ML20086F502
| ML20086F502 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/30/1995 |
| From: | Broughton T, Heysek W GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| C311-95-2282, NUDOCS 9507130146 | |
| Download: ML20086F502 (7) | |
Text
,
GPU Nuclear Corporation NhhIhhf Route 441 South P.O. Box 480 Middletown, Pennsylvania 17057-0480 (717)944-7621 Writer's Direct Dial Number:
(717) 948-8005 July 7,1995 C311-95-2282 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.
20555 Gentlemen:
Subject:
Three Mile Island Nuclear Station, Unit I (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Monthly Operating Report for June 1995 Enclosed are two copies of the June 1995 Monthly Operating Report for Three Mile Island Nuclear Station, Unit 1.
Sincerely,
$UN A
T. G. Broughton v
Vice President and Director, THI WGH Attachments cc: Administrator, Region I TMI Senior Resident Inspector T95001 j
9507130146 950630
~
PDR ADOCK 05000289 i
R PDR GPU Nuclear Corporation is a subsidiary of General Public Utilities Corporation
OPERATIONS
SUMMARY
June 1995 The plant entered the month operating at 100% power and remained at that level throughout the entire month. Net unit electrical output averaged approximately 796 MWe during June.
MAJOR SAFETY RELATED MAINTENANCE The following is a summary of major safety related maintenance items accomplished during the month.
Heat Trace Panel 4A/B Heat Trace Panel 4A/B was removed from service when investigation of an alarm light revealed the heat trace at Decay Heat valve DH-V-29 was burned open.
Insulation was removed to permit replacement of the heat tracing.
Reinstallation of the insulation will continue into July.
Boric Acid Injection Pumps CA-P-1A/B Pumps CA-P-1A/B were removed from service because of a ruptured Pulsatrol diaphragm.
The Pulsatrol surge chamber was disassembled and the diaphragm replaced. After reassembly, CA-P-1A/B were satisfactorily tested and returned to service.
Decay Heat Removal Pump DH-P-1B DH-P-1B was removed from service to attempt to change the natural frequency of the pump and reduce vibration. A dial indicator was set up on the pump shaft in the area of the coupling to measure possible defection expected to result from the addition of lead weight to the pump casing and bearing housing support bracket.
No rubbing or deflection of the shaft was evident when the shaft was hand rotated after initially loading the casing with 123 pounds of lead or after adding an additional 60 pounds.
After four lead bricks, weighing approximately 26 pounds each were attached to the pump bearing housing support bracket, the dial indicator showed 0.001" vertical shaft movement. When the pump was hand rotated, rubbing and chatter was evident at one area (approx. 45 degrees of rotation). The rubbing, chattering and shaft deflection remained after all lead was removed from DH-P-1B.
After repositioning the bearing housing support bracket and further hand rotation, little or no rubbing was noted.
Following an alignment check of the pump / motor with acceptable results, the pump was started and a high pitched noise was heard. The pump was immediately shutdown.
The pump ran for approximately 15 seconds and achieved a maximum flow 400 GPM.
Pump amps, vibration and coastdown were observed to be acceptable. The pump was hand rotated again and moderate rubbing and chattering was evident in the same angular area as before.
Dur.ing a meeting, an evaluation of pump characteristics was done and since vibration was normal and the noise dissipated during the end of the 15 second run, it was decided to run the pump for one minute observing amp, flow, and vibration parameters and listen for unusual noises.
The noise recurred but stopped after approximately 4 seconds when DH-P-1B was restarted. The pump run continued for the duration of the performance of in-service testing.
After the pump was shutdown, the shaft hand rotated freely with no evidence of rubbing.
DH-P-1B was declared operable after satisfactory completion of the 1
in-service testing, l
1
l
~4 OPERATING DATA REPORT DOCKET NO.
50-289
-DATE July 7,1995 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:
JUNE 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 1
- 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:
i THIS MONTH YR-TO-DATE CUMMULATIVE
- 11. HOURS IN REPORTING PERIOD (HRS) 720.0 4343.0 182568.0
- 12. NUMBER OF HOURS REACTOR WAS CRITICAL (HRS) 720.0 4343.0 105931.7
- 13. REACTOR RESERVE SHUTDOWN HOURS (HRS) 0.0 0.0 2284.0
- 14. HOURS GENERATOR ON-LINE (HRS) 720.0 4343.0 104798.1
- 15. UNIT RESERVE SHUTDOWN HOURS (HRS) 0.0 0.0 0.0
- 16. GROSS THERMAL ENERGY GENERATED (MWH) 1846495 11141833 257058104
- 17. GROSS ELECTRICAL ENERGY GENERATED (MWH) 606630 3738424 86421678
- 18. NET ELECTRICAL ENERGY GENERATED (MWH) 573186 3532206 81187624
- 19. UNIT SERVICE FACTOR
(%)
100.0 100.0 57.4
- 20. UNIT AVAILABILITY FACTOR
(%)
100.0 100.0 57.4
- 21. UNIT CAPACITY FACTOR (USING MDC NET) 101.3 103.5 56.6
- 23. UNIT CAPACITY FACTOR (USING DER NET) 97.2 99.3 54.3
- 23. UNIT FORCED OUTAGE RATE
(%)
0.0 0.0 36.6 UNIT FORCED OUTAGE HOURS (HRS) 0.0 0.0 60761.2
- 24. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE AND DURATION OF EACH):
j Re-fueling outage / September 8, 1995 / 45 days j
- 25. IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP:
2
AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
50-289 i
UNIT TMI-1 DATE Julv 7.1995 COMPLETED BY W G HEYSEK TELEPHONE (717) 948-8191 MONTH:
JUNE DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-NET)
(MWe-NET) 1 800 17 799 2
794 18 796 3
795 19 790 4
799 20 784 5
800 21 789 6
798 22 797 7
791 23 798 8
790 24 797 9
803 25 791' 10 799 26 789 11 793 27 794 12 801 28 801 13 806 29 796 14 801 30 790 15 804 31 NA 16 800 5
3
a 9
DOCKET NO.
