ML19330A938

Monthly Operating Rept for June 1980

ML19330A938
Person / Time
Site:	Three Mile Island
Issue date:	07/15/1980
From:	Mitchell D METROPOLITAN EDISON CO.
To:
Shared Package
ML19330A935	List: ML19330A934 ML19330A938
References
NUDOCS 8007290805
Download: ML19330A938 (6)

Text

.

OPERATING

SUMMARY

) 315' 2" Icvel range.

On May 30, 1980, the Reactor Coolant System was drained to the 314' 10" to

(

This level is 51 to 56 inches above the top of the core.

The drain down was made to enable the following work:

1.

High Pressure Injection cross tie (Restart Modification RM-14).

2.

Examination of various primary system check valves performed as a result of IE Circular 78-15 (also see TMI-I LER 80-03).

While in this condition, several precautions were taken to ensure adequate core cooling:

1.

Actual icvel indication of each loop was monitored using tygon tubing.

2.

Both decay heat loops have remained operable while drained down.

3.

Special operating procedures instituted for reinitiating decay heat removal in the event level would go below the decay heat removal line tie-ins.

On June 27, 1980, a spill occurred in the TMI-I Reactor Building as a result of an inadvertant actuation of the Emergency Safety Actuation System. The actuation occurred when vital bus A was reenergized following a transfer of the inverters-This opened valve DH-VSA allowing bor'ted water to enter the Reactor Coolant System (RCS).

The additional water caused the RCS level to increase to the point where primary coolant spilled through open manways in the Once Through Steam Generators. The total volume of the spill was approximately 10,000 gallons. There were no over exposures or contaminated individuals and air sample activities less than maximum permissible concen-trations.

The spill area was promptly decontaminated and normal activities in the Reactor Building resumed.

MAJOR SAFETY RELATED MAINTENANCE While the Unit remained in cold shutdown, with restart modifications in progress, the following major maintenance was performed:

1.

The Emergency Diesel Generator (EG-Y-1A) inspection and repairs were performed with satisfactory results. Tech Spec Surveillance 1301-8.2 was performed as the inspection item.

Repair included inspection and repair of Cenerator bearing and insulating; replaced turbocharger gaskets; replaced ex-pansion joint gaskets; repaired water bypass fittings; repair plenum; and painted muffler.

This work was performed with the "B" Decay heat loop service.

2.

Eddy Current testing of the B OTSG was performed during June.

Setup work included:

a)

Building clean room around upper and lower primary side manways.

b) Removal of upper handhole and manucy c) Building eddy current test platform for test equipment.

d) Service (breathing air supply, electrical, lighting).

3.

The pressurizer code safety valve (RC-RV-1B) was removed from the pressurizer for Tech Spec Surveillance testing. Testing and reinstalling valve back on pressurizer is expected to ocent in July.

After the valve was removed, a cover was installed ove; obe open pipe.

)

8007290

4.

RM-14, High Pressure Injection Cross-Connect - prefabrication work was performed prior to draining the RCS down.

Work performed during the drain down included:

a) cutting pipe b) pipe preparation c) fitup of pipe d) hot / root pass e) information x-ray of hot / root pass f) hot / root pass repair as required g) weld out of weld h) final x-ray 1) flush of piping The x-rays and flushing were perforned wit".. satisfactory results.

5.

Stop Check Valve Inspections - (Ref:

IE Circular 78-15 and LER 80-03).

The High Pressure Injection valves (MU-V-101 A/B/C/D; MU-V-86 A/B; MU-V-94, 95) work included:

a) Removing bonnet b) internal inspection / measurements c) remove internals for modification d) bonnets reinstalled for raising of RCS IcVel During the inspection, measutement and modification of the internals, it was determined that new internals were required to properly repair the check valves.

Because of plant conditions and the mode of operation, it was determined to re-install the bonnets at this time to raise RCS level and install new internals later this year (projected date of August,1980).

The Decay Heat stop check valves (DH-V-22 A/B work was also performed on one (1) valve at a time with the valve retest being performed prior to removing the other valve. Work included:

4 a) removing bonnet b) inspection / measurements of valve internals c) removal of valve internals d) modify valve internal (hinge pin and lock device) e)

reinstalling valve internals f) inspection g) reinstall valve bonnet All work was performed with satisfactory results and with one (1) operable decay heat loop in operation.

6.

The Borated Water Storage Tank (BWST) 24" flange work included:

replacing existing bolts - one at a time; and reinjecting furmanite to stop leak. -Repairs were performcd with satisfactory results.

OPERATING DATA REPORT DOCKET NO.

