ML20247H211
| ML20247H211 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 08/31/1989 |
| From: | Block M, Fuller C PUBLIC SERVICE CO. OF COLORADO |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| P-89364, NUDOCS 8909190245 | |
| Download: ML20247H211 (17) | |
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16805 WCR 19 1/2, Platteville, Colorado 80651 September 15, 1989 Fort St. Vrain Unit No. 1 P-89364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.
20555 Docket No. 50-267 SUPJECT:
AUGUST 1989 MONTHLY OPERATIONS REPORT
REFERENCE:
Facility Operating License Number DPR-34
Dear Sir:
Enclosed, please find the Monthly Operations Report for the month of August 1989, submitted per the requirements of Fort St.
Vrain Technical Specification AC 7.5.1.
Nuclear Operations at Fort St. Vrain were terminated August 29, 1989.
If you have any questions, please contact Mr. M. H. Holmes at (303) 480-6960.
. Sincerely, 25 C. H. Fuller Manager, Nuclear Production CHF/sih Enclosure cc:
Regional Administrator, Region IV ATTN:
Mr. T. F. Westerman, Chief Projects Section B
[c2Y Mr. Robert Farrell Senior Resident Inspector
,f l Fort St. Vrain 8909190245 890831 I
PDR ADOCK 05000267
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PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION MONTHLY OPERATIONS REPORT NO. 187 l
August, 1989 l
l
This report contains the highlights of the Fort St. Vrain, Unit No. ], activities operated under the provisions of the Nuclear L
Regulatory Commission Operating License Nd. 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 August,.1989.
1.0 NARRATIVE SUPEARY OF OPERATING EXPERIENCE AND 'MAJ0P SAFETY RELATEU MAINTENANCE On August 1,1989, reactor power was reduced fce RT-500 testing.
Testing was performed through a core differential pressure of 4.70 psid.
RT-500 is a period of verification that the core
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fluctuation problems encountered in the late 1970's will not recur.
On-August 2,
- 1989,
'D' circulator tripped during work on the circulator instruments. Reactor power was reduced to recover the circulator and return to power comenced.
1 During the month of August, the new Security Diesel was successfully test run.
l-On August 11, 1989, FSC determined that 12 cable trays in the i
Auxiliary Equipment Room had been filled to greater than their design criteria.' This condition constituted operation beyond the plant's design criterla and was reported to the Nuclear Regulatory Comission in Licensee Event Report 89-013.
On August 18, 1989, PSC discovered that 4 fire barrier penetration seals in Building 10 had been improperly, installed and were inoperable.
Two additional seals were subsequently discovered and this condition is being reported to the Nuclear Regulatory Commission in Licensee Event Report 89-014.
On August 18, 1989, the ccetrol rod pair in region 39 wes tcund to be incapable of insertion during weekly partial scram testing.
The reactor was shut down per interim Technical Specification requirements. The immcycble control rod condition was caused by the failed head of a clevis pin that had become lodged between an absorber canister and its guide tube.
The root cause of the clevis pin failure is under investigation. This event is being reported to the Nuclear Regulatory Commission in Licensee Event Report 89-015.
On August 18, 1989, during the reactor shut down sequence for the Region 19 control rods, 'C' and 'D' circulators tripped on loss
- of bearing water, which resulted from low water level in the Loop II bearing water surge tank, causing a Loop II automatic shut cown.
This event is being reported to the Nuclear Regulatory Commission in Licensee Event Report 89-016.
On August 18, 1989, after the plant shut oown was completed, a reactor scram was received from the inadvertent actuation of Wide Range Channels I?I and V.
These actuations were caused by electronic noise. The scram is being reported to the Nuclear Regulatory Commission in Licensec Event Report 89-017.
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On August 25, 1989, a leak was discovered in one of the Loop I steam generator module main steam ring headers.
Further L
investigations of the other eleven steam generator. modules revealed 37 more cracks-of various sizes in two modules.
The nature and cause of the cracks are under, metallurgical investigation. This ccedition is being reported to the-Nuclear Regulatory Commission in Licensee Event Report 89-018.
The decision to terminate nuclear operations at Fort St. Vrain was announced August 29, 1989. The repair effort associated with the cracks'in the stean generator module main steam ring headers and the probability of further clevis pin problems associated with the control rods, were considered too extensive to justify.
continued operation of the plant.
