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LASALLE NUCLEAR POWER STATION
LASALLE NUCLEAR POWER STATION UNIT 1 MONTHLY PERFORMANCE REPORT MARCH 1985 COIMONWEALTH EDISON COMPANY t
.                        UNIT 1 MONTHLY PERFORMANCE REPORT MARCH 1985 COIMONWEALTH EDISON COMPANY t
NRC DOCKET NO. 050-373 i
i NRC DOCKET NO. 050-373 LICENSE NO. NPF-11 i
LICENSE NO. NPF-11 i
V i
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. 8504190611 850331 DR ADOCK0500g3
8504190611 850331 ADOCK0500g3 DR


a 3 TABLE OF CONTENTS I. INTRODUCTION II. MONTHLY REPORT FOR UNIT ONE A. Sununary of Operating Experience B. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE
a 3
: 1. Amendments to Facility License or Technical Specifications
TABLE OF CONTENTS I.
: 2. Facility or Procedure Changes Requiring NRC Approval
INTRODUCTION II.
: 3. Tests and Experiments Requiring NRC Approval
MONTHLY REPORT FOR UNIT ONE A.
: 4. Corrective Maintenance of Safety Related Equipment C. LICENSEE EVENT REPORTS D. DATA TABULATIONS
Sununary of Operating Experience B.
: 1. Operating Data Report
PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE 1.
: 2. Average Daily Unit Power Level
Amendments to Facility License or Technical Specifications 2.
: 3. Unit Shutdowns and Power Reductions E. UNIQUE REPORTING REQUIREMENTS                               l
Facility or Procedure Changes Requiring NRC Approval 3.
: 1. Main Steam Relief Valve Operations
Tests and Experiments Requiring NRC Approval 4.
: 2. ECCS Sptem Outages                                   l
Corrective Maintenance of Safety Related Equipment C.
: 3. Off-Site Dose Calculation Manual Changes             ,
LICENSEE EVENT REPORTS D.
: 4. Major Changes to Radioactive Waste Treatment         '
DATA TABULATIONS 1.
l                  System i
Operating Data Report 2.
Average Daily Unit Power Level 3.
Unit Shutdowns and Power Reductions E.
UNIQUE REPORTING REQUIREMENTS l
1.
Main Steam Relief Valve Operations 2.
ECCS Sptem Outages l
3.
Off-Site Dose Calculation Manual Changes 4.
Major Changes to Radioactive Waste Treatment l
System i
l
l


I. INTRODUCTION The LaSalle Nuclear Power Station is a Two Unit Facility Located in Marseilles, Illinois. Each Unit is a Boiling Water Reactor with a designed electrical output of 1078 MWe net. The Station is owned by Comunonwealth Edison company. The Architect / Engineer was Sargent & Lundy, and the primary construction contractor was Commonwealth Edison Company.
I.
INTRODUCTION The LaSalle Nuclear Power Station is a Two Unit Facility Located in Marseilles, Illinois. Each Unit is a Boiling Water Reactor with a designed electrical output of 1078 MWe net. The Station is owned by Comunonwealth Edison company. The Architect / Engineer was Sargent & Lundy, and the primary construction contractor was Commonwealth Edison Company.
The condenser cooling method is a closed cycle cooling pond. Unit One is subject to License Number NPF-ll, issued on April 17, 1982. The date of initial criticality was June 21, 1982. Unit Two is subject to license number NPP-18, issued on December 16, 1983. The date of initial criticality was March 10, 1984.
The condenser cooling method is a closed cycle cooling pond. Unit One is subject to License Number NPF-ll, issued on April 17, 1982. The date of initial criticality was June 21, 1982. Unit Two is subject to license number NPP-18, issued on December 16, 1983. The date of initial criticality was March 10, 1984.
s This report was compiled by Richard J. Rohrer telephone number (815)357-6761, extension 575.
s This report was compiled by Richard J. Rohrer telephone number (815)357-6761, extension 575.
: 11. MONTHLY REPORT FOR UNIT ONE A. SUlWWtY OF OPERATING EXPERIENCE FOR UNIT ONE March 1-3 March 1, 0001 hours Reactor Critical at 99% Power, generator on-line.
 
11.
MONTHLY REPORT FOR UNIT ONE A.
SUlWWtY OF OPERATING EXPERIENCE FOR UNIT ONE March 1-3 March 1, 0001 hours Reactor Critical at 99% Power, generator on-line.
March 3, 1500 hours Reactor Power at 93%.
March 3, 1500 hours Reactor Power at 93%.
March 3, 1654 hours Reactor Scram, Turbine Control Valve Fast Closure due to an offsite fault on-transmission line 0101. The reactor had been critical for 64 hours and 54 minutes in March.
March 3, 1654 hours Reactor Scram, Turbine Control Valve Fast Closure due to an offsite fault on-transmission line 0101. The reactor had been critical for 64 hours and 54 minutes in March.
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March 14, 0400 hours Reactor power decreased to 74% in order to manipulate control rods.
March 14, 0400 hours Reactor power decreased to 74% in order to manipulate control rods.
March 15, 0700 hours Reactor power restored to 99%.
March 15, 0700 hours Reactor power restored to 99%.
March 21, 1317 hours Reactor scram, Low Vessel Level and High Reactor Pressure signals due i                                                              to valving error during an Envionmental Qualification Modification. The reactor was critical for a total of 451 hours and 47 minutes in March.
March 21, 1317 hours Reactor scram, Low Vessel Level and High Reactor Pressure signals due to valving error during an i
Envionmental Qualification Modification. The reactor was critical for a total of 451 hours and 47 minutes in March.
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l         Document 0043r/0005r
l Document 0043r/0005r
  , . - -    -  ,          . . , . . , . - - . . - , - . -        ---....--_-.,,--,.,,.,,,,..,.-.,n-,,_.,.
---....--_-.,,--,.,,.,,,,..,.-.,n-,,_.,.
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n.r--
,,n-,


B. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS AND SAFETY RELATED MAINTENANCE.
B.
: 1. Amendments to facility license or Technical Specification.
PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS AND SAFETY RELATED MAINTENANCE.
1.
Amendments to facility license or Technical Specification.
There were no amendments to the facility license or Technical Specification during this reporting period.
There were no amendments to the facility license or Technical Specification during this reporting period.
: 2. Facility or procedure changes requiring NRC approval.
2.
Facility or procedure changes requiring NRC approval.
There were no facility or procedure changes requiring NRC approval during this reporting period.
There were no facility or procedure changes requiring NRC approval during this reporting period.
: 3. Tests and Experiments requiring NRC approval.
3.
Tests and Experiments requiring NRC approval.
There were no tests or experiments requiring NRC, approval during this reporting period.
There were no tests or experiments requiring NRC, approval during this reporting period.
: 4. Corrective maintenance of safety related equipment.
4.
Corrective maintenance of safety related equipment.
The following table (Table 1) presents a sununary of safety-related maintenance completed on Unit One during the reporting period. The headings indicated in this summary include: Work Request number, Component Name, Cause of Malfunction, Results and Effects on Safe Operation, and Corrective Action.
The following table (Table 1) presents a sununary of safety-related maintenance completed on Unit One during the reporting period. The headings indicated in this summary include: Work Request number, Component Name, Cause of Malfunction, Results and Effects on Safe Operation, and Corrective Action.


TABLE 1
TABLE 1
                                                      ' CORRECTIVE MAINTENANCE OF
' CORRECTIVE MAINTENANCE OF SAFETY RELATED EQUIPMENT WORK REQUEST COMPONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION ON SAFE OPERATION J L44677 Main Steam High Contacts out of Adjustment Inconsistent Alarm function.
,                                                        SAFETY RELATED EQUIPMENT WORK REQUEST       COMPONENT             CAUSE OF MALFUNCTION           RESULTS AND EFFECTS             CORRECTIVE ACTION ON SAFE OPERATION J L44677           Main Steam High       Contacts out of Adjustment Inconsistent Alarm function.       Adjusted Contacts.
Adjusted Contacts.
Radiation Relay L46942           HCU Accumulator for   Leaking valve.               Potential to cause failure to   Replaced valve.
Radiation Relay L46942 HCU Accumulator for Leaking valve.
CRD 30-55.                                           scram this rod if combined with other events.
Potential to cause failure to Replaced valve.
L46947           HCU Accumulator for   Leaking Valve.               Potential to cause failure to   Replaced valve.
CRD 30-55.
CRD 34-59.                                           scram this rod if combined with other events.
scram this rod if combined with other events.
L46974           HCU Accumulator for   Leaking Valve.               Potential to cause failure to   Replaced valve.
L46947 HCU Accumulator for Leaking Valve.
CRD 58-31                                           scram this rod if combined with other events.
Potential to cause failure to Replaced valve.
L46986           HOU Accumulator for   Leaking Valve.               Potential to cause. failure to Replaced Valve.
CRD 34-59.
CRD 42-07.                                           scram this rod if combined with other events.
scram this rod if combined with other events.
L47009           Diesel Generator 0     Oil Leak.                     Possible degraded pump         Replaced Bearing.
L46974 HCU Accumulator for Leaking Valve.
Cooling Water Pump.                                 Performance.
Potential to cause failure to Replaced valve.
L47038           HCU Accumulator for   Leaking Valve                 Potential to cause failure to   Replaced valve.
CRD 58-31 scram this rod if combined with other events.
CRD 58-19                                           scram this rod if combined with other events.
L46986 HOU Accumulator for Leaking Valve.
L47219           No. 2 Turbine control I.imit Switch Arm was         Loss of indication of valve's   Replaced limit switch Valve.                 broken off.                   position,                       arm.
Potential to cause. failure to Replaced Valve.
L47274           "A" Chlorine Detector Glass orifice which           Loss of redundancy.             Replaced orifice and on "A" Control Room   supplies electrolyte                                         wick.
CRD 42-07.
HVAC.                 was clogged.
scram this rod if combined with other events.
L47009 Diesel Generator 0 Oil Leak.
Possible degraded pump Replaced Bearing.
Cooling Water Pump.
Performance.
L47038 HCU Accumulator for Leaking Valve Potential to cause failure to Replaced valve.
CRD 58-19 scram this rod if combined with other events.
L47219 No. 2 Turbine control I.imit Switch Arm was Loss of indication of valve's Replaced limit switch Valve.
broken off.
: position, arm.
L47274 "A" Chlorine Detector Glass orifice which Loss of redundancy.
Replaced orifice and on "A" Control Room supplies electrolyte wick.
HVAC.
was clogged.
DOCUMENT 0044r/0005r
DOCUMENT 0044r/0005r


TAILE 1 CORRECTIVE MAINTENANCE OF SAFETY RELATED EQUIPMENT WORK REQUEST   COMPONENT               CAUSE OF MALFUNCTION           RESULTS AND EFFECTS           CORRECTIVE ACTION ON SAFE OPERATION L47679       Division II Batteries. Excessive gaps between       Possible seismic deficiency. Installed material batteries and side rails.                                     to r2ke gaps narrower.
TAILE 1 CORRECTIVE MAINTENANCE OF SAFETY RELATED EQUIPMENT WORK REQUEST COMPONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION ON SAFE OPERATION L47679 Division II Batteries. Excessive gaps between Possible seismic deficiency.
L47691       Inlet damper to "A"   Faulty microswitch in       Actuator would not move damper. Cleaned microswitch Control Room INAC     actuator,                                                     and verified proper Return Fan.                                                                         operation.
Installed material batteries and side rails.
to r2ke gaps narrower.
L47691 Inlet damper to "A" Faulty microswitch in Actuator would not move damper. Cleaned microswitch Control Room INAC
: actuator, and verified proper Return Fan.
operation.
I DOCUMENT 0044r/0005r
I DOCUMENT 0044r/0005r


