ML19310A321
| ML19310A321 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 05/30/1980 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML19310A316 | List: |
| References | |
| NUDOCS 8006110084 | |
| Download: ML19310A321 (25) | |
Text
. .- .. . - . . -. .-. .
t
- LO QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 i
MONTHLY PERFORMANCE REPORT l
j MAY 1980 COMMONWEALTH EDISON COMPANY
' ~
AND IOWA-ILLINOIS GAS & ELECTRIC COMPANY l
.NRC DOCKET ~NOS. 50-254 and 50-265 LICENSE NOS. DPR-29 and DPR-30 1
4
}
i 8 0 0 611- 0 @Y f
TABLE OF CONTENTS
- 1. Intreduction
- 11. Summary of Operating Experience A. Unit One B. Unit Two lli. Plant or Procedure Changes, Tests, Experiments, and Safety Related Maintenance A. Amendments to Facility License or Technical Specifications B. Facility or Procedure Changes Requiring NRC Approval C. Tests and Experiments Requiring NRC Approval D. Corrective Maintenance of Safety Related Equipment IV. Licensee Event Reports V. Data Tabulations VI. Unique Reporting Requirements A. Main Steam Relief Valve Operations B. Control Rod Drive Scram Timing Data -
Vll. Refueling information Vill. Glossa ry 4
s a
I i
A e v -
- l
- l. INTRODUCTION Quad-Cities Nuclear Power Station is composed of two Boiling Water Reactors, each with a Maximum Dependable Capacity of 769 Mwe net, located in Cordova, lilinois. The Station is jointly owned by Commonwealth Edison Company and lowa-Illinois Gas & Electric Company. The Nuclear Steam Supply Systems are General Electric Company Boiling Water Reactors. The Architect / Engineer was
~
Sargent & Lundy, Inc. and the primary construction contractor was United Engineers & Constructors. The condenser cooling method is a closed-cycle spray canal, and the Mississippi River is the condenser cooling water source.
The plant is subject to license numbers DPR-29 and DPR-30, issued October 1, 1971, and March 21, 1972 respectively, pursuant to Docket Numbers 50-254 and 50-265 The date of initial reactor criticalities for Units 1 and 2 respec-tively were October 18, 1971 and April 26, 1972. Commercial generation of I power began' on February 18, 1973 for Unit I and March 10, 1973 for Unit 2.
This report was compiled by Becky Brown and Tom Hafera, telephone number 309-654-2241, extensions 245 and 176.
II.
SUMMARY
OF OPERATlHG EXPERIENCE R
A. UNIT ONE May 1-11: Unit One began the reporting period holding load at 703 HWe, and held an average load of 694 MWe. At 2300 on May 10, load was re-i duced at a rate of 200 MWe/ hour to shutdown the unit in order to repair the 18 Recirc Motor-Generator Set field breaker. At 2045 on May 11, the Unit One Main Generator was manually tripped off-line. At 2105 the reactor mode switch was placed in the HOT STANDBY position.
May 12: At 0810, following replacement of the IB Recirc MG Set field breaker trip coII, the reactor mode switch was placed in the RUN position, and at 0900, the generator was placed on-line. At 1425 load was reduced at a rate of 10 MWe/ hour to prepare for a drywell entry. At 1800 the rate of load reduction was increased to 50 MWe/ hour, at 2145 the load reduction was stopped at 210 MWe. At this load, a drywell entry was made to re-place the holding coil on the trip mechanism on the 3B electromatic relief valve. At 2210, after the maintenance was completed, the load was in-creased at a rate of 100 MWe/ hour, at 2400 the rate was changed to 50 MWe/ hour.
May 13: At 0100 the load increase was stopped in order to calibrate the Local Power Range Monitors. At 1000, after the monitors had been calibrated, the load was increased at a rate of 50 MWe/ hour. At 1200 the rate increase was changed to 8 MWe/ hour 6 l May 14: At 0400 the load increase was stopped in order to change con-densate demineralizers. At 0600 the load was increased at a rate of 8 MWe/ hour, until at 1650 a full load of 729 MWe was achieved.
