ML20052G546
| ML20052G546 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 05/07/1982 |
| From: | Mcbride L PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML20052G456 | List: |
| References | |
| NUDOCS 8205180452 | |
| Download: ML20052G546 (10) | |
Text
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' 1 PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION MONTHi_Y, OPERATIONS REPORT NO. 100 April , 1982 i
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! 8 205180'tTD
i-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 DPR-34. This report is for the month of April, 1982.
1.0 NARRATIVE
SUMMARY
OF OPERATING EXPERIENCE AND MAJOR SAFETY RELATED MAINTENANCE 1.1 Summary The reactor was brought critical five times during the month. It spuriously scrammed four times and was manually scrammed once to leave the reactor shut down at month!s -
end.
At the beginning of the month, reactor power was limited by high moisture and then by Loop 2 steam generator penetration leakage.
On April 12, the modification to maintain steam generator
_ penetration pressure at slightly above cold reheat pressure was put in service (see Amendment 26 to the Technical Specifications).
The turbine generator was placed on line on April 14, but faulty servo assemblies on turbine valves prevented increasing load.
Early in the month, some activity was detected in the water of Loop 1, System 46. Analysis indicated that some of the isotopes present were short-lived, indicating a leak of primary coolant to the system. The leak was temperature dependent and disappeared as reactor power was increased.
It was subsequently determined that due to high moisture levels wnich had been experienced in the primary coolant, it was neces:ary to demonstrate the functionability of the reserve shutdown system (Nuclear Regulato ry Commission requirement). The control rod drive in region 19 was selected for the reserve shutdown functional test due to its apparent reaction to the high primary coolant moisture.
Therefore, on April 20, 1982, reactor power was decreased to remove and replace control rod drive 19. During this power reduction, the System 46 leak was located, and a core support floor tube was isolated by the end of the month.
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1.2 Ooerations At the beginning of the month, the prestressed concrete reactor vessel was being pressurized after having been evacuated to less than 1 psia for moisture removal. Forced cooling was shut down and circulator static seals were set, while the prestressed concrete reactor vessel pressure was below atmospheric pressure, to prevent further ingress of moisture.
Comi support floor cooling tubes were cut in and forced cooling was established by 0350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br /> on April 1.
The reactor was arought critical on April 2, but scrammed about three hours later while performing a surveillance on the plant protective system. It was brought to critical again the same day.
Early on April 3, a continuous air monitor in the Reactor Building alarmed. An investigation revealed leaks from the bellows of the moisture monitor sample return valve in B-2 and B-4 penetrations. The tertiary covers, which had been removed for work on the moisture monitors, were replaced,
. and the tertiary space vented to the gas waste system.
Subsequently, leaks were found in B-3 and B-5 penetrations resulting in four of the six low lovel moisture monitors being inoperable.
On April 5, a noise spike caused one of the high level moisture monitors to trio. A: a result, the reactor scrammed and Loop 2 was shut down and dumped. Loop 2 was recovered, and the reactor was brought to critical again the same day.
On April 7, some low level contamination was discovered in the water of Loop 1, System 46.
On April 10, work was started on the final portions of a design modification which provides capability to control the pressure of Loop 2 steam generator penetration relative to cold reheat pressure and to monitor the penetration.
Work was completed on Easter weekend, and the system was placed in service on April 12.
On April 13, with four of the six low level moisture monitors manually tripped, a high level moisture monitor r spuriously tripped, causing a reactor scram and Loop 2 shutdown and dump. The loop was recovered, and the reactor brought back to critical again the same day.
2 3 While the reactor was still cool, an alarm was received
, indicating high pressure in a top barrel subheader of l .
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Loop 1, System 46. The pressure in Loop 1 was indeed found to be abnormally high. The surge tank was depressurized to normal, and an investigation of the leak began.
On April 14, reactor power was increased to 30%, and the turbine generator was synchroni:ed and load was increased to 70 MWe. The number three control valve went shut while on full arc admission and the intercept valves were very erratic. On April 15, at 1545 hours0.0179 days <br />0.429 hours <br />0.00255 weeks <br />5.878725e-4 months <br />, while investigating these problems, the intercept valves closed, causing the circulators' speed to decrease resulting in a programmed high reactor pressure scram, Loop 2 shutdown, and dump.
The loop was recovered, and the reactor again brought critical early on April 16.
Over the next several days, attempts were made to clear the turbine electro hydraulic control system of problems, but with new servo assemblies not available, the problems were not resclved.
At 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> on April 20, with the plant load at 70 MWe, the cooling tower drift accumulated on the main bus insulators and caused the north bus to trip. The bus was back in service later the same day. The turbine was shut down and reactor power decreased on April 20 to functionally test the reserve shutdown system.
