ML20245B706
| ML20245B706 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/31/1989 |
| From: | Dupree D, Fox C, Shawn Smith TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8904260235 | |
| Download: ML20245B706 (55) | |
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TENNESSEE VALLEY AUTHORITY NUCLEAR POWER GROUP SEQUOYAH NUCLEAR PLANT 1
l MONTHLY OPERATING REPORT TO THE NUCLEAR REGULATORY COMMISSION l
MARCH 1989 l
UNIT 1 DOCKET NUMBER 50-327 LICENSE NUMBER DPR-77 UNIT 2 DOCKET NUMBER 50-328 LICENSE NUMBER DPR-79 Submitted by:
S. J. gt61th, Plant Manager y
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l TABLE OF CONTENTS Page I.
Operational Summary Performance Summary 1
Significant Operational Events 2-4 Fuel Performance 5
Spent Fuel Storage Capabilities 5
PORVs and Safety Valves Summary 5
Special Reports 5-6 Licensee Events 7-9 Radwaste Summary 10 Offsite Dose Calculation Manual Changes 10 II.
Operating Statistics A.
NRC Reports Unit 1 Statistics 11-13 Unit 2 Statistics 14-16 B.
TVA Reports Nuclear Plant Operating Statistics 17 Unit 1 Outage and Availability 18 Unit 2 Outage and Availability 19 Unit 1 Reactor Histogram / Analysis 20-21 Unit 2 Reactor Histogram / Analysis 22-23 III. Maintenance Summary Maintenance 24-34 Modifications 35-38 IV.
Glossary Common Abbreviations and Systems of Sequoyah Nuclear Plant 39-41 Operational Modes 42
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OPERATIONAL
SUMMARY
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PERFORMANCE
SUMMARY
March 1989 The following summary describes the significant operational activities for the month of March.
In support of this summary, a chronological log of significant events is included in this report.
Unit 1 The unit continued to operate well and was online the entire month.
For the month, it generated 865,210 MWh with a capacity factor of 98.30 percent.
Unit 1 has been in continuous operation for 47 days as of March 31, 1989.
Unit 2 On~ March 2,1989, at 1725 (EST), all fuel assemblies were installed in the reactor vessel.
This was a significant step during the cycle 3 refueling outage.
A major milestone was reached on March 9,1989, at 0001 (EST), when the unit entered mode 5, bringing to conclusion the cycle 3 refueling activities.
Extensive testing, system alignment, and various other activities are now in progress for power operation scheduled in April 1989.
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SIGNIFICANT OPERATIONAL EVENTS Unit 1 Date Time (EST)
Event 03/01/89 0001 Reactor at 100 percent power, 1,173 MW.
1B-2 e
waterbox tagged and drained for condenser tube ir.spection.
0620 IB-2 waterbox back in service.
1030 Condenser waterbox 1B-1 removed from service for tube inspection.
2334 1,176 MW,
e 03/02/89 1145 IB-1 condenser waterbox is back in service.
1300 Reactor at 100 percent power, 1,183 MW.
e 03/10/89 1230 Began power decrease to 75 percent for maintenance on 1C unit board.
1530 Reactor power at 75 percent, 890 MW.
e 1821 Started power increase to 100 percent.
03/12/89 0005 Holding at 90 percent power, 1,092 MW for reactor e
power verification.
0038 Test complete, starting load increase to 100 percent.
0600 Reactor at 100 percent power, 1,188 MW.
e 1738 Control rod E-13 RPI inoperable.
1819 Control rod E-13 operable.
03/15/89 1555 RPI B-10 inoperable.
1620 RPI H-8 inoperable.
03/18/89 0844 Isolated main steam valves to 1C MSR to repair a steam leak on MSR, 1,176 MW.
e 1120 Placing 1C MSR back in service.
Steam leak repaired.
1300 Reactor at 100 percent power, 1,183 MW.
e 1730 RPI E-13 RPI drifting up and down, declared inoperab,1,e.
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SIGNIFICANT OPERATIONAL EVENTS Unit 1 Date Time (EST)
Event 1
03/19/89 0015 RPI E-13 back operable.
1140 Received main turbine-trip alarm and various associated multiple annunciators - verified no l
03/20/89 1040 l
Received turbine-trip alarm and various alacms -
verified again no turbine / reactor trip.
1057 Received turbine-trip alarm and various turbine alarms - verified no trip.
03/21/89 1703 RPI E-13 inoperable.
Drifts up and down.
03/22/89 0024 RPI E-13 operable after maintenance.
03/25/89 0734 100 percent reactor power, 1,190 MW -
e 03/29/89 0805 RPI E-13 inoperable, drifting again.
1719 RPI E-13 operable after maintenance.
03/30/89 1500 RPI H-8 began to drift low.
Declared inoperable.
1800 RPI H-8 operable after maintenance.
03/31/89 2400 Reactor at 100 percent power, 1,185 MW.
e Unit 2 03/01/89 0001 Fuel movement in progress. Approximately 40 assemblies have been loaded in the reactor vessel.
.03/02/89 0600 Assemblies (153) have been loaded.
1725 Fuel loading completed - 193 assemblics were loaded.
03/03/89 0514 Started latching control rods.
0720 All control rods latched.
03/04/89 0555 Started reducing RCS water level.
03/06/89 2300 Lifted head for final inspecticn.
2309 Head lowered back on vessel after inspecting stud holes.
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p SIGNIFICANT OPERATIONAL EVENTS Unit 2'
_Date-Time (EST)
Event 03/08/89 0513 Reactor vessel. head tensioning in progress.
2400 All reactor head tensioning complete.
1 03/09/89 0001 Entered mode S.
l 03/10/89 0615 Started raising RCS level.
0842 Stabilized RCS level.
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1551 RCS at 118 F.
03/11/89 0129 Began venting, sweeping, and pressurizing the RCS at-various times.
03/13/89 0000 Venting and sweeping at various times and pressures continued.
2 03/22/89 1315 RCS at 325 lb/in, venting and sweeping continues.
03/25/89 1059 Received an' unanticipated reactor-trip signal with an SI signal while performing maintenance on high steam flow instrumentation.
1552 Received another reactor-trip signal with an SI signal while performing maintenance on high steam flow instrumentation.
1717 Steam bubble established in pressurizer.
03/26/89 0103 Started RCS heatup to 170 F.
0 1640 Mode 5, various maintenance and startup activities are in progress.
2 2008 RCS pressure 383 lb/in, RCS temperature 1750F, 03/27/89 0300 Began RCS cooldown.
0630 The pressurizer heatup rate exceeded 1000F/ hour while cooling down the RCS.
Terminated cooldown.
1639 Venting and sweeping continue, along with various maintenance activities being performed on the RCS.
03/30/89 1647 Mode 5, RCS at 1330F, 8 lb/in.
2 1700 Began RCS heatup.
03/31/89 2400 Mode 5, RCS at 178 F, 350 lb/in. Maintenance
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activities continue for startup.
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4 FUEL PERFORMANCE Unit 1 The core average fuel exposure accumulated during March was 1,172 mwd /MTU, with a total accumulated core average fuel exposure of 3,500 mwd /MTU.
Unit 2 The unit 2 cycle 3 refueling outage continues.
SPENT FUEL PIT STORACE 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,
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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
Unit 1 No PORVs or safety valves were challenged in March.
