration (Refuelirg Outage), - Unit 2." m e root cause is un h n. It could have drifted out of tolerance from age of equipment and/or wear from cyclirg. Recalibrates to desired tolerance, performM postmaintenance test and returned to normal. (SI-88.1).
01/28/89, 2-LT-068-0326C, 03/21/89 Durirg unit outage the RCS pressurizer level transmitter was found out of calibration while performing SI-88.2. We 2xot cause is unknown. It could have drifted out of tolerance from age of equipment and/or wear from cycling. Recalibratal to desired tolerance, performed pcsbnaintenance test and returned to service.
(SI-88.1).
02/02/89, 2-PI'068-0322-G, 03/10/89 During unit 2 cycle 3 outage while perfomirg SI-94.88, " Channel Calibration of Pressurizer Pressure Channel 4, Rack 12 Icop P-68-322 (P-458) - Unit 2," the RG pressurizer pressure transmitter was found out of operational tolerance. S e cause of failure was unknown but may have been because of transmitter drift or equipment age. m e transmitter was recalibrates and returned to service during the performance of SI-94.88. EQ maintenance was performed in accordance with QMDS requirements.
02/03/89, 2-PI'068-0334-E, 03/08/89 Durirg unit 2 cycle 3 outage while perfoming SI-94.86, " Channel Calibration of Pressurizer Pressure Channel 2, Rack 5 Icop P-68-334 (P-456) - Unit 2," the RCS pressurizer pressure transmitter was found out of operational tolerance. We cause of failure was tmknown but may have been because of transmitter drift. m e transmitter was recalibrates arxl returned to service durity SI-94.86. Inbricant was applied to the new inner and outer o-rings then installed. We cover was torqued in accordance with QMDS requirements. (SI-94. 86) .
02/06/89, 2-IS-068-0339D, 03/14/89 Durirg unit outage the RCS pressurizer level switch was found trippire at different setpoints while performing SI-94.1, " Reactor Trip Instrumentation Refueling Outage Channel Calibration (RCS PRZ Press and IcVels), - Units 1 and 2." 'Ihe root cause is unkncrai. It may have been because of cycling fatigue and/or a deficiency in design. Tried a diffemnt wiring scheme and calibrated switch.
Nuclear Engineering is evaluating the problem and will provide a ivc.uuuandation at a later date. ReturnM to service. (WR B769345) .
Instrument Maintenance Report
,03/09/89, 2-PP-068-0340-D, 03/08/89 During unit 2 cycle 3 outage while perfoming SI-94.85, " Channel Calibration of Pressurizer Presure Clannel 1, Rack 1 Loop E68-340 (P-455)' - Unit 2," the RCS pressurizer pressure transmitter was found out of operational tolerance. me cause of failure was unknown but may have been because of equipment aging. Se transmitter was recalibrates and returned to service during the perfomance of SI-94.85. Inbricant was applied on new inner and outer o-rings then installed. m e cover was torqued in accordance with QMDS requirements. (SI-94.85).
02/10/89, 2-FT-001-0021A-D, 03/22/89 During unit 2 cycle 3 outage while performing SI-98.1, " Channel Calibration for Engineered Safety Feature Instrumentation (Steam Flow & Pressure), - Units 1 and 2," the main steam system steam generator 3 main steam header flow channel 1 transmitter was found out of operational tolerance. S e cause of failure was unknown but may have been because of transmitter drift. W e transmitter was recalibrates and returned to service during the performance of SI-98.1. EQ maintenance performed in accordance with QMDS requirements. (SI-98.1).
-02/13/89, 2-PT-068-0323-F, 03/08/89 During unit 2 cycle 3 outage while performing SI-94.87, " Channel Calibration of Pressurizer Pressure Channel 3, Rack 9 locp P-68-323 (P-457) - Unit 2," the RCS pressurizer pressure transmitter was found out of operational tolerance he cause of failure was unknown but may have been bemum of transmitter drift because of equipment age. h e transmitter was recalibrates and returned to service during SI-94.87. Inbricant was applied to new inner and outer o-rings then installed. W e cover was torqued in accordance with QMDS requirements. (SI-94.87).
02/13/89, 2-R4-068-0323B-F, 03/08/89 During unit outage the RCS pressurizer pressure nodifier was fourd out of tolerance while perfoming SI-94.87. he root cause is unknown. It may have been bx:ause of the age of the equipment ard/or cycling fatigue. Recalibrates controller to desired tolerance and returned to service. No PRO was initiated. (SI-94.87) .
03/02/89, 2-H4-003-0155A, 03/28/89 During unit outage the main and auxiliary feedwater system auxiliary feedwater to steam generator 2 flow modifier was fourxi out of tolerance while performirry SI-97.2, " Calibration of Auxiliary Feedwater Flow Rate for Remote Shutdown and Accident Monitoring (550 Days), Unit 2." h e modifier was out of calibration, possibly because of the age of equipment and/or wear. Recalibrates to desired tolerance ard returned to service. (SI-97.2) .
l 1
Ins' m t Maintenance Report
'03/02/89, 2-FM-003-0170A,
.03/28/89. During unit outage the main and auxiliary feedwater system steam generator 4 flow modifier was found cut of tolerance while performing SI-97.2. Se modifier was out of calibration, possibly because of age of equipment and/or wear. Recalibrates to desired
. tolerance and returned to service. (SI-97.2) .
