ML20011E784
| ML20011E784 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 01/31/1990 |
| From: | Hukill H, Smyth C GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| C311-90-2014, NUDOCS 9002220467 | |
| Download: ML20011E784 (8) | |
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i GPU Nuclear Corporat6on .! NggIg{ Post Office Box 480 Route 441 South i j Middletown, Pennsylvania 17057 0191 717 944 7621 , TELEX 84 2386 Writer's Direct Dial Number: Febnnry 15, 1990 - C311-90-2014 , l j U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.- 20555 :
Dear-Sir:
j Three Mile Island Nuclear Station, Unit I (TM1-1) Operating License No. DPR-50 Docket No. 50-289 Monthly Operating Report January 1990 ,
' Enclosed are two copies of the January 1990 Monthly Operating Report for Three Mile Island Nuclear Station, Unit 1. ;
Sincerely,
. D. u ill i Vice President & Director, TMI I !
HDH/WGH:2014 ; cc: W.; Russell, USNRC F. Young, USNRC Attachments { P I/ ADOCK 05 900222046790gs;1g9 {DR GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation
. OPERATIONS SUtHARY [
JANUARY 1990 i It was previously reported that the plant was in a slow power reduction since December 15th, coasting down for the BR refueling outage. The plant was operated at reduced power until approximately 2200 hours on January 5th, when 1 the main generator was taken off line and the BR outage comenced. Refueling ; of the reactor commenced on January 26, 1990 with control rod shuffle and " continued into February. MAJOR SAFETY RELATED MAINTENANCE During January, the following major safety related maintenance activities were performed: Integrated Leak Rate Testing (ILRT) ; With the plant in cold shutdown, preparations for the Integrated Leak Rate Test were performed. The preparation activities included calibration and installation of test instruments, inspection and test of the six air compressors to be used to pressurize the Reactor Building, and performance of ; local leak rate testing of the Reactor Building purge valves (AH-V-1A/B/C/D). Repairs were required for AH-V-1C to eliminate a gasket leak before - satisfactory local leak rate test results were obtained for the valve. Bonnet leaks on'four Feedwater valves (FW-V-55A, 80A, 81A, and 82B) were sealed by injection of Furmanite compound. The Leak Rate system was aligned for pressurization on completion of the preparations and the Reactor Building pressurized to 50.6 psi. After a four hour stabilization period, the first twenty-four hour test was begun. This initial-ILRT was declared invalid due to leakage into the OTSGs. Several valves on the secondary plant outside of the ILRT valve lineup were found to be open and were closed. No significant leakage was identified as a result of inspection of the containment isolation valves and boundaries. Unstable ; Reactor Building temperature conditions caused a second ILRT to be declared " invalid at approximately the ten hour mark. A third test was started after securing the Reactor Building Industrial Cooler system. This third ILRT and a six hour superimposed test were completed and their results declared satisfactory. The Reactor Building was depressurized and systems returned to normal. Fuel Handling Systems Repair and modification to the Fuel Handing equipment was performed in preparation for its use in fuel handling activities. Maintenance and repairs included repairing the telescopic cylinder down light and hose replacement on the control rod-mast of the spent fuel bridge, replacing a ten amp breaker on ; the Reactor Building auxiliary bridge, and replacing hoses with stainless steel tubing on the Reactor Building upenders. The modification involved installation of pointers to aid the operators in positioning both the Spent Fuel Pool and Reactor Building Fuel Transfer Canal fuel upenders to ensure a full vertical position. Divers installed the pointers on the Fuel-Pool fuel upenders. E : I '
. I * . l Reactor Vessel (RC-T-1)
In preparation for refueling the reactor vessel, the following work items were accomplished. The reactor head micsile shields were removed and the Fuel Transfer Canal stairway installed. The reactor vessel head insulation was removed and the RV studs detensioned. The RV studs were parked and the guide j studs installed. A modification to install permanent reactor vessel head ; lifting pendants was started with the attachment of the pendants to support head lift. The remainder of the modification work will be completed with the head on the head stand. The Reactor Coolant Inventory Tracking system lines were disconnected and the head fans removed. The "A" to "B" cables, "P!" tube cables and thermocouple cables were disconnected from the head and all CRDM l position indicators were then removed. The CRDM cooling water lines were disconnected and the APSR/CRDMs were uncoupled and the lead screws parked. The : flood line cover plate and the fuel transfer tube flanges were removed. The ' head was removed from the vessel and transferred to the head stand. The indexing fixture was installed and the plenum removed from the vecsel and stored in the deep end of the canal. The stud holes were sealed and the vessel i seal plate installed and sealed. The reactor dosimetry was replaced and the "NI" cover plate reinstalled and sealed. The reactor vessel internal vent , valves were exercised and the vessel incore detectors parked. And finally the ; fuel canal was flooded in preparation for fuel movement. Control Rod Drive Mechanisms (CRDM) During preparations for lifting and moving the reactor vessel head, leakage from several of the CRDM nozzle flanges was identified. After investigation, fourteen leaking CRDMs were repaired. The stators from the leaking CRDMs were removed while the head was on the head stand. A shielded platform was constructed at the head to facilitate removal of the CRDM hold down bolts. The CRDM motor tubes were removed to a vertical storage rack. Drilling of the CRDM split nut rings and additional repair work will continue into February. Once Through Steam Generators (RC-H-1A/B) Once Through Steam Gerrator (OTSG) work completed in January included opening of the upper primary m9nways and handboles on both "A" and "B" OTSGs and ventilating both units through the handholes. Bubble test rigs were installed and the bubble tests performed. Though two tubes were rolled as a result of the bubble tests in the "A" OTSG, the tests were deemed inconclusive and the OTSGs drained. The lower primary mant:ays were removed in preparation for Installation of the Cold 1.eg Dams. Main Steam Safety Valves The Main Steam Safety Valve work performed during January resulted in seven valves being repaired. The following is a summary of that work:
- 1. MS-V-17B, MS-V-17D, and MS-V-20D had seats lapped,
- 2. MS-V-20B, MS-V-21A, and MS-V-21B had discs replaced and seats lapped,
- 3. and MS-V-18A had the disc and stem replaced and seats lapped The valves were reassembled and prepared for testing when the plant achieves a hot shutdown condition.