Unit 2 PORV PCV-68-340 momentarily popped open when its block valve was opened.
This was because the RCO was pressurized and the block valve prohibited the pressure from reaching the PORV. When the block valve was opened, the surge of RCS pressure was great enough to actuate the PORVs.
This event occurred on March 26, 1989.
_SPECIAL REPORTS 89-04 Between February 10 and February 28, 1989, with unit 2 in mode 6, a fire barrier (shield building wall, between el 698 and el 772 and azimuth 172 degrees and 355 degrees) was nonfunctional for an interval greater than seven days.
Between F ^ ruary 3 and February 21, 1989, within the area of the above specified shield building wall section, 14 mechanical penetration seals of various sizes were removed in accordance with WP 7095-01. Removal of the seal material from the penetrations resulted in the fire barrier being nonfunctional; and as required by LCO 3.7.12, an hourly firewatch was established within one hour and will be maintained until the penetrations are restored to functional status.
Because of scheduling difficulties, the fire barrier remained nonfunctional for greater than seven days. Modifications to the chield building mechanical penetration seals are being performed to fulfill commitments to the NRC.
(reference LER 50-326/87040).
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SpECIAL REPORTS 89-04 A roving firewatch was established to inspect the affected area (cont.) of the shield building wall on an hourly basis, as required by the action statement of LCO 3.7.12.
The existing fire-detection system on one side of the affected penetrations was verified operable and would have activated in the event of a fire. Therefore, there was no danger to safety-related equipment.
89-05 On March 14, 1989, with unit 1 at 100 percent power and unit 2'in mode 5, a fire barrier (door A-158a, Auxiliary Building, el 734 ECTS filter-housing room) was nonfunctional for an interval greater than seven days. The EGTS is common to both units, and both trains j
j of filter housing are located in the same room. The door _is a I
three-hour-rated, overhead, rolling fire door.
During performance of SI-261, " Visual Inspection of Technical Specification Fire Doors on a Periodic Basis," the latch lever broke and rendered the fire
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door inoperable. The SI ensures that in the event of a fire, the overhead rolling fire door will operate as designed. This is done by disconnecting the fusible link (that melts during a fire) from the latch lever mechanism and allows the door to clcse. While disconnecting the fusible link, the latch lever slipped after the door started closing and reengaged into the gears causing the latch lever to break.
A WR was issued to repair the door, and an hourly firewatch was established, as required by TS LCO 3.7.12.
Subsequently, the door was rolled up and is being kept open by inserting a lock bolt through the gear mechanism. Maintenance Support determined that the replacement (latch lever) parts could not be procured in seven days because the original manufacturer of the fire door is no longer in business.
As a result, the fire door could
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not be repaired and returned to service in seven days, which requires a special report in accordance with TS LCO 3.7.12.
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room on an hourly basis, as required by the action statement of LCO 3.7.12.
The existing fire-detection and fire-suppression systems for the room are operable and would activate in the event of a fire.
In addition, the door opening has a heavy equipment door similar to other heavy equipment doors that have been evaluated as equal to a three-hour fire door. This heavy equipment door is controlled by the plant security key card system. Therefore, there is no danger to safety-related equipment.
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1 A vendor, who possibly could supply the original latch level, has i
been located; but the parts delivery has not been confirmed.
A procurement request has been initiated to purchase the required parts, and efforts are presently being pursued to procure the parts from available sources. WR B-282544 will be used to document the I
door repair.
It is anticipated that the door will be repaired and returned to functional status by June 15, 1989 N,
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LERs The following LERs were transmitted to the Nuclear Regulatory Commission in March 1989.
Description of Event LER i
1-89003 on February 19, 1989, at 0004 (EST), with unit 1 at 100 percent reactor power, the 1A-A MDAFWP started, following the momentary interruption of control power to the 6.9-kV shutdown board 1A-A.
Before thir event occurred, Transmission and Customer Services personnel initiated SI-621, " Periodic Functional Test of Loss of Voltage Relays on Shutdown Board." At 2300 (EST), on February 9,1989, S1-621 was stopped because the normal feeder breaker 1718 for 6.9-kV shutdown board 1A-A did not trip as required.
A ground was discovered on the positive side of the 125-V de alternate control bus, which was verified by the operator by checking the ground indicator on 125-V de vital battery board III. While attempting to isolate the ground, breaker 204 on 125-V dc vital battery board III was opened, which removed power from the alternate control bus.
Breaker 204 was closed after verifying that the ground was not on the shutdown board. When power was restored on the alternate control bus, the BOY relays operated as they would in blackout condition and the UVY relay reenergized.
This completed the needed permissives that caused the MDAFWP breaker to close. The breaker closure for centrifugal charging pump 1A-A, ERCW pump J-A, and component cooling system pump 1A-A was initiated by the UVY and BOY relays in a similar manner as that of the MDAFWP; however, these pumps were already running so that the breaker closure signal was undetected. Operations personnel attempted to ensure that operation of breaker 204 would not adversely impact plant equipment. However, because of the complexity of the circuit, it was not realized that operating breaker 204 would cause ESFA.
As immediate corrective action, Operations personnel verified that the IA-A HDAFWP started while they were searching for ground on 125-V de vital battery board III.
Subsequently, the pump was shut down, the tripped relays were reset, and the pump was realigned for standby operation. To prevent recurrence of this event, TVA will issue procedures specifying the method to search and isolate grounded circuits.
1-89005 On February 10, 1989, with unit 1 at 100 percent reactor power, a reactor trip occurred at 2036 (EST).
The trip signal was a result of S/G steam flow to feedwater flow mismatch of greater than 40 percent of the nominal value of steam flow at full power coincident with a low S/G level (25 percent) signal on E/G loop 3.
Two IM technicians were implementing WR-B238429 on flow recorder 2-200/201, " Condenser Bypass / Makeup Flow," which is located on MCR panel 1-M-3.
The recorder pen needed to be restrung, which required it to be removed from the case.
The technicians fully removed the recorder in its case, which required lifting the power supply leads.
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Description of Event LER 1-89005 Arter reinstallation, the technicians reterminated the power-supply (cont.)
leads, at which time one technician determined the terminating leads were too close to each other.
The technician inadvertently shorted a screwdriver between the terminals, tripping open breaker 39 on 120-V ac vital instrument board I-II, which is the power supply to the mold supplying the recorder. The plug mold is also the common power supply to the FIC-3-35
-90, and -103 for S/G loops 1, 3, and 4.
FIC-3-48 for KFWRV on S/G loop 2 is powered from I-III.
The three MFWRVs closed, isolating MFW to the S/Gs.
This resulted in the SSPS reactor trip signal. The root cause of the reactor trip signal was personnel error, in that appropriate precautions were not taken in performing terminations of energized equipment.
To prevent recurrence of this event, the event was discussed with IM planners, technicians, and engineers to familiarize them with the root cause of the event.
Additionally, open WRs and scheduled pHs on MCR recorders were retrieved by the Work Control / Outage Group for replanning to address precautions on equipment with common power supplies.
1-89006 On February 16, 1989, with unit 1 at 100 percent power and unit 2 in mode 6, it was discovered that the Auxiliary Building waste-packaging area door A111 (el 706) had been breached without issuance of a breaching permit.
Door A111 functions both as a TS fire-barrier door and as part of the ABSCE.