May 15-31: Unit one held an average load of 655 MWe, and finished the 1 reporting period holding a load of 624 MWe. l l
B. UNIT TWO May 1: Unit Two began the reportirg period holding load at 687 MWe. At 1217, while performing the Flow Control Line Determination test, the 2B Feedwater Regulating Valve failed closed, and caused a low water level reactor scram. Af ter all startup procedures were carried out, the 1 reactor was brought critical at 1845 !
l May 2: At 0027 the reactor mode switch was placed in the RUN position, and at 0110, the generator was placed on line. At 0115 the load was increased at a rate of 55 MWe/ hour; at 0625 the rate was changed to 50 MWe/ hour, and at 0730 the rate was changed to 5 MWe/ hour. l 1
( ,
1
May 3: At 1300 the rate of load increase was changed to 8 MWe/ hour. At 2305 the load was held at a constant level in order to perform weekly turbine generator tests and was held by request of the Load Dispatcher.
May 4: _ At 0700 the load was increased at a rate of 8 MWe/ hour. At 1055 the load was held at 725 MWe, and at 1130 the load was reduced at a rate of 100 MWe/ hour due to a high vibration indication on the high pressure turbine. At 1200 the load was held at 670 MWe. At 1625 the load was decreased 50 MWe in 15 minutes due to a high vibration indication on the high pressure turbine. It was discovered that the high vibration in-dication was an incorrect reading from a sensor which had failed. At 0245 on May 5, after the high vibration on the high pressure turbine was confirmed as false, the load was increased at a rate of 5 MWe/ hour.
May 6: At 0310 load was decreased at a rate of 100 MWe/ hour in order to change Condensate Demineralizers. At 0410 load was held at 680 and the change-over was performed. At 0505 the load was increased at a rate of 20 MWe/ hour; at 0600 the rate of increase was changed to 5 MWe/ hour. At 1300 the load increase was stopped in order to move up the limit stops on the Recirc Motor Generator Set scoop tubes. The load increase was resumed at 2000 at a rate of 5 MWe/ hour. At 0130 on May 7 full core flow was achieved and load was held at 800 MWe.
May 7-8: At 2200 on May 7 the Flow Control Line Determination Test began; during this test load was reduced 50 MWe in one minute three times with a two minute analysis period between each drop. At 2210 load was decreased
, at a rate of.50 MWe/ hour to prepare for drywell entry to inspect the oil level on the 2B Recirc pump; the level Indication was reading high. At 0200 the load was held at 400 MWe, the drvwell was entered and the 2B Recirc pump oil level Indicator was repaired. At 0330 the load was in-creased at a rate of 50 MWe/ hour, at 0715 the rate of load increase was
! changed to 5 MWe/ hour. At 1545 the load increase was stopped to backwash and precoat the 2D Condensate Demineralizer. At 2205 the load was in-creased at a rate of 5 MWe/ hour.
May 9-11: At 1610, on May 10, full load was achieved at 812 MWe. At 0600, on May 11, load was decreased at a rate of 100 MWe/ hour to prepare for main condenser flow reversal. At 0700 load was held at 703 MWe to reverse main condenser flow. At 0950 the load was increased at a rate of 50 MWe/
hour, ~ this rate was then changed to' 25 MWe/ hour at 1050, and to 8 MWe/ hour at 1150. At 1840 full load was achieved at 815 MWe.
May 12: At 2250 the load was increased at a rate of 8 MWe/ hour using.
Recirc pumps. At 2355 full load.was achieved at 810 MWe.
May 13-14: Unit Two held load at 810 MWe until 0350 on May 14 when the load was dropped 25 MWe due to a high Condensate Demineralizer differential pressure.
May 15-16: Unit Two held load until at 0735 on May 16; the load was increased 8 MWe/ hour until 0800. At 2230 load was decreased 200 MWe/ hour in order to shutdown the unit due to main condenser tube leaks.
May 17: At 0247, the turbine was manually tripped. At 0305 the reactor mode switch was placed in the STARTUP mode; at 1123 the SHUTDOWN MODE.
May 19: The unit remained shutdown to repair the main condenser until 0155 on May 19; the reactor mods switch was placed in the STARTUP mode.
Reactor criticality was achievea at 0447, the generator came on-line at 2125 and load was increased.
May 20: At 0330 load was leveled due to high differential pressure on a condensate demineralizer. At 0530 load was increased 15 MWe/ hour; this rate was changed to 75 MWe/ hour at 0755 At 0915 load was held at 500 MWe for Xenon soaking of the fuel.
May 21: At 0230 load was increased at a rate of 25 MWe/ hour, at 0330 the rate of increase was changed to 8 MWe/ hour. At 0110 on May 22 load was held at 740 MWe due to low flow through the condensate demineralizer.
May 22-23: Unit Two held load until at 2135 on May 23; load was decreased 100 MWe in one hour due to a bad condensate demineralizer.
