Text

Monthly Operating Rept 171 for Apr 1988
	ML20154C274
Person / Time
Site:	Fort Saint Vrain
Issue date:	04/30/1988
From:	Fuller C, Novachek F; PUBLIC SERVICE CO. OF COLORADO
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	P-88168, NUDOCS 8805180103
	Download: ML20154C274 (15)
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3 k MONTHLY OPERATIONS REPORT i

NO. 171 ,

April, 1988 i

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4 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. DPR-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,1988.

1.0 NARRATIVE

SUMMARY

OF OPERATING EXPERIENCE AND MAJOR SAFETY RELATED MAINTENANCE On April 4, 1988, at 1420 hours0.0164 days 0.394 hours 0.00235 weeks 5.4031e-4 months , the turbine generator automatically tripped as a result of a disturbance in the regional electric distribution system. The reactor was manually scrammed at approximately 1421 hours0.0164 days 0.395 hours 0.00235 weeks 5.406905e-4 months as a precautionary measure when "A" and "C" boiler feed pumps began to slow down due to a reduction of steam supply. An Unusual Event was declared at 1515 hours0.0175 days 0.421 hours 0.0025 weeks 5.764575e-4 months when the core support floor vent relief V-6389 lifted, resulting in an unplanned, monitored release. The Unusual Event 1 was terminated at 2100 hours0.0243 days 0.583 hours 0.00347 weeks 7.9905e-4 months when the release was isolated. Due l to the low primary coolant activity levels and vent system flow limitations, the activity released was significantly below Technical Specification limits. This event has been investigated and reported to the Nuclear Regulatory Commission in Licensee Esent Report 88-004.

On April 4, 1988, and again on April 12, 1988, neutron flux rate-of-change high reactor scram actuations occurred due to electrical noise induction in the wide range nuclear

) instrumentation channels. The source of electrical noise on April 4 was a malfunction in Linear Power Channel 7. The source 1

of electrical noise on April 12 was chattering of Electro-Hydraulic Control relay KT-873. These events have been investigated and reported to the Nuclear Regulatory Commission in Licensee Event Report 88-005.

The reactor was taken critical on April 5,1988, at 1336 hours0.0155 days 0.371 hours 0.00221 weeks 5.08348e-4 months .

The turbine generator was returned to service on April 6, at 1205 hours0.0139 days 0.335 hours 0.00199 weeks 4.585025e-4 months .

On April 7, 1988, at 0509 hours0.00589 days 0.141 hours 8.416005e-4 weeks 1.936745e-4 months , the reactor was manually scrammed following trips of 1A and 18 circulating water pumps.

The expansion joint on the discharge of circulating water pump 1A i had ruptured, resulting in the flooding of the circulating water [

j pump pit and suosequent circulating water pump trips. An Unusual

Event was declared at 0650 hours0.00752 days 0.181 hours 0.00107 weeks 2.47325e-4 months when the core support floor vent relief V-6389 lifted, resulting in an unplanned, monitored

, release. The Unusual Event was terminated at 0820 hours0.00949 days 0.228 hours 0.00136 weeks 3.1201e-4 months when the i

release was isolated. The release was calculated to be 5.3% of i

the Technical Specification ELCO 8.1.1(a) limits. This event has

been investigated and reported to the Nuclear Regulatory >

Commission in Licensee Event Report 88-006, a ,

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J On April 9, 1988, it was determined that the Technical Specification frequency limit for performance of Surveillance SR-4.1.8.c.I/2/3-Q (Reserve Shutdown Hopper, ACM Otsconnect and Low Pressure Alarm Test) had been exceeded. The root cause was inadequate verification of the computer program through software controls. This event has been investigated and reported to the ,

Nuclear Regulatory Commission in Licensee Event Report 88-007. l The reactor was taken critical on April 21, 1988, at 2123 hours0.0246 days 0.59 hours 0.00351 weeks 8.078015e-4 months .

The turbine generator was returned to service on April 23, at '

0242 hours0.0028 days 0.0672 hours 4.001323e-4 weeks 9.2081e-5 months .