Subsequent investigation into the incident found that the door had been breached open from 1741 (EST) to 1847 (EST), and from 1900 (EST) to 2130 (EST), on February 15, 1989, and for an undetermined period of time on February 16, 1989. The root cause of this event is considered to be inadequate training.
The contributing causes of this event are incomplete door identification to alert personnel that door Alli is both a fire door and part of the ABSCE boundary, inadequate programmatic controls to ensure that doors remain uniquely identified, and personnel oversight.
For immediate corrective sution, door Alli was closed on February 16, 1989.
personnel ecsronsible for the breaches on February 15, 1989, were counseled on the requirements for ABSCE boundary and fire doors.
A plant-wide dispatch was issued on February 23, 1989, explaining to personnel the importance of adherence to requirements associated with fire doors so that compliance with regulations can be ensured.
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explained to plant personnel that a fire door may be identified by either its red color or a sign stating that it is a fire door.
For long-term corrective action, TVA will evaluate increasing the scope and frequency of site-specific training associated with the requirements for breaching fire and ABSCE boundary doors.
TVA will uniquely identify all doors that are utilized as part of the ABSCE boundary. TVA will perform an evaluation of current programmatic controls to determine where changes can be made to ensure that doors retain their unique identification.
TVA has revised SOI-77.3, " Waste processing," to provide a precaution to personnel that door A111 must have a breaching permit to allow the door to be blocked open.
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Dtacription of Event LER 2-89001 At 0021 (EST), on February 12, 1989, unit 2 NIS SR channel N-31 received a spike of sufficient magnitude to actuate the high-flux trip bistable and generated a reactor trip signal. Unit 2 was in mode 6 at the time of the event, although no core alterations were in progress. At 0030 (EST), and again at 0039 (EST), channel N-31 spiked high enough to generate two more reactor trip signals.
These spikes were of a nature that could not have been caused by an actual increase in neutron flux. The unit 2 ASOS attempted to identify tha source of the interference and found welding in progress in the Auxiliary Building near the CCW system heat exchansers.
Although the actual voltage source could not be located for this event, it is well documented that welding can cause sufficient electromagnetic interference to generate a reactor trip signal on SR instrumentation.
The root cause of,the SR spiking is attributed to the susceptibility of the NIS design to electrostatic, electromagnetic, and frequency interferences. Therefore, it is concluded that the occurrence of these events had no significant adverse affect on the health and safety of the general public. Reduction of SR trip signals, caused by noise, will require long-term system hardware changes to reduce system susceptibility to noise interference. This will be accomplished by the upgrade of the existing Westinghouse NIS with Gamma-Metrics equipment.
2-83097 This revision provided updated information on the planned corrective (RA) actions for the containment sump pump level transmitters.
Because of the continued problems with air infiltration, even on double 0-ring-type transmitters, TVA has decided to replace the Barton transmitters with a different type of transmitter. Unit 2 transmitters are presently being replaced.
The design packages to replace unit 1 transmitters will be issued by June 1, 1989.
.e 4-RADWASTE
SUMMARY
March 1989 1.
Total volume of solid waste shipped offsite:
'A.
Dry active waste:
1313.2 ft3 Activity:
10.373 curies B.
Spent resins, sludges, bottoms:
170.0 ft3 Activity: 1193.74 curies Shipped:.Barnwell, Inc.
- March 17, 1989 (2)
March 23, 1989 (2)
March 28, 1989 (1) 2.
Radwaste onsite and awaiting shipment:
A.
Resin in storage:
_ 109.0 ft3 B.. Estimate resin that will be generated:
39.0 ft3 C.
Dry active waste awaiting shipment:
1143.0 ft3
.1 - Dry active waste 2 - Spent resin OFFSITE DOSE CALCULATION MANUAL CHANGES No changes were-made to the Offsite Dose Calculation Manual'for the month of March.
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OPERATING STATISTICS (NRC REPORTS) l m9 ee#
GPERATi;T, DATA RETiORT DOCKET NO, D0-027 DATE APRIL. 04.
PM9 COMPLETED DY D. I..:UPP! 1-TELEPHONE ( 615 ) N.0-67;';'
ERATING STATUS.
1.
/JIT NAME: GEGUOviaH NUCLEAP ?LANT, UNIT t
NOTE 5; 2.
WPORT PERICD: NARCH 1989
.' CENSED THERMAL POWER (MWT*
2411.O
'3.
4 aMEPLATE RATING (GROSS MWE).
1220.,6 5.
EdSIGN ELECTRICAL RATING U;ET MWE):
1148.0 6.
MXIMUM DEPENDABLE CAPACITY (GROSS MWE):
1180.0 7.
c'!AXIMUM DEPENDABLE CAPACITY ' NET MWE):
1148.O
- O.
- F CHANGES OCCUR IN CAPACIT'Y RATINGS (ITEMS NUMBERS 3 THROUGH 7)OINCE LAST PEPORT. GIVE REASONS: _,
9.
'CWER LEVEL TO WHICH RESTRICTED,IF ANY(NET MWE);
- 10. lEASONS FOR RESTRICTIONS,IT ANY:
THIS NONTH YR, - TO-DATE CUMULATIVE 11.
-CURS IN REPCRTING PERIDD 744.00 2160.00 67945.00
- 12. NUMBER OF HOURS REACTOR WA2 'RITICAL 744.00 2111.25 26935.69 13.
MACTOR RESERVE SHUTDOWN HC'JRE O.00 O. CO O.00
- 14. HCURS GENERATOR ON-LINE 744.00 2085.e6
- '6149.54 15.
NIT RESERVE SHUTDOWN HOURE O.00 O. CO O.00
- 16. " JOSS THERMAL ENERGY GENERATED (11WH) 2505304.05 6967571.73 84U44508.60
- 17. GFOSS ELECTRICAL ENERGY GEN. ;MWH) 864810.00 2400610.CO 20WlOSS6.00
- 10. NET ELECTRICAL ENERGY GENEitATED (MWH) 836290.00 2322021.00 27;?43449.00
- 19. UNIT SERVICE FACTOR 100.00 96.56 28.49
- 20. UNIT AVAILABILITY FACTOR 100.00 96.56 38.49
- 21. UNIT CAPACITY FACTOR (USING MDC NET) 97.91 93.64 34.93
- 22. UNIT CAPACITY FACTOR (USING DCR NET) 97.91 93.64 34.93
- 23. UNIT FORCED OUTAGE RATE O.00 3.44 55.60
- 24. 3HUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, AND DURATION Of-EACH):
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- 25. I.~- CHUTDOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP' NOTE THAT THE YR. -TO-DATE AND CUMUL.hTIVE VALUES HAVE DEEN UPDATED.
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,o SEQUOYAH NUCLEAR PLANT AVERAGE DAILY POWER LEVEL i
DOCKET NO. : 50-327 l
-UNIT : ONE DATE : APRIL 04,1989 COMPLETED BY :-D.C.DUPREE TELEPHONE : (615)843-6722 MONTH: MARCH 1989 AVERAGE DAILY POWER LEVEL AVERAGE DAILY POWER LEVEL DAY (MWe Net)
DAY (MWe Net) 01 1132' 17 1138 02 1132 18 1138 03-1133 19 1138 04 1139 20 1141 05 1142 21 1139 06 1142 22 1140 07 1140 23 1141 08 1141 24 1139 09-1143 25 1139 10,
1025 26 1139 11 877 27 1139 12 1128 28 1139 13 1140 29 1138 14 1139 30 1138 15 1138 31 1139 16 1138
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UPERAT U G OATA ' REPORT I
DOCKET NO. D0-328
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DATE APRI). 04, 1969 COMPLETED DY D. G. DUPRM-TELEPHONE (615)340-67PP PERATING STATUS 1.
