ML20086A308
| ML20086A308 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/31/1991 |
| From: | Broughton T, Heysek W GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| C311-91-2147, NUDOCS 9111190032 | |
| Download: ML20086A308 (8) | |
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GPU Nuclear Corporation 1 Nuclear
- =e "rs:reo Middletown, Pennsylvania 17057 0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Number:
(717) 948-8005 November 15, 1991 C311-91-2147 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.
20555 Gentlemen:
Subject:
Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Operating License No. DPR 50 Docket No. 50-289 Monthly Operating Report October 1991 Enclosed are two copies of the October,1991 Monthly Operating Report for Three Mile Island Nuclear Station, Unit 1.
Sincerely, W
i T. G. Bro ghton Vice President & Director, TMI-1 WGH:
Attachments cc: Administrator, Region I THI Senior Resident Inspector I C 0,'~ /1 r 91111900$b911031 d
PDR ADOCK 05000289 IG R
PDR 7
GPU Nuclear Corporation is a subs %ary of General Pubhc Utdibes Corporation
(
OPERATIONS
SUMMARY
OCTOBER 1991 l
The unit was shutdown and cooled down on Decay Heat Removal as the month began. The Cycle 9 refueling outage was in progress.
The OTS0s were eddy current tested with no increase in testing scope at a result of initial sample results. Tube sleeving was completed in both OTSGs in the lane wedge high probability tube leak areas.
Prior to offload of the Reactor Vessel on 10/8/91 a fuel assembly was lifted prior to the establishment of containment integrity.
This event was reported in accordance with 10 CFR 50.73.
A core offload was completed and the Reactor Vessel 10 year inspection was performed.
The core was then reloaded. A chemical cleaning process was completed on both OTSGs that should prevent power output limitations due to high secondary side operating range steam generator levels. The RCS was in a partially drained down condition on decay heat removal at the close of this reporting period.
Cycle 9R workscope is continuing with restart scheduled for the second week in
- November, tLMOR SAFETY RELATFD MAINTENANCE During October, the following major safety related maintenance activity was performed:
Refuelina Activities (Reactor Vessel RC-T-1)
The following activities were performed in support of rieactor refueling and the 10 year Reactor Vessel (RV) examination activities.
The RV head missile shields, fans and insulation were removed.
The Fuel Transfer Canal (FTC) seal plate was installed and the RV studs detensioned and parked. The "A" to "B" cables, thermocouple and "PI" cables were uncoupled.
The fuel transfer tube and flood line cover plates were removed. After all preparati: 3 were complete the reactor head was rigged, removed and transported to the head stand. The "NI" cover plate and reactor seal ring were sealed and the FTC was flooded.
Installation of the indexing fixture, removal and storage of the plenum in the shallow end of the FTC and removal of the indexing fixture followed. Operations off-loaded all fuel from the vessel to the spent fuel pool and the Core Support Assembly (CSA) was removed from the RV and stored in the deep end of the FTC. The Automated Reactor Inspection System (ARIS) was then installed on the vessel to perform the 10 year ISI examination. U)on completion of the ISI ir:,,ection, the ARIS equipment was removed from tie vessel and the CSA reinstalled. After the core was reloaded with fuel, the indexing fixture and the plenum were reinstalled in the RV.
The indexing fixture was then removed and the FTC drained.
While the RV head was in the head stand, the following work items were completed:
twenty CRDHs were removed and the nozzle gaskets replaced, eight support skirt access ports were machined and the skirt and bolted covers installed.
2
4 Qace Throuah Steam Generators Eddy Current (ECT) inspection of 3% of the tubes and all tubes in the lane wedge of each OTSG was completed. As a result both OTSGs were categorized as "C-1" and examination of a larger sample population was not required.
As a result of the tube plug ECT examination, 15 B&W Inconel 600 rolled plugs in the "A" 0TSG upper tubesheet were replaced with inconel 690 rolled plugs and 5 rolled plugs (2 in "A" and 3 in "B" OTSG) were installed in the lower tubesheets below B&W explosive slugs with pressure boundary ETC indications.
Though the ETC indications in tie 5 lowar plugs were not the same plug crack indications found at Oconet, ^ hey were conservatively replugged.
On completion of drip testing and (.hemical cleaning fills and pressure tests, an additional 30 rolled plugs were installed below leaking explosive plugs (27 in the "A" lower tubesheet and 3 in the "B" lower tubesheet). A leaking Westinghouse rolled plug was drilled out and replaced with a manual welded, tapered, tubesheet plug.
Inconel 690 tube sleeves were installed in high cycle fatigue susceptible lane wedge tubes in both the "A" and "B" OTSGs (125 in each).
EPRI SGOG Chemical Cleaning was performed on both OTSGs. The dissolution process removed approximately 3000 pounds of Fe 0. (magnetite) from each OTSG.
