ML20236W368
| ML20236W368 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 10/31/1987 |
| From: | Robey R, Schmidt K COMMONWEALTH EDISON CO. |
| To: | Lieberman J NRC |
| References | |
| RAR-87-51, NUDOCS 8712070423 | |
| Download: ML20236W368 (29) | |
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Commonwealth Edison ouad Cities Nuclear Power Station 22710 206 Avenue Norta Cordova, Illinois 61242 Telephone 309/654-2241 f
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G RAR-87-51 November 4, 1987 U.S. Nuclear Regulatory Commission Washington, D. C.
20555 Attn:
J. Lieberran M/S MN88-4100
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Enclosed for your information is the Monthly Performance Report covering the operation of Quad-Cities Nuclear Power Station, Units
}1 One and Two, during the month of October,1987.
T Respectfully, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION i
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D R. A. Robey Services Superintendent vk
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OCTOBER, 1987
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lummary of thehting Experience 3
A.
Unit One B.
Unit Two
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Plant or Procedure, Changes, Tests, Expe:Iments, and Safety I'
Related Mainten3Nej r
- f.. Amendments to Facility License 'r Technical Specifications o
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B.
FacilityorProcedur),dangesRequiringNRCApproval
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Tests and Experiments; Requiring NRC Approval Corrective Maintena.rea H}. 5afety Rela, tad Equipment D.
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IV.
Licensee Event Reports,
V.
DataTabujottors t
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A.
Operating Data Report Average Daily #olt; h ar i.evel j
B.
J C.
Unit Shatdowns and Power Reductic:is 1
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unique ReportinIl Requiremerns
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Main Steam Relief Valve Operations 4
B.
Control Rod Drive Scram T/..iing Data VII.
Refueling Information i
VIII.
Glosse.ry.
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0027H!0061Z
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I '. ' INTRODUCTION 4
. Quad-CitiesNuclearPowerStationiscomposedoftwoBoilingbtero
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Reactors,eachwithaMaximumDependableCapacityof763HNe-(tt)locatedin j
r Cordova, Illinois.
The Station is jointly owned by Commm wealth Edison a
Company ard Iowa-Illinois Gas & Electric Company.
The Nuclear Steam Supply-Systenis are General Electric Company Bolling Water Reactors.
The
- Architect / Engineer was Sargent & Lundy, Incorporated, and thE primary'
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construction contr,1ctor was United Engineers & Constructors.
The Missis?,ippt
[v River is'the condenser cooling water source.
The plant it subject to license
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4 numbers DPR-29 and DPR-30 issued October 1, 1971. r.nd March 21, 1972, f
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respectively; pursuant to Docket Numbers 50-254 and 50-265.
The date of
- initial Reactor criticalities for Units One and Two, respectively were October 18, 1971, and April 26, 1972. Commercial generation..of power began en February 18, 1973 for Unit One and March 10, 1973 for Unit Two.
This report was compiled by Verna'Koselka and Kurt Schmidt, telephone number 309-654-2241, extensions 2240 and 2147.
0027H/00612 L
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II.
SUMMARY
OF OPERATING EXPERIENCE A.
Unit One October 1-31 Unit One continues in the end of cycle nine refueling and maintenance shutdown which began on September 12, 1987 i
B.
Unit Two October 1-15 Unit Two began October while reducing load to meet Load Dispatcher (LD) requirements. A minimum load of 550 MWe was reached at 0200 on October 1 and held until 0545. Power was then slowly increased and the unit reached full load at 1300. Full load was held until 2345 when it was reduced for EGC.
The unit was operated in EGC from 0000 until 0950 on October 2 when
,[
the unit was increased to full load. At 1605 power was reduced and the 4
unit placed back in EGC at 1625 on October 2.
The unit operated in EGC
(
with'only minor interruptions for surveillance and maintenance until October 10.