MIT NAME: SEGUOYAH NUCLEAR FLANT, UNIT 2
NOTES:
2.
~ ::. PORT PER IOD: MARCH 1989 3.
- CENSED THERMAL POWER (MWT ) :
3411.0 4.
N6MEPLATE RATING (GROSS MWE):
1220.6 5.
DESIGN ELECTRICAL RAT!NG (NET MWE):
1148.0 6.
MAXIMUM DEPENDABLE CAPACITY (GROSS MWE):
1133.0 7.
NAXIMUM DEPENDABLE CAPACITY (NET MWE):
1148.0 8.
IF CHANGES OCCUR IN CAPACITY RATINGS (ITEMS NUMBERS 3 THROUGH 7)SINCE LAST REPORT, GIVE REASONS:
9.
.$WER LEVEL TO WHICH RESTRICTED,IF ANY(NET MWE):.
- 10. EEASONS FOR RESTRICTIONS, IF ANY:
THIS MONTH YR.-TO-DATE
(:UMULATIVE
- 11. HOURS IN REPORTING PERIOD 744.00 2160.00 09905.00
- 12. NUMBER OF HOURS REACTOR WAS CRITICAL O.00 429.25 27615.89
- 13. 1EACTOR RESERVE SHUTDOWN HOURS 0.00 0.00 0.00 14.
iC.URS GENERATOR ON-LINE O.00 420.97 P7020.94
- 15. ; NIT RESERVE SHUTDOWN HOURS 0.00 0.00 0.00
- 16. GEOSS THERMAL ENERGY GENERATED (MWH) 0.00 1014187.44 82738285.50
- 17. ? JOSS ELECTRICAL ENERGY GEN. (MWH) 0.00 342270.00 20030990.00
- 18. NET ELECTRICAL ENERGY GENERATED (MWH)
-6543.00 315851.00 26706947.00
- 19. UNIT SERVICE FACTOR O.00 19.66 45.11
- 20. UNIT AVAILABILITY FACTOR O.00 19.86 45.11
- 21. UNIT CAPACITY FACTOR (USING MDC NET) 0.00 12.74 30.83
- 22. UNIT CAPACITY FACTOR (USING DER NET) 0.00 12.74 38.83
- 23. UNIT FORCED OUTAGE RATE O.00 0.00 49.28
- 24. 9HUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE. AND DURATION Di-EACH):
4
- 25. }F SHUTDOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF' STARTUP:
CYCLE THREE REFUELING OUTAGE BEGAN JANUARY 18, 1989. THE STARTUP OF UNIT-2 IS SCHEDULED FOR APRIL 1989.
NOTE THAT THE YR. -TO-DATE AND CUMULATIVE VALUES HAVE BEEN UPDATED.
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SEQUOYAH NUCLEAR PLANT AVERAGE DAILY POWER LEVEL DOCKET NO. : 50-328 UNIT : TWO DATE : APRIL 04,1989 COMPLETED BY : D.C.DUPREE TELEPHONE : (615)843-6722 MONTH: MARCH 1989 AVERAGE DAILY POWER LEVEL AVERAGE DAILY POWER LEVEL DAY (MWe Net)
DAY (MWe Net) 01
-7 17
-7 02
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NUCLEAR PLANT OPERATING STATISTICS Sequoyah Nuclear Plant i
Period Hours 744 Month March 19 89 f tem No.
Unit No.
UNIT ONE UNIT TWO PLANT 1
Average Housiv Gross Load. kW 1,162,379 N/A 1,162,379 2
Maximum Hour Net Generation. MWh 1,152 N/A 1,152 3
Core Thermal Energy Gen, GWD (t)2 104.3877 0
104.3877 4
Steam Gen. Thermal Energy Cen., GWD (t)2 104.7812 0
104.7812 8
5 Gross Electrical Gen., MWh 864,810 0
864,810 5
6 Station Use. MWh 28,520 6.543 35,063 h
7 Net Electrical Gen., MWh 836,290
-6,543 829,747 8
Station Use, Percent 3.3 N/A 4.05 9
Accum Core Avg. Exposure, MWD /Toni 3,500 N/A 3,500 6
10 CTEG This Month,10 BTU 8,550,603 N/A 8,950,603 11 SGTEG This Month,106 BTU 8,582,840 N/A 8,582,840 12 13 Hours Reactor Was Critical 744.0 0.0 744.0 14 Unit Use, Hours Min.
744:00 0:00 744:00 15 Capacity Factor, Percent 98.26 0.0 49.13 S
16 Turbine Avail. Factor, Perrent 100.0 0.0 50.0 17 Generator Avail. Factor. Percent 100.0 0.0 50.0 0
18 Turbogen. Avail. Factor. Percent 100.0 0.0 50.0 h
19 Peactor Avad. Factor. Percent 100.0 00 50.0 0
20 Unit Avail. Factor, Percent 100.0 0.0 50.0 21 Turtune Startuns 0
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22 Reactor Cold Startuns 0
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Gross Heat Rate Blu/kWh 9.890 N/A 9,890 j
25 Net Heat Rate,8tu/kWh 10.220 N/A 10,300 5
26 Gross Heat Rate. Btu /kWh (w/ oil) 9,890 C
?7 Net Heat Ra te. Bru /kWh (w/cil )
10.300 28 Throttle Pressure, psig 848.5 N/A 848.5 5
29 Throttle Temperature, 'F 526.9 N/A 526.9
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30 Exhaust Pressure. InHg Abs.
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31 Intake Water Temo., *F 51.0 N/A 51.0 32 33 Main Feedwater, M tb/hr 15.2 N/A 15.2 34
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35 36 37 Full Power Capacity, EFPD 404.86 411.6 816.46 38 Accum. Cycle Full Power Days. EFPD 91.41 0.0 91.41 N
39 Oil Fired for Generation, Gallons 1,980 5
40 OH Heatino 6,lue. Btu / Gal.
138,000 41 Diesel Generation. MWh 30 4?
Max. Hour Net Gen.
Max. L ty Net Gen.
Load MWh Time Date MWh Date f actor, %
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43 1,152 1900 3-19-89 27,424 3-9-89 9 6. 81 */.
O Remarks: IFor BFNP this value is MWD /STU and for SQNP and WBNP this value is MWD /MTU.
E 2(t) indicates Thermal Energy.
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MAINTENANCE
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ELECTRICAL MONTHLY REPORT 01/11/88, 2-HV0P-070-0075-B, 01/31/88 During unit 2 outage, a surveillance test indicated residual heat removal heat exchanger "B" return header isolation valve operator did not move valve properly. The close torque switch was set too low i
(approximately 3.25 as read on the torque limiter ple;.e). Adjusted l
close torque switch setting to approximately 4.5 as read on the torque limiter plate and verified proper valve movement. (WR B285537). UR B275336 was submitted to be scheduled for grease replacement in the operator.