03/03/89, 2-LT-063-0082,
'03/06/89 . Durirq unit outage the CVCS level transmitter was found with the
" strain gage" wire broken while performing SI-161. Se Itx:rt cause is unknown. It could have been from age of equipnent, vibration and/or fatigue. Replaced " strain gage" arxi calibrated to desired tolerance.
(WR B769963).
03/11/89, 2-LT-068-0339-D, 03/27/89 During unit 2 cycle 3 outage while performing SI-94.1, the RG pressurizer level channel 1 transmitter was found out of operational tolerance. H e cause of failure was unknown but may have been because of setpoint drift due to equipment age. Se transmitter was recalibrates and returned to service during the performance of SI-94.1. Lubricant was applied to new inner and outer o-rings and installed. H e cover was torqued in accordance with QMDS requirements.
03/15/89, 2-PC-068-0340A, 03/17/89 During unit refueling the RG pressurizer pressure controller was found out of tolerance while performirg SI-94.90, " Channel Calibration of Pressurizer Pressure Control arx1 Operational Test of PORV's PCV-68-340a (PCV-455A) and (PCV-456) - Unit 2." he root cause is unknown. It may have been because of age of equipment and/or cycling fatigue. Recalibrates the controller to desired tolerance and returned to service. (SI-94.90) .
03/16/89, 2-FM-002-0035A, 03/23/89 During unit outage the condensate system gland seal heat excharger recirculating flow control valve was found modulating open. Modifier was moving around from vibration. Reinstalled a new support stand adjacent to valve to reduce vibration. Mechanical maintenance is working WR B795767 on valve. (WR B795768).
03/16/89, 2-lM-003-0172-A, 03/16/89 During unit outage the main and auxiliary feedwater system steam generator 3 level modifier input resistor was found cracked which caused the output to read low. m is was attributed to the age of equipment and heat stress. Replaced the input resistor arxi the 100 microfarads capacitor. Bench calibrated and returned to service.
(WR B757383).
t
-3s-
t-R
- - *" Instrument Maintenance Report 03/21/89, 2-FCV-062-0093A, 03/21/89 During unit outage the CVCS charging header flow controller would not stroke valve properly. The cam was worn out-fran age and use.
Replaced cam- and calibrated controller. Verified stroke and returned to service. (WR B757196) .
e
4 4
MAINTENANCE
SUMMARY
(MECHANICAL)
_ - - _ _ _ _ _ _ - - - - ^-
MECHANICAL MAINTENANCE MONTHLY REPORT FOR APRIL 1989 Unit 1
- 3. Flushed oil system on HDTP 3.
- 4. Replaced BAE rupture disc A.
- 5. Completed repairs on various valves (systems 27, 61, and 90).
Unit 2
- 3. Rebuilt bus duct fan 2A and 2B,
- 4. Replaced and adjusted packing in condenser vacuum pumps.
- 5. Balanced RCPs.
- 6. Replaced V-belts on the east main steam vsult air handling unit.
- 7. Completed repairs on various valves (systems 1, 3, 43, 62, 70, and 74).
Unit 0 a
- 1. Rebuilt control air compressor D.
- 2. Completed repair on waste gas compressor A.
- 3. Rebuilt control air compressor B.
- 4. Rebuilt Auxiliary Building floor / equipment sump pump.
- 5. Completed work on floor drain collector tank.
- 6. Rebuilt CDWE pump.
- 7. Replaced refueling water purification filters.
- 8. Completed repairs on v rious valves (systems 27 and 67).
Other
- 1. Completed closure of various CAQRs, etc.
=______-..___---__-_ --
t i
Mechanical Maintenance Report )
08/12/85, 1- M -062-0070-A, 12/28/88 With unit 1 shutdown and during routine observation, the CVCS reactor coolant loop 3 letdown flow control valve was discovered to be leakig through when closed to isolate letdown. 'Ihe cause was determined to be dirty internals and steam cut on plug and seat.
Boltig is also too short. InstallM new plug, stem aM seat with new gaskets and replaced four studs with new studs. (WR B529490) .
09/10/87, 2- M -072-0023-A, 02/23/89 With unit 2 in operation, the containment spray pump A flow control valve from containment sump was discovered to have lots of boron buildup around bonnet of the valve. Packig had been adjusted until there was no adjustment left. New packiry was installed. NOVATS to be performed on preventive maintenance 3851. (WR B276615).
09/14/87, 1- M -062-0093-A, 10/05/87 Duriry routine observation and with unit 1 shutdown, the CVCS chargirg header flow control va'"e was discovered to be allowirg f1w with valve closed and in ranual. Erosion of plug and cage had been caused by operating condition in exteMed mode 5 outage.
Fabricated and installed new stem, plug and cage. (WR B288515). ,
01/12/88, 2- M -062-0083,
> 02/24/89 During unit 2 shutdown the CVCS RHR letdown flw control valve was discovered leaking during routine observation. This was attributed to normal wear. Stem threads were staked, new packing was installed and nut was tightened. (WR B257608).