C. LICENSEE EVENT REPORTS The following is a tabular sununary of all licensee event reports .for LaSalle Nuclear Power Station, Unit One, occurring during the reporting period, March 1 through March 31, 1985. This information is provided pursuant to the reportable occurrence reporting requirements as set forth in 10CFR 50.73.
C.
Licensee Event Report Number                                 Date Title of Occurrence 85-015-00                                                 2-7-85 "C" Vacuum Breaker Cycled 85-016-00                                                 2-8-85 Unsecured High Radiation Door 85-017-00                                                 2-8-85 Unit Scram 85-018-00                                                 2-16-85 Missed Offgas Hydrogen Sample Frequency 85-019-00                                                 2-12-85 High Rad Door Unsecured 85-020-00                                                 2-14-85 Unsecured High Radiation Door 85-021-00                                                 2-20-85 Unsecured High Radiation Area Door 85-022-00                                                 2-22-85 RCIC Steam Line High Flow Isolation During Warm-up.
LICENSEE EVENT REPORTS The following is a tabular sununary of all licensee event reports.for LaSalle Nuclear Power Station, Unit One, occurring during the reporting period, March 1 through March 31, 1985. This information is provided pursuant to the reportable occurrence reporting requirements as set forth in 10CFR 50.73.
:    85-023-00                                                 2-23-85 Unsecured High Radiation Area l
Licensee Event Report Number Date Title of Occurrence 85-015-00 2-7-85 "C" Vacuum Breaker Cycled 85-016-00 2-8-85 Unsecured High Radiation Door 85-017-00 2-8-85 Unit Scram 85-018-00 2-16-85 Missed Offgas Hydrogen Sample Frequency 85-019-00 2-12-85 High Rad Door Unsecured 85-020-00 2-14-85 Unsecured High Radiation Door 85-021-00 2-20-85 Unsecured High Radiation Area Door 85-022-00 2-22-85 RCIC Steam Line High Flow Isolation During Warm-up.
l   85-024-00                                                 3-3-85 Reactor Scram l
85-023-00 2-23-85 Unsecured High Radiation Area l
85-025-00                                                 3-4-85 Drywell Vacuum Breaker Cycled l   85-026-00                                                 3-13-85 Passed Critical Date on Monthly Functional Test 85-027-00                                                 3-13-85 Unsecured High Radiation Area.
l 85-024-00 3-3-85 Reactor Scram l
85-025-00 3-4-85 Drywell Vacuum Breaker Cycled l
85-026-00 3-13-85 Passed Critical Date on Monthly Functional Test 85-027-00 3-13-85 Unsecured High Radiation Area.
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t D.               DATA TABULATIONS                                                                                                                                                                         i The following data tabulations are presented in this report:
t D.
: 1.           Operating Data Report
DATA TABULATIONS i
: 2.           Average Daily Unit Power Level
The following data tabulations are presented in this report:
: 3.           Unit Shutdowns and Power Reductions 3
1.
Operating Data Report 2.
Average Daily Unit Power Level 3.
Unit Shutdowns and Power Reductions 3
1 6
1 6
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: 1. OPERATING DATA REPORT DOCKET NO. 050-373 UNIT LaSalle One DATE April 10, 1985 COMPLETED BY Richard J. Rohrer TELEPHONE (815)357-6761 OPERATING STATUS
 
: 1. REPORTINJ PERIOD: March, 1985     GROSS HOURS IN REPORTING PERIOD: 744
1.
: 2. CURRENTLY AUTHORIZED POWER LEVEL (MWt):3323 MAX DEPEND CAPACITY (IWe-Net): 1036 DESIGN ELECTRICAL RATING (MWe-Net):1078
OPERATING DATA REPORT DOCKET NO. 050-373 UNIT LaSalle One DATE April 10, 1985 COMPLETED BY Richard J. Rohrer TELEPHONE (815)357-6761 OPERATING STATUS 1.
: 3. POWER LEVEL TO WHICH RESTRICTED (IP'ANY) (MWe-Net):                       N/A
REPORTINJ PERIOD: March, 1985 GROSS HOURS IN REPORTING PERIOD: 744 2.
: 4. REASONS FOR RESTRICTION (IF ANY):N/A THIS' MONTH YR TO DATE                   CUMULATIVE 5     NUMBER OF HOURS REACTOR WAS CRITICAL   451.8       1770.1                       8051
CURRENTLY AUTHORIZED POWER LEVEL (MWt):3323 MAX DEPEND CAPACITY (IWe-Net): 1036 DESIGN ELECTRICAL RATING (MWe-Net):1078 3.
: 6. REACTOR RESERVE SHUTDOWN HOURS           41.5               71.0                 1237
POWER LEVEL TO WHICH RESTRICTED (IP'ANY) (MWe-Net):
: 7. HOURS GENERATOR ON LINE                 443.7       1721.3                       7777
N/A 4.
: 8. UNIT RESERVE SHUTDOWN HOURS               0.0               0.0                   0.0
REASONS FOR RESTRICTION (IF ANY):N/A THIS' MONTH YR TO DATE CUMULATIVE 5
: 9. GROSS THERMAL ENERGY GENERATED (MWH)   1339056'   4841667                     21664956
NUMBER OF HOURS REACTOR WAS CRITICAL 451.8 1770.1 8051 6.
: 10. GROSS ELEC. ENERGY GENERATED (MWH)     449172     1604050                     7074693
REACTOR RESERVE SHUTDOWN HOURS 41.5 71.0 1237 7.
: 11. NET ELEC. ENERGY GENERATED (MWH)         431306     1541242                     6736304
HOURS GENERATOR ON LINE 443.7 1721.3 7777 8.
: 12. . REACTOR SERVICE FACTOR                   60.7%     81.8%                       73.4%
UNIT RESERVE SHUTDOWN HOURS 0.0 0.0 0.0 9.
: 13. REACTOR AVAILABILITY FACTOR                 66.3%     84.3%                       84.7%
GROSS THERMAL ENERGY GENERATED (MWH) 1339056' 4841667 21664956 10.
: 14. UNIT SERVICE FACTOR                         59.6%     78.8%                       70.9%
GROSS ELEC. ENERGY GENERATED (MWH) 449172 1604050 7074693
: 15. UNIT AVAILABILITY FACTOR                   59.6%     78.8%                       70.9%
: 11. NET ELEC. ENERGY GENERATED (MWH) 431306 1541242 6736304
: 16. UNIT CAPACITY FACTOR (USING MDC)           56.0%     68.1%                       59.3%
: 12.. REACTOR SERVICE FACTOR 60.7%
: 17. UNIT CAPACITY FACTOR (USING DESIGN MWe)                                   53.8%       65.5%                       57.0%
81.8%
: 18. UNIT FORCED OUTAGE RATE                 40.4%       20.1%                       13.8%
73.4%
: 19. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, AND DURATION OF EACH)
13.
        -Unit one is scheduled for a refueling, maintenance, modification, and surveillance outage beginning September 3, 1985 and lasting 26 weeks.
REACTOR AVAILABILITY FACTOR 66.3%
: 20. IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP:                     4-5-85.
84.3%
84.7%
14.
UNIT SERVICE FACTOR 59.6%
78.8%
70.9%
15.
UNIT AVAILABILITY FACTOR 59.6%
78.8%
70.9%
16.
UNIT CAPACITY FACTOR (USING MDC) 56.0%
68.1%
59.3%
17.
UNIT CAPACITY FACTOR (USING DESIGN MWe) 53.8%
65.5%
57.0%
18.
UNIT FORCED OUTAGE RATE 40.4%
20.1%
13.8%
19.
SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, AND DURATION OF EACH)
-Unit one is scheduled for a refueling, maintenance, modification, and surveillance outage beginning September 3, 1985 and lasting 26 weeks.
20.
IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP:
4-5-85.
Document-0043r/0005r
Document-0043r/0005r
: 2. AVERAGE DAILY UNIT POWER LEVEL DOCKET NO: 050-373 UNIT: LASALLE OWE DATE: April 10, 1985 COMPLETED BY: Richard J. Rohrer TELEPHONE: (815) 357-6761 MONTH: MARCH, 1985 DAY AVERAGE DAILY POWER LEVEL     DAY AVERAGE DAILY POWER LEVEL (MWe-Net)                           (MWe-Net)
: 2. AVERAGE DAILY UNIT POWER LEVEL DOCKET NO: 050-373 UNIT: LASALLE OWE DATE: April 10, 1985 COMPLETED BY: Richard J. Rohrer TELEPHONE: (815) 357-6761 MONTH: MARCH, 1985 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
: 1.           1068             17.             1086
(MWe-Net) 1.
: 2.           1069 '           18.             1083 3             730             19.             1075
1068 17.
: 4.           -18             20.             1068
1086 2.
: 5.             15           21.               584
1069 '
: 6.           473             22.               -15                 f
18.
: 7.           797             23.               -15
1083 3
: 8.           856             24.               -15
730 19.
: 9.           1050             25.               -14
1075 4.
: 10.           1064             26.               -15
-18 20.
: 11.           1004             27.               -14
1068 5.
: 12.           1053             28.               -15
15 21.
: 13.           1000             29.               -15
584 f
: 14.           975             30.               -15
6.
: 15.           1072             31.               -15
473 22.
: 16.           1009
-15 7.
797 23.
-15 8.
856 24.
-15 9.
1050 25.
-14 10.
1064 26.
-15 11.
1004 27.
-14 12.
1053 28.
-15 13.
1000 29.
-15 14.
975 30.
-15 15.
1072 31.
-15 16.
1009


ATTACHMENT E
ATTACHMENT E
: 3. ~ UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 050-373 UNIT NAME LaSalle One DATE March 10i 1985 REPORT MONTH February 1985                   COMPLETED BY Richard J. Rohrer TELEPHONE (815)357-6761 METHOD OF TYPE                                             SHUTTING DOWN F: FORCED       DURATION                             THE REAC'!OR OR       CORRECTIVE NOo                                   DATE   S: SCHEDULED     (HOURS)           REASON           REDUCING POWER       ACTIONS /COf9 TENTS 5                                     850303 P               49.60             A                 3                     Reactor Scram on Turbine Control Valve Fast Closure due to an offsite fault on transmission line 0101.
: 3. ~ UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 050-373 UNIT NAME LaSalle One DATE March 10i 1985 REPORT MONTH February 1985 COMPLETED BY Richard J. Rohrer TELEPHONE (815)357-6761 METHOD OF TYPE SHUTTING DOWN F: FORCED DURATION THE REAC'!OR OR CORRECTIVE NOo DATE S: SCHEDULED (HOURS)
6                                   850321 F               250.72             G                 3                     Reactor Scram on Low Vessel Level and High Vessel Pressure Signals due to a valving error during installation of an Environmental Qua1ification Modifiction.
REASON REDUCING POWER ACTIONS /COf9 TENTS 5
850303 P
49.60 A
3 Reactor Scram on Turbine Control Valve Fast Closure due to an offsite fault on transmission line 0101.
6 850321 F
250.72 G
3 Reactor Scram on Low Vessel Level and High Vessel Pressure Signals due to a valving error during installation of an Environmental Qua1ification Modifiction.
DOCUMENT 0044r/0005
DOCUMENT 0044r/0005