On April 22, 1988, it was determined that the Technical Specification frequency limit for performance of Surve111ances SR-5.4.1.2.8A-M (Primary Coolant Moisture Low Level Test), and SR-5.4.1.2.88-M (Primary Coolant Mo:i:ure Low Level Channel Test) had been exceeded. This event will be investigated and reported i to the Nuclear Regulatory Commission in Licensee Event Report 88-008.

2.0 SINGLE RELEASES OF RADIOACTIVITY OR RADIATION EXPOSURE IN EXCESS OF 10% OF THE ALLOWABLE ANNUAL VALUE None 3.0 INDICATION OF FAILED FUEL RESULTING FROM IRRADIATED FUEL EXAMINATIONS None 4.0 MONTHLY OPERATING DATA REPORT Attached 5.0 CONTROL R00 DRIVE PARTIAL SCRAM TEST RESULTS AND MAXIMUM OAILY I T~EMPERATURE REPORT Attached l

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OPERATING DATA REPORT 00CKET N0. 50-267 DATE May 13. 1988 COMPLETED BY F. J. Novachek TELEPHONE (303) 620-1007 OPERATING STATUS h0TES I

1. Unit Name: Fort St. Vrain. Unit No. 1

2. Reporting Period: 880401 through 880430

3. Licensed Thermal Power (Fdt): 842 4 Nameplate Ratirig (Gross MWe): 342 ,

5. Design Electrical Rating (Net MW,): __ 330

6. Maximum Dependable Capacity (Gross MWe): 342

7. Maximum Dependable Capacity (Net Kde): 330

8. If Changes Occur In Capacity Ratings (Items Number 3 Through 7) Since Last Report. Give Reasons:

None

9. Power Level To Which Restricted. If Any (Net Kee): 270.6

10. Reascns For Restrictions. If Any: Reanalysis of safe shutdown coolina followino a 90 minute interruotion of forced cooling.

This Month Year To Date Cumulative

11. Hours In Reporting Period 719 2.903 77.448

12. Number Of Hours ReactJr Was Critical 343.6 2.492.7 35.939.4

13. Reactor Reserve Shutdown Hours 0.0 0.0 0.0

14. Hours Generator On-Line 291.8 2.394.1 23.97.9.6

15. Unit Reserve Shutdown Hours 0.0 0.0 0.0

16. Gross Thermal Energy Generated (MWH) 160.677.2 1.259.896.3 12.193.230.9

17. Gross Electrical Energy Generated (PdH) 57.601.0 458.761.0 4.000.981.0

18. Net Electrical Energy Generated (MWH) 5?,36R_0 430.646.0 1.520.912.0

19. Unit Service Factor 40.6 82.5 31.0

20. Unit Availability Factor 40.6 82.5 31.0

21. Unit Capacity Factor (Using MDC Net) ?? 1 45.0 11.R

22. Unit Capacity factor (Usir.g DER Net) 22.1 45.0 13.8

23. Unit Forced Outage Rate 59.4 17.5 6?.5 24 Shutdowns Scheduled Over Next 6 Months (Type. Date, and Duration of Each): Helium Circulator Reoairs. 880705. 2.184 hourJ;

25. If Shut Down At End Of Report Period. Estimated Date Of Startup: N/A

26. Units in Test Status (Prior To Commercial Operation): Forecast Achieved INITIAL CRITICALITY N/A N/A INiiIAL ELECTRICITY N/A N/A C0HMERCIAL OPERATION N/A N/A

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AVERAGE DAILY UNIT POWER LEVEL ,

Docket No. 50-267 +

Unit Fort St. Vrain Unit No. 1 !

Date May 13, 1988 !

Completed By E J. Novachek i Telephone (303) 620-1007 i

Month APRIL DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL. !

(MWe-Net) (MWe-Net) 1 195.4 17 0.0 2 222.5 18 0.0 3 206.5 19 0.0 1 4 130.3 20 0.0 l 3 ,

i 5 0.0 21 0.0 i 1

6 38.0 22 0.0

7 30.3 '

23 48.6 8 0.0 24 73.9 i 9 0.0 25 88.4 10 0.0 26 213.1 i

11 0.0 27 249.3 '

! 12 0.0 28 250.5' !