May 24-27: Unit Two held load at 620 MWe until at 0930,on May 27 load .
! was increased 8 MWe/ hour. At 1220 load was decreased at a rate of 100 I i MWe/ hour using the Recirc pumps to change condensate demineralizers. i Load was held at 1330 and the change-over was performed. At 1545 load was increased at a rate of 25 MWe/ hour, this rate was then changed to 8 MWe/ hour at 1645 May 28: At 1305 the load increase was stopped to work on the 28 Recirc Motor Generator set. At 1320 the load was increased at a rate of 8 MWe/ hour until a full load was achieved of 806 MWe at 2300.
May 29-31: Unit Two ended the reporting period holding an average load of 800 MWe.
I l
I
Ill. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE A. Amendments to Facility License or Technical Specifications There were no ' Amendments to Facility License or Technical Specifications during the reporting period.
B. Facility or Procedure Changes Requiring NRC Approval There were no facility or procedure changes requiring NRC approval.
C. Tests and Experiments Recuiring NRC Approval There were no tests or experiments performed during the reporting period.
D. Corrective Maintenance of Safety Related Equipment The following represents a tabular summary of the safety related maintenance performed on Unit One and Unit Two during the reporting period. The headings indiciated in this summary include Work Request Numbers, LER Numbers, Components, Cause of Malfunctions, Resuits and Effects on Safe Operation', and Action Taken to Prevent Repetition.
j .
t i
l l
1 I
l
'I
UNIT ONE HAINTENANCE
SUMMARY
CAUSE RESULTS & EFFECTS W.R. LER OF . ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q05197 1/2 D.G. The oil filter was The oil filter had to The oil filter was re-dirty. be replaced. placed and the D.G.
was tested.
he Q0$196 Reactor The coil was open. The relay was found .The coil was replaced Bldg. Vent open during trouble- and the system was Manifold High shooting of the alarm tested.
Rad. Aux. Relay circuit.
(1-1705-102)
.Q05248 80-15/03L Electromatic The control relay The valve would not The relay and shorting Re's.e/ Valve was bad and a open from the Control bar were replaced and (1-203-38) shorting bar was Room. the valve was tested.
missing.
UNIT TWO MAINTENANCE
SUMMARY
CAUSE RESULTS & EFFECTS OF ON ACTION TAKEN TO W.R. LER NUMBER COP.PONENT HALFUNCTION SAFE OPERATION PREVENT REPETITION NUMBER Q04908 CPD L-7 The scram inlet The .od was drifting The scram inlet valve's (42-77) valve was, leaking in. stroke was adjusted.
through. ,
QO3257 CRD With- The line was scraped N/A The line was repaired draw Riser with a grinder. and tested.
Line (2-0336A-3/4"-A)
Q04869 RBH Ch. 8 The count trip Bypassing LPRM card The count trip level level was out of for 16-250 had no was adj us ted and . tes ted.
adj us tment. effect on count reading of RBH with rod C-7 selected.
Q04883 Vacuum The disc was out of The valve was leaking. The disc was adjusted Breaker (AO- adj ustment, and the valve was 2-1601-32C) tested.
Q04884 Vacuum The disc was out of The valve was leaking. The disc was adjusted Breaker (AO- adjustment. and the valve was 2-1601-32F) tested.
Vacuum The disc was out of The valve was leaking. The disc was adjusted Q04885 Breaker (AO- adjustment.. and the valve was 2-1601-33D) tested.
Q04949 80-13/03L Suppression Aux.-interlock con- The valve would not The aux. Interlock con-Chamber Dump tact.CR105X100N was close. tact was replaced and Valve (2-1001- bad, the valve was stroked 36A) three times.
CRD (50-27) The seals were worn. The CRD had to be - The CRD was replac5d and QOll84 replaced, tested.
L. .__
UNIT TWO MAINTENANCE
SUMMARY
CAUSE RESULTS & EFFECTS OF ON ACTION TAKEN TO W.R. JLER NUMBER COMPONENT HALFUNCTION SAFE OPERATION PREVENT REPETITION NUMBER.
CRD (46-31) The seals were worn. The CRD had to be The CRD was replaced and QOl193 replaced. tested.
CRD (22-39) The seals were worn. The CRD had to be The CRD was replaced and Q01194 replaced, tested.
CRD (06-47) The seals were worn. The CRD had to be The CRD was replaced and Q01195 replaced. tested.
CRD (18-39) The seals were worn. The CRD had to be The CRD was replaced and QO1196 replaced, tested.