01/14/88, 1-VLV-063-0640-S, 07/15/88 During unit 1 shutdsn the RHR heat exchanger discharge check valve was discovered with excessive boron buildup on the valve body. 'Ihis was found during incidental observation ard was because of agirg of bonnet gasket. Installed new bonnet gasket and new studs and retorqued valve. (WR B281341) .
03/16/88, 2- M -062-0084, 02/03/89 Durirg unit 2 shutdown the CVCS charging flw to RCS spray valve was discovered during routine observation to be leaking borated water.
'Ihere was no adjustment left on the pacPdng. Cleaned boron off of valve and installed new packdrg. (WR B238217) .
03/16/88, 2- M -062-0085-B, 02/02/89 During unit 2 shutdwn the CVCS chargiry flow RCS cold leg loop 1 valve was discovered durirg routine observation to be leaking borated water. No adjustment was left on packing. Cleaned boron off of valve, added packing, ard adjusted. (WR B238217) .
1 l
I
Mechanical Maintenance Report I
03/16/88, 2-VLV-062-0564-S, 02/24/89 While performing SI-146.1, " Reactor Coolant System Isak Test, Unit 2," durirg unit 2 shutdown, the CVCS number 1 seal bypass isolation valve was leaking borated water. No adjustment was left on packing. l New packing was installed. (WR B223346) .
03/16/88, 2-VLV-062-0566-S, 02/24/89 While performing SI-146.1 during unit 2 shutdown, the CVCS seal water injection isolation valve was leaking borated water. No adjustment was left on packing. New packing was installed.
(WR B223346).
03/16/88, 2-VLV-062-0596-S, 02/24/89 While performing SI-146.1 during unit 2 shutdown, the CVCS number 1 seal bypass isolation valve was discovered leaking borated water. No adjustnent was left on packing. New packing was installed.
(WR B223346).
03/16/88, 2-VLV-062-0608-S, 02/24/89 While performing SI-146.1 during unit 2 shutdown, the CVCS number 1 seal leakoff bypass valve was leaking borated water. No adjustment was left on packing. New packing was installed. (WR B223346).
e 03/16/88, 2-VLV-062-0660-S, 02/01/89 While performing SI-146.1 during unit 2 shutdown, the alternate charging check valve was discovered leaking borated water. 'Ihis valve is in the CVCS. 'Ihe bonnet stud was bent and bonnet seal was old. A new bonnet seal-ring and stud were installed. (WR B279034).
03/16/88, 2-VLV-062-0717-S, 01/31/89 During unit 2 shutdown and performance of SI-146.1, the CVCS alternate charging check valve was discovered leaking borated water because of age. A new bonnet seal-ring was installed. (WR B279034).
03/16/88, 2-VLV-068-0535-S, 03/12/89 'Ihe RCS loop 3 cold leg manifold isolation valve was discovered leaking borated water during performance of SI-146.1. No adjustment was left on packing. Cleaned stuffing box and installed new packing.
(WR B279038).
03/16/88, 2-VLV-063-0583-S, 01/30/89 During unit 2 shutdown while performing SI-146.1, the SIS boron 1 injection valve cold Icg loop 2 was found to be leaking borated '
water. No adjustment was left on packing. Baron was cleancd fram valve, and new packing was installed. (WR B279040).
l l l- -)
Mechanical Maintenance Report j l
\
04/08/88, 2-VLV-063-0610-S, j 02/16/89 By routine observation during unit 2 shutdown, the SIS accu:m.11ator 1 j l fill isolation valve was discovered leaking borated water. No j adjustnent was left on packing, so new packing was installed.
(WR B279524).
04/26/88, 2-KN-063-0111, !
03/05/89 By routine observation during unit 2 shutdown, the SIS check valve leak test isolation valve was discovered leaking through. Cause was unknown. No apparent problem was found when valve was disassembled.
A new bonnet gasket aM packing were insta.11ed. (WR B271239) .
07/15/88, 2 ' ICV-067-0158, 08/13/88 During unit 2 operation the ERCW shutdown board room air conditioner condensate supply control valve failed closed aM will not open.
'Ihis valve needs to be open to perform SI-566, "ERCN Flcw Verification Test, - Units 0, 1 and 2." Valve had sludge above the top surface of the top of the diaphragm. Valve was dirty and corroded. Installed new diaphragm and o-rings in valve. Irstalled 1/2" plug where control tubirq normally is. Did not install pilot valve internals. (WR B261668).
08/15/88, 2-K.V-001-0015-A, 03/04/89 During unit 2 shutdown aM routine observation, the main steam system AEW pu::p turbine steam supply valve from steam generator 1 was discovered to be leaking. No adjust 2nent was left on packing.
Cleaned stuffing box ard installed new packing. (WR B789196) .