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13 0.0 29 252.6 f

14 0.0 30 250.0 a [

1 1 15 0.0 31 N/A i 16 0.0 i

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Generator on line but no net generation.

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TSP-3 Attachemeqt TSP-3C lssue 2 UM11_}HMlQQW8l M D POWER REDUCTIONS DOCKET NO. 50-267 UN I T NAME f o rt St. Vra in Unit No. 1 DATE May 13. 1988 COMPLElED BY [Lanis J. NovaCtHrk TELEPHONE (3031 620-1007 REPORT MONTH A PR I L. 1988 i i I I i IMETHOD OFl l 1 l l l l 1 l 1 ISHUTTING l I ( l CAUSE AND CORRECTIVE ACTION I i NO. I DATE ITYPEI DURATION l REASON lDOWN l LER # ISYSTEMICOMPONENTI TO PREVENT RECURRENCE I l l l l l l REACTOR 1 10000 .. . lGQQE I I I I I I I I I I I l l 188-071 880404 i F l 45.7 Hr i H l 2 l 88-004 I ZZ l ZZZZZZ l Turbine Trip And Manuel Scram Due To I 1 l I I i l I l l l Regional Electric Distribution System i I i i l I I I I I I Upset. I I I I I I I I I I I 188-081 880807 l F l 381.5 H r i 4 A I 2 1 88-006 l NN I EXJ l Manual Scram Due To Rupture Of 1 A M.a in l-i l l 1 l l l l l l l Circulating Water Pump Discharge Expansion I I i l l l l l l l l Joint. Replaced The failed Expansion i l l i l l l I l l l Joint. All Similar Expansion J6ints Have 1 I I I I I 1 I I i l Been Or Soon will Be Inspected And i I I I I i 1 i i i i Replaced I f Necesca ry. I I l l l I I I i 1 1 I I I I I I I I I I I I I I I I I I I I i 1 1 I I 1 i I i l I I I I i i i i l i i I i 1 1 I I I i i 1 1 I I I I I I I i I i i i i I i 1 i I i 1 1 1 1 I I I i 1 i l i l i i l I l l I I I I I I I I I i 1 1 I i 1 I I I I i 1 I I I I i 1 1 1 1 I l 1 i I i l I I i l i I i 1 1 1 1 1 I I I i 1 1 1 I I i 1 1 I I i l l 1 1 1 I I I I I I I I I I I I I I I I I I I I I I i 1 1 I I I I I I I I I I 1 1 I I I I I I 1 I i i i I I i I I I I I I I I I i l i i l I i 1 1 I i 1 1 I I I I i 1 I I I 1 I I I I I I I I i 1 I l l l 1 1 I i i i I I l

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l REFUELING INFORMATION l l l l 1. Name of Facility l Fort St. Vrain Unit No. 1 l l l I l 2. Scheduled date for next l l l refueling shutdown. I April 4, 1989 l

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i I l 3. Scheduled date for restart l May 27, 1989 l l following refueling. l l

l l l l l 4. Will refueling or resumption ofl No l '

l operation thereafter require a l l l technical specification change l l l or other license amendment? I

_, 1 ,

I l l l If answer is yes, what, in l ---------------- l l general, will these be? 1 l I l l l If answer is no, has the reloadl l l fuel design and core configura-l l l tion been reviewed by your l l l Plant Safety Review Committee l No 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 I I I l If no such review has taken l 1988 1 l place, when is it scheduled? l l

1 l l l 5. Scheduled date(s) for submit- l l !

l ting proposed licensing action l ---------------- l 1 l and supporting information. I l .

I I I l 6. Important licensing considera- l l l tions associated with refuel- 1 i l ing, e.g., new or different l l i l fuel design or suppliur, unre- l ---------------- l l viewed design or performance l 5

l l analysis methods, significant l l !

l changes in fuel design, new l l l operating procedures. I l

l l l l 7. The number of fuel assemblies l l l (a) in the core and (b) in the I a) 1482 HTGR fuel elements (

l spent fuel storage pool. I b) 0 spent fuel elements l

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d REFUELING INFORMATION (CONTINUED) t I I I ,

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 l (approximately 500 HTGR elements).l l planned, in number of fuel l No change is planned. l l assemblies. 1 l l . I I l 9. The projected date of the last l 1996 under Agreements AT(04-3)-633l 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. I Company, and 00E.* l The 1996 estimated date is based on the understanding that spent fuel discharged during the term of the Agreements will be stored by DOE at the Idaho Chemical Processing plant. The storage capacity has evidently been sized to accommodate eight fuel segments. It is estimated that the eighth fuel segment will be discharged in 1996.