E. UNIQUE REPORTING REQUIREMENTS
E.
UNIQUE REPORTING REQUIREMENTS
: 1. Safety / Relief valve operations for Unit One.
: 1. Safety / Relief valve operations for Unit One.
VALVES             NO & TYPE         PLANT           DESCRIPTION DATE             ACTUATED           ACTUATION         CONDITION       OF EVENT 3-3-85           IB21-F013E 1                           1075 psig     Actuated to relieve pressure following reactor scram.
VALVES NO & TYPE PLANT DESCRIPTION DATE ACTUATED ACTUATION CONDITION OF EVENT 3-3-85 IB21-F013E 1 1075 psig Actuated to relieve pressure following reactor scram.
3-3-85           IB21-F013U 1                           1075 psig     Actuated to relieve pressure following reactor scram.
3-3-85 IB21-F013U 1 1075 psig Actuated to relieve pressure following reactor scram.
3-3-85           IB21-F013S 1                           1075 psig     Actuated to relieve pressure following reactor scram.
3-3-85 IB21-F013S 1 1075 psig Actuated to relieve pressure following reactor scram.
3-21-85         IB21-F0138 1                         1060 psig       Actuated following reactor scram.
3-21-85 IB21-F0138 1 1060 psig Actuated following reactor scram.
3-27-85         IB21-F013S 3                         510 psig       Unitentional actuations during pressure test of instrument lines following installation of an Environmental Qualification Modification.
3-27-85 IB21-F013S 3 510 psig Unitentional actuations during pressure test of instrument lines following installation of an Environmental Qualification Modification.
3-27-85         IB21-F013J 1                         510 psig         Unintentional Actuation during pressure test of instrument lines following installation of an Environmental Qualification Modification.
3-27-85 IB21-F013J 1 510 psig Unintentional Actuation during pressure test of instrument lines following installation of an Environmental Qualification Modification.
3-27-85         IB21-F013B 1                         510 psig         Unintentional Actuation during pressure test of l                                                                       instrument. lines following installation of an Environmental Qualification Modification.
3-27-85 IB21-F013B 1 510 psig Unintentional Actuation during pressure test of l
instrument. lines following installation of an Environmental Qualification Modification.
l l
l l
Document 0043r/0005r i
Document 0043r/0005r i
: 2. ECCS Systems Outtgaa The following outages were taken on ECCS Systems during the reporting period.
OUTAGE NO.                  EOUIPMENT                PURPOSE OF OUTAGE l-196-85                    1DOO3P                  Lubrication 1-206-85                    LPCS Pump Discharge      Environmental Qualification Flow Switch              Modification (EQ Mod)
    .1-207-85                    LPCS Pump Discharge      EQ Mod.
Flow Switch 1-208-85                    LPCS Pump ADS            EQ Mod Permissive Switch 1-209-85                    LPCS Pump ADS            EQ Mod Permissive Switch 1-210-85                    RHR Shutdown Cooling    EQ Mod 135 PSIG Pressure Switch 1-211-85                    RHR Shutdown Cooling    EQ Mod 135 PSIG Presure Switch 1-213-85                    RCIC Low Suction        EQ Mod Pressure Switch 1-214-85                    1E51-N022A,B,C,D        EQ Mod 1-215-85                    1E21-F011                Technical Specification Requirement 1-216-85                    RCIC System              Telchnical Specification Requirement 1-220-85                    1E51-N006                EQ Mod 1-224-85                    IB Diesel Generator      Lubrication Fuel Oil Transfer Pump l-230-85                    1E22-N012                EQ Mod 1-231-85                    1E22-N012                EQ Mod 1-232-85                    1E22-N006                EQ Mod 1-233-85                    1822-N006                EQ Mod Document 0043r/0005r


OUTAGE NO-             EQUIPMENT                               PURPOSE OF OUTAGE 1-247-85             HPCS pump tad 1B                         Modification Diesel Generator 1-249-85             "B" RHR Pump Running                     EQ Mod ADS Permissive Switch 1-250-85             "B" RHR Pump Running                     EQ Mod ADS Permissive Switch 1-251-85             "C" RHR Pump Running                     EQ Mod
2.
,                                ADS Permissive Switch 1-252-85             "C" RHR Pump Running                     EQ Mod ADS Permissive Switch 1-270-85             IE12-N010A                               EQ Mod 1-271-85             lE12-N010A                               EQ Mod 1-272-85             1812-F064A                               Technical Specification Requirement 1-275-85             lE12-N010C                               EQ Mod 1-276-85             lE12-N010C                               EQ Mod 1-277-85             1E12-F064C                               Technical Specification Requirement 1-283-85             lE12-N010B                               EQ Mod 1-284-85             lE12-N010B                               EQ Mod t         1-285-85             1E12-F064B                               Technical Specification Requirement 1-308-85             1A Diesel Generator                     Repair Motor Cooling Water Pump i
ECCS Systems Outtgaa The following outages were taken on ECCS Systems during the reporting period.
: 3. Off-Site Dose Calculation Manual There were no changes to the off-site dose calculations manual during this reporting period.
OUTAGE NO.
: 4. Radioactive Waste Treatment Systems.
EOUIPMENT PURPOSE OF OUTAGE l-196-85 1DOO3P Lubrication 1-206-85 LPCS Pump Discharge Environmental Qualification Flow Switch Modification (EQ Mod)
Several Procedures were revised to allow solidification of spent resin using a 20% Free-Standing Water formula. Previously,
.1-207-85 LPCS Pump Discharge EQ Mod.
,              a 10% Free-Standing Water Formula was used.
Flow Switch 1-208-85 LPCS Pump ADS EQ Mod Permissive Switch 1-209-85 LPCS Pump ADS EQ Mod Permissive Switch 1-210-85 RHR Shutdown Cooling EQ Mod 135 PSIG Pressure Switch 1-211-85 RHR Shutdown Cooling EQ Mod 135 PSIG Presure Switch 1-213-85 RCIC Low Suction EQ Mod Pressure Switch 1-214-85 1E51-N022A,B,C,D EQ Mod 1-215-85 1E21-F011 Technical Specification Requirement 1-216-85 RCIC System Telchnical Specification Requirement 1-220-85 1E51-N006 EQ Mod 1-224-85 IB Diesel Generator Lubrication Fuel Oil Transfer Pump l-230-85 1E22-N012 EQ Mod 1-231-85 1E22-N012 EQ Mod 1-232-85 1E22-N006 EQ Mod 1-233-85 1822-N006 EQ Mod Document 0043r/0005r
 
OUTAGE NO-EQUIPMENT PURPOSE OF OUTAGE 1-247-85 HPCS pump tad 1B Modification Diesel Generator 1-249-85 "B" RHR Pump Running EQ Mod ADS Permissive Switch 1-250-85 "B" RHR Pump Running EQ Mod ADS Permissive Switch 1-251-85 "C" RHR Pump Running EQ Mod ADS Permissive Switch 1-252-85 "C" RHR Pump Running EQ Mod ADS Permissive Switch 1-270-85 IE12-N010A EQ Mod 1-271-85 lE12-N010A EQ Mod 1-272-85 1812-F064A Technical Specification Requirement 1-275-85 lE12-N010C EQ Mod 1-276-85 lE12-N010C EQ Mod 1-277-85 1E12-F064C Technical Specification Requirement 1-283-85 lE12-N010B EQ Mod 1-284-85 lE12-N010B EQ Mod t
1-285-85 1E12-F064B Technical Specification Requirement 1-308-85 1A Diesel Generator Repair Motor Cooling Water Pump i
3.
Off-Site Dose Calculation Manual There were no changes to the off-site dose calculations manual during this reporting period.
4.
Radioactive Waste Treatment Systems.
Several Procedures were revised to allow solidification of spent resin using a 20% Free-Standing Water formula. Previously, a 10% Free-Standing Water Formula was used.
i Document 0043r/0005r
i Document 0043r/0005r


LASALLE NUCLEAR POWER STATION UNIT 2 MONTHLY PERFORMANCE REPORT MARCH 1985 COOGONWEALTH EDISON COMPANY NRC DOCKET NO. 050-374                                                                                   ,
LASALLE NUCLEAR POWER STATION UNIT 2 MONTHLY PERFORMANCE REPORT MARCH 1985 COOGONWEALTH EDISON COMPANY NRC DOCKET NO. 050-374 LICENSE NO. NPF-18 i
LICENSE NO. NPF-18 i
l 1
l 1
5 DOCUMENT ID 0036r/0005r
5 DOCUMENT ID 0036r/0005r


TABLE OF CONTENTS I.       INTRODUCTION II. MONTHLY REPORT FOR UNIT TWO A.               Sunnary of Operating Experience B.             PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE
TABLE OF CONTENTS I.
: 1.             Amendments to Facility License or Technical Specifications
INTRODUCTION II.
: 2.               Facility or Procedure Changes Requiring NRC Approval
MONTHLY REPORT FOR UNIT TWO A.
: 3.             Tests and Experiments Requiring NRC Approval
Sunnary of Operating Experience B.
: 4.               Corrective Maintenance of Safety Related Equipment C.               LICENSEE EVENT REPORTS D.               DATA TABULATIONS
PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE 1.
: 1.             Operating Data Report
Amendments to Facility License or Technical Specifications 2.
: 2.               Average Daily Unit Power Level
Facility or Procedure Changes Requiring NRC Approval 3.
: 3.               Unit Shutdowns and Power Reductions E.               UNIQUE REPORTING REQUIREMENTS
Tests and Experiments Requiring NRC Approval 4.
: 1.             Safety / Relief Valve Operations
Corrective Maintenance of Safety Related Equipment C.
: 2.             ECCS System Outages
LICENSEE EVENT REPORTS D.
: 3.             Off-Site Dose calculation Manual Changes
DATA TABULATIONS 1.
: 4.               Major Changes to Radioactive Waste Treatment System l
Operating Data Report 2.
Average Daily Unit Power Level 3.
Unit Shutdowns and Power Reductions E.
UNIQUE REPORTING REQUIREMENTS 1.
Safety / Relief Valve Operations 2.
ECCS System Outages 3.
Off-Site Dose calculation Manual Changes 4.
Major Changes to Radioactive Waste Treatment System l
l 4
l 4
l DOCUMENT ID 0036r/0005r
l DOCUMENT ID 0036r/0005r


I. INTRODUCTION The LaSalle Nuclear Power Station is a Two Unit Facility Located in Marseilles, Illinois. Each Unit is a Boiling Water Reactor with a designed electrical output of 1078 MWe net. The Station is owned by Commonwealth Edison company. The Architect / Engineer was Sargent & Lundy, and the primary construction contractor was Commonwealth Edison Company.
I.
INTRODUCTION The LaSalle Nuclear Power Station is a Two Unit Facility Located in Marseilles, Illinois. Each Unit is a Boiling Water Reactor with a designed electrical output of 1078 MWe net. The Station is owned by Commonwealth Edison company. The Architect / Engineer was Sargent & Lundy, and the primary construction contractor was Commonwealth Edison Company.
The condenser cooling method is a closed cycle cooling pond. Unit One is subject to License Number NPF-ll, issued on April 17, 1982. The unit commenced commercial generation of power on January 1, 1984. Unit Two is subject to license number NPF-18, issued on December 16, 1983. The date of initial criticality was March 10, 1984.
The condenser cooling method is a closed cycle cooling pond. Unit One is subject to License Number NPF-ll, issued on April 17, 1982. The unit commenced commercial generation of power on January 1, 1984. Unit Two is subject to license number NPF-18, issued on December 16, 1983. The date of initial criticality was March 10, 1984.
This report was compiled by Richard J. Rohrer,, telephone number (815)357-6761, extension 575.
This report was compiled by Richard J. Rohrer,, telephone number (815)357-6761, extension 575.
DOCUMENT ID 0036r/0005r i
DOCUMENT ID 0036r/0005r i