3 This is approximately 95% of the availabl magnetite.
OTSG Sludge Lancing removed and filtered out an additional 118 pounds of insoluble material from the "B" OTSG.
Sledge Lancing of the "A" 0TSG was still in progress at the end of the month.
Make-Vo Pump MU-P-lc Make-up pump MU-P-10 repair work continued in October with the disassembly of the pump.
The rotating assembly was found to have evidence of metal-to-metal contact and was removed from the casing. A new rotating assembly was procured from Florida Power Corporation and installed in the casing.
Wearing ring clearances were checked and the pump bearings replaced. The casing cover was reinstalled and the rotating assembly rotated to ensure free movement.
Fuel Transfer Valves FH-V-1 A/B During the underwater inspection of the spent fuel pool utilizing a submarine camera, it was discovered the yoke nuts on fuel transfer tube gate valves FH-V-1 A/B had indications of cracking. Divers replaced the manganese / bronze bushings with stainless steel nuts per vendor (Stearns-Rogers) instructions.
Fuel Transfer Carriaae FH-A-4A During the Reactor Core fuel off-load and transfer, the east fuel carriage became inoperable due to the drive chain coming off the sprocket.
Divers found that the west chaft locking collar had become loose and allowed the drive shaft to move in a lateral direction throwing the chain off.
The divers 3
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realigned the sprockets, reinstalled the chain and tightened all fasteners.
The transfer carriage was cycled with 3 further discrepancies found.
Inspection of the west carriage drive i the divers found no problems.
Incore Detector ReplaceED1 Incore detector replacement work accomplished in October included the removal and cutting of all fif ty-two incores.
The incore tubes were flushed in preparation for the installation of the new detectors. During an inspection under the Reactor Vessel at the incore tubes, the walls and tubes were found covered with boron crystals. To ensure the integrity of the incore tubes, a pressure test involving freeze sealing six tubes at the vessel and pressurizing to 1800 psig from the incore table was performed.
No leakage was evident and the boron buildup was attributed to FTC seal ring leakage.
The fifty-two new incore detectors were installed in the tubes and instrument testing started.
Auxiliary Egel Bridae FH-A-2 During preventive maintenance on the Auxiliary Fuel Bridge, the inner mast extended to the fuel transfer canal floor when the cable slipped through the restraint during mast load brake replacement. The mast was removed from the Reactor Building for inspection and repair.
The piston cylinder was replaced and a rebuilt fuel grapple installed.
The fuel transfer canal was liquid dye penetrani, tested, vacuum box tested, and visually examined with no indications found. The inner mast was reinstalled on the bridge and testing is in progress.
Pressurizer Safety Relief Valves The Pressurizer Pilot Operator Relief Valve (PORV), RC-RV-2 and the Code Safety Valves, RC-RV-1 A/B were replaced with spare tested valves. The valves that were removed were shipped to Wyle Labs for testing and inspection.
tiain Steam Safety Relief Valves Main Steam Safety Relief valve work accomplished included the disassembly, cleaning, lubrication and lapping the seats on the following valves: MS-V-17A, MS-V-20A, MS-V-21A, MS-V-188, MS-V-19B, MS-V-18C, MS-V-200, MS-V-19D and MS-V-200.
Q.ther Maintenance Activi_tjm.
The following is a summary of various refueling outage activities accomplished: 10A of 448 Valve repacks, H of M welded component replacements, H of E Local Leak Rate Tests,_H of E Snubber testing / seal replacements and M of E Baseline and 1Q of n D/P M0 VATS Valve Operator Tests.
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OPERATING DATA REPORT DOCKET NO.
50-289 OATg November 15, 1991 COMPLETED BY W G HEYSEK OPERATING STATUS TELEPHONE (717) 948-8191
- 1. UNIT NAME:
THREE MILE ISLAND UNIT 1 NOTtS:
- 2. REPORTING PERIOD:
OCTOBER 1991
- 3. LICENSED THERMAL POWER:
2568
- 4. NAMEPLATE RATING (GROSS MWe):
871
- 5. DESIGN ELECTRICAL RATING (HET HWe):
819
- 6. MAXIMUM DEPENDABLE CAPACITY (GROSS MWe):
856
- 7. MAXIMUM DEPENDABLE CAPACITY (NET MWe):
808
- 8. IF CHANGES OCCUR IN (ITEMS 3-7) SINCE LAST REPORT, GIVE REASONS:
- 9. POWER LEVEL TO WHICH RESTRICTED, IF ANY (NET MWe):
- 10. REASONS FOR RESTRICTIONS, IF ANY:
THIS MONTH YR-TO-DATE CUMMULATIVE
- 11. HOURS IN REPORTING PERIOD (HRS) 745.0 7296.0 150457.0
- 12. NJMBER OF HOURS REACTOR WAS CRITICAL (HRS) 0.0 6427.1 75591.3
- 13. REACTOR RESERVE SHUTDOWN HOURS (HRS) 0.0
-0.0 2245.6
- 14. HOURS GENERATOR ON-LINE (HRS) 0.0 6421.5 74541.2
- 15. UNIT RESERVE SHUTDOWN HOURS (HRS) 0.0 0.0 0.0
- 16. GROSS THERMAL ENERGY GENERATED (MWH) 0 14694918 181165249
- 17. GROSS ELECTRICAL ENERGY GENERATED (MWH) 0 5124291 61085195
- 18. NET ELECTRICAL ENERGY GENERATED (MWH)
-3442 4816526 57305946
- 19. UNIT SERVICE FACTOR (4) 0.0 88.0 49.5
- 20. UNIT AVAILABILITY FACTOR
(%)
0.0 81.0 49.5
- 21. UNIT CAPACITY FACTOR (USING MDC NET) 0.0 81.7 48.5
- 22. UNIT CAPACITY FACTOR (USING DER NET) 0.0 80.6 46.5
- 23. UNIT FORCED OUTAGE RATE
(%)
0.0 0.8 44.9 UNIT FORCED OUTAGE HOURS (HRS) 0.0 51.9 60648.7
- 24. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE AND DURATION OF EACH):
- 25. IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP:
