ML20247A896

Monthly Operating Rept for Apr 1989 for Fort St Vrain Nuclear Generating Station

ML20247A896
Person / Time
Site:	Fort Saint Vrain
Issue date:	04/30/1989
From:	Block M, Fuller C PUBLIC SERVICE CO. OF COLORADO
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
P-89185, NUDOCS 8905230315
Download: ML20247A896 (15)

Text

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PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION MONTHLY OPERATIONS REPORT NO. 183 April, 1989 i

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8905230315 890430 /f gDR ADOCK0500g7

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s This report contains the highlights of. the Fort St. Vrain, Unit- .

No. 1, activities' operated under the provisions of the Nuclear Regulatory - Commission Operating License No. .OPR-34. This report includes.the monthly partial scram / maximum temperature reports for-control rod drive and orificing assemblies. This report is for the month of April,1989.

1.0 NARRATIVE-

SUMMARY

OF OPERATING ~ EXPERIENCE AND MAJOR SAFETY RELATED MAINTENANCE The reactor remained critical at low power for the first part of the month for'the . continuation of moisture' removal. Reactor power was increased as-dewpoint continued its downward trend.

The Turbine Generator was initially placed on line at 0032 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />, April . 9, 1989. Due to adjustments needed to the Electro Hydraulic Control (EHC) System, the turbine generator was taken off-line and was.again placed on line at 2045 hours0.0237 days <br />0.568 hours <br />0.00338 weeks <br />7.781225e-4 months <br /> on April 9, 1989.

On April 19,.1989, reactor power was reduced to comply with FPOR-16 due to the 'ACM' and 'B' diesel generators being declared inoperable. The diesel generator units are in opposite Safe Shutdown Cooling Trains. The 'B' diesel generator set was returned to service on April 20, 1989, alleviating the need to continue the power reduction.

On April 21, 1989, a Loop I shutoewn occurred white taking the

'C' battery charger out of service. During this process the 'C' essential instrument bus became deenergized and the turbine generator tripped from approximately 157 MW at 0623 hours0.00721 days <br />0.173 hours <br />0.00103 weeks <br />2.370515e-4 months <br />. Both Loop I helium circulators tripped on low programmed speed, causing t loop shutdown. This event is being reported to the Nuclear Regulatory Commission in Licensee Event Report 89-007.

The reactor remained at low power for training startups on April 23 and 24, 1989. Plant return-to power was begun on April 26, 1989.

On April 27, 1989, PSC determined that the trip setpoint for RT-4801, the backup reactor building ventilation exhaust stack particulate monitor, had been in error since October 1986, due to a calculational error. RT-4801 had been relied upon in lieu of the primary particulate monitor on two separate occasions during that time. No ' activity rehases in excess of allowable values resulted from this incorrect trip setpoint. This condition is being reported to the Nuclear Regulatory Commission in Licensee Event Report 89-008.

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.0n April 27, 1989, the control rod pair in Region 3 failed to scram during the weekly partial scram surveillance test. The Turbine Generator had been placed on line at 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> April 27, 1989, but was manually tripped at 0220 hours0.00255 days <br />0.0611 hours <br />3.637566e-4 weeks <br />8.371e-5 months <br /> April 28, 1989. An orderly manual reactor shutdown commenced at 0136 hours0.00157 days <br />0.0378 hours <br />2.248677e-4 weeks <br />5.1748e-5 months <br /> April 28, 1989. The Region 3 control rod pair was driven into its fully inserted position. Shutdown was completed at 0654 hours0.00757 days <br />0.182 hours <br />0.00108 weeks <br />2.48847e-4 months <br /> April 28, 1989. This event is being reported to the Nuclear Regulatory Commission in Licensee Event Report 89-009.

Replacement of the Regions 3 and 7 control rods is being planned and is scheduled for completion in early May,1989. The control rod drive in region 7 has been running hot and is being replaced-to. avoid problems in the future.

Reactor criticality is expected during the first week of May, 1989.

