ML20197H657
| ML20197H657 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/30/1986 |
| From: | Hukill H, Smyth C GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF ADMINISTRATION (ADM), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| 5211-86-2089, NUDOCS 8605190232 | |
| Download: ML20197H657 (8) | |
Text
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OPERATIONS
SUMMARY
APRIL, 1986 The unit entered April in a cold shutdow'1 condition on "A" Decay Heat Removal in a partial drain down condition.
Eddy Current Testing was in progress.
On April 1,1986, the turbine generator support systems were shutdown to commence disassembly of the Main Generator. This was due to a leak detected in the main generator stator coolant system. Following reassembly of the main generator, a leak test was performed with satisfactory results.
On April 3, 1986, an EFW nozzle was inspected and showed signs of circumfurential cracking around the entire nozzle.
Details of this inspection are contained in the Maintenance Summary below.
The Eddy Current Outage completion proceeded with a fill and vent of the RCS on April 19, 1986. The next day vacuum was obtained in the main condenser and feedwater cleanup commenced. Hot shutdown was achieved on April 21, 1986.
Shortly after reaching a hot shutdown condition, a partial loss of offsite power occurred. This occurred at 0935 and was a result of a fault in the "A" Auxiliary transformer feeder breaker to the "D" 4160V ES bus. The unit was stabilized and returned to hot shutdown af ter power was restored to plant buses.
The reactor was first taken critical following the Eddy Current Outage on April 22,1986. The next day when escalating power beyond 8% while transferring the operating feed pump from auxiliary to main steam, a high pressure reactor trip occurred due to inadequate feedwater to the OTSGs. The plant was quickly stabilized at hot shutdown following performance of emergency actions in accordance with the Reactor Trip procedure.
The same day the reactor was taken critical and power was escalated to 57% with the main generator on line. On April 24,1986, at 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> a steam leak was reported to the control room. The steam leak was the result of one of the drain lines downstream of the turbine stop and control valves breaking at a welded connection. Power was reduced and the turbine was taken off line with the reactor at 5% power to repair the broken line.
Subsequently, power escalation occurred and the unit achieved 100% stable' power ope:ation on April 26, 1986.
The unit remained at 100% steady state power at the close of this reporting period.
8605190232 860430 PDR ADOCK 05000289 R
es MAJOR SAFETY RELATED MAINTENANCE During the month of April, TMI-l performed the following major maintenance items.
Once Through Steam Generators RC-H-1 A/B During the month of April, Eddy Current testing of the A & B Once Through Steam Generators (OTSGs) was completed. Testing data was analyzed and a list of tubes that were to be removed from service developed. Ten tubes were removed from service in the "A" 0TSG.
B8W rolled plugs (2 with stabilizers) were installed in the upper tube sheet and B&W ribbed plugs were installed in the lower tube sheet. Fif teen tubes were removed from service in the "B" 0TSG. B&W rolled plugs (9 with stabilizers) were installed in the upper tube sheet and B&W ribbed plugs installed in the lower tube sheet.
Close out inspections were performed in the OTSG primary openings and handhole and manway covers reinstalled.
The "A" 0TSG secondary side inspection port drilling completed with two ports installed. A secondary side video inspection was performed followed by removal of sludge samples from the tube bundle and support plates. All inspection data and sludge samples were turned over to Plant Engineering (samples currently at B&W) for analysis.
Inspection port covers were installed and insulated.
Emergency Feedwater Nozzle, Header and Riser Work While repairing an Emergency Feedwater Header / Riser flange that had been injected with Furmanite, the associated EFW nozzle (A-2) on the "A" OTSG was liquid penetrant inspected.
The inspection revealed cracks on tre nozzle collar weld.
Due to this finding, the remaining five nozzles on the "A" OTSG and all six nozzles on the "B" 0TSG were removed for inspection. The nozzles on the "A" OTSG (A-1, A-2, A-4 and A-5) required weld repairs. The indications were ground out and rewelded.
Additional linear indications were also found on a few of the remaining nozzles and were removed by minor grinding.
No weld repairs were required as they were determined to be surface defects from construction.
