ML20034E729

Annual Rept for LaSalle County Nuclear Power Station for 1992

ML20034E729
Person / Time
Site:	LaSalle
Issue date:	12/31/1992
From:	Spedl G COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
NUDOCS 9303010254
Download: ML20034E729 (91)

Text

. _ _ _ _ _ - _ _ _ _

s Commonwealth Edison LaSalle County Nuclear Station 2601 N. 21st. Rd.

Marseilles,I!!inois 61341 Telephone 815/357-6761 February 24, 1993 Director of Nuclear Reactor Regulation United States Nuclear Regulatory Comunission Mall Station F1-137 Washington, D.C.

20555 ATIN: Docuenent Control Desk Gentlessen:

Enclosed for your information is the annual report covering LaSalle County Nuclear Power Station for the period covering January, 1992 through December, 1992.

Very truly yours e

4x h Gary F. Spedl Station Manager LaSalle County Station Enclosure zc:

A. B. Davis, NRC, Region III D. E. Hil.is, NRC Resident Inspector LaSalle J. L. Roman, IL Dept. of Nuclear Safety B. Stransky, NRR Project Manager M. J. Wallace, Ceco D.

L. Farrar, CECO INPO Recor<. Center D.

R. Eggutt, NED P. D. Doverspike, GE Resident T. K. Schuster, Manager of Nuclear Licensing T. A. Rieck, Nuclear Fuel Services Manager J. E. Lockwood, Regulatory Assurance Supervisor W. P. Pietryga, QA/NS Off Site Review Station File 1-9303010254 921231

((

l DR ADOCK 0500 3

1

TABLE OF CGETENTS I.

INTRODUCTIGt II. ANNUAL REPORTABLE DOCUMENTATIGE FOR UNITS 1 AND 2 A.

Susunary of Operating Experience B.

Unit Outage or Power Reductions C.

Radiation Exposure D.

Indications of Failed Fuel Elesments E.

Tests and Experiments Not Covered in the Safety Analysis Report F.

Changes to Procedures Covered in the Safety Analysis Report G.

Sumanary of Changes to the Facility Which Are Described in the Safety Analysis Report B.

Summaary of Safety Related Modifications I.

Sminnary of ECCS Outages J.

Survey of Evaluation Results of Chlorine Shipsnents by Barge on the Illinois River K.

Susunary of Events Violating Technical Specification 3.4.5 Primary Coolant Iodine Spiking Exceeding Allowable Limits

I.

INTRODUC71CM The LaSalle County Nuclear Power Station is a two-Unit facility owned by Cormnonwealth Edison Company and located near Marseilles, Illinois.

Each unit is a Boiling Water Reactor with a designed net electrical output of 1078 Hegawatts. Maste heat is rejected to a man-nade cooling pond using the Illinois river for make-up and blowdown. The architect--engineer was Sargent and Lundy and the contractor was Commonwealth Edison Company.

Unit one was issued operating license number NPF-11 on April 17, 1982. Initial criticality was achieved on June 21, 1982 and comrnercial power operation was commenced on January 1, 1984.

Unit two was issued operating license number NPF-18 on December,16 1983. Initial criticality was achieved on March 10, 1984 and consnercial power operation was commenced on October 19, 1984.

This report was compiled by Michael J. Cialkowski, telephone number (815) 357-6761, extension 2427.

1 II. Annual Raportchle Documentation for Unit 1 and 2 A.

S umm ary_g f_Qpe rating _EFECLILCD C C I

The summary of the operating experience has been reported monthly in LaSa11e's NRC Monthly Reports (Section II.A) dated January 1992 through December 1992. For safety related maintenance (non-outage related) performed during the period of January 1992 thru December 1992, see Attachment A.

B.

Unit Outages _and Power _ Reductions For unit outages, see Attachment B.

For unit power reductions see Attachment C.

C.

Radiatinn_ErPOEur.e This information is reported annually for 1992 in the respective sections relating to numbers compiled for LaSalle Unit 1 and 2 in the 10 CFR 20.407 annual report submitted under a different cover.

D.

Jaidications of Failed Fuel Elements This section has been reported monthly in LaSa11e's NRC Monthly Reports (Section II.F.5) dated January 1992 through December 1992.

E.

• rests and Experiments not egycred in the Safety Analysis Report During this reporting period, January 1,1992 through December 31,1992 there were no tests or experiments conducted.

F.

ChADggs to Procedures Covered in the Safety Analysis Report LLP-91-069, Unit 2 Division II Temporary 125VDC Power Supply This special procedure provided a method of maintaining the Unit 2 Division II 125Voc bus energized and operable while its battery was being replaced in accordance with M-1-2-88-003.

This change allowed the Unit 2 Control Room Heating, Ventilation, and Air Conditioning, Standby Gas Treatment, and the Unit 2 Hydrogen Recombiner systems to remain operable to meet the Unit 1 LCO.

The installation of a temporary cable j

was non-seismic which did not meet Updated Final Safety Analysis Report, Section 3.1.2 requirements. The cable was safety related material with most of cable run on the floor and roped off.

Open fire doors required by procedure did not meet the fire protection requirements of Updated Final Safety Analysis Report, Section 9.4.1.

A continuous fire watch was established for the open fire doors. During the replacement, the reactor was defueled. A failure of this temporary system had been analyzed to be only as likely as a failure of the original system. The system was returned to normal after the battery replacement.

~

I F.

Ch;nges to Proc?durcs Cov2 red in th7 Snfety Analysis Report-(continued) l I

LLP-91-108 Unit 2 Division 3 Temporary DC Power During Battery / Charger Replacement This special procedure was written to provide DC power to Bus 213 while the Division 3 Battery and charger were replaced per modifications 1-1-90-009 and 1-1-90-010.

The procedure also allowed the new Division 3 battery and charger to be tested without de-energizing Bus 213.

The temporary charger was connected to Motor Control Center 232Y-2 cubical D3.

Updated Final Safety Analysis Report, Section 8.3.2, requires three independent class IE DC Power systems. Division 3 DC system did not meet the requirements of Updated Final Safety Analysis Report, Section 8.3.2, during this procedure. The temporary DC charger received AC power from Motor Control Center 232Y-2 which is fed from non-essential Bus 252.

The additional AC load to Motor Control Center 232Y-2 was found to be acceptable per ELMS-AC.

The system was returned to normal upon completion of the procedure.

t LLP-91-136, Defeating The Unit 2 Mode Switch To Shutdown Scram i

This special procedure was written to momentarily jumper-out the mode switch to the shutdown scram contacts while the mode switch was taken to the Shutdown position. This was done when all Control Rods were full-in and the jumpers were installed during the process of placing the mode switch to shutdown. This minimized " accumulator-pressurized" scrams during the refuel outage.

Such pressurized scrams with the Control Rods at position 00, place high stresses on the Control Rod Drive mechanisms and disturbs reactor water chemistry. This Special procedure constituted a temporary change to the facility and its operation as described in the Updated Final Safety Analysis Report, but did not increase the risk to the plant.

LLP-92-029, Reactor Core Isolation Cooling System Pump Operability And Valve Inservice Tests In Conditions 1, 2, and 3 This Special Procedure was developed to remove the In-Service Tests for 1E51-F019 and 1E51-F021, by changing the procedure for shutting down the system at the end of the operability tests. This was done because the motor function of the minimum flow valve was inoperable. All automatic functions for the valve were not available. System shutdown was to be accomplished without the use of the minimum flow line. The primary containment integrity for the minimum flow valve remained intact, since the valve failed closed, and the valve remains out-of-service closed. The minimum flow bypar. line to the suppression pool is provided to protect the pump during startup with no discharge valves open, shutdown performance evaluation, and/or testing. This protects the pump from the overheating that would be caused by discharging to a deadhead. The flow path for pump cooling during startup or shutdown of the system was still provided.

l i

F.

Changes to Proradures Covared if the Snfaty Analysis _R2ppI_t-(continusd)

LLP-92-032, Reactor Core Isolt. tion Cooling System Cold Quick Start In Condition 1, 2, and a With 1E51-F019 0.O.S.

Closed.

This Special Procedure changed th0 procedure of shutting down the system at the end of the cold quick start.

The motor function of the minimum flow valve was inoperable.

All automatic functions for the valve were not available. System shutdown was accomplished without the use of the minimum flow line. The minimum flow bypass line to the suppression pool was provided to protect the pump during startup with no discharge valves open, and shutdown during performance evaluation and/or testing. The flow path for pump cooling during startup or shutdown of the system was provided using this procedure.

The operability status of the Reactor Core Isolation Cooling System during normal unit operation condition was not compromised. The system was returned to normal following repairs to the 1E51-F019 valve motor.

LLP-92-055, RHR System Radiation Flush This procedure allowed the flushing of various piping lines in the Residual Heat Removal System. The Residual Heat Removal Steam Condensing steam lines were filled with water, the "A" and "B" Residual Heat Removal loops were crosstied, and flow was diverted through the steam lines in the reverse direction in an attempt to reduce the general area radiation dose rates.

This procedure operated the Residual Heat Removal system in a configuration not previously approved or evaluated in the Updated Final Safety Analysis Report. This procedure was performed when the reactor was defueled and the Residual Heat Removal system was not required to be operable in any mode.

The system was returned to normal upon completion of the work.

LLP-92-072, Controlled Start of the Reactor Core Isolation Cooling System in the CST Test Mode I

with IE51-F019 Out-of-Service Closed.

This Special Procedure was developed to modify procedure LOP-RI-06, Revision 10, to use the IE51-F022 and 1E51-F059 valves as a minimum flow path.

This was to provide minimum flow with valve IE51-F019 out-of-service.

The use of valves 1E51-F022 and IE51-F059 as a minimum flow path was previously evaluated by LLP-92-029, Reactor Core Isolation Cooling (RCIC)

System Pump Operability And Valve Inservice Tests In Conditions 1, 2, and 3.

l LLP-92-082, Temporary DC Power for Relay House System 1 This procedure allowed temporary DC power to be provided to the Relay House System I so that maintenance on the battery and charger could be i

performed. The temporary battery's capacity was rated at 100 amp-hours instead of the 200 amp-hours described in the Updated Final Safety Analysis Report, Section 8.1.2.5.

The temporary System I battery had less capacity, i

and therefore would not have been able to maintain loads if the temporary charger should have become deenergized. The switchyard is non-safety related.

I and is not required for reactor safety.

l l

l l

l i

F.

Changes to ProceMgIss Covered in the Sifety Analysis Report-(continued)

LLP-92-101, Control Blade Pin And Roller g

Replacement Demonstration Procedure This Special Procedure describes the assembly, operation, and disassembly of the ABB Equipment required for the removal of the pin and roller and installation of the replacement button for the control rod blades. This process took place in the fuel pool cask area with everything secured to the area walls.

The Control Rod Blade selected for the demonstration was previously damaged and was to be returned to its original storage area afterwards. The purpose of the demonstration was to replace the stellite pin and rollers on the control blades with stainless steel buttons to reduce the amount of activated cobalt released to the plant.

Considerations were given for chemical compatibility of the various process fluids hnd components, filtering of the process residue, and storage of the components af terwards.

All measures for protection of the individuals and the site were followed to ensure radioactive material control.

LLP-92-121, Defeating The Unit 1 Mode Switch To Shutdown Scram This special procedure momentarily jumpered-out the mode switch to the shutdown scram contacts while the mode switch was taken to the Shutdown position. This was only done when all Control Rods were full-in and the jumpers were installed during the process of placing the mode switch to shutdown.

This minimized " accumulator-pressurized" scrams during the refuel outage. Such pressurized scrams with the Control Rods at position 00, place high stresses on the Control Rod Drive mechanisms and disturbs reactor water chemistry. This Special procedure constituted a temporary change to the facility and its operation as described in the Updated Final Safety Analysis Report, but did not increase the risk to the plant.

LLP-92-143, Unit 1 Division 3 Temporary DC Power During Battery / Charger Replacement This special procedure provided DC power to Bus 113 while the Division 3 Battery and charger were replaced per modifications 1-1-90-011 and 1-1-90-012.

The procedure also allowed the new Division 3 battery and charger to be tested without de-energizing Bus 113.

The temporary charger was connected to Motor Control Center 132B-1 cubical E6.

The Moisture Separator Reheater 2nd Stage Reheat Steam High Load Valve RSHLV1 was disconnected to allow the temporary charger to be connected to Motor Control Center 132B-1 cubical E6.

The Moisture Separator Reheater 2nd Stage Reheat Steam High Load Valves were closed. The Moisture Separator Reheater 2nd Stage Reheat is not required during plant operations and was not currently being used.

The Updated Final Safety Analysis Report, Section 8.3.2 requires three independent class IE DC Power systems. Division 3 DC system did not meet the requirements of the Updated Final Safety Analysis Report, Section 8.3.2 during this procedure.

The temporary DC charger received AC power from Motor Control Center 132B-1 which is fed from non-essential Bus 152.

The additional AC load to Motor Control Center 132B-1 was found to be acceptable per ELMS-AC.

The system was returned to normal upon completion of the procedure.

l

F.

Changes to Procedums_Cnygied_Jn_t.be Snfety Annivsia_Empar_t-(continued)

LLP-93-014, Temporary DC Power for Relay House System 2 This procedure will allow temporary DC power to be provided to the Relay House System 2 so that maintenance on the battery and charger can be performed. The temporary battery's capacity will be rated at 100 amp-hours instead of the 200 amp-hours described in the Updated Final Safety Analysis Report, Section 8.1.2.5.

The temporary System 2 battery will have less capacity, and therefore will not be able to maintain loads as long if the temporary charger should become deenergized. The switchyard is non-safety related and is not required for reactor safety.

LST-92-086, Relay House Battery BITE Test This procedure provided a methodology for passing a 60 Hz AC signal through the Relay House Battery and measuring the AC voltage drop across each cell.

The cell impedance is calculated from the measured voltage drop.

The measurement of cell impedance will help in determining the adequacy of each cell.

This procedure attaches an AC signal source to a non-safety related battery. The Updated Final Safety Analysis Report does not address this AC signal added to the battery, therefore this is a test.

The loss of either relay house battery would not cause a loss of offsite power per Updated Final Safety Analysis Report, Section 8.1.2.5.

The redundant protective relaying system protected faulted equipment.

Operation of the battery being tested was unaffected by this test.

Battery low impedance will prevent AC circulating

)

currents from flowing into the loads or charger. Updated Final Safety l

Analysis Report, Section 15.2.6 covers the loss of AC power, which bounds the failure of the switchyard. This test did not affect the ability of the battery to perform its designed function.

LST-92-088, Division 2 Battery Charger 1(2)DC16E Capacity Test This procedure provided the methodology for performing an eight hour capacity test on battery charger 1(2)DC16E. The battery charger was inspected to verify proper operation. The charger 1(2)DC16E was tested while charger 1(2)DC17E maintained the Division 2 loads. This procedure verified that the battery charger 1(2)DC16E met the Technical Specification surveillance requirements for an operable Division 2 Battery charger. The loading on the 1A(2A) Diesel Generator per Updated Final Safety Analysis Report, Table 8.3-1, was increased. An additional 10 kW of load was added to bus 136X-3(236X-3) while the capacity test was being performed.

Battery charger 1(2)DC16E was isolated from the Division 2 DC Bus 3B(2B) by a class IE breaker. All Technical Specifications relating to Division 2 AC and DC systems were adhered to during the test.

The additional load for 136X-3(236X-3) was found to be acceptable per ELMS AC.

The additional load on the 1A(2A) Diesel Generator was acceptable. The Division 2 battery was unaffected by this test.

l l

t F.

Changes to Procedures Covered in the Safety Analysis R' port-(continu:d)

LST-92-119, Startup, Testing And Shutdown of the Unit 1 VT Evaporative Cooler This Special Test Procedure is for the startup, testing and shutdown of the Unit 1 Turbine Building Ventilation Evaporative Coolers during warm weather. This procedure will be used to allow the station to evaluate the feasibility of operating the Turbine Building Vent!.lation evaporative Coolers during warm weather. The Updated Final Safety Analysis Report indicates that the evaporative coolers are supplied with domestic water.

This change will allow the use of MC water to supply the evaporative coolers. The evaporative i

coolers have no impact on plant safety.

LST-92-126, Division 2 Battery Charger 1(2)DC16E Capacity Test This Special Test Procedure provided the methodology for performing an eight hour capacity test on battery charger 1(2)DC16E. The battery charger was inspected to verify proper operation.

Charger 1(2)DC16E was tested while charger 1(2)DC17E maintained the Division 2 loads. This verified that battery charger 1(2)DC16E met the Technical Specification surveillance requirements for an operable Division 2 Battery Charger. This maintained the Division 2 DC system operable while testing the battery charger 1(2)DC16E. The loading on the 1A(2A) Diesel Generator per the Upde.ted Final Safety Analysis Report was increased. An additional 10 kW of load was added to bus 136X-3(236K-3) while the capacity test was being performed.

Battery Charger 1(2)DC16E was isolated from Division 2 DC Bus 1B(2B) by a class 1E breaker, therefore the operation of charger 1(2)DC16E did not affect Division 2 DC.

All Technical Specifications relating to Division 2 AC and DC were adhered to during this test.

LST-92-128, MSIV Leakage Control System Pressure Test For The 2E32-F009 Valve The purpose of this special test procedure was to pressure test the 2E32-F009 Main Steam Isolation Valve Leakage Control Outboard Depressurization Valve.

The 2E32-F008 valve was opened during startup at approximately 35 psi steam pressure in order to allow the 2E32-F009 valve to experience the main steam line pressure for visual inspection. Normal operation of the outboard leakage control system does not allow the valves to open above 35 psi main steam line and reactor vessel pressure.

The system was returned to a normal configuration after the test.

F.

Changes to Prpecedures Cover _ed_in_the_S h ty_Analylin_t Part-(continusd)

LST-92-134, RHR System Pressure Response With Water Leg Pump Isolated This special procedure was used to record the system pressure response of i

the 1A Residual Heat Removal loop with the waterleg pump isolated. Response was recorded by use of a strip chart recorder connected across the inputs to control room indicator 1E12-R505A.

The performance of this procedure required that the waterleg pump be isolated, which differs from the system description in the Updated Final Safety Analysis Report.

This procedure resulted in a single loop being declared inoperable, which is an analyzed condition. This procedure would have be exited and the system would have be filled and vented immediately if a low pressure alarm condition was created.

i LST-92-141, Verification Of Main Turbine Shut-Of f Valve Setpoints This special test procedure is to outline a method to set the shut-off valve for the main turbine intermediate stop valves. The Electro-Hydraulic Control will be placed on line and the turbine reset per station procedures.

This procedure also governs the adjustment of system pressure and defeats low pressure trips. The resetting of the turbine normally opens the Intermediate Stop Valves. This procedure adjusts the Electro-Hydraulic Control pressure

)

and will control the opening of these valves. The actions described in this procedure have no impact on reactor safety while the unit is shutdown.

Electro-Hydraulic Control is non-safety related and is not required with the unit down.

The low hydraulic turbine trips are defeated to allow for low pressure testing of the Electro-Hydraulic Control trip system. The reactor is in cold shutdown and the Electro-Hydraulic Control system is being operated within its design capability.

LOP-AP-08, Removing System Auxiliary Transformer Sat 142 (242)

From Service With Unit 1 (2) In Shutdown This is a temporary change to this procedure to allow livening 2W Main Power Transformer without auto deluge available. The fact that the i

transformer does not have auto deluge is in conflict with Section 9.5.1.2.3 of the Updated Final Safety Analysis Report. Manual Deluge is still available as is portable fire equipment. The Deluge Valve is currently out-of-service closed due to an identified problem with the deluge clapper. The system will be returned to normal following repairs.

LOS-CS-01, Secondary Containment Damper operability Test The change to procedure LOS-CS-01 added guidance for the installation of a jumper to bypass the Reactor Water Cleanup Area Temperature isolation signals for up to I hour while the Reactor Building Ventilation system is shutdown and isolated. The use of the jumper is limited to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, which is the time required by the Technical Specifications to return the system to normal. The system will be returned to normal upon completion of this surveillance.

i l

F.

Changes to Procedures Covsred in th7 S2fety Aanlysis Report-(continued)

LOP-NB-01, Reactor Vessel Leakage Test This procedure change provided steps to allow the performance of Scram Insertion Timing during the vessel leakage test.

This change also provided steps that provide guidance if a full reactor scram occurs during scram timing. During this evolution, the reactor will be in Mode 3 or 4, and the Reactor Mode Switch will be in Refuel with the head tensioned. The accumulators are not required. This procedure can trap hydro pressure in the accumulators and mask failure to achieve full travel of the accumulator piston.

Controls are established in the scram timing procedure to ensure detection of over-charged accumulators which do not bottom out the accumulator piston. The accumulators are charged and operable prior to pulling a rod for scram time testing.

LOP-RP-04, RPS Bus B Transfer The change to this procedure added a caution that taking the Group 1 Condenser Low Vacuum Bypass Keys from Bypass to Normal will/can cause a Group 1 Isolation with low main condenser vacuum present. This change is in accordance with the original intent of the procedure. There will be no effect with the Condenser Low Vacuum Eypass Keys in Bypass because the Group 1 Primary Containment Isolation System is interlocked to only be bypassed on low vacuum if 1) the mode switch in not in run, 2) the turbine stop valves are closed, and 3) reactor vessel pressure is less than 1043 psig.

This change is consistent with the Updated Final Safety Analysis Report and the Technical Specifications.

LOP-WE-01, Waste Collector Tank Processing To The Waste Sample Tank The change tv this procedure eliminated the requirement for sampling prior to processing. Sampling in the waste collector is not required, by the Updated Final Safety Analysis Report, to ensure processed water will be of cycled condensate quality. All water transferred into the collector from other tanks has been previously sampled and normal inputs are expected to be of a quality that allows processing to bring water chemistry within the specifications described in LCP-110-1, Chemical Analysis and Corrective Action Schedule, for cycled condensate.

A short recirculation and monitoring of the in line conductivity cell will still be required so abnormal inputs should be noticed prior to processing. A precaution was added to monitor the Waste Filter differential pressure and Demineralizer outlet conductivity for abnormal trends to protect these components from abnormal input.

LCP-110-1, Chemical Analysis and Corrective Action Schedule, limits will be enforced for cycled condensate so cycled condensate chemistry is ensured.

F.

Th_ang's to Procedurss Covered in the Srfaty Annlysis R2 port-(continusd)

LOP-WR-02, Startup and Operation of the Reactor Building Closed Cooling Water System This precedure change authorized the startup of the Reactor Building Closed Cooling Water system without Service Water, in unusual circumstances.

The system is normally operated with Service Water in operation and functional.

Extraordinary circumstances may require the Shift Engineer to waive the prerequisite to have Service Water in operation with Reactor Building Closed Cooling Water Heat Exchangers valved in.

The Reactor Building Closed Cooling Water system is not an Engineered Safety Feature and does not have a Safety Design Basis. This procedure change allows Reactor Building Closed Cooling Water system recovery before Service Water is restored. During extended Unit Shutdown, normal plant heat loads are not present, and operation without Service Water is permissible.

LOP-VG-02, Shutdown of the Standby Gas Treatment System (SBGT) i

]

This procedure revision changed the minimum run time of the Standby Gas Treatment Train from 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> with the Standby Gas Treatment Wide Range Gas Monitor sample flow at or above 1.3 cubic feet per minute. The 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of operation per month of the Standby Gas Treatment Train with the heaters on to dry out the charcoal filter is satisfied by LOS-VG-M1, Chemistry supervision has verified the minimum sample requirements can be met with 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of operation with the Standby Gas Treatment Wide Range Gas Monitor sample flow rate at or above 1.3 cubic feet per minute. This revision did not change how the equipment operates, the alignment of the equipment, or the way the equipment is operated. This revision is an a@ministrative change to an administrative limit that does not alter the design or function of any system.

a j

LaSalle OnSite Review 92-039 This LaSalle OnSite Review determined the required review (whether Technical Review or Onsite Review) of each type of procedure. The basis for determination is Technical Specification amendment 86 (U-1) and 70 (U-2), and Regulatory Guide 1.33, Rev.

2.

Each Department Head signature indicates concurrence with the evaluation of the department's procedures. Many procedures that previously required onsite review will now require Technical Review per the referenced Technical Specification amendment. Technical Reviewers must meet the same qualifications as an Onsite Review participant, but only those reviewers required by the procedure change / content are required 4

to review a department's procedures. Technical Review requires only one reviewer, other than the writer and Department Head; an Onsite review requires a minimum of 2 reviewers. Technical Review is required to evaluate the need for either additional review disciplines or Onsite Review as needed, based on the particular change being made to a given procedure.

l

i F.

Chaages to Procedures Covarrd in th?,_Er,1cty AnMysis Reparl-(continu d)

~

LaSalle OnSite Review 92-041, Revision 0 This LaSalle OnSite Review dealt with a Temporary Waiver of Compliance for Unit 2 Technical Specifications 3.6.3, Action Statement a.2, which requires that Unit 2 be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This is required because the Primary Containment Isolation Valve 2G33-F040, Reactor Water Cleanup Return Isolation l

Valve, is inoperable due to an inadequate Local Leak Rate Test.

The 2G33-F040 valve requires a satisfactory Type C test in order to be an operable Primary Containment Isolation Valve.

The Type C test (Local Leak Rate Test) was determined to be performed with a potentially inadequate vent path during the test.

The Updated Final Safety Analysis Report takes credit for the 2B21-F010A&B and 2B21-F032A&B Feedwater check valves, not the 2G33-F040 Reactor Water Cleanup Return Isolation Valve, which affects only a design leakage barrier (long term leakage). This test is to be performed during the next Unit 2 cold shutdown which has a duration of 2 weeks or more, i

LaSalle OnSite Review 92-042, Revision 0 This LaSalle OnSite Review dealt with a Technical Specification amendment request to Unit 2 Technical Specification 3.6.3 for valve 2G33-F040, Reactor Water Clean-Up Return to Feedwater Stop Valve.

This change will add a footnote to Table 3.6.3-1 for valve 2G33-F040, waiving the requirement for the Type C test to be current.

It will also allow for its leakage not to be included in the totals for Type B and C Containment leakage, as required by Technical Specification 3.6.1.1 and 3.6.1.2, for the remainder of the Unit 2 Cycle 5, or until the first outage in which Unit 2 is in Cold Shutdown for a duration of two weeks or more.

The 2G33-F040 valve requires a satisfactory Type C test in order to be an operable Primary Containment Isolation Valve.

The Type C test (Local Leak Rate Test) was determined to be performed with a potential inadequate vent path during its last test.