CRD (22-11) The seals were worn. The CRD had to be The CRD was replaced and Q01197 replaced. tested.
CRD (26-23) The seals were worn. The CRD had to be The CRD was replaced and QO1199 replaced. tested.
CRD (46-27) The seals were worn. The CRD had to be The CRD was replaced and Q01202 replaced. tested.
CRD (38-07) The seals were worn. The CRD had to be The CRD was replaced and Q01203 replaced. tested.
CRD (26-35) The seals were worn. The CRD failed The CRD was replaced and QO3798 friction testing and the replacement CRD was would withdraw past 48 tested, but not overtravel.
SJAE Suction The solenoid was The solenoid for the The solenoid was replaced Q04725-Valve (AO worn. operator was Icaking and the valve was tested.
5402A) through. .
IJNIT ,,IWO_ , ilAINTENANCI: Si!MilAltY ,
CAUSE RESULTS & EFFECTS W.R. LER OF ON ACTION TAKEN TO
'8UMB E R NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION'
'Q05218 Refuel Bridge The assembly was The main hoist load The main hoist load and-worn. ',
and pulley assembly pulley assembly was would not work. switched with U-l and
tested.
SBLC Heat A temperature Fuses kept opening. The potentiometer was Q04951 Trace potentiometer was adjusted.
Circuits out of adjustment.
QO3896 Valve None found. The valve would not The valve was tested (201001-168) open. three times.
BRM Ch. 12 Four diodes, two The channel was The parts were replaced Q02767 capacitors and one spiking Hi-Hl. and the channel was transistor in the tested.
pre-amp were bad.
Vacuum The discs were out The valves were The discs were adjusted Q04796 Q04797 Breakers of adjustment. leaking through. and the valves were Q04871 (AO-2-1601- tested.
Q04882 33B, 33C, 33F, Q04886 32B,32A)
Q05189 LPRM (48-33B) The LPRM was out of The LPRM was reading The calibration.was calibration.. high. adjusted.
.Q05362 CRD 22-59 The accumulator The accumulator was The fitting was (2-305 pressure switch losing pressure. tightened.
- 59) impulse line fitting was lose.
LPRM (24-170) The LPRM card The LPRH was reading The connector was Q05272 connector was intermittently. , cleaned.
dirty..
IV. LICENSEE EVENT REPORTS The following is a tabular summary of all license event reports for Quad-Cities Units One and Two. occurring during the reporting period, pursuant to the reportable occurrence reporting requirements as set forth in sections 6.6.B.1. and 6.6.B.2. of the Technical Speci fications.
UNIT ONE Licensee Event Date of Report Number Occurrence Title of Occurrence 4-15-80 Reactor Water Level 80-13/03L Switch Drift LIS 263-57A 4-15-80 orywell Pressure 80-14/03L Swi tch Dri f t PS 1001-88A 5-11-80 Electromatic l-203-80-15/03L 3B failed to Operate it - -
UNIT TWO l 5-17-80 MSIV Closure Times 80-10/03L Out of Adjustment i l
l l
l l
l 1
.j i
i f
.i ,
V .' DATA TABULATIONS The following ' data tabulations are presented in this report.
A. L0perating Data Report
-B. ' Average Daily Unit Power Level- ,
C. . Unit Shutdowns and Power Reductions u
I e
p' .
l
- l i
l l
1 4
.I l
+ 4 y '
-j U_ '
F OPERATING DATA REPORT DOCKET NO. 50-254 UNIT ONE DATE June 2, 1980 COMPLETED BY T. Hafera TELEPHONE 309-654-2241, Ext.
176 C ERATING STATUS 0000 050180 L. Raoorting period:2400 053180 Gross nours in reporting period: 744 1 Currently authorized power level (MWt): 2511 Max.Depa.nd capacity
/ :2-Net): 769* Design electrical rating (MWe-Net): 789 .
