ML20064H246

Monthly Oper Rept for July 1978

ML20064H246
Person / Time
Site:	Fort Saint Vrain
Issue date:	07/31/1978
From:	Hillyard H PUBLIC SERVICE CO. OF COLORADO
To:
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ML20064H243	List: ML20064H240 ML20064H246
References
NUDOCS 7812180245
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Text

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4 PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION MONTHLY OPERATIONS REPORT NO. 55 Jt1Y, 1978 it i

i 78121802'[X 'ET

l

( , This report contains the highlights of the Fort St. Vrain, Unit No. 1 activi-ties, operated under the provisions of the Nuclear Regulatory Commission Operating License, DPR-34. This report is for the month of July,1978.

1.0 NARRATIVE

SUMMARY

OF OPERATING EXPERIENCE AND MAJOR SAFETY RELATED MAINTENANCE The investigation of the loss of insc$ ament power bus 2, reported last month, was inconclusive. The two blosu fuses were replaced and no further problems have been encountered.

The buffer helium dryer bypass modification described in the previous report, was installed.

The. reactor was taken critical June 30, 1978, and returned to power as rapidly as the moisture concentration in the primary coolant permitted.

The turbine generator was resMchronized with the system on July 4,1978, at a reactor power of 28%. P. ant operation was restricted by the avail-ability of only one helium circulator in each loop. Helium circulator lA was out of service due to an inoperable operator on the inlet steam isolation valve, and circulator 1C in Loop 2 was indicating high buffer-mid-buffer differential pressure with the main drain controller in auto-matic.

The buffer-mid-ouffer problem on "C" circulator was cleared by increasing

,the circulator speed. Helium circulator inlet steam isolation valve was manually opened and the circulator returned to service on July 6,1978.

The reactor power level was increased to 40% (110 We) on July 6,1978, and to 52% (162 We) on July 8,1978.

On July 7,1978, a localized fire occurred in the turbine building at a hydraulic valve operator that leaked hydraulic fluid on the hot reheat steam piping. The fire was readily extinguished using hand held fire extinguishers. No permanent damage was done by the fire.

The reactor power was increased to 58% (175 We) on July 11, 1978, and operated at this level until July 14, 1978, when an unplanned trip of circulator ID occurred during surveillance testing of the Plant Protective System. Failure of an integrated circuit chip caused this trip. During the recovery of 1D circulator, circulator 1C spuriously tripped causing loss of primary coolant flow through the Loop 2 steam generators. The reactor was manually scrammed following this occurrence.

The plant was returned to power, the turbine generator resynchronized and power increased to 58% (175 We) on July 17, 1978.

Problems with high primary coolant total oxidants are continuing.

Two trips of instrument bus inverter 1C occurred during this period, (July 13,19 78, and July 14, 1978). These trips were caused by failed silicon controlled rectifiers.

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• During this period, additional thermocouples were installed in the core region outlet thermocouple assemblies. Two modified control rod drive

, and orificing assemblies were received on site for insertion into the core at the first refueling if cdditional information is required in the investigation of the primary systa temperature fluctuations.

The plant operated between 45% (133 We) and 65% (195 We) until July 31, 1978. During the evenings, with 1 swer system power demands, the power level was decreased to 45% to reduce the core outlet temperature below 1,200*F. This mode of operation reduces the total time of operation above the 1,200*F, which is the threshold for the 10 ppa total oxidants condi-

tional limit on primary coolant impurity levels.

The cause for the high primary coolant imourity levels continues to be inves tiga ted. Operation with normal circulator buffer helium flows re-suits in an increase in primary coolant impurity levels. However, with equal buffer helium supply and return flows, reduced but higher than ex-pected impurity levels are still experienced.

On July 31, 1978, a winding failure and insulation fire occurred in 450 volt transformer 1A and the plant was manually scranuned. The loss of the 480 volt transformer caused an upset in the buffer helium system for Loop 1, resulting in ingress of an unknown amount of moisture into the primary system. .

It is anticipated that the plant will be returned to power in early August with the increase in power level being limited by the removal of the moisture injected into the primary system during the July 31, 1978, upset.