11/13/91
3 AVERAGE DAILY UNIT POWER LEVEL DOCKET No.
50-289 UNIT THI-1 DATE flovember 15, 1991 COMPLETED BY W G HEYSEK TELEPHONE (717) 948-8191 MONTH:
OCTOBER DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-NET)
(HWe-NET) 1
-5.
17
-4 2
-6 18
-4 3
-5 19
-4 4
-6 20
-4 5
-5 21
-4 6
-5 22
-3 7
-5 23
-5 8
-5 24
-3 9
-5 25
-5 10
-5 26
-4 11
-5 27
-5 12
-5 28
-5 13
-4 29
-4 14
-4 30
-4 15
-4 31
-4 16
-4 6
UNIT SHUTDOWNS AND PCWER REDUCTIONS DOCKET NO.
50-289 UNIT NAME TMI-1 REPORT MONTH Octcber 1991 DATE November is, 1991 COMPLETED BT W. G. Heysek TNT.RPHONE (717) 948-8191 i
Method of Licensee System.orsponent Cause & Corrective 80-Date Type' Duration Reason" shutting Event Code Code Action to (Mours)
Down Report #
Prevent Recurrence Reactor'
- ga 2
91-IC/01/91 S.
745 C
1 & 3*
91-003 The unit is shutdown for refueling. The 9R 04 outage commenced on 09/27/91. The total number of outage hours as of this report period is 824.6.
During a scheduled turbine valve test the reactor tripped from 154 power.
L 1
2 3
4 F Forced Reason m thod Exhibit C - Instructions for e
s schedated A-Equipment Falture (Explain) 1-Ma m t preperation of Data Entry sheets 8-Maintenance or Test 2-Manuel Scram for Licensee Event Report (LER)
C-Refueling 3-Autoretic Screra File (NUREG-0161) j D-Regulatory Restriction 4-Other (Explain)
E-operator Training & Licensing Examination 5 Exhibit 1 m source J
F-Adninistrative i
G-Operational Error (Explain) 6 Actuelty used exhibits F & It NUREG C161 i
R-Other (Explain) j 7
. a
REFUELING INFORMATION RE0 VEST 1.
Name of Facility:
Three Mile Island Nuclear Station, Unit 1 2.
Scheduled date for next refueling shutdown:
September 17, 1993 (10R) 3.
Scheduled date for restart following current refueling: November 13, 1991 (9R) 4.
Will refueling or resumption of operation thereafter require a technical specification change or other license amendment? No.
If answer is yes, in general, what will these be?
If answer is no, has the reload fuel design and core configuration been reviewed by your Plant Safety Review Committee to determine whether any unreviewed safety questions are associated with the core reload (Ref. 10 CFR Section 50.59)? Yes.
If no such review has taken place, when is it scheduled?
5.
Scheduled date(s) for submitting proposed licansing action and supporting information:
None planned.
6.
Important licensing considerations associated with refueling, e.g. new or different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures:
GPU Nuclear has installed four Westinghouse Lead Test Assemblies during the reload of the TMI-1 core for cycle 9 operation.
Westinghouse fuel technology will be utilized to the extent possible.
7.
The number of fuel assemblies (a) in the core, and (b) in the spent fuel storage pool:
(a) 177 (b) 521 8.
The present licensed spent fuel pool storage capacity and the size of any increase in licensed storage capacity that has been requested or is planned, in number of fuel usemblies:
The present licensed capacity is 752.
Planning to increase licensed capacity through fuel pool reracking is in progress.
9.
The projected date of the last refueling that can be discharged to the l
spent fuel pool assuming the present licensed capacity:
1991 is the last refueling discharge which allows full core off-load j
capacity (177 fuel assemblies).
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