ML20093E673

Monthly Operating Rept for Sept 1995 for TMI-1

ML20093E673
Person / Time
Site:	Three Mile Island
Issue date:	09/30/1995
From:	Heysek W, James Knubel GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
C311-95-2405, NUDOCS 9510170027
Download: ML20093E673 (9)

Text

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l GPU Nuclear Corporation OM soute 44' so"t" P.O. Box 480 l

Middletown, Pennsylvania 17057-0480 (717)944-7621 Writer's Direct Dial Nurnber; (717) 948-8005 October 10, 1995 l 1

C311-95-2405 l U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 i

Gentlemen:

Subject:

Three Mile Island Nuclear Station, Unit I (TMI-1)

Operating License No. DPR-50 Docket No. 50-289 l Monthly Operating Report for September 1995 Enclosed are two copies of the September 1995 Monthly Operating Report for Three Mile Island Nuclear Station, Unit 1.

Sincerely, J. Knube Vice President and Director, TMI WGH Attachments cc: Administrator, Region I THI Senior Resident Inspector T95001 l

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- OPERATIONS

SUMMARY

September 1995 The plant entered the month operating at 100% power and remained at that level until 1700 on September 8 when reactor power was reduced to initiate the scheduled 11R refueling outage. The main generator breaker was opened at 1900. Net unit electrical output averaged approximately 770 MWe while the unit was on-line during September.

MAJOR SAFETY RELATED MAINTENANCE The following is a summary of major safety related maintenance items accomplished during the month.

10R Refueling Activities On September 8, at approximately 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />, the plant shutdown commenced for the scheduled 11R refueling outage. The Turbine was taken off line at 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br /> and Reactor shutdown occurred at 2057 hours0.0238 days <br />0.571 hours <br />0.0034 weeks <br />7.826885e-4 months <br />. Activities completed since the start of the outage include:

Refuelina Preparations were completed in the Fuel Transfer Canal and on the reactor head service structure for the reactor head removal. The full core was offloaded and ultrasonic inspection of all the off-loaded fuel assemblies was performed.

Cycle 10 operating reactor coolant radiochemistry data estimates of the number of defective fuel rods was confirmed by the ultrasonic test (UT) results.

l Nine failed fuel rods confined to four assemblies (three of the assemblies l

involved once-burned fuel) were identified by UT. An additional failed rod was found by eddy current testing.

Visual examination of some first burned fuel assemblies showed unusual distinctive crud patterns (DCP) on the top two spans. During reconstitution of the first once-burned assembly, eddy current testing identified that six rods had through wall defects. Eddy current testing of the intact fuel rods with DCP showed some to have cladding wall loss. A systematic program was initiated to determine the extent of the degradation and actions to assure the integrity of fuel assemblies to be reused in Cycle 11. Twenty fuel assemblies scheduled for use in Cycle 11 containing fuel rods having DCP were reconstituted using a maximum of ten stainless steel rods and donor rods of equivalent fuel burnup if more than ten rods per assembly were involved. The Cycle 11 core was redesigned to incorporate eight new fuel assemblies thereby limiting the extent of the reconstitution effort. Further examination to determine the cause of the DCP and associated fuel rod failures will occur during cycle 11.

Control Rod Drive Mechanisms On September 8, 1995 during pre-llR testing conducted concurrent with plant shutdown, control rod trip insertion times were obtained in accordance with 1

6 Surveillance Procedure 1303-11.1. Seven control rods exceeded the trip insertion time limit of 1.66 seconds from fully withdrawn to 3/4 insertion as specified in the TMI-1 Technical Specifications. The reactor was placed in cold shutdown for the 11R outage and 27 CRDMs and their thermal barriers were removed and replaced with new open flow path thermal barriers. These barriers are dimensionally similar to those used to replace the four barriers in June 1994; in that they have larger clearances in the region of the ball check valves, however the open flow path barriers are different in that one of the four check valve balls was eliminated. Inspection of the thermal barriers removed showed crud deposition in the area of the thermal barrier ball check

- valves that restricted movement of the balls and in the lead screw guide-bushing. These results further confirm the deposits as the cause of the slow rod insertion times.