? Fower level to which restricted (if any)(MWe-Net): NA
- 4. Rcosons for restriction (if any):
This Month Yr.to Date Cunulative I. Number of hours reactor was critical 744,0 3647.0 58361.1
- . Recctor reserve shutdown hours 0.0 0.0 3421.9 3
. Hours generator on line 731.8 3634.8 55675.7
- 5. Jnit reserve shutdown hours. 0.0 0.0 909.2,
- Oross thermal energy generated (MWH) 1456691 8098710 112642691 i:. Gross electricci energy generated (MWH) 465302 2638383 36261437 L:. Net electrical energy generated (MWH) 423479 2474620- 33890156
- 11. deoctor service factor 100.0 100.0 82.6 L3. React or availability f act or 100.0_ 100.0 87.5 14, 'Jnit service factor 98.4 99.7 78.8 LE. Unit avoilcbility factor 98.4 99.7 80.1
. Unit ccpacity. Factor (Using MDC) 72.1 86.0 60.8
.iT Unit capacity factor (Using Des.MWe) 74.0_ 88.2 62.4~
- 13. Unit. forced outage rate 1.6 .3 7.7 19..-Shutdowns scheduled over'next 6'nonths (Type,Date,and Durntion of each) -
2 0 '. - I f shutdown-atiend of report period,estinated dcte of startup ,,__pA________
l
- C t S' .ov- de loser then 769.We dt.aing periods of high chicnt tenparature due
- *: .' .treel perhe.nate of the spra coul.'
OPERATING DATA REPORT DOCKET NO. 50-265 UNIT T!JO DATE June 2, 1980 20MPLETED BY T. Hafera TELEPHONE 309-654-2241, Ext.
~~~i76 0;EiATING STATUS 0000 C50190 in reporting period: 744
- 1. Rap;rti.ng period:2400 053190_ Gross nours E. Currently cuthorized power level (MWt): 2511 Mux. Depend ccpacity
" W.. a - N e t ) : 769* Design electrical roting (MWe-Net): 789 NA
- 3. Power level t o wh ich res tricted(if any )(MWe-Ne t ) :
- 4. Reasons f or restriction (if any):
This Month Yr.to Date Cumulative 696.i_ 849.4_ 53978.3_
- 3. Nunber of hours reactor was critical 0 . 0_ , 0 . 0_ '2985.8
- b. Recctor reserve shutdown hours 664.5_ 736.2 51529.0
- 7. Hours generator . - on line 0.0 0.0 702.9_
O. Unit reserve shutdown hours.
1448078_ 1615422_ 104631531 Y Gross thernal energy generated (MWH) 462506 509491 33372541
- 14. Gross electrical energy generated (MWH) 440847 442575 31185100_
11 Net electrical energy generated (MWH) 23.3 7 7 . 4_
93.6_
- 12. Reactor service factor 81.7 93.6_ 23.3 L3. Reactor availability factor 39.3 20.2 73.9
- . 4 . Unit service factor 09.3. 20.2 74.9
- 45. Unit availability factor 75.i_ 15,4_ 56.7
- 14. Unit cupucity factor (Using MDC) 7 7 . i_- 15.8 58.2 L7 Unit capacity Factor.(Using Des.MWe) 21.2_ 9.4 10.7_
- 13. Unit forced outage rate
- 17. Shutdowns scheduled over next 6 Months (Type,Date,and Duration af each):
p e r i o d , e s t inct ed d a t e o f s t ar t u p ,, ___N_A_____,,__
20.LIf. shutdown ot,end-of report IIht ?.iG M/ 2t lower!the 769 tiUt during priods of high ebient tenpercturt det M *nt tr.ited perf 4rmnct ci the sprG7 CCMI. . hP (P p .y ,,
APPENDIX D AVER AGE DAILY UNIT POWER LEVEL -
DOCKET NO. 50-254 m
IJNIT ONE DATE June 2, 1980 COMPLETED BY T. Hafera TELEPHONE 309-654-2241, Ext.176
" .: i, T ;i Mov 1980
. Ai AVERAGE DAILY PCWER LEVEL DAY AVERAGE DAILY POWER LEVEL
( M'.J e -N e t )
(MWe-Net) 636.0 17, 631.4 634.9 18. 583.6 631.2 19, 596.7 3.
4, 637.3 20. 595.5 632.2 21. 589.9 3.
- - 632.2 22, 593.8 j
. 6.
628.4 23. 590.3
?.
3, 627.2 24. 587.8
'. 636.4 25. 582.2 590.1 26. 588.3 U.
l 133.3 27. 573.2
- 1. i
- 28. 569.6 !
