ML20024A448

Monthly Operating Repts for May 1983.W/830606 Ltr

ML20024A448
Person / Time
Site:	Quad Cities
Issue date:	06/06/1983
From:	Kalivianakis N, Misak A, Weinfurter E COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
References
NJK-83-205, NUDOCS 8306170314
Download: ML20024A448 (22)

Text

_ _ _ _ _ _ _ _ . _ _ _

0 QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PEREDRMANCE REPORT MAY 1983 COMMONWEALTH EDISON COMPANY AND IOWA-ILLINOIS GAS & ELECTRIC COMPANY NRC DOCKET NOS 50-254 AND 50-265 LICENSE NOS DPR-29 AND DPR-30 8306170314 830606 g i PDR ADOCK 05000254 \

R PDR

TABLE OF CONTENTS I, Introduction II. - Summary of Operating Experience A. Unit One B, Unit Two III, 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 A, Operating Data Report B. Average Daily Unit Power Level C, Unit Shutdowns and Power Reductions VI, Unique Reporting Requirements A, Main Steam Relief Valve Operations B. Control Rod Drive Scram Timing Data VII, Refueling Infonnation VIII, Glossary.

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Q I, 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, Illinois. The Station is jointly owned by Commonwealth Edison Company and Iowa-Illinois Gas & Electric Company. The Nuclear Steam Supply Systems are General Electric Company Boiling Water Reactors. The Architect / Engineer was Sargent & Lundy, Incorporated, and the primary construction contractor was United Engireers & 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 critica11tles for Units 1 and 2 respectively were October 18, 1971, and April 26, 1972.

Commercial generation of power began on February 18, 1973 for Unit 1 and March 10, 1973 for Unit 2.

This report was complied by Becky Brown and Alex Misak, telephone number 309-654-2241, extensions 127 end 194.

II.

SUMMARY

OF OPERATING EXPERIENCE A. UNIT ONE May 1-20: : Unit One began the- month operating at full power,- and continued to operate at this level throughout the period, except for three occasions when the unit dropped load to perform weekly Turbine tests. At 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br />, on May 20, the unit dropped load 200 MWe/ hour to 250 MWe in anticipation of unit shutdown for a weekend maintenance outage. While holding load at 250 MWe, control rod drive scram. .

" timing was perforned. At 2340 hours0.0271 days <br />0.65 hours <br />0.00387 weeks <br />8.9037e-4 months <br />, control rod insertion began for unit shutdown.

May-21-31: At' 0105 hours0.00122 days <br />0.0292 hours <br />1.736111e-4 weeks <br />3.99525e-5 months <br />, on May 21, the unit was shutdown to repair a leak on the continuous Reactor head vent line.. At 1535 hours0.0178 days <br />0.426 hours <br />0.00254 weeks <br />5.840675e-4 months <br />, on May 22, control rod withdrawal began, and at 1833 hours0.0212 days <br />0.509 hours <br />0.00303 weeks <br />6.974565e-4 months <br /> the Reactor was critical. During Reactor startup, with the Reactor at 60 psig, 3

at 1937 hours0.0224 days <br />0.538 hours <br />0.0032 weeks <br />7.370285e-4 months <br />, the "E" Main Steam Electromatic ReIIef Valve inadvertently opened. The Reactor was manually scrammed, per

= p rocedu re . The relief valve, pilot valve had stuck open, and was replaced. - Repairs to the Electromatic Relief Valve were completed on May 23, and the Reactor was critical at 0355 hours0.00411 days <br />0.0986 hours <br />5.869709e-4 weeks <br />1.350775e-4 months <br />.

! At 1427 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.429735e-4 months <br />, the Genera' tor was placed on-line, and a nornal load increase was initiated.