Reactor Coolant Pump Maintenance.

The Reactor Coolant Pumps RC-P-1A/1B/1C and ID were inspected during hot shutdown. The high pressure oil lift hoses on all four pumps were replaced with new flexible stainless steel hoses and the motor oil leaks identified during operation were repaired. The RC-P-10 motor was thoroughly inspected as part of a routine preventive maintenance activity. The upper thrust bearing shoes were replaced in RC-P-1A, 1B and 10. The seal injection line for RC-P-1B was modified to increase the capacity of the temporary drain path 1 to support seal maintenance by keeping the water level below the seal cavity.

Both RC-P-1B and 1C were converted from the conventional seal package to a  ;

cartridge package. The conversion included modification of the seal leakoff l lines to accommodate the design differences associated with the seal cartridges. Vibration monitoring equipment was upgraded on all four RCPs during the outage.

Main Steam Safety Relief Valves i Main Steam Safety Relief Valve relief testing was performed while taking the plant off line for the 11R refueling outage. Although the results of all l valve tests were satisfactory, four valves (MS-V-17A, MS-V-17C, MS-V-21A and  ;

MS-V-218) were found to be leaking past the seat. Each valve was refurbished  :

and will be retested during plant startup.

Pressurizer Code and Safety Relief Valves Pressurizer Code and Safety valve RC-V-1A was satisfactorily tested in place.

It was replaced with a rebuilt and factory tested spare due to the original

, valve being outside the 1% Technical Specification "as left" tolerance. The Pressurizer Pilot Operated Relief Valve RC-RV-2 was replaced with a rebuilt tested spare. The valves that were removed will be sent off-site to the vendor for rebuilding and post maintenance testing after the 11R outage.

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i Once'Throuah Steam Generators RC-H-1A/B Maintenance'and Inspection Once Through Steam Generator work completed included removing both the 'A' and

'B' OTSG upper handhole and manway covers and installing ventilation equipment in support of eddy current testing (ECT) and tube plugging. The lower manway

covers were removed from both generators and the 'B' lower handhole cover was removed to repair a minor leak.

ECT.of 21% of the tubes and approximately half of the sleeves in the lane l wedge of each OTSG was performed. As a result the ' A' OTSG was categorized as "C-1" and examination of a larger sample population was not required. Based on the ECT results, OTSG 'B' was categorized as "C-2" due to two adjacent defective tubes from the 12% sample (a portion of the 21% total). Immediately adjacent unplugged tubes were additionally inspected to bound the defective tubes. No additional increase in sample population was required.

As'a result of the tube plug eddy current examination, 21 B&W Inconel (I) 600

. rolled plugs in the 'A' OTSG upper tube sheet were replaced with 21 I-690 rolled plugs and eight I-600 rolled plugs in the 'B' OTSG upper tube sheet were replaced with seven I-690 rolled plugs and one I-690 welded plug. On l completion of drip testing, all Mk3 explosive plugs remaining in service were i repaired with welded plugs (380 in the 'A' lower tube sheet and 79 in the 'B' lower tube sheet). One leaking I-690 rolled plug installed below a cable stabilizer in the 'B' lower tube sheet was repaired with a remote weld plug.

One Westinghouse rolled plug in 'B' lower was replaced and one Mk1 explosive plug in the 'A' lower was replugged as a result of drip testing.