- 2. 151.8 3, 437.6 29. 563.0
- . 620,5 30, 562.4 634.1 31. -564.8
.3.
is. 604.4 l 1
1 i
INSTRUCTIONS
- - this forn, list the average daily unit power level in Mk-Net for ecch day in the reperting nonth.Corpute to the i:rtst eole neemtt.
tit i: pres aiIl at used to olst c gec: 5 for tech reporting conth. Note thst . hen noxim de:enicle coaccito is there not be cccesions whta the daily everece power level excteds the
.n:
- ';i:e Letthe(;rr.et theelectrical restrictedretir.g state of level the snit,In line).such casts,tne average daily unit pmr cotpit sheet shasid se ;
3:str.;ted to etalain the cpparent entnoly 1
APPENDIX B AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-265 UNIT TWO DATE June 2. 1980 COMPLETED BY T. Hafera
'f TELEPHONE 309-6%-2241. Ext. 176
b7H, Mov 1980 __
DAY AVERAGE DAILY POWER LEVEL
..f A'JER AGE DAILY POWER LEVEL (MWe-Net)
(MWe-det)
- 17. 30.0
~'
317.1 gg* -6.6 386.4 6.1 583.6 19.
~. .
20.
419.5
- 4. 636.5 21, 589.9
- 5. 665.0
- 22. 689.3 6.
727.4
.! 23. 677.9 7, . - 731.3
- 24. 589.3 3.
533.5
- 25. 606.4 9.
661.1
- 26. 625.9 13, 716.8
- 27. 656 0 J 11, 743.6
- 28. 694.2 2 777.5 i 29, 754.4 759.8 l
- 13. !
- 30. 754.9 D. 759.6 i 761.7 l 31.
1E. 779.4 !
15 741.3 Cr. t.nsINSTRUCTI0 tiscurcie dcily unit puer __ level in M'Jediet ict each day in the reporting an icr i,11st the i l
' (~4W6NfE'+y/dM.,,#7;$6M$,f
- daceti:.ub le n g tt ? a@lNSU[M"/d@$%j@a'iEM$#ff0"Se,%8$$$
t
. - x f E 1 .
? 4 p
- 2 i
- a 2
, r' C r - t
\ 3 e 4 ,
rY f 5 S r.- 5 8 eB 1
' 7 a 6 T kD H - N n9 9 o E aN
- 0 o1 0 t eA r 0i T 3 M rT 3 sh 7 M bS i c O P vr C dT
) T ea Y / l O
~- QRM B S eH .
N i
-. D E O f n E N I i T O T t E H C ed
~- 1 L P A S e P E c
- M L E G a .
i O E V - l C T , I Mp T cs C
3 E ra R i w R c .
L O er C R o t ;
B c 5c S I a er c
a; e
N l l O pi I
T eo .
R c C
N U D
E R 0 .
8 9 R E
9 1 g8 A
) .
E- DW N i .
O Y s5 g g:8 a .
XP A i n
1 M f
- IDD .
bp ( .
NN .
A -
E A .
N
- 1 P I .
g8 -
PS i AN T xpy* '
W N 9
r O
D O
M . .
T .
T E O lV i R ETN S O SN _
i P NET . _
T E EVR A . . _
R -
CEO t" I N I P N O U L E R
J
~
~
[
$b6S gR _
e pt - 5 1 E @gy M ' .
i
"* E z86" A .
N O
E^ 0 N)
S 8 OS
. 9 I R 5 5
I 1 TU 2. .
- T ,
AO 2 I
RH _
"0 4 5
C 2 U(
1 D
2 D e
- A n m eo '
0 u u 9 5 q J Ey F
- L ,
)
. 1 O E E 1 1
T
^ '( M N 5 A A 0 "E T E
N D 8
0 K T E C I T
t O N A
' D U D * .
~ O N 7
'I
r 1 p ,a p.4 g .w.3 g,~5 gw , 7. -y p-$ p3 "r' 1 E9 P'"'1 M n '
.-- l
. '~~ ~ 1 s'~~
T " ' '- . ,
( \ f1 h APPENDIX D QTP 300 ".13 '
UNIT SilVTDOWNS AND POWER REDUCTIONS Revision 5 DOCKET tt0. 50-265 March 1978 UNIT HAME QUAD-CITIES TWO CCHPLETED BY T. Hafera DATE June 2, 1980 REPORT MONTil TELEPHONE 309-654-2241, Ext.
MAY 1980
.. l'16 f,s w $ s e = Eb z 5 Ee 8 S LICENSEE pg gg . ,
pou. DURATION $ jc 3{ $
- EVENT p8 @:8 ,
" m CORRECTIVE ACTIONS /COMMEllTS NO. DATE (1100RS) y 5 5- REPORT 110. 8 R
i
.3 800501 F 12.9 A 3 NA NA NA Reactor scram from low vessel water level due to
. 2B Feedwater Regulation valve falled closed.