08/15/88, 2-FCV-001-0016-A, 03/04/89 During unit 2 shutdown aM routine observation, the main steam system AEW pt.mp turbine steam supply valve from steam generator 4 was discovered to be leaking. No adjustment was left on packing. ;
Cleaned stuffing box aM installed new packing. (WR B789196) . l 09/27/88, 1-VLV-070-0696A-A, 11/23/88 During unit 1 shutdown and routine observation, the CCS return valve from the RCP oil cooler was discovered to be leaking. 'Ihe packi.g had no more adjustment left. Installed new packing, cleaned all parts ard lubricated. (WR B789865) .
I i
Mechanical Maintenance Report 09/27/88, 1-VLV-070-0696B-A, 11/23/88 During unit 1 shutdown the CCS return valve from RCP motor 2 oil cooler,was discovered leaking during routine observation. Packing had no nore adjustment left. Installed new packing, cleaned all parts and lubricated. (WR B789865) .
10/24/88, 1-VLV-001-0512, 03/27/89 During unit 1 shutdown while performing SI-111, "Tes.ing and Setting of Main Steam Safety Valves, - Units 1 and 2," the main steam safety valve was discovered leaking past the seat. Leakage was because of I scale in system. Also, there was nozzle damage during disassembly of !
valve. Installed new nozzle and lapped disc insert and nozzle. m is valve was not reinstalled; it was returned to power stores to bomm a spare. A new valve was installed. (WR B769839).
01/19/89, 2-FCV-062-0070-A, 02/04/89 During unit 2 cycle 3 refueling outage, routine observation discovered that the CVCS reactor coolant loop 3 letdown flow control valve was not opening because of a blown diaphragm. Operator age and condition may have contributed. Installed new diaphragm and replaced packing in operator. (WR B797646).
01/19/89, 2-FCV-063-0080-A, 03/05/89 During unit 2 cycle 3 refueling outage, the SIS accumulator tank 3 flow isolation valve was discovernd to be leaking borated water.
Packing had been adjusted down. Cleaned stuffing box and added packing to valve and adjusted. (WR B238800) .
01/19/89, 2-VLV-063-0582-S,
'01/30/89 During unit 2 cycle 3 refueling outage, routine observation of the SIS boren injection valve cold leg loop 1 detected leakage. No adjustment was Jeft on packing. Cleaned boron from valve and repacked valve. (WR B238799).
01/23/89, 2-FCV-068-0305-A, 02/20/89 During unit 2 cycle 3 refueling outage, the pressurizer relief tank was routinely observed increasing in pressure. W e RCS nitrogen manifold flow control valve was leaking through because of normal wear. Installed air jumper, rebuilt valve per maintenance instruction (MI) 11.7.2, " Air Operated Grinnell Diaphragm Valve Rebuilding With Air Operator Model No. 3225 (Air 'Ib Open) For All Systems." Removed air jumper and reinstalled air line.
(WR B238868).
I I
Mechanical Maintenance Report 01/27/89, 2-FCV-067-0088-B, 03/09/89 During unit 2 cycle 3 refueling outage while performing SI- 158.1,
" Containment Isolation Valve Leak Rate '1bst, - Units 1 aM 2," the ERCW lower containment cooler A discharge isolation valve outside of containment failed leak rate test. 'Ihere was a cut on the rubber seat and limitorque was out of adjustment. Cleaned seat with fine sa@por. Replaced grease in limitorque operator and MOVATS test was performed on WR B292072. (hR B753972).
01/31/89, 2-VLV-061-0692, 02/02/89 During unit 2 cycle 3 refueling outage, the ice condenser system ,
glycol bypass check valve failed leak rate test while performing I SI-158.1. Seating surface had trash on it, and the spring was bad.
Iapped seat, installed new spring and hcilder assembly. (WR B271353) .
01/31/89, 2-VLV-061-0745, 02/02/89 During unit 2 operation while performing SI-158.1, the ice condenser glycol bypass check valve failed leak rate test because of trash on seat. Die. nee:ennhled top of valve, removed spring, installed red rubber disc to block valve closed. Lapped seat and installed new spring ard holder assembly. Pone.e. ambled valve. (WR B271354).
02/02/89, 2-VLV-067-0585B-B, o
03/25/89 During unit 2 cycle 3 refueling outage while performing SI-158.1, the ERCW return upper containment cooler check valve failed leak rate test Wance of rust on seat and disc. Cleaned body and internals, lapped seat and disc and installed spring. (WR B271355).
02/06/89, 2-VLV-072-0513-B, 02/15/89 During unit 2 cycle 3 outage while performing SI-164, " Testing Setpoint of Safety Relief Valves (ASME Section XI Category C Valves)
Unit 0, 1 and 2," the containment spray system pump suction relief valve was discovered leaking through because of dirty internals.
Cleaned internals, lapped seating surface and performed SI-164 as post maintenance test and it was acceptable. (WR B784038).
02/24/89, 2-VLV-063-0643-S,
- 03/06/89 During unit 2 cycle 3 refueling outage, the RHR heat exchanger discharge check valve was discovered to be leaking borated water because of age. Installed new bonnet gasket and cleaned boron off valve body. (WR B759785).
a- - , - - - - - - , _ - - - - - - - - - - - - - - - - - - - - -
Mechanical Maintenance Report 03/27/89, 2-VLV-068-0530-S, 03/14/89 Durig unit 2 cycle 3 refuelirg outage, the RCS loop 3 hot leg manifold isolation valve was discovered to be leakirg borated water.