I

i CONTROL R0D ORIVE PARTIAL SCRAM TEST RESULTS AND MAXIMUM DAILY I TEMPERATURE REPORT i

L REPORT PERIOD:

) APRIL 1, 1988 - APRIL 30, 1988 n-1 .

Prepaxed bvi #//, G ATuane L. Spiker /

Senior Technica1 Services Engineering a

Technician r

Approved by: D , u L rQC'adley ~ ( /-4 83 et M Plant Engineering Supervisor Public Service Company of Colorado '

Fort St. Vrain Unit No. 1 i

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Page 1

, 1. ABSTRACT '

This report summarizes the' partial scram test results and the 1 maximum daily temperature of those control rods with motor !

temperatures above 215 degrees Fahrenheit. It is prepared to l satisfy the Fort St. Vrain Interim Technical Specification Surveillance Requirement 4.1.1.A.I.a.

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The time period covered L

by this report is April 1, 1988, through April 30, 1988. ,

II. BACKGROUND l Because proper operation of the Control Rod Drive Mechanism (CRDM) is critical for safe operation of the reactor, a series j of qualification tests were cond7cted 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 ;

1 during the final destgr. testing, which is many times that j expected over the normal life of a CRDM. The operating i temperature of the CRDM is limited by the motor's Class H ;

4 insulation which is de-rated to 272 degrees Fahrenheit to i account for motor temperature rise, frictional torque l increase, and winding life expectancy. '

i In order to monitor CRDM temperatures, Resistance Temperature ;

Devices (RTD's) are mounted on the closure plate, orifice j 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 at 212 degrees Fahrenheit and 247 degrees Fahrenheit.

A CRDM motor temperatures are monitored at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to verify that they are less than 250 degrees Fahrenheit, which is a limiting condition for operation. If one or mere CRDM motor temperature (s) is found to be greater 2

than 235 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 l hours. These surveillances ensure that CRDM motor '

temperatures exceeding 215 degrees Fahrenheit de not degrade the CRDM's reliability to perform its design function when

} required.

4 III.

SUMMARY

OF RESULTS

{

i l During the month of April, 1988, Regions 12 and 30 CRDM motor l temperatures exceeded 215 degrees Fahrenheit in two separate l time intervals, April 1, through April 4; and April 26, through April 30. There were no high CRDM motor temperatures i

' in the intervening period. April 5 through April 25, due to reactor shutdown or operation at low power levels.

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Page 2 H

. The 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 (surveillance value and Back EMF calculated 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 calculated by surveillance. The surveillance 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 extrapo?;ted scram time calculated by the Back EMF program applies a correction factor to provide a projected full length scram time. Therefore, the surveillance scram times can only be compared for trends if the control rod position is approximately the same. The Back EMF calculated scram time is more readily trended since the value should be independent of control rod position.

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 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.

Page 3 Iy, 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.

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Company of Colorado 16805 WCR 19 1/2, Plattefille, Colorado 80651 May 13, 1988 Fort St. Vrain Unit No. 1 P-88168 U. S. Nuclear Regulatory Comission ATTN: Document Control Desk Washington, D.C. 20555 l

Docket No. 50-267

SUBJECT:

APRIL 1988 MONTHLY OPERATIONS REPORT !

REFERENCE:

Facility Operating License No. OPR-34 l l

Dear Sir:

)

Enclosed, please find the Monthly Operations Report for the month of April 1988, 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, C. H, Fuller Manager, Nuclear Production I i

Enclosure j cc: Regional Administrator, Region IV ATTN: Mr. T. F. Westerman, Chief j Projects Section B '

Mr. Robert Farrell Senior Resident Inspector Fort St. Vrain pV CHF:djm /

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ML20154C274

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