0.0 A NA NA NA NA Load reduction due to false sensor reading on 4 800504 F high pressure turbine vibration.
5 800507 F 0.0 A NA NA NA NA Load reduction due to feedwater flow reduction test and drywell entry to repair 2B Recirc pump oil level indication.
6 800517 F 66.'6 B 1 NA. NA NA Shutdown to repair main condenser tube leaks.
i 0
(finai)
VI. UNIQUE REPORTING REQUIREMENTS The following items are included in this report based on prior commitments to the commission:
A. .iAIN STEAM RELIEF VALVE OPERATIONS Relief valve operations during the reporting period are summarized in the following table. The table includes information as to which relief valve was actuated, how it was actuated, and the circumstances resulting in its actuation.
VALVES NO. & TYPE PLANT DESCRIPTION UNIT DATE ACTUATED ACTUATIONS CONDITIONS OF EVENTS 1 5-11-80 1-203-3A 1 Manual Rx Press Surveillance 1-203-3B Failed to Open 960 T.S. 4.5.D.1.b.
1-203-3C 1 Manual 1-203-30 1 Manual Defective 1-203-3E 1 Hanual solenoid hold switch' replaced 5-12-80 1-203-3B 3 Manual Rx Press Post Maintenance 960 B. CONTROL ROD DRIVE SCRAM TIMING DATA FOR UNITS ONE AilD TWO The basis for reporting this data to the Nuclear Regulatory Commission are specified in the survelliance requirements of Technical Specificiations 4.3.C.1 and 4 3.C.2.
The following table is a complete summary of Units One and_Two Control Rod )
Drive Scram Timing for the reporting period. All scram timing was per- !
formed with reactor pressure greater than 800 psig.
I l
l I
l l
l l
RESULTS OF SCRAM TlHING HEASUREMENTS PERFORMED ON UNIT ONE CONTROL ROD DRIVES, FROM 5-1-80 TO 5-31-80 AVERAGE TIME IN SECONDS AT % Max. Time INSERTED FROM FULLY WITHDRAWN For 90%
Insertion DESCRIPTION NUMBER 5 20 50 90 Technical Specification 3.3.C.I &
,DATE 0.375 0.900 2.00 3.5 7 sec. 3.3.C.2 (Average Scram insertion Time)
OF RODS 5-11 88 0.29 0.67 1.45 2.56 M-Il 2.96 Seq. A Hot Scram Timing, Unit 1 o
'l .
l l
I l
VIi. REFUELING INFORMATION The following information'about future reloads at Quad-Cities Station was requested in a January 26, 1978 licensing memorandum (78-24) from D. E.
O'Brien to C. Reed, et. al., titled "Dresden, quad-Cities, and Zion Station -
NRC request for refueling information dated January 18, 1978.
b
.l e
(LL y :au u-w.
Revision 1 - - -
~ ~ . . mrec.4 .1973 ' -i l . . ..
~~~~.-~ . ~ QUND-CIT I ES' 'REFUELI!iG - .
!!! FORMATION REQUEST --
. . .: y .- _
l.~_ . .. . . _ .
.- 4 _ Cycle: -
5 :. .
- 1. Unit: 2
__ Reload:
. - ~ ': _ -
- - : f- --:-l- . " . -;- L ~_
~ ~ " O 11-M79 -(Shutdown- EOC4'.
2.
Scheduled date for'iiext'~refdelirig sh'dtdoWn:
i ; 1-17 80 (Startuo EOC5) 3 Scheduled date for restart following refueling: '
- 4. Will refueling o'r resumption of operation thereaf No, ter require Plan 10CFR50.59 a technical-Reloads specification change or other license ' amendment: The review will be conducted-for future cycles of Quad Cities Unit 2. .
i by early September,1979 ..
Scheduled date(s) for' submitting proposed licensing action and supporting '
~
5 a gu
. information: Early ' August,1979 for 10CFR50.59 related. changes , ,
days prior to shutdown.
e.g., new or I
I 6. Important licensing considerations associated dith r'efueling,
" ' different fuel desig*n or supplier, unreviewed design 'or performance analysis -
' methods, significant changes in fuel design, new operating procedures:
New Fuel Design: Retrofit 8x8 fuel (180) '
i a nat. U at bundle top and' bottora b)) two larger water rods ,. :
l' '
- d' "
. cJ)prepressurized 6
new enrichment distribution
~
! This fuel design was 'previously used for Quad Cities Unit: 1 Cycle 5 *
! and Dresden Unit 2 Cycle 7 .
~
.- l i*
7 ,
The number of fuel assemblies. . . .