Cause was attributed to age. Installed new bonnet gasket, replaced valve and cleaned' boron from valve. (WR B233093).
03/03/89, 2-VLV-070-0553A-A, l 03/04/89 During unit 2 cycle 3 outage, the CCS return camponent cooling ptmp mechanical seal cooler water f1w was not being detected on the flow indicator (2-FI-70-146) . 'Ihis was found during routine observation ,
and was caused by scale and corrosion from piping in amals. Removed valve internals, cleaned all parts. IJubed and reasseZDled valve.
Torqued to 400 foot pouMs and installed new diaphragm. No maintenance was performed on flow indicator. (WR B2E "354) .
03/14/89, 2-VLV-062-0660-S, 03/22/89 During unit 2 cycle 3 refueling outage, the CVCS alternate charging check valve was discovered to be leaking at the bonnet. Were were scuff rarks on internal parts and one on seatirg surface. Were was also minor pitting on bonnet. Cleaned valve internals, lightly polished seating surfaces on valve bonnet and valve body.
Reinstalled valve bonnet with new seal ring. (WR B275583) .
03/16/89, 2-FCV-002-0035A, 03/19/89 During unit 2 cycle 3 refuel outage, the coMensate system glaM seal heat exchanger recirculating flow control valve was discovered to be leaking from valve body. Se body of the valve had a crack in it. Removed valve from system, welded body and reinstalled valve. (WR B795767).
03/17/89, 2-VLV-067-0575A-A, 03/22/89 During unit 2 cycle 3 refuel outage, the ERCW return level control valve cooler check valve was discovered leaking thruugh during routine observation. S e seating surface was dirty. Di w e bled valve, installed rubber blocks, reinstalled top of valve.
Disassembled valve, removed rubber block, cleaned seating surface aM reassembled valve. (WR B775973).
i 03/19/89, 2-VLV-067-0573A-A, l 03/24/89 During unit 2 cycle 3 outage while performing SI-156, " Containment Integrated Icak Rate Test - Units 1 and 2," the ERCN return level control valve cooler valve was discovered leaking through because of dirty internals. Iapped seat aM nozzle, cleaned internals and '
l testod valve with water. (WR B775733) .
{ l L____-__-__--_-_.---------.--------_------------------_-_---------- - - - - - - - - - - - - - - - - - - - -
I i
Mechanical Maintenance Report 03/20/89, 2-TRB-001-0034, 03/21/89- During unit 2 cycle 3 refuel outage, the main feedwater pump 2A turbine was discovered during routine observation to have a broken ruptured disc. mis was attributed to normal wear. Installed new ruptured disc. (WR B769462).
03/22/89, 2-INP-002-0020, 03/25/89 During unit 2 cycle 3 refueling outage, the condensate ' system hotwell pump 2C was discovered during routine observation to be leakirq at the seal. me retainer ring was bent out of place. A new mechanical seal was installed. (WR B256624). ]I 03/26/89, 2-VLV-062-0649-S ,
03/29/89 During. unit 2 cycle 3 refueling outage while performirg SI-632.4.3,
" Auxiliary Building Chemical and Volume Control System Unit 2 Train A External leakage - Unit 2," the CVCS seal water heat exchanger relief valve was leaking. Se seat was slightly dirty, arrl the outlet was full of water. Cleaned internals, lightly lapped seats and installed new bellows assembly. (WR B775280).
04/02/89, 2-FLV-001-0182-B, . .
04/03/89 During unit 2 cycle 3 rufueling outage while performiry SI-166.6,
"'Dasting of Category 'A' and 'B' Valves After Maintenance or Upon Release From a . Hold Order - Units 1 and 2," the main steam system steam generator 2 blowdown isolation valve inside containment failed stroke time. W e stem nut was loose. his is a generic Masonellan valve rotation problem. Tightened nat and installed locktite.
(WR B256255) .
04/03/89, 2-FLV-062-0093, 04/04/89 During unit 2 cycle 3 refueling outage, the CVCs charging header flow control valve was suspected to have a packing leak. m is was fourri during routine observation of temperature readings on the packing leakoff line. W e cause was attributed to the fact that the packing had no' adjustment left. Installed new packing. (WR B256257).
04/18/89, 2-INP-003-0118, 04/18/89 During unit 2 cycle 3 refueling outage,the MDAFWP 2A-A was discovered durirg incidental observation to be leaking at the outboard end of pump. No adjustment was left on packing. Cleaned stuffing box and installed new packing (WR B797429).