II. MONTHLY REPORT FOR UNIT TWO A. SUlttARY OF OPERATING EXPERIENCE FOR UNIT TWO MARCH 1       The. reactor was subcritical, and the generator was off-line for the entire month of March. Unit Two is in a schedled outage for maintenance, testing, and modifications.
II.
MONTHLY REPORT FOR UNIT TWO A.
SUlttARY OF OPERATING EXPERIENCE FOR UNIT TWO MARCH 1 The. reactor was subcritical, and the generator was off-line for the entire month of March. Unit Two is in a schedled outage for maintenance, testing, and modifications.
1 DOCUMENT ID 0036r/0005r
1 DOCUMENT ID 0036r/0005r


a B. PLANT OR PROCEDURE CHANGES, TESTS,~ EXPERIMENTS AND SAFETY RELATED MAINTENANCE.
a B.
: 1. Amendments to facility license or Technical Specifications.
PLANT OR PROCEDURE CHANGES, TESTS,~ EXPERIMENTS AND SAFETY RELATED MAINTENANCE.
1.
Amendments to facility license or Technical Specifications.
Amendment 8 was made to the Unit Two Operating license. This Amendment extends the deadline for full compliance to 10CFR50.49 from March 31, 1985 to November 30, 1985.
Amendment 8 was made to the Unit Two Operating license. This Amendment extends the deadline for full compliance to 10CFR50.49 from March 31, 1985 to November 30, 1985.
: 2. Facility or procedure changes requiring NRC approval.
2.
Facility or procedure changes requiring NRC approval.
There were no facility or procedure changes requiring NRC approval during the reporting period.
There were no facility or procedure changes requiring NRC approval during the reporting period.
: 3. Tests and experiments requiring NRC approval.
3.
Tests and experiments requiring NRC approval.
There were no tests or experiments requiring NRC approval during the reporting period.
There were no tests or experiments requiring NRC approval during the reporting period.
: 4. Corrective Maintenance of Safety Related Equipment.
4.
Corrective Maintenance of Safety Related Equipment.
The following table (Table 1) presents a summary of safety-related maintenance completed on Unit one during the reporting period. The headings indicated in this summary include: Work Request number, Component Name, cause of malfunction, results and effects on safe operation, and corrective action.
The following table (Table 1) presents a summary of safety-related maintenance completed on Unit one during the reporting period. The headings indicated in this summary include: Work Request number, Component Name, cause of malfunction, results and effects on safe operation, and corrective action.
i l         DOCUMENT ID 0036r/0005r
i l
DOCUMENT ID 0036r/0005r


T TABLE 1 i
T TABLE 1 CORRECTIVE MAINTENANCE OF i
CORRECTIVE MAINTENANCE OF SAFETY RELATED EQUIPMENT WORK REQUEST     COMPONENT                         CAUSE OF MALFUNCTION             RESULTS AND EFFECTS             CORRECTIVE ACTION ON SAFE OPERATION
SAFETY RELATED EQUIPMENT WORK REQUEST COMPONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION ON SAFE OPERATION
, L37828           Tailpipe temperature             Faulty thermocouple.         Sporadic indication.             Replaced thermocouple.
, L37828 Tailpipe temperature Faulty thermocouple.
Sporadic indication.
Replaced thermocouple.
element for SRV "S".
element for SRV "S".
L40429         HCU Accumulator for               Leaking disconnect           Potential to cause failure to     Replaced disconnect
L40429 HCU Accumulator for Leaking disconnect Potential to cause failure to Replaced disconnect CRD 22-55.
:                  CRD 22-55.                       fitting.                     scram this rod if combined with fitting.                   ,
fitting.
,                                                                                  other events.
scram this rod if combined with fitting.
L45891         Suppression Pool                 Faulty lug on Control         Erratic readings.                 Replaced faulty lug.
other events.
Water Temperature                 Center element.
L45891 Suppression Pool Faulty lug on Control Erratic readings.
L46488         Division II                       Instrument out of             Incorrect indication.             Recalibrated Instrument.
Replaced faulty lug.
Suppression Chamber               calibration. Reading Pressure Indication.             +1.5 psig.
Water Temperature Center element.
3 L46608           Refrigeration Unit               Oil Level too high.           Refrigeration Unit periodically Lowered oil level.
L46488 Division II Instrument out of Incorrect indication.
for "B" Auxiliary                                               tripped on high oil temperature.
Recalibrated Instrument.
Electrical Equipment                                           Possible inadequate ventilation.
Suppression Chamber calibration. Reading Pressure Indication.
i                 Room HVAC.
+1.5 psig.
l i L46824         HCU Accumulator for               Leaking Valve.               Potential to cause failure to     Replaced valve.
3 L46608 Refrigeration Unit Oil Level too high.
CRD 26-35                                                       scram this rod if combined with
Refrigeration Unit periodically Lowered oil level.
                                                            .                      other events.
for "B" Auxiliary tripped on high oil temperature.
l L46832 HCU Accumulator for               Leaking Valve.               Potential.to cause failure to     Replaced valve.
Electrical Equipment Possible inadequate ventilation.
CRD 26-39.                                                     scram this rod if combined with other events.
i Room HVAC.
-f L46837         Electrical Power                 Faulty Card.                 Non-repeatable readings for       Removed card from Unit Monitoring Card for                                             overvoltage trip setpoint.       I and installed in Unit RPS System.                                                                                       2.
l i L46824 HCU Accumulator for Leaking Valve.
L46870         Diesel Generator "2B"             Air supply lines reversed. Loss of Redundancy.                 Corrected arrangement Air Start Motor.                                                                                 of air supply lines.
Potential to cause failure to Replaced valve.
l l
CRD 26-35 scram this rod if combined with other events.
4
l L46832 HCU Accumulator for Leaking Valve.
,  DOCUMENT 0044r/0005r i
Potential.to cause failure to Replaced valve.
CRD 26-39.
scram this rod if combined with other events.
-f L46837 Electrical Power Faulty Card.
Non-repeatable readings for Removed card from Unit Monitoring Card for overvoltage trip setpoint.
I and installed in Unit RPS System.
2.
L46870 Diesel Generator "2B" Air supply lines reversed. Loss of Redundancy.
Corrected arrangement l
Air Start Motor.
of air supply lines.
l 4
DOCUMENT 0044r/0005r i


s TABLE 12 CORRECTIVE MAINTENANCE OF SAFETY RELATED BQUIPMENT WORK REQUEST     COMPONENT             CAUSE OF MALFUNCTION             RESULTS AND EFFECTS                       CORRECTIVE ACTION ON SAFE OPERATION L46886       HCU Accumulator for     Leaking Valve.               Potential to cause failure to               Replaced valve.
s TABLE 12 CORRECTIVE MAINTENANCE OF SAFETY RELATED BQUIPMENT WORK REQUEST COMPONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION ON SAFE OPERATION L46886 HCU Accumulator for Leaking Valve.
CRD 26-43.                                           scram this rod if combined with-                                                                   '
Potential to cause failure to Replaced valve.
other events.
CRD 26-43.
L46889       Division II 125 Volt   Faulty input card.           Relay chattered in and out.                 Replaced input card.
scram this rod if combined with-other events.
Breaker Trip Alarm                                   No alarm was indicated.
L46889 Division II 125 Volt Faulty input card.
Relay chattered in and out.
Replaced input card.
Breaker Trip Alarm No alarm was indicated.
Relay.
Relay.
L42893       RHR Temperature         Dirty Selector Switch,     - Sporadically printed upscale.               Cleaned selector switch, Recorder.               faulty ink wheel,                                                         replaced ink wheel.
L42893 RHR Temperature Dirty Selector Switch,
L46927       HCU Accumulator for. Leaking Valve.               Potential to cause failure to             ' Replaced valve.
- Sporadically printed upscale.
CRD 14-55.                                           scram this rod if' combined                                     ,
Cleaned selector switch, Recorder.
with other events.
faulty ink wheel, replaced ink wheel.
L47073       Electrical Power       Faulty Breaker.               Breaker would not reclose.                 Installed Breaker Monitoring Breaker                                                                               removed from unit one.
L46927 HCU Accumulator for.
Leaking Valve.
Potential to cause failure to
' Replaced valve.
CRD 14-55.
scram this rod if' combined with other events.
L47073 Electrical Power Faulty Breaker.
Breaker would not reclose.
Installed Breaker Monitoring Breaker removed from unit one.
for RPS.
for RPS.
L47300       Main Steam Isolation   Torque switch out of         Valve would not cycle.                     Adjusted torque switch.                               t Valve Leakage Control Adjustment.
L47300 Main Steam Isolation Torque switch out of Valve would not cycle.
Adjusted torque switch.
t Valve Leakage Control Adjustment.
Inboard Valve.
Inboard Valve.
L47377       HPCS Full Flow Test     Broken lug at Motor Control Valve could not be cycled                     Replaced lugs on motor Valve.                 Center.                       remotely.                                 connection.
L47377 HPCS Full Flow Test Broken lug at Motor Control Valve could not be cycled Replaced lugs on motor Valve.
L47440       2B RHR Service Water   Shorted motor.               Pump Feed Breaker Tripped. Pump Replaced Motor.                                                   '
Center.
Pump.                                                 Inoperable.                                                                                       ;
remotely.
L47557       "2B" Diesel Generator Ground on Bridge.               Loss of position indication.               Repaired ground.
connection.
L47440 2B RHR Service Water Shorted motor.
Pump Feed Breaker Tripped. Pump Replaced Motor.
Pump.
Inoperable.
L47557 "2B" Diesel Generator Ground on Bridge.
Loss of position indication.
Repaired ground.
Fuel Oil Transfer Pump.
Fuel Oil Transfer Pump.
L47640       Unit Two Hydrogen       Dirty Contacts.               No indication of " closed"                 Cleaned contacts.                                     j Recombiner Cross-tie                                 position.
L47640 Unit Two Hydrogen Dirty Contacts.
* valve.
No indication of " closed" Cleaned contacts.
L47719       Division I Battery     Faulty Connections on         Periodic Tripping of breaker.               Replaced breaker and Charger.               line side of breaker.                                                     breaker trip unit.
j Recombiner Cross-tie position.
valve.
L47719 Division I Battery Faulty Connections on Periodic Tripping of breaker.
Replaced breaker and Charger.
line side of breaker.
breaker trip unit.


C. LICENSEE EVENT REPORTS The following is a tabular summary of all licensee event reports for LaSalle Nuclear Power Station, Unit Two, occurring during the reporting period, March 1 through March 31, 1985. This information is provided pursuant to the reportable occurrence reporting requirements as set forth in 10CFR 50.73.
C.
Licensee Event Report Number     Date   Title of Occurrence 85-008-00                       2-13-85   Ammonia Detector Actuation 85-009-00                       3-1-85   Reactor Water Cleanup Differential Flow Isolation when Reactor Vessel Head Vent Opened.
LICENSEE EVENT REPORTS The following is a tabular summary of all licensee event reports for LaSalle Nuclear Power Station, Unit Two, occurring during the reporting period, March 1 through March 31, 1985. This information is provided pursuant to the reportable occurrence reporting requirements as set forth in 10CFR 50.73.
85-010-00                       2-15-85   Hydraulic Control Unit Accumulator Pressure Switch Failures 85-011-00                       2-26-85   . Failure of Type C Leak Rate Test DOCUMENT ID 0036r/0005r
Licensee Event Report Number Date Title of Occurrence 85-008-00 2-13-85 Ammonia Detector Actuation 85-009-00 3-1-85 Reactor Water Cleanup Differential Flow Isolation when Reactor Vessel Head Vent Opened.
85-010-00 2-15-85 Hydraulic Control Unit Accumulator Pressure Switch Failures 85-011-00 2-26-85
. Failure of Type C Leak Rate Test DOCUMENT ID 0036r/0005r