2.0 SINGLE RELEASES OF RADI0 ACTIVITY OR RADIATION EXPOSURE IN EXCESS OF 10% OF THE ALLOWABLE ANNUAL VALUE None

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3.0 INDICATION- 0F. FAILED FUEL RESULTING FROM IRRADIATED FUEL EXAMINATIONS None 4.0 MONTHLY OPERATING DATA REPORT Attached.

5.0 CONTROL ROD DRIVE - PARTIAL SCRAM TEST RESULTS AND MAXIMUM DAILY TEMPERATURE REPORT Attached

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b OPERATING DATA REPORT DOCKET NO. 50-267 DATE -

May 15, 1989 I$_ COMPLETED BY M.L. Block'

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TELEPHONE. -(303)620-1180 l

.. OPERATI'NG STATUS .l -

NOTES

- 1. Unit Name: Fort St. Vrain Unit No. 1 0 '2. Reporting Period: 890401 through 890430 '

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, 3. Licensed Thermal Power (MWt): 842

. '4 . Nameplate Rating (Gross NWe): 342

5. Design Electrical Rating.(Net MWe): 330

6. Maximum Dependable Capacity (Gross MWe): 342

7. Maximum Dependable Capacity (Net' MWe): 330

8. If Changes Occur In Capacity Ratings -(Items Number 3 Througn 7) Since Last Report,' Give Reasons

- None

9. Power Level To Which Restricted, If Any (het MWe): 270.6 (82%)

-10. Reasons For Restrictions, If Any: Reannalysis of safe shutdown cooling following a 90 minute interruption of forced cooling.

This Month Year To Date Cumulative-

11. Hours In Reporting Period . 719.0 2,879.0 86,208.0-

12. Number Of. Hours Reactor Was Critical 627.1 819.0 38,064.7

13. Reactor Reserve Shutdown Hours 0.0 0.0 0.0

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l'4 . Hours Generator On-Line 292.0 292.0 ,

-25,364.8

15. Unit Reserve Shutdown Hours 0.0 0.0 0.0

16. Gross Thermal Energy Generated (MWH) 110,'28!.2 135,403.5 13,003,113.6

17. Gross Electrical Energy Generated (MWH) 28,513.0 28,513.0 4.288,917.0

18. het Electrical Energy Generated (MWH) 24,691.0 15,219.0 3,765,213.0

-19. Unit Service Factor 40.6 10.1 29.4'

.20. Unit Availability Factor 40.6 10.1 29.4

21. Unit Capacity factor (Using MDC Net) 10.4 1.6 13.2

22. Unit Capacity Factor (Using DER Net) 10.4 1.6 13.2

23. -Unit Forced Outage Rate 23.9 87.6 64.8

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

None 25, If Shut Down At End Of Report Period, Estimated Date Of Startup: May 5, 1989

26. Units In Test Status (Prior To Commercial Operation): Forecast Achieved INITIAL CRITICALITY N/A N/A INITIAL ELECTRICITY N/A N/A COMMERCIAL OPERATION N/A N/A L

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AVERAGE DAILY UN1T POUER LEVEL i

.l Docket No. 50-267 )

Unit Fort St. Vrain Unit No. 1  !

Date May 15, 1989 Completed By M. L. Block ] '

Telephone (303) 620-1180 l

Month APRIL j

DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 0.0 17 116.6 2 0.0 18 131.8 3 0.0 19 145.8 4 0.0 20 108.1 5 0.0 21 21.0 6 0.0 22 0.0 7 0.0 23 0.0 8 0.0 24 0.0 9 36.9 25 0.0 10 55.9 26 0.0 11 70.0 27 0.0*

i 12 72.8 28 0.0 I? 74.4 29 0.0 14 74.4 30 .__ 0.0 15 89.1 31 N/A 16 103.8 l

• Generator on line but no net generation.