The non-conservative Type C test method for the 2G33-F040 valve puts in question the Containment Maximum Path Leakage Limit of 0.6 L and thus the amount of radioactive a

effluent that could be released during analyzed accidents. The path used in its last test, Valves 2G33-F040 and 2G33-F039, is likely to provide a leakage barrier if needed. With no credit taken for the 2G33-F040 valve, the containment leakage through the feedwater lines still meets Containment Leakage Criteria in Appendix J in the event of a single failure.

i

G.

Eummary of Changer,_to the Fecility Which cro Drsribed_itt,he Halc1LAnalysis_ Report Temporary System Change 1-1719-88 The Unit 1 Off Gas High Flow alarm setpoint was changed from 20 scfm to 80 scfm.

The High High Flow alarm was bypassed.

The present Off Gas Flow is 58 scfm, rendering the High Flow alarm useless and the alarm is a nuisance in the Control Room.

The Updated Final Safety Analysis Report states that the High Flow setpoint is 20 scfm and the High High Flow alarm setpoint is 256 scfm.

Neither alarm is Technical Specification related. The Off Gas Post Treatment Radiation Monitor isolates valve IN62-F057, Off Gas Discharge to the Stack, upon excess radiation release.

The flow rate was evaluated to be acceptable per LTP-900-1.

Raising the setpoirt for the High Flow alarm will not affect the safety function of the Off Gas Post Treatment Radiation Monitor. A Modification and Setpoint Change Request were initiated to eliminate this Temporary System Change.

Temporary System Change 1-0090-92 This Temporary System Change defeated the trips of the Unit 2 "A" and "B" Control Room Ventilation Radiation Monitors, each one on a different occasion. This was done by lifting leads at panel OPM15J to prevent an auto start of the "B" Emergency Make Up Train due to spurious instrument spikes.

These spurious spikes resulted in Engineered Safety Feature equipment actuation. The "A" and "B" Moniters still indicate and the F.mergency Makeup trains could still start from a high radiation signal from the other detectors. This change was still in compliance with Technical Specifications since the minimum number of channels was still operable and the related equipment was still operable. The "A" and "B" Radiation Monitors were returned to service after repairs.

i Temporary System Change 1-0335-92 This Temporary System Change defeated the Unit 1 Drywell Floor Drain Sump Trouble Alarm. This was done by defeating R-point 0065 for the Drywell Floor Drain Sump Trouble Alarm to prevent spurious nuisance alarms. These alarms were caused by the failing of the Drywell Floor Drain Sump Fillup Rate Transmitter. The Drywell Floor Drain Sump inleakage was still monitored by floor drain totalizers which backup the Drywell Floor Drain Sump Fillup Rate Transmitter alarm. The Control Room monitored the totalizers every four hours. In addition, there were other Containment instruments still available with their alarm units.

The Drywell Floor Drain Sump Trouble Alarm was returned to service after repairs.

C.

fiummsrv of Changes to the Freility Which nre described in th's Safety Analysis Report-(Continuedl Temporary System Change 1-0386-92 i

This Temporary System Change is to defeat the trip of the Unit 2 "C" Control Room Ventilation Radiation Monitor. This was done by lifting a lead j

at panel OPM15J to prevent an auto start of the "B" Emergency Make Up Train j

l due to spurious instrument spikes. These spurious spikes resulted in j

Engineered Safety Feature equipment actuation. The "C" Monitor still l

indicates and the Emergency Makeup train can still start from a high i

radiation signal from the other detectors. This change was still in compliance with Technical Specifications since the minimum number of channels was still operable and the related equipment was still operable.

The "C" Radiation Monitor was returned to service after repairs.

l Temporary System Change 1-0402-92 This Temporary System Change installed a switched jumper at the IFPO4JA i

panel to allow for taking detection Zone 1-33 out-of-service during welding l

operations for Work Request L71269 to prevent false alarms. The fire I

j detection zone 1-33 was only out-of-service when work was being performed for work request L71269. When zone 1-33 was out-of-service, a continuous fire watch was provided. The fire loading in the detection zone did not increase and manual fire protection equipment wis still available.

The area was returned to normal af ter the completion of welding.

Temporary System Change 1-0404-92 This Temporary System Change defeated the trip of the Unit 1 "D" Control Room Ventilation Radiation Monitor. This was done by lifting a lead at panel DPM14J to prevent an auto start of the "A" Emergency Make Up Train due to spurious instrument spikes.

These spurious spikes resulted in an Engineered Safety Feature equipment actuation.

The "D" Monitor still gave j

indication and the Emergency Makeup train could still start from a high radiation signal from the other detectors. This change was still in compliance with Technical Specifications since the minimum number of channels was still operable and the related equipment was still operable.

The "D" Radiation Monitor was returned to service after repairs.

l Temporary System Change 1-0565-92 This Temporary System Change installed a switch jumper to bypass the Unit 1 "A" and "B" Reactor Recirculation Flow Control Valve Actuator Drain Alarm. The Unit 1 "A" and "B" Reactor Recirculation Flow Control Valve Actuator Drain alarms were up solid and masking other alarm signals.

By bypassing these alarms the other alarms could initiate the annunciator in l

the control room identifying other problems. The alarm signal is initiated from a leaky actuator. The system was returned to normal af ter repairs.

G.

Summary _of Changes to the_f0Cility Which cre described in th2 Sainty Analysis Report-(continug41 Temporary System Change 1-0609-92 This Temporary System Change prevented moisture intrusion into the Standby Gas Treatment Wide Range Gas Monitor by temporarily adding a mechanical moisture separator to the sample line. The Standby Gas Treatment Wide Range Gas Monitor stack particulate and Iodine sampling was inoperable with this Temporary System Change in place. The Noble Gas activity monitor was also inoperable with this Temporary System Change in place. The Wide Range Gas Monitor does not have a safety function as it is only a monitoring system. Radioactive release was not affected by this Temporary System Change. Sampling was still maintained in accordance with the Technical Specifications.

Temporary System Change 1-0617-92 This Temporary System Change installed switch jumpers at the 1FPO4JA panel for Detection Zones 1-30, 1-31, and 1-32 to aid in taking detectors out-of-service during welding operations in these detection zone. Welding in the vicinity of the in-service detector could result in false alarm at the fire protection control panels. A one hour fire watch was maintained in the area while the detection zones were out-of-service.

A continuous fire watch was posted when welding, cutting, or grinding was occurring. This Temporary System Change was removed upon completion of the welding and grinding in these fire detection zones.

Temporary System Change 1-0634-92 This Temporary System Change prevented moisture intrusion into the i

Standby Gas Treatment Wide Range Gas Monitor by temporarily adding a mechanical moisture separator to the sample line. The Standby Gas Treatment Wide Range Gas Monitor stack particulate and Iodine sampling was inoperable with this Temporary System Change in place. The Noble Gas activity monitor was also inoperable with this Temporary System Change in place. The Wide Range Gas Monitor does not have a safety function as it is only a monitoring system. Radioactive release was not affected by this Temporary System i

Change. Sampling was still maintained in accordance with the Technical Specifications.

+

,r-

G.

Summuy_of Changes to the_Ircility Which nre described in the Safety _ Analysis Report-(continsedl Temporary System Change 1-0648-92 This Temporary System Change defeated the trip of the Unit 1 "A" Control j

Room Ventilation Radiation Monitor. This was done by 13fting a lead at panel OPM14J to prevent an auto start of the "A" Emergency Make Up Train due to spurious instrument spikes. These spurious spikes resulted in an Engineered Safety Feature equipment actuation.

The "A" Monitor still gave indication and the Emergency Makeup train could still start from a high radiation signal from the other detectors. This change was in compliance with Technical Specifications since the minimum number of channels was still operable and the related equipment was still operable. The "A" Radiation Monitor was returned to service after repairs.

Temporary System Change 1-0681-92 This Temporary System Change plugged various floor drains on the Turbine Deck 768' elevation.

This assisted in keeping debris out of the floor drains system during the turbine outage.

These drains are not a flood protection barrier and not all drains in the area were plugged. The floor drains were returned to normal upon completion of the turbine outage.

Temporary System Change 1-0687-92 This Temporary System Change installed water supply piping and pumps for mechanically cleaning the Main Condenser tubes.

The water supply was tapped from IWS158, Generator Hydrogen and Stator Winding Cooler Influent Low Pressure Drain. The Hydrogen coolers were inoperable during this period, but this was acceptable since the Hydrogen Coolers were in an outage condition. The system was returned to normal following this evolution.

Temporary System Change 1-0693-92 This Temporary System Change defeated the trip of the Unit 1 "D" Control Room Ventilation Radiation Monitor. This was done by lifting a lead at panel OPM14J to prevent an auto start of the "B" Emergency Make Up Train due to spurious instrument spikes. These spurious spikes resulted in an Engineered Safety Feature equipment actuation. The "D" Monitor still gave indication and the Emergency Makeup train could still start from a high radiation signal from the other detectors. This change was in compliance with Technical Specifications since the minimum number of channels was still operable and the related equipment was still operable. The "D" Radiation l

Monitor was returned to service after repairs.

I

G.

Summ ny of Chamges to the._Eccility_Which ere described in the Hafety Analysis Report-(continued)

Temporary System Change 1-1112-92 1

This Temporary System Change allowed the OVD04Y, OVD05Y, DVD40Y, and 1

OVD41Y ventilation dampers, located within the "0" Diesel Generator Room, to be wired in the open position. These dampers were wired open to allow for the "0" Diesel Generator to be operated with Carbon Dioxide Fire Protection System inoperable. Normally the "0" Diesel Generator Room Ventilation dampers would be closed by its electro-thermal links, if the Carbon Dioxide system was initiated.

These electro-thermal links required replacement.

The dampers were wired open to allow repairs to these electro-thermal links j

at the same time that the "0" Diesel Generator was undergoing maintenance i

runs.

A fire watch was put in place until the Temporary System Change was

]

removed.

Temporary System Change 1-0027-93 This Temporary System Change wired the IVXO8Y Recirculation damper in the open position after being disconnected from the actuator. The Operating Department adjusted the damper manually as needed per the Shif t Engineer.

This system was still able to perform its function. This Temporary System Change was removed when the repairs were completed.

Temporary System Change 2-0214-92 This Temporary System Change utilized switch jumpers to jumper out the trip signals from the Reactor Building Ventilation and Fuel Pool Ventilation Radiation Monitors. This prevented a Group IV Primary Containment Isolation System isolation during replacement and maintenance of the 2D18-K609A, B, C, and D and 2D18-K615A, B,

C, and D detectors. The High Radiation Trip Signal was bypassed on the channel whose detector is being replaced via a switch type jumper. While the jumpers were installed and the switches are c1csed, no irradiated fuel moves or core alterations, or operations that present the possibility for draining the reactor vessel took place. The system was restored to normal af ter the work was complete.

I l

l l

l

G.

Epmmyy_of__ Changes to tbp_Rqility Which cre denctibed in ths Sainty_ Analysis Report-(EDAtinued)

Temporary System Change 2-0403-92 This Temporary System Change will wire the breaker contacts off the 2B Diesel Generator output breaker, ACB 2433, in series with the power directional relay, 2E22B-K32. The Power Direction Relay is tripping the 2E22B-K1 Lockout Relay when the 2B Diesel Generator is coasting down after a shutdown. This trip is due to an adjustment made to the VAR protection supplied by the relay that affects the frequency setting of the relay.

By wiring the breaker contacts in, the relcy will be disabled when the breaker is open.

The addition of the auxiliary breaker contacts into the Reverse Power relay logic will prevent the relay from energizing unless the 2B Diesel Generator is synchronized to the power grid. The Updated Final Safety Analysis Report, Table 8.3-2, item 3i states that the Reverse Power relay will send a trip signal to ACB 2433.

With this change installed, the Reverse Power relay will send a trip signal to ACB 2433 only when ACB 2433 is closed such as described for the Loss of Field relay.

This change does not affect the operation of the 2B Diesel Generator during the time it is required by Technical Specifications to function in the event of an emergency.

Temporary System Change 2-0412-92 This Temporary System Change disconnected the sight glass on the "2B" Reactor Recirculation Pump Below Seat Drain Line to facilitate draining. A Chicago fitting with a blank flange was installed upstream and a blank flange was installed downstream.

The sight glass was believed to be plugged. This will facilitate draining of the Reactor Recirculation system and troubleshooting of the sight glass. The reactor was defueled, the fuel pool gates were installed, and the "2B" Reactor Recirculation system was isolated. The system was returned to normal upon completion of the work.

Temporary System Change 2-0776-92 This Temporary System Change installed a switch jumper to bypass the Unit 2 "A" Reactor Recirculation Flow Control Valve Actuator Drain Alarm.

The Unit 2 "A" Reactor Recirculation Flow Control Valve Actuator Drain alarms were up solid and masking other alarm signals. By bypassing these alarm s, the other alarms can initiate the annunciator in the control room identifying other problems. The alarm signal was initiated f rom a leaky actuator. The system was returned to normal after repairs.

G.

Summary of Chan_ges to the Freility Which nre described in thm Sately_ Analysis Report-f continuedl Temporary System Change 2-0812-92 This Temporary System Change defeated the trip of the Unit 2 "D" Control Room Ventilation Radiation Monitor. This was done by lifting a lead at panel OPM15J to prevent an auto start of the "B" Emergency Make Up Train due te spurious instrument spikes. These spurious spikes resulted in an Engineered Safety Feature equipment actuation. The "D" Monitor still gave indication and the Emergency Makeup train could still start from a high radiation signal from the other detectors. This change was in compliance with Technical Specifications since the minimum number of channels was still operable and the related equipment was still operable.

The "D" Radiation Monitor was returned to service after repairs.

Temporary System Change 2-0867-92 This Temporary System Change defeated the trip of the Unit 2 "A" Control Room Ventilation Radiation Monitor.

This was done by lifting a lead at panel OPM15J to prevent an auto start of the "B" Emergency Make Up Train due to spurious instrument spikes. These spurious spikes resulted in an Engineered Safety Feature equipment actuation. The "A" Monitor still gave indication and the Emergency Makeup train could still start from a high radiation signal from the other detectors.

This change was in compliance with Technical Specifications since the minimum number of channels was still operable and the related equipment was still operable. The "A" Radiation Monitor was returned to service after repairs.

Temporary System Change 2-0908-92 This Temporary System Change defeated the trip of the Unit 2 "B" Control Room Ventilation Radiation Monitor. This was done by lifting a lead at panel OPM15J to prevent an auto start of the "B" Emergency Make Up Train due to spurious instrument spikes. These spurious spikes resulted in an Engineered Safety Feature equipment actuation.

The "B" Monitor still gave indication and the Emergency Makeup train could still start from a high radiation signal from the other detectors. This change was in compliance with Technical Specifications since the minimum number of channels was still operable and the related equipment was still operable.

The "B" Radiation Monitor was returned to service after repairs.

G.

Summary _pLChnoges to the Freility Wilich era dfsgribed 11 tb3 S_Ofety_ Analysis _ Report-(snatinuedl Temporary System Change 2-0910-92 This Temporary System Change defeated the trip of the Unit 2 "C" Control Room Ventilation Radiation Monitor. This was done by lifting a lead at panel OPM15J to prevent an auto start of the "B" Emergency Make Up Train due to spurious instrument spikes. These spurious spikes renulted in an Engineered Safety Feature equipment actuation. The "C" Itonitor still gave indication and the Emergency Makeup train can still start. from a high radiation signal from the other detectors. This changearas in compliance with Technical Specifications since the minimum number of channels was still operable and the related equipment was still operable. The "C" Radiation Monitor was returned to service after repairs.

Temporary System Change 2-0969-92 This Temporary System Change installed an oil pressure gauge in the supply line to pressure switches 2PS-TO-018A(B), and also an oil pressure gauge in the supply line to pressure switches 2PS-TO-019A(B).

The gauges were connected to the pennanently installed tees in place of the installed plugs. An isolation valve was installed between the tee and the gauge on each of the supply lines. These gauges monitor Turbine Oil Pressure while i

operating. The system was returned to normal after this check was completed.

Temporary System Change 2-1006-92 This Temporary System Change is to install a switch jumper to bypass Unit 2 "A" and "B" Reactor Recirculation Flow Control Valve Actuator Drain Alarm.

The Unit 2 "A" and "B" Reactor Recirculation Flow Control Valve Actuator Drain alarms are up solid and masking other alarm signals.

By bypassing these alarms, the other alarms can initiate the annunciator in the control room identifying other problems.

The alarm signal is init3ated from a leaky actuator. The system will be returned to normal after repairs.

Temporary System Change 2-1007-92 This Temporary System Change is to defeat the trip of the Unit 2 "B"

Control Room Ventilation Radiation Monitor. This was done by lifting a lead at panel OPM15J to prevent an auto start of the "B" Emergency Make Up Train due to spurious instrument spikes. These spurious spikes resulted in an Engineered Safety Feature equipment actuation. The "B" Monitor still indicates and the Emergency Makeup train can still start from a high radiation signal from the other detectors. This change was still in compliance with Technical Specifications since the minimum number of channele was still operable and the related equipment was still operable.

I The "B" R6diation Monitor was returned to service after repairs.

1

k r

f i

j G.

Summery _ni_Ctt;moes to the Faci}ity %Dtich ere described int _the i

Safety Analysis Report-(continuedl I

i l

Temporary System Change 2-0015-93 l

1 a

This Temporary System Change will install switch jumpers to allow either j

one or both Drywell Equipment Drain Sump Transfer Pumps to be run j

continuously to provide additional cooling.

The installation of the Switch l

Jumper will bypass the High Temperature Interlock of valve 2RE035, to allow l

j emergency sump pump down with a high temperature in the sump.

Section i

9.3.3.2.1 of.the Updated Final Safety Analysis Report states that the sump l

1 pumps auto start on high sump level. This will not be true while the switch f

jumper is installed and closed. This condition will exist until the l

automatic temperature control circuit can be repaired.

l 5

l Safety Evaluation, Construction Of Temaorary

[

(

Wooden Duct In U-2 Steam Tunnel f

i i

This Safety Evaluation was to allow the construction of a temporary wooden duct immediately upstream of damper 2VT79YC to test the resistance j

j characteristics of 2VT79YC when enclosed in ductwork. The construction of l

this duct increased the fire loading in the affected zone above those 4

j allowed for in the area, even though this additional loading is below a l

negligible fire load as defined by tLe Updated Final Safety Analysis Report, j

Section H.l.1.h.

The wooden duct is to be removed upon completion of the 5

test.

i l

i i

Safety Evaluation, Facilities Improvement Program j

Security Fence Relocation j

This Safety Evaluation was performed for rerouting the security fence-i 1

and to install the intrusion and alarm assessment equipment for the new Main Access Facility. The new equipment was fully compatible with the existing

[

security system. The temporary fence relocation was to allow for the construction of the new Main Access Facility outside the protected area.

j This change required the LaSalle Security Plan to be revised. This change j

to the security system was not described in the Updated Final Safety Analysis Report, and does not affect anything provided by the Updated Final Safety Analysis Report.

l

~!

l I

t i

t t

I l

l,

G.

Summ ry_oLChangento_thLricility which era omrliked in the Safety _ Analysis _Repo r t- ( co n tinued}

Operability Evaluation, Operation Of The 1A Emergency Diesel Generator Without Automatic Prelubing This Operability Evaluation does describe a change to the Updated Fin.a1 Safety Analysis Report. The 1A Diesel Generator AC Soakback Pump will be out-of-service because its control transformer failed and must be repaired.

During this period, the DC Soakback Pump will be used.

This evaluation addressed the total loss of prelubing. This change in no way affects the ability of the 1A Diesel Generator to perform its design function. The DC Soakback Pump is capable of providing lube oil to the Diesel Generator turbocharger bearings in the same capacity as the AC Soakback Pump.

The DC Soakback pump load current is already assumed in the Loss-of-Offsite/Onsito AC Power Division 2 battery load profile as described in the Updated Final Safety Analysis Report. The prelubing system is used to prevent long term wear caused by frequent dry starting of the engine, therefore this subsystem is not required to meet the definition of operability as defined by the Technical Specifications.

Operability Evaluation, 1CM025A Post LOCA Sample V61ve This Operability Evaluation addressed the results of taking the containment valve 1CM025A, Post Loss Of Coolant Accident Sample Valve, out-of-service open.

The purpose of the change is to ensure that no adverse affects exist on safety functions for the valve and system. The Updated Final Safety Analysis Report requires the system to automatically actuate on the occurrence of a Loss Of Coolant Accident and to remain in operation after initiation unless turned off with a handswitch. With the ICM025A l

valve in the open position, sampling of the primary containment with the H /0 Monitor is allowed and therefore operable.

Primary containment 2 2 integrity will not be affected since the system is a closed loop with the isolation valves in the open position and the system was designed to operate on the occurrence of and during a Loss Of Coolant Accident.

Operability Evaluation, 2CM025A Post LOCA Sample Valve This Operability Evaluation addressed the results of taking the Containment Valve 2CM025A, Post Loss Of Coolant Accident Sample Valve, out-of-service open.

The purpose of the change is to ensure that no adverse affects exist on safety functions for the valve and system. The Updated Final Safety Analysis Report requires the system to automatically actuate on the occurrence of a Loss Of Coolant Accident and to remain in operation after initiation unless turned off with a handswitch. With the 2CM025A valve in the open position, sampling of the primary containment with the H /O; Monitor is allowed and therefore operable.

Primary containment 2

integrity will not be affected since the system is a closed loop with the j

isolation valves in the open position and the system was designed to operate on the occurrence of and during a Loss Of Coolant Accident.

l G.

Summ;ry_QLQLanges to the_F2cility Which cro described in the

)

Safety Analy11s_ Report-(continuedi Operability Evaluation, Actions For Inoperable Water Tight Door This Operability Evaluation is to specify the actions to be taken when a water-tight door which is not specifically addressed in the Technical Specifications is inoperable. This interpretation is intended to provide guidance to the plant reactor operators for compliance with Station Technical Specifications. This guidance will allow the operators a maximum degree of operational flexibility while helping to ensure that the Technical Specification requirements are adhered to.

The Updated Final Safety Analysis Report describes the safety evaluation of the Core Spray Cooling System cooling water systems affected by the doors. The worst case failure could result in the loss of redundant divisions and this would require going to a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time clock Limiting Condition for Operation.

Operability Evaluation, Standby Gas Treatment Train Operable With Cooling Fan Inoperable This Operability Evaluation is to allow the Standby Gas Treatment Train to be considered operable while the IVG02C Cooling Fan is inoperable. The Standby Gas Treatment Cooling Fan is described in the Updated Final Safety Analysis Report. The Cooling Fan is not required for Technical Specification operability requirements per Technical Specification 3.b.5.3.

The Cooling Fan is classified as Auxiliary Safety and is not referred to in the Technical Specifications. The inoperable Cooling Fan on the Standby Gas Treatment Train will not effect the performance of the train during an accident. The Standby Gas Treatment Train can still be operated for f

Technical Specification Surveillances and will not be prevented from auto starting should the need arise.

Operability Evaluation, Unit 2 Diesel Generator Operability Test This Operability Evaluation is to allow the 2B Diesel Generator to be used while the thermocouple from the # 10 cylinder is missing. The # 10 Cylinder thermocouple monitors the Diesel Generator exhaust temperature.

The exhaust temperature will not be monitored until a permanent repair can be performed. The old thermocouple can not be completely removed from the exhaust manifold at this time.

Therefore, the hole created by this partially removed thermocouple will be welded shut until a permanent fix can be done.

The Updated Final Safety Analysis Report states in section 7.3.6.2 that thermocouples were provided to monitor cylinder exhaust temperatures.

The thermocouples are used for indication only and do not affect any of the designed functions of the Diesel Generator.

G.

Summary _cLChanges to the_Rrility_Mhich era dessribed in the Saf1ty_ AnalJala_Eeport-frontinumd)

Out Of Service 1-1499-92 This Out-Of-Service took 1E12-F068B out-of-service open as an administrative control during valve maintenance. One division of Residual Heat Removal was still available for decay heat removal.

The Contrr,1 Room Operator wasunable to operate the valve, but it could be isolated from the l

l Service Water tunnel for flood control. Detection and isolation of a tube leak via Process Radiation Monitor was not available, but this was not a l

change to its normal idle configuration. The system was returned to normal 6

l upon completion of the valve maintenance.

1 Out-Of-Service 1-1905-92 The Safety Evaluation for this Out-Of-Service determined the operability of the

'A' Residual Heat Removal Service Water and Shutdown Cooling with the l

1E12-F068A valve taken Out-Of-Service open.

The Out-Of-Service applied to l

the electrical operation of the valve only.

This change prevented the IE12-F068A valve from being operated remotely. The capability to manually operate this valve was not affected. This change was only in effect during Refuel or Cold Shutdown conditions. The Suppression Pool Cooling mode of t

I l

Residual Heat Removal was not affected.

l l

Out-Of-Service 2-1079-92 l

l This Out-Of-Service allowed opening the 2E12-F068A Valve for repacking.

The Residual Heat Removal Service Water Heat Exchanger Outlet Valve was taken out of service open against its backseat to allow for replacement of the valve packing. The out-of-service boundary consisted of valves 2E12-F003A (full closed manually), 2E12-F047, 2E12-F052A, and 2E12-F014A, all which were closed, and 2E12-F068A, which was open.

This change did not affect the operation of the Residual Heat Removal system as described in the i

Updated Final Safety Analysis Report, however valve 2E12-F068A is described as a normally closed valve. This change was made with Unit 2 defueled and I

l with no potential flow paths through the Residual Heat Removal heat

{

exchanger. The system was be returned to normal upon completion of the work.

r UFSAR Chapter 3 - Design of Structures, Components, Equipment and Systems Table 3.2-1 of the Updated Final Safety Analysis Report (UFSAR) was updated to reflect an administrative change to delete reference to the Control Rod Drive System water being returned to Feedwater. The actual flowpath is described in the System Description portion of the UFSAR Section 4.6.1.1.2.4.2.5 and has been approved by the Nuclear Regulatory Commission.

This change has been fully evaluated in a previously approved Evaluation.

t

G.

Summary _of_Ch oges to the_rtcility_Mhicham_ described _in_the SAIRLy_ Analysis _Repo r t-f. continued)

UFSAR Chapter 3 - Design of Structures, Components, Equipment and Systems Saction 3.9.1.2.2 of the Updated Final Safety Analysis Report (UTSAR) j was updated to include a new section, Section 3.9.1.2.2.13 which describes the Piping Analysis Program SUPERPIPE, related to the the Snubber Reduction Modification, M-1-2-90-008, for 2MSO9, This is a comprehensive computer program developed for the structural analysis and design checking of piping systems. The snubber reduction involved non-safety related snubbers and will not affect the mo0e of operation.