On May 26, the unit began to experience a derating of approximate 1y' 60 MWe due to tube leaks in a "D" Feedwater Heater. .The unit continued to operate at about 770 MWe for the remainder of the month, except for two occasions when the unit dropped load for weekly Turbine tests and control rod pattern changes, i

j. B. UNIT TWO Unit Two continued to be derated throughout the month due to End of

. Cycle Fuel Depletion.. Load was held steady throughout the month until

- 0130 hours0.0015 days <br />0.0361 hours <br />2.149471e-4 weeks <br />4.9465e-5 months <br /> on May 30 when the unit dropped load to 400 MWe as requested by the Load Dispatcher due to low system demand. At 0635 hours0.00735 days <br />0.176 hours <br />0.00105 weeks <br />2.416175e-4 months <br />, the unit increased load normally until, at 1430 hours0.0166 days <br />0.397 hours <br />0.00236 weeks <br />5.44115e-4 months <br />, load was 630 MWe.

i r

...__,...___.,,__,.....,,...m _. , _ _ . . , . . . . . , . . . .

III, PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE A. Amendments to Facility License or Technical Specifications 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 for the reporting period.

D. Corrective Maintenance of Safety Related Equipnent The following represents a tabular summary of the major safety related maintenance performed on Unit One and Unit Two during the reporting period. This summary includes the following headings: Work Request Numbers, LER Numbers, Components, Cause of Malfunctions, Results and Effects on Safe Operation, and Action Taken to Prevent Repetition.

L UNIT ONE MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R. LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q26331 83-23/03L 1-203-3E Electro- Pilot valve stuck When the electromatic The pilot valve was matic open. ,went open the situation replaced l ike-for-l ike.

was identified & the Reactor scrammed.

Reactor was in Startup at 60 psig.

Q26348 83-22/03L l-1301-16 RCIC A wire was loose on When the valve was The wire was tightened inboard Steam the' indicating given an OPEN signal, and contacts cleaned.

Supply switch at the valve. it would start to open then appear to close.

Since it was an indication problem, the valve would have-operated i f requi red.

Q26092 83-20-/03L I-1601-33F The shaft of the The Drywell-Torus The valve was lubricated Suppression vacuum breaker Separation test was and the switches Chamber to Dry- was binding performed to determine adj us ted.

well Vacuum slightly. that the valve was Breaker fully closed.

Indication Switch e

?

==

UNIT- TWO MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R. LER OF ON . ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q26272 83-8/03L Southeast RHR The door's bottom The door would not The latch was Vault Submarine latch arm was bent, seal. This made one straightened.

Door preventing the loop of Containment operation of the Cooling mode of RHRS handwheel. Inoperable. The required surveillances were perforned.

Q24643 Rod Block The G8 integrated Because Unit 2 was in The G8 integrated Mon i tor circuit on.the a limiting control rod circuit was replaced

Channel 7 null sequence pattern, the channel 7 and the monitor was board was failed. RBM was jumpered to tested functionally.

provide a continuous 4

rod block until the RBM could be repaired.

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1

4 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.B.1, and 6.6.B.2. of the Technical Specifications.

UNIT ONE i Licensee Event Report Number Date Title of Occurrence 83-20/03L 5-10-83 1-1601-33F Vacuum Breaker

Lost Division ll Indication 83-21/0lT 5-20-83 Continuous Reactor Head Vent Pinhole Leak I-0214-2"-B 83-22/03L 5-23-83 1-1301-16 Valve Out of Service for Repairs 83-23/03L 5-22-83 3E Electromatic Relief Valve Stuck Open UNIT TWO 83-7/01T 5-9-33 Control Rod Drives Inoperable 83-8/03L 5-20-83 "B" RHR Loop Inoperable 4

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V. DATA TABUIATIONS The following data tabulations are presented in this report:-

A. Operating Data Report' B,. Average Daily Unit Power Level C, Unit Shutdowns 'and Power Reductions

OPERATING DATA REPORT DOCKET NO. 50-254 UNIT ONE DATEJune 06 1983 COMPLETED BYErich Weinfurter TELEPHONE 309-654-2241Xi91 OPERATING STATUS 0000 050183

1. Reporting period 2400 053183 Gross hours in reporting period: 744

2. Currently authorized power level (MWt): 2511 Max. Depend capacity (MWe-Net): 769* Design electrical rating (MWe-Net): 789