2.0 SINGLE RELEASES _0F RADI0 ACTIVITY OR RADIATION EXPOSURE IN EXCESS
- 0F 10% OF THE ALLOVA'BI'E' ANNUAL VALUE None 3.0 INDICATION OF FAILED FUEL RESULTING FROM IRRADIATED FUEL EXAMINATION 5 None 4.0 MONTHLY OPERATING DATA REPORT Attached 5.0 CONTROL R0D DRIVE PART,IAL SCRAM TEST RESULTS AND MAXIMOM DAILY itMPERATURE REPORT Not Recuired this month.
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l 3.0 INDICATION OF FAILED FUEL RESULTING FROM IRRADIATED FUEL EXAMINATIONS None 4.0 MONTHLY OPERATING DATA REPORT
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Attached l
5.0 CONTROL ROD DRIVE PARTIAL SCRAM TEST RESULTS AND MAXIMUM DAILY TEMPERATURE REPORT Not Required this month.
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AVERAGE DAILY UNIT POWER LEVEL Dock.st No. 50-267 Unit Fort St. Wain Unit No. 1 Date September 15, 1989 Completed By M. L. Block Telephone (303) 620-1180 Month AUGUST DAY AVERAGE DAli.Y POWER. LEVEL DAY AVERAGE DAILY POWER LEVEL l
(MWe-Net)
(MWe-Net) l l.
1 228.0 17 249.7 2
168.0 18 60.3
.3 234.4 19 0.0 **
4 244.8 20 0.0 **
5 245.6 21 0.0 **
6 239.2 22 0.0 **
7-247.1 23 0.0 **
8 245.0 24 0.0 **
9 241.3' 25 0.0 **
10 250.1 26 0.0 **
11 245.9 27-0.0 **
12 242.6 28 0.0 **
13 243.7 29 0.0 **
14 244.2 30 0.0 **
15 244.2 31 0.0 **
16 245.1 1
- Generator on line but no net generation.
- Nuclear Operations at Fort St. Vrain were terminated August 29, 1989.
et l
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OPERATING DATA REPORT DOCKET NO.
50-267 DATE Seotenber 15. 1989 COMPLETED BY M. L. Block TELEPHONE (303) 620-1180 OPERATING STATUS NOTES 1.
Unit Name:
Fort St. Vrain, Unit No. 1 2.
Reporting Per!od:
890801 thrcugh 890831
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3.
Licensed Thermal Power (MWt):
842 4.
Nameplate Rating (Gross MWe):
342 5.
Design Electrical Rating (Net MWe):
330 6.
Maximum Dependable Capacity (Gross MWe):
342 4
7.
Maximum Dependable Capacity (Net MWe):
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 MWe):
270.6 (82%)
- 10. Reasons For Restrictions, if Any:
Reanalysis of safe shutdown cooling followinct a 90 minute interruption of forced cooling.
This Month Year To Date Cumulative
- 11. Hours In Reporting Period 744.0 5.831.0 89.160.0
- 12. Number Of Hours Reactor Was Critical 423.9 3,331.9 40,576.7
- 13. Reactor Reserve Shutdown Hours 0.0 0.0 0.0
- 14. Hours Generator On-Line 418.5 2,704.6 27.777.4
- 15. Unit Reserve Shutdown Hours 0.0 0.0 0.0
- 16. Gross Thermal Energy Generated (MWH) 272.875.1 1,562,941.1 14.450.651.2
- 17. Gross Electrical Energy Generated (MWH) 103.966.0 576,430.0 4.836.834.0
- 18. Net Electrical Energy Generated (MWH) 97,239.0 531,819.0 4,281.813.0
- 19. Unit Service Factor 56.3 46.4 31.2
- 20. Unit Availability Factor 56.3 46.4 31.2
- 21. Unit Capacity Factor (Using MDC Net) 39.6 (48.3)*
27.6 (33.7)*
14.6 (17.8)*
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- 22. Unit Capacity Factor (Using DER Net) 39.6 (48.3)*
27.6 (33.71*
14.6 (17.8)*
- 23. Unit Forced Outage Rate 43.8 52.9 63.2
- 24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):
Nuclear Operations at Ibrt St. Vrain were terminated August 29, 1989.
- 25. If Shut Down At End Of Report Period, Estimated Date Of Startup:
N/A Forecast Achieved
- 26. Units In Test Status (Prior To Commercial Operation):
INITIAL CRITICALITY N/A N/A INITIAL ELECTRICITY N/A N/A COMMERCIAL OPERATION N/A N/A
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- Figures in parenthesis represent Net Capacity Factor based on 82%/270.6 me.
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REFUELING INFORMATION l
l l
l 1.