At 0910, the unit was taken out of EGC and load held at 800 MWe due to difficulties with the station computer which performs core thermal limit calculations.
Following the computers return to service at 1810, power was reduced to within EGC limits and at 1842 the unit was o
placed in EGC.
Again, the unit operated in EGC with only minor interrup-tions until October 13.
At 0340 the unit was taken out of EGC and the units j
' load reduced at the request of the LD.
Load was held at a minimum of 650 j
MWe from 0405 to 0440. Power ascent was begun, and at 0505 the unit was returned to EGC. At 1130 the unit was taken out of EGC and raised to full load which it held from 1155 until 1400.
At that time, problems with level control in a moisture separator drain tank required a slight load reduction to correct. Load was held at 790 MWe until 0625 on October 14.
An ascent to full load was begun, but the reoccurrence of level instabilities allowed only 805 MWe which was held from 0630 until 1510 when the unit was placed 3
in EGC.
The unit was operated in EGC, again with only brief, routine 3
interruptions from 1510 on October 14 until October 19.
{
Optober 16-31 On October 16, 17 and 18, the unit was operating in EGC. At 0951 on October 19, EGC was tripped due to a damaged circulating water pump (CWP).
j The 2A CWP was taken off line at 0956 and at 1003 the unit was returned to EGC.
At 1959 on October 19, a reactor scram occurred due to low reactor water level. The icw wa:er level occurred due to a loss of the 23 bus which was i
i 0027H/0061Z j
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W caused when an equipment operator inadvertently racked out the wrong circuit breaker thus de-energizing the bus and causing extensive damage to electrical i
l equipment. Falling reactor water level caused an automatic (Group I) isola-tion of the'r2 actor and an automatic initiation of the High Pressure Coolant Injection and Reactor Core Isolation Cooling systems. The core was kept
)
covered, feedpumps were restarted and following restoration of normal reactor water level control,'the emergency systems were returned to normal condition.
Shutdown cooling was started at 2345. At 0040 on October 20, the 2C CWP l
was taken off line. At 0.s0 the reactor head vents were.open and at 0210 l
the unit was in a normal shutdown status. Drywell deinerting began at 0425 and was complete at 0950.
The plant spent the remainder of the month I
in a shutdown condition performing repairs on the equipment damaged in the October 19 event. Electrical cross-tie connections and the emergency diesel generator were used extensively during this period to augment the damaged electrical equipment. The October 19 event is detailed in LER 2-87-013.
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e III.
PLANT OR PRDCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE A.
Amendments to Facility License or Technical Specification.
There were no Amendments to the Facility License or Technical Specifications for the reporting period.
B.
Facility or Procedure Changes Requiring NRC Approval There were no Facility or Procedure changes requiring NRC approval for the reporting period.
C.
Tests and Experiments Requiring NRC Approval There were no Tests or Experiments requiring NRC approval g
for the reporting period.
D.
Corrective Maintenance of Safety Related Equipment 7
The following represents a tabular summary of the major safety related maintenance performed on Units One and Two during the reporting period. This summary includes the fcilowing: Work Request Numbers, Licensee Event Report Numbers, Components, Cause of Malfunctions, Results and Effects on Safe Operation, and Action Taken to Prevent Repetition.
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0027H/0061Z
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I
UNIT 1 MAINTENANCE
SUMMARY
WORK REQUEST NO.: Q54096 LER NUMBER: NA COMPONENT: System 300 - Installed washers on holddown bolts, changed lockwashers and defective bolts on HCU's.
Torqued to 50 ft. Ibs. +5.
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CAUSE OF MALFUNCTION: The cause of the missing or loose mounting bolts with no flatwashers on the Hydraulic Control Units (HCU's) is due to installation error.
RESULTS & EFFECTS ON SAFE OPERATION: According to a study performed by Sargent and Lundy, the event does not impair the operability of the HCU's.