3. Power level to which restricted (if any)(MWe-Net): NA

4. Reasons for restriction (if any):

This Month Yr.to Date Cumulative

5. Number of hours reactor was critical 694.2 3415.6 78586.8

6. Reactor reserve shutdown hours 0.0 0.0 3421.9

7. Hours generator on line 682.6 3375.5 75462.2

8. Unit reserve shutdown hours. 0.0 0.0 909.2

9. Gross thernal energy generated (MWH) 1615467 8120041 154333032

~10. Gross electrical energy generated (MWH) 530944 2670536 49792417

11. Ne 1i electrical energy generated (HWH) 2515130

~

500652 46344038

12. Reactor service factor 93.3 94.3 81.1

13. Reactor availability factor 93.3 94.3 84,6

14. Unit service factor 91.8 93.2 77.9

15. Unit avo11ob111ty factor 91.8 93.2 78.8

16. Unit capacity factor (Using HDC) 87.5 90.3 62.2

17. Unit copocity factor (Using Des.MWe) 85.3 88.0 60.6

18. Unit forced outage rate 1.2 2.1 6.5

19. Shutdowns scheduled over next 6 months (Type,Date,ond Duration of each):

20. If shutdown at end of report period,estincted date o f s t ar t u p ____N A________

1

• The MDC nay be lower than 769 MWe during perleds of high anblent tenperatere due to the thernal perfernance of the spray canal.

$UN0FFICIAL COMANY NUMBERS ARE USED IN THIS REPORT

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' OPERATING DATA REPORT I DOCKET NO. 50-265 UNIT TWO DATEJune 06 1983 COMPLETED BYErich Weinferter TELEPHONE 309-654-2241X191 OPERATING STATUS 0000 050183

1. Reporting period:2400 053183 Gross hours in reporting period 744

2. Currently authorized power level (MWt): 2511 Max. Depend capacity (MWe-Net): 769* Design electrical rating (MWe-Net): 789

3. Power level to which restricted (if any)(MWe-Net): NA

4. Reasons for restriction (if any):

This Month Yr.to Date Cumulative

5. Number of hours reactor was critical 744.0 3359.0 75622.4

'6. Reactor reserve shutdown' hours 0.0 0.0 2985.8

7. Hours generator on line 744.0 3329.7 72917.8

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8. Unit ~ reserve shutdown hours. 0.0 0.0 702.9

9. Gross thermal energy generated (HWH) 1489762 7311495 151902989

10. Gross electrical energy generated (MWH) 477002 2343436 48380971 s

ii. Het ~ electrical energy generated (MWH) 445056 2195616 45379183

12. Reactor service factor 100.0 92.7 78.8

13. Reactor availability factor 'i00.0 92.7 81.9

14. Unit service factor 100.0 91.9. 76.0

15. Unit avullability factor 100.0 91.9. 76.7

16. Unit capacity factor- (Using MDC) 77.8 78.8 61.5

17. Unit copocity' factor (Using Des.MWe) 75.8 76.8 59.9

18. Unit forced outage rate 0.0 3.1 8.9

19. Shutdowns scheduled over next 6 nonths (Type,Date,and Duration of each):

20. If shutdown at end of report period,estinated date o f s t ar t u p __ NA,_____,,,

8The IOC not be lever then 769 Nie dering perleds of high adlent temperatore due

'to the thernel perfernance of the sprey canal.