This modification does not create a transient / accident of a different type from those previously evaluated in the Final Safety Analysis Report or the Updated Final Safety Analysis Report.

UFSAR Chapter 4 - Reactor An administrative change was made to replace the incorrect figure 4.2-4 of the Updated Final Safety Analysis Report (UFSAR) with the correct Tigure 4.2-4 to reflect the actual number of boron carbide tubes per wing of a control rod.

The Safety Evaluation concluded that there is no new possibility for an accident or malfunction of a dif ferent type: the correction is of a typographical nature only which solely affects the Figure.

UTSAR Chapter 5 - Reactor Coolant System and Connecting Systems The Updated Final Safety Analysis Report (UFSAR) was updated to reflect deletion of the Residual Heat Removal Steam Condensing Mode through administrative controls and to provide for an alternate method of shutdown cooling which will be controlled by procedures. The probability of an accident or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the UFSAR have not increased.

UFSAR Chapter 6 - Engineered Safety features m p.

Chaptor 8 and Appendix H of the Updated Final Safety Analysis Report (UFSAR) were revised due to Minor Change P01-2-91-513, Changing the Operating Gear Ratio of Valve 2E21-r012. This Minor Change replaced the motor pinion gear and worm shaft gear to change the operator gear drive ratio to improve valve reliability.

The Safety Evaluation concluded that the change does not create or increase the probability of an occurrence or the consequences of an accident or malfunction previously evaluated in the UTSAR.

I I

J J

I l

l G.

Swummrv of Chanaes to the Fmellity Which nre deceribed in the j

Safety Analysis Report-(continued]

UFSAR Chapter 6 - Engineered Safety Fertures Table 6.2-21 of the Updated Final Safety Analysis Report (UFSAR) was i

updated to reflect an administrative change to permit local leak rate

{

testing of the Unit I reactor water cleanup return valve IG33-F040 in the

" reverse" direction. The Safety Evaluation for this administrative change concluded that no unreviewed safety question exists.

j i

Table 6.2-21, Sheet 1 of 24, of the Up6ated Final Safety Analysis Report (UFSAR) was revised to reflect the normal valve positions for 1(2)B21-F067A,B,C,D from " Closed to "Open".

The valves basic function j

remains unchanged and they will continue to operate as described in the l

UFSAR. The probability of an accident or the consequences of an accident or i

malfunction of equipment important to safety es previously evaluated in the l

UFSAR have not increased.

l I

A revision was required to Table 6.2-21 and Figure 6.2-31 of the Updated 1

Final Safety Analysis Report (UFSAR) due to Component Replacements CR-90-124, CR-90-125, CR-90-126 and CR-90-127 which replaced the 2B21-F067A/B/C/D Main Steam Drain Globe Valves with Disk Gate Valves. This was done to reduce through-seat leakage, increasing the ability to meet the l

1eakage requirements stipulated in 10CFh50 Appendix J, and to decrease f

maintenance requirements. The Safety Evaluation concluded that the l

probability or consequences of an accident previously evaluated will not be increased, nor will it create an accident or malfunction of a different type than any previously analyzed in the UFSAR.

i UFSAR Chapter 7 - Instrumentation and Control systems

-l Figure 7.3-7 of the Updated Final Safety Analysis Report (UFSAR) has been revised to reflect the removal of miscellaneous electrical items associated with the gland seal leak-off lines for 2B33-F067A/B which were removed by Modifications M-1-2-84-067 and M-1-2-89-001.

The items removed i

were non-safety related and non-functional conduit and cable which were l

previously abandoned in place.

1 An administrative change was made to Sections 7.6.2.2.4 and 7.6.2.2.5 as well as Figure 7.3-7 of the Updated Final Safety Analysis Report (UFSAR) to clarify the description of the RCIC and RHR leak detection systems. The changes were identified during review for modification M01-1-88-052, Deletion of RHE and RCIC Line Break Switches 1E31-N007AB/BB, 1E31-N012AB/BB, and 1E31-N013AB/BB. The change is not due to the installation of the j

modification. The Safety Evaluation concluded that there were no unreviewed j

safety questions.

j l

l l

~

i x

,,__m,..,,_,.,

..~-_-m,,%y

T G.

Sumin2xy of Chages to the Fmcility Which Ors _ described _in_the Saicty_ Analysis Report-(snatiguedl UFSAR Chapter 7 - Instrumentation and Control Systems A revision was required to Figure 7.3-7 of the Updated Final Safety Analysis Report (UFSAR) due to Minor Change P01-2-91-506 which replaced the body and internals of the 2G33-F001 Reactor Water Cleanup Valve from a flexible wedge gate to a double disk gate design. This was done to reduce through-seat leakage, increasing the ability to meet the leakage requirements stipulated in 10CFR50 Appendix J, and to decrease maintenance requirements. The Safety Evaluation concluded that the probability or consequences of an accident previously evaluated will not be increased, nor will it create an accident or malfunction of a different type than any previously snalyzed in the UFSAR.

Section 7.3 of the Updated Final Safety Analysis Report (UFSAR) required a revision due to modification M01-2-90-002, Replacement of 2B Diesel Generator Instrument Tubing. The 1/4 inch copper instrument tubing was replaced with stainless steel tubing. The Safety Evaluation concluded that the probability or consequences of an accident previously evaluated will not be increased, nor will it create an accident or malfunction of a different type than any previously analyzed in the UFSAR.

Section 7.3 of the Updated Final Safety Analysis Report (UFSAR) required revision due to modifications M01-0-90-001B, M01-1-90-001 and M01-1-90-002, Replacement of the 0, 1A and IB Diesel Generators Instrument Tubing. The 1/4 inch copper instrument tubing was replaced with stainless steel tubing.

The Safety Evaluation concluded that the probability or consequences of an accident previously evaluated will not be increased, nor will it create an accident or malfunction of a different type than any previously analyzed in the UFSAR.

i Chapter 7 of the Updated Final Safety Analysis Report (UFSAR) required revision to add guidance for the installation of a switched jumper to bypass the Reactor Water Cleanup System (RWCU) Delta Flow signals for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a

during startup and shutdown conditions.

This is to prevent undesirable RWCU isolations during startup and shutdown due to density differences from normal operating conditions. The Safety Evaluation concluded that the probability or consequences of an accident previously evaluated will not be increased, nor will it create an accident or malfunction of a different type than any previously analyzed in the UFSAR.

]

UFSAR Chapter 8 - Electric Power Section 8.3.3.1 of the Updated Final Safety Analysis Report (UFSAR) was updated to reflect the installation of new level instrumentation for the radwaste Ultrasonic Resin Cleaner / Waste Sample Tank / Spent Resin Tank using control cables lower in voltage than those described in UFSAR 8.3.3.1.b.

The higher voltage used in the UFSAR is more conservative from a safety evaluation perspective, and there is therefore no possibility for an accident or malfunction of a dif ferent type than any previously evaluated.

C.

Summary _nL_ Changes to the Freility Which cro dracribed in the 5afely_ Analysis Report-f coniinued]

UFSAR Chapter 8 - Electric Power An administrative change was made to Sections 8.2.3.2.2 of the Updated Final Safety Analysis Report (UTSAR) to revise the minimium acceptable starting voltages supplied from 480 VAC MCC's to reflect the findings of the Degraded Voltage Steady State Analysis (Chron #196647). The Safety Evaluation concluded that there were no unreviewed safety questions.

Section 8.3.2 of the Updated Final Safety Analysis Report (UFSAR) was revised to reflect the replac< ment of the 60 lead-antimony FPS-15 Plante battery cells of the 125 VDC Division II battery with 58 GNB NCX-17 lead-calcium cells per modification M01-2-88-003.

The battery capacity will increase from 581 A-hr. to 1128 A-hr.

A new battery rack arranged in two tiers was installed to accomodate the larger cells. Ammeter scales in the Main Control Room and at the bus will be replaced to be consirtent with other station battery ammeters due to human factors concerns. This requires replacement of the signal converter and recalibration of the DC alarm unit.

The new batteries and modified racks meet all the requirements specified by the UFSAR.

Table 8.3 of the Updated Final Safety Analysis Report (UFSAR) required

{

revision due to modification M-01-2-90-009, Replacement of the 125 Volt Division III Battery, Rack and Main Feed Breaker. This was due to increased DC loads as well as to improve breaker reliability and coordination. The Safety Evaluation concluded that the change does not create or increase the probability of an occurrence or the consequences of an accident or malfunction previously evaluated in the UFSAR.

UFSAR Chapter 9 - Auxiliary Systems An update was made to Section 9.1.3.2.1 of the Updated Final Safety Analysis Report (UFSAR) to correct a numerical error in the value of the Maximum Normal Heat Load for the Unit 2 Fuel Pool. This error correction is not a change to the plant, but a correction to the description of the plant.

The plant was previously analyzed with the correct information.

The probability of an accident or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the UFSAR have not increased.

l l

Chapter 9 of the Updated Final Safety Analysis Report (UFSAR) has been l

updated due to Minor Change POI-1-91-532, Replacement of the Unit 1 Refuel Platform Triangular Refueling Mast with a GE NF-500 Circular Mast.

The Safety Evaluation concluded that the change does not create a new or different kind of accident or malfunction from any accident previously evaluated.

I i

l l

I G.

Summ ry_nLChanges to the Fccility Which are described _in_the Safely _Analynig Report-(continuedl UFSAR Chapter 9 - Auxiliary Systems Section 9.5.4.2 of the Updated Final Safety Analysis Report (UFSAR) has been updated to reflect the exception to ANSI N-195.

This change is administrative only.

This exception involves the permanent connection between the High Pressure Core Spray (HPCS) Storage Tank and the Fire Protection Diesel Fuel Oil Transfer System. This permanent connection and its function is described in the UFSAR Section 9.5.4.2.

The permanent connection does not affect the HPCS Emergency Diesel Generator seven day fuel oil inventory Technical Specifications requirement. The Safety Evaluation concluded that there were no unreviewed safety questions.

Section 9.4.1.2 of the Updated Safety Analysis Report (UFSAR) was updated to reflect the deletion of the reference to humidity control and I

isolation dampers for the computer room.

The dampers and humidity control I

were never installed. Humidity control is not necessary because the f

computer room does not take in outside air.

Isolation dampers are not required because there are no habitability requirements for the computer room.

The Safety Evaluation concluded that there were no unreviewed safety questions.

Section 9.5.3 of the Updated Final Safety Analysis Report (UFSAR) has been updated to reflect the installation of Emergency Lighting Battery Packs i

l (ELBP's), lamps, and associated hardware throughout the Reactor, Auxiliary, and Turbine Buildings. This change is required in accordance with 10 CFR 50 l

Appendix R,Section III.J which requires 8-hour battery packs to illuminate Safe Shutdown equipment along with access and egress routes to and from this equipment.

The scope of work in this change does not result in a functional design change which could adversely affect any safety system, nor does it create an accident of a different type from those previously evaluated in the Final Safety Analysis Report or the Updated Final Safety Analysis.

Chapter 9 of the Updeted Final Safety Analysis Report (UFSAR) specifying the manual method of resin regeneration requires revision. An unreviewed safety question does not exist.

UFSAR Chapter 10 - Steam and Power Conversion System A revision to Chapter 10 of the Updated Final Safety Analysis Report (UFSAR) to clarify the present function of the regeneration subsystem and provide flexibility when ultrasonic resin cleaners are used was required. An unreviewed safety question does not exist, i

G.

Sumary_oLChanges to th2_recility_Mhichars_ described _in_t.h3 Saf e ty_Analysi s_Repor t- ( co n tianved]

UFSAR Chapter 11 - Radioactive Waste Management Section 11.4 and Table 11.4-1 of the Updated Final Safety Analysis Report (UFSAR) was updated to reflect an administrative change referencing on-site use of the dry active waste interim storage facility. The Safety Evaluation for the interim storage facility concluded that no unreviewed safety question exists.

Section 11.4.2.9 to the Updated Final Safety Analysis Report (UFSAR) was added for an Over the Wall Modification which added concrete cubicles to the intermediate storage level, processing penetrations in the outside wall, and removed upper blocks of walls to facilitate movement of containers. All this to help provide an ef ficient manner for processing radwaste liners and high integrity containers. The Safety Evaluation concluded that there were no unreviewed safety questions.

UFSAR Chapter 13 - Conduct of Operations Changes were made to Section 13.1.2.2 of the Updated Final Safety Analysis Report (UFSAR) to reflect changes in the function, responsibilities, and authorities of the Plant Manager, Production Superintendent, and Technical Superintendent; also several changes to some LaSalle organizational titles. These are administrative changes not effecting plant safety. Therefore, the probability of an accident or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the UFSAR have not increased.

A revision to Chapter 13 of the Updated Final Safety Analysis Report (UFSAR) to update the description of the station Training Department and its programs was required. An unreviewed safety question does not exist.

Page 13.0-lii of the Updated Final Safety Analysis Report (UFSAR) required an administrative change to remove the reference to Table 13.5-5.

Table 13.5-5 was removed from the UFSAR in the Revision 8 update.

The Safety Evaluation concluded that the change does not create or increase the probability of an occurrence or the consequences of an accident or malfuction previously evaluated in the UFSAR.

UFSAR Chapter 15 - Accident Analyses Section 15.7.4 of the '

ted Final Safety Analysis Report (UFSAR) required an administrati n,'ange to correct a typographical numerical error in the number of fuel roer st fail as a result of a fuel handling accident. The correct number of failed fuel rods as a result of this accident was correctly stated in a previous discussion in this section.

This is a correction to the plant description and not a change to the plant itself. The Safety Evaluation concluded that the possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created.

G.

Summary _gf_ Changes to th!_Eccility Which erce desCIIbed in the Sainly_AnalyAis Report-(continued)

Appendix G - Reactor Recirculation System A revision to Appendix G.3.1.2.2 of the Updated Final Safety Analysis Report (UFSAR) due to the Anticipated Transient Without Scram -

Recirculation Pump Trip (ATWS-RPT) Logic Change installed by modifications M-1-1-89-026 and M-1-2-89-021.

These modifications renlaced the 1-out-of-4 logic with 1-out-of-2 taken twice for tripping each Reactor Recirculation Pump during an ATWS condition.

The Safety Evaluation concluded that the change does not create or increase the probability of an occurrence or the consequences of an accident or malfuction previously evaluated in the UFSAR.

A revision to Appendix G of the Updated Final Safety Analysis Report (UFSAR) due to minor change P01-1-91-529 which replace the RR Flow Control Valve actuators hard piped hydraulic lines with flexible hoses. The double-block drain valve located near each connection will be replaced by a lighter single glove valve.

These actions will reduce vibration.

The Safety Evaluation concluded that the probability or consequences of an accident previously evaluated will not be increased, nor will it create an accident or malfunction of a different type than any previously analyzed in the UFSAR.

A revision to Appendix G of the Updated Final Safety Analysis Report (UFSAR) due to minor changes MC1-1-90-028 and MCl-2-90-013 which addr* e clamp and beam assembly to jet pumps 85 and #15 to prevent vibration imeuued fatigue failure in the lower support brackets.

The Safety Evaluation concluded that the probability or consequences of an accident previously evaluated will not be increased, nor will it create an accident or malfunction of a different type than any previously analyzed in the UFSAR.

UFSAR Appendix H - Fire Hazards Analysis An administrative change to Appendix H of the Updated Final Safety Analysis Report (UFSAR) was required due to the installation of modifications M1-0-90-008, M1-1-90-014 and M1-1-90-015 which removed Solenoid Valves OD0004, 1D0004, and 1DOO14 from the from the Diesel Fuel Oil Transfer System and replaced them with 1.5" diameter piping. The Safety Evaluation concluded that the possibility of an accident or malfunction of a different type than any previously evaluated in the UFSAR doesn't increase.

An administrative change was required to Section H.3 of the Updated Final Safety Analysis Report (UFSAR) to reflect the installation of Minor Change P01-1-90-571, Unit 1 CRD HCU Drain Header Minor Plant Change Modification. This Minor Change installed drain hoses at all Control Rod Drive Hydraulic Control Unit Accumulator Drain Valves which are routed to local RF floor drains to minimize the spread of contamination from water spills which occur when the accumulators are drained for maintenance or nitrogen charging. The Safety Evaluation concluded that the probability of an accident or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the UFSAR have not increased.

G.

SummxyJf__Ch ;pges to theErility_Which cre degrtihed is the nafetv_ Analysis Report-(egntinuedl UFSAR Appendix H - Fire Hazards Analysis Section H.3.5.1 was updated to reflect the removal of the 3-hour fire rating for steel beams supporting slabs at elevation 768 feet-0 inch. The fireproofing on these beams was removed under Work Request L08569 and the beams now are protected to a 1-hour fire rating.

This change was necessary because the fireproofing on these beams continued to fall off and onto the Turbine Drive Reactor Feed Pump.

The Safety Evaluation concluded there was no unreviewed safety question since the equipment is non-safety related and feedwater is not assumed to be available.

An administrative change to Section H.3.7 of the Updated Final Safety Analysis Report (UFSAR) to make it consistent with the as-built configuration of the sprinkler systems for Diesel Generator Day Tank Room fire zones was required. The possibility of an accident or malfunction of a different type than previously evaluated has not been created because the configuration of the sprinkler systems and barrier remain unchanged.

Sections H.3.1, H.3.2, and H.3.5 of the Updated Final Sefety Analysis Report (UFSAR) required revision to reflect the derating of non-Tech Spec fire walls.

The probability of an accident or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the UFSAR have not increased.

Section_li 3.1.1 - Currently there is no separation between Fire Zone 1 at Elevation 843'-6" and 2B and 3B at Elevation 820*-6" due to non-rated floor and open stairwells. The stairwell walls, elevator walls, and vestibule walls, therefore, are not providing any added protection between Fire Zone 1 and the adjacent Fire Zones.

Section H.3.2.1 - Currently there is no separation between Fire Zones 2A and 3A at Elevation 832'-0" and Fire Zone 1 at Elevation 843'-6" or Fire Zones 2B2 and 3B2 at Elevation 820*-6" due to non-rated floor and ceiling slabs at Elevation 832'-0" within Fire Zones 2A and 3A.

The north, south and east walls at Elevation 832*-0" within Fire Zones 2A and 3A are not providing any added protection between Fire Zones 2A and 3A at Elevation 832'-0" and the adjacent Fire Zones.

Enstion H.3.5 21 - The doors which are currently installed are U.L.

labeled "A" fire doors and the frames have been rated as a result of the U.L.

testing program.

An administrative change to Section H.3.8 of the Updated Final Safety Analysis Report (UFSAR) for wording changes to state that the floor / ceiling separating Fire Zone 8A1 from BB1, and Fire Zones 7A1, 7A2, and 7A3 are three hour fire rated except for an untested exhaust stack fire penetration seal.

This seal is of the same materials and construction as other rated penetration seals, but must be considered unrated because it is of a larger size than what has been tested. The probability of an accident or the consequences of an accident or malfunctiot

>f equipment important to safety as previously evaluated in the UFSAR have et increased.

G.

Summny_oLClenges to the_Itcility_Mbich_cre_ described _in_the I

Saftty_ Analysis _REPDLt-in utinued) 1 UFSAR Volume XI l

A revision to Drawing M-142 Sheet 2 of the Updated Final Safety Aralysis Report (UFSAR) was required due to Minor Change P01-2-90-039, Installation i

of Four Hydrolating Ports.

Hydrolazing ports were installed on RHR iniection lines 2RH40CB-16 and 2RH40AB-12 and Globe valves at the low point of 2RH40CB. This will permit the flushing of water and crud from the RHR pipe to the floor drain or a collection tank, and to reduce the dose associated with these activities. The Safety Evaluation concluded that the probability or consequences of an accident previously evaluated will not be increased, nor will it create an accident or malfunction of a different type than any previously analyzed in the UFSAR.

Appendix H - Fire Hazards Analysis i

Chapter 8 - Electric Power Chapter 8 and Appendix H of the Updated Final Safety Analysis Report (UFSAR) required revision due to the replacement of the existing 50 amp C &

D power systems battery charger with a new Power Conversion Products charger also rated at 50 amps (Modification M01-2-90-010).

The Safety Evaluation concluded that the change does not create or increase the probability of an occurrence or the consequences of an accident or malfunction previously evaluated in the UFSAR.

A revision to Chapter 8 and Appendix H of the Updated Final Safety Analysis Report (UFSAR) was required due to the eddition of the 125 Volt Division II Battery Charger of a 200 Ampere rating (Modification M01-2-88-002).

This Modification consisted of upgrading to a larger capacity charger for the 125 Volt Division II Battery.

The Safety Evaluation concluded that the change does not create or increase the probability of an occurrence or the consequences of an accident or malfunction previously evaluated in the UFSAR.

Appendix H - Fire Hazards Analysis Corrective Action Record 037 LaSalle Corrective Action Record (CAR)91-037 was issued to resolve the discrepancy of plugged floor drains under the Turbine Driven Reactor Feed Pump Room (TDRFP) which have no record of documentation as to why they are plugged. Under this CAR the floor drains will be maintained in their existing plugged condition on a permanent basis. The plugged drains will prevent lubrication oil from the Unit 1 and Unit 2 TDRFP Rooms from entering the Radweste Liquid Processing System. The rooms do not contain safety related equipment and have a low fire loading. They are protected with an automatic sprinkler system. No new accidents or malfunctions are created.

l

G.

Summ ry_oLChmges to the_Iccility_Whinb3re_Jescribed_is_t.lO Safely _ADitlyXis Report-(comiinuell UFSAR Chapter 3 - Design of Structures, Components, Equipment and Systems UFSAR Chapter 10 - Steam and Power Conversion System Chapters 3 and 10 of the Updated Final Safety Analysis Report (UFSAR) required revision due to Component Replacements90-157 and 92-013.

Replacement of the Unit 1 and 2 Low Pressure Turbine Rotors. Both Unit I and 2 low pressure turbine rotors were replaced. The Safety Evaluation concluded that the probability or conseguences of an accident previously evaluated will not be increased, nor will it create an accident or malfunction of a different type than any previously analyzed in the UFSAR.

UFSAR Chapter 5 - Reactor Coolant System and Connected Systems UFSAR Chapter 7 Instrumentation and Control Systems l

Chapters 5 and 7 of the Updated Final Safety Analysis Report (UFSAR) required a revision due to ECCS Testable Check Valve Upgrades by Partial Modifications M-1-1-87-098-02, 03, 04, 05, 06, 07, 08, 09, M-1-2-87-087-01, 02, 05, and 06.

These Modifications involved the following, where applicable: valve packing to reduce valve shaft friction; leakoff line removed; stuffing box connection plugged; limit switch replacement. The Safety Evaluation concluded that the probabi3ity or consequences of an accident previously evaluated will not be increased, nor will it create an accident or malfunction of a different type than any previously analyzed in the UFSAR.

t UFSAR Chapter 8 - Electric Power Appendix H - Fire Hazards Analysis An administrative change to Tables 8.3-14, H.3-2, H.4-73 and page H.3-136 of the Updated Final Safety Analysis Report (UFSAR) to reflect the upgrading of the capacity of the 125 Volt Division 3 Battery per modification M01-1-90-011 was required.

The probability of an accident or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the UFSAR have not increased.

I UFSAR Chapter 9 - Auxiliary Systems UFSAR Appendix H - Fire Hazards Analysis I

I Section 9.5 and Appendix H of the Updated Final Safety Analysis Report (UTSAR) was updated to reflect an administrative change making the fire hazards analysis consistent with present plant conditions.

Plant safety is not affected by these administrative changes.

The Safety Evaluation concluded that an unreviewed safety question does not exist.

1 G.

Sunun:ry_9LChanges to the FCCiMiy_.Which ere described in the 4

S af.nty_An alysi s_ReRo r t- ( continund) l UFSAR Chapter 4 - Reactor i

UFSAR Chapter 7 - Instrumentation and Control Systems l

A revision to Chapters 4 and 7 of the Updated Final Safety Analysis Report (UFSAR) due to modification M-1-2-87-083, Replacement of Scram Discharge Volume Level Transmitter was required. This Modification replaced the existing Rosemont Transmitters with the 1153 Capillary Style transmitter 4

which have sealed reference legs to provide more reliable level indication.

4 The Safety Evaluation concluded that the change does not create or increase the probability of an occurrence or the consequences of an accident or

]

malfunction previously evaluated in the UFSAR.

UFSAR Chapter 5 - Reactor Coolant System and Connected Systems UFSAR Chapter 6 - Engineered Safety Features UFSAR Chapter 7 Instrumentation and Control Systems UFSAR Chapter 8 - Electric Power UFSAR Chapter 9 - Auxiliary Systems UFSAR Appendix H - Fire Hazards Analysis Chapters 5, 6,

7, 8,

9, and Appendix H of the Updated Final Safety Analysis Report (UFSAR) were revised due to modifications M-1-1-86-072, M-1-2-86-049, M-1-1-87-095 and M-1-2-87-082. These Modifications abandoned the High Pressure Core Spray (HPCS) system suction and return lines to the CST and aligned the RCIC system to the Suppression Pool.

The Safety Evaluation concluded that the probability or consequences of an accident previously evaluated will not be increased, nor will it create an accident or malfunction of a different type than any previously analyzed in the UFSAR.

i j

Y

\\

4 i

i H.

Summary of Safety Related Modifications M01-0-84-031 This modification involved rerouting the backflow of condensate drainage from the Control Room and Auxiliary Electric Equipment Room HVAC cooling coils to the equipment drains instead of the common header floor drains.

This required extensive piping changes from the evaporator to the equipment drain header.

This modification also changed the solenoid valves to fail closed instead of fail open by changing the wiring at the affected control panels and by replacing the valve solenoids and internals to the normally closed type. The purpose of modifying the drainage is to lessen the burden on radwaste which had difficulty in handling the existing supply of 1

water from the floor drains. The Safety evaluation concluded that there were no unreviewed safety questions. There were no UFSAR or Technical Specification changes required.

I 1

M01-0-89-003-01 M01-0-89-003-02 These modifications removed the initiation function of the Train A and l

Train B VC/VE recirculation mode upon detection of high ammonia concentration by disconnecting the interlocks between ammonia detectors and the VC/VE dampers at panel OPL15 (16) J.

These modifications also installed a manual pushbutton with protective collar ring and an indicating light for manual actuation of the recirculation mode when a detector alaoms on Control Room panel 1(2)PM05J.