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l' 4 REFUELING INFORMATION' l .

l l l 1. Name of Facility l Fort St. Vrain Unit No. 1 l E

l. l l l c 2. Scheduled date for next l None, no further refueling at l l refueling shutdown. I FSV is expected. l l- l l l 3. Scheduled date for restart l N/A l l following refueling. l l; I l l j 4. Will refueling or resumption ofl N/A l-

,] operation thereafter require a l l l technical specification change l l l- or other license amendment? I l l .

l- l l If answer is yes, what, in l ---------------- l l general, will these be? l l l 1 l l If answer is no,' has the reloadj [

l fuel design and core configura-l l l tion been' reviewed by your l .l

-l . Plant Safety Review Committee l N/A 'l l to determine whether any unre- l l l viewed safety questions are l l

l. associated with the core reload l l l (Reference.10'CFR Section l :l l' 50.~59)? l l l .

I l-l If no such review has taken l N/A l i l place, when is it scheduled?. l _ l 1 I I l 5. Scheduled date(s) for submit- l l l ting proposed licensing action l ---------------- l l and supporting information. l l l l l l l 6. Important licensP , .onsidera- l l l tions associated with refuel- l l l ing, e.g. , new or different l l 1 fuel design or supplier, unre- l ---------------- l l viewed design or performance l l )

l analysis methods, significant l l 4

l changes in fuel design, new l l l operating procedures. l l l l l l - 7. The number of fuel assemblies l l l l (a) in the core and'(b) in the l a) 1482 HTGR fuel elements l l spent fuel storage pool. I b) 0 soent fuel elements l i

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REFUELING INFORMATION (CONTINUED)

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l l 1 l 8. The present licensed spent fuell l l pool storage capacity and the l l l size of any increase in l Capacity is limited in size to l l licensed storage capacity that l about one-third of core l l has been requested or is ] (approximately 500 HTGR elements).1 l planned, in number of fuel l No change is planned. l l assemblies. l l l l .

l l 9. The projected date of the last l 1993 under Agreements AT(04-3)-633]

l l refueling that can be dis- l and DE-SC07-791001370 between l l charged to the spent fuel pool l Public Service Company of l l assuming the present licensed l Colorado, and General Atomic l l capacity. 1 Company, and DOE.* l No further refueling is anticipated at Fort St. Vrain. Spent fuel discharged during the defueling process will be stored by DOE at the Idaho Chemical Processing Plant. The storage capacity is presently sized to accommodate eight fuel segments. It is estimated that the eighth fuel segment will be discharged in 1993.

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CONTROL ROD DRIVE PARTIAL SCRAM TEST RESULTS AND MAXIMUM DAILY TEMPERATURE REPORT REPORT PERIOD:

April 10,1989 - April 20,1989 Prepared by: M Rob Gappa e'"

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System Engineer Systems Engineering Approved by: x .O _ f. E

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Steve Jones Preventive intenance Supervisor Systems Engineering Public Service Company of Colorado Fort St. Vrain Unit No. 1

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Page 1 I. ABSTRACT This' report summarizes the partial scram' test resu'lts and the maximum daily temperature of those control rods ~ with motor.

temperatures 'above 215 degrees Fahrenheit.' It is prepared to {

1 satisfy the: Fort St. Vrain Interim' Technical Specification

~ Surveillance Requi rement .4.1.1. A.1.a. The time period covered.

byfthis report is April 10, 1989, through April 20, 1989.

II. BACKGROUND Because- proper operation of the Control Rod Drive Mechanism

'(CRDM) is critical- for safe operation of the reactor, a series -

of qualification tests .were conducted over a'one year time-span to demonstrate its capability. 'The motor temperature during these tests varied between 200 degrees Fahrenheit and 230 degrees Fahrenheit, and averaged 215 degrees Fahrenheit.

A total -of 130,000 jog cycles plus 1600 scrams was logged during the final design. testing, which is -many times that expected; over the normal life of a. CRDM. The operating temperature of the CRDM is limited by the motor's: Class H insulation which is de-rated to 272 degrees Fahrenheit to account for motor temperature rise, frictional torque increase, and winding life expectancy.

In' order to monitor CRDM temperatures, Resistance Temperature Devices (RTD's) are mounted on the closure plate, orifice valve motor plate, and CRDM motor as shown on Figure 1. All CRDM's installed in the Reactor are equipped with RTD's.