8W10FFICIAL COlFAllY NUlWERS ARE USD 111 THIS REPORT

APPENDIX B AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-254 UNIT ONE DATEJune 06 1983 COMPLETED BYErich Weinfurter TELEPHONE 309-654-2241X191 MONTH Mov 1983 DAY AVERAGE DAILY POWER LEVEL DAY AVER AGE DAILY POWER LEVEL (MWe-Net) (MWe-Net)

1. 697.3 17. 774.7

2. 783.8 18. 779.2

3. 785.0 19, 780.9

4. 782.5 20. 530.2 5, 783.3 21. -7.i

6. 782.2 22. -11.4

7. 774.7 23. 104.4

8. 752.0 24. 526.0

9. 781.1 25. 702.7 10, 783.2 26. 721.6

~ii. 779.3 27. 714.0

12. 778.4 28. 734.3

13. 776.6 29. 720.7

14. 781.9 30. 723.6

15. 749.8 31. 710.8

16. 784.3 INSTRUCTIONS On this forn nearest dele,negemet list thet. eserege daily enit power level in Itle-Net fer each day in the reperting nonth.Conpete to the These figeres will be esed to plot a graph for each reporting nenth. Note that when notinen dependable copecity is esed for the net electrical rating of the enit there not be eccesiens When the delly everage power level exceeds the ill! line (or the restricted power level line),.In sich cases,the estrage dolly snit power setpet sheet shoold be feetnoted to esplein the apperent onenely

l APPENDIX B AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-265 UNIT TWO DATEJune 06 1983 COMPLETED BYErich Weinfurter TELEPHONE 309-654-2241Xi91 MONTH Mov 1983 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net)

1. 612.0 17, 594.3

2. 636.0 18. 595.0

3. 640.2 19, 594.2 4, 633.3 20. 585.5

5. 630.1 21. 577.8

6. 626.4 22. 575.5

7. 623.1 23. 579.1

8. 627.3 24. 580.0

9. 623,6 25. 574.0

10. 620.3 ,_

26, 574.5 ii. 615.3 27, 568.6

12. 615.5 28, 568.0

13. 599.7 29. 570.7

14. 606.6 30. 528.5 15, 603.3 31, 564.7

16. 601.3 INSTRUCTIONS On this fern list the overage daily snit power level in IWe-Net for each day in the reporting nenth.Conpete to the nearest whole ,nepeatt.

These figures whl be esed to plot a graph for each reporting nonth. Hete that when notinen dependable capacit, is sled for the net electrical rating of the init there may be occasions when the daily overage power level exceels the 1991 line (or the restricted power level lind,.In such cases,the overage daily snit power setpet sheet shield be Aa.tnoted to erplein the apparent onenely

M M M M M M M M M M M M M M M p ID/5A APPENDIX D QTP 300-S13 UNIT SilUTDOWNS AND POWER REDUCTIONS Revision 6 DOCKET NO. 050-254 August 1982 UNIT NAME Quad-Cities Unit 1 COMPLETED BY Alex Misak, ext 194 DATE June 6, 1983 REPORT tt0NTil MAY 1983 TELEPilONE 309-6S4-2241 m 5 g w"

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U$M LICENSEE gm m@

Wm NO. DATE w DURATION d EVENT REPORT NO.

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(IlOURS) CORRECTIVE ACTIONS /C0titlENTS ca 83-28 830501 S 0.0 B 5 HA XXXXXX Reduced load to perform weekly Turbine test e

,. 83-29 830507 5 0.0 B 5 HA XXXXXX Reduced load to perform weekly Turbine test 83-30 830515 S 0.0 B 5 HA XXXXXX Reduced load to perform weekly Turbine test 83-31 830520 s 41.47 B 2 CJ PIPEXX Unit shutdown for maintenance on con-tinuous Reactor head vent line 83-32 830522 F 8.3 B 2 CC VALV0P Unit shutdown for Relief Valve problems 83-33 830529 5 0.0 B 5 HA XXXXXX Reduced load to perform weekly Turbine test 83-34 830531 F 0.0 A 5 RB XXXXXX Reduced load to change rod pattern APPROVED AUG 101982 (final) ygg3g

M M P""1 M M M -

M M M M M M p ID/5A APPENDIX D QTP 300-S13 UNIT SilUTDOWNS AND POWER REDUCTIONS Revision 6 DOCKET NO. 050-265 August 1982 UNIT NAME Quad-Cities Unit 2 COMPLETED BY Alex Hisak, 194 DATE June 6, 1983 REPORT NONTil MAY 1983 TELEPil0NE 309-654-2241 w 5 g w" 5 $$ gm Ww w DURATION Q

d

{d LICENSEE EVENT m@

gu {Ou NO.