Also, a selector switch and two indicating lights are provided to allow either one of the ammonia detector to be bypassed if, by reading the recorder tapes of each ammonia detector, a control room high ammonia alarm is t etermined to have been activated. The Safety Evaluations concluded that there were no unreviewed safety questions.

However there were Technical Specification Amendments, Numbers 61 and 42 to Facility Operating License Number NPF-ll and NFF-18. The Technical Specifications have been revised as described in the Amendments. The changes to the UFSAR Chapter 6,

" Engineered Safety Features", subsection 6.4.3 " System Operational Procedures" addresses the automatic actuation function of the Control Room HVAC system upon detection of high ammonia. This subsection

]

required revision to reflect the removal of the automatic initiation function of the recirculation mode.

M01-0-89-006 This modification involved upgrading the existing 0 Diesel Generator Room Carbon Dioxide system from an unsupervised detection system to a class "A" supervised detection system. This also provides protection of the DG CO2 system against fire / water damage in order to prevent a single fire in i

the DG Corridor from causing a CO2 discharge in both the division 1 and i

division 2 DG rooms.

The existing control panels were replaced with l

seismically qualified NEMA Type 4 CO2 control panels. The alarm horns, I

pushbutton stations, and electrol manual pilot cabinets were also replaced i

with NEMA Type 4 alarm horns, pushbutton stations, and electrol manual pilot j

cabinets. The conduits containing associated cables were fire wrapped so if damaged, would not spuriously initiate a CO2 discharge or adversely affect their respective DG ventilation system. The Safety evaluation concluded that there were no unreviewed safety questions. There were no UFSAR or Technical Specification changes required.

I 1

D

H.

Summarv of Safety Related Modifications l

M01-0-89-010 I

I This modification involved rerouting portions of the Fire Protection piping because the new structures, i.e. service building, warehouse, etc.,

were located over the existing buried fire protection yard loop.

This modification also provided two supply lines for the new service building, two for the new main warehouse, one for the MAT building, and one for the new receiving warehouse.

Additional hydrants were installed for manual fire fighting capability. A main fire protection system flushing line has been provided to enhance station FP flushing performance. The Safety evaluation concluded that there were no unreviewed safety questions.

There were no UFSAR or Technical Specification changes required.

M01-0-89-017-A 1

l This modification involved the installation of conduit, supports, and antenna cable between locations in the Auxiliary buidings, Turbine buildings and the south end of the Unit 1 Turbine Building Trackway. A junction box joined the raceway conduit and the new FIP underground ductbank system.

This provided communication cables to the new service building, warehouse, and MAF.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no UFSAR or Technical Specification changes required.

M01-0-89-018-This modification involved a revision to the fuel filter restricted alanm circuit for the "O" Diesel Generator to meet the " blackboard" concept.

The design eliminated spurious activation of this alarm. Fuel filter high differential pressure switch OPDS-DG051 will be interlocked with 50-second time delay relay K33 which is energized when DG engine speed reaches 150 rpm.

This interlock prevents the alarm from annunciating for 50 seconds after a DG start. This alarm will only appear when fuel oil flow through the fuel filter becomes restricted during DG operation. The main function of the alarm remains unchanged. The Safety evaluation concluded that there were no unreviewed safety questions. There were no UFSAR or Technical Specification changes required.

M01-0-90-001-A M01-0-90-001-B M01-0-90-001-C This modification involved the upgrade of the 1/4 inch instrumentation tubing connected to the "0" Diesel Generator. The modification was split into three (3) separate partials:

Partial A, removal and replacement of portions of the air box pressure tubing and lube oil tubing, removal and replacement of plastic tie wraps with cushioned metal tie wraps, adding a tubing support to the air box pressure tubing, adding the crankcase pressure gauge mounting plate and mounting the gauge in place.

Partial B, removal and replacement of crankcase pressure tubing to the new pressure gage OPI-DG122.

Partial C, installation of the crankcase pressure gauge mounting plate, mounting the gauge in place, and addition of a support to the existing air box pressure tubing.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no UFSAR or Technical Specification changes required.

H.

Summary of Safety Related Modificatiens l

M01-0-90-006 This modification involved the upgrade of the ventilation and fire protection / detection systems for Building 30, QA Records Storage Vault.

This included installation of conventional four ton HVAC system, comprised of a 1600 CFM air handline unit and a condensing unit.

This will control the temperature and humidity levels inside the Vault by a fully automatic, climate control system. An extension of the fire detection signal from the vault's Halon panel and Building 30 detectors to the plant's main fire detection panel, 2FF04JA, in the Auxiliary Electrical Equipment Room.

Also, three (3) new Building 30 detectors were added to monitor for smoke outside the Vault so that appropriate measures could be established to prevent outside fires from propagating to inside the Vault.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no UFSAR or Technical Specification changes required.

M01-0-90-008 This modification involved the enhancement of operation for the diesel fuel system by eliminating the unreliable operation of the solenoid valves for the "0" Diesel Generator. These valves have had a history of sticking open, requiring excessive maintenance, and resulting in down-time for the diesel generators.

In addition to the valve removal work, antisyphoning holes were drilled into the inlet piping inside the day tanks.

The antisyphoning holes replaced the intended function of the valves. These holes break the vacuum created inside the fuel transfer pipe during the fuel transfer operation. This will eliminate the differential pressure across the Day Tanks and the Storage Tanks thereby preventing backflow from the Day Tanks to the Storage Tanks which was the purpose of the solenoid valves.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification changes zequired.

However, changes to the FSAR/UFSAR Appendix H.4,

" Safe Shutdown Analysis" l

required update.

M01-1-86-072 This modification involved modifying the High Pressure Core Spray (HPCS) system so that is is aligned to the Suppression Pool (SP), and r

abandonment of the suction and test return lines to the Condensed Storage Tank (CST) due to failure in the test return line to the CST.

The failure of the test return line was attributed to microbiological cortosion, primarily of the weld metal. Major scope included but was not limited to:

installing a new blind flange on HPCS test return line to the CST, de-tena control cables to nearest junction box for the full flow test return valves, remove control switches, position indicators, indicating lights, computer input from control room panel 1H13-P601, and deletion of digital inputs fram plant computer systems. No unreviewed safety questions associated with the modification exists.

Changes to the Technical Specification, Amendment

1. 81, have been completed. Changes required to the UFSAR include sections for Chapters 1, 3,

5 thru 9,

12, 13, and 15. Appendix H, J,

and L also required revisions.

r i

i H.

Summarv of Safety Related Modifications M01-1-87-098-1, M01-1-87-098-2, M01-1-87-098-3 i

M01-1-87-098-4, M01-1-87-098-5, M01-1-87-098-6 M01-1-87-098-7, M01-1-87-098-8, M01-1-87-098-9 These modifications involved increasing the valve reliability for the ECCS Test able Check Valves.

The maintenance work history indicated that the ECCS Testable Check Valves have experienced rotational resistance during low flow conditions. This resistance can prevent the valve from fully closing after low flow testing.

To correct these problems the following was completed: 1)

Replacement of the valve packing with modified ring packing / carbon spacer arrangement, 2) replacement of the Namco limit t

switches with mircoswitches to reduce valve shaft friction. The leakoff line was removed since it is no longer required due to replacement of the valve stem packing.

3) removal of the leakoff solenoid valve, manual valve, thermocouple, flexible hose, sightglass and removal of associated wiring and conduit. This scope of work was completed in nine (9) partial modifications:

Partial 1 modified the HPCS 1E22-F005 ECCS Testable Check Valve.

Partial 2 modified the LPCS 1E21-F006 ECCS Testable Check Valve.

Partial 3 modified the RHR 1E12-F041A ECCS Testable Check Valve.

Partial 4 modified the RHR lE12-F041B ECCS Testable Check valve.

Partial 5 modified the RHR 1E12-F041C ECCS Testable Check Valve.

Partial 6 modified the RHR 1E12-F050A ECCS Testable Check Valve.

Partial 7 modified the RHR 1E12-F050B ECCS Testable Check Valve.

Partial 8 modified the RCIC 1E51-F065 ECCS Testable Check Valve.

Partial 9 modified the RCIC lE51-F066 ECCS Testable Check Valve.

The Safety evaluations concluded that there were no unreviewed safety questions. There were no Technical Specification changes required.

However, changes to the LaSalle FSAR section 5.2.5.2 and figure 7.3-7 sheet 2 were required.

l M01-1-88-026 This modification involves the installation of two (2) fuses (in series) for the RHR Pressure Switches lE12-NO32A/B and IE12-NO33A/B. During the steam condensing mode of RHR, steam is routed from the reactor pressure vessel to the RHR heat exchanger via the RCIC system. During this mode, RHR pressure switches assist in protecting the RHR heat exchanges from over pressurization by providing a signal to close the bypass valves.

If one of the pressure switches were to fail, it would cause an electrical short to ground at one of the non-safety related components being supplied from the same DC bus.

This could have disabled the RHR control logic for essential equipment required during and/or after a postulated design event.

The Safety evaluation concluded that there were no unreviewed safety questions.

1 There were no UFSAR or Technical Specification changes required.

l

\\

l 1

-n

,,an,

l H.

Summary of Safety Related Modifications M01-1-88-052 This modification involved the deletion of line break / flow switches IE31-N007AB/BB, 1E31-N012AB/BB, and 1E31-N0013AB/BB. These switches are SOR type 103AS-B203 differential pressure (DP) switches which have a high diaphragm failure rate and are subject to additional surveillance requirements due to NRC Commitments. These switches initiate an isolation of the RHR heat exchanger and RCIC steam supply lines, respectively, when i

their setpoints are exceeded.

Eliminating these switches will also improve the availability of these systems by reducing the probability of spurious i

system trips.

The change constituted removal of six (6) DP switches and their respective instrument lines. The instrumentation ports of the manifold valves were plugged and abandoned in place.

The removal of the cables between the DP switches and the junction boxes on the instrument racks and the flexible conduit that is attached to these switches has been completed.

Cables have also been removed and abandoned in place between the instrument racks and the computer input cabinet. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification changes required.

However, upon review of the UFSAR Sections 7.6.2.2.4, 7.6.2.2.5 and Figure 7.3-7 concluded that a revision to clarify the description of the RCIC and RHR leak detection systems was j

required.

M01-1-89-030 This modification involved a revision to the fuel filter restricted alarm circuit for the "1A" Diesel Generator to meet the " blackboard" concept.

The design eliminated spurious activation of this alarm. Fuel filter high differential pressure switch 1PDS-DG051 will be interlocked with 50-second time delay relay K33 which is energized when DG engine speed reaches 150 rpm.

This interlock prevents the alarm from annunciatins for 50 seconds after a DG start.

This alarm will only appear when fuel oil flow through the fuel filter becomes restricted during DG operation. The main function of the alarm remains unchanged.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no UTSAR or Technical Specification changes required.

M01-1-89-033-5 l

This modification involved the installation of Dmergency Lighting Battery Packs (ELBPs) for the MSIV Rooms.

This work included installation of a battery pack with remote ammeter outside the MSIV room and the installation of two (2) remote lamps inside the MSIV Room.

This was l

completed per 10CTR50, Appendix R,Section III.j, adequate illumination to assure operators can bring the plant to a safe shutdown (SSD) in the event of a fire which may disable the normal station lighting system. The Safety evaluation concluded that there were no unreviewed safety questions. There are no Technical Specification changes required.

However, this modification did not change any commitments to the FSAR/UFSAR Section 9.5.1 (Fire Protection System) or Appendix H.4 (SSD Analysis). Section 9.5.3 (Lighting Systems) did require a revision to reflect the addition of the ELBP's. In addition, the report entitled " Fire Protection Documentation and Review of the Assessment of 10CTR50, Appendix R, LaSalle County Station" required a revision to reflect the additional emergency lighting.

1 l

l r

i h

I f

H.

Smmmarv of Safety Relat ed Modifications M01-1-90-001 This modification involved the upgrade of the 1/4 inch instrumentation tubing connected to the "1A" Diesel Generator. This modification was initiated in response to the High Pressure Core Spray (HPCS) Safety System Functional Inspection (SSFI) held in 1989. This modification upgrades any non-stainless steel 1/4 inch tubing and components to stainless steel.

In addition the removal and replacement of crankcase pressure tubing to the new pressure gage 1PI-DG122 was performed. The Safety evaluation concluded that there were no unreviewed safety questions.

There were no Technical Specification changes required. However, changes to the FSAR/UTSAR section 7.3.6.2 required update to include the crankcase pressure gauge to the list of local instrumentation.

M01-1-90-002 This modification involved the upgrade of the 1/4 inch instrumentation tubing connected to the "1B" Diesel Generator. This modification was initiated in response to the High Pressure Core Spray (HPCS) Safety System Functional Inspection (SSFI) held in 1989.

This modification upgrades any non-stainless steel 1/4 inch tubing and component to stainless steel.

In addition, crankcase pressure, air box pressure, water jacket pressure and motor driven fuel oil pump filter inlet pressure gauges were installed.

The Safety evaluation concluded that there were no unreviewed safety questions.

There were no Technical Specification changes required.

However, changes to the FSAR/UFSAR section 7.3.6.2 required update to include the addition of the added pressure gauges to the list of local instrumentation.

M01-1-90-005 This modification involved the replacement of the RHR Shutdown Cooling High Flow Isolation Relays with a time delay relay.

During RHR shutdown cooling startups, shutdowns, and flowrate adjustments, large differential pressure (dp) spikes are originating from the shutdown cooling (SDC) high flow instruments lE31-N012AA/BA. The dp spikes are being sensed at the shutdown cooling suction flow elbows. The existing SDC high flow isolation relays 1B21H-K74/77 (Agastate GP) were replaced with a time delay relay (Agastat E7024AB) set for a 1 second time delay. This time delay relay will prevent differential pressure fluctuations seen by isolation switches from tripping the SDC isolation logic. The Safety evaluation concluded that there were no unreviewed safety questions. There were no UFSAR or Technical Specification changes required.

M01-1-90-009 This modification involved increasing the nominal clearance between the Control Rod Drive (CRD) housing supports and the lower surface of the CRD flange cap screws from approx. I to 1.5 inches. This will facilitate undervessel maintenance during the removal and reinstallation of the CRD's during outage periods. This will also decrease the dose associated with this maintenance activity. The Safety evaluation concluded that there were no unreviewed safety questions. There were changes required to the Technical Specification. Reactivity Control systems, Section 3/4.1.3 Control Rods.

Changes to the UFSAR Section 4.6.1.2.3, 4.6.2.3.1.2.1, 4.6.2.3.1.2.4 and 4.6.2.3.3.1 were revised to reflect this modification.

l l

i

H.

Summary of Safety Related Modifications M01-1-90-011 This modification involved the upgrade of the 125 Volt Division III Batteries and rack.

The batteries were replaced due to the addition of DC loads since fuel load.

The 125V division III system has reached its maximum load for the existing size battery.

In addition, these batteries have been replaced because the existing capacity margin to accommodate degraded capacity due to aging and low temperature operation is very small. The battery physical configuration did not change significantly although the new racks are longer and sightly wider.

The Safety evaluation concluded that there were no unreviewed safety questions. There were changes to Technical Specification section 3/4.8.2.3.2 which was based on the existing cells' nominal specific gravity of 1.210.

The new replacement cells have a nominal epecific gravity of 1.215.

The changes to the UFSAR Section 8.3 and the Fire Hazard Analysis were also completed.

M01-1-90-012 This modification involved the replacement of the 125 volt Division III battery charger. This existing charger was replaced because it had become unreliable and spare parts were no longer available. The physical location of the battery charger did not change.

This change upsized the DC cable in order to increase the ampacity of the cable.

The AC cable was also replaced with a longer cable of the same size.

The current and voltage monitoring i

instrumentation for the battery charger remained the same.

The Safety evaluation concluded that there were no unreviewed safety questions.

There were no Technical Specification changes required. A revision to the UFSAR Section 8.3 and Appendix H was required.

M01-1-90-014 M01-1-90-015 These modifications involved the enhancement of operation for the diesel fuel system by eliminating the unreliable operation of the solenoid valves for the 1A Diesel Generator. These valves have had a history of sticking open, requiring excessive maintenance, and resulting in down-time for the diesel generators.

In addition to the valve removal work, antisyphoning holes were drilled into the inlet piping inside the day tanks.

The antisyphoning holes will replace the intended function of the valves.

These holes break the vacuum created inside the fuel transfer pipe during the fuel transfer operation. This will eliminate the differential pressure across the Day Tanks and the Storage Tanks thereby preventing backflow from the Day Tanks to the Storage Tanks which was the purpose of the solenoid valves.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification changes required.

However, changes to the FSAR/UFSAR Appendix H.4,

" Safe Shutdown Analysis" required update.

l

I f

H.

Summary of Safety Related Modifications M01-2-84-031 i

i This modification involved modifying the control circuitry to prevent the closure of the 2A Diesel Generator breaker onto a faulted bus.

This change will also trip the breaker upon bus fault iadication as long as a ECCS signal is not present. This change was based on Station Electrical Engineering Department's recommendation letter which resulted from review of the loss of 4160 volt ESF bus (OPEX No. 83-13).

The physical changes resulting from this modification are strictly minor wiring changes inside the 2422 and 2423 breaker control circuitry cabinets. The safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification revisions required.

However, changes to UFSAR were required to section 8.3-10.

r M01-2-84-141 This modification involved the removal of the isolation and alarm function of the Kiley Modules for the Reactor Water Cleanup Pump Room Temperature Leak Detection (LD) system.

This was completed by removing the wires from the output switch contacts of Riley Modules 1/2E31-N600/601A-F at the Main Control Room Panel and rewire of the output switch contacts of Riley Modules 1/2E31-N600/601G-K at the Main Control Room Panels. The alarm and isolation functions of the RWCU Heat Exchange Room Temperature leak detection are still maintained. This has improved the performance of the leak detection monitoring system, since it has eliminated spurious isolation and alarm signals. The Safety evaluation concluded that there were no unreviewed safety questions. Technical Specification revisions to section 3.3.2 to amend the deletion of the RWCU pump trips was required. The UFSAR required revisions to delete leakage detection equipment as a result of placing the RWCU pumps downstream of the heat exchangers.

6 M01-2-86-017 i

This modification involves the reslope of the instrument line for the Suppression Pool Level Gauge 2CM01M. The level gauge 2CM0lM had a low point

+

in the high leg of its sensing line.

The presence of this low point caused incorrect level indication by the level gauge.

The instrument line 2CM62AB was reconfic=. aid at two (2) locations to eliminate the low points. The Safety evaluation concluded that there were no unreviewed safety questions.

There were no Technical Specification changes required, However, revisions to the UFSAR Chapter 5, Reactor Coolant System and Connected Systems, and Table 5.2.1, Item #10 were amended to include the CM Instrument line for the suppression pool level gauge.

M01-2-86-032 This modification involved the reconfiguration of portions of the Division I and II Suppression Pool instrument sensing lines. These sensing lines contained low spots and/or incorrect pipe slopes which allowed condensation to trap in them, causing the erroneous readings. The lines were resloped from the PC Vacuum breaker piping back to the condensate pots towards the suppression pool penetration using a minimum slope of one-half inch per foot. The Safety evaluation concluded that there were no unreviewed safety questions. There were no UFSAR or Technical Specification changes required.

i

)

L e

w

=

H.

Summary of Safety Related Modifications M01-2-86-049 This modification involved modifying the High Pressure Core Spray (HPCS) system so that is is aligned to the Suppression Pool (SP) and abandonment of the suction and test return lines to the Condensent Storage Tank (CST) due to a failure in the test return line to the CST. The failure of the test return line was attributed to microbiological corrosion, primarily of the weld metal.

Major scope included but was not limited to:

installing a new blind flange on HPCS test return line to the CST, de-term control cables to nearest junction box for the full flow test return valves, remove control switches, position indicators, indicating lights, computer input from control room panel lH13-P601, and deletion of digital inputs from plant computer systems. No unreviewed safety questions associated with the modification exists.

Changes to the Technical Specification, Amendment

1. 81, have been completed.

Changes required to the UFSAR include sections for Chapters 1, 3, 5 thru 9,

12, 13, and 15. Appendix H, J,

and L also required revisions.

M01-2-87-082 c

This modification involved modifying the Reactor Core Isolation Cooling (RCIC) system so that it could be aligned to the Suppression Pool in the event the RCIC suction and return lines to the Condensate Storage Tank fail.

This included adding a new water leg pump suction from the suppression pool, adding a full flow test return line to the suppression pool, removing the standpipe from the 2CYO1A line of the CY tank which reserves 135,000 gallons af water for RCIC and HPCS use, and adding a keylock switch and an alarm to the control circuit of valve 2E51-F022.

The new RCIC full flow test line was tied into the LPCS full flow test return line.

The Safety evaluation concluded that there were no unreviewed safety questions. Technical Specification changes for 3/4.3.5, 3/4.7.3, B3/4.7.3 and B3/4.5 associated with the RCIC modification are to add containment isolation provisions for the new full flow test return line to the suppression pool. The UFSAR changes to Chapters 1, 3,

5, 6,

7, 8,

9, 12, 13, and 15, also Appendix H, J.

and L consisted of revising descriptions and tables associated with the RCIC system.

M01-2-87-083 This modification involved replacing the Scram Discharge Instrument Volume (SDIV) level transmitter 2C11-N012A,B,C,D.

The Differential Pressure Transmitters monitor the water level in the scram discharge volume to assure that sufficient volume is available to accomplish a scram.

This is in response to concerns by GE where inadvertent trips and/or incorrect infonmation may be indicated concerning the level in the Scram Discharge Volume.

Four (4) Rosemount capillary style model 1153 transmitters were replaced with "hard piped" existing transmitters. This consisted of removing a section of pipe at each SDV and adding a flange for attachment of the respective capillary tube (reference leg), tracing the capillary tubing and mounting the transmitters on the existing stands and re piping between the transmitters " sensing leg" and original 5 valve manifold. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification changes required.

However, changes to the UFSAR sections 4.6.1.1 and 7.2.2.4.7 along with Tables 7.2-1 and 7.3-4 were required to provide more detailed information concerning the transmitters.

i

i H.

Summary of Safety Related Modifications M01-2-86-087 This modification involved replacing the existing relief valves on the shell side of the Reactor Water Cleanup (RWCU) Regenerative heat exchanges.

The hydraulic transient created by valve cycling, coupled with pipe thermal expansion, is believed to be resulting in the repeated cracking of the fillet welds at the connection of the heat exchangers and the inlet of the relief piping. To correct the concerns, the existing Lonergan relief valves 2G33-F340A/B were replaced with a slow-opening Dresser relief valve.

This will minimize the hydraulic transients.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no UFSAR or Technical Specification changes required.

M01-2-87-087-1, M01-2-87-087-2 M01-2-87-087-5, M01-2-87-087-6, These modifications involved increasing the valve reliability for the ECCS Testable Check Valves.

The maintenance work history indicated that the ECCS Testable Check Valves have experienced rotational resistance during low flow conditions. This resistance can prevent the valve from fully closing after low flow testing. To correct the problems the following was completed: 1)

Replacement of the valve packing with modified ring packing / carbon spacer arrangement, 2) replacement of the Namco limit switches with milcoswitches to reduce valve shaft friction, 3) removal of the leakoff line since it is no longer required due to replacement of the valve stem packing,

4) removal of the leakoff solenoid valve, manual valve, thermocouple, flexible hose, sightglass and removal of associated wiring and conduit. This scope of work was completed in four (4) partial modifications:

Partial 1 modified the HPCS 2E22-F005 ECCS Testable Check Valve.

Partial 2 modified the RCIC 2E51-F066 ECCS Testable Check Valve.

Partial 5 modified the RHR 2E12-F041C ECCS Testable Check Valve.

Partial 6 modified the RHR 2E12-F041B ECCS Testable Check Valve.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification changes required.

However, changes to the LaSalle FSAR section 5.2.5.2 and figure 7.3-7 sheet 2 were required.

M01-2-88-002 I

This modification involved the replacement of the 125 Volt Division II battery charger. This existing charger was replaced to accommodate the new 125V division II, GNB type NCX-17, batteries.

The existing charger has remained as a back up charger and will be used in the event of the primary 5

charger. failure. This backup charger will not be used to recharge the batteries after a discharge.

The remote instrumentation and alarms of the existing charger were relocated onto the new battery charger. The new

[

battery charger required instrument changes due to the higher charger capacity. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification changes required.

i However, a revision to the UFSAR Section 8.3 and Appendix H was required.

i i

l I

-~

l H.

Summarv of Safety Related Modifications M01-2-88-003 This modification involved the replacement of the 125 Volt Division II Batteries and rack.

These batteries have been replaced because the existing capacity margin to accommodate degraded capacity due to aging and low temperature operation is very small. The battery physical configuration did not change significantly although the new racks are longer and sightly wider.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no changes to the Technical Specification. The changes to the UFSAR Section 8.3 and the Fire Hazard Analysis were have been completed.

M01-2-88-029 This modification involved the installation of two (2) fuses (in series) for the RHR Pressure Switches 2E12-NO32A/B and 2E12-NO33A/B.

During the steam condensing mode of RHR, steam is routed from the reactor pressure vessel to the RHR heat exchanger via the RCIC system. During this mode, RHR pressure switches assist in protecting the RHR heat exchanges from over pressurization by providing a signal to close the bypass valves.

If one of the pressure switches were to fail, it would cause an electrical short to ground at one of the non-safety related components being supplied from the same DC bus.

This could have disabled the RHR control logic for essential equipment required during and/or after a postulated design event.

The Safety evaluation concluded that there were no unreviewed safety questions.

There were no UFSAR or Technical Specification changes required.

M01-2-88-030 This modification involved replacement of the 2B Diesel Generator Fuel Pump Alarm to eliminate nuisance alarms resulting from spurious activation of the 2B DG alarm circuit due to diesel generator electrical noise.

Several attempts to suppress the electrical noise have failed to prevent the spurious alarms.

This modification removed four (4) SCR diodes, two capacitors and two resistors and installed one HFA relay and two blocking diodes at the diesel generator room panel 2E22-P301B.

This will improve the performance of the DG alarm circuits without changing their function or affecting other control circuits. A failure in the alarm circuit will not affect the operation of the diesel generator due to the electromechanical separation provided between the relay coil and the contacts. The Safety evaluation concluded that there were no unreviewed safety questions.

There were no UFSAR or Technical Specification changes required.

M01-2-89-008 This modification has reduced the possibility of a reactor scram due to inadvertent / spurious RCIC initiations by the addition of a time delay to the turbine trip logic. This time delay provides a four minute span for the operator to determine if the start was spurious and secure RCIC if the system is not needed to maintain vessel inventory. The start of the delay period is signalled by a new control room annunciator which alerts the operator to the RCIC start.