Reactor Coolant Inventory Trendina System (RCITS) 4 In accordance with Technical Specification section 3.24, a report in the March 1995 Operating Report identified that the affected channels of the RCITS were

< isolated. The action was taken because of a 15 to 20 gpm reactor coolant leak

, caused by an-instrument tubing / fitting separation. The affected channels of RCITS were not restored to an OPERABLE status prior to the shutdown for the

. 11R outage. Since the shutdown, all tubing joints were inspected, the system has been repaired and will be hydrostatically tested before being restored to l operation next month. Procedural guidance for work on mechanical joints when pressure is greater than 200 psig and temperature is greater than 200'F has been appropriately incorporated.

l Fuel Transfer System Repairs

. Fuel Transfer System repairs performed involved the use of divers to adjust

! 'the underwater limit switches on the east fuel carriage ana the installation 2

of additional weight on the east fuel basket to ensure it would latch in the down position. The drive chain on the west fuel carriage was found broken and repair work is scheduled for a later date. l l

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1 Cold Leo Drain Line Reactor Coolant Leak A leak of approximately 20 drops per second was found in a weld on a nonisolable two inch diameter cold leg drain line. Metallurgical evaluations have determined that the failure was fatigue induced and that the fatigue occurred over a long period. The most probable cause for growth of the crack from an initial flaw is reactor coolant turbulent flow penetration into the drain line compounded by thermal stratification causing fatigue cycles in the weld. The four foot section of 'B' RCP suction drain line and the reducing elbow were replaced. Piping supports for all four RCP drain lines were reconfigured to reduce thermal stresses.

Electrical Bus Outages Bus outages were completed on the following electrical equipment:

• 10, ID and 1E 4160V busses

• AB-E; TRA and TRB; VBA, VBB and VBC; ATA and ATB distribution panels

• IH, IL, IP, IR, 1S and IT 480V busses

• 1A and 18 Auxiliary Transformers

• IE Inverter bus

• 1A and IB Control Rod Drive busses Miscellaneous Outage Work The following is a summary of various refueling outage activities accomplished:

• Completed 22 of 30 motor operated valve diagnostic tests (M0 VATS /V0TES).

• Completed 41 of 46 motor operated valve repairs and 22 of 27 Generic Letter 89-10 motor operated valve repairs.

• Completed 97 of 101 valve repacking actions.

• Completed 40 of 69 Local Leak Rate Tests.

• Completed 19 of 19 Technical Specification snubber tests.

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OPERATING DATA REPORT DOCKET NO. 50-289 DATE COMPLETED BY W G HEYSEK OPERATING STATUS TELEPHONE (717) 948-8191

1. UNIT NAME: THREE MILE ISLAND UNIT 1 l NOTES: l

. 2. REPORTING PERIOD: SEPTEMBER 1995 l l

3. LICENSED THERMAL POWER: 2568 l l

4. NAMEPLATE RATING (GROSS MWe): 872 l l S. DESIGN ELECTRICAL RATING (NET MWe): 819 l l

6. MAXIMUM DEPENDABLE CAPACITY (GROSS MWe): 834 l l

7. MAXIMUM DEPENDABLE CAPACITY (NET MWe): 786 l l l l

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) 720.0 6551.0 184776.0

12. NUMBER OF HOURS REACTOR WAS CRITICAL (HRS) 188.9 6019.9 107608.6

13. REACTOR RESERVE SHUTDOWN HOURS (HRS). -0.0 0.0 2284.0

14. HOURS GENERATOR ON-LINE (HRS) 187.0 6018.0 106473.1

15. UNIT RESERVE SHUTDOWN HOURS (HRS) 0.0 0.0 0.0

16. GROSS THERMAL ENERGY GENERATED (MWH) 478264 15351915 261268186

17. GROSS ELECTRICAL ENERGY GENERATED (MWH) 156749 5135569 87818823

18. NET ELECTRICAL ENERGY GENERATED (MWH) 144888 4848443 82503861

19. UNIT SERVICE FACTOR (%) 26.0 91.9 57.6

20. UNIT AVAILABILITY FACTOR (%) 26.0 91.9 57.6

21. UNIT CAPACITY FACTOR (USING MDC NET) 25.6 94.2 56.8

22. UNIT CAPACITY FACTOR (USING DER NET) 24.6 90.4 54.5

23. UNIT FORCED OUTAGE RATE (%) 0.0 0.0 36.3 UNIT FORCED OUTAGE HOURS (HRS) 0.0 0.0 60761.2

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: __ October 11, 1995 5

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AVERAGE DAILY' UNIT POWER LEVEL' i.