55g 0 v

DATE (IIOURS) o REPORT NO. CORRECTIVE ACTIONS /COMilENTS ca 83-35 830530 F 0.0 F 5 Load reduction requested by Load Dispatcher I

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APPROVED AUG 1 G 1982

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, v s VI. UNIQUE REPORTING REQUIREMENTSs -

V ,

,, The following items are included in this report based on prior -

commitments to the commission: s , , ,

s s

. s. .

A. MAIN STEAM RELIEF VALVE OPERATl0NS . -

Relief ' valve operations during'th'e reporting period are summarized s in the followingfticle.- The table includes information as to . '

'* c \

l which relie5, valve-was actuatdd, how it was actuated, and the., '

circumstances resulting in~'its actuation. <

, 1 s -

s .

VALVES NO. & TYPE PLANT \

UNIT DATE ,' ACTUATED ACTUATIONS CONDITIONS > DESCRIPTION.0F EVENTS 3

,W 1-203-3E l Manual i Rx Press Post-liaintenance - --

'i 1 ('{5-23-83 N', ' -

s 1000 (Replaced pilot valve) [s

% s B. CONTROL R0D DRIVE SCRAM TIMING DATA FOR UNIT 5 ONE AND TWO -

q~, ,

i g lhe, basis for reporting this data to the Nuc!dar Regulatory Commission are specified ~iri the surveillance 're'quirements of '

Technical Sp'eci fication 4.3'.C.1 and 4.3.C.2. '~ '1

.\ +

s 6 a.-> complete summary of Uni ts drie and' Tw\o '*

The followirig table :Is ,

Control Rod Drive Scram Timing sfor the reporting phrlod. A$1['g.

< i scram timing was performed'with Reactor pressure greater 4than

,- 800 psig.

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RESULTS OF SCRAM TIMING MEASUREMENTS PERFORMED ON UNIT 1&2 CONTROL

, ROD DRIVES, FROM l-1 TO 12-31-83 AVERAGE TIME IN SECONDS AT % Max. Time INSERTED FROM FULLY WITHDRAWN For 90%

Insertion DESCRIPTION c '; NUMBER 5 20 50 90 Technical Specification 3.3.C.I s DATE' 0F RODS 0.375 0.900 2.00 3.5 7 sec. 3.3.C.2 (Average Scram insertion Tine) 5-23 89 0.28 0.65 1.41 2.49 2.79 Uni t One Hot Scram Time "B" (H-0) Sequence

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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.

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QTP 300-S32 1 Revision 1

[' '

QUAD-CITIES REFUEllHG ttorch 1978 1

( INFORttATION REQUEST l

_: 1. Unit: Q1 Reload: 6 Cycle: 7 2.

[ 3 Scheduled date for next refueling shutdown:

Scheduled date'for restart following refueling:

9-6-82 12-18-82 -

l F

h. Will refueling or resumption of operation thereafter require a technical specification change or other IIcense amendment: Yes E

L 5 Scheduled date(s) for submitting proposed IIcensing action and supporting

. Information: 8-19-82: Te ct.. Spec. changes submitted to the NRC.

6. Important IIcensing considerations associated with refueling, e.g., new or

' different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures:

a) All Tx7 fuel assemblies will be removed from the core.

, b) MAPLHGR curves for fuel types in the core are being extended to h0,000 MWD /ST.

c) MCPR limits vill be detemined by GE's ODYN computer code.

o d) The vessel pressure safety limit is being modified to acec=modate the i~ potential for higher reactor pressures as c' alculated by ODYN.