This modification was performed as part of the continuing effort to reduce the number of scrams at LaSalle. The Safety evaluation concluded that there were no unreviewed safety questions.

There were no UFSAR or Technical Specification changes required.

H.

Smmmary of Safety Related Modifications M01-2-89-021 This modification revised the ATWS RPT logic from a one-out-of-two to one-out-of-two taken twice using the Alternate Rod Insertion (ARI) level and pressure instrumentation. The modified circuit also can be tested at full power up to and including the relays that energize the trip coils for the recirculation pump circuit breakers. This change was made to comply with ATWS Rule 10CFR50.62. The Safety evaluation concluded that there were no unreviewed safety questions.

This modification revised Technical Specifications 3.3.4.1, 3.3.4.1-1, 4.3.4.1-1, 4.3.4.1-1, and 3.3.4.1-2.

UFSAR sections 7.6.4.2.1, 15.8 and Appendix G.3.1.2.2 were revised to show the changes in logic to conform to 10CFR50.62 " Requirements for Reduction of Risk from Anticipated Transients Without Scram (.ATWS) Events for Light Water-Cooled Nuclear Power Plants.

M01-2-89-027 This modification involved the revision of the fuel filter restricted alann circuit for the "2A" Diesel Generator to meet the " blackboard" concept. The design eliminated spurious activation of this alarm.

Fuel filter high differential pressure switch 2PDS-DG051 was interlocked with 50-second time delay relay K33 which is energized when DG engine speed reaches 150 rpm.

This interlock prevents the alarm from annunciating for 50 seconds after a DG start.

This alarm will only appear when fuel oil flow through the fuel filter becomes restricted during DG operation. The main function of the alanm remains unchanged. The Safety evaluation concluded that there were no unreviewed safety questions. There were no UFSAR or Technical Specification changes required.

M01-2-90-002 This modification involved the upgrade of the 1/4 inch instrumentation tubing connected to the "2B" Diesel Generator. This modification was initiated in response to the High Pressure Core Spray (HPCS) Safety System Functional Inspection (SSFI) held in 1989. This modification upgrades any non-stainless steel 1/4 inch tubing and components to stainless steel.

In addition, crankcase pressure, air box pressure, water jacket pressure and motor driven fuel oil pump filter inlet pressure gauges were installed. The Safety evaluation concluded that there were no unreviewed safety questions.

There were no Technical Specification changes required.

However, changes to the FSAR/UFSAR section 7.3.6.2 required update to include the addition of the new pressure gauges to the list of local instrumentation.

M01-2-90-005 This modification involved the replacement of the RHR Shutdown Cooling High Flow Isolation Relays with a time delay relay.

During RHR shutdown cooling startups, shutdowns, and flowrate adjustments, large differential pressure (dp) spikes are originating from the shutdown cooling (SDC) high flow instruments 2E31-N012AA/BA. The dp spikes'are being sensed at the shutdown cooling suction flow elbows. The existing SDC high flow isolation relays 2B21H-K74/77 (Agastate GP) were replaced with a time delay relay (Agastat E7024AB)' set for a 1 second time delay.

This time delay relay will prevent differential pressure fluctuations seen by isolation switches from tripping.the SDC isolation logic. The Safety evaluation concluded that there were no unreviewed safety questions. There were no UFSAR or Technical Specification changes required.

l

f

)

i i

(

?

H.

Summarv of Safety Related Modificationg l

?

M01-2-90-009 i

This modification involved the upgrade of the 125 Volt Division III Batteries and rack.

The batteries were replaced due to the addition of DC loads since fuel load.

The 125V division III system has reached its maximum load for the existing size battery. In addition, these batteries have been j

replaced because the existing capacity margin to accommodate degraded capacity due to aging and low temperature operation is very small. The l

battery physical configuration did not change significantly although the new racks are longer and sightly wider.

The Safety evaluation concluded that f

there were no unreviewed safety questions. There were changes to Technical Specification section 3/4.8.2.3.2 which was based on the existing cells' nominal specific gravity of 1.210.

The new replacement cells have a nominal specific gravity of 1.215.

The changes to the UFSAR Section 8.3 and the Fire Hazard Analysis were also completed.

)

M01-2-90-010 This modification involved the replacement of the 125 Volt Division III battery charger. This existing charger was replaced because it had become unreliable and spare parts were no longer available. The physical location of the battery charger did not change. This change upsized the DC cable in order to increase the ampacity of the cable. The AC cable was also replaced with a longer cable of the same size.

The current and voltage monitoring 1

instrumentation for the battery charger remained the same.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification changes required.

However, a revision to j

the UFSAR Section 8.3 and Appendix H was required.

1 CR 88-076 4

This component replacement involved replacing'the seat ring, disc and i

stem of the RHR full flow test valve 2E12-F024B with multi-stage / multi-path, l

high pressure drop valve trim.

The RHR full flow test valve is an 18" 300 lb globe valve.

This valve is a Anchor / Darling Valve which.has been I

l reported to fail under throttle flow conditions with the valve positioned at 20% open or less.

This new multi-stage / multi path valve trim will have less cavitation problems. The Safety evaluation concluded that there were no j

unreviewed safety questions. There were no Technical Specification or FSAR/UFSAR changes required.

i l

CR 90-124, CR 90-125 CR 90-126, CR 90-127 l

1 These component replacements involved replacing existing Anderson-Greenwood bellows sealed globe valves on the Main Steam Drain valves, 2B21-r F067A/2B21-F067B/2B21-F067C/2B21-F067D with an Anchor / Darling double disk l

gate valve.

The existing motor operator was not replaced. The purpose of these replacements were to install a valve that is better able to meet through-seat leakage limits while requiring less maintenance than the existing valves. The Safety evaluations concluded that there were no unreviewed safety questions. There were no Technical Specification changes I

required. However, UFSAR Table 6.2-21 was revised to indicate the change in valve type (from globe to gate). The table also reflects the Technical Specification stroke time limit of 23 seconds, rather than " standard" stem speed.

l

H.

Summary of Safety Related Modifications ECR 92-922 (L18827) i l

This exempt change request relocated the digital point 1D #1304 from contact 14-14C to contact 5-SC to allow the open torque switch bypass to be set for the Reactor Building Close Cooling Water Supply Valve LWR 179.

This change allowed more accurate indication of valve position for the control room from the digital points.

It also allowed the open torque switch bypass to be set to the requirements of NOD-MA.1 without sacrificing the accuracy of the digital points. The Safety evaluation concluded that there were no l

unreviewed safety questions. There were no Technical Specification or l

FSAR/UFSAR changes required.

ECR 92-922 (18826)

This exempt change request relocated the digital point ID #1312 fram contact 14-14C to contact 5-5C to allow open torque switch bypass to be set for the Reactor Building Close Cooling Water Supply Valve IWR029. This change allowed more accurate indication of valve position for the control room from the digital points.

It also allowed the open torque switch bypass to be set to the requirements of NOD-MA.1 without sacrificing the accuracy of the digital points.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or TSAR /UFSAR changes required.

P01-0-90-021 This minor plant change involved modifying the attachment of the 6 GPM AC Lube Oil pump / motor assembly to the "0" Diesel Generator Skid from its present welded attachment to a bolted connection. This will allow the maintenance departments to replace the pump / motor without altering the pump / motor alignment and having to realign the assembly after each replacement. The present welded attachment forces the pump / motor assemble to be removed from its skid so that the welds from the pump / motor skid to Diesel Generator skid can be completed. The necessity to remove the pump / motor from it's as-supplied, as-aligned condition results in the need to perform post installation realignment procedures which are difficult and time consuming. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UTSAR changes required.

P01-1-90-039 P01-1-90-044 These minor plant changes involved the installation of high temperature wires to replace deteriorated lead wires of the Contaimment Monitoring Suppression Pool Inlet Isolation Valves, 1CM027/ LCM 031.

The damaged field conductors were replaced with $14 AWG Rockbestos Radiation Resistant Firewall SR lead wires.

The spliced connections were insulated with Raychem WCSF-N-ll5 heat shrink tubing. The Safety evaluations concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UFSAR changes required.

i H.

Summarv of Safety Related Modifications P01-1-90-045 P01-1-90-047 These minor plant changes involved installation of high temperature wires to replace deteriorated lead wires of the Containment Monitoring Return to Suppression Pool Isolation valves, 1CM025A/ LCM 026A.

The damaged field conductors were replaced with #14 AWG Rockbestos Radiation Resistant Firewall SR lead wires.

The spliced connections were insulated with Raychem WCSF-N-ll5 heat shrink tubing. The Safety evaluations concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UFSAR changes required.

P01-1-90-049 P01-1-90-052 These minor plant changes involved the installation of high temperature wires to replace deteriorated lead wires of the Containment Monitoring Hum.

Monitor "A" and dB" Inlet Isolation Valves, ICM017A/ LCM 018B.

The damaged field conductors were replaced with #14 AWG Rockbestos Radiation Resistant Firewall SR lead wires.

The spliced connections were insulated with Raychem WCSF-N-115 heat shrink tubing. The Safety evaluations concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UFSAR changes required.

P01-1-90-053 P01-1-90-055 These minor plant changes involved the installation of high temperature wires to replace deteriorated lead wires of the Containment Monitoring Hum.

Monitor "A" Outlet Isolation Valves, ICM019A/ LCM 020A. The damaged field conductors were replaced with #14 AWG Rockbestos Radiation Resistant Firewall SR lead wires.

The spliced connections were insulated with Raychem WCSF-N-ll5 heat shrink tubing. The Safety evaluations concluded that there tere no unreviewed safety questions. There were no Technical Specification or FSAR/UFSAR changes required.

P01-1-90-054 P01-1-90-056 These minor plant changes involved the installation of high temperature wires to replace deteriorated lead wires of the Containment Monitoring Hum.

Monitor "B" Outlet Isolation valves, ICM019B/ LCM 020B. The damaged field conductors were replaced with #14 AWG Rockbestos Radiation Resistant Firewall SR lead wires.

The spliced connections were insulated with Raychem WCSF-N-ll5 heat shrink tubing. The Safety evaluations concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UFSAR changes required.

i 1

I t

l I

H.

Summary of Safety Related Modifications i

P01-1-90-062 This miner plant change involved the installation of two (2) hydrolazing ports for the "B" RHR LPCI Injection line.

The 16-inch diameter RHR injection line, 1RH40CB, which is located outside primary containment contains high radiation levels which are not sufficiently reduced by nornal

}

flushing of the line.

These hydrolazing ports, which are 1 inch NPS, were welded to the pipes and have removable plugs to provide access into the lines for ef fective flushing as part of the dose rate reduction program.

Two (2) 1 inch diameter globe valves were also installed at the low point of IRH40CB to allow flushed water and crud from the pipe to drain into a floor drain or collection tank.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UFSAR changes required.

P01-1-90-069 P01-1-90-070 These minor plant changes involved modifying the attachment of the 6 GPM AC Lube Oil pump / motor assembly to the "1A" and "1B" Diesel Generator Skid from its present welded attachment to a bolted connection. This will allow the maintenance departments to replace the pump / motor without altering

[

the pump / motor alignment and having to realign the assembly after each replacement. The present welded attachment forces the pump / motor assemble to be removed from its skid so that the welds from the pump / motor skid to Diesel Generator skid can be completed. The necessity to remove the i

pump / motor from it's as-supplied, as-aligned condition results in the need to perform post installation realignment procedures which are difficult and time consuming. The Safety evaluations concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UTSAR changes required.

P01-1-91-513 This minor plant change rerouted the conduit for the RPS cable 1RPO42 which interferes with the opening of the hinged cover of the Main Turbine l

Control Valve #3's limit switch enclosure. This change also included 6

installation of a new junction box and installation of conduit and cable l

between the new and existing junction boxes.

Changes to existing supports i

and removal of supports was also required. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical j

Specification or FSAR/UTSAR changes required.

P01-1-91-516 i

1 This minor plant change replaced the existing flexible wedge gate valve for the Reactor Water Cleanup 1G33-F001 valve with a double disk gate valve.

The valve was replaced so that it would be able to meet its through-seat leakage limits while requiring less maintenance that the existing valve.

The new valve was furnished with live loaded packing and did not require a leakoff line, therefore, the existing leakoff line was cut, capped and abandoned in place. Also solenoid valve IE31-r005r1 and temperature element lE31-F016F1, which provided isolation and leak detection for the leakoff line, were abandoned in place.

The control switch for the soleneid valve was removed from the Main Control Room panel.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification changes requirsd.

However, Figure 7.3-7 sheet 2 of the UYSAR was revised to indicate that valve 1G33-r001 no longer required a leakoff line, i

I i

l i

H.

Summarv of Safety Related Modifications P01-1-91-517 This udnor plant change replaced the existing flexible wedge gate valve for the Reactor Water Cleanup 1G33-FOO4 valve with a double disk gate valve.

The valve was replaced so that it would be able to meet its through-seat leakage limits while requiring less maintenance that the existing valve.

The Safety evaluation concluded that there were no unreviewed safety I

questions. There were no Technical Specification or FSAR/UFSAR changes required.

P01-1-91-518 This minor plant change replaced the existing flexible wedge gate valve for the Reactor Core Isolation Cooling System (RCIC) lE51-F008 valve with a double disk gate valve.

A function of this valve is to provide containment isolation in the unlikely event t hat the RCIC steam supply line should break or rupture. The existing valve exhibits unpredictable behavior when r

subjected to blowdown conditions of differential pressure and flow.

The new l

Anchor / Darling double-disc parallel slide gate valve was installed because they have been successfully type-tested under simu3ated blowdown conditions.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UTSAR changes i

required.

j i

- P01-1-91-530

{

P01-1-91-531 These minor plant changes involved repairing the seal between the RHR f

heat exchangers, IE12-F001B/lE12-F001A partition plate end tube sheet groove. This involved welding two sill plates, one on each side of the RHR heat exchanger partition plate.

FURMANITE FSC-N-3B was injected between the e

sill plates.

The sill plate and the Furmanite configuration will function j

as a seal to minimize bypass flow between the inlet and outlet plenums of the heat exchanger thus improving heat exchanger performance. The Safety evaluation concluded that there were no unreviewed safety questions.

There j

were no Technical Specification or FSAR/UTSAR changes required.

l P01-1-91-532 This minor plant change involved removing the existing triangular i

refueling mast and installing the NF-500 tubular refueling mast.

The NF-500 l

refueling mast is mounted to the fuel grapple and provides a torsionally and externally rigid structure by which the grapple head's motion can be controlled. This cylindrical mast also provides attachment points for the hoisting cables, air lines, and electrical connectors supplying the fuel grapple. The Safety evaluation concluded that there were no unreviewed safety questions. A revision to the Technical Specification Sections 3.9.6 and 4.9.6 due to the increased weight of the mast sections was required. A 4

safety evaluation consisting of a Significant Hazards Analysis (SEA) under i

the Fuel Handling Accident (FHA) was campleted. There were no revisioris required to the FSAR/UTSAR.

I l

1 i

l 1

I I

H.

Summarv of Safety Related Modifications P01-1-91-539 This minor plant change removed the existing PSA mechanical snuober NB13-1001S and replaced it with a new Lisega hydraulic anubber.

Due to the reactor vessel head environment, the NB13-1001S PSA mechanical snubber has experienced a high failure rate.

Snubber failures have led to increased e

maintenance, inspection time, repair / replacement cost along with increased l

radiation exposure to plant personnel. The Hydraulic snubber is better

[

suited to withstand the reactor vessel head conditions and thus will reduce maintenance, inspe ctions, and exposure time.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UFSAR changes required.

P01-1-91-546 This minor plant change replaced the existing 5 ft-lb DC motor on the l

Reactor Core Isolation Cooling (RCIC) Full Flow Test Downstream Stop valve IE51-F059 with a 7.5 ft-lb DC motor.

The existing thermal overload relay l

was replaced with a Westinghouse type AN23P overload relay with an FH24 heater.

The thermal overload setting was reset to 100% and the MCC circuit l

breaker magnetic setting was reset.

This change increased the thrust that i

the actuator delivers to the valve stem.

The Safety evaluation concluded i

that there were no unreviewed safety questions. There were no Technical l

Specification or FSAR/UTSAR changes required, t

i i

P01-1-91-551 This minor plant change replaced the existing 2 ft-lb motor on the "B" RHR Supply from Fuel Pool emergency Makeup Pump Downstream Stop valve, IE12-F093, with a 5 ft-lb motor.

This also required the changing of the MCC thermal overload setting. The IE12-F093 valve indicated that additional l

actuator thrust was required to ensure the valve's reliability under design i

basis flow and differential pressure conditions. This change increased the thrust that the actuator delivers to the valve stem.

The Safety evaluation j

concluded that there were no unreviewed safety questions. There were no i

Technical Specification or FSAR/UFSAR changes required.

l P01-1-91-552 P01-1-91-053 These minor plant changes replaced the Limitorque gearing overall ratio (OAR) to 82.0 and change the spring pack to model 0301-111 on the "lC" and j

"1B" RHR Pump Min. Flow Valves lE12-F064C/IE12-F064B. This change increased the thrust that the actuator delivers to the valve stem by replacing the motor pinion gear and the worm shaft gear to increase the j

operator's overall gear ratio.

This replacement did not alter the actuator i

weight.

The Safety evaluations concluded that there were no unreviewed safety questions. There were no Technical Specification changes required.

i However, Table 6.2-21 of the UFSAR was revised to properly show the valves stoke timing.

i i

?

\\

l

)

I e

H.

Summary of Safety Related Modifications P01-1-91-556 i

This minor plant change replaced the existing 2 ft-lb motor on the 1A RHR Blowdown Isolation valve lE12-F049A with a 5 ft-lb motor.

The thermal overload setting was reset.

Additional actuator thrust is required to ensure the valves's reliability under design basis flow and differential j

pressure conditions This has increased the thrust that the actuator g

delivers to the valve stem.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or 4

FSAR/UTSAR changes required.

P01-1-91-559 This minor plant change involved the addition of a set of piping j

flanges in the RCIC Head Spray Piping Line (1RI24B) just upateam of the N-7 RCIC head Spray nozzle on the Reactor Pressure Vessel. The installation was intended to ease the process of RPV disassembly and reassembly during refueling by alleviating the need to withdraw the " snout" portion of the existing spray nozzle from the RPV during removal of the Head Spray piping.

The existing configuration has resulted in extended reassembly times and unnecessary dose.

It has also introduced the possibility of damaging both the vessel nozzle and the snout due to the difficulties involved in making up the vessel to snout flange connection. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UTSAR changes required.

P01-1-91-561 This minor plant change installed an interposing relay in the Diesel Generator "0" Main Feed Breaker (1413) Closing Circuit, thus providing a greater design margin between the minimum acceptable pickup voltage of the circuit breaker closing coil and the expected voltage across the coil during emergency operation. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UFSAR changes required.

P01-1-92-511 i

This minor plant change replaced the undervoltage relays 1427-AP270A/B in the EST Division I degraded voltage protection circuit. The main function of these undervoltage relays is to monitor the voltage on the Division I 4KV bus and initiate an automatic transfer from of f-site power to the associated diesel generator when a degraded voltage condition is detected. This was essentially a "like-for-like" replacement, and the scope of work consisted of only minor electrical changes. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UTSAR changes required.

l i

l l

i i

I 1

H.

Summarv of Safety Related Modifications i

j P01-1-92-512 i

l This minor plant change replaced the undervoltage relays 1427-AP271A/B in the EST Division II degraded voltage protection circuit. The main function of these undervoltage relays is to monitor the voltage on the Division II 4KV bus and initiate an automatic transfer from off-site power to the associated diesel generator when a degraded voltage condition is detected.

This was essentially a "like-for-like" replacement, and the scope of work consisted of only minor electrical changes. The Safety evaluation j

concluded that there were no unreviewed safety questions. There were no i

Technical Specification or FSAR/UTSAR changes required.

P01-1-92-513 4

This minor plant change replaces the undervoltage relays 1427-AP272A/B in the EST Division III degraded voltage protection circuit. The main l

function of these undervoltage relays is to monitor the voltage on the Division III 4KV bus and initiate an automatic transfer from off-site power to the associated diesel generator when a degraded voltage condition is d

detected. This was essentially a "like-for-like" replacement, and the scope of work consisted of only minor electrical changes. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UFSAR changes required.

a P01-1-92-514 This minor plant change replaced the 120 VAC contactor coils with 480 VAC coils in the open and close motor control circuits of HPCS Injection Valve lE22-F004.

Currently, each open and close motor contactor coil is i

connected in parallel with an auxiliary relay, and both devices are directly a

interlocked with the other switch and relay contacts in the control circuit.

4 This change removes the existing 120 VAC contactor coils from the control circuit and connects the new 480 VAC coils to the 480 VAC power supply through an auxiliary relay contact.

To accommodate this change, replacement auxiliary relays were also required since the contacts of the existing P&B relays are not rated for 480 VAC.

The circuit logic remained unchanged since the contractor coils will be interlocked with the control circuit thorough the auxiliary relays. The Safety evaluation concluded that there I

were no unreviewed safety questions. There were no Technical Specification or FSAR/UFSAR changes required.

P01-1-92-519 This minor plant change installed an interposing replay to each of the I

open and close control circuits of the motor operator for the RHR Suction Cooling Inboard Isolation Valve 1E12-F009.

Previously, the open and close motor starter (or contactor) coils were directly interlocked with the rest of the switch and relay contacts in the control circuit. This change has each contactor coil indirectly interlocked with the control circuit through an interposing relay. This replaced the contactor coils with relays in the open and close circuits and reconnected the contactor coils to a contact from these new interposing relaye. The other side of these relay contacts are wired to the "line" side of the 120 VAC power supply.

The logic of the circuit remained unchanged. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UFSAR changes required.

i r-

,,,,_,__F,__

_,,u,_,,,,

v.

m W.

H.

Summary of Safety Related Modifications P01-2-90-013 This minor plant change removed the temporary support on Jet Pump Sensing Lines (#5 and #15), added new supports and reinforced the existing lower bracket support and changed the natural frequecy of the line by.

changing the free-span length. The new sensing line clamp and beam are designed to rigidly restrain the jet pump sensing line against the lower support bracket and at a point 7.75 inches lower, reinforcing the original bracket attachment to the sensing line.

The clamp assemble consists of a beam which supports the sensing line, a ring which encircles the jet pump diffuser and sensing line, a wedge block which wedges the slide against the jet pump diffuser, and a wedge adjustment screw which drives the wedge block down to provide the necessary clamping force against the beam. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification changes required.

However, a change to the UFSAR to clarify the new clamp arrangement was required.

P01-2-90-023 This minor plant change installed a cmall drain line to the upper and lower bearing oil reservoirs of the LPCS pump motor.

This drain line replaced an oil plug, and consists of a 1/2-inch manually operated globe valve and connecting piping. The addition of the drain line will facilitate oil sampling activities associated with the pump motor. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UFSAR changes required.

P01-2-90-030 This minor plant change installed a vent to the bonnet for the EHR suppression Pool valves 2E12-F004B. This included installation of two (2) manual gate valves (2E12-F428B and 2E12-F429B) and connecting pipe to the bonnet of valve 2E12-F004B and return co the process header, 2RH01AB.

Addition of the vent line has relieved the hydraulic locking problems associated with this valve.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification changes required.

However, the UFSAR Table 6.2-21 was revised to add the 2E12-F428B valve as a containment isolation valve.

P01-2-90-035 This minor plant change installed a vent to the bonnet for the RHR suppression Pool Valve 2E12-F004A.

This included installation of two (2) manual gate valves (2E12-F428A and 2E12-F429A) and connecting pipe to the bonnet of valve 2E12-F004A and return to the process header, 2RH01AA.

Addition of the vent line has relieved the hydraulic locking problems associated with this valve.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification changes required.

However, changes to the UFSAR Table 6.2-21 were revised to add the 2E12-F428A valve as a containment isolation valve.

t 3

i H.

Summary of Safety Related Modifications P01-2-90-034 i

This minor plant change added transient suppressors or varistors across the line and neutral of the 120 VAC feeds to the GE NUMAC power supplies for the nain steamline radiation monitors, 2D18-K610A, B,

C, and D.

The varistors were installed in control room panels 2H13-P635 and 2H13-P636.

This change was recommended by GE in SIL No. 499 to protect the power supplies from low energy transients such as voltage spikes.

Thus, this

[

change will improve the reliability of these monitors. This change did not alter the function of these monitors which is to initiate a reactor scram on high main steamline radiation. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification and FSAR/UFSAR changes required.

P01-2-90-036 This minor plant change provided a lifting device for the CRD pumps.

This change installed a lifting beam with support steel to assist the acintenance departments with performance of maintenance activities. The Safety evaluation concluded that there were no unreviewed safety questions.

There were no Technical Specification and FSAR/UTSAR changes required.

P01-2-90-039 This minor plant change involved installation of two (2) hydrolazing ports for the "B" RHR LPCI Injection line.

The 16-inch diameter RHR injection line, 2RH40CB, which is located outside primary containment contains high radiation levels which are not sufficiently reduced by normal flushing of the line.

These hydrolazing ports, which are 1 inch NPS, were welded to the pipes and have removable pJ ugs to provide access into the lines for effective flushing as part of dose rate reduction program. Two (2) 1 inch diameter globe valves were also installed at the low point of 2RH40CB to allow flushed water and crud from the pipe to drain into a floor drain or collection tank.

The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UTSAR changes required.

P01-2-90-043 This minor plant change installed a small drain line to the upper and lower bearing oil reservoirs of the 2C RHR pump motor.

This drain line replaced an oil plug, and consists of a 1/2-inch manually operated globe valve and connecting piping. The addition of the drain line will facilitate oil sampling activities associated with the pump motor.

The Safety i

evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UFSAR changes required.

P01-2-90-053 This minor plant change involved the installation of high temperature wires to replace deteriorated lead wires of the Containment Monitoring Suppression Pool Common Outlec Isolation valve, 2CM033. The damaged field conductors were replaced with #14 AWG Rockbestos Radiation Resistant Firewall SR lead wires.

The spliced connection was insulated with Raychem WCSF-N-115 heat shrink tubing. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UTSAR changes required.

l

\\

l

H.

Summary of Safety Related Modifications i

P01-2-90-058 This minor plant change involved the installation of high temperature wires to replace deteriorated lead wires of the Containment Monitoring Suction From Drywell Isolation valve, 2CM022A. The damaged field conductors i

were replaced with #14 AWG Rockbestos Radiation Resistant Firewall SR lead wires.