I l
Mechanical: Maintenance Report
'04/19/89, '2-FMP-003-0128, t 04/20/89 During the restart of unit 2 following an automatic trip because of feedwater level control, MDMWP 2B-B was discovered during incidental observation with no packing adjustment left. 'Ihe packirg blew out before the. pump was removed from service for repacking. Ioa .
flow characteristic as a result of plant corditions cane.ui the pump discharge check valve to chatter. 'Ihis valve chatter caused higher than normal pressures in the outboard stuffing box causing the packing to fail. Installed new packing and added oil to inboard and outboard bearirgs. Added 12 ozs. of oil. (WR B282552) .
t l !
l 4
1 4
N MAINTENANCE
SUMMARY
(MODIFICATIONS)
L e__-____-_-_____-____-_____=-________-____________-______-____-_-_____-__---__________________--___-____-
l.-
p Maior Capital Projects:
PN7108: ECN 6720 - Crane Consistency Program Unit 1 polar crane limit switch weights remain to be painted.
Completion'is scheduled for unit 1 cycle 4 (UlC4).
Completed cranes . unit 1 Turbine Building 15-ton' crane, unit 2
. polar crane, unit 2. Turbine Building 200-ton crane, unit 2 i
' Turbine Building 15-ton' crane, Service Building 5-ton crane, and Turbine Building 10-ton hatchway crane, i
I Remaining listed crancs-are to be modified after unit 2 cycle 3
-(U2C3) - Auxiliary; Building 125-ton crane, waste packaging crane,Leallroad bay crane, and unit 1 Turbine Building 200-ton crane.
I' PN7130: ECN 6180 - Postaccident Monitoring Work is currently being held for material.
PN7132: DCN 0026 - Sewage Treatment Facility and Civil Upgrade All work is complete.
PN7161: ECN 5855 - Replacement of Doors A56 and A57 '
WP 09679 remains on hold and is partially complete, i i
Other Items:
ECN 5111 - Provide Permanent Power to Manholes 42-46 Work has stopped because of lack of funding. The cable has been run ,
from breaker 4E at the 480-V common board in the Turbine Building through !
manhole 1 to manhole 42. All material.was purchased before the job was -
stopped for lack of funding. . Work'is 45 to 50 percent complete. l
.i ECN 5503 - Evacuation Alarms O&PS/ Fire Detection O&PS WP 12482 - Work has stopped because of lack of funding.
ECN 5552 - Condensate Demineralized Modifications and High Crud Filter WP 5552 Fieldwork is complete.
i 1
Other Items (cont.):
ECN 5609 - Alteration to the Makeup Water Treatment Plant WP 12387 - Work is 90 percent complete. WP is now in work, t
i WP 12576 - Work is complete. WP revision is required before closing.
WP 12633 - Work is approximately 90 percent complete because of redesigns.
WP 12665 - Work is field complete. WP is in final closure.
WP 12682 - WP is 80 percent complete. Awaiting installation of a pump i for the alum sludge pond.
WP 12684 - WP is field complete. Equipment calibration and functional tests remain. j
)
WP 12731 - WP is approximately 97 percent complete. MODS needs approval of FCR 8211 by Systems Engineering.
ECN 5626 - Containment Ladders, Unit 1 MODS needs additional design information to complete. NE needs to issue all drawings listed on this ECN. Work has not begun because of this holdup.
ECN 5841 - Hot Shop Fire Protection / Evacuation Alarm WP 12360 is field complete. Awaiting drawings to be updated.
ECN 5911 - Waste Disposal Piping Addition System tie-ins are in progress and scheduled to be completed by May 15, 1989.
ECN 5916 - Replacement of Cask Decon Collector Tank Pumps Fabrication and installation of waste disposal piping is in progress.
ECN 5935 - Correct Power Block Lighting Deficiencies WP 12437 is complete. WP 12275 is complete. WP 5935-01 is field complete. Waiting on secondary drawings to be revised. ;
WP 5935-02 is field complete.
ECN 5977 - Install Steam Generator Blowdown Demineralized System tie-ins are planned for refueling outage.
n-------_-_-_-
Other Items (cont.):
ECN 6357 - ERCW Roof Access and Rails for Security Equipment original design for WP 12238 was rejected by Operations. NE to rework design to comply with Operations' needs and attempt to salvage existing work.
ECN 6388 - Hydrogen Monitors in Switchyard WP.12223 - Work has been stopped because of a lack of funding.
ECN 6429 - Component Cooling Heat Exchanger Replacement Work on heat exchanger B is complete, except for insulation and painting. Work on heat exchanger C will resume as soon as funding is approved.
ECN 6815 - 500-kV Switchyard Addition A 30-liter air replenishing tank had to be purchased to replace a defective tank on PCB 5018. This tank is scheduled to be delivered by May 15, 1989. Receipt of this tank will allow continuation of PCB testing. Software changes remain to be done for the data logger.
Testing remains for the switchyard data acquisition. A decision has been made to retire the 161-kV equipment in bay 20 by leaving it in place.
This decision will require revision of design drawings by means of an FCR. FCR 8087 has been submitted to NE to revise the drawings. Awaiting FCR approval.
DCN 214 - AFW Tap Rotation Work is complete, PMT required.
ECN 7328 - Installation of Backflush Line Spent Rosin Storage Tank Drain Line WP is in review cycle.
ECN 7349 - Removal of Temporary Bull Hose from CDWE to Floor Drain Collector Tank Fabrication of new piping is in progress and is scheduled to be completed May 19, 1989.