D.               DATA TABULATIONS The following data tabulations are presented in this report:
D.
DATA TABULATIONS The following data tabulations are presented in this report:
: 1. Operating Data Report
: 1. Operating Data Report
: 2. Average Daily Unit Power Level
: 2. Average Daily Unit Power Level
Line 280: Line 438:
i i
i i
8 DOCUMENT ID 0036r/0005r
8 DOCUMENT ID 0036r/0005r
: 1. OPERATING DATA REPORT DOCKET NO. 050-374 UNIT LaSalle Two DATE April 10, 1985
: 1. OPERATING DATA REPORT DOCKET NO. 050-374 UNIT LaSalle Two DATE April 10, 1985 COMP'.ETED BY Richard J. Rohrer TEr.EPHONE (815)357-6761 OPERATING STATUS 1.
,                                                      COMP'.ETED BY Richard J. Rohrer TEr.EPHONE (815)357-6761 OPERATING STATUS
REPORTING PERIOD: March, 1985 GROSS HOURS IN REPORTING PERIOD: 744 _
: 1. REPORTING PERIOD: March, 1985 GROSS HOURS IN REPORTING PERIOD: 744 _
2.
: 2. CURRENTLY AUTHORIZED POWER LEVEL (MWt):3323 MAX DEPEND CAPACITY (MWe-Net): 1036   DESIGN ELECTRICAL RATING (MWe-Net):1078
CURRENTLY AUTHORIZED POWER LEVEL (MWt):3323 MAX DEPEND CAPACITY (MWe-Net): 1036 DESIGN ELECTRICAL RATING (MWe-Net):1078 3.
: 3. POWER LEVEL TO WHICH RESTRICTED (IP ANY) (MWe-Net):         N/A
POWER LEVEL TO WHICH RESTRICTED (IP ANY) (MWe-Net):
: 4. REASONS FOR RESTRICTION (IF ANY):
N/A 4.
THIS MONTH YR TO DATE     CUMULATIVE 5     NUMBER OF HOURS REACTOR WAS CRITICAL     0.0         1399.8       3011.6
REASONS FOR RESTRICTION (IF ANY):
: 6. REACTOR RESERVE SHUTDOWN HOURS           0.0       0.0           125.3
THIS MONTH YR TO DATE CUMULATIVE 5
: 7. HOURS GENERATOR ON LINE                   0.0         1397.3       2934.7
NUMBER OF HOURS REACTOR WAS CRITICAL 0.0 1399.8 3011.6 6.
: 8. UNIT RESERVE SHUTDOWN HOURS               0.0         0.0           0.0
REACTOR RESERVE SHUTDOWN HOURS 0.0 0.0 125.3 7.
: 9. GROSS THERMAL ENERGY GENERATED (MWH)     0.0       4387385       8894977
HOURS GENERATOR ON LINE 0.0 1397.3 2934.7 8.
: 10. GROSS ELEC. ENERGY GENERATED (MWH)         0.0       1460387       2945373
UNIT RESERVE SHUTDOWN HOURS 0.0 0.0 0.0 9.
: 11. NET ELEC. ENERGY GENERATED (MWH)         -10422       1400555       2792672
GROSS THERMAL ENERGY GENERATED (MWH) 0.0 4387385 8894977
: 12. REACTOR SERVICE FACTOR                     0.0%       64.1%         76.1%
: 10. GROSS ELEC. ENERGY GENERATED (MWH) 0.0 1460387 2945373 11.
: 13. REACTOR AVAILABILITY FACTOR                 0.0%       64.1%         79.2%
NET ELEC. ENERGY GENERATED (MWH)
: 14. UNIT SERVICE FACTOR                         0.0%       64.0%         74.1%
-10422 1400555 2792672 12.
: 15. UNIT AVAILABILITY FACTOR                   0.0%       64.0%         74.1%
REACTOR SERVICE FACTOR 0.0%
: 16. UNIT CAPACITY FACTOR (USING MDC)_         -1.4%       61.9%         68.1% _,
64.1%
: 17. UNIT CAPACITY FACTOR (USING DESIGN MWe)                                   -1. 3% _     59.5%         65.4%
76.1%
: 18. UNIT FORCED OUTAGE RATE                   0.0%         0.0%         6.0%
13.
: 19. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, AND DURATION OF EACH):
REACTOR AVAILABILITY FACTOR 0.0%
64.1%
79.2%
14.
UNIT SERVICE FACTOR 0.0%
64.0%
74.1%
15.
UNIT AVAILABILITY FACTOR 0.0%
64.0%
74.1%
: 16. UNIT CAPACITY FACTOR (USING MDC)_
-1.4%
61.9%
68.1% _,
17.
UNIT CAPACITY FACTOR (USING DESIGN MWe)
-1. 3% _
59.5%
65.4%
18.
UNIT FORCED OUTAGE RATE 0.0%
0.0%
6.0%
19.
SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, AND DURATION OF EACH):
An outage for maintenance and surveillance was begun at 0520 on February 28, 1985.
An outage for maintenance and surveillance was begun at 0520 on February 28, 1985.
: 20. IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP 5-30-85 DOCUMENT ID 0036r/0005r
20.
: 2. AVERAGE DAILY UNIT POWER LEVEL' DOCKET NO: 050-374 UNIT: LASALLE TWO DATE: April 10, 1985 COMPLETED BY: Richard J. Rohrer TELEPHONE: (815) 357-6761 MONTH: March 1985 DAY AVERAGE DAILY POWER LEVEL       DAY AVERAGE DAILY POWER LEVEL (MWe-Net)                               (MWe-Net)
IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP 5-30-85 DOCUMENT ID 0036r/0005r
: 1.             -14               17.                 -13
: 2. AVERAGE DAILY UNIT POWER LEVEL' DOCKET NO: 050-374 UNIT: LASALLE TWO DATE: April 10, 1985 COMPLETED BY: Richard J. Rohrer TELEPHONE: (815) 357-6761 MONTH: March 1985 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
: 2.             -14               18.                 -13 3             -15               19.                 -12
(MWe-Net) 1.
: 4.             -17               20.                 -11
-14 17.
: 5.             -15               21.                 -14
-13 2.
: 6.             -14               22.                 -15
-14 18.
: 7.             -14               23.                 -14
-13 3
: 8.             -15               24.                 -14
-15 19.
: 9.             -14               25.                 -13
-12 4.
: 10.             -14               26.                 -14
-17 20.
: 11.             -14               27.                 -13
-11 5.
: 12.             -14               28.                 -14
-15 21.
: 13.             -14               29.                 -14
-14 6.
: 14.             -14               30.                 -14
-14 22.
: 15.             -13               31.                 -14
-15 7.
: 16.             -14 4
-14 23.
-14 8.
-15 24.
-14 9.
-14 25.
-13 10.
-14 26.
-14 11.
-14 27.
-13 12.
-14 28.
-14 13.
-14 29.
-14 14.
-14 30.
-14 15.
-13 31.
-14 16.
-14 4
DOCUMENT ID 0036r/0005r
DOCUMENT ID 0036r/0005r


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E. UNIQUE REPORTING REQUIREMENTS
E.
: 1.     Safety / Relief Valve Operations for Unit Two.
UNIQUE REPORTING REQUIREMENTS 1.
DATE         VALVES         NO & TYPE                           PLANT     DESCRIPTION ACTUATED       ACTUATIONS                           CONDITION OF EVENT There were no safety relief valves actuated during this reporting period.
Safety / Relief Valve Operations for Unit Two.
DATE VALVES NO & TYPE PLANT DESCRIPTION ACTUATED ACTUATIONS CONDITION OF EVENT There were no safety relief valves actuated during this reporting period.
d i
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DOCUMENT ID 0036r/0005r
DOCUMENT ID 0036r/0005r
: 2. ECCS Systems Outtgis-The following outages were taken on ECCS Systems during the reporting period.
 
OUTAGE NO.                               EQUIPMENT                                             PURPOSE OF OUTAGE 2-201-85                                 High Pressure Core                                     Repack Valves 2E22-F023 and Spray                                                 2E22-F026 2-206-85                                 HPCS Water Leg Pump                                   Lubrication 2-209-85                                 2B Diesel Generator                                   18 Month Inspection 2-210-85                                 2B Diesel Generator                                   18 Month Inspection 2-216-85                                 2B Diesel Generator                                   Lubrication 2-226-85                                 RCIC Steam Supply                                     Repack Valve Valve 2-247-85                                 2B Diesel Generator                                   Repair Breaker output Breaker 2-248-85                                 2B Diesel Generator                                   Instrument Calibration 2-266-85                                 RCIC                                                   Surveillance 2-317-85                                 HPCS Pull Flow Test                                   Replace Broken Lug on Valve                                                 MCC 2-322-85                                 RCIC                                                   Electrical Work 2-327-85                                 RCIC Head Spray Line                                   Remove Spool and Blind Flange 2-328-85                                 RCIC                                                   Replace Steam Trap 2-334-85                                 HPCS Pump Breaker                                     Replace Puffer Piston 2-337-85                                 2B RHR Service Water                                   Repair Motor Pump 2-351-85                                 2B RHR Service Water                                   Replace Motor Pump 2-353-85                                 RCIC                                                   Repair Coupling, Turbine Trip and Throttle Valve, and Steam Trap 2-356-85                                 2B Diesel Generator                                   Work on Motor Driven Air Compressor 2-363-85                                 2E12-F009                                             Inspect Limitorque i
2.
ECCS Systems Outtgis-The following outages were taken on ECCS Systems during the reporting period.
OUTAGE NO.
EQUIPMENT PURPOSE OF OUTAGE 2-201-85 High Pressure Core Repack Valves 2E22-F023 and Spray 2E22-F026 2-206-85 HPCS Water Leg Pump Lubrication 2-209-85 2B Diesel Generator 18 Month Inspection 2-210-85 2B Diesel Generator 18 Month Inspection 2-216-85 2B Diesel Generator Lubrication 2-226-85 RCIC Steam Supply Repack Valve Valve 2-247-85 2B Diesel Generator Repair Breaker output Breaker 2-248-85 2B Diesel Generator Instrument Calibration 2-266-85 RCIC Surveillance 2-317-85 HPCS Pull Flow Test Replace Broken Lug on Valve MCC 2-322-85 RCIC Electrical Work 2-327-85 RCIC Head Spray Line Remove Spool and Blind Flange 2-328-85 RCIC Replace Steam Trap 2-334-85 HPCS Pump Breaker Replace Puffer Piston 2-337-85 2B RHR Service Water Repair Motor Pump 2-351-85 2B RHR Service Water Replace Motor Pump 2-353-85 RCIC Repair Coupling, Turbine Trip and Throttle Valve, and Steam Trap 2-356-85 2B Diesel Generator Work on Motor Driven Air Compressor 2-363-85 2E12-F009 Inspect Limitorque i
DOCUMENT ID 0036r/0005r
DOCUMENT ID 0036r/0005r
              , - - - - - . , . - ,,w<,,   , , , - - ,  , , , - - - - , - , , , , - , + . - , -     ww       -, , , - -  ---w,-. .m-,   w -
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OUTRGE NO.                 BQUIPMENT                       PURPOSE OF OUTAGE 2-377-85                   23 DiO001 Generator             Troublechooting Motor Driven Air compressor 2-381-85                   28 Diesel Generator             Inspect Air Start Motor Driven Air                 Motor Compressor 2-388-85                   2A Diesel Generator             Repair Puffer Feed to Bus 242Y               , Piston 2-391-85                   2A Diesel Generator /           18 Month Inspection 2-394-85                   2E21-F023                       Technical Specification Requirement 2-407-85                   RCIC                             Ldcal Leak Rate Testing 2-409-85                   2E22-F023                       Repack Valve 2-424-85                   2A RHR Service Water           Replace Seal Process Radiation Monitor 2-433-85                 '2E51-F065                       Repair Hinge Pin 2-435-85                   2E22-F004                       Inspect Motor for jumper 2-438-85                   2E51-F013                       Inspect Motor 2-443-85                   2B12-F009                       Environmental Qualification Modifiction (EQ Mod) 2-451-85                   Low Pressure Core Spray Surveillance 2-453-85                   High Pressure Core               EQ Mod Spray 1   2-468-85                   2E22-N012                       EQ Mod l
OUTRGE NO.
2-469-85                   2E22-N012                       BQ Mod I
BQUIPMENT PURPOSE OF OUTAGE 2-377-85 23 DiO001 Generator Troublechooting Motor Driven Air compressor 2-381-85 28 Diesel Generator Inspect Air Start Motor Driven Air Motor Compressor 2-388-85 2A Diesel Generator Repair Puffer Feed to Bus 242Y Piston 2-391-85 2A Diesel Generator /
l   2-470-85                   2822-N006                       BQ Mod l
18 Month Inspection 2-394-85 2E21-F023 Technical Specification Requirement 2-407-85 RCIC Ldcal Leak Rate Testing 2-409-85 2E22-F023 Repack Valve 2-424-85 2A RHR Service Water Replace Seal Process Radiation Monitor 2-433-85
l   2-471-85                   2E22-N006                       EQ Mod i
'2E51-F065 Repair Hinge Pin 2-435-85 2E22-F004 Inspect Motor for jumper 2-438-85 2E51-F013 Inspect Motor 2-443-85 2B12-F009 Environmental Qualification Modifiction (EQ Mod) 2-451-85 Low Pressure Core Spray Surveillance 2-453-85 High Pressure Core EQ Mod Spray 1
: 3. Off-Site Dose Calculation Manual There were no changes to the off-site dose calculations manual during this reporting period.
2-468-85 2E22-N012 EQ Mod l
A    Radioactive Waste Treatment Systems.
2-469-85 2E22-N012 BQ Mod I
Several procedures were revised to allow solidification of spent resin using a 20% Free-Standing water formula. Previously, a 10%
l 2-470-85 2822-N006 BQ Mod l
l 2-471-85 2E22-N006 EQ Mod i
3.
Off-Site Dose Calculation Manual There were no changes to the off-site dose calculations manual during this reporting period.
Radioactive Waste Treatment Systems.
A Several procedures were revised to allow solidification of spent resin using a 20% Free-Standing water formula. Previously, a 10%
Free-Standing water formula was used.
Free-Standing water formula was used.
DOCUMENT ID 0036r/0005r p
DOCUMENT ID 0036r/0005r p