Simultaneously, a further investigation into the source of the primary coolant in-leakage into Loop 1 of System 46 was initiated. The leak appeared to be temperature dependent so reactor power was decreased until the pressure began to increase in the Loop 1, System 46 surge tank. This occurred at about 10%. The location of the leak was later determined to be a tube in the core support floor. High pressure feedwater heater number 6 was also found to be leaking from the tube to shell side. On April 26, while Loop 2 was cleared out for work on the main steam safety valves, Loop 1 circulators tripped, and the reactor was manually scrammed.
On April 28, the prestressed concrete reactor vessel was depressuri:ed, and on April 29 and April 30, the leaking System 46 tube was isolated, the control rod drive in region 19 was changed out, and a reserve shutdown hopper test was performed. The design modification to replace K 2 relay on the control rod drive in region 1 was completed. .
2.0 SINGLE RELEASES OF RADI0 ACTIVITY OR RADIATION EXPOSURE IN EXCESS OF 10% OF THE ALLOWABLE ANNUAL VALUE
!o None '
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3.0 INDICATION OF FAILED FUEL RESULTING FRCM IRRADIATED FUEL EXAMINATIONS There were no indications of failed fuel particles or fuel rods during the report period. However, on April 26, 1982, while performing an inspection of elements that had been removed during the second refueling, a crack on one face of element 1-2415 was observed. Note that the fuel was in no way affected.
The crack extended across the minimum cross-section between a coolant hole and one face of the element (approximately one-half inch), vertically down the full length of the block (31.2 inches), and terminated at the lower exit of the coolant hole.
Engineering / physics calculations are being performed by General Atomic Company wnich will address the conditions specifically experienced by the fuel element. A decision concerning the ultimate disposition of this fuel element will be made after the i various possibilities have been evaluated. (
4.0 MONTHLY OPERATING DATA REPORT Attached l
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- s?-3 Atttchment-33 Issue :
r cP!2AMNC OAn REPORT DOC 3:IT NO. 50-267 1
! :AIE 820507 l l
COMPLE!!D 3T L. M. McBride l
- n,:rsoNE (303) 785-2224 ortRAnNe snTes sons
- t. cait 3 ,,, Fort St. Vrain
- 2. Reporting Period 820401 through 820430
- 3. I.icensed thermal Power () Sic): 842 4 Nameplate Racing (Gross !Sle): 342
- 5. ign nectrical Racing cret se): 330
- 6. Faw=.= Dependable Capacity (Cross !Sie): 342
- 7. F= H == Oependable Cgacity (pet !sie): 330
- 8. If Changes occur in capacity Ratings (Itaas Nummer 3 Through 7) Since Last Report, Give Reasons:
None
?. Power Level To tJhich Restricted. If Any (Net !$ie): 211
- 10. Reasons for Rastrictions If any: NRC rastriction of 70% pendine resolution of temperature fluctuations.
This Stach Year to Date Cu-dative
- 11. Hours in Kacorting Period 3 719.0 2.979.0 2&_gan.n
- 12. Nummer of sours anaccoe was cr.tical 535.3 637.2 15.215.6
- 13. Reactor Raserve Shutdown Hours O.0 0.0 _ 0.0
- 14. R2urs Generator on-l.ine 86.2 86.2 9.994.5
- 13. Unit Reser re Shutdown ifours 0.0 0.0 0.0
- 16. Cross !hermal Energy Generated C*A) 62,491.6 62,491.6 4,996,436.4 17, ccess nectrical Energy Cenerated c51) 5,691 5,691 1.697.047 1s. Net Electrical Eneray cenerated (255) 160 -7,205 1.547.054
- 19. Unit Sernce Factor 12.0 3.0 40.2
- 20. Unic A .11 ability ractor 12.0 3.0 40.2
- 21. Unit Capacity Factor (Using MDC Nat) O.1 0.0 18.9
- 22. Unit Capacity Factor (Cstng OE1 Net) 0.1 0.0 18.9 "J. Cait Forced Outage Rate 38.2 38.2 34.1 :
- 24. Shutdowns Scheduled Over Next 6 Months (Type. Data, and Duration of Each): None
- 23. If Shut Down at End of Raport Period. Estimated Date of Startup: 5/3/82
- 26. Units In Test Stact.J (Prior to Cocenercial Operation): Forecast Achieved IN nAL Cn n :AL T N/A N/A IN:nAL "_ IC RIC::T N/A N/A C0!w1 CIA:. OPERAncs N/A N/A
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TS?-3 Attachraent-3A AVERAGE DAILY CIIT POWER LEVEL Issue 2 Page 1 of 1 Docket No.30-267 Unic Fort St. VrMn Date A?nsn7 Completed By L. M. McBride Telephone (103) 735-2224 Month Anril 1982 i DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL. .
l (MWe-Net) (MWe-Net) 1 0.0 17 50.5 2 0.0 18 23.3 3 _
0.0 19 41.1 4 0.0 20 23.4 5 0.0 21 0 . 0' _
6 0.0 22 0.0 7 0.0 23 0.0 3 0.0 24 0.0 9 0.0 25 0.0 10 0.0 25 0.0 11 0.0 27 0.0 12 0.0 23 0.0 13 0.0 29 0.0 14 0.0 30 0.0 15 18.2 31 N/A 16 0.0
- Generator on line but no net generation.