DCN 341 - Modify Pressurizer Safety Valves to Accept Steam Trim and Install Loop Seal Drains - Unit 1 All work is complete except touchup paint and grouting of one baseplate (to be done during UIC4 outage).
. . ;;. .1
)
- 'Other Items (cont.): ,
. DCN 550'- Modify' Pipe Support
' System 68, Unit 1
--All work is complete'except touchup paint (to be donc during UIC4 outage).
DCN 703 . Modify RCP Seals - System 68, Unit 2 All work'is complete on-RCP Nos. 3 and 4.
DCN'704 - Modify Pipe Support - System 68, Unit 1 All work is complete.except touchup paint'(to be done during UIC4 outage).
DCN 943 - Addition of Stiffeners to Penetrations - System 61, Unit 2'.
All work is complete.
~
DCN 1045 - Modify Pressurizer Safety Valve Discharge Pipe Support - Unit 1 All work.is. complete except for touchup paint (to be done during UIC4 outage)' .
1 l
L 1
l l-c_ - - -
k, '
,.. .i
{ .s ,..
I l
l I
[.
1, e
GLOSSARY
GLOSSARY OF VARIOUS ABBREVIATIONS Page 1 of 3
- 2. ABSCE .- Auxiliary Building Secondary Containment Enclosure
- 3. AB(I) - Auxiliary, Building (Isolation)
- 4. AFW - Auxiliary Feedwater j
- 5. AOI - Abnormal Operating Instruction
- 6. ASOS - Assistant Shift Operations Supervisor ,
- 7. AUO - Assistant Unit Operator
- 8. BAE - Boric Acid Evaporator jl
- 9. BAT - Boric Acid Storage Tank
- 10. BIT - Boron Injection Tank
- 11. CAQR - Condition Adverse to Quality Report
- 12. CAR - Corrective Action Report
- 13. CCP - Centrifugal Charging Pump
- 14. CCS ~ - Component Cooling System
- 15. CCW - Component Cooling Water
- 16. CDWE - Condensate Demineralized Waste Evaporator
- 17. CRI - Control Room Isolation
- 18. CREVS - Control Room Emergency Ventilation System
- 19. CSS (CS) - Containment Spray System
- 20. CVCS - Chemical Volume and Control System
- 21. CVI 22.
- Containment Ventilation Isolation D/G(s) - Diesel Generator (s)
- 23. DCN - Design Change Notice
- 24. DCR - Design Change Request
- 25. DR - Discrepancy Report
- 27. ECN - Engineering Change Notice
- 28. EGTS - Emergency Gas Treatment System l
- 29. EM - Electrical Maintenance
- 30. EMI - Electromagnetic Interference
- 31. EQ - Environmentally Qualified / Environmental Qualification
- 32. ERCW - Essential Raw Cooling Water
- 33. E/ES - Emergency Insttvetion
- 34. ESF - Engineered Safety Feature
- 35. ESPA - Engineered Safety Feature Actuation
- 36. FCR - Field Change Request
- 37. FCV - Flow Control Valve
- 38. FDCT - Floor Drain Collector Tank
- 39. FDS - Flow Differential Switch
- 40. FIC - Flow Indicating Controllers
- 41. FSAR - Final Safety Analysis Report
- 42. FS - Flow Switch
- 43. FWI - Feedwater Isolation
- 44. COI - General Operating Instruction
- 45. CPM - Gallons Per Minute
- 46. HDTP - Heater Drain Tank Pump
- 47. HO - Hold Order
- 48. IM
- 49. IMI
- Instrument Mechanic / Instrument Maintenance
- 50. - Instrument Maintenance Instruction LCV - Level Control Valve I
GLOSSARY OF VARZOUS ABBREVIATIONS Page 2 of 3
- 51. LER - Licensing Event Report
- 52. LCO - Limiting Condition for operation
- 53. LOCA - Loss Of Coolant Accident
- 54. LS - Level Switch
- 55. M&TE - Measuring and Test Equipment 56, mA - Milliampere
- 57. MAST - Maximum Allowable Stroke Time
- 58. MCR - Main Control Room
- 59. MDAFWP - Motor-Driven Auxiliary Feedwater Pump
- 61. MWF - Main Feedwater
- 62. MFWRV - Main Feedwater Regulating Valves
- 64. MI - Maintenance Instruction
- 65. MODS - Modifications
- 66. MOV - Motor Operated Valve
- 67. MSI - Main Steam Isolation
- 69. MSR - Moisture Separator Reheaters
- 70. NE
- Nuclear Engineering (formerly Division of Nuclear Engineering)
- 71. HIS - Nuclear Instrumentation System
- 72. NMUDI - New Makeup Deionized System
- 73. NSS - Nuclear Security Service
- 74. NSSS - Nuclear Steam Supply Systems
- 75. O&PS - Office and Power Stores Building
- 76. PM - Preventive Maintenance
- 77. PMT - Postmodification Test
> 78. PORC - Plant operations Review Committec
- 79. PORV - Power-Operated Relief Valve
- 80. PRO - Potential Reportable Occurrence