The spliced connection was insulated with Raychem WCSF-N-ll5 heat j

shrink tubing. The Safety evaluation concluded that there were no unreviewed i

safety questions. There were no Technical Specification or FSAR/UTSAR changes required.

t P01-2-90-061 This minor plant change involved the installation of high temperature l

wires to replace deteriorated lead wires of the Containment Monitoring Suppression Pool Return Isolation valve, 2CM025A. The damaged field I

conductors were replaced with #14 AWG Rockbestos Radiation Resistant Firewall SR lead wires.

The spliced connection was insulated with Raychem WCSF-N-115 heat shrink tubing. The Safety evaluation concluded that there i

were no unreviewed safety questions. There were no Technical Specification or FSAR/UTSAR changes required.

P01-2-90-067 This minor plant change involved the installation of high temperature wires to replace deteriorated lead wires of the Containment Monitoring i

Suppression Pool Inlet Isolation Valve, 2CM018B. The damaged field conductors were replaced with #14 AWG Rockbestos Radiation Resistant Firewall SR lead wires.

The spliced connection was insulated with Raychem WCSF-N-ll5 heat shrink tubing. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UTSAR changes required.

P01-2-90-070 i

P01-2-90-071 1

These minor plant changes involved modifying the attachment of the 6 l

GEM AC Lube Oil pump / motor assembly on the "2A" and "2B" Diesel Generator Skids from their present welded attachment to a bolted connection. This will allow the raaintenance departments to replace the pump / motor without altering the pamp/ motor alignment and having to realign the assembly after each replacement. The present welded attachment forces the pump / motor

(

assemble to be removed from its skid so that the welds from the pump / motor skid to Diesel Generator skld can be completed. The necessity to remove the l

pump / motor from it's as-supplied, as-aligned condition results in the need to perform post installation realignment procedures which are difficult and time consuming. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UTSAR changes required.

H.

Summ=rv of Safety Related Modificatiens P01-2-90-072 This minor plant change involved the installation of high temperature wires to replace deteriorated lead wires of the Containment Monitoring 7

Suppression Pool Purge Isolation Valve, 2INO31. The damaged field conductors were replaced with #14 AWG Rockbestos Radiation Resistant Firewall SR lead wires.

The spliced connection was insulated with Raychem WCSF-N-115 heat shrink tubing. The Safety evaluation concluded that there were no unreviewed safety questions.

There were no Technical Specification or TSAR /UTSAR changes required.

P01-2-91-506 t

This minor plant change replaced the existing flexible wedge gate valve I

for the Reactor Water Cleanup 2G33-F001 valve with a double disk gate valve.

The valve was replaced so that it would be able to meet its through-seat leakage limits while requiring less maintenance. The new valve was furnished i

with live loaded packing and did not require a leakoff line, therefore, the existing leakoff line was cut, capped and abandoned in place. Also solenoid j

valve 2E31-!005F1'and temperature element 2E31-701671, which provided isolation and leak detection for the leakoff line, was abandoned in place.

The control switch for the solenoid valve was removed from the Main Control Room panel.

The Safety evaluation concluded that there were no unreviewed i

safety questions. There were no Technical Specification changes required.

However, Figure 7.3-7 sheet 2 of the UFSAR was revised to indicate that valve 2G33-r001 no longer required a leakoff line.

P01-2-91-507 i

This ndnor plant change replaced the exist!ng flexible wedge gate valve for the Reactor Water Cleanup 2G33-F004 valve with a double disk gate valve.

f The valve was replaced so that it would be able to meet its through-seat leakage limits while requiring less maintenance. The Safety evaluation concluded that there were no unreviewed safety questions. There were no i

Technical Specification or FSAR/UFSAR changes required.

I P01-2-91-511 This minor plant change replaced the existing 15 ft-lb motor for the RHR Shutdown Cooling Isolation Valve 2E12-T008, with a 25 ft-lb motor.

The thermal overload setting was reset.

Additional actuator thrust is required to ensure the valves's reliability under design basis flow and differential pressure conditions. This will increase the thrust that the actuator i

delivers to the valve stem.

The Safety evaluation concluded that there were l

no unreviewed safety questions. There were no Technical Specification or l

FSAR/UTSAR changes required.

j

l i

l 5

f H.

Summary of Safety Related 1iodifications f

P01-2-91-512 i

P01-2-91-513 l

l These minor plant changes replaced the Limitorque gearing overall ratio (OAR) to 82.0 and change the spring pack to model 0301-111 on the 2A RHR Pump Min. Flow Valve 2E12-F064A and the LPCS Pump Full Flow Valve 2E21-F012.

This change will increase the thrust that the actuator delivers to the valve stem by replacing the motor pinion gear and the worm shaft gear to increase the operator's overall gear ratio.

This replacement did not alter the l

actuator weight. The Safety evaluations concluded that there were no unreviewed safety questions. There were no Technical Specification changes required. However, Table 6.2-21 of the UFSAR was revised to properly show l

the valves stoke timing.

i P01-2-91-514, P01-2-91-515 P01-2-91-516, P01-1-91-517 i

These minor plant changes replaced the existing 2 ft-lb motor for the Reactor Building Closed Cooling Water Isolation Valves 2WR040/2WR179 2WR029/2WR180, with a 5 ft-lb motor.

The thermal overload setting was reset.

Additional actuator thrust is required to ensure the valves's reliability under design basis flow and differential pressure conditions 3

This will increase the thrust that the actuator delivers to the valve stem.

l The Safety evaluations concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UTSAR changes required.

P01-2-91-524 P01-2-91-525 These minor plant changes involved repairing the seal between the RHR heat exchangers, 2E12-F001B/2E12-F001A partition plate end tube sheet groove. This involved welding two sill plates, one on each side of the RHR heat exchanger partition plate. FURMANITE FSC-N-3B was injected between the sill plates. The sill plate and the Furmanite configuration will function as a seal to minimize bypass flow between the inlet and outlet plenums of the heat exchanger, thus improving heat exchanger performance. The Safety evaluations concluded that there were no unreviewed safety questions. There were no Technical Specification or FSAR/UTSAR changes required.

P01-2-91-555 i

i This minor plant change installed an interposing relay in the Diesel Generator "0" Main reed Breaker (2413) Closing Circuit, thus providing a l

greater design margin between the minimum acceptable pickup voltage of the circuit breaker closing coil and the expected voltage across the coil during l

emergency operation. The Safety evaluation concluded that there were no unreviewed safety questions. There were no Technical Specification or i

FSAR/UTSAR changes required.

}

t l

i t

.,. - - -..l

~

I.

Symm'ry of ECCS OutD9c5 This information has been reported monthly in LaSalle's NRC Monthly Reports (Section II.F.2) dated January 1992 through December 1992.

J.

Survey _nf Evaluation _ItesnLtE_nf_ChlDrine__ Shipments _by_Ratge_ppe_1.he 11Linoisl izan Not required for 1992, the survey was last completed in 1991.

K.

Sumary_pf EventE_.Vio}ating_ Technical __ Specification 3.4.5 - Primary Englant_loding_Spiling Exceeding Allowable Lim _i 1 During this reporting period, January 1, 1992 through December 31, 1992, there were no violations of Technical Specification 3.4.5, Primary Coolant Iodine Spikes Exceeding Allowable Limits.

ATTACHMENT A SAFETY-RELATED MAINTENANCE COMPLETED j

(NON-OUTAGE RELATED) j

---

UNIT =0 ---------------------------------------

WRNUM SYSTEM EPN DESCRIPTION LOO 609 VC ORG047 RECEIVER INLET STOP VALVE LEAKING FREON

[

LO2193 VC OPDS-VC051 RECALIBRATE SUPPLY FAN FLOW SWITCH LD5347 DG ODG01S CLEAN & INSPECT AIR COMPRESSOR SKID LO6050 DG ODOO2T CLEMI DAY TANK VENT SCREENS LO8772 DG ODOO1P FUEL PUMP OILER LEAKING OIL LO9622 VC 0FR-VC068 REMOUNT EMERGENCY MAKEUP DIFF PRESS RECORDER j

L10433 DG DJT-DG065 REPLACE WATT TRANSDUCER FOR EDSFI L10615 NR OC51-000 ICPS UNSTABLE, WILL NOT STAY WITHIN TOLERANCE L10672 DG ODG02JB DG PANEL REPLACE CAPACITOR L10744 VC OPI-VC133 A VC COMPRESSOR LEAKING SUCTION PRESSURE GAUGE L11805 DG ODG023B INSPECT AIR COMPRESSOR CHECK VALVE LI2015 VC 0FZ-VC001F OVC30YA DAMPER WOULD NOT OPEN L12454 DC ODC19E REPLACE RSH 125V BATTERY CHARGER UUGS L13115 DG ODG08DA INSPECT & VERIFY PROPER OPERATION OF AIR DRYER L13116 DG ODGOBDB INSPECT & VERIFY PROPER OPERATION OF AIR DRYER L13126 DG ODG08CA INSPECT AIR COMP SUCTION AND DISCHARGE VALVE L13127 DG ODG08CB INSPECT AIR COMP SUCTION AND DISCHARGE VALVE 3

L13144 DG OSI-DG28B REPLACE DG FREQUENCY METER, READS LOW

{

L13149 DG ODG08DA VERITY AIR DRYER DRAIN TRAP OPERATION L13150 DG ODGOBDB VERITY AIR DRYER DRAIN TRAP OPERATION L13178 AP OAP09E PERFORM RSH BUS BREAKER CUBICLE INSPECTION f

L13403 VC OTT-VC145 AIR INLET TEMPERATURE STICKING AT 175 DEG L13650 VC 0XY-VC165A DETECTOR LOCK INOPERABLE L13745 VC 0XY-VC125A DETECTOR CASSETTE CARTRIDGE INOPERABLE l

L13754 AP OAP09E CLEAN & INSPECT BUS 041 GROUND / TEST DEVICE L13782 VC OVCO2SB HUMIDIFIER SCALE BUILD UP IN VESSEL L13879 VC 0FZ-VC047A ODOR EATER INLET VALVE FAILED OPEN i

L14113 RD OC11-000 DATA DRIVER CARD BAD L14121 AP OAP49E REPLACE BROKEN HANDLE ON MCC 031 L14332 VC OTZ-VC043 REPLACE TEMPERATURE CONTROLLER ON ACTUATOR L14617 DC OPL96J REPLACE CABLE ON STANDBY BATTERY PACK I

L14721 DG ODG01K ADJUST GOVENOR MOTOR FRICTION CLUTCH L14981 DG OTS-DG104 REPLACE TEMPERATURE SWITCH, FAILED CAL L14982 DG OTS-DG117 REPLACE SENSOR, FAILED CALIBRATION L15095 DG ODC20E ADJUST TSC DIESEL FLOAT VOLTAGE L15272 VC OKY-VC125B AMMONIA DETECTOR ALARMS AT 15 PPM L15385 VC OVC01MA HUMIDITY VALVE FAILED TO CYCLE j

L15479 DG OPDS-DG036 REPLACE COOLING WATER PUMP STRAINER L15600 VC OKY-VC087B AIR DUCT DETECTOR FAILED TO ALARM L15732 AP ODG01P DG COOLING WATER PUMP INSTALL TEED BREAKER L15762 VC OXY-VC125B REPLACE AMMONIA DETECTOR TAPE DRIVE L15797 VC OVCO2SB REPLACE STEAM GENERATOR RELIEF VALVE L15875 DG OPDS-DG051 REPLACE FUEL FILTER SWITCH REVERSE TUBING L15884 DG OD0004 DISCHARGE VALVE FULL OPEN LIGHT INOPERABLE L16004 VC OKY-VC125A AMMONIA DETECTOR ALARMING AT 12.5 PPM L16130 DG ODG000 INSPECT DG HACR-1 TEST SWITCH L16246 DG ODG09K TSC DIESEL WATER JACKET HEATER INOPERABLE s

C

-e

,--r--

~

~n-r-,---

_.,, _ _e

i i

I ATTACHMENT A SAFETY-RELATED MAINTENANCE COMPLETED (NON-OUTAGE RELATED) l I

---

~~----------------UNIT =0---------------------------------------

l WRNUM SYSTEM EPN DESCRIPTION l

L16248 DC OPL96J DC-DC CONVERTER DRAWING EXCESSIVE CURRENT L16309 VC OXY-VC165A MOISTURE IN AMMONIA DETECTOR FLOW METER L16401 VC OVCO2CB VENT RETURN FAN TRIPS ON THERMALS i

L16507 DG ODG044 DRAIN LINE U-BOLT NOT SECURED ON INLET HEADER L16591 VC OXY-VC165B AMMONIA DETECTOR ALAPJ4ING j

L16723 DC ODC21E REPLACE TSC 125 BATTERY CHARGER TIMER HANDLE L16737 VC OVC01SA REPLACE TERMINAL BOX COVER L16885 DG OPS-DG042A AIR SKID SWITCH DRIFTING, ALARM RECIEVED L16920 DG ODG01K INSTALL 20 TEMP CYLINDER TEST VALVES L16926 DG OLI-DOOO2B REPLACE DG LEVEL GAUGES l

L16964 VC OXY-VC165B AMMONI A DETECTOR TAPE NOT ADVANCING I

L17047 VC OVC01TB INSPECT SUPPLY AIR FILTER UNIT DAMPER SUPPORT l

L17189 DC ODC20E CORROSION ON TSC BATTERY CELL TERMINALS l

L17552 AP OAP49E REPLACE BROKEN HANDLE, MCC031 CUBICLE C-4 i

L17912 VC OXY-VC087A OXY-VCOB7B FAILED CALIBRATION L17971 DG ODG09K TSC DIESEL GEN VENTILATION DAMPERS INOPERABLE t

L18111 DC ODC18E RSH 125 VDC BATTERIES CELLS 55 & 58 LOW L18146 NR ONR000 VOLTAGE PREREGULATOR NEGATIVE OUTPUT L19057 DG ODGOBCB AIR COMPRESSOR ATTERCOOLER AIR LEAK t

L19197 VC 0FSY-VC004X REPLACE DEFECTIVE RELAY I

i f

I t

[

I i

1 l

l ATTACHMENT A SAFETY-RELATED MAINTENANCE COMPLETED (NON-OUTAGE RELATED)

UNIT =1 _______________________________________

WRNUM SYSTEM EPN DESCRIPTION

?

LO4672 AP IE32-F008 REPLACE 136X-2 CUBICLE F5 LO6049 DG IDOOOO CLEAN DG DAY TANK VENT SCREENS LO6051 DG IDOOOO CLEAN DG DAY TANK VENT SCREENS l

LO6377 RH PDI-1E12-N029B RHR INJECTION LINE SWITCH DRIFTS LO9362 DC 1ER-DC015 DIV I GRND DETECTOR +15V GROUNND ON RECORDER LO9594 RD LT-1C11-N012A INSPECT CRD SCRAM DISCH LEVEL TRANSMITTER LO9700 RH ZI-1E12-R608A CONTROL RM INDICATOR DOES NOT RESPOND LO9705 NB RE-1B13-D193FB LPRM 32-25D HIGH SPIKES LO9806 NR INR000 1D APRM HIGH NEUTRON TRIP RECEIVED LO9863 RD TR_1C11-R018A CRD 30_11 TEMPERATURE PEGS UPSCALE LO9953 RI IE51-C003 TAKE RCIC WATER LEG PUMP CURRENT READINGS L10012 HP IE22-C001 PERFORM HPCS MOTOR MEGGER L10013 VG IVG02C MONITOR SBGT FAN LINE C'RRENT ON MOTOR L10108 AP 1AP76E-F5 REPLACE MCC 135Y-2 COhTROL TRANSFORMER I

L10435 DG 1E22-5001 CALIBRATE IB DG METERS FOR EDSFI L10662 DG ILS-DOOO4 SECURE DAY TK LEVEL SWITCH L10752 DG 1DG01K SECURE CONDUIT ON DIESEL SKID L10841 RD 1C11-000 RMCS ACTIVITY CONTROLS INACCURATE INDICATION L10967 DG 1PS-DG047 REPLACE CRANKCASE PRESSURE SWITCH L11141 AP 1AP58E MCC 132X-1/D2 REMOVE OVERLOAD RELAY L11834 RD 1C11-D001-100 HCU 30-55 ACCUMULATOR LEAKING L12319 RD IC11-D001-004 REPL1"E LIGHT ASSEMBLY FOR CRD 26-55 L12438 DG 1D0024 FUEL OlL SUCTION VALVE FAILED TO OPEN L12451 DG ITI-DG004 REPLACE DG COOLING WATER DISCH INDICATOR L12470 RI IE51-C003 RCIC WATER LEG PUMP LOSING OIL L12632 DC 1ER-DC009 GROUND DETECTOR RECORDER NO INDICATION L12673 NB 1H13-P629 REPLACE LPCS/RHR RELAY FUSE L12674 NB 1H13-P629 REPLACE LPCS/RHR RELAY FUSE L12708 RD UR-1C11-R611 INSPECT SCRAM TIME RECORDER L12795 RI 1E51-F045 RCIC TURBINE SUCTION VALVE PACKING LEAK L12914 RI FCK_1E51-R600 REPLACE RCIC FLOW CONTROLLER THUMBWHEEL L12940 RH 1E12-F073A RHR VENT INDICATOR DOES NOT RESPOND L13019 DG IDG08CA REPLACE STARTING AIR COMPRESSOR BELTS L13045 RH PS_1E12-N018 INSTRUMENT STOP VALVE LEAKS L13111 DG 1DG09DA INSPECT FOR PROPER OPERATION OF AIR DRYER L13112 DG IDG09DB INSPECT FOR PROPER OPERATION OF AIR DRYER L13114 DG 1DG08DB INSPECT FOR PROPER OPERATION OF AIR DRYER L13137 DC 1ER-DC009 NO CHART INDICATION L13151 DG IDG0BDA INSPECT OPERATION OF AIR DRYER DRAIN TRAF L13152 DG 1DG08DB INSPECT OPERATION OF AIR DRYER DRAIN TRAP L13153 DG IDG09DA INSPECT OPERATION OF AIR DRYER DRAIN TRAP L13154 DG IDG09DB INSPECT OPERATION OF AIR DRYER DRAIN TRAP L13620 DG 1E22-5001 DG FUEL PUMP FAILURE ALARMS r

L13621 DC 1EI-DC056 CALIBRATE THE 250VDC CHARGER VOLTMETER I

L13687 AP 1AP38E REPLACE BREAKER C5 HANDLE I

L13786 NR 1C51-000 REPLACE APRM'S K18 RELAYS I

ATTACHMENT A SAFETY-RELATED MAINTENANCE COMPLETED (NON-OUTAGE RELATED)

---

UNIT =1 ---------------------------------------

WRNUM SYSTEM EPN DESCRIPTION

{

L13792 DG IDGOBCA INSPECT AIR COMP SUCTION & DISCHARGE VALVE

{

L13793 DG 1DG08CB INSPECT AIR COMP SUCTION & DISCHARGE VALVE L13837 DG PI-1E22-R514 REPLACE FUEL FILTER GAUGE L13938 RD 1C11-D001-010 REPLACE HCU 14-51 DIRECTIONAL CONTROL VALVE f

?

L14256 MS LIS-1B21-N702B LOW WATER LEVEL INDICATOR STICKING L14370 DG IDG09DA AIR DRYER TEMPERATURE CONTROLLER INOPERABLE 1DG09DB AIR DRYER TEMPERATURE CONTROLLER INOPERABLE f

L14371 L14514 NB HK-1B21-R613 JET PUMP TOTAL FLOW INDICATES LOW L14553 DG 1DG01K LOW LUBE OIL PRESSURE ALARM FAILED TO CLEAR L14722 DG IPS-DG042A AIR BANK DIGITAL ALARM UP IN CONTROL ROOM L14725 DC IDC36E CHARGER TRIPS AC CIRCUIT BREAKER L14938 RI 1E51-C002 RCIC TURBINE TRIPPED ON OVERSPEED l

L14983 RI 1E51-C002 RCIC TURBINE TRIPPED ON OVERSPEED L15064 HP IE22-F015 KESET SUCTION VALVE LIMIT SWITCH ROTOR l

L15065 HP 1E22-F012 RESET MIN FLOW LIMIT SWITCH ROTOR L15073 RI 1E51-F045 RCIC INLET VALVE PACKING LEAK L15077 NR INR000 LPRM 32-17B DOWNSCALE LAMP FAILED TO LIGHT L15106 VY ITI-VY0l6 RHR ROOM TEMP READS LOW L15221 RI 1E51-C003 REPLACE RCIC WATER LEG PUMP MOTOR BEARINGS L15269 HG IPA 12J HYDROGEN RECOMB LIGHT FAILED L15310 VG IVG01C REPLACE SBGT SUPPLY BROKEN TERMINAL BLOCK L15612 DG IE22-C302A AIR COMPRESSOR, REMOVE DUCT TAPE & WIRES L15731 AP ODG01P REPLACE COOLING WATER PUMP FEED BREAKER L15743 RD IC11-D001-011 INSTRUMENT BLOCK CAP LEAKING NITROGEN L15785 DG 1AP80E REPLACE DG OIL CIRC PUMP OVERLOAD RELAY L15787 RH PI-1E12-R002C RHR PUMP GAUGE STICKING L15807 RD IC11-D001-038 HCU 06-35 NITROGEN LEAKING L15819 RH 1E12-C002B RHR PUMP SEAL COOLER, REPLACE BLOWN GASKET l

L15820 RH IE12-C002C RHR PUMP SEAL COOLER, REPLACE BLOWN GASKET L15883 AP 1APO4E D/G BREAKER NOT CLOSING L15898 NR RY-1C51-K605GM "A" APRM ALARMING ABNORMALLY L15938 RD 1C11-D001-000 REPLACE SCRAM PILOT VALVES AT HCU 06-27 L16034 NR RR-1C51-R603D RECORDER DIGITAL DISPLAY INOPERABLE L16044 HR RR-1C51-R603A SECURE GREEN PEN CARRIAGE r

i L16132 DG IE22-S001 DG HACR-1 RELAY INSPECTION L16202 DG 1DG01K REPLACE SPACE HTR OVERLOAD RELAY L16210 DG 1DG01K DG SOAKBACK OIL PUMP FAILS TO RUN L16243 DG IPS-DG042A 1A DG AIR COMP, LOW AIR PRESS ALARM UP I

L16403 RD IC11-D001-012 HCU 14-55 NITROGEN LEAK L16543 RD 1Cll-D183-126 HCU 42-07 VALVE LEAKAGE L16638 RH TE-1E12-C002BB RER PUMP ERRATIC TEMPERATURE INDICATION l

L16723 RD IC11-F022A REPLACE CRD FILTER VENT HANDWHEEL L16788 RH 1E12-F409A RER STRAINER CLOGGED NEEDLE VALVE L16918 DG 1DG01K INSTALL 20 TEMPERARY CYLINDER TEST VALVES L16919 DG IE22-S001 INSTALL 20 TEMPERARY CYLINDER TEST VALVES j

L16923 DG ILI-DOOO9B REPLACE LEVEL GAUGES l

L16927 DG ILI-DOOO2B REPLACE LEVEL GAUGES f

i

ATTACHMENT A l

SAFETY-RELATED MAINTENANCE COMPLETED I

(NON-OUTAGE RELATED)

{

--.----- ----_----_----_-----------------------~~- UNIT =1 ---------------------------------------

f l

WRNUM SYSTEM EPN DESCRIPTION l

l L17068 NR RE-1B13-D193BG FULL CORE DISPLAY LPRM 24-25B D/S IS LIT L17115 AP 1AP57E REPLACE 131Y-1 C4 BREAKER POSITIG4 IND SLEEVE j

L17565 DC ER-1E22-R542 DIV 3 BATT GROUND RECORDER PEN NOT WORKING L17696 DG 1DGOBCA AIR COMP CIRCUIT. BREAKER TRIPPED MAGNETICS l

L17796 DC IDC09E CALIBRATE DIV II CHARGER AMMETER l

L18103 NR 1C51-000 APRM D 5 VOLT POWER SUPPLY PS-25 HAS 175 MV F-P i

L18125 NR JI-1C51-R901B B APRM SPIKED UPSCALE L18278 AP 1APG4E REPLACE CONTROL POWER TRANSPORMER SCREW i

L18298 AP 1AP41E REPLACE MCC 131B-4 COMPT B4 BREAKER HANDLE L18795 DG 1DG049A DG AIR COMP CHECK VALVE STUCK OPEN l

L18836 DG IDG000 AIR START MOTOR REGULATOR SET TO LOW L18892 AP 1AP60E REPLACE MCC 132Y-2 A-2 BREAKER HANDLE L18897 DG IE22-S001 SECURE K1 LOCKOUT RELAY

}

L19171 VG IVG02C SBGT COOLING FAN STARTED G7 RPS BUS TRANSPER l

L19318 RD IC11-D001-076 REPLACE CRD 10-15 L19717 DG IDG08CB REPLACE

• B' AIR COMPRESSOR BELTS L19721 DG IDG03J REPLACE 1A DG K3' RELAY-L19781 DC 1DC38E 24/48 SYS B BATTERY CHARGER INOPERABLE 9

t 5

i i

l i

l r

i i

l 6

i

?

l ATTACHMENT A j

SAFETY-RELATED MAINTENANCE COMPLETED l

(NON-OUTAGE RELATED) l l

---

UNIT =2 --------------------------------------- l I

WRNUM SYSTEM EPN DESCRIPTION i

LOS235 RD 2C11-D001-139 REPLACE HCU 58-31 STOP VALVE HANDWHEEL LO6052 DG 2DOO2M CLEAN DAY TANK VENT SCREENS LO6053 DG 2D0000 CLEAN DAY TANK VENT SCREENS l

LO6603 DG 2E22-5001 VERITY DG WIRING j

LO7005 RH 2RH000 REPLACE RHR WATERTIGHT DOOR LATCHING MECH LOB 632 DC 2DC000 PERFORM 24/48V BATTERY ICV'S LO9172 DC 2DC06E REPLACE 221Y CROSSTIE BREAKER AUX SWITCH LO9954 HG 2HG01A MONITOR HYD RECOMB MOTOR LINECURRENT L10014 HP 2E22-C003 TAKE HPCS WATER LEG PUMP CURRENT READINGS L10015 RI 2E51-C003 TAKE RCIC WATER LEG PUMP CURRENT READINGS L10016 VG 27G02C MONITOR SBGT COOLING FAN LINE CURRENT L10017 RH 2VYOIC MONITOR RHR SUPPLY LINE CURRENT L10018 HP 2VYO2C MONITOR HPCS SUPPLY LINE CURRENT I