Name of Facility l Fort St. Vrain Unit No. 1 l
l l
1 l 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
-l l
l l 4.
Will refueling or resumption ofl N/A l
l operation thereafter require a l l
l technical specification change l l
l or other license amendment?
l l
l.
l l
l l-If answer is yes, what, in l ----------------
l l
general, will these be?
l l
l l
l l
l If answer is no, has the reload l l
l fuel design and core configura-1 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
l 50.59)?-
l l
l l
l l
If no such review has taken l N/A l
l place, when is it scheduled?
l-l l
l l
l 5.
Scheduled date(s) for submit-l l
l ting proposed licensing action l ----------------
l l
and supporting information.
l l
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l l 6.
Important licensing considera-l l
l tions associated with refuel-l l
l ing, e.g., new or different l
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fuel design or supplier, unre-l ----------------
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viewed design or performance l
l l
analysis methods, significant l l
l changes in fuel design, new l
l l
operating procedures.
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-l l
l 7.
The number of fuel assemblies 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 spent fuel elements l
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-_.___.___._,_-___.__m___.____
m____________
f.
7 REFUELING INFORMATION (CONTINUED) 1:
l.
l l
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 rc: quested or is l (approximately 500 HTGR elements).l l
planned, in number of fuel l No change is planned.
l l
assemblies.
l l
1-l l
l 9.
The projected date of'the last l No further refueling is l
L l
refueling that can be dis-l anticipated at Fort St. Vrain.
- l l
charged to the spent fuel pool l
- l l
assuming the present licerded l l
l capacity.
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Under Agreements AT(04-3)-633 and DE-SC07-791001370 between Public Service Company of Colorado, General Atomic Company, and DOE, spent fuel discharged during the defueling process will be stored by DOE at the Idaho Chemical Processing Plant.
The storage capacity i s presently sized to accommodate eight fuel segments.
It is estimated that the eighth fuel segment will be discharged in 1992. Discussions concerning the disposition of ninth fuel segment are in progress with DOE. Nuclear Operations at Fort St. Vrain were terminated August 29, 1989.
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l.
CONTROL ROD DRIVE PARTIAL SCRXM TEST RESULTS.'AND MAXIMUM DAILY
. TEMPERATURE REPORT REPORT PERIOD:
August 1, 1989 - August 17, 1989 l
.hh Prepared by:
Bruce Brunsdon System Engineer Systems Engineering
/d.m Approved by:
m.
Steve Jones 7 ~
Preventive Maintenance Supervisor Systeras Engineering Public Service Company of Colorado Fort St. Vrain Unit No. 1 6
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.m_.-....mm-__
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Page 1 l
I.
ABSTRACT L
This report summarizes the partial scram, test results and the l
maximum daily temperature of those control rods with motor l
temperatures above 215 degrees Fahrenheit.
It is prepared to satisfy the Fort St.
Vrain Interim Technical Specification Surveillance Requirement 4.1.1.A.I.a.
The time period covered by this report is August 1, 1989, through August 17, 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 l-230 degrees Fahrenheit, and averaged 215 degrees Fahrenheit.
l A total of 130,000 jog cycles plus 1600 scrams was logged l
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 i
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 at 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 verify that they are less than 250 degrees l
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 CP.DMs 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 />.
These surveillance ensure that CRDM motor temperatures exceeding 215 degrees Fahrenheit do not degrade the CRDM's reliability to perform its design function when required.
III.
SUMMARY
OF RESULTS i
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During the month of August,1989, reactor operations continu d from the beginning of the month through the 17th.
On August 17, 1989, during the weekly CRD operability surveillance SR 4.1.1.B-1/2W, a slack cable indication was received on region 19.
Attempts to free the. jammed rod were unsuccessful, so a controlled reactor shut down was initiated, in accordance with LCO 3.1.1.A.
The shut down was completed within twelve hours.
No further surveillance per SR 4.1.1.A.1.a were performed between August 17, 1989, and August 31, 1989.
While the reactor was operating, CRDM shim motor temperatures l
in five reactor regions exceeded 215 degrees Fahrenheit.
In accordance with Technical Specification 4.1.1.A.1.a.
the maximum daily temperature exceeding 215 degrees was recorded on a daily basis. These temperatures are reported in Table 1.
Furthermore, a partial scram test was performed on the highest temperature region once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Results from these tests are shown in Table 2.
Note that all CRDM shim motors remained below the limiting condition of operation temperature of 250 degrees Fahrenheit for the entire reporting period.