DOCKET NO. 50-289-UNIT .TMI-l' DATE COMPLETED BY W G HEYSEK TELEPHONE (717) 948-8191-

, MONTHi SEPTEMBER DAY ' AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-NET) (MWe-NET) i

1 790 17

-4

.2 801 18 -4 3 801 19 -4 4 799 20 -4 ,

i;~ 5 795 21 -3

, 6 795 22 -4 7 792 23 -4 l 8 587 24 -4 9 -32 25 -4 10 -14 26 -4 11 -5 27 -4 12 -4 28 -4 13 -4 29 -4 14 -4 30 -4 15 -4 31 NA 16 -4 1

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DOCKET N0. 50-289 UNIT MANE TMI-1  ;

REPORT MONTH September 1995 DATE  :

COMPLETED BY W. G. Heysek TELEPHONE (717) 948-8191 -

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e. meti,od ef ";;""* Sg',:= c - nent c"",c't .ct"t.'" "

Prevent Recurrence Type' Duration Reason

• shutting Reporte Code Date (Hours) Down *&*

Reactor * . ,

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l 9-08-95 S 533 C 1 95-002 AB AA From =100% power, the unit was shutdown to begin the

"" "'li"8 *"d"*i"**"*" "**8'- S"'**ilI*"

01 procedure 1303-11.1 was performed to obtain control rod trip insertion times. Seven rods exceeded the TS specified limit of 1.66 seconds. The thernet barriers on these control rods and others determined to be affected by the crud deposition in the area of the l

thermal barrier batt check valves that restricted

' movement of the balls and in the lead screw guide bushing were replaced during the 11R outage.

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1 2 3 4 F Forced Exhibit G - Instructions for Reason Method preparation of Date Entry sheets s scheduled A-Equipment Fa11ere (Emplatn) 1-Manual for Licensee Event Report (LEE)

B-Maintenance er Test 2-Manual scram File (NUREG-0161)

C-Refueling 3-Automatic serem D-Reguietory Restriction 4-O*her (Explatn) C E-Operator Training & Licensing Examinetten d Exhibit 1 same source F-Administrative

' " b Actually used exhibits F & II NUREG c161 H- r E a 7

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l REFUELING INFORMATION REQUEST

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1. Name of Facility: Three Mile Island Nuclear Station, Unit 1

2. Scheduled date for next refueling shutdown: NA

3. Scheduled date for restart following current refueling: October 11, 1995

4. Will refueling or resumption of operation thereafter require a technical specification change or other license amendment? No.

5. Scheduled date(s) for submitting proposed licensing action and supporting information: NA

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:

a) TMI is using the new Mark B10 fuel assembly in the Cycle 11 reload batch. The Mark B10 design meets all current BWFC fuel design criteria and is in use at other B&W 177 FA plants.

b) TMI also will use four new Westinghouse Lead Test Assemblies (LTA) in the Cycle 11 reload batch. Their planned operation is for three consecutive cycles with discharge at end-of-Cycle 13. The LTAs will meet current W fuel design criteria while operating within TMI core operating limits. LTA enrichment and core location will ensure that an LTA will not be the lead (hot) assembly at any time during the cycle and will not set any safety or operating limits. The LTAs will remain bounded by existing UFSAR safety analyses results.

l 7. The number of fuel assemblies (a) in the core, and (b) in the spent fuel storage pool: (a) 177 (b) 683 I

8. The present licensed spent fuel pool storage capacity and the size of any l increase in licensed storage capacity that has been requested or is planned, in number of fuel assemblies:

The present licensed capacity is 1990. Phase 1 of the reracking project to incr97 spent fuel pool storage capacity permits storage of 1342

' assemb5 ;. Upo;: completion of Phase II of the reracking project, the full lifensed capacity will be attained. Phase II is expected to be started in 2002.

9. The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity:

Completion of Phase I of the reracking project permits full core off-load (177 fuel assemblies) through the end of cycle 14 and on completion of the rerack project full core off-load is assured through the end of the current operating license and beyond.

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