L10020 LP 2VYO4C MONITOR LPCS SUPPLY LINE CURRENT

[

L10434 DG 2E22-S001 CALIBRATE DG METERS FOR EDSFI L10514 VY 2VYO4A REMOVE INSULATION FOR CHEMICAL CLEANING l

L10538 AP 2APABE MCC 231B-7 BREAKERS TRIPPED l

I L10644 DG 2E22-D300 DG COOLING WATER DISCHARGE INSUL DAMAGED i

l L10651 DG 2DGOBDA REPLACE DG AIR DRYER INDICATING LIGHT L10714 DG 2DG08CA REPLACE DG AIR COMP SHEET METAL SCREWS L10836 DG 2PG02JB INSTALL COVER FOR COMPONENT 86 t

L10840 DG 2DG02JA TIGHTEN LOOSE TERMINALS ON COMPONENT 59N L10852 RD TE-2C11-C001AA REPLACE PUMP MOTOR BEARING THERMOCOUPLE L11153 DG 2DG09DA AIR DRYER FAILED TO OPERATE l

L11539 NR 2C51-000 REPLACE LPRM POWER SUPPLY / CIRCUIT CARDS L11849 NR RY-2C51-K605GS 2E APRM RECIEVED HIGH HIGH NEUTRON TRIP L12032 NR 2C51-000 FABRICATE DETECTOR PIGTAILS l

L12089 DG 2J'I-DG024 CHECK VAR TRANSDUCER & METER MEASUREMENT

'i l

L12090 DG 2DO20M REMOVE RECIRCULATION ORIFICE l

L12259 NR RE-2C51-N003D TIP D READ LOW DURING TRANSVERSE TRACE L12431 RD TI-2C11-R902A REPLACE TEMPERATURE INDICATOR L12448 DC 2DC09E DIV I CHARGER OSCILLATING APPROX 2 AMPS l

L12462 DG 2DG08CA INSPECT AIR COMP SUCTION DISCHARGE VALVES L12463 DG 2DGOBCB INSPECT AIR COMP SUCTION DISCHARGE VALVES l

i L12472 RH 2E12-F411B RHR CONTROL SWITCH, INCORRECT INDICATION L12493 AP 2AP85E NO LIGHT INDICATIONWHEN BUS ENERGIZED f

L12679 AP 2AP64E REPLACE MCC 234X-1 RESET BUTTONS L12707 hD UR-2C11-R611 INSPECT SCRAM TIME RECORDER

[

LI2857 RH 2TS-VY001 REPLACE CUBICLE COOLER FAN TEMP SWITCHES I

L12863 D';

2AP07E REPLACE BREAKER PLUNGER' INTERLOCK CLIP f

L13098 NR 2NR000 LPRM 16-17 BINDING L13117 DG 2DG09DA INSPECT AIR DRYER OPERATION j

L13118 DC 2DG09DB INSPECT AIR DRYER OPERATION L13120 DC 2DG08DB INSPECT AIR DRYER OPERATION f

L13125 4G 2DGOBCA INSPECT AIR COMP SUCTION AND DISCHARGE VALVE i

l 1

[

4 I

j ATTACHMENT A 1

SAFETY-RELATED MAINTENANCE COMPLETED I

(NON-OUTAGE RELATED)

]

1

---

_------------------_-----_--_-------- -- UNIT =2 ---------------------------------------

WRNUM SYSTD4 EFN DESCRIPTION t

L13145 DG 2DG08DA INSPECT AIR DRYER DRAIN TRAP OPERATION

{

L13165 DC 2DC000 REPLACE CONDUIT SUPPORT BOLT L13171 DG 2DG01K REPLACE BROKEN HEAT SHRINK L13179 DG PS-2E22-N508 CORRECT FUEL PUMP PRESS SWITCH WIRING l

L13334 HP 2E22-P301B REPAIR CUT CABLE l

L13335 HG 2PA12J REPLACE HYD RECOMBINER PANEL LIGHT f

L13372 DG 2E22-F362A AIR DRYER CHECK VALVE FAILED

[

L13507 DG 2PDS-DG051 2A DG FUEL FILTER RESTRICTION ALARM UP L13720 DC ZIP 03E TSC UPS FAILED TO RETURN TO NORMAL l

L13787 NR 2NR000 REPLACE APRM'S K18 RELAYS l

L14298 AP 25A0]C REPLACE SWGR 241X TERMINAL BLOCK i

L14324 RD 2C11-D001-094 REPLACE CRD 38-59 VALVE HANDWHEEL l

L14437 DG TS-2E22-N901B CALIBRATE REFRIG AIR DRYER TEMP SWITCH L14518 DG 2DG08CB INSPECT AIR COMP SUCTION & DISCHARGE VALVE i

L14547 AP 2AP79E REPLACE 243-1 COMPT 2C LUG t

L14598 DG 2DG01F COOLING WATER STRAINER INOPERABLE i

L14610 VY 2TZ-VYO23A DAMPER ACTUATOR INOPERABLE F

LJ4611 VY 2TZ-VYO23B DAMPER ACTUATOR INOPERABLE L14615 NR RY-2C51-K605BX LPRM 16-17C READ BETWEEN 50%

L14622 AP 2APS6E REPAIR CUT CABLES L14881 UR 2NR000 D APRM SPIKED UP TO 11% & SLOWLY DRIFTED L15007 RI 2E51-C005 VACUUM PUMP WATER SPRAYED ON MOTOR L15008 RI 2E51-C005 VACUUM PUMP SEAL LEAKAGE L15059 AP 2AP20E PERFORM SWGR 235Y HFA RELAY CALIBRATION L15084 RI 2DC066 LOSS OF POWER TO 2E15-F019 L15097 RI 2E51-C002 RCIC TURBINE OVERSPEED ASSD4BLY INOPERABLE L15100 RI 2E51-F045 RCIC ISOLATION VALVE PACKING LEAK L15162 RD 2C11-000 SCRAM VALVE DELAY ACTUATING L15265 RD 2C11-D003-A CRD FILTER, STRIPPED DRAIN LINE UNION L15299 AP 2AP03E 237X/Y FEED BREAKER, NO CLOSED INDICATION L15301 PC 2FS-CM902 NO LOW FLOW SWITCH ALARM INITIATED L15318 NB RY-2C51-K605FF LPRM 48-25D DOWNSCALE O 50% POWER L15423 NR RY-2C51-K605GS 2E' APRM CANNOT CALIBRATE L15447 DG 2DG08CA AIR COMPRESSOR MOUNTING BOLT SHEARED OFF L15544 DG 2E22-F369B HPCS D/G RECEIVER LEAKS L15603 DC 2ER-DC020 125 VDC BUS PEN IS NOT WORKING L15746 DG 2DG049B INSPECT AIR COMPRESSOR CHECK VALVE L15853 RI 2E51-F331 STEAM LINE TRAP, LEAK AT FLANGE L15872 LC FT-2E32-N053A REPLACE FLOW TRANSMITTER TERMINAL STRIP I

L15914 VG 2VG003 INSPECT SBGT SUCTION DAMPER L15915 VG 2VG01C MONITOR LINE CURRENT ON SBGT FAN MOTOR L15997 MS TR-2B21-R816 TEMP RECORDER, CHART PAPER FAILED TO ADVANCE L16023 RD TE-2C11-N730 HCU 14-47 THERMOCOUPLE READS UPSCALE L16051 HP PSH-2E22-N014 REPLACE SUCTION PRESS INSTRUMENT STOP VALVE L16053 NR RY-2C51-K605GS SUPURIOUS ALARMS RECEIVED i

i l

l

{

^

ATTACHMENT A SAFETY-RELATED MAINTENANCE COMPLETED (NON-OUTAGE RELATED)

UNIT =2 ---------------------------------------

---

~~-------------------

WRNUM SYSTEM EPN DESCRIPTION J

I L16116 NR RY-2C51-J600 D TIP DRAWER BOTTOM READING L16134 DG 2DG000 INSPECT DG HACR-1 TEST SWITCH l

L16198 RD XY-2C11-K919F CRD 10-19 TRIPPED RMCS L16200 DC 2PM01J 250V BATTERY DISCHARGE HIGH L16201 RI 2E51-F086 236Y-2 BREAKER TRIPPED ON THERMALS L16206 RI 2E51-F086 LUBRICATE ISOLATION VALVE STEM L16209 AP 2AP000 SWGR 231B AUTO TRIPPED ON NEUTRAL OVERCURRENT f

L16341 RH PIS-2E12-N022C DISCHARGE HEADER INDICATOR, BENT NEEDLE L16396 AP 2AP000 SECURE PRESSURE REGULATOR IN PANEL l

L16438 DG 2DG08CA INSPECT AIR COMP SUCTION & DISCHARGE VALVE L16439 DG 2DGOBCB INSPECT AIR COQ SUCTION & DISCHARGE VALVE l

L16449 DG 2DGOBCA AIR COMP c'Ld5 188 OF OIL PRESSURE j

L16500 RD 2C11-C001B CLEAN CRD PUMP SCREENS L16577 RD eui-C001A REPL7.CE CRD PUMP MOTOR VENT SCREEN L16639 RH 2E12-F024A FULL FLOW TEST VALVE TRIPPED ON MAGNETICS l

L16650 RD 2C11-D001-052 ACCUM 10-27 NITROGEN LEAKAGE L16677 RI 2E51-F086 KCIC VACUUM BREAKER VALVE PACKING LEAKAGE L16678 RH 2E12-F024A REPLACE ANTI ROTATION LOCKING DEVICE L16766 RI 2E51-C002 TRIP & THROTTLE FAILED TO RESET L16876 NR RE-2C51-N003A TIP A DAMAGED DURING TIP SET L16924 DG 2D0009B REPLACE LEVEL GAUGES L16925 DG 2LI-DOOO2B REPLACE LEVEL GAUGES L16946 RI 2E51-F086 LUBRICATE VALVE YOKE i

L17093 RI 2E51-F008 BREAKER TRIPPED ON MAGNETICS L17262 RI 2E51-F059 FULL FLOW TEST STROKE TIMES L17271 RD 2C11-C001B HIGH CRD PUMP VIBRATION IN GEAR BOX L17291 RD ZS-2C11-F001A FLOW CONTROL VALVE DOES NOT SHOW FULL OPEN L17334 NB RE-2B13-D193BV LPRM 16-17A SPIKING AT 100%

L17335 NR RY-2C51-K601H UNABLE TO ADJUST L17340 NR RY-2C51-K605GM RECONNECT RECORDER TO APRM OUTPUTS L17360 RH PIS-2E12-N022B REPLACE RHR PUMP PRESSURE COVER PLATE f

L17387 DG 2DG08CB INSPECT AIR COMPRESSOR L17471 MS MS01-2877S REPLACE MECH SNUBBER PIPE CLAMP BOLT L17632 RI 2E51-C005 REPLACE BAROMETRIC CONDENSER PIGTAIL L17668 HR RY-2C51-K601E IRM E DID NOT RESPOND WHEN PULLING RODS I

L17717 NB LT-2B21-N403D CALIBRATE REACTOR VESSEL LEVEL TRANSMITTER L17901 RP 2C71-S003E REPLACE RPS LOGIC CARD l

L18123 DG 2DG08CB STARTING AIR COMPRESSOR REMAINS RUNNING L18188 RD 2RM000 RMCS TRIPPED AND WILL NOT RESET L18238 RH TRS-2E12-R601 SPURIOUS RHR TEMPERATURE RECORDER ALARM L18253 RD XY-2C11-K921 RMCS TRIPS ON ROW 30-11 L18379 LP FS-2E21-N004 SWITCH FOUND OUTSIDE OF REJECT LIMIT L18697 DC 2DC000 DIV II GROUND DETECTOR HAS 460 VDC GROUND L18731 NB LR-2B21-R884A RX LEVEL / PRESS RECORDER TAKE UP SPOOL L18864 RD 2C11-000 RMCS TRIPPED ON ROD 30-11

)

L18907 NB PS-2B21-N413B RPV LO PRESS & INJ LINE B/C LPCI f

r 4 --..

.. -.~ -.

i e

ATTACHMENT A SAFETY-RELATED MAINTENANCE COMFLETED (NON-OUTAGE RELATED) i

___..____________________________________.-------- UNIT =2 --------------------------------------- l j

WRNUM SYSTEM EPN DESCRIPTION L18982 DG 2DG01K SPACE HEATER NOT OPERATING L19140 RD 2C11-C001B CRD PUMP OIL COOLER LEAKAGE

{

L19157 DG 2DG000 REPLACE DG SMALL BORE TUBING l'

L19178 RH 2E12-F024A FULL FLOW TEST VALVE MOTOR TRIPPED L19246 RD 2C11-D001-066 HCU 18-19 CHARGING NIPPLE LEAKS L95782 PC 2PL75J REPLACE CAM PART ASSEMBLY T-HANDLE L98683 AP 2AP07E REPLACE BUS 243 CUB 5 SBM SWITCH L98833 RH 2E12-F362C REPLACE RHR PUMP DISCH VENT HANDWHEEL

{

i a

e 5

i i

?

i l

i

i ATTACEMEET B II.B UNIT SHUTDOWNS (UNIT 1)

DATE: 920101__

GENERATOR OFF_LINE: 94.25 OUTAGE TYPE:

Forced (LIF15) 1 (YYMMDD)

(Hours)

REASON: Automatic reactor scram due to loss of main condenser vacuum.

Condenser vacuum was lost when the steam seal evaporator failed resulting in a loss of steam to the main turbine seals.

1 i

CRITICAL ACTIVITY PATH:

Troubleshoot, repair and test the steam seal evaporator.

Correct limit switch problems on the main turbine valves.

Repair leaks and correct valve indication problems on the heater drain system.

Replace four fire seals in steam sensitive areas.

Repair radiation monitors on the

'B' control room HVAC system.

CORRECTIVE ACTIONS (DVR/LERf if applicable):

DVR5 1_1_92-0014 LER# 92-003_00 RADIOACTIVITY RELEASE / EXPOSURE OVER 10% ALLOWABLE VALUES: None.

I SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED:

WR_ HUM SXS EPN DE6CRIPTION LO8982 RP 1C71_N006F Replace the #3 Turbine Stop Valve Limit switch i

L12797 MS 1B21_CV4 Control problem at EHC panel L13591 RD IC11-K946BK Control Rod 30_11 full out indication L13646 VC OREY-VC081XB Radiation Monitor relay failed to energize L13872 RD JY-1C22-K601A Division 1 Alternate Rod Insertion trouble alarm i

DATE:_211003 GENERATOR OFF-LINE: 2160.5 OUTAGE TYPE; Scheduled (L1R05) l (YYMMDD)

(Hours) i REASON: Refueling Outage.

CRITICAL ACTIVITY PATH: See Appendix A.

I CORRECTIVE ACTIONS (DVR/LER$ if applicable): None.

RADIOACTIVITY RELEASE / EXPOSURE OVER 10% ALLOWABLE VALUES: None.

]

SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED: See Appendix B.

v

>-T's u

me ww sp ww

-e=="

i f

APPENDIX A CRITICAL PATH ACTIVITY REFUEL OUTAGE (L1R05)

DESCRIPTION MAIN TURBINE OFF LINE 4

INSPECT 1B21-RSDV2 VALVE INSPECT IPS-FWOB8 FOR OIL LEAK REMOVE 1PCM111 PERSONNEL ACCESS HATCH STRONGBACKS l

INSTALL IPCM111 PERSONNEL ACCESS HATCH STRONGBACKS COOL REACTOR TO COLD SHUTDOWN BEGIN UNIT AUXILIARY TRANSFORMER MAINTENANCE COMPLETE REACTOR VESSEL DISASSEMBLY BEGIN CIRCULATING MATER WINDOW MAINTENANCE BEGIN TURBINE GENERATOR WINDOW MAINTENANCE BEGIN DIESEL GENERATOR B WINDOW MAINTENANCE MRI BELLOWS INSULATION MODIFICATION WALKDOWN-(P01-1-91-528)

MRI BELLOWS INSULATION INSTALLATION-(P01-1-91-528)

BEGIN DC BATTERIES (DIV III) MAINTENACE BEGIN HIGH PRESSURE CORE SPRAY MAINTENANCE l

UNLOAD REACTOR CORE BEGIN SAFETY / RELIEF VALVES (SRV/ ADS) MAINTENANCE BEGIN RESIDUAL HEAT REMOVAL B MAINTENANCE j

BEGIN RX CORE ISOLATION COOLING MAINTENANCE I

INSPECT 1HD059A HEATER DRAIN VALVE INSPECT 1HD056A HEATER DRAIN VALVE BEGIN CONTROL ROD DRIVE MAINTENANCE BEGIN REACTOR WATER CLEAN-UP MAINTENANCE BEGIN DIESEL GENERATOR A MAINTENANCE SIGNATURE TRACE, SET LIMIT SWITCHES ON VALVE 1WR180 BEGIN DIESEL GENERATOR 0 MAINTENANCE BEGIN LOW PRESSURE CORE SPRAY MAINTENANCE REPAIR 1B21-1BSFV1 MSR STEAM FEED VALVE INDICATION REPLACE 1CD01PB CD/CB PUMP COUPLING COMPLETE CORE RELOAD INSPECT AND TEST INTEGRATED LEAK RATE TEST INSTRUMENT CABLES INSPECT IVP01CB VP CHILLER HYDROTEST 1HD059A HEATER DRAIN VALVE PERFORM SIGNATURE TRACE, SET LIMIT SWITCHES ON IVP053B VALVE PERFORM VOTES TEST ON IVP053B VALVE REASSEMBLE 1HD059A HEATER DRAIN VALVE COMPLETE CONTROL ROD DRIVE TIMING & FRICTION TESTS COMPLETE REACTOR REASSEMBLY MODIFICATION TEST, CLEAN CONDENSATE CHECK VALVES-(P01-1-90-026)

REINSTALL IHD056A HEATER DRAIN VALVE INSULATION REINSTALL 1HD059A HEATER DRAIN VALVE INSULATION

APPDTDIX B REFUELING OUTAGE (L1R05)

SAFETY RELATED CORRECTIVE MAINTENANCE i

WRNUM SYSTEM EPN DESCRIPTION i

L19453 HP 1E22-F012 INCREASE HPCS PUMP FLOW BYP VALVE THRUST I

L19461 RD IC11-D001-002 REPAIR CRD 22-59 FULL IN/OUT LIGHT SOCKET l

L19481 RH IE12-F024A FULL FLOW TEST STOP VALVE FAILED TO CLOSE f

L19509 RI 1E51-F063 REPLACE SEAL ON LIMITORQUE OPERATOR LO4716 RH 1E12-F09BC RHR MANUAL DISCHARGE VALVE LEAKING LO6124 LC IE32-F001E REBALANCE ISOLATION VALVE TORQUE SWITCH l

LO6940 DC 1DCO2E 250V DC BUS UNABLE TO CLOSE BREAKER WITH DOOR LO715B NR RY-1C51-K601D 1D SRM SHORT PERIOD ALARM SPORADIC LO7282 DC IDC01E 250 VDC CELL li61 ICV BELOW LIMIT LO8824 NR RY-1C51-K601D IRM (D) READING HIGH LO9602 RP 1RP000 1C71-K7A MSL HI RAD SCRAM & ISOL RELAY NOISY LO9846 MS IB21-F502B SJAE PCV UPSTREAM STOP VALVE MECH POS IND L10656 DG IDG01K REPLACE ENGINE TERMINAL BOX CABLE LUG L10658 DG IDG01K IDG02JB PANEL, LUGS BENT IN EXCESS OF 45 DEG L10659 DG IDG01K 1DG05J PANEL, LUGS BENT IN EXCESS OF 45 DEG L10660 DG 1DG01K 1DG03J PANEL, LUGS BENT IN EXCESS OF 45 DEG L10661 DG 1DG01K 1DG02JA PANEL, LUGS BENT IN EXCESS OF 45 DEG L10666 DG 1DG01K 1DG03J PANEL, LOL7E TERMINAL BOARD SECTION L10700 DG IDG01K OIL LEAK AROUND AIR BOX DRAIN LINE FITTING L10759 DG 1E22-P301B PANEL RELUG WIRE AT TERMINAL 106 OF TB11 L10824 DG 1DG04T DG DAYTANK MANWAY FIANGE L10825 DG 1E22-S001 FUEL OIL FLANGE LEAK L11177 RD 1C11-F109A CRD PUMP VENT LEAKING L11800 RD XY-1C11-K946BH CR 46-19 NO POSITION INDICATED WHEN AT (B L11896 DG ODG000 INSPECT AIR COMP SUCTION AND DISCHARGE VALVES L11897 DG ODG000 INSPECT AIR COMP SUCTION AND DISCHARGE VALVES L11908 NR RY-1C51-K605EZ REPAIR LPRM DETECTOR 32-25B L13046 MS PS-1B21-N015B MSL LOW PRESSURE STOP VALVE LEAKS BY L13740 NR RY-1C51-K605CK LPRM 24-09C CAUSE UPSCALE ALARMS L13959 RD XY-1C11-K939AX HCU 26-39 DOES NOT HAVE FULL OUT INDICATION L14571 RH 1E12-F094 VALVE LEAKS BY SEAT L15187 RD 1RD000 REPLACE SCRAM BOXES COVER CLIPS L15254 RH 1E12-F068A RHR HK OUTLET VALVE PACKING LEAK L15348 RH 1AP82E REPAIR IVYO6C VENT FAN CONDUCTOR JACKETS L15387 LP 1E21-F012 FULL FLOW TEST BYPASS VALVE PACKING LEAK L15767 RH 1E12-F068B RHR HX OUTLET VALVE PACKING LEAK L15826 RH 1E12-F333C RER PUMP VALVE LEAKING L15827 RH IE12-F334C RER PUMP VALVE LEAKING L15961 LC PT-1E32-N061A MSL PRESS INSTRUMENT STOP VALVE LEAKS L15962 LC PT-1E32-N060 REPLACE RX PRESSURE TRANSMITTER TUBE FITTING L15963 LC PT-1E32-N050 REPLACE RX PRESSURE PIPE THREAD L16199 RD PDI-IC11-N015 CRD SUCTION FILTER DP INDICATOR READS LOW L16320 ED 1C11-D001-078 HCU 30-11 INSTRUMENT AIR FITTING LEAK i

L16345 DG ODG01P REPAIR ODG01P COOLING WATER PUMP CABLE L17019 DG PDI-1E22-N502 1B DG COOLING WATER PUMP DP GAUGE PEGGED LOW L17328 RD 1C11-D001-002 ACCUM 22-59 VALVE PACKING LEAK L17870 RP IC71-S003E RECALIBRATE L17877 RD PDI-1C11-N002 CED DRIVE FILTER DP ALARM UP j

L17924 RD IC11-F002B CRD FLON CONTROL VALVE PACKING LEAK j

[

APPENDIX B REFUELING OUTAGE (L1R05)

SAFETY RELATED CORRECTIVE MAINTENANCE WRNUM SYSTEM EPN DESCRIPTION L17998 RD 1C11-F385 CRD DISCHARGE STOP VALVE PACKING LEAK L18037 RH 1E12-AK070A REPLACE LOOSE BANANA JACK PLUG L18121 NR RY-1C51-K600BA SRM B SETPOINTS ERRATIC L18148 NR RY-1C51-K602A IRM FAILED DOWNSCALE L18181 MS 1B21-2BSVV-2 BLANKET STEAM VENT VALVE LEAK L18193 RH IE12-F040B RHR HX BLOWDOWN VALVE, NO CLOSED INDICATION L18196 RH ZS-1E12-F041B NO VALVE POSITION INDICATION IN CONTROL ROOM L18198 RH 1E12-F412B RHR SAMPLE VALVE FAILED TO CYCLE L18206 RH ZS-1E12-F050A RER RETURN CHECK VALVE, INDICAATION INOPERABLE L18219 NR RY-1C51-K601C C IRM SPIKES L18222 NR RY-1C51-K601H H IRM SPIKES L18241 MS 1B21-F029B AIR ACCUM CHECK VALVE LEAK L18285 NB TE-1B21-H050A1 DRYWELLHEAD TEST PENETRATION LEAK L18288 MS 1B21-F022B MSIV B AIR LEAKING OUT OF EXHAUST PORT L18309 RD 1C11-000 CRD 10-23 SCRAM LITE INOPERABLE L18329 MS IB21-F024D MSIV ACCUMULATOR CHECK VALVE OUT OF TOLERANCE L18335 RH TRS-1E12-R601 RER HX DISCHARGE COOLING WATER TDiP HIGH L18375 RH 1E12-F090B RHR MANUAL RETURN VALVE, DUAL IND WHEN CLOSED L18434 RI 1E51-F086 1AP83E-E3 MCC 1364-2 RESET BUTTON INOP L18521 RH 1E12-F359A RER SDC EXCESS FLOW CHECK VALVE FAILED L18522 RH 1E12-F360B RER SDC EXCESS FLOW CHECK VALVE FAILED L18523 RH IE12-F359B RER SDC EXCESS FLOW CHECK VALVE FAILED L18526 RH IE12-F360A RER SDC EXCESS FLOW CHECK VALVE FAILED L18552 DG ODOO1P OILER LEAKS SLIGHTLY L18553 RD 1C11-000 REPLACE HCU 46-43 HOSE FITTING 0 107 VALVE L18559 RH LT-1B21-N403A INSPECT RX VESSEL LOW WATER LEVEL 3 LOGIC L18575 AP 1AP07E SWGR 143 CUB 6 A PHASE ARC CHUTE PROT COATING L18576 AP 1AP07E REPAIR SWGR 143 CUB 3 SPACE HEATER BOLT L18635 RD HS-1C11-BS003A REPLACE A CRD PUMP CONTROL SWITCH L18645 RH RH23-1001V REALIGN VARIABLE SUPPORT FOR 1E12-F055B L18667 LC FE-1E32-N006A BLEED OFF FLOW TRANSMITTER LOOP FAILED f

L18676 NB 1B13-D003 INSPECT SHROUD ACCESS HOLE COVER l

L18704 AP 1AP80E 13tX-Z BUS BAR 1AP80E LOCK WASHER BROKE L18706 MS PDS-1E31-N008A MSL HIGH FLOW SWITCH EXCESS ELECTRICAL NOISE L18707 RD PS-1C11-N001B CRD PUMP SUCTION PRESS SWITCH RESET ABOVE NORM l