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lanT ms. 50-267 _ _ _ .
UNIT NAHE Fort St. Vrain DA1E 820507 coeu EETus eY ldAgrMe . _ _
mEruar lumniJ_A ril,,_1982 TELErm>HE _(30__J)_Jgi--222_4 HE1110D OF SituTTING ImlWN SYSTDI COMPONENT No. DATE TYPE 14*ATION REASON REACTOR EER # Q)DE UlDE cAllSE AND CukitECTIVE A(* TION TO PitEVENT StF4:Hitkt.NCE _
81-026 820401 S 331.3 B 2 N/A CBI XXXXXX Loop-split modification.82-001 820415 F 35.3 F 3 N/A IBil INSTRU liigh pressure scrain - PPS.
82--002 820418 F 10.0 11 4 N/A ilBD INSTRU Turbine manually tripped due to electro-hydraulic control system up-set during maintenance. Reactor remained critical.82-003 820418 F 5.0 A 4 N/A ilBD VAINEX Turbine manually tripped for maint-enance. Reactor remained critical.82-004 820419 F 2.9 A 4 N/A IIBD INSTRU Turbine trip due to low hydraulic control pressure. Reactor remained ,
critical. l 82-005 820420 S 251.7 B 2 N/A CJB XXXXXX Manual shutdown to change-out control rod drive (CRD) in region 19 as per request of NRC.
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- T_ad data for saz: refueling shutdova. Ceceber 1. 1983 .
- 3. Scheduled data 'or .astar:
foll= wise sfual1=r. w +.- 1. toa3
. 4 T" :st:ali=g er :ssump:1sn of ,,
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c?aracian .harsaf:ar : squire a
., woe-*'
spaad*d-*: ion change or other 11cause ==p ht? Yes U answer is yes, what, in tensr21, vill dese be? Use of evne H-451 grauhite.
If answer is =m, has the :sicad fuel design and es:s c=nfign n-
, .d_an been :sriswed by ycur ?lan=
l Sadecy Zaviaar Cc:mai::se .s desar-21=a wha: hse any :===sviewed sade:7 quae:1:ss a:s associa:ad ud-% -J:a cars sisad (2afarsnes l' 10C71 See:1:n 50.391? l U =a sucit :sviar has .akan elace. when is ': schedulad?
- 3. Scheduled da:s(s) 'or suimmt.--' --
proposed L1 cans 1=g actica and succor d ? 1 f:r=a:1su. No: scheduled at this time: to be determined.
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- 5. :=oor:an: 11:sns1=g c=nsiders-tions associa:ad w1_h sf=e11=g, e.3., Saw cT Ad ##Ersat fuel da=
sign or suppliar, ungsviewed design or perfor_ance analysis
( nachods, sig=1fican: changss i=
l fuel design, saw opers:1=g pre-l . endurss.
- 7. The number of f a1 assemblias laa2 COR fuel elanen:s (a) in .he cc:s and (*s) in ,,50 spen: s.GR :.ual elenen:s
.%.a seen ,u a_, secrzgo ecol.
- 3. " e prssen:.11cansed spene fuel '
- col s
- crage capac1:7 and de Capac1:7 is id d:ad i si:e to abou: cee-sima of a=7 inc aasa is licensed .ht::i of cars (appr xi=a:aly 500 3.'01 s:orage capac1:7 :ha: has been ele =en:s). :To change is ;t '- ad.
- squestad or is planned, in et=roer of fuel assenbliss.
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R*TUELING INFOUfATION (CONTINL"2D) i
- 9. The projected date of the - 1992 under Agreements AT(04-3)-633 and ~~
last refueling . hat can be '
DE-SC07-791D01370 between Public Service'~~
discharged to the spent fuel Company of Colorado, General Atomic pool assuming the present Company, and DOE.*
7y
- The 1992 estimated date is based en the understanding that spent fuel discharged during the term of the Agreements will be scored by DOE.at the Idaho Chemical Processing Plant. The storage capacity has evidently been sized to accomodate eight fuel segments. It is estimated that the
_ _ . . . _ . _ - . - . __. eighth fuel segment will be discharged in 1992. - ----
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