ACTION TAKEN TO PREVENT REPETITION: The missing mounting bolts will be replaced and all bolts for both units will be tightened. Also, flatwashers will be installed on all mounting bolts as designed.
,I WORK REQUEST NO.:
Q58075 E'
LER NUMBER: NA L
COMPONENT: System 1300 - Taped butt splices in local pull boxes of RCIC Temp Switches to meet EQ qualifications.
CAUSE OF MALFUNCTION: There was no malfunction of the RCIC Temperature Switches.
RESULTS & EFFECTS ON SAFE OPERATION: Operability assessment done by Bechtel Eastern Power Corporation stated that the results and effects on safe operation were minimal.
ACTION TAKEN TO PREVENT REPETITION: The butt splices for the switches were taped and the station is now maintaining better documentation on splices.
_ _ _ _ _ _ _ _ _ _ _ _ _ - _ - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' - - - - - - - - - - - - - - - - - - - - - - - - - - - ' - - - ' - - - ' - - ~ ~ - - - - - ' ~ ~ ~ ~ ~ ~ ~
i WORK REQUEST No.: Q58191 LER NUMBER: NA COMPONENT:
System 595 - Found wires shorted on 261-18D switch. Replaced with new calibrated switch.
Recalibrates 261-17D.
CAUSE OF MALFUNCTION: Half of a Group I Isolation occurred due to the failure of temperature switch 1-261-18D. The failure was caused by grounding due to the wiring insulation being rubbed off on the temperature switch support during excessive vibration.
RESULTS & EFFECTS ON SAFE OPERATION: The safety implications were minimal since the other area monitor 2 were operating properly and the temperature switch failed in the conservative direction -- a half Group I isolation occurred.
ACTION TAKEN TO PREVENT REPETITION:
The temperature switch was replaced like-for-like and supported to prevent damage due to excessive vibrations. As this i
is an isolated incident, no further action is necessary.
j
- 7.
L WORK REQUEST NO.: Q58308 I
LER NUMBER: 87-012 1
d COMPONENT:
System 1700 - Removed coil from relay 500F57 and installed in relay 1-1700-100B.
CAUSE OF MALFUNCTION: The Reactor Building and Control Room Ventilation systems isolated due to relay (1-1701-100B) coil failure attributed to age.
RESULTS & EFFECTS ON SAFE OPERATION: As both trains of SBGTS were operable and there was a redundant channel of the process radiation monitor available, safety implications are minimal.
ACTION TAKEN TO PREVENT REPETITION:
The coil for relay 1-1701-100B was replaced like-for-like and operability verified. Also, modification M-4-1(2)-85-17 has
_I been initiated to replace similar model relays in the 901(2)-40 and 41 panels to prevent further recurrence of this problem.
4 12 '
WORK REQUEST NO.: Q58588 LER NUMBER: NA COMPONENT:
System 1700 - Replaced fuse on 1-1705-3B Off Gas Rad Monitor with a safety-related one.
CAUSE OF MALFUNCTION: The cause of the IB SJAE radiation monitor (1-1705-3B)-
'to give a downscale signal and blow its fuse was a loor high voltage connector.
RESULTS & EFFECTS ON SAFE OPERATION: As the chimney radiation monitors were available at all times during this. event, safety implications are minimal.
ACTION TAKEN TO PREVENT REPETITION: The high voltage connector was tightened and the fuse replaced.- The monitor was then successfully tested and returned to service.
WORK REQUEST No.: Q60109 e.
O LER NUMBER: NA h
r COMPONENT:
System 0201 - Replaced spacer #2 on fuel bundle LYJ449 by G.E.
damaged during movement into new fuel vault.
I CAUSE OF MALFUNCTION: The failure of spacer #2 on fuel bundle LYJ449 was due to a misaligned fuel rack guide which, caused the spacer to catch while being placed
. in the racks.
RESULTS & EFFECTS ON SAFE OPERATION: There were no effects on safety as the fdamage was immediately noted and corrected.