50-289

"'Y

' I980 DATE COMPLETED 11Y

...D. G v t tchel1 TELEPHONE (215) 921-6579 Ol'El(ATING STATUS Three Mile Island Nuclear Station. Unit I

1. Unit Name:

3""" l980

2. Iteporting Period:

2515

3. Licensed Thermal Power (MWI1:

871

4. Nameplate Rating (Gross MWe):

819

5. Design Electrical Rating (Net MWel:

840

h. Masimum Dependable Capacity (Grou MWe).. 776

7. Masimum Dependable Capacity (Net MWe):

h. If Changes Occur in Capacity Ratings (Items Number 31hrough 7i Since Last Report.Gise Reasons:

9. Power Lesel To which Restricted lf Any(Net MWes:

I

10. Reasons For Restrictions.lf Any:

This Month Yr to-Date Cumularise 720.0 436 7.0 s1006.0

11. Hours in Reporting Period 0.0 0.0 31731,8

12. Number Of Hours Reactor Was Critical 0.0 0.0 839,5

13. Reactor Reserse Shutdown Hours U.0 0.0 31180.9

14. Ilours Generator On.Line 0.0 0.0 0.0

15. Unis Rewrve Shutdown Hour,

16. Grow Thermal Energy Generated IMWil

[1. 0 0.0 76b110'il.

O.

u.

O.

17

(, row Electrical Energy Generated (MWill u.

2384003T.

Ih. Net Electrical Energy Generated (MWil)

U.0 0.0 g,,g

19. Unit Service Factor

_070 u,o 7,

20. Unit Asailability Factor 0.0 0.0 59,3

21. Unit Capacity Factor iJsing MDC Net 1 0.0 0.0 57.0

22. Unit Capacity Factor (Using DER Neil 100.0 100.0 28.6

23. Unit Forced Outage Rate

24. Shutdowns Scheduled Oser Nest 6 Months (Type.Date,and Duration of Eacht:

25. If Shut Demn At End Of Report Period. Estimated Date of Startup:

26. Units in Test Status (Prior o Commercial Operations:

Forecast Achiesed INITIA L CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPER ATION tu/7" s n.

AVER AGL 0411T trN1g POhER LEVEL DOCKET NO.

50-289 TMI-I S'IT DATE I"I.V_ Ij, 1980 COMPLETED BY D-G. Mitcliell TELEPHONE (215) 921-6579 MONTH lune 1980 DAY AVER AGE DAll P

'ER LEVEL 1)W AVER AGE DAILY POWER LEVEL (MWe-Net i I

O I7 0

0 0

33 l

o 3

0 0

4 0

20 0

5 0

0 6

0 22 7

0 O

0 0

x 0

25 0

0 10 0

0 33 27 0

p-2h 0

33 29 0

I4 -

30 0

0 15 31 0

.i 0

is 9

y e

50-289 DOCKET NO.

UNIT SilUTDOWNS AND POWER REDUCTIONS UNIT N A\\lE Dil-1 iulv 15. loso DATE COMPLETED BY D. C Mitcholl REPORT MONTH June 1980 TELEPHONE (215) 921-6579 "5

C

~;

e es 24

• e.

Cause & Correctise

.i ?

g 3sx Licensee j e, g

N...

Date

(

4 2sE Event g?

['8 A~

Action to

-=

$5 E'

j di, g Report e mO Present Recurrence H

d ep a y estraint Order 1

6/1/80 F

720 D

1 i

2 3

4 F: Forced Reason:

Method:

Edubit G Instructions 5: Scheduled A Equipment Failure (Explain) 1 Manual for Preparation of Data B-Maintenance of Test 2 Manual Seram.

Entry Sheets fe Licensee C-Refueling

.t-Autonuti. Scram.

E-ent Repart (LERI File int:RLG-D-Regulatory Restriction 4-Other ( E splain) 016!i E Operator Training 1 Licente i sami:ution F-Administrati.e 5

G-Operational Error (Esplain t Esh. bit ! Sam S urce Il Other (Esplain t

• a *. 'd REFUELING INF0,lptATijoN_ REOUEST 1.

Name of Facility:

Three Mlle Island Nuclear Station, Unit I 2.

Scheduled date for next refueling shutdown:

Unknown 3.

Scheduled date for restart followinn refueling:

Unknown 4.

Will refueling or resumption of operatiam thereaf ter require a technical k.,

specification change or other license amendment?

If answer 18 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 Keview Conanittee to determine whether any unreviewed safety questions are associated with the core reload (Ref. 10 CFR Section 50.59)?

If no such review has taken place, when is it scheduled?

Amendment No. 50, Cycle 5 reinau, was approved on 3-16-79.

5.

Scheduled date (s) for submit ting proposed licensing action and support ing information:

N/A 6.

Important licensing considerations associated with refueling, e.g. new or different fuel design er supplier, unreviewed design or performance analysis methods, significant channes in fuel design, new operating procedurca:

N/A 7.

The number of fuel assemblies (a) in the core, and (b) in the spent fuel l

storage pool:

(a) 177 (b) 208 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 i

planned, in number of fuel assemblies:

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:

1986 is the last refueling discharge which allows full core off-load i

capacity (177 fuel assemblies).

h-i