The control rod drive performance parameters calculated from the back-EMF voltage data include actual scram
- time, extrapolated scram time, and starting acceleration.
The extrapolated scram time is an estimate of the amount of time it would take a particular control rod pair to insert from the fully withdrawn position.
It is calculated based of the initial acceleration and steady state velocity of the rods measured during the partial scram test.
This parameter provides a means to verify CRD0A operability against Technical Specification performance criteria without conducting a full stroke scram test, and also provides a basis to compare data from partially withdrawn rods to data from fully withdrawn rods.
The starting acceleration parameter is a measure of the free rotation of the CRDM to a steady state velocity under a known applied torque.
The difference between the actual acceleration and the theoretical maximum acceleration serves as an indication of the mechanical drag present in the drive mechanism. The amount of torque available to initiate a scram varies as a function of control rod withdrawl
- height, affecting the theoretical maximum acceleration.
For scram tests performed on partially withdrawn rods, the projected starting acceleration is the theoretical maximum acceleration the mechanism could experience with the available starting torque.
For fully withdrawn rods, 'the projected starting acceleration is the actual. starting acceleration.
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.' CONCLUSIONS The ' data, indicates' that the performance of control rodLdrive,
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serial. number.5 did not degrade-during the reporting period.
The projected ' scram. time calculated varied by no more than a
. tenth of a ;second for! the entire month, and only minor i?
variations were seen in the starting acceleration. 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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C.D5unE PLATE f
(CONTACT UNDERSIDE) l Q
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MOTOn PLATE l
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m Figure 1 Control Rod Drive Mechanism Temperature Sensor Locations
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l TABLE 1.
MAXIMUM DAILY Q,Rpji M MOTOR TEMPERATURES REG. 09 REG. 10 REG..25 REG. 26 REG. 30 DATE CRDM 18 CRDM 36 CRDM'21 CRDM 29 CRDM 05 l
8/1/89 231.0 226.9-L 8/2/89 220.9 235.7 218.6 216.8 238.2 l,
8/3/89, 215.5 234.7 219.6 241.5 l
8/4/89 234.5 219.9 242.0 8/5/89 234.0 219.6 241.3 8/6/89 231.8 218.0 239.2 8/7/89 231.4-217.9 238.8 8/8/89 231.3 217.7 238.5 l
8/9/89 231.9 218.3 239.3 8/10/89 231.8 218.1 239.2 l
l 8/11/89 232.5 218.8 215.0 239.8 l'
'8/12/89-232.2 218.7 215.0 239.5 8/13/89 231.8 218.4 239.5 8/14/89 231.9 219.6 215.5 239.3 8/15/89 232.3 220.1 215.9 240.0
'8/16/89 232.1 220.3 215.9 239.7 8/17/89 232.2 220.3 216.0 239.3 8/18/89 230.0 218.5 236.7 8/19/89 8/20/89 8/21/89 8/22/89 8/23/89 8/24/89 8/25/89 8/26/89 l
8/27/89 8/28/89 8/29/89 i
8/30/89 8/31/89 NOTE:
All temperatures reported are in degrees Fahrenheit.
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TABLE 2.
CRDM PERFORMANCE FARAMETERS FROM DAILY FARTIAL SCRAM TESTS SCRAM TIME ACCELE?.ATION (seconds).
-(radians /sec/sec)
DATE REGICN EE2M TEST PROJECTED ACTUAL PROJECTED 8/1/89 12 36 10.0 145.4 115.9 8/2/39 30 05 10.0 135.1 115.2 8/3/89-30 05 10.0 135.2 116.5 8/4/89 30 05.
10.0 135.2 116.0 S/5/89 30 05 10.0 135.2 115.3 N
-8/6/29 30 05 10.0 135.2 115.5 S/7/39 30 05 10.0 135.2 115.6 8/8/39 30 05 10.0 135.2 116.3 8/9/39 30 05 10.0 135.2 116.0 8/10/89 30 05 10.0 135.1 116.7 8/11/89 30 05 10.0 135.2 116.7 8/12/89~
30 05 10.0 135.2 115.9 3/12/89 30 05 10.0 135.2 115.3 8/14/39.
30 05 10.0 132.2 115.5 8/15.'59 30 05 10.0-135.2 116.4 8/16/39 30 05 10.0 135.1 116.4 8/17/89 30 05 10.0 135.2 115.6 Notes:
No projected starting acceleration values are reported because all scram tests were performed frc= the fully withdrawn position.
No nigh shim motor temperature surveillance were perf:rmed af ter 8/*7/89, due to a reactor shutdown.
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