724
! a. Plumber of assemblies in core:
! 492
- b. Ilumber of assemblies in spent fuel pool: .
i
! The present licensed spent fuel pool storage capacity and the size of any S. j
} increase in licensed storage capacity that has been requested or i~s planned ,
in number of fuel assemblies: . '
t, -
1460 . .
I a. Licensed storage capacitv for' spent fuel: .
None .
I b. Planned increase in licensed storage: _
9 The projected date of the last refueling that can licensed be discharged capacity: March,' 1986 to the
? spent' fuel pool assuming the present -
(End of batch discharge. capability)
! K P P R O V.E [l
< .... a 1
~ 'APR 2 01973 j
.- Q.C.O.S.R.l1 1
1 4 _
' Revision 1 Harch 1973 QbAD-CITIESREFUEll!!G .
ll! FORT %T10!! REQUEST *
- (
s.
_ cycle: 6 .
1 Reload: 5
- h. Unit: _
8 9CRn . (he...a (goc 5)
Scheduled date for next refueling sliutdown: " .
- 2. . 12-; 80(Startun BOC6)'..
- : following refueling:
Scheduled date for restart 3 . l
- 4. Vill refueling o'r resumption of operation thereaf ter require a technic specification change or other 1icense amendment:The review will' be conducted '
for future cycles of Quad Cities Unit 1. . . .
in June, 1980. i -
Scheduled date(s) for submitting proposed. licensing action and support ng ,
5 June,1980 for .100FR50.59 related changes ' a 90 days
[
. information: '
~
prior to shutdown. .
e.g., new or
- 6. ' dif Important licensing considerations associated w'ith '
refueling, f snalysis ferent fuel desigh or supplier, unreviewed design or per ormance d methods, significant changes in fuel design, new operating proce ures:
(^- 224)
New fuel designs: Retrofit 8x8' fuel 2.65 w/o% . .
. ~- ,
l, -
. g l
The number of fuel assemblies. * - ,
7 .
l 724
~ - ,
- a. !! umber.of assemblics in core: .
l 596
- b. !! umber of assemblies in spent fuel pool: __
- S. The present licensed spent fuel pool storage capacit -
in number of fuel assembi es: , . I j
1460 . ,,- .
a.
Licensed storage capacity for' spent fuel: -
.. . ~. .
I
' None .
b.
Planned' increase in licens'ed storage: .
~
l refueling that can be discharged to the
- 5. .The projected date of the last licensed capacity: Septeinber, 1985 pent'of fuel ool assu:ning the present l' end ba ch discharge" capability) '
WPPROVEC APR 2.01973 Q.C.O.S.R.
I f . . . . .
Vill. GLOSSARY The following abbreviation which may have been used in the Monthly Report, are defined below:
CRD - Control Rod Drive System SBLC -
Standby Liquid Control System MSiv -
Main Steam Isolation Valve RHRS - Residual Heat Removal System RCIC - Reactor Core Isolation Cooling System HPCI -
High Pressure Coolant injection System SRM -
Source Range Monitor I RM -
Intermediate Range Monitor LPRM -
Local Power Range Monitor APRM -
Average Power Range Monitor TlP - Traveling incore Probe .
( - - RBCCW -
Reactor Building Closed Cooling Water System TBCCW -
Turbine Building Closed Cooling Water System RWM -
Rod W o rth Minimizer SBGTS -
Standby Gas Treatment System HEPA .
High-Efficiency Particulate Filter RPS -
Reactor Protection System IPCLRT -
Integrated Primary Containment Leak Rate Test LPCI -
Lo'w Pressure Coolant injection Mode of RHRS RBM. -
Rod Block Monitor BWR -
Bolling Water Reactor.
ISI -- In-Service inspection MPC -
Maximum Permissible Concentration
PCI - Primary Containment isolation SDC - Shutdown Cooling Mode of RHRS LLRT - Local Leak Rate Testing MAPLHGR - Maximum Average Planar Linear Heat Generation Rate R.O. - Reportable Occurrence DW -
Drywell 1
i RX
- Reactor EHC - Electro-Hydraulic Control System MCPR - Minimum Critical Power Ratio i
PC10MR - Preconditioning Interim Operating Management Recommendations i - Licensee Event Report LER ANSI -
American ilational Standards Institute NIOSH - National Institute for Occupational Safety and f - - Health ACAD/ CAM
- Atmospheric Containment Atmospheric Dilution /
' Containment Atmospheric Monitoring I
1 f
m u -
,