During th~e EFW riser / nozzle removal, pitting of the OTSG shell gasket seating surface and the mating EFW riser flange were found. Pitting was also found in the lower EFW riser flange and mating EFW header flanges.
Shell gasket seating surfaces were refinished by Plant Personnel with the exception of the A-5 and A-6 surfaces.
B&W contractor personnel set up inline machining equipment to resurface the A-5 and A-6 gasket surfaces. Plant Personnel also milled / dressed pits from the riser flanges and EFW header flanges. All nozzles and risers were reinstalled and the secondary sides of both OTSGs pressure tested. The "A" OTSG tested sati sfactory.
Leakage was detected at a secondary handhole cover on the "B" OTSG. Minor steam cuts were dressed out of the shell gasket seating surface and a new cover installed. The "B" 0TSG was then retested satisfactorily.
t
Station Battery Replacement During April, the "A" Bank Station Battery change out was completed and retested satisfactorily per SP 1303-11.11.
M0 VATS Valve Testing Motor Operated Valve Analysis and Testing (M0 VATS) was completed during April. Thirty-three valves were tested and test data is being analyzed.
Control Rod Drive Mechanism Leaf Spring-Inspection A CRDM leadscrew nut leaf spring inspection was performed in April. All sixty-nine CRDM leadscrew nut leaf springs were inspected with no defects found.
Local Leak Rate Testing Local Leak Rate Testing was performed on ninety-four valves in April. Valves RB-V-7 and IC-V-3 failed initial testing and were repaired and retested satisfactory.
All other valves tested satisfactorily.
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OPERATING DATA REPORT DOCKET NO.
50-289 DATE 4/30/86 COMPLETED BY C.W. Smyth TELEPHONE (717) 948-8551 OPERATING STATUS NOTES
- 1. UNIT NAME:
THREE MILE ISLAND UNIT 1
- 2. REPORTING PERIOD:
APRIL
,1986.
- 3. LICENSED THERMAL POWER (MWT) :
2535.
- 4. NAMEPLATE RATING (GROSS MWE) :
871.
- 5. DESIGN ELECTRICAL RATING (NET MWE) :
819.
- 6. MAXIMUM DEPENDABLE CAPACITY (GROSS MWE) :
824.
- 7. MAXIMUM DEPENDABLE CAPACITY (NET MWE) :
776.
- 8. IF CHANGES OCCUR IN (ITEMS 3-7) SINCE LAST REPORT, GIVE REASONS:
- 9. POWER LEVEL TO WHICH RESTRICTED, IF ANY (NET MWE) i10. REASONS FOR RESTRICTIONS, IF ANY:
THIS MONTH YR-TO-DATE CUMMULATIVE
- 1. HOURS IN REPORTING PERIOD 719.
2879.
102216.
2'. NUMBER OF HOURS REACTOR WAS CRITICAL 181.7 1864.1 35680.7
- 3. REACTOR RESERVE SHUTDOWN HOURS 537.2 980.4 1863.7
- 4. HOURS GENERATOR ON-LINE 148.5 1814.5 34848.4
'5.
UNIT RESERVE SHUTDOWN HOURS 0.0 0.0 0.0
- 6. GROSS THERMAL ENERGY GENERATED (MWH) 343138.
4481474.
83831871.
- 7. GROSS ELECTRICAL ENERGY GENERATED (MWH) 110111.
1494635.
27886788.
- 8. NET ELECTRICAL ENERGY GENERATED
(. WH) 95619.
1393101.
26044813.
M
- 9. UNIT SERVICE FACTOR 20.7 63.0 34.1
- 0. UNIT AVAILABILITY FACTOR 20.7' 63.0 34.1
- 1. UNIT CAPACITY FACTOR (USING MDC NET) 17.1 62.4 32.6
- 2. UNIT CAPACITY FACTOR (USING DER NET) 16.2 59.1 31.1
- 53. UNIT FORCED OUTAGE RATE 14.7 11.7 63.0 Q4. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, AND DURATION OF EACH NONE l
%5. IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP:
l
r AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
50-289 UNIT TMI-1 DATE. 4/30/86 COMPLETED BY C.W. Smyth TELEPHONE (717) 948-8551 MONTH:
APRIL 1
I DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWE-NET)
(MWE-NET) 1
-6.