1 l
1
Commonwrith Edison                                            I LaSatte County Nuclear Station
@ Telephone 815/357-6761 Commonwrith Edison LaSatte County Nuclear Station Rural Route #1, Box 220 Marse:lles,lilinois 61341 April 10, 1985 Director, Office of Management Information and Program Control United States Nuclear Regulatory Commission Washington, D.C.
      @ Telephone 815/357-6761 Rural Route #1, Box 220 Marse:lles,lilinois 61341 April 10, 1985 Director, Office of Management Information and Program Control United States Nuclear Regulatory Commission Washington, D.C. 20555 ATTN: Document Control Desk Gentlemen:
20555 ATTN: Document Control Desk Gentlemen:
Enclosed for your information is the monthly performance report covering LaSalle County Nuclear Power Station for the period covering March 1 through March 31, 1985.
Enclosed for your information is the monthly performance report covering LaSalle County Nuclear Power Station for the period covering March 1 through March 31, 1985.
Very truly yours, r#                   J
Very truly yours, r#
                                                    ' G. J. Diederich   #/h-S perintenient LaSalle County Station GJD/RJR/cth Enclosure l
J
l xc:   J. G. Keppler, NRC, Region III NRC Rt.sident Inspector LaSalle Gary kright, Ill. Dept. of Nuclear Safety D. P. Galle, CECO
' G. J. Diederich
;-      D. L. Farrar, CECO l       INPO Records Center Ron A. Johnson, PIP Coordinator SNED J. E. Ellis, GE Resident J. M. Nowicki, Asst. Comptroller l
#/h-S perintenient LaSalle County Station GJD/RJR/cth Enclosure l
!        H. E. Bliss, Nuclear Fuel Services Manager l
l xc:
J. G. Keppler, NRC, Region III NRC Rt.sident Inspector LaSalle Gary kright, Ill. Dept. of Nuclear Safety D. P. Galle, CECO D. L. Farrar, CECO l
INPO Records Center Ron A. Johnson, PIP Coordinator SNED J. E. Ellis, GE Resident l
J. M. Nowicki, Asst. Comptroller H. E. Bliss, Nuclear Fuel Services Manager l
I o
I o
6}}
6}}

Latest revision as of 00:19, 13 December 2024

Monthly Operating Rept for Mar 1985
ML20115F648
Person / Time
Site: LaSalle 
Issue date: 03/31/1985
From: Diederich G, Rohrer R
COMMONWEALTH EDISON CO.
To:
NRC OFFICE OF ADMINISTRATION (ADM)
References
NUDOCS 8504190611
Download: ML20115F648 (30)
v • d • e


Text

.

LASALLE NUCLEAR POWER STATION UNIT 1 MONTHLY PERFORMANCE REPORT MARCH 1985 COIMONWEALTH EDISON COMPANY t

NRC DOCKET NO. 050-373 i

LICENSE NO. NPF-11 i

V i

8504190611 850331 ADOCK0500g3 DR

a 3

TABLE OF CONTENTS I.

INTRODUCTION II.

MONTHLY REPORT FOR UNIT ONE A.

Sununary of Operating Experience B.

PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE 1.

Amendments to Facility License or Technical Specifications 2.

Facility or Procedure Changes Requiring NRC Approval 3.

Tests and Experiments Requiring NRC Approval 4.

Corrective Maintenance of Safety Related Equipment C.

LICENSEE EVENT REPORTS D.

DATA TABULATIONS 1.

Operating Data Report 2.

Average Daily Unit Power Level 3.

Unit Shutdowns and Power Reductions E.

UNIQUE REPORTING REQUIREMENTS l

1.

Main Steam Relief Valve Operations 2.

ECCS Sptem Outages l

3.

Off-Site Dose Calculation Manual Changes 4.

Major Changes to Radioactive Waste Treatment l

System i

l

I.

INTRODUCTION The LaSalle Nuclear Power Station is a Two Unit Facility Located in Marseilles, Illinois. Each Unit is a Boiling Water Reactor with a designed electrical output of 1078 MWe net. The Station is owned by Comunonwealth Edison company. The Architect / Engineer was Sargent & Lundy, and the primary construction contractor was Commonwealth Edison Company.

The condenser cooling method is a closed cycle cooling pond. Unit One is subject to License Number NPF-ll, issued on April 17, 1982. The date of initial criticality was June 21, 1982. Unit Two is subject to license number NPP-18, issued on December 16, 1983. The date of initial criticality was March 10, 1984.

s This report was compiled by Richard J. Rohrer telephone number (815)357-6761, extension 575.

11.

MONTHLY REPORT FOR UNIT ONE A.

SUlWWtY OF OPERATING EXPERIENCE FOR UNIT ONE March 1-3 March 1, 0001 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Reactor Critical at 99% Power, generator on-line.

March 3, 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> Reactor Power at 93%.

March 3, 1654 hours0.0191 days <br />0.459 hours <br />0.00273 weeks <br />6.29347e-4 months <br /> Reactor Scram, Turbine Control Valve Fast Closure due to an offsite fault on-transmission line 0101. The reactor had been critical for 64 hours7.407407e-4 days <br />0.0178 hours <br />1.058201e-4 weeks <br />2.4352e-5 months <br /> and 54 minutes in March.

MARCH 5-21 March 5, 1024 hours0.0119 days <br />0.284 hours <br />0.00169 weeks <br />3.89632e-4 months <br /> Reactor critical.

March 5, 1830 hours0.0212 days <br />0.508 hours <br />0.00303 weeks <br />6.96315e-4 months <br /> Generator was synchronized to the grid.

March 7, 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> Reactor power at 79%.

March 8, 0015 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> Reactor power decreased to 55% in order to manipulate control rods.

March 8, 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br /> Reactor power restored to 91%.

March 14, 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> Reactor power decreased to 74% in order to manipulate control rods.

March 15, 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> Reactor power restored to 99%.

March 21, 1317 hours0.0152 days <br />0.366 hours <br />0.00218 weeks <br />5.011185e-4 months <br /> Reactor scram, Low Vessel Level and High Reactor Pressure signals due to valving error during an i

Envionmental Qualification Modification. The reactor was critical for a total of 451 hours0.00522 days <br />0.125 hours <br />7.457011e-4 weeks <br />1.716055e-4 months <br /> and 47 minutes in March.

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B.

PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS AND SAFETY RELATED MAINTENANCE.

1.

Amendments to facility license or Technical Specification.

There were no amendments to the facility license or Technical Specification during this reporting period.

2.

Facility or procedure changes requiring NRC approval.

There were no facility or procedure changes requiring NRC approval during this reporting period.

3.

Tests and Experiments requiring NRC approval.

There were no tests or experiments requiring NRC, approval during this reporting period.

4.

Corrective maintenance of safety related equipment.

The following table (Table 1) presents a sununary of safety-related maintenance completed on Unit One during the reporting period. The headings indicated in this summary include: Work Request number, Component Name, Cause of Malfunction, Results and Effects on Safe Operation, and Corrective Action.

TABLE 1

' CORRECTIVE MAINTENANCE OF SAFETY RELATED EQUIPMENT WORK REQUEST COMPONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION ON SAFE OPERATION J L44677 Main Steam High Contacts out of Adjustment Inconsistent Alarm function.

Adjusted Contacts.

Radiation Relay L46942 HCU Accumulator for Leaking valve.

Potential to cause failure to Replaced valve.

CRD 30-55.

scram this rod if combined with other events.

L46947 HCU Accumulator for Leaking Valve.

Potential to cause failure to Replaced valve.

CRD 34-59.

scram this rod if combined with other events.

L46974 HCU Accumulator for Leaking Valve.

Potential to cause failure to Replaced valve.

CRD 58-31 scram this rod if combined with other events.

L46986 HOU Accumulator for Leaking Valve.

Potential to cause. failure to Replaced Valve.

CRD 42-07.

scram this rod if combined with other events.

L47009 Diesel Generator 0 Oil Leak.

Possible degraded pump Replaced Bearing.

Cooling Water Pump.

Performance.

L47038 HCU Accumulator for Leaking Valve Potential to cause failure to Replaced valve.

CRD 58-19 scram this rod if combined with other events.

L47219 No. 2 Turbine control I.imit Switch Arm was Loss of indication of valve's Replaced limit switch Valve.

broken off.

position, arm.

L47274 "A" Chlorine Detector Glass orifice which Loss of redundancy.

Replaced orifice and on "A" Control Room supplies electrolyte wick.

HVAC.

was clogged.

DOCUMENT 0044r/0005r

TAILE 1 CORRECTIVE MAINTENANCE OF SAFETY RELATED EQUIPMENT WORK REQUEST COMPONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION ON SAFE OPERATION L47679 Division II Batteries. Excessive gaps between Possible seismic deficiency.

Installed material batteries and side rails.

to r2ke gaps narrower.

L47691 Inlet damper to "A" Faulty microswitch in Actuator would not move damper. Cleaned microswitch Control Room INAC

actuator, and verified proper Return Fan.

operation.

I DOCUMENT 0044r/0005r

C.

LICENSEE EVENT REPORTS The following is a tabular sununary of all licensee event reports.for LaSalle Nuclear Power Station, Unit One, occurring during the reporting period, March 1 through March 31, 1985. This information is provided pursuant to the reportable occurrence reporting requirements as set forth in 10CFR 50.73.

Licensee Event Report Number Date Title of Occurrence 85-015-00 2-7-85 "C" Vacuum Breaker Cycled 85-016-00 2-8-85 Unsecured High Radiation Door 85-017-00 2-8-85 Unit Scram 85-018-00 2-16-85 Missed Offgas Hydrogen Sample Frequency 85-019-00 2-12-85 High Rad Door Unsecured 85-020-00 2-14-85 Unsecured High Radiation Door 85-021-00 2-20-85 Unsecured High Radiation Area Door 85-022-00 2-22-85 RCIC Steam Line High Flow Isolation During Warm-up.