50-289 UNIT NAME TMI-1 REPORT MONTH June 1995 DATE July 7,1995 COMPLETED BY W. G. Heysek TELEPHONE (717) 948-8191 Licensee System Cause & Corrective
%+
Method of Event Code Component Action to Shuttfag Report #
Code prevent Recurrence Type' Duretten Reasona Date (Hours)
Down Reactor
- e &*
None 1
2 3
4 F Forced Enhtbit G - Instructions for Reason Method preparation of Data Entry Sheets S Schedeled A. Equipment Fallure (Esplein) 1-Manual for Licensee Event Report (LER)
B Maintenance or Test 2-Manual Scram Ftte (faiREG-0161)
C-Refueling 3-Automatic Scram D-Regulatory Restriction 4-Other (Emplain)
EJ Exhibit 1 same source E-Operator Trotning & Licenstag Enamination F. Administrative (Emplain) 2 0 *jg ;jainy e b Actually used enhibits F & !! NUREG 0161 4
REFUELING INFORMATION RE0 VEST
- 1. Name of Facility:
Three Mile Island Nuclear Station, Unit 1
- 2. Scheduled date for next refueling shutdown:
September 8, 1995
- 3. Scheduled date for restart following current refueling: NA
- 4. Will refueling or resumption of operation thereafter require a technical specification change or other license amendment?
YES.
See 6.c and d below.
- 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 will use the new Mark B10 fuel assembly in the Cycle 11 reload batch which is an upgraded design of the Mark B9 assembly used in Cycle
- 10. The Mark B10 provides a leaf-type cruciform assembly holddown spring to replace the previous coil spring design which has experienced random failures during operation and requires visual inspection each outage. 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 four M LTAs inserted in Cycle 9 were discharged at EOC-9 due to detection of fuel rod failures caused by grid-to-rod fretting similar to that seen in M Vantage 5H fuel designs. The Cycle 11 LTAs will use the generic H recommended design fix of rotated intermediate spacer grids to minimize flow-induced fuel vibrations and thus eliminate fretting. A prototype LTA was flow-tested to demonstrate the effectiveness of the fix. The production LTA will use ZIRLO fuel rod cladding, guide tubes and instrumentation tube and intermediate grids in place of Zircaloy 4 materials used for the Cycle 9 LTAs.- Otherwise, the Cycle 11 LTA design is basically the same as the Cycle 9 design.
The LTAs will meet current M 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.
c) GPUN plans to place two types of BWFC advanced non-zircaloy cladding in TMI-1 Cycle 11; eight rods each. The two types will be equally distributed in two Mark B10 fuel assemblies; one rod of each material 5
~,,;
t-in,each of the four peripheral rows per assembly. These cladding materials are also being irradiated in the McGuire reactor and other international reactors with no negative performance observed. Use of
~
the cladding requires NRC acceptance of an exemption request regarding 10CFR50.46 which was submitted as-Attachment II to Technical-Specification Change Request number 251.
Exemption approval was requested no later than July 21, 1995 to support final fuel delivery schedules for the 11R Outage.
d) GPUN submitted Technical Specification Change Request number 252 to relocate the volume and boron concentration requirements of TS section 2.3 for the chemical addition system boric acid mix tank and reclaimed boric acid storage tanks to the existing TMI-1 Core Operating Limits.
7 Report (COLR). The proposed change is consistent with the intent of NRC Generic Letter 88-16 guidance. The change will preclude the need for a TSCR to revise the parameters to meet Cycle 11 specific requirements.
- 7. The number of fuel assemblies (a) in the core, and (b) in the spent fuel storage pool:
(a) 177 (b) 601
- 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 1990.
Phase 1 of the reracking project to increase spent fuel pool storage capacity permits storage of 1342 assemblies. Upon completion of Phase II of the reracking project, the i
full licensed capacity will be attained.
Phase II is expected to be
[
sta.ted 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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