ACTION TAKEN TO PREVENT REPETITION: The spacer was replaced by the vendor and' a work request was written to inspect and repair new fuel racks to prevent recurrence.
UNIT 2 MAINTEwNCE
SUMMARY
WORK REQUEST NO.: Q56048 l
LER NUMBER: NA COMPONENT: System 263 - Installed new 2-263-72D Lo-Lo Reactor Water Level Switch.
CAUSE OF MALFUNCTION: Ihe reactor water level indicating switch (LIS) 2-263-72D was inspected by the Instrument Maintenance department, but no problems were noted.
The LIS had previously not closed when tested, but operated successfully three times after that.
i RESULTS & EFFECTS ON SAFE OPERATION: As the logic is set up in a one-out-of-two twice arrangement and since the other three switches (2-263-72B, C, and D) functioned properly during this surveillance test, the safety significance of this event is considered minimal.
ACTION TAKEN TO PREVENT REPETITION: As the inspection found no problems, a mercury switch associated with LIS 2-263-72D was replaced with a new one as a preventative measure and then functionally tested satisfactorily.
3
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WORK REQUEST NO.: Q57390 LER NUMBER: NA r
COMPONENT:
System 263 - Replaced magnet in Reactor to Lo Level Switch 2-263-72D, calibrated and functionally tested.
CAUSE OF MALFUNCTION: The reactor water level indicating switch (LIS) 2-263-72D did not close again and was later discovered to be a seal leak causing atmospheric contamination which made the switch operats intermittently.
RESULTS & EFFECTS ON SAFE OPERATION: As the logic is set up in a one-out-of-two twice arrangement and since the other three switches (2-263-72B, C, and D) functioned properly during this surveillance test, the safety significance of this event is considered minimal.
~..
ACTION TAKEN TO PREVENT REPETITION: This work request replaced the magnet for the switch in 2-263-72D which did not correct the problem as later discovered.
The defective switch was replaced under W.R. Q58188.
l
WORK REQUEST NO.: Q57392 LER NUMBER: NA COMPONENT:
System 1700 - Calibrated 2-1743A Fuel Pool Rad Monitor and functionally tested.
CAUSE OF MALFUNCTION: The 2-1705-16A, Fuel Pool Radiation Monitor, failed due to instrument calibration drift.
RESULTS & EFFECTS ON SAFE OPERATION: The logic for this to initiate safety systems is one-out-of-two.
Since radiation monitor 2-1705-16B was functioning properly, the safety implications of this event were minimal.
ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action was to recalibrates the monitor. A change to the Technical Specification setpoint limit is in progress to prevent repetition.
71 WORK REQUEST No.: Q58358 i-LER NUMBER: NA 1
COMPONENT:
System 6600 - Installed new coupling on 2-6601 Diesel Generator Soak j
Back Pump.
CAUSE OF MALFUNCTION: The cause of the coupling failure on the Oil Recirculating Pump, 2-6601, was due to misalignment.
RESULTS & EFFECTS ON SAFE OPERATION: The pump was available until it was removed from service for repairs and all surveillance required for an inoperable Diesel Generator were successfully performed.
ACTION TAKEN TO PREVENT REPETITION:
Immediate Corrective Action was to replace the coupling like-for-like.
Also, alignment -- jackscrews are being installed for ease of alignment and alignment support.
(W.R.'s Q59465,Q59466, and Q59467)
T-
.i WORK REQUEST NO.: Q59021 l
LER HUMBER: NA COMPONENT: System 10M - Replaced two lugs on torque switch on MO 2-1001-36A RHR Torus Pump.
CAUSE OF MALFUNCTION:' Tha lug connections to the torque switch were broken due to normal vibration and wear resulting in dual indication and failure of valve 2-1001-36A to open.
RESULTS & EFFECTS ON SAFE OPERATION: The 'B' loop of RHR Containment Cooling was proven operable at the time, therefore safety implications were minimal.
ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action was to install new lugs. The valve was cycled and determined.co be operable. As this is an isolated incident, no further action is necessary.
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c; IV.
LICENSEE EVENT REPORTS The following is a tabular summary of all licensee 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 6.6.8.1. and 6.6.8.2. of the Technical Specifications.
UNIT 1 Licensee Event Report Number Date Title of Occurrence 87-020 10-6-87 ESF Actuation RBV Radiation Monitor Relay UNIT 2 e
87-013 10-19-87' Scram and ECCS
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Initiation s
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V.
DATA TABULATIONS l
The.following data tabulations are presented in this report:
A.
Operating Data Report 8.
Average Daily Unit Power.' Level l
L C.
Unit Shutdowns and Power Reductions l
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m-3 APPENDIX C.
OPERATING DATA REPORT -
"=
DOCKETNO.
~.S.f O ^Id UNIT 2.-
/O Jd d[
DATE COMPLETED BY
'6f8 d> '.
809 - 0 *IM-??Q TELEPHONE OPERATINGSTATUS C000
/CC M 7 1.' REPORTING PERIOD: 2#
'O 3/U GROSS HOURS IN REPORTING PER10D:
3 5// MAX. OEPEND. CAPACITY (MWe Neti:
bY
- 2. CURRENTLY AUTHORIZED POWER LEVEL (MWt):
DESIGN ELECTRICAL MATING (MWo. Net):
WV
- 3. POWER LEVEL TO WHICH RESTRICTED (IF ANYI (MWe Neti:
- 4. REASONS FOR RESTRICTION (IF ANYl:
THl3 MONTH YR TO DATE CUMULATIVE.
Uous
/0643
- a. o
- 5. NUMSER OF HOURS REACTOR WAS CRITICAL.
N
- 6. REACTOR RESERVE SHUTOOWN HOURS...................
C
6C/G.7 '
M S* T 3 /.?.
- 7. HOURS GENER ATOR ON LINE............
CO Q.O 9M2
- 8. UNIT RESERVE SHUTDOWN HOURS..................
0
/ Y/73 U 2-27124CV *iC
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- 9. GROSS THERMAL ENERGY GENER ATED (MWHI d' d 96 IE. I/
72 39 3dB
- 10. GROSS ELECTRICAL ENERGY GENERATED (MWHi...........
CO N Y 'M/ 6~/8 77 W 3
- 11. NET ELECTRICAL ENERGY GENERATED (MWH)..............
- 12. r EACTOR SERVICE F ACTOR...................,.......
_ b O f-O N27 I
O 827 8.2.7 t
- 13. R EACTOR AV AI LA BILITY F ACTOR.......................
C' 824 7^
- 14. UNIT SERVIC E F ACTOR...........................
- ]
- 15. UNIT AV AILABILITY F ACTOR..........................
7YO 5
d'
- 16. UNIT CAPACITY FACTOR (Using MDC)
OU 770 0
- 17. UNIT CAPACITY F ACTOR (Using Design MWel..............'...
3O U,4
'I Y
- 18. UNIT FORCED OUTAGE RATE..........................
- 19. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE. DATE, AND DURATION OF EACHl:
I
!#- &87
- 20. IF SHUT DOWN AT END OF REPORT PER100, ESTIMATED DATE OF STARTUP:
FORECAST ACHIEVED
- 21. UNITS IN TEST STATUS (PRIOR TO COMMERCI AL OPERATION):
INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 1.16 9
4 m
APPENDlX C OPERATING DATA REPORT 00CKET NO.
"bd.__
UNIT
'd dd ks DATE CA 4' C.