85-023-00 2-23-85 Unsecured High Radiation Area l

l 85-024-00 3-3-85 Reactor Scram l

85-025-00 3-4-85 Drywell Vacuum Breaker Cycled l

85-026-00 3-13-85 Passed Critical Date on Monthly Functional Test 85-027-00 3-13-85 Unsecured High Radiation Area.

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I Document 0043r/0005r

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t D.

DATA TABULATIONS i

The following data tabulations are presented in this report:

1.

Operating Data Report 2.

Average Daily Unit Power Level 3.

Unit Shutdowns and Power Reductions 3

1 6

f

1.

OPERATING DATA REPORT DOCKET NO. 050-373 UNIT LaSalle One DATE April 10, 1985 COMPLETED BY Richard J. Rohrer TELEPHONE (815)357-6761 OPERATING STATUS 1.

REPORTINJ PERIOD: March, 1985 GROSS HOURS IN REPORTING PERIOD: 744 2.

CURRENTLY AUTHORIZED POWER LEVEL (MWt):3323 MAX DEPEND CAPACITY (IWe-Net): 1036 DESIGN ELECTRICAL RATING (MWe-Net):1078 3.

POWER LEVEL TO WHICH RESTRICTED (IP'ANY) (MWe-Net):

N/A 4.

REASONS FOR RESTRICTION (IF ANY):N/A THIS' MONTH YR TO DATE CUMULATIVE 5

NUMBER OF HOURS REACTOR WAS CRITICAL 451.8 1770.1 8051 6.

REACTOR RESERVE SHUTDOWN HOURS 41.5 71.0 1237 7.

HOURS GENERATOR ON LINE 443.7 1721.3 7777 8.

UNIT RESERVE SHUTDOWN HOURS 0.0 0.0 0.0 9.

GROSS THERMAL ENERGY GENERATED (MWH) 1339056' 4841667 21664956 10.

GROSS ELEC. ENERGY GENERATED (MWH) 449172 1604050 7074693

11. NET ELEC. ENERGY GENERATED (MWH) 431306 1541242 6736304
12.. REACTOR SERVICE FACTOR 60.7%

81.8%

73.4%

13.

REACTOR AVAILABILITY FACTOR 66.3%

84.3%

84.7%

14.

UNIT SERVICE FACTOR 59.6%

78.8%

70.9%

15.

UNIT AVAILABILITY FACTOR 59.6%

78.8%

70.9%

16.

UNIT CAPACITY FACTOR (USING MDC) 56.0%

68.1%

59.3%

17.

UNIT CAPACITY FACTOR (USING DESIGN MWe) 53.8%

65.5%

57.0%

18.

UNIT FORCED OUTAGE RATE 40.4%

20.1%

13.8%

19.

SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, AND DURATION OF EACH)

-Unit one is scheduled for a refueling, maintenance, modification, and surveillance outage beginning September 3, 1985 and lasting 26 weeks.

20.

IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP:

4-5-85.

Document-0043r/0005r

2. AVERAGE DAILY UNIT POWER LEVEL DOCKET NO: 050-373 UNIT: LASALLE OWE DATE: April 10, 1985 COMPLETED BY: Richard J. Rohrer TELEPHONE: (815) 357-6761 MONTH: MARCH, 1985 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 1.

1068 17.

1086 2.

1069 '

18.

1083 3

730 19.

1075 4.

-18 20.

1068 5.

15 21.

584 f

6.

473 22.

-15 7.

797 23.

-15 8.

856 24.

-15 9.

1050 25.

-14 10.

1064 26.

-15 11.

1004 27.

-14 12.

1053 28.

-15 13.

1000 29.

-15 14.

975 30.

-15 15.

1072 31.

-15 16.

1009

ATTACHMENT E

3. ~ UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 050-373 UNIT NAME LaSalle One DATE March 10i 1985 REPORT MONTH February 1985 COMPLETED BY Richard J. Rohrer TELEPHONE (815)357-6761 METHOD OF TYPE SHUTTING DOWN F: FORCED DURATION THE REAC'!OR OR CORRECTIVE NOo DATE S: SCHEDULED (HOURS)

REASON REDUCING POWER ACTIONS /COf9 TENTS 5

850303 P

49.60 A

3 Reactor Scram on Turbine Control Valve Fast Closure due to an offsite fault on transmission line 0101.

6 850321 F

250.72 G

3 Reactor Scram on Low Vessel Level and High Vessel Pressure Signals due to a valving error during installation of an Environmental Qua1ification Modifiction.

DOCUMENT 0044r/0005

E.

UNIQUE REPORTING REQUIREMENTS

1. Safety / Relief valve operations for Unit One.

VALVES NO & TYPE PLANT DESCRIPTION DATE ACTUATED ACTUATION CONDITION OF EVENT 3-3-85 IB21-F013E 1 1075 psig Actuated to relieve pressure following reactor scram.

3-3-85 IB21-F013U 1 1075 psig Actuated to relieve pressure following reactor scram.

3-3-85 IB21-F013S 1 1075 psig Actuated to relieve pressure following reactor scram.

3-21-85 IB21-F0138 1 1060 psig Actuated following reactor scram.

3-27-85 IB21-F013S 3 510 psig Unitentional actuations during pressure test of instrument lines following installation of an Environmental Qualification Modification.

3-27-85 IB21-F013J 1 510 psig Unintentional Actuation during pressure test of instrument lines following installation of an Environmental Qualification Modification.

3-27-85 IB21-F013B 1 510 psig Unintentional Actuation during pressure test of l

instrument. lines following installation of an Environmental Qualification Modification.

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Document 0043r/0005r i

2.

ECCS Systems Outtgaa The following outages were taken on ECCS Systems during the reporting period.

OUTAGE NO.

EOUIPMENT PURPOSE OF OUTAGE l-196-85 1DOO3P Lubrication 1-206-85 LPCS Pump Discharge Environmental Qualification Flow Switch Modification (EQ Mod)

.1-207-85 LPCS Pump Discharge EQ Mod.

Flow Switch 1-208-85 LPCS Pump ADS EQ Mod Permissive Switch 1-209-85 LPCS Pump ADS EQ Mod Permissive Switch 1-210-85 RHR Shutdown Cooling EQ Mod 135 PSIG Pressure Switch 1-211-85 RHR Shutdown Cooling EQ Mod 135 PSIG Presure Switch 1-213-85 RCIC Low Suction EQ Mod Pressure Switch 1-214-85 1E51-N022A,B,C,D EQ Mod 1-215-85 1E21-F011 Technical Specification Requirement 1-216-85 RCIC System Telchnical Specification Requirement 1-220-85 1E51-N006 EQ Mod 1-224-85 IB Diesel Generator Lubrication Fuel Oil Transfer Pump l-230-85 1E22-N012 EQ Mod 1-231-85 1E22-N012 EQ Mod 1-232-85 1E22-N006 EQ Mod 1-233-85 1822-N006 EQ Mod Document 0043r/0005r

OUTAGE NO-EQUIPMENT PURPOSE OF OUTAGE 1-247-85 HPCS pump tad 1B Modification Diesel Generator 1-249-85 "B" RHR Pump Running EQ Mod ADS Permissive Switch 1-250-85 "B" RHR Pump Running EQ Mod ADS Permissive Switch 1-251-85 "C" RHR Pump Running EQ Mod ADS Permissive Switch 1-252-85 "C" RHR Pump Running EQ Mod ADS Permissive Switch 1-270-85 IE12-N010A EQ Mod 1-271-85 lE12-N010A EQ Mod 1-272-85 1812-F064A Technical Specification Requirement 1-275-85 lE12-N010C EQ Mod 1-276-85 lE12-N010C EQ Mod 1-277-85 1E12-F064C Technical Specification Requirement 1-283-85 lE12-N010B EQ Mod 1-284-85 lE12-N010B EQ Mod t

1-285-85 1E12-F064B Technical Specification Requirement 1-308-85 1A Diesel Generator Repair Motor Cooling Water Pump i

3.

Off-Site Dose Calculation Manual There were no changes to the off-site dose calculations manual during this reporting period.

4.

Radioactive Waste Treatment Systems.

Several Procedures were revised to allow solidification of spent resin using a 20% Free-Standing Water formula. Previously, a 10% Free-Standing Water Formula was used.

i Document 0043r/0005r

LASALLE NUCLEAR POWER STATION UNIT 2 MONTHLY PERFORMANCE REPORT MARCH 1985 COOGONWEALTH EDISON COMPANY NRC DOCKET NO. 050-374 LICENSE NO. NPF-18 i

l 1

5 DOCUMENT ID 0036r/0005r

TABLE OF CONTENTS I.

INTRODUCTION II.

MONTHLY REPORT FOR UNIT TWO A.

Sunnary of Operating Experience B.

PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE 1.

Amendments to Facility License or Technical Specifications 2.

Facility or Procedure Changes Requiring NRC Approval 3.

Tests and Experiments Requiring NRC Approval 4.

Corrective Maintenance of Safety Related Equipment C.

LICENSEE EVENT REPORTS D.

DATA TABULATIONS 1.

Operating Data Report 2.

Average Daily Unit Power Level 3.

Unit Shutdowns and Power Reductions E.

UNIQUE REPORTING REQUIREMENTS 1.

Safety / Relief Valve Operations 2.

ECCS System Outages 3.

Off-Site Dose calculation Manual Changes 4.

Major Changes to Radioactive Waste Treatment System l

l 4

l DOCUMENT ID 0036r/0005r

I.

INTRODUCTION The LaSalle Nuclear Power Station is a Two Unit Facility Located in Marseilles, Illinois. Each Unit is a Boiling Water Reactor with a designed electrical output of 1078 MWe net. The Station is owned by Commonwealth Edison company. The Architect / Engineer was Sargent & Lundy, and the primary construction contractor was Commonwealth Edison Company.

The condenser cooling method is a closed cycle cooling pond. Unit One is subject to License Number NPF-ll, issued on April 17, 1982. The unit commenced commercial generation of power on January 1, 1984. Unit Two is subject to license number NPF-18, issued on December 16, 1983. The date of initial criticality was March 10, 1984.

This report was compiled by Richard J. Rohrer,, telephone number (815)357-6761, extension 575.

DOCUMENT ID 0036r/0005r i

II.

MONTHLY REPORT FOR UNIT TWO A.

SUlttARY OF OPERATING EXPERIENCE FOR UNIT TWO MARCH 1 The. reactor was subcritical, and the generator was off-line for the entire month of March. Unit Two is in a schedled outage for maintenance, testing, and modifications.

1 DOCUMENT ID 0036r/0005r

a B.

PLANT OR PROCEDURE CHANGES, TESTS,~ EXPERIMENTS AND SAFETY RELATED MAINTENANCE.

1.

Amendments to facility license or Technical Specifications.

Amendment 8 was made to the Unit Two Operating license. This Amendment extends the deadline for full compliance to 10CFR50.49 from March 31, 1985 to November 30, 1985.

2.

Facility or procedure changes requiring NRC approval.

There were no facility or procedure changes requiring NRC approval during the reporting period.

3.

Tests and experiments requiring NRC approval.

There were no tests or experiments requiring NRC approval during the reporting period.

4.

Corrective Maintenance of Safety Related Equipment.

The following table (Table 1) presents a summary of safety-related maintenance completed on Unit one during the reporting period. The headings indicated in this summary include: Work Request number, Component Name, cause of malfunction, results and effects on safe operation, and corrective action.

i l

DOCUMENT ID 0036r/0005r

T TABLE 1 CORRECTIVE MAINTENANCE OF i

SAFETY RELATED EQUIPMENT WORK REQUEST COMPONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION ON SAFE OPERATION

, L37828 Tailpipe temperature Faulty thermocouple.

Sporadic indication.

Replaced thermocouple.

element for SRV "S".