05/13/88, 2-IGN-268-0213-A, 01/27/89 During unit 2 outage, surveillance instruction (SI)-305.2 indicated hydrogen igniter #213 was not working. Igniter burned out due to normal wear. Replaced hydrogen igniter #213. Performed SI-305.2 and returned igniter to service. (WR B259979).
11/21/88, 1-MV0P-003-0020, 01/05/89 During unit 1 outage, an operational abnormality indicated heater B1 outlet flow control valve motor thermals out when attempting to close. Moisture and corrosion causing motor leads to ground out.
Replaced grease in limit switches and replaced motor. Sealed conduit with RTV to prevent water intrusion, set J1mits and performed motor operated valve analysis testing.. (WR B283319).
01/20/89, 2-MTRA-068-0031, 02/10/89 During unit 2 refueling outage, an operational abnormality indicated reactor coolant pump #2 tripped during a start. Possibly, the current relay magnets were dirty. The 51 (IAC77) overcurrent relay did not fully reset. Tested and cleaned magnets in each current relay. Performed breakaway torque test, 250 v DC and 500v megger test and returned to service. (WR B797649).
02/19/89, 2-BCTD-067-0298-B, 03/14/89 During unit 2 refueling outage, a motor operated valve analysis test indicated upper containment vent cooler i discharge valve, motor starter contacts, did not make up properly. Failure due to dirty front auxiliary contacts. Cleaned front auxiliary contact on right side of contactor and reterminated wires. (WR B292073).
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I ELECTRICAL !!ONTHLY REPORT l
J 1
02/23/89, 0-BATB-250-QX-F, 03/03/89 During unit 2 refueling outage and Unit 1 operation, vital battery III did not meet surveillance instruction (SI)-100 acceptance criteria. Specific gravity was below 1.195 probably due to age.
Placed vital battery on equalize charge for seven days. Added water to cells and performed SI-100. (WR B795105).
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(INSTRUMENTATION) i
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- e INSTRUMENT MONTHLY REPORT-i 10/06/88, 1-FT-062-0093C, 01/23/89 During unit shutdown, the chemical and volume control system charging header flow transmitter was found out of tolerance while performing surveillance instruction.(SI)-208. Root cause unknown. It may have been due to the extended time period between calibrations.
'l Recalibrates the transmitter.to desired tolerance and returned to service. (SI)-208.
11/17/88, 2-LT-068-0312C, 03/01/89 During. unit operation, the reactor coolant. system pressurizer relief tank level transmitter readings did not agree with the readings from 2-LI-068-0300. The zero point on the transmitter shifted.
i Recalibrates the transmitter to desired tolerance and returned to normal. Performed post maintenance test and returned to service.-(WR B769675).
-12/21/88, 2-TIC-067-0085-A, 02/22/89-During unit operation, the essential raw cooling water system control rod drive vent cooler "A" temperature indicating controller was not responding to setpoint controller. Root cause unknown. It may have been due to wear and/or cycling fatigue. The controller was calibrated to desired tolerance and returned to service.
(WR B299693).
12/25/88, 1-FT-030-0242, 02/21/89 During unit shutdown, the ventilation system shield building exhaust vent flow transmitter was found not reading properly. Root cause unknown. It was probably due to age of equipment. Recalibrates to desired tolerance and returned to normal. A potential reportable occurrence report (PRO)# 1-89-48 was written and submitted for review. (ER B772109).
01/22/89, 2-FM-063-0173C, 03/04/.89 During unit outage, the safety injection system residual heat removal injection flow modifier was found out of calibration while performing surveillance instruction (SI)-202. The modifier was found with a defective circuit board. Replaced printed circuit board and bench calibrated to desired tolerance. Reinstalled modifier'and performed post maintenance test. (WR B757739).
- 4; W4d t
j INSTRUIENT !!ONTHLY REPORT 01/26/89,-
2-LT-003-0098 02/20/89 During a refuel outage while performing surveillance instruction (SI)-70.3,-the main and auxiliary-feedwater system steam generator
- 3 level transmitter was found out of desired operational tolerance low throughout the entire range. The cause of failure'is unknown but may have been due to. equipment age and/or normal wear. The l
transmitter uas recalibrates and returned to service during the performance of SI-70.3. A new cover and housing o-rings were lubricated and installed and the cover torqued per QHDS requirements. (SI-70.3) 01/26/89, 2-LT-003-0043, 02/20/89 During a refuel outage, while performing surveillance' instruction (SI)-70.3, the main and auxiliary feedwater system steam generator
- 1 level transmitter was found out of desired operational tolerance range low. The cause of failure is unknown. It may have been due to equipment age and/or normal wear. The transmitter was recalibrates and returned to service. During the performance of SI-70.3, a new cover and housing o-rings were lubricated.and installed and the cover torqued per QMDS requirements. (SI-70.3) 01/26/89, 2-LT-003-0111,.
02/23/89 During a refuel outage, while performing surveillance instruction (SI)-70.3, the main and auxiliary feedwater system steam generator
- 4 level transmitter was found out of desired operational tolerance low throughout the entire range. The cause of failure is unknown but may have been due to equipment ags and/or normal wear. The transmitter was recalibrates and returned to service. During the performance of SI-70.3, a new cover and housing o-rings were lubricated and installed and the cover torqued per QMDS requirements. (SI-70.3) 02/10/89, 2-PDIC-065-0080-A, 03/01/89 During unit outage while in manual, the emergency gas treatment system cor.tainment annulus delta P indicating controller would not go and stay to zero percent. Drifts up to approximately 10%. Found that the amplifier circuit board had failed probably due to age and deterioration, Replaced amplifier circuit board and verified operability on the bench. Reinstalled controller and returned to service. Verified proper manual operation. (WR B757209).
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INSTRUMENT MONTHLY REPORT 02/11/89, 2-TM-068-0001C, 02/14/89 During unit outage, the reactor coolant loop #1 hot leg temperature modifier was found out of tolerance high while performing surveillance instruction (SI)-91.21. Root cause unknown. It may have been due to wear and/or age of equipment. Recalibrates the modifier to desired tolerance and returned to service. (SI-91.21).
-02/11/89, 2-TM-068-0024C-D, 02/14/89 During unit outage, the reactor coolant loop #2 hot leg temperature modifier was found out of tolerance high while performing surveillance instruction (SI)-91.22. Root cause of failure unknown, It may have been due to uear and/or age of equipment. Recalibrates the modifier to desired tolerance and returned to service.
(SI-91.22).
02/11/89, 2-TM-068-0043C, 02/13/89 During unit outage, the reactor coolant loop #3 hot leg. temperature modifier was found out of tolerance while performing surveillance instruction (SI)-91.23. Root.cause of failure unknown. It may have been due to wear _from equipment cycling. Recalibrates the modifier to desired tolerance. Potential reportable occurrence (PRO)# 2-89-32 was submitted for analysis. (SI-91.23).
02/11/89, 2-TM-068-0065C, 02/14/89 I'uring unit outage, the reactor coolant loop #4 hot leg temperature modifier was found out of tolerance high while performing surveillance instruction (SI)-91.24. Root cause of failure unknown.
It may have been due to wear and/or age of equipment. Recalibrates the modifier to desired tolerance and returned to service.
(SI-91.24).