2.0 SINGLE RELEASES OF RADIOACTIVITY OR RADIATION EXPOSURE IN EXCESS'OF 10% OF IRE ALLOWABLE ANNUAL VALUE None 3.0 INDICATIONS OF FAILED FLTL RESULTING FROM IRRADIATED FUEL EXAMINATIONS None 4.0 MONTHLY OPERATING DATA REPORT Attached i .

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k OprnATrne NATA REronT nocrET No. .0-267 DAfg 780301 COMP *.ETED BY H. W. tUllyard, Jr.

. TELErnoxE (303) 755-2253 OPERATINC STATUS '

NOTES-

1. nest n==e: Fort St. Vrain, Unit No.1

2. neporting period: 780701 to 780731

3. 1.icensed Thermal Power (Wt): 842 .

4. Nameplate Rating (Cross We): 342 -

3. Design Electrical Ratins Otet Wa): 330 l

6. M=== Dependable Capacity (Cross We): 342 l

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7. M== Dependable CapacJty Olet We): 330 ,-

, 8.

If Giangas occur in Capacity Ratings (Items Number 3 Through- 7) Since Last Report, Cive Reasons:'

None l .

9. Fever Level To Which Restricted. If Any (!!at We): 231 *

10. Reasons for Restrictions If Any: NRC Restriction (70%) pending resolution of cerchin

. Final Safety Analysis Report and Technical Specification Bases discrepancies.

., This unit is in the power ascension phase of startup testing program..

This Month Y' ear to Date Cumulative *

u. no.rs in r.cyort ns Porioa 744 - 5,087 -

12. Number of Hours Reactor was Critical 713.3 3,881.0 10.515.5

13. Reactor Reserve Shutdown sours 0 . 0 0

14. sous t .nerator on-t.ine 581.9 2,584.3 5,922.3

15. Dr t Reserse shutdown nours 0 0 0

16. Cross Thernal Energy Cenerated (tsal). 274,444 1,183,219 2,342,007 *

17. cr e tiectrica' Enersr cenorated Osat) 91,391 389,693 674,570

18. net Exactrical Energy cenerated (tini) 85.839 362,407 604,766

19. Unit Service Factor N/A N/A N/A

20. thit Availability Tactor N/A N/A N/A

~

21. Unit Capacity Factor (Using MDC Net) N/A N/A N/A

22. thit Capacity Factor (Using DER Net) N/A N/A N/A

23. thit Forced Outage Rate N/A N/A N/A 24.

Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each): Scheduled for tests.

mid-September, 1978, 10 days. Rofueling, January, 1979, 2 months.

25. If Shut Down at End of Report Period Estimated Date of Startups Early August

26. Units In Test Status (Prior to Cornercial Cperation): Toreenst Achieved INITIAL CRITICALITY 740201 740131 INITIAL r.LtCIKICIU 7612, 761211

, CDP:tENCIAL Ort 1\ TION

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' WIT Sh 00WNS AND POWER REDUcTONS .

DoCast u0 50-267 UNIT uAnt Fort St. Vrain, Unit No.1 nATE 780801

. . CON LETED af H. W. Hillyard, Jr.

REPORT HONTH July, 1978 .

m TEt.EPuour (303) 785-2253 l

HETHOD OF ShuTTIxc DOWN SYSTEM CON oWENT No. DATE TYPE - DURATIOM s REASON REACTOR LER # CODE CODE cAUSE AND CORRECTIVE Ac" .4 TO PREVENT RECURRENCE -

78-19 780629 F 83.8 A 3 N/A N/A N/A Reactar scram and turbine trip due to loss of instrument bus.

78-20 780704 F 1.1 'M 4 N/A N/A . N/A Turbine-generator trip due to electrica:

. noise spike. Reactor not shutdown.

78-21 780714 F 60.9 A 2 N/A N/A N/A 'C' circulator tripped when 'B' logic bus was powered up.' 'D' circulator was

~

not running. The loop shutdown caused

. a feedwater upset. The turbine gener-ator was manually tripped, and the reactor manually scrammed.