/
COMPLETED BY AO SY~
?l TELEPHONE r
OPERATING STATUS ggo foof sp MdO
/d 3/ U GROSS HOURS IN REPORTING PER100:
- 1. REPORTING PERIOD:
N// MAX. OEPEND. CAPACITY (MWo.Neti:
M9
- 2. CURRENTLY AUTHORIZED POWER LEVEL (MWti:
7M DESIGN ELECTRICAL RATING (MWe Net):
- 3. POWER LEVEL TO WHICH RES1 RICTED liF ANY) (MWo. Net):
'l e
- 4. REASONS FOR RESTRICTION IP ANYl:
THIS MONTH YR TO DATE CUMULATlVE 2* O
- ~ 2 '7' E
'O 3 I
- 5. NUMSER OF HOURS REACTOR WAS CRITICAL...............
- 6. REACTOR RESERVE SHUTDOWN HOUPS..................
OO O
'I
'" # S' i
- 7. HOURS GENER ATOR ON LINE.......................
c 8, UNIT RESERVE SHUTOOWN HOURS......................
D D
............/O'
/M/M f /NO 3/
- 9. GROSS THERMAL ENERGY GENERATEQ(MWH)
Y//- d[./M i
- 10. GROSS ELECTRICAL ENERGY GENERATED (MWH).......
M
- 11. NET ELECTRICAL ENERGY GENERATED (MWM)
A f*. ?
M8 76.6,
- 12. r EACTOR SE RVICE F ACTOR.......................
b
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N$
7 P
- 13. R E ACTOR AVAILABILITY F ACTOR.......................
N'!
Y 7 N
- 14. UNIT SERVIC E F ACTOR...............................
O T4 ~I
'2 P
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- 15. UNIT AV AILASILITY F ACTOR..........................
7' 5-Y0 0
- 16. UNIT CAPACITY FACTOR (Using MDC).....................
Ec M y'
' 'g* 7
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- 17. UNIT CAPACITY FACTOR lusing Design MWel..............'..
N
/ Y S-
- 18. UNIT FORCED OUT AGE RATE..........................
- 19. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE. DATE, AND OURATION OF EACHl:
- -n uxy,.,.. a. + :,
- 20. IF SHUT DOWN AT END OF REPORT PERIOD. ESTIMATED DATE OF STARTUP:
FORECAST ACHIEVED 21, UNITS IN TEST STATUS (PRIOR TO COMMERCIAL OPERATIONI:
INiflAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 1.16 9
APPENDIX 8
. AVERAGE DAILY UNIT POWER LEVEL
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N'U U DOCKET NO.
UNIT CA' 6-
/M A'S' 8/
DATE COMPLETED BY #4 S', wor TELEPHONE tY-6 W-22 W OCNdd NA'7 MONTH DAY. AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MWe. Net)
- 3. 0 3S 37 s
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13 2
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2'9 13 3
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- 5~ [i 11 C,C y,
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12
-65 OO 29 13
-36
-l 0,O 30 14 l
-33 CJ, 5 39 15
-29 1s 1
INSTRUCTIONS On this form, list the average daily unit power level in MWe Net for each day in the reporting month. Compute to the neascst whole megawatt, lhew figures will be used to plot a graph for cach reporting month. Note that when maximum dependable capacity is u>cd for the net cleetncal rating of the umt, there may be occasions when the daily average power level exceeds the 800'A line (or the restrwted power Icvel line). In such cases. the average daily unit power output sheet should be footnoted to explain the apparent anomaly.
1.16-8 O
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APPENDIX 8 AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. SC' ?6 F UNIT DATE -<> NJ' 8 ?
COMPLETED BY #d Gmv TELEPHONE N-6Se/ 'N/
0 MEd /WP MONTH DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe Net)
(MWe-Net) 72R R 7 2 Y' 5 11 1
77 p 8 M '5 ~ *i gg 2
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'2-27 11 0- Q
- ? W. 5 2s 12
- O ~7 72d,3 29 13
-33 1
738 3 30 14
-9 N D/3 31 15 22 /. 2 to On this form, list the average daily unit power level in MWe Net for each day in the reporting m INSTRUCTIONS nearcst whole megawatt.