2
Pr'essurizer Code Safety Valve (RC-RV-1A1 Pressurizer Code Safety valve RC-RV-1A was replaced with a rebuilt / tested valve. The valve removed was packaged and shipped to Wyle Laboratories for inspection and testing. Results of the tests are due in February. l , Miscellaneous Work Items i Following is a summary / status of various other work items being performed j during the BR outage:
- 1. Valve repacks - 307 of 427
- 2. Valve replacements - 66 of 175
- 3. Large valve seat repairs - 2 of 13 ,
- 4. Local leak rate tests - 102 of 115 l MOVATS valve operator tests - 9 of 52
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- 6. Conex wire inspections - 53 of 53
- 7. Molded case breaker tests - 65 of 154
- 8. 4160V breaker tests - 8 of 8
- 9. 480V breaker tests - 40 of 44 '
- 10. Tech Spec snubber seal replacements - 28 of 37 '
- 11. Tech Spec snubber functional tests - 20 of 20 ,
- 12. Non-Tech Spec snubber functional tests - 1 of 38 Note: snubber numbers reflect completed testing and not reinsta11ation.
pr " c f 4 OPERATING DATA REPORT DOCKET NO. 50-209 DATE 01/31/90 COMPLETED BY C.W. Smyth TELEPHONE (717 F948-8551._ OPERATING STATUS , l NOTES l le UNIT NAME: THREE MILE ISLAND UNIT 1 1 I 2; REPORTING PERIOD: JANUARY ,1990. I I 3.; LICENSED THERMAL POWER (MWT): 2560. I I
& NAMEPLATE RATING (GROSS MWE): 871. I I 52 DESIGN ELECTRICAL RATING (NET MWE): 819. 1 I MAXIMUM DEPENDABLE CAPACITY (GROSS MWE): 05G. 1 I
-7; MAXIMUM DEPENDADLE CAPACITY (NET MWE): 800. l .I i i Oc IF. CHANGES OCCUR IN (ITEMS 3-7) SINCE LAST REPORT, GIVE REASONS:
9 POWER LEVEL TO WHICH RESTRICTED, IF ANY (NET MWE) ___
- 10. CEASONS FOR RESTRICTIONS, IF ANY:
a d THIS MONTH YR-TO-DATE CUMMULATIVE 11, HOURS IN REPORTING PERIOD 744.- 744. 135145.
-12 NUMBER OF= HOURS REACTOR WAS CRITICAL 118.1 118.1 62116.7 13 -REACTOR RESERVE SHUTDOWN HOURS 0.0 0.0 2002.8.
14 2-HOURS GENERATOR ON-LINE 117.1 117.1 61113.3 [ 15. UNIT RESERVE SHUTDOWN HOURS 0.0 0.0 0.0: 1 ;. GROSS THERMAL ENERGY GENERATED (MWH) 235434. 235434. 149607329. [171_ GROSS ELECTRICAL ENERGY GENERATED (MWH) 82799. 82799. 50389468. 104 NET ELECTRICAL ENERGY GENERATED (MWH) 72500. 72508. 47258997. 103 -UNIT SERVICE FACTOR 15.7 15.7 45.2 , 20. UNIT AVAILABILITY. FACTOR ~15.7 15.7 45.2 21e UNIT CAPACITYLFACTOR (USING'MDC NET) 12.1 12.1 44.G
- 22. UNIT CAPACITY FACTOR (USING DER NET) 11.9 11.9 42.7 23e UNIT FORCED OUTAGE RATE O.0 0.0 49.7 i:
h 24, SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE. AND DURATION OF EACH-Presently shutdown for 8R Outage. . fll~25.. IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP: 2/25/90 4 L e La 1 - l
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11 Delece (Liplom t 6 L Actiselly emed Exhibitm l F & H NUREC 016I [
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, . . i REFUELING INFORMATION REQUEST ;
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- 1. Name of Facility: Three Mile Island Nuclear Station, Unit 1 l
- 2. Scheduled date for next refueling shutdown: October 4,1991(9R)
- 3. Scheduled date for restart following current refueling: February 25, 1990 (8R) !
l 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 Seccion 50.59)? No If no such review has taken place, when is it scheduled? t February 22, 1990
- 5. Scheduled date(s) for submitting proposed licensing action and supporting information: None Planned ;
- 6. Important licensing considerations associsted with refueling, e.g. new or i different fuel design or supplier, unreviewed design or performance i analysis methods, significant changes in fuel design, new operating .
l_ procedures: None
- 7. The number of fuel assemblies (a) in the core, and (b) in the spent fuel storage pool: (a) 177 (b)440
- 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 assemblies: The present licensed capacity is 752. Planning to increase licensed capacity through fuel pool reracking is in process. I
- 9. The projacted 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 core off-load capacity (177 fuel assemblies).
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