1 ll 7 The number of fuel assemblies.

a. Number of assemblies la core: 724

b. Number of assemblies in spent fuel pool: 800 9

d 8. The present licensed spent fuel pool storage capacity and the size of any increase in licensed storage capacity that has been requested or is planned n in number of fuel assemblies:

a. Licensed storage capacity for spent fuel: 3657 3 b. Planned increase in IIcensed storage: O d

9 The projected date of the last refueling that can be discharged to the I

a spent fuel pool assuming the present licensed capacity: 2003 V WPPROVED E APR 2.01978

~

y Q.c.o.S.R.

a

E QTP 300-S32 n

. Revision 1

, QUAD-CITIES REFUELING March 1978 t ,

INFORMATION REQUEST 2, 1. Unit: Q2 Reload: 6 Cycle: 7 p 2. Scheduled date-for next refueling shutdown: 9-5-83 3 Scheduled date for restart following refueling: 11-12-83 e~

j"' 4. Will refueling or resumption of operation thereafter require a technical specification change or other license amendment:

P No, however, a change to the Technical Specifications is being submitted

{ (see below).

5 Scheduled date(s) for submitting proposed licensing action and supporting P' information:

o June 14, 1983 (scheduled) 1

! r, lI" 6. Important IIcensing considerations associated with refueling, e.g., new or

' different fuel design or supplier, unreviewed design or performance analysis

, methods, significant changes in fuel design, new operating procedures:

a) All new fuel assemblies will be of barrier design; MAPLHGR curves will l" be re-labeled to include the barrier designation.

b) The use of improved assumptions in the load reject without bypass analysis l resul ted in a much improved MCPR operating limit. Technical Specifications

'llj ' are being changed to provide this additional operating margin.

l l1'

'i a-r 7 The number of fuel assemblies.

L' a. Number of assemblies in core: 724

b. Number of assemblies in spent fuel pool: 204 e

8. The present IIcensed spent fuel pool storage capacity and the size of any increase in IIcensed storage capacity that has been requested or is planned

, in number of fuel assemblies:

a. Licensed storaga capacity for spent fuel: 3897

b. Planned increase in licensed storage: 0 9 The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 2003

(

4PPROVED

, APR 2 01973

,. Q.C.O.S.R.

1 1$

l

VIII, GLOSSARY The following abbreviations which may have been used in the Monthly Re port, are defined below:

ACAD/C AM ' -

Atmospheric Containment Atmospheric Dilution / Containment Atmospheric Monitoring ANSI -

American National Standards Institute APRM -

Average Power Range Monitor ATWS -

Anticipated Transient Without Scram BWR -

Boiling Water Reactor-CRD -

Control Rod Drive

. EHC -

Electro-Hydraulic Control System EOF -

Bnergency Operations Facility GSEP -

Generating Stations Emergency Plan HEPA -

High-Ef ficiency Particulate Filter HPCI -. High Pressure Coolant Injection System 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 Ceneration Rate MCPR~ -

Minimum Critical Power Ratio

. MFLCPR - Maximum Fraction Limiting Critical Power Ratio MPC -

Maximum Permissible Concentration MSIV ' -

Main Stean Isolation Valve L 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 Protection System RRM -

Rod Worth Minimizer SBGTS -

Standby Gas Treatment System SBLC -

Standby Liquid Control SDC - Shutdown Cooling Mode of RHRS SDV - Scram Discharge Volume

, SRM -

Source Range Monitor l TBCCW - Turbine Building Closed Cooling Water System

, TIP -

Traversing Incore Probe TSC -

Technical Support Center 1

m Commonwealth Edison O Quid Citts Nucl2tr Power Stttion gj - 22710 206 Avenue North Cordova, Illinois 61242 Telephone 309/654-2241 NJK-83-205 June 6, 1983 D1 hector, Office of Inspection & Enforcement United States Nuclear Regulatory Commission Washington, D, C. 20555 Attention: Document Control Desk Gentlemen:

Enclosed for your information is the Monthly Performance Report covering the operation of Quad-Cities Nuclear Power Station, Units One and Two, during the month of May 1983.

Very truly yours, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATTON i

d.7. h p N. J, Kalivianakis Station Superintendent bb Enclosu re i

[t\I

._ . _ _ _ _ . . _ . _ _. _ . _ _ . _ . _ . . _ _ _ . _ _ . _ _ _