02/11/89, 1-FCV-001-0103, 02/11/89 During unit operation, the main steam cool down dump valve to condenser 1A, pops open and closed. Feedback arm fell off because of improper installation. Reconnected feedback arm and verified stroke.
Returned to service. (WR B797963).
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INSTRUIENT !!ONTHLY REPORT 02/12/89, 2-PDIC-065-0082-B, 02/17/89 During unit refueling, the emergency gas treatment system containment annulus delta P indicating controller manual pushbuttons would not respond with controller in manual. Broken internal wire.
Repaired broken wire and verified operability of controller on bench. Returned controller to service and verified operability.
(NR B757026).
02/13/89, 2-TI-068-0043C, 02/13/89 During unit outage, the reactor coolant loop #3 hot leg temperature indicator would not calibrate. Cause due to age of indicator or wear. Replaced indicator and calibrated to desired tolerance.
Performed post maintenance test and returned to service.
(WR B757442).
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4 MAINTENANCE
SUMMARY
(MECHANICAL)
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Mt;CHANTCAL MONTHLY REPORT 07/15/88, 2-VLV-062-0901, 02/23/89 During unit 2 shutdown, the chemical and volume control system mixed bed demineralization on "A" inlet valve was discovered leaking by routine observation. Normal wear on diaphragm and o-rings.
Disassembled valve and installed new o-rings, diaphragm and stem.
reassembled valve. (WR B780646).
07/15/88, 2-VLV-062-0908, 02/23/88 During unit 2 shutdown the chemical and volume control system mixed bed demineralization "B" inlet valve was discovered leaking by routine observation. Normal wear on diaphragm and o-rings.
Disassembled valve, installed new o-rings and diaphragm. reassembled valve. (WR B780646).
09/12/88, 2-FCV-063-0063, 02/19/89 During unit 2 shutdown the safety injection system accumulator tank 4 nitrogen makeup valve was showing red and green light when valve is closed. Discovered during routine observation. Loose stem nut.
Generic Masoneillan stem rotation problem. Aligned limit switch with actuator arm, removed diaphragm cover tightened stem nut, staked thread, repacked valve and returned to normal. (WR B789495).
11/04/88, 2-VLV-061-1187, 03/14/89 During unit 2 operation, the ice condenser system g]ycol return block valve was discovered leaking during routine observation.
Normal wear on diaphragm. Installed new diaphragm and o-rings.
(WR B237113).
01/17/89, 2-VLV-001-0522, 02/18/89 During unit 2 operation while performing surveillance instruction (SI)-759, the main steam safety valve was found to be leaking past the seat. Seat was discolored and scratched. Removed valve bonnet and lapped seat and reinstalled bonnet. (WR B253492).
01/17/89, 2-VLV-001-0523, 02/18/89 During unit 2 operation while perfonning surveillance instruction (SI)-759, the main steam safety valve was found to be leaking past the seat. Seat was discolored and scratched. Removed valve bonnet and lapped seat and reinstalled bonnet. (WR B253492).
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!!ECHANICid. !!ONTHLY REPORT 01/17/89, 2-VLV-001-0529, 02/18/89 During unit 2 operation while perfonning surveillance instruction (SI)-759, the main steam safety valve.was found to be leaking past the seat. Seat was discolored and scratched. Removed valve bonnet and lapped seat and reinstalled bonnet. (WR B253492).
01/17/89,
'2-VLV-001-0530, 02/18/89 During unit 2 operation while performing surveillance instruction (SI)-759, the main steam safety valve was found to be leaking past the seat. Seat was discolored and scratched. Seat was discolored and scratched. Removed valve bonnet and lapped seat and reinstalled ~
01/17/89, 2-VLV-001-0531, 02/18/89 During unit 2 operation while performing surveillance. instruction (SI)-759, the main steam safety valve was found to be leaking past the seat. Seat was discolored and scratched. Removed valve bonnet and lapped seat and reinstalled bonnet. (WR B253492).
01/26/89, 2-HV0p-001-0023, 02/07/89 During unit 2 refuel outage, the steam generator 3 main steam header pressure relief control valve was discovered during routine observation as having an air leak around the diaphragm. Bad diaphragm due to normal use. Disassembled operator, removed old diaphragm and installed new diaphragm. Reassembled operator.
(WR B280985).
01/27/89, 2-VLV-067-0575A-A, 03/03/89 During unit 2 outage, while performing surveillance instruction (SI)-158.1, the essential raw cooling water return level control valve cooler check valve failed leak rate test. Seats were dirty.
Disassembled valve bonnet, cleaned valve' internals and reassmbled valve. (WR B753974' 02/07/89, 2-VLV-062-0518-S, 02/25/89 During unit 2 outage, the chemical and volume control system reciprocating charging pump discharge relief valve was discovered leaking following the performance of surveillance instruction (SI)-164. Appeared to have arc strike on seating surface.
Disassembled valve, cleaned internals, lapped seat and reassembled valve and installed lead seal (WR B784043).
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MECHANICAL !!ONTHLY REPORT 02/07/89, 2-VLV-062-0649-S, 02/25/89 During unit 2 outage, the chemical and volume control system seal water heat exchanger relief valve was discovered leaking following the performance of surveillance instruction (SI)-164. Seats were scuffed and scratched. Disassembled valve, cleaned internals, lapped seats, lapped tail / disc bushing, set rings and reassembled valve.
(WR B784043).
~
02/07/89, 2-VLV-062-0505-S, 02/25/89 During unit 2 outage, the chemical and volume control system charging pump suction relief valve was discovered leaking following the performance of surveillance instruction (SI)-164. Leaking thru seats, causing rust and boron buildup on bellows assembly.
Disassembled valve, cleaned internals and lapped seats. Reassembled valve. (WR B784043).
02/07/89, 2-VLV-062-0675-S, 02/25/89 During unit 2 outage, the chemical and volume control system letdown relief valve was discovered leaking following the performance of surveillance instruction (SI)-164. Seats were scuffed and rough.
Disassembled valve, cleaned internals, lapped seats, found bellows i
ruptured, installed new bellows, lapped new bellows tailpiece to j
disc bushing, lightly polished seats and reassembled valve.
(WR B784043).
02/22/89, 2-VLV-062-0684-S, 03/01/89 During unit 2 outage, the chemical and volume control system letdown check valve was discovered leaking boron around the bonnet during routine observation. Gasket was worn due to age. Disassembled valve and installed new gasket. Reassembled valve and retorqued per maintenance instruction (MI)-11.4. (WR B775269).
03/11/89, 2-FCV-030-0007-A, 03/23/.89 During unit 2 cycle 3 outage, the ventilation system upper
{
compartment purge isolation valve failed its desired response time when actuated by slave relay K622A Valve stem was dirty. Cleaned valve and had operations stroke valve several times and valve operated properly. (WR B256670).
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!ECHANICAL !!ONTHLY REPORT 03/14/89, 1-FCV-087-0024-G, 03/16/89 During unit 1 operation, the upper head injection system isolation flow control valve was discovered during routine observation leaking nitrogen. Schrader valve was worn out. Removed accumulator from upper head injection valve in machine shop and replaced bladder to use on valve ~if needed. Installed new schrader valve. (WR B775052).
During the repairs to the valve operator 2 significant events occurred: 1) Argon was used to charge the accumulator. This was replaced with nitrogen. 2) A scaffold support was found obstructing the operator coupling inside the valve yoke. The initial investigation report is attached with this WR package.