- 81. PDS - Pressure Differential Switch
- 82. QMDS - Qualification Maintenance Data Sheet
- 83. RCS/(P) - Reactor Coolant System /(Reactor Coolant Pump)
- 84. RMR - Residual Heat Removal
- 85. RM - Radiation Monitor (RAD Monitor / RAD MON)
- 86. RPI - Rod Position Indicator
- 87. RWST - Refueling Water Storage Tank
- 88. SCR - Significant Condition Report
- 89. S/D - Shutdown
- 90. SFP - Spent Fuel Pit
- 91. S/G(s) - Steam Generator (s)
- 92. SI - Surveillance Instruction /or Safety Injection
- 93. SMI - Special Maintenance Instruction
- 94. SOS - Shift Operations Supervisor
- 95. SOI - System operating Instruction
- 96. SQN - Sequoyah Nuclear Plant
- 97. SR - Surveillance Requirement / Source Range
- 98. SSPS - Solid State Protection System
- 99. TACF - Temporary Alteration Control Form 100. TI - Technical Instruction 1
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ __ _ i
GLOSSARY OF VARIOUS ABBREVIATIONS Page 3 of 3 101. TS(s) - Technical Specification (s) 102. TVA - Tennessee Valley Authority 103. UHI - Upper Head Injection 104. UO/(S)RO - Unit Operstor/(Senior) Reactor Operator 105, VLV - Valve -
106. ' WP . - Workplan 107. WR - Work Request I
l 1
1 j
1 i ,
li GLOSSARY OF VARIOUS SYSTEMS OF SEQUOYAH NUCLEAR PLANT SYSTEM CODE SYSTEM TITLE
.1
- Main Steam System (Turbine) -(MSR) -
2 Conde'nsate System (FW Heaters) 3 Main and Auxiliary Feedwater System 5 Extraction Steam System 6 Heater Drains and Vents System 14 Condensate Demineralized 15 Steam Generator Blowdown System.
- 24. Raw Cooling Water System 27 Condenser Circulating Water System' 30 Ventilating System 35 Generator Cooling Systems 36 Feedwater/ Secondary Treatment System 37' Gland Seal. Water System 46 Main / Auxiliary Feedwater Control System 47 Turbogenerator Control System 54 Injection Water System-58 Generator Bus Coolins System 61 Ice Condenser System 62 Chemical and Volume Control System 63 Safety Injection System 64 Ice Condenser Containment System 65 Emergency Gas Treatment System 67 Essential Raw Cooling Water System 68 Reactor Coolant System (Steam Generator) 70 Component Cooling System 74 Residual Heat Removal System 82 Standby Diesel Generator System
-87 Upper Head Injection System 90 Radiation Monitoring System 268 Hydrogen Mitigation System
= _ _ _ _ _ _ _ _ - - _ _ - _ _ _ _ _ _ _ _ _ _
OPERATIONAL MODES (
% RATED AVERAGE COOLANT MODE TIIERMAL POWER TEMPERATURE
- 1. POWER OPERAT' ION > 5% > 3500F
- 2. STARTUP < 5% > 3500F
- 3. liOT STANDBY 0 > 3500F
- 4. HOT SliUTDOWN O 3500 F >Tavg
> 2000F
- 5. COLD SHUTDOWN O < 2000F
- 6. REFUELING 0 < 1400F t
m1___ _ _ __:-- _m
- e. ,
e
- h. .
i THIS PAGE INTENTIONALLY LEFT BLANK I
1
~
v
' M;: A _
l l
l
~
b TENNESSEE VALLEY AUTHORITY q q
L ,
5N 157B Lookout Place j e i 1 NAY 121989 L ,
i- U.S., Nuclear Regulatory Commission. 3 ATINs . Document' Control Desk- l Washington,'D.C. '20555 j 1
Gentlemen:.
In'the Matter of- ) Docket Nos. 50-327 !
Tennessee Valley Authority ) 50-328 I SEQUOYAH NUCLEAR PLANT (SQN) - APRIL 1989 MONTHLY OPERATING REPORT Enclosed.is the April 1989 Monthly Operating Report as required by SQN- )
Technical Specification-'6.9.1.10. i l
If you have any questi"ns concerning this matter, please call R. R. Thompson at (615) 843-7470.
1 Very truly yours, )
l TENNESSEE VALLEY AUTHORITY l v Manager.s uclear. Lice sing and Regulatory Affairs Enclosure l cc (Enclosure):
Director, Region II INPO Records Center ,
Nuclear Regulatory Commission Suite 1500 i Office of Inspection and Enforcement 1100 circle 75 Parkway- q Suite 3100 Atlanta, Georgia 30323 l 101 Marietta Street Atlanta, Georgia 30339 Sequoyah Resident Inspector Sequoyah Nuclear Pla~nt Depucy Executive Directa 2600 Igen Ferry Road for Regional Operations Soddy Daisy, Tennessee 37379 :
Nuclear Regulatory Commission Washington, D.C. 20555 Mr. T. Marston Electric-Power Research Institute P.O. Box 10412 Palo Alto, California 94304 ..
J. G. Partlow, Director TVA Projects Division Office of Special Projects Washington, D.C. 20555 An Equal Opportunity Employer oi