These figures will be used to plot a graph for each reporting month. Note that when max d h i
used for the net electrical rating of the unit. there may be occasions when the daily average powe 100'# line (or the restrwted power level line). In such cases, the aver.sge daily urut power foutnoted to explam the apparent anomaly.
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n VI. UNIQUE REPORTING REQUIREMENTS The following items are included in this report based on prior commitments to the commission:
A.
Main 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.
Unit: Two Date: October 19, 1987
\\
Valves Actuated No. & Type of Actuation 6
2-203-3B 1 Manual Plant Conditions:
Reactor Pressure - 1000 psig Description of Events:
Scrammed due to racking out Bus 23.
Wanted to control pressure by actuating the electromatic relief valve.
so they could pressure withia ?/tesonable limits.
B.
Control Rod Drive Scram Timing Data for Units One and Two There was no Control Rod Drive Scram Timing Data for Units One and Two for the reporting period.
I
~l l
0027H/0061Z J
1&
y, VII.
REFUELING INFORMATION l-The'following information about future reloads a' Quad-Cities Station was t
requested in a' January 26, 1978, licensing memorandum (78-24) from D. E.
1,
-O'Brien to C. Reed, et al.. titled "Dresden, Quad-Cities, and Zion Station--NRC Request for Refueling Information", dated January 18, 1978.
I-1 o
0027H/00612
QTP 300-S32 QUAO-CITIES REFUELING R:visicn 1 INFORMATION REQUEST March 1978 1.
Unit:
01 Reload:
8 Cycle:
9 i
i 2.
Scheduled date for next refueling shutdown:
6-24-89 3.
Scheduled date for restart following refueling:
12-18-87 4.
Will refueling or resumption of operation thereaf ter require specification change or other license amendment:
a technical AMENDMENT TO MOVE SINGLE LOOP
{
E l
TO MCPR-LIMIT AND OPERATION AT INCRE CH6NGE 5.
information:
. REDUCTION.
SUBMITTED TO NUCLEAR LICENSING AUGUST 31, 1987 FOR TRANSMITTAL TO NRC 6.
Important IIcensing considerations associated with refueling
- different fuel design or supplier, unreviewed de
, e.g., new or nce analysis e
FIRST RELOAD OF GENERAL ELECTRIC, GE8E FUEL WITH 4 WATE LIMIT OF 14.4 KW/FT.
I 7.
The number of fuel assemblies.
Number of assemblies in core:
a.
0 6.
Number of assemblies in spent fuel pool:
249
_,4 B.
The present increase in licensed storage capacity that has been in number of fuel assemblies:
anned 1.lcensed storage capacity for spent fuel:
a.
3657 b.
Planned increase in licensed storage:
0 9.
The projected date of the last spent fuel pool assuming the presentrefueling that can be discharged to the licensed capacity:
20_088 XPPROVED
, APR 2 01978 Q.C.O.S.R.
6
.. - - - - - - - - - - - - ^
o QTP 300-532 Rsvision i QUAD-CITIES REFUELING
~ March 1978 INFORMATION REQUEST 1.
Unit:
02 Reload:
8
_ Cycle:
9 2.
Scheduled date for next refueling shutdown:
3-14-88 3.
Scheduled date for restart following refueling:
s-22-88 I
fe.
Will refueling or resumption of operation thereafter requi specification change or other license amendment:
re a technical NOT AS YET DETERMINED.
5.
Scheduled date(s) for submitting proposed licensing actio
.i information:
)
n and supporting DECEMBER 14, 1987 6.
Important licensing considerations associated with refueli
- different fuel design or supplier, unreviewed ng, e.g., new or ormance analysis e ures:,
t NONE AT PRESENT TIME.
i
\\
I 7
The number of fuel assemblies.
Number of assemblies in core:
a.
724 b.
Number of assemblies in spent fuel pool:
1311 8.