17
-15.
2
-6.
18
-15.
3
-5.
19
-15.
4
-5.
20
-18.
5
-5.
21
-36.
6
-6.
22
-42, 7
-5.
23 44.
8
-5.
24 173.
9
-6.
25 162.
10
-6.
26 727.
11
-6.
27 752.
12
-6.
28 788.
13
-6.
29 788.
14
-6.
30 790.
15
-6.
31 N/A-16
-14.
4 i
1 i
0 i
i i
4
i 50-289 UNIT SilUTDOWNS AND POWER REDUCTIONS DOCKET NO.
UNIT NAME TMI-I _
DATE April av, nuo APRIL COMPLETED BY C.W. Smyth REPORT MONTH TELEPHONE (/1/) 948-8551 j
e e
.! ?
3 hY Licensee P
Cause & Corrective No.
Date g.
3g
(
jg5 Event g?
g3 Action to j
$5 5
jgg Repor:
g' Prevent Recurrence H
mU c5 1
3/21/86 S
532 B
1 N/A CC HTEXCH Eddy Current Outage 2
4/23/86 F
6 G
3 86010 CH N/A Failure to Maintain Feedwater Requirements 3
4/23/86 F
14 A
N/A N/A HA TURBIN Restricted to 57% Power Due to Turbine Bearing Vibration 4
4/24/86 F
' 22 A
N/A N/A HA PIPE A Steam Drain Line Break i
2 3
4 F: Forced Reason:
Methat:
Exlubit G Instructions S: Scheduled A fquipment Failure (Explain) 1-Manual for Preparation of Data B Miintenance of Test 2 Manual Scram.
Entry Sheets for Licensee C Refueling 3-Autonutic Scram.
Event Report (LE R) File INUREG-D-Regulatory Restriction 4-Other (lisplam) 0161)
E-Operator Training & License I xamination F-Administrative 5
G Operational Eirus (Esplain t E stubit 1 - Same Source li-Other (E xplain) 6 Actually used Exhibits F & 11 NUREG 0161
REFUELING I W ORMATION REQUEST 1.
Name of Facility:
Three Mile Island Nuclear Station, Unit 1 2.
Scheduled date for next refueling shutdown:
Deceder 6,1986 3.
Scheduled date for restart following refueling:
May 4, 1987 4.
Will refueling or resumption of operation thereafter require a technical specification change or other license anendnent?
Yes If answer is yes, in general, what will these be?
Conversion to 18 month fuel cycle with associated changes i
to Power Idalance, Quadrant Tilt and Rod Insertion Limits.
If answer is no, has the reload fuel design and core configuration been reviewed by your Plant Safety Review Committee to deteralno whether any unreviewed safety i
questions are associated with the core reload (Ref.10 CFR Section 50.59)?
No If no such review has taken place, when is it scheduled?
To be determined.
5.
Scheduled date(s) for submitting proposed ifcensing action and supporting infornation:
Noveder,1986 i
l 6.
Inportant ifcensing considerations associated with refueling, e.g. new or different fuel design or supplier, unreviewed design or perfornance analysis methods, significant changes in fuel design, new operating I
procedures:
None 7.
The nuder of fuel asse211es (a) in the core, and (b) in the spent fuel storage pool:
(a) 177 (b) 208 8.
The present Itcensed 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 assedlies:
The present licensed capacity is 752. There are no planned increases at this time. '
9.
The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity:
1991 is the last refueling discharge which allows full f
core off-load capacity (177 fuel assemblies).
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GPU Nuclear Corporation Nuclear
- ,ome::,v8o s
Middletown, Pennsylvania 17057 0191 717 944 7621 TELEX 84-2386 Writer's Direct Dial Number:
May 13, 1986 5211-86-2089 Director, Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Document Control Desk
Dear Sir:
Three Mile Island Nuclear Station, Unit I (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Monthly Operating Report April, 1986 Enclosed please find two (2) copies of the April,1986 Monthly Operating Report for Three Mile Island Nuclear Station, Unit-1.
Si ncerely, l
H. D.
ill Director, TMI-1 HDH:DVH:spb cc: Dr. T. E. Murley i
Attachments
)
0015C t
GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation l
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