78-22 780726 S N/A H 4 78-18/03-L-0 CB XXXXXX Power reduction necessary due to high oxidant concentration. Cause of

+ -

, oxidant problem currently under investi" .

g'a tion .

78-22 780731 F' 16.3 A 2 78-28/01-T-0 ED BATTRY ' Fault in 480V transformer lA resulted

. in loss in instrument bus 3. Turbine generator automatically tripped.

Reactor was manually scrammed.

SUN!ARY: Following recovery from shutdown 78-23, plan to continue operation for electric production until release to I

continue power ascension testing above 70% received from Nuclear Regulatory Commission.

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1

AVEl: ACE DAILY UNIT POWER LEVEL Docket No. '50-267 Unit Fort St. Vrain, Unit No.1

'Date 780801

'.. Completed By H. W. Hillyard, Jr.

Telephone (303) 785-2253

. Month July, 1978 .

DAY AVERACE DAILY POWER LEVEL- DAY AVERAGE DAILY POWER LEVEL (We-Ne t) .

(Mie-Ne t)

{ . 1 0 . 17 45 2 0' 18 162 3 0 19 164 4 18 20 171

• i . 5 55 21 182

)

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6 '90 [ 22 186 7 133 23 185 8 154 24 184 9 154 25 184 30 159 -

2'6 1 173 6

11 164 .

27 152 15 165 28 163 -

13 163 29 125 .

14 136 . .

30 123

, 15 0 31 35 16 0 l

• Cencrator on line but no net generation, ,

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REF'JEI.ING INFORMATION

1. Name of Facility. Fort St. Vrain, Unit No.1 .

2. Scheduled date for next refueling .

shutdown. January 1, 1979 3, Scheduled date for restart ~

following refueling. March 1, 1979 i . 4. Will refueling or resumption of '

operation thereafter require a

, technical specification change Yes*

1 or other license amende:ent?

. If answer is yes, what, in To facilitate insertion of eight fuel test general, will these be? elements.

If answer is no,'has the reload

• NOTE: If Technical Specification change fuel design and core configura- approval is not received from NRC in time tion been reviewed by your Plant for refueling, then the answer to. #4 is NO Safety Review Committee to deter- and the reload fuel and graphite design 'have j mine whether any unreviewed been reviewed.

. safety questions are. associated *

' with the core reload (Reference * .

10CFR Section 50.59)?

If no such review has taken '

place, when is it scheduled?

5. Scheduled date(s) for submitting

, proposed licensing action and supportine: information. January 9, 1978

6. Important licensing considera- Eight test fuel elements to allow; i ,

tions associated with refueling, e.g., new or different fuel de- 1) Different fuel particle design.

sign or supplier, unreviewed '

design or performance analysis 2) To qualiff near isotropic graphite.

methods, significant changes in

• Tuel design, new operating pro-j .

i ,. cedures. ..

7. The number of fuel assemblics a) 1482 HTCR elements.

I (a) in the core and (b)' in j

the spent fuel storage pool. b) 0 in fuel storage wells.

8. The present licensed spent fuel Capacity is limited by size only to about pool storage capacity and the one third of core (approximately 500 HTCR size of any increase in licensed elements). No change is' planned.

storage capacity that has been requested or is planned, in number of fuel assemblics.

9

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'Y=Y ,

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REFUELIUC INFOIU!ATION (CONTINUED) .

i

9. Le projected date of the 1986 under the nree Party Agreement (Con-

• 1ast refueling that can be tract AT (04-3)-633) between DOE, Public discharged to the spent fuel Service Company of Colorado (PSCO), and pool assuming the present' General Atomic Company.*

licensed capacity.

• The' 1986 date is based on the understanding that spent fuel discharged during the term of the Three Party Agreement will be shipped to the Idaho National Engineering Laboratory for storage by DOE at the Idaho Chemical Processing Plant (ICPP) . The storage capacity has evidently been sized to accomodate fuel which is expected to be discharged during the eight year period covered by the Three Party Agreement. .

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