The present s
increase in licensed storage capacity that has been re in number of fuel assemblies:
y or is planned 4
Licensed storage capacity for spent fuel:
{
a.
\\
3897 b.
Planned increase in licensed storage:
0 9.
The projected date of
\\
spent fuel pool assuming the presentthe last refueling that can be discharged to th licensed capacity:
2,00,8 APPROVED APR 2 01978 Q.C.O.S.R.
--_-_-_n___
?
t VIII. GLOSSARY The following abbreviations which may have been used in the Monthly Report, are defined below:
ACAD/ CAM Atmospheric Containment Atmospheric Dilution / Containment Atmospheric Monitoring ANSI American National Standards Institute APRM Average Power Range Monitor ATHS Anticipated Transient Without Scram BWR Boiling Water Reactor CRD Control Rod Drive EHC Electro-Hydraulic Control System EOF Emergency Operations facility GSEP Generating Stations Emergency Plan HEPA High-Efficiency Particulate Filter g
HPCI High Pressure Coolant Injection System 8
HRSS High Radiation Sampling System IPCLRT Integrated Primary Containment Leak Rate Test IRM Intermediate Range Monitor ISI Inservice Inspection
[
LER Licensee Event Report LLRT Local Leak Rate Test LPCI Low Pressure Coolant Injection Mode of RHRS LPRM Local Power Range. Monitor MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio MFLCPR Maximum Fraction Limiting Critical Power Ratio M?C Maximum Permissible Concentration MSIV Main Steam Isolation Valve NIOSH National Institute for Occupational Safety and Health PCI Primary Containment Isolation PCIOMR Preconditioning Interim Operating Management Recommendations RBCCW Reactor Building Closed Cooling Water System RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling System RHRS Residual Heat Removal System RPS Reactor Protecticn System RWM Rod Worth Minimizer SBGTS Standby Gas Treatment System SBLC Standby Liquid Control SOC Shutdown Cooling Mode of RHRS SOV Scram Discharge Volume SRM Source Range Monitor TBCCW Turbine Building Closed Cooling Water System TIP Traversing Incore Probe TSC Technical Support Center 0027H/00612
F'..
Of (Q O Telephone 309/654-2241 Commonwealth Edison i
ound Citles Nuclear Power Station 22710 206 Avenue North Corcova, Illinois 61242 RAR-87-52 November 4, 1987 U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attn:
J. Lieberman M/S MNBB-4100 Enclosed please find a listing of those changes, tests, and experiments completed during the month of October,1987, for Quad-Cities Station Units 1 and 2. DPR-29 and DPR-30. A summary of the safety evaluation is l
being reported in compliance with 10 CFR 50.59.
Thirty-nine copies are provided for your use.
Respectfully, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION RA.
R. A. Robe Services Superintendent vk Enclosure cc:
- 1. Johnson T. Watts /J. Galligan 0027H/00612 f}.]-l-/~lbI
o' i
SPECIAL TEST 2-78 I
j 1
[
Special test 2-78 was completed on October 21, 1987.
The purpose of this
)
[
test was to simulate a Reactor Low Water Level initiation of the High Pressure Coolant Injection (HPCI) system using the manual initial pushbuttons in order to verify that HPCI valves and equipment actuations perform as designed.
Safety Evaluation 1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because this test was performed with the reactor in a cold shutdown and therefore could not cause any actual water injection into the
)
reactor vessel.
Actuations of valves and turbine auxiliary equipment do not affect plant operation, therefore this test does not increase the probability of occurrence, or consequences of an accident.
2.
The possibility for an accident or malfunction of a different type i
4 than any previously evaluated in the Final Safety Analysis Report is not created because this test is similar to the HPCI logic test (QMS 700-4) which is performed every refueling outage and has been previously evaluated.
3.
The margin of safety, as defined in the basis for any Technical Specification is not reduced because this rest was performed during a cold shutdown condition when the HPCI system does not affect the margin of safety.
i J