03/23/89,.
2-FCV-001-0029-T, 03/23/89 During unit 2 cycle 3 outage, the steam generator 4 main steam header isolatica valve would not operate from local or remote stations. Discovered during routine observation. Failure due to lack of lubrication of stem. Adjusted packing, lubricated stem and checked for leaks. Post maintenance test will be performed on work request B734028.
(WR B256435).
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MECHANICAL MAINTENANCE MONTHLY REPORT FOR MARCH 1989 Unit 1 1.
Installed new air-side seal oil pump at 1-PMP-35-267.
2.
Installed new stator cooling water pump at 1-PMP-35-88.
3.
Replaced several screens on CCW traveling screen 1-SCR-26-7.
4.
Rebuilt boric acid transfer pump 1B.
5.
Rebuilt Reactor Building floor and equipment drain sump pumps.
Unit 2 1.
Rebuilt S/G layup pump 2-PMP-41-17.
2.
Continued work on snubbers.
3.
Removed center missile shield for draining RCS.
4.
Installed solenoid valves on RM-90-114, -115, -116, -108B, -109B, and
-110B.
5.
Completed inspection on D/G 2B-B and 2A-A.
6.
Replaced seals on lower inlet doors on ice condenser.
7.
Completed repairs on System 43 containment isolation valves.
8.
Installed new reactor coolant drain tank pump.
9.
Completed several WRs on System 68.
10.
Replaced air-side seal oil pump.
11.
Replaced UHI spool pieces.
12.
Repacked'2-FCV-63-98 isolation valve to the SI accumulator tank.
13.
Installed two main steam safety valves.
14.
Rebuilt boric acid transfer pump 2B.
15.
Completed RCP carriage seal work.
16.
Reinstalled vessel head and retensioned studs.
17.
Replaced seal on boardroom chiller 2B.
Unit 0 1.
Congleted maintenance on various System 62 valves.
2.
Installed two new nitrogen pressure regulators.
3.
Rebuilt radiation monitor pump 0-RM-90-206.
4.
Rebuilt emergency bearing lube water pump.
5.
Rebuilt waste gas compressor B.
Other 1.
Continued closure of various CAQRs, CARS, DRs etc.
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O 4'
e MAINTENANCE
SUMMARY
(MODIFICATIONS) i L,
]
j t
O a
SUMMARY
OF WORK COMPLETED _
MODIFICATIONS - CURRENT STATUS MARCH 1989
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_ _ - - _ _ _ _ _ _ = _ _ _ _ _ - - - _ _ _ _ - - _ _ _ _ - - _. - _ _ - _.
Major Capital Projects:
ECN 6720 - Crane Consistency Program Unit 1 polar crane limit switch weights. remain to be painted.
Completion is scheduled for unit 1 cycle 4 (U1C4).
Completed cranes - unit 1 Turbine Building 15-ton crane, unit 2 polar crane, unit 2 Turbine Building 200-ton crane, unit 2 Turbine Building 15-ton crane, Service Building 5-ton crane, and Turbine Building 10-ton hatchway crane.
Remaining listed cranes are to be modified after unit 2 cycle 3 (U2C3) - Auxiliary Building 125-ton crane, waste-packaging crano, railroad-bay crane, and unit 1 Turbine Building 200-ton crane.
ECN 6180 - Postaccident Monitoring This work is now scheduled on unit 1 by UIC4 and unit 2 cycle 4.
DCN 0026 - Sewage Treatment Facility and Civil Upgrade WP 026-02 is physically complete and in the closure cycle.
ECN 5855 - Replacement of Doors A56 and A57 WP 09679 remains on hold and is partially complete.
DCR 1898 - ECNs 6832 and 6596 - Dry Active Waste (DAW) Building Equipment installed by WP 12477 has been removed from the DAW building as mandated by Water and Waste Processing Croup. All work is complete.
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 manho'e 1 to manholo 42.
All material was purchased before the job was stopped for lack of funding. Work is 45-to 50-percent complete.
ECN 5503 - Evacuation Alarms O&PS/ Fire Detection O&PS WP 12482 - Work has stopped because of lack of funding.
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_ _ _ _ _ _ _ _ _ _ _ _ _.. _ _ _ _ _ ~ _ _ _ _ _ _.. _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _
Other Items (cont.):
ECN 5552 - Condensate Demineralized Modifications and High Crud Filter Upgrade to higher-range instrumentation for condensate demineralized system neutralization and nonreclaimable waste pumps.
WP 5552 Fieldwork is complete.
ECN 5609 - Alteration to the Makeup Water. Treatment Plant -
WP 12387 - Work is 90 percent complete. Activity is on hold because of other priorities.
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 receipt of a pump for the alum sludge pond.
WP 12684 - WP is field-complete. Equipment calibration and functional tests remain.
l WP 12731 - WP is approximately 97 percent complete. MODS needs connection diagram revisions, FCRs, and drawings for heat trace installation from NE.
l 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.
ECH 5911 - Waste Disposal Piping Addition Work was temporarily halted for U2C3 outage.
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 Domineralizer Work was temporarily halted for U2C3 outage.
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othtr Item, (cont.):
i l
ECH 6183 - Replace Instrumentation Root Valves - System 1, Unit 2 1
Work is complete except for insulation and leak check in mode 3.
ECN 6196 - Pressurizer Hangers and Valves Unit 1 is complete.
Related unit 2 work is being performed under DCN 1099.
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.
~
Etni 6429 - Component Cooling Heat Exchanger B Replacement Work required for U2C3 is complete.
Some painting and insulation work remain-to be completed.
ECN 6455 - Upgrade CU-3 Box Battery Packs WP 12295 is field-complete. Waiting for secondary drawings to be. revised.
ECN 6689 - Relocation of Main Steam PORV WP 12172 - Field complete. Awaiting secondary drawings to be updated.
ECN 6706 14 Support Enhancement / Lost Calculations Postrestart work is in progress.
ECN 6739 - Alternate Analysis Unit 2 postrestart work is in progress.
ECN 6815 - 500-kV Switchyard Addition All work is complete in the 500-kV switchyard except for high potential testing of transformer 5018.
Software changes remain to be donc 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.
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.Othnr Items (cont.):
ECN 7329 - Replace Check Valves with Motor-Operated Butterfly Valves -
System 67, Unit 2 Field complete.
Awaiting secondary drawings to be updated.
ECN 7331 - Replace Check Valves with Motor-Operated Butterfly Valves -
System 70, Unit 2 Field complete. Awaiting secondary drawings to be updated.
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 UICA outage).
l DCN 550 - Modify Pipe Support - System 68. Unit 1 All work is complete except touchup paint (to be donc during UIC4 outage).
DCN 645 - Replace Main Condenser Circulating Water Piping Bellows -
System 27, Unit 2 Work is in progress.
DCN 703 - Modify RCP Seals - System 68, Unit 2 Work is in progress.
DCN 704 - Modify Pipe Support - System 68 Unit 1 All work is complete except touchup paint (to be done during UIC4 outage).
DCN 794 - Main Cenerator Hydrogen and Stator Cooling Water System Modifications - System 35, Unit 2 Work planned for U2C3 is complete.
Remaining work is held because of lack of funding.
DCN 933 - Replace Ice Condenser Drain Piping Bellows - System 61 Unit 2 Complete.