Three recorders located in the control room record each of these temperatures for all 37 CRDM's. .CRDM motor temperatures are - alarmed a t' 212 degrees Fahrenheit and 247 degrees Fahrenheit.

CRDM motor temperatures are monitored at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to veri fy that they are less than 250 degrees Fahrenheit, which is a limiting condition for operation. If one or more CRDM motor temperature (s) is found to be greater than 215 degrees Fahrenheit during the daily surveillance, the motor temperature of all CRDMs exceeding 215 degrees Fahrenheit is recorded and a partial scram test is performed on the CRDM with the highest motor temperature once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Thesa surveillance ensure that CRDM motor temperatures exceeding 215 degrees Fahrenheit do not degrade the CRDM's rel'iability to perform its design function when required.

l III.

SUMMARY

OF RESULTS Region 7 motor temperature exceeded 215 degrees Fahrenheit from April 10 through April 20, 1989. Regions 12 and 30 CRDM motor temperatures exceeded 215 degrees Fahrenheit from April 16 through April 19, 1989. On April 18 and April 19, 1989, Region 7 exceeded 250 degrees Fahrenheit requiring this CRDM to be declared inoperable.

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Thel. partial scram-test results and maximum daily temperature

.of those CRDMs with motor temperatures "above 215 degrees '

Fahrenheit are presented on Table 1. The pertinent parameters

are

. reactor region involved,' control rod ' position, maximum

' daily ; . motor' temperature, extrapolated scram times (actual a

value and projected value) .and . starting acceleration. The control rod position is presented because of its relationship to available starting torque which affects the starting acceleration and extrapolated scram time. 'The extrapolated scram time is affected by control. rod position because it ' is calculated using distance travelled versus time. Since the acceleration is slower at lower control rod positions, a longer period of time is required to reach steady state speed.

This longer acceleration period :results in a- decrease in distance travelled over a given period of time, thereby indicating a longer extrapolated scram time. The atrapolated scram time calculated by the Back EMF program applies a correction factor to provide a projected full length scram time.

.7 The starting acceleration values presented are calculated by

. the Back EMF program. These values reflect the CRDM's freedom of rotation to' accelerate to steady state speed and therefore are considered to be valuable performance indicators. Two values are presented: The actual acceleration measured during a scram from the indicated control rod position 'and the projected acceleration if the control rods were at the full t

out position (greater than 188 inches). If' the control rod

. position is greater than.188 inches, no data is presented for a projected starting acceleration since the actual and projected are the same. As indicated with scram times, the projected starting acceleration'is more readily trended- since the value should be independent of control rod position.

IV. CONCLUSIONS The performance of CRDMs with motor temperatures in excess of 215 degrees Fahrenheit did not show any significant trends or

. degradation during .this report period. Therefore, it is concluded that CRDM motor temperatures exceeding 215 degrees Fahrenheit did not degrade the CRDM's capability to perform its design function when required. Region 7 CRDM motor temperature exceeded 250 degrees Fahrenheit for two days.

Although this CRDM was declared inoperable, testing continued.

No significant trends or degradation was observed with this CRDM.

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I Figure 1 Control Rod Drive Mechanism Temperature Sensor Locations

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O Pubiic service- ===-- 16805 WCR 19 1/2, Platteville, Colorado 80651 May 15, 1989 Fort St. Vrain Unit No. 1 P-89185 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Docket No. 50-267

SUBJECT:

APRIL 1989 MONTHLY OPERATIONS REPORT

REFERENCE:

Facility Operating License Number DPR-34

Dear Sir:

Enclosed, please find the Monthly Operations Report for the month of April 1989, submitted per the requirements of Fort St. Vrain Technical Specification AC 7.5.1.

If you have any questions, please contact Mr. M. H. Holmes at (303) 480-6960.

Sincerely, A lb C. H. Fuller Manager, Nuclear Production CHF/jm Enclosure cc: Regional Administrator, Region IV ATTN: Mr. T. F. Westerman, Chief Projects Section B Mr.. Robert Farrell Senior Resident Inspector Fort St. Vrain i