L40429 HCU Accumulator for Leaking disconnect Potential to cause failure to Replaced disconnect CRD 22-55.

fitting.

scram this rod if combined with fitting.

other events.

L45891 Suppression Pool Faulty lug on Control Erratic readings.

Replaced faulty lug.

Water Temperature Center element.

L46488 Division II Instrument out of Incorrect indication.

Recalibrated Instrument.

Suppression Chamber calibration. Reading Pressure Indication.

+1.5 psig.

3 L46608 Refrigeration Unit Oil Level too high.

Refrigeration Unit periodically Lowered oil level.

for "B" Auxiliary tripped on high oil temperature.

Electrical Equipment Possible inadequate ventilation.

i Room HVAC.

l i L46824 HCU Accumulator for Leaking Valve.

Potential to cause failure to Replaced valve.

CRD 26-35 scram this rod if combined with other events.

l L46832 HCU Accumulator for Leaking Valve.

Potential.to cause failure to Replaced valve.

CRD 26-39.

scram this rod if combined with other events.

-f L46837 Electrical Power Faulty Card.

Non-repeatable readings for Removed card from Unit Monitoring Card for overvoltage trip setpoint.

I and installed in Unit RPS System.

2.

L46870 Diesel Generator "2B" Air supply lines reversed. Loss of Redundancy.

Corrected arrangement l

Air Start Motor.

of air supply lines.

l 4

DOCUMENT 0044r/0005r i

s TABLE 12 CORRECTIVE MAINTENANCE OF SAFETY RELATED BQUIPMENT WORK REQUEST COMPONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION ON SAFE OPERATION L46886 HCU Accumulator for Leaking Valve.

Potential to cause failure to Replaced valve.

CRD 26-43.

scram this rod if combined with-other events.

L46889 Division II 125 Volt Faulty input card.

Relay chattered in and out.

Replaced input card.

Breaker Trip Alarm No alarm was indicated.

Relay.

L42893 RHR Temperature Dirty Selector Switch,

- Sporadically printed upscale.

Cleaned selector switch, Recorder.

faulty ink wheel, replaced ink wheel.

L46927 HCU Accumulator for.

Leaking Valve.

Potential to cause failure to

' Replaced valve.

CRD 14-55.

scram this rod if' combined with other events.

L47073 Electrical Power Faulty Breaker.

Breaker would not reclose.

Installed Breaker Monitoring Breaker removed from unit one.

for RPS.

L47300 Main Steam Isolation Torque switch out of Valve would not cycle.

Adjusted torque switch.

t Valve Leakage Control Adjustment.

Inboard Valve.

L47377 HPCS Full Flow Test Broken lug at Motor Control Valve could not be cycled Replaced lugs on motor Valve.

Center.

remotely.

connection.

L47440 2B RHR Service Water Shorted motor.

Pump Feed Breaker Tripped. Pump Replaced Motor.

Pump.

Inoperable.

L47557 "2B" Diesel Generator Ground on Bridge.

Loss of position indication.

Repaired ground.

Fuel Oil Transfer Pump.

L47640 Unit Two Hydrogen Dirty Contacts.

No indication of " closed" Cleaned contacts.

j Recombiner Cross-tie position.

valve.

L47719 Division I Battery Faulty Connections on Periodic Tripping of breaker.

Replaced breaker and Charger.

line side of breaker.

breaker trip unit.

C.

LICENSEE EVENT REPORTS The following is a tabular summary of all licensee event reports for LaSalle Nuclear Power Station, Unit Two, occurring during the reporting period, March 1 through March 31, 1985. This information is provided pursuant to the reportable occurrence reporting requirements as set forth in 10CFR 50.73.

Licensee Event Report Number Date Title of Occurrence 85-008-00 2-13-85 Ammonia Detector Actuation 85-009-00 3-1-85 Reactor Water Cleanup Differential Flow Isolation when Reactor Vessel Head Vent Opened.

85-010-00 2-15-85 Hydraulic Control Unit Accumulator Pressure Switch Failures 85-011-00 2-26-85

. Failure of Type C Leak Rate Test DOCUMENT ID 0036r/0005r

D.

DATA TABULATIONS The following data tabulations are presented in this report:

1. Operating Data Report
2. Average Daily Unit Power Level
3. Unit Shutdowns and Power Reductions i

i i

8 DOCUMENT ID 0036r/0005r

1. OPERATING DATA REPORT DOCKET NO. 050-374 UNIT LaSalle Two DATE April 10, 1985 COMP'.ETED BY Richard J. Rohrer TEr.EPHONE (815)357-6761 OPERATING STATUS 1.

REPORTING PERIOD: March, 1985 GROSS HOURS IN REPORTING PERIOD: 744 _

2.

CURRENTLY AUTHORIZED POWER LEVEL (MWt):3323 MAX DEPEND CAPACITY (MWe-Net): 1036 DESIGN ELECTRICAL RATING (MWe-Net):1078 3.

POWER LEVEL TO WHICH RESTRICTED (IP ANY) (MWe-Net):

N/A 4.

REASONS FOR RESTRICTION (IF ANY):

THIS MONTH YR TO DATE CUMULATIVE 5

NUMBER OF HOURS REACTOR WAS CRITICAL 0.0 1399.8 3011.6 6.

REACTOR RESERVE SHUTDOWN HOURS 0.0 0.0 125.3 7.

HOURS GENERATOR ON LINE 0.0 1397.3 2934.7 8.

UNIT RESERVE SHUTDOWN HOURS 0.0 0.0 0.0 9.

GROSS THERMAL ENERGY GENERATED (MWH) 0.0 4387385 8894977

10. GROSS ELEC. ENERGY GENERATED (MWH) 0.0 1460387 2945373 11.

NET ELEC. ENERGY GENERATED (MWH)

-10422 1400555 2792672 12.

REACTOR SERVICE FACTOR 0.0%

64.1%

76.1%

13.

REACTOR AVAILABILITY FACTOR 0.0%

64.1%

79.2%

14.

UNIT SERVICE FACTOR 0.0%

64.0%

74.1%

15.

UNIT AVAILABILITY FACTOR 0.0%

64.0%

74.1%

16. UNIT CAPACITY FACTOR (USING MDC)_

-1.4%

61.9%

68.1% _,

17.

UNIT CAPACITY FACTOR (USING DESIGN MWe)

-1. 3% _

59.5%

65.4%

18.

UNIT FORCED OUTAGE RATE 0.0%

0.0%

6.0%

19.

SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, AND DURATION OF EACH):

An outage for maintenance and surveillance was begun at 0520 on February 28, 1985.

20.

IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP 5-30-85 DOCUMENT ID 0036r/0005r

2. AVERAGE DAILY UNIT POWER LEVEL' DOCKET NO: 050-374 UNIT: LASALLE TWO DATE: April 10, 1985 COMPLETED BY: Richard J. Rohrer TELEPHONE: (815) 357-6761 MONTH: March 1985 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 1.

-14 17.

-13 2.

-14 18.

-13 3

-15 19.

-12 4.

-17 20.

-11 5.

-15 21.

-14 6.

-14 22.

-15 7.

-14 23.

-14 8.

-15 24.

-14 9.

-14 25.

-13 10.

-14 26.

-14 11.

-14 27.

-13 12.

-14 28.

-14 13.

-14 29.

-14 14.

-14 30.

-14 15.

-13 31.

-14 16.

-14 4

DOCUMENT ID 0036r/0005r

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UNIQUE REPORTING REQUIREMENTS 1.

Safety / Relief Valve Operations for Unit Two.

DATE VALVES NO & TYPE PLANT DESCRIPTION ACTUATED ACTUATIONS CONDITION OF EVENT There were no safety relief valves actuated during this reporting period.

d i

DOCUMENT ID 0036r/0005r

2.

ECCS Systems Outtgis-The following outages were taken on ECCS Systems during the reporting period.

OUTAGE NO.

EQUIPMENT PURPOSE OF OUTAGE 2-201-85 High Pressure Core Repack Valves 2E22-F023 and Spray 2E22-F026 2-206-85 HPCS Water Leg Pump Lubrication 2-209-85 2B Diesel Generator 18 Month Inspection 2-210-85 2B Diesel Generator 18 Month Inspection 2-216-85 2B Diesel Generator Lubrication 2-226-85 RCIC Steam Supply Repack Valve Valve 2-247-85 2B Diesel Generator Repair Breaker output Breaker 2-248-85 2B Diesel Generator Instrument Calibration 2-266-85 RCIC Surveillance 2-317-85 HPCS Pull Flow Test Replace Broken Lug on Valve MCC 2-322-85 RCIC Electrical Work 2-327-85 RCIC Head Spray Line Remove Spool and Blind Flange 2-328-85 RCIC Replace Steam Trap 2-334-85 HPCS Pump Breaker Replace Puffer Piston 2-337-85 2B RHR Service Water Repair Motor Pump 2-351-85 2B RHR Service Water Replace Motor Pump 2-353-85 RCIC Repair Coupling, Turbine Trip and Throttle Valve, and Steam Trap 2-356-85 2B Diesel Generator Work on Motor Driven Air Compressor 2-363-85 2E12-F009 Inspect Limitorque i

DOCUMENT ID 0036r/0005r

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OUTRGE NO.

BQUIPMENT PURPOSE OF OUTAGE 2-377-85 23 DiO001 Generator Troublechooting Motor Driven Air compressor 2-381-85 28 Diesel Generator Inspect Air Start Motor Driven Air Motor Compressor 2-388-85 2A Diesel Generator Repair Puffer Feed to Bus 242Y Piston 2-391-85 2A Diesel Generator /

18 Month Inspection 2-394-85 2E21-F023 Technical Specification Requirement 2-407-85 RCIC Ldcal Leak Rate Testing 2-409-85 2E22-F023 Repack Valve 2-424-85 2A RHR Service Water Replace Seal Process Radiation Monitor 2-433-85

'2E51-F065 Repair Hinge Pin 2-435-85 2E22-F004 Inspect Motor for jumper 2-438-85 2E51-F013 Inspect Motor 2-443-85 2B12-F009 Environmental Qualification Modifiction (EQ Mod) 2-451-85 Low Pressure Core Spray Surveillance 2-453-85 High Pressure Core EQ Mod Spray 1

2-468-85 2E22-N012 EQ Mod l

2-469-85 2E22-N012 BQ Mod I

l 2-470-85 2822-N006 BQ Mod l

l 2-471-85 2E22-N006 EQ Mod i

3.

Off-Site Dose Calculation Manual There were no changes to the off-site dose calculations manual during this reporting period.

Radioactive Waste Treatment Systems.

A Several procedures were revised to allow solidification of spent resin using a 20% Free-Standing water formula. Previously, a 10%

Free-Standing water formula was used.

DOCUMENT ID 0036r/0005r p

1

@ Telephone 815/357-6761 Commonwrith Edison LaSatte County Nuclear Station Rural Route #1, Box 220 Marse:lles,lilinois 61341 April 10, 1985 Director, Office of Management Information and Program Control United States Nuclear Regulatory Commission Washington, D.C.

20555 ATTN: Document Control Desk Gentlemen:

Enclosed for your information is the monthly performance report covering LaSalle County Nuclear Power Station for the period covering March 1 through March 31, 1985.

Very truly yours, r#

J

' G. J. Diederich

  1. /h-S perintenient LaSalle County Station GJD/RJR/cth Enclosure l

l xc:

J. G. Keppler, NRC, Region III NRC Rt.sident Inspector LaSalle Gary kright, Ill. Dept. of Nuclear Safety D. P. Galle, CECO D. L. Farrar, CECO l

INPO Records Center Ron A. Johnson, PIP Coordinator SNED J. E. Ellis, GE Resident l

J. M. Nowicki, Asst. Comptroller H. E. Bliss, Nuclear Fuel Services Manager l

I o

6

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