DCN 943 - Addition of Stiffeners to Penetrations - System 61 Unit 2 Work is complete except for leak-check.
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).
j GLOLSARY
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GLOSSARY OF VARIOUS ABBREVIATIONS Page 1 of 2 1.
- Auxiliary Building Gas Treatment System 2.
- Auxiliary Building Secondary Containment Enclosure 3.
AB(I)
- Auxiliary Building (Isolation) 4.
- Auxiliary Feedwater 5.
AOI
- Abnormal Operating Instruction 6.
ASOS
- Assistant Shift Operations Supervisor 7.
- Assistant Unit Operator 8.
BAT
- Boric Acid Storage Tank 9.
BIT
- Boron Injection Tank 10.
CAQR
- Condition Adverue to Quality Report 11.
- Corrective Action Report 12.
- Centrifugal Charging Pump 13.
- Component Cooling Water 14.
CDWE
- Condensato Demineralized Waste Evaporator 15.
CRI
- Control Room Isolation 16.
- Control Room Emergency Ventilation System 17, CSS (CS) - Containment Sprsy System 18.
- Containment Ventilation Isolation 19.
D/C(s)
- Diesel Generator (s) 20.
DCN
- Design Change Notice 21.
- Design Change Request 22.
DR
- Discrepancy Report 23.
- Emergency Core Cooling System 24.
- Engineering Change Notice 25.
- Emergency Cas Treatment System 26.
- Electromagnetic Interference 27.
- Environmentally Qualified / Environmental Qualification 28.
- Essential Raw Cooling Water 29.
E/ES
- Emergency Instruction 30.
- Engineered Safety Feature 31.
ESFA
- Engineered Safety Feature Actuation 32.
FCR
- Field Change Request 33.-
- Flow Control Valve 34.
- Floor Drain Collector Tank 35.
- Flow Differential Switch 36.
- Flow Indicating Controllers 37.
- Final Safety Analysis Report
,38.
FS
- Flow Switch 39.
FWI
- Feedwater Isolation 40.
GOI
- General Operating Instruction 41.
GPM
- Callons Per Minute 42.
HDTP
- Heater Drain Tank Pump 43.
H0
- Hold Order 44.
IM
- Instrument Mechanic / Instrument Maintenance 45.
IMI
- Instrument Maintenance Instruction 46.
- Level Control Valve i
47.
LER
- Licensing Event Report 48.
LCO
- Limiting Condition for Operation 49.
- Loss Of Coolant Accident 50.
- Level Switch a* ~ '
t CLOSSARY OF*/ARIOUS ABBREVIATIONS Tage 2 of 2 51.
- Measuring and Test Equipment
$2.
- Maximum Allowable Stroke Time 5 3..
- Main Control Room 54.
- Motor-Driven Auxiliary Feedwater Pump 55.
MFI
- Main Feedwater Isolation 56.
NWF.
- Main Feedwater 57.
MFWRV
- Main Feedwater Regulating Valves 58.
- Main Feedwater Pump 59.
MI
- Maintenance Instruction 60.
MODS
- Modifications l
61.
- Motor Operated Valve 62.
MSI
- Main Steam Isolation 63.
- Main Steam Isolation Valve 64.
- Moisture Separator Reheaters 65.
NE
- Nuclear Engineering (formerly Division of Nuclear Engineering) 66.
NIS
- Nuclear Instrumentation System 67.
NMUDI
- New Makeup Deionized System 68.
- Nuclear Security Service 69.
- Nuclear Steam Supply Systems 70.
O&PS
- Office and Power Stores Building 71.
- Preventive Maintenance 72.
- Postmodification Test 73.
- Plant Operations Review Committee 74.
PORY
- Power-Operated Relief Valve 75.
PRO
- Potential Reportable Occurrence 76.
- Pressure Differential Switch 77.
RCS/(P)
- Reactor Coolant System /(Reactor Coolant Pump) 78.
- Residual Heat Removal 79.
- Radiation Monitor (RAD Monitor / RAD MON) 80.
- Rod Position Indicator 81.
- Refueling Water Storage Tank 82.
- Significant Condition Report 83.
- Spent Fuel Pit 84.
S/G(s)
- Steam Generator (s) 85.
- Surveillance Instruction /or Safety Injection 86.
SMI
- Special Maintenance Instruction 87.
SOS
- Shift Operations Supervisor 88.
SOI
- System Operating Instruction 89.
- Sequoyah Nuclear Plant 90.
SR
- Surveillance Requirement / Source Range 91.
SSPS
- Solid State Protection System 92.
TACF
- Temporary Alteration Control Form 93.
TI
- Technical Instruction 94.
TS(s)
- Technical Specification (s) 95.
- Tennessee Valley Authority 96.
UHI
- Upper Head Injection 97.
UO/(S)RO - Unit Operator /(Senior) Reactor Operator 98.
- Valve 99.
WP
- Workplan 100. WR
- Work Request w*
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l GLOSSARY OF VARIOUS SYSTEMS OF SEOUOYAH NUCLEAR PLANT l
SYSTEM CODE SYSTEM TITLE 1
Main Steam System (Turbine) (MSR) 2 Condensate 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 l
30 Ventilating System i
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 cooling 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 I
87 Upper Head Injection System 90 Radiation Monitoring System 268 Hydrogen Mitigation System l
l I
i
, _ - - _ - - - - _ - _ _. - _ = -
l' s '
\\
OPERATIONAL MODES
% RATED
' AVERAGE C00LANT' l:
_ MODE THERMAL POWER
' TEMPERATURE 1.'
POWER'0PERATION~
>-5%
> 3500F 2.
-STARTUP
< 5%
> 3500F 0
3.
HOT STANDBY 0
> 350 F 1.
L 4.
HOT SHUTDOWN-0 350 F >Tave 0
L
> 2000F 0
5.
COLD SHUTDOWN 0.
f 200 F 0
6.
REFUELING 0
< 140 F e
e e
1. _..
.i
l THIS PAGE INTENTIONALLY LEFT BLANK I
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1 16 y
' TENNESSEE VALLEY AUTHORITY CH ATTANOOGA, TENNESSEE 37401 5N 1578 Lookout Place APR 181989 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.
20555 Gentlemen:
In the Mattar of
)
Docket Nos. 50-327 Tennessee Valley Authority
)
50-328 SEQUOYAH NUCLEAR PLANT (SQN) - MARCH 1989 MONTHLY OPERATING REPORT Enclosed is the March 1989 Monthly Operating Report as required by SQN technical specification 6.9.1.10.
?
If you have'any questions concerning this matter, please call R. R. Thompson at (615) 843-7470.
Very truly your,
{
o, r.,
ice President and Nuclear Tech cal Director Enclosure cc (Enclosure):
Director, Region II INPO Records Center Nuclear Regulatory Commission Suite 1500 Office of Inspection and Enforcement 1100 Circle 75 Parkway Suite 3100 Atlanta, Georgia 30323 101 Marietta Street i
Atlanta, Georgia 30339 Sequoyah Resident Inspector Sequoyah Nuclear Plant Deputy Executive Director 2600 Igou 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
>V Palo Alto. California 94304 1
Dennis Crutchfield, Director
[i TVA Projects Division Office of Special Projects i
Washington, D.C.
20555 I
t l
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