L18710 RD PI-1C11-R008 CRD PUMP DISCHARGL PRESSURE GAUGE STICKS L18711 RD TI-1C11-R902A CRD PUMP LUBE OIL TEMP INDICATOR UNABLE TO CAL L18736 RD 1C11-D046-112 SDV INLET ISOL VALVE HCU 02-31 LEAKS L18743 HG IHG002A RECORDER SUCTION VALVE MOTOR CURRENTS HIGH L18801 DG 1E22-S001 1B DG RESISTOR R1 IS CRACKED L18839 NR RY-1C51-K600C UNABLE TO CALIBRATE 1C SRM DRAWER L18841 HP INB25A CORRECT REFERENCE LEG PIPING SLOPE L18842 HP IB21-D004C REATTACH CONDENSING POT HANGER ASSEMBLY L18849 NR RY-1C51-K605ET LPRM 56-33 MACHINE DOWN FOR INSTALLATION L18850 NR RY-1C51-K605T LPRM 48-49 MACHINE DOWN FOR INSTALLATION L18853 AP 1AP03E ACB 1411 UNABLE TO CLOSE FROM CONTROL ROOM L18948 DG ODG09MB AIR COMPRESSOR FLEXIBLE OUTLET PIPING LEAK l

L19019 DG ODG01K DAMAGED LOWER MAIN BEARINGS #2 AND #10 L18869 DG 1PI-DG086 AIR PRESS INDICATION SENSING LINE BROKEN

APPENDIX B REFUELING OUTAGE (L1R05)

SAFETY RELATED CORRECTIVE MAINTENANCE WRNUM SYSTEM EPN DESCRIPTION L18876 NR RY-1C51-K605GR COMP PT B680 FOUND OUT OF TOLERANCE L18896 NR RY-1C51 -K605ET MACHINE DOWN LPRM FOR INSTALLATION L18905 HP 1E22-C001 BREAKER PIN BENT, JAMS UP ON GUIDE ARM L18909 DG PI-1E22-R546 REPLACE IB DG CRANKCASE PRESS GAUGE SNUBBER L18910 RD 1C11-F002B CRD FLOW CONTROL VALVE DUAL INDICATION L18915 HP 1E22-F038 HPCS MANUAL INJECTION VALVE INDICATION BAD L18928 DC ITCU-VD013 REPAIR DIV III SWGR ROOM TEMP CONTROLLER L18942 DG IE22-P301B DG PANEL LOCAL ALARM HORN FAILED L18959 RH IE12-F050B REPAIR TESTABLE CHECK VALVE CONDUIT L18983 DG IE22-S001 INSPECT DG SPACE HEATER L18991 LP 1E21-F331 SUCTION DRAIN VALVE UNABLE TO CLOSE L18995 RD PS-1C11-N077 REPAIR CRD HCU 02-23 ELECTRICAL CABLE L19003 RH IE12-D300A RHR STRAINER REPAIR LEAK L19015 LP 1TI-DG032 REPLACE PUMP TEMPERATURE INDICATOR GLASS L19033 DG 1E22-S001 DG OVERSPEED TRIP RELAY FAILED L19037 RD 1C11-D001-019 HCU 26-43 AIR SUPPLY LINE LEAK L19039 RD 1C11-D001-144 HCU 38-31 ACCUMULATOR FITTING LEAK L19043 RD IC11-D001-177 HCU 42-11 NITROGEN LEAKS L19056 RP 1RP000 1A RPS MG SET IND & RELAYS VOLTAGE CYCLING L19071 RD IC11-D001-009 HCU 18-51 ACCUMUALTOR CARTRIDGE VALVE LEAKS L19076 DC 1DC07E DIV 1 125V BATTERY CORROSION ON CELLS 50 & 51 L19077 RI 1H13-P601 REPAIR TERMINAL BLOCK DIVIDER L19097 RD 1C11-D001-111 HCU VALVE LEAKS L19098 RD 1C11-D001-111 HCU 26-07 VALVE FAILED TO CLOSE L19102 RD KY-1C71-AK016C SCRAM RESET INTERLOCK RELAY BUZZING L19106 RD IC11-D001-072 HCU ACCUM 26-15 VALVE STEM BROKEN L19107 RD 1C11-D001-015 HCU 18-47 VALVE LEAKS L19116 DC IDC11ECB17 DIV I 125VDC +75V GROUND ON CIRCUIT BOARD L19133 RH 1TI-VY019 RECALIBRATE RHR DUCT TEMPERATURE INDICATOR L19145 AP 1APO4E INSPECT SWGR 141Y CUB 1 MAIN FEED BREAKER L19146 RH 1E12-F073B REPAIR RHR HK VENT VALVE CONTROL ROOM IND L19148 MS 1B21-F022A INBOARD MSIV WOULD NOT SLOW CLOSE L19149 MS IB21-F022B INBOARD MSIV WOULD NOT SLOW CLOSE L19170 MS ITG01TB LEVEL TRANSMITTER DRAIN PIPE PLUG STUCK L19174 LP 1DG036 REMOVE MOTOR COOLER DISCH CHECK VALVE SPRING L19212 RD 1C11-000 HCU 34-19 REPLACE FULL-IN/-OUT LIGHT SOCKET L19220 AP 1AP03E ACB 1412 SAT FEED TO 141Y FAILED TO OPEN L19245 NR RY-1C51-K601D REPLACE D IRM POWER SUPPLY l

L19255 RH IVYO1C A RHR ROOM COOLER BREAKER FOUND TRIPPED I

L19256 RH 1H13-P629 INSPECT PANEL EXTERNAL WIRING L19266 RD JY-1C22-K601A POWER SUPPLY FAILED L19267 MS 1MSO4BL INSPECT SRV DOWNCOMER SUPPORT MSO4-1027C L19271 RH 1E12-D300A DIV 1 RHR STAINER LEAK ON FLANGE L19279 RD LS-1C11-N372 HCU 38-15 ACCUM LEVEL SWITCH CONTACTS INOPERABLE l

L19282 RD 1C11-D001-140 REPLACE HCU 54-31 FLEXIBLE CONDUIT L19283 RD IC11-D001-160 REPLACE HCU 34-23 FLEXIBLE CONDUIT L19287 RD IH13-P603 REPLACE RMCS WITHDRAW BUTTON L19303 NR INR000 TIP TUBING A8 UNABLE TO VERIFY LOCATION L19413 RH ZS-1E12-F090B-C REPLACE LIGHT INDICATION CONDUIT l

l ATTACHMENT B II.B UNIT SHUTDOWNS (UNIT 2) l DATE._920104 GENERATOR OFF-LINE: 2392.2 OUTAGE TYPE:

Scheduled (L2R04) 1 I

(YYMMDD)

(Hours)

REASON: Refueling Outage.

I CRITICAL ACTIVITY PATH: See Appendix C.

CORRECTIVE ACTIONS (DVR/LER# if applicable): None.

RADIOACTIVITY RELEASE / EXPOSURE OVER 10% ALLOWABLE VALUES: None.

j l

SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED: See Appendix D.

(

DATE:_220412 GENERATOR OFF-LINE:

7.67 OUTAGE TYPE:

Forced (L2F15)

(YYMMDD)

(Hours) l l

REASON: Hanual Turbine trip due to Main Turbine bearing high vibrations.

The vibrations were due to ' rubs' on the newly installed low pressure turbine rotors. The reactor remained critical during the outage.

CRITICAL ACTIVITY PATH: None.

CORRECTIVE ACTIONS (DVR/LER$ if applicable): None.

RADIDACTIVITT RELEASE / EXPOSURE OVER 10% ALLOWABLE VALUES: None.

SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED: None.

DATE: 920413 GENERATOR OFF-LINE:

6.92 OUTAGE TYPE:

Forced (L2F16)

(YYMMDD)

(Hours)

REASON: Manual Turbine trip due to Main Turbine bearing high vibrations.

The vibrations were due to ' rubs' on the newly installed low pressure turbine rotors. The reactor remained critical during the outage.

CPITICAL ACTIVITY PATH: None.

CORRECTIVE ACTIONS (DVR/LER# if applicable): None.

RADIOACTIVITY RELEASE / EXPOSURE OVER 10% ALLOWABLE VALUES: None.

SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED: None.

APPENDIX C CRITICAL PATH ACTIVITY REFUEL OUTAGE (L2R04)

DESCRIPTION MAIN TURBINE GENERATOR OFF LINE COMPLETE REACTOR DISASSEMBLY REPLACE MAIN TURBINE ROTORS-(CR 90-157)

MAIN STEAM SNUBBER REDUCTION MODIFICATION-(M01-2-90-008)

BEGIN CIRCULATING WATER MAINTENANCE ACTIVITIES BEGIN DIESEL GENERATOR B MAINTENANCE ACTIVITIES BEGIN HIGH PRESSURE COP.E SPRAY MAINTENANCE AND TESTING BEGIN DC BATTERIES (DIV III) MAINTENANCE ACTIVITIES 2FWO1KA TDRFP UN/RECOUPLE TURBINE FOR OVERSPEED TEST O/G RECOMB TEMPERTURE RECORDER-(M01-2-87-023)

COMPLETE REACTOR CORE UNLOAD BEGIN LOW PRESSURE CORE SPRAY MAINTENANCE AND TESTING PLACE SYSTEM AUXILIARY TRANSFCRMER OUT OF SERVICE PERFORM CONTROL ROD DRIVE HYDRAULIC CONTROL UNIT MAINTENANCE BEGIN RESIDUAL HEAT REMOVAL A MAINTENANCE BEGIN 24/48 BATTERY MAINTENANCE ACTIVITIES REPLACE LOW POWER RANGE MONITORS BEGIN REACTOR CORE ISOLATION COOLING MAINTENANCE REMOVE JET PP #15 TEMP CLAMP-(P01-2-90-013)

BEGIN REACTOR WATER CLEAN UP VALVEMAINTENANCE BEGIN DC BATTERIES (DIV II) MAINTENANCE ACTIVITIES INSPECT 2FWO1KB TDRFP LP BEARING RCIC TURBINE TIME DELAY TRIP BYPASS-(M01-2-89-008)

BEGIN RESIDUAL HEAT REMOVAL B MAINTENANCE PERFORM 2A DIESEL GENERATOR /DIV II TESTING REPLACE 26B HP HEATER FIRE SEAL REPAIR 2HD171A HD REJECT TO 2A COND ACTUATOR COMPLETE CORE RELOAD REPLACE 2CBOIPA PUMP COUPLING REPAIR H2 COOLING SUPPORT WS14-2036R REPAIR H2 CCOLING SUPPORT WS13-2019R REPAIR 2HD026D NORM DRN TO 25A ACTUATOR REPLACE FIRE RATED PENETRATION SEAL MK-2TB-138 2B21-F514B DETERM AND RETERM FOR MMD COMPLETE REACTOR REASSEMBLY REPLACE VP SYSTEM DUCTWORK REINSULATE SRVS RESET 2FWOO3 ISOLATION VALVE LIMIT SWITCH REPAIR 2B33-F023A PUMP SUCTION VALVE INDICATION INSPECT 2FWOIKA THRUST BRG PERFORM SIGANTURE TRACE ON VALVE 2CD035F REPLACE FIRE RATED PENETRATION SEAL MK-2TB-087 GROUT TURBINE PENETRATION INTO LOW PRESS HEATER BAY REPAIR VALVE 2B21-F069 LEAKAGE UNPIN CD SYSTEM PIPING INSPECT VALVE 2FWOO3

=_

i i

4

~

APPENDIX C i

CRITICAL PATH ACTIVITY j

REFUEL OUTAGE (L2R04) 1 I

l-

____-____________________________________________________________.______.--____~______________-_

J DESCRIPTION 1

REBUILD VALVE 2B21-F065A j;

REPAIR 2E12-F059B DISCHARGE STOP VALVE FEEDWATER PUMP PRESSURE SWITCH TESTING-(M01-2-89-009)

INSPECT 2LS-RF003 D/W FLOOR DRAIN PUMP LEVEL SWITCH INSPECT 2RE002 DRYWELL EQUIPMENT DRAIN PUMP j

INSPECT 2FP012 AUTO DELUGE VALVE INSPECT MAIN POWER TRANSFORMER DELUGE I

PERFORM VESSEL HYDRO REPAIR 2B21-F013N N SRV LEAKAGE REMOVE SCAFFOLDING FOR VALVE 2E32-F008 REPAIR 2B33-F338A INBOARD STOP VALVE INDICATION REPAIR 2HD-1SRDCHA NORM DRAIN VALVE REPLACE 2DG03J 2A FUEL FILTER TIGHTEN 2VP111B VALVE CAP TIGHTEN 2VP102B VALVE CAP PERFORM INTEGRATED LEAK RATE TEST REMOVE HYDROGEN COOLER INSULATION i

REPAIR FIRE BARRIER SEAL REPAIR 2FWO1KA FEED PUMP TURB OIL LEAK REMOVE 59A/59B VALVE INSULATION INSPECT GENERATOR BEARING PUMP #10 PERFORM SIGNATURE TRACE ON VALVE 2B21-F514B INSPECT AND TEST 2PL18J VOLT SWITCH

[

INSTALL AND REMOVE SA SPOOL PIECE AFTER INTEGRATED LEAK RATE TEST REPAIR HEATER BAY AIR LEAKS RETERM 2B21F514B AFTER MAINTENANCECOMPLETE COMPLETE FINAL REVIEW FOR VALVE 2B21-F514B REPAIR 2B33-F079A LEAKOFF CONTROL VALVE RCIC WATER LEG PUMP MOD TEST-(M01-2-87-082)

FW MINIMUM FLOW VALVE MOD TESTING-(M01-2-88-033)

TROUBLESHOOT AND REPAIR TIPS

[

T.S.I. CABLE UPGRADE.(M01-2-69-002 l

3 INSTALL 2HD-01PD PUMP MOTOR REPAIR RY_2C51-K605CB LPRM LOCAL DOWNSCALE LIGHT REPAIR 2E12-F041C TESTABLE CHECK VALVE INDICATION AMP CHECK VALVE 2B21-F422A CLEAR OUT OF SERVICE FOR 2C51-J011 CONNECTION PERFORM ROTATION CHECK ON 2B HD PUMP

[

INSPECT / TEST 2GS-SSAFV VALVE

[

SEND SRV 2B21-F013R TO WYLE LABS, TEST AND REFURBISH INSPECT 2CB018B CB MINIMUM FLOW VALVE INSPECT 2B21-F073 VALVE THERMAL OVERLOAD RCIC TURBINE TIME DELAY TRIP BYPASS TEST (M01-2-89-008)

TURBINE LUBE OIL CONTROL VALVE TEST-(CR 90-017) i REPAIR 2B21-F001 REACTOR HEAD VENT VALVE i

REPLACE 2G33-F001 VALVE MOTOR i

2G33-F001 VALVE, SET LIMITS & SIGNATURE TRACE e

r r,

c,,,,

,. + - -.

--we w

.-r~

~

-=

APPENDIX C CRITICAL PATH ACTIVITY REFUEL OUTAGE (L2R04)

DESCRIPTION PERFORM VOTES TESTING ON THE 2G33-F001 VALVE RESET LIMITS AND SIGNATURE TRACE 2G33-F001 REBUILD 2G33-F001 VALVE ACTUATOR FINAL IN SERVICE TESTING REVIEW, 2G33-F001 2G33-F004 SET LIMITS & SIGNATURE TRACE PERFORM 2G33-F004 VOTES TEST REBUILD 2G33-F004 VALVE ACTUATOR FINAL REVIEW OF 2G33-F004 INSPECT 2CYO3P CY JOCKEY PUMP FOR VIBRATION REPLACE 2PAI6J STARTREC PANEL VALIDYNE CARD REPAIR H2/CO2 SELECTOR VALVE 2TG02KB LPB TURBINE NEW MONOBLOCK ROTOR 2TG02KC LPC TURBINE NEW MONOBLOCK ROTOR 2TG02KA LPA IURBINE NEW MONOBLOCK ROTOR REPAIR VALVE 2B33;r015B 2FWOIKA REMOVE / REINSTALL INSTRUMENTATION FOR PUMP REBUILD 2FWO1KB REMOVE / REINSTALL INSTRUMENTATION FOR PUMP REBUILD INSPECT 2A MSR IST STAGE DRAIN VALVE TO DRAIN TANK REPAIR 2HD-MSDCV-HB REPAIR 2B21-MOVSV-5 DRAIN VALVE RECALIBRATE 2PSL-GS031 STEAM PACKING EXHAUST SWITCH INSTALL 2PS-TOO31 TURBINE LUBE OIL MANOMETER REPLACE HD10-2406X RIGID STEUT END PADDLES TORQUE 2A TDRFP BOLTS REPLACE 2G33-F001 VALVE THERMAL OVERLOAD REPLACE 2G33-F004 VALVE MOTOR 2G33-F004 REPLACE THERMAL OVERLOAD ADJUST 2B21-F067D MSL DRAIN VALVE CLOSE LIMIT SWITCHES ADJUST 2B21-F067A MSL DRAIN VALVE CLOSE LIMIT SWITCHES ADJUST 2B21-F067B MSL DRAIN VALVE CLOSE LIMIT SWITCHES ADJUST 2B21-F067C MSL DRAIN VALVE CLOSE LIMIT SWITCHES INSTALL MAIN TURBINE BEARING VIERATION DETECTOR-(M01-2-89-002 )

REPLACE 2B EXTRACTION STEAM SUPPLY LINE DRAIN VALVE-(CR 90-093)

REPLACE 2B TDRFP LOW PRESS LINE DRAIN VALVE-(CR 90-086)

REPLACE 2G33-F001 RWCU SUCT LINE ISOL VALVE-(CR 90-057)

CLEAR 005 FOR 2G33-F001 VALVE PERFORM PLANT HEA7UP RECOUPLE RCIC TURBINE TO PUMP AFTER OVERSPEED TEST TAKE 2E51-C002 OUT OF SERVICE LST 92-073 RI FULL FLOW TO S/P @ 150# MOD TEST-(M01-2-87-082)

RCIC & LPCS SIMULTANEOUS RUN @ 150# - MOD TEST-(H01-2-87-082)

RCIC FULL FLOW TO SUP POOL @ 1000# MOD TEST-(M01-2-87-082) l i

1 5

t APPENDIX C l

l CRITICAL PATH ACTIVITY REFUEL OUTAGE (L2R04) 7 I

i i

DESCRIPTION l

SYNCHRONIZE GENERATOR TO GRID

{

B TDRTP AND C MDRTP MIN TLOW MOD TEST-(M01-2-88-033) l A TDRFP MINIMUM TLOW VALVE MOD TEST-(M01-2-88-033)

{

RECOUPLE A TURBINE DRIVEN REACTOR FEED PUMP 2TWOIKB TDRTP UN/RECOUPLE TURBINE FOR OVERSPEED TEST l

UNCOUPLE *B' TURBINE DRIVEN REACTOR FEED PUMP COUPLE

• B* TURBINE DRIVEN REACTOR FEED PUMP l

t

)

l I

I t

t t

i f

l r

I 1

1 l

H

---m

--,em

---

e w

-mcy..

..-.r

+

APPENDIX D i

REFUELING OUTAGE (L2R04)

]

SAFETY RELATED CORRECTIVE MAINTENANCE j

l WRNUM SYSTEM EPN DESCRIPTION l

LOO 514 DG 2TE-DG07710 DG CYLINDER HEAD TEMPERATURE INDICATION LOW i

LO1096 AP 2E22-S001 DG OUTPUT BREAKER TUSE HOLDER INTERFERENCE l

LO1147 AP 2AP000 6.9 KV BUS DUCT WTR DRIPPING FROM SEAM ON BUS l

LO2158 DG 2DOO4T EPCS DG FUEL OIL DAY TANK LEAKING l

LO2259 LP 2E21-F012 LPCS PUMP FULL FLOW BYPASS VALVE PACKING LEAK I

LO2527 MS 2B21-F020 EQ'UALIZATION LINE STOP VALVE PACKING LEAK LO2539 RD 2C11-D001-SCRAM VALVE AIR SUPPLY FOR 50-35 LEAK LO2764 RD 2C11-D001-054 CONTROL ROD 02-27 TULL IN LIGHT INDICATION LO3048 NR RY-2C51-K601E IRM 2E INDICATES HIGH j

LO3077 RD 2C11_000 HCU

• S RECEIVED DATA FAULTS LO3905 MS 2TT041BA PACKING LEAK

[

LO3906 MS 2TG03TB INPECT HSR DRAIN TANK FOR FLANGE LEAKS I

LO4347 MS 2B21-F021 MSL DRAIN VALVE WON'T FULL CLOSE LD4360 DG 2DG009 2VYO3A COOLER ISOLATION VALVE LEAKS j

f LO4361 DG 2DG008 2VYO3A COOLER ISOLATION VALVE LEAKS LO4383 NR RE_2B13-D193DR LFRM 56-25A READING DOWNSCALE j

LO4937 DG 2DOO64B DIESEL FUEL OIL VALVE LEAKS i

LO5384 RI 2E51-F008 RCIC STEAM SUPPLY VALVE LEAK i

LO5671 MS 2TT002AB STOP VALVE PACKING LEAK LOS673 MS 2B21-MSBPV1 TURBINE BYPASS SEAL LEAKOFF LEAKING f

LO6194 MS 2B21-RSSV2 DUAL INDICATION WHEN CLOSED

[

f LO6416 RH 2E12-C300A RER PUMP OIL LEAK LO6879 RH 2E12-F053A REBUILD RHR INJECTION VALVE & VOTES TEST LDB061 RH 2E12-F053B ADJUST RHR RETURN VALVE LIMIT SWITCHES LOB 812 DC 2DC01E 250V BATTERY REPLACE CELL #114 f

LO9371 NR RE-2B13-D193DX REPLACE LPRM STRING 48-41 LO9372 NR RE-2B13-D193N REPLACE LPRM STRING 32_49

[

LO9373 NR RE-2B13-D193GD REPLACE LPRM STRING 32-09

[

f LO9375 NR RE-2B13-D193CH REPLACE LPRM STRING 24-09 LO9377 NR RE_2B13-D193DN REPLACE LPRM STRING 16-41 LO9378 NR-RE-2B13-D193EU REPLACE LPRM STRING 16-25

[

LO9379 NR RE_2B13-D193BY REPLACE LPRM STRING 16-17 LO9380 NR RE-2B13-D193X REPLACE LPRM STRING 08-41 f

LO9383 NR RE-2B13-D193CW REPLACE LPRM STRING 32-57

[

LO9385 NR RE-2B13-D193EY REPLACE LPRM STRING 32-25 LO9387 NR RE-2B13-D193AP REPLACE LPRM STRING 16-33 LO9391 NR RE-2B13-D193CM REPLACE LPRM STRING 40-09 l

LO9392 NR RE_2B13-D193EK REPLACE LPRM STRING 40-33 i

LO9895 RI 2E51-F025 STEAM SUPPLY OUTLET VALVE PACKING LEAK j

LO9969 MS 2MS14A STEAM LEAK ON FLANGE f

L10025 RD 2C11-D001-111 CED 42-47 DOUBLE NOTCHED

{

L10323 AP 2AP27E SWGR 231A CUBICLE 101B DAMAGED L10352 RH 2E12_F325A RER STOP VALVE LEAKS LIO460 PC 2PC001A INSPECT VACUUM BREAKER LIMIT SWITCH l

L1047B NR RY_2C51-K601D D IRM ERRATIC DURING STARTUP L10526 MS 2AP39E REPAIR HIGH LOAD VALVE MOTOR LEADS L10619 MS 2B21-F409A MSR VALVE LEAKAGE

[

L10675 DG ODG03J PANEL LUGS BENT IN EXCESS OF 45 DEG f

L10676 DG ODG02JA PANEL LUGS BENT IN EXCESS OF 45 DEG l

l

APPENDIX D REFUELING OUTAGE (L2R04) i SAFETY RKLATED COERECTIVE MAINTENANCE WRNUM SYSTEM EPN DESCRIPTION L10677 DG ODG02JB PANEL LUGS BENT IN EXCESS OF 45 DEG L10689 DG 2DG035 SAMPLE STOP VALVE LEAKING BY SEAT l

L10709 DG 2DG005 DG COOLER INLET ISOLATION VALVE CORRODED L10763 DG 2E22-P301A NO INSULATION ON LUGS AT TERMINALS 31 + 32 L10770 DG 2E22-P301A PANEL LUGS BENT IN EXCESS OF 45 DEG I

L10831 DG 2DG02JA REPLACE BENT LUGS L10833 DG 2DG05J REPLACE CORRODED TERMINAL SCREWS i

L11086 RD 2C11-D001-113 HCU 50-31 ACCUM CHARGING WATER VALVE LEAKS L11147 MS 2B21-F418A AUX SUPPLY STEAM STOP VALVE PACKING LEAK L11148 MS 2B21-MOVSV-7 SEAT DRAIN VALVE PACKING LK L11231 RD 2C11-000 BCU 18-03 BAD INDICATION AT POSITION 04 L11386 RD 2C11-D001-075 SPURIOUS ACCUM 14-15 WATER LEVEL ALARMS L11520 MS 2B21-F481B ROOT VALVE PACKING LEAKS L11657 RD 2C11-D177-113 CRD 42-11 DISC SEPARATED FROM STEM L11706 RH 2E12-F021 FULL FLOW TEST VALVE LEAKING OIL L11867 NR RE-2B13-D193EB REPLACE LPRM SING 08-33 L11883 MS PS-2B21-N015B LOW PRESS INS 7 UMENT VENT VALVE LEAKS L12030 RI 2E51-F031B PERFORM PUMP EsCTION STOP VALVE VOTES TEST L12067 LP 2E21-F009 LPCS CHECK VALVE INCORRECTLY ORIENTED L12099 RD 2C11-DOO1-179 REPLACE HCU 34-11 ACCUMULATOR L12148 RD 2C11-D001-065 REPLACE HCU 22-19 ACCUMULATOR

[

L12149 RD 2C11-D001-052 REPLACE CONTROL ROD DRIVE 10-27 L12150 RD 2C11-D001-006 REPLACE CONTROL ROD DRIVE 06-31 L12151 RD 2C11-D001-030 REPLACE CONTROL ROD DRIVE 10-39 L12152 RD 2C11-D001-143 REPLACE CONTROL ROD DRIVE 42-31 L12153 RD 2C11-D001-120 REPLACE CONTROL ROD DRIVE 38-43 L12154 RD 2C11-D001-095 REPLACE CONTROL ROD DRIVE 34-59 L12155 RD 2C11-D001-065 REPLACE CONTROL ROD DRIVE 22-19 L12156 RD 2C11-D001-051 REPLACE CONTROL ROD DRIVE'14-27 L12428 RD 2APO4E CRD PUMP BREAKER ARCING CONTACT CRACKED l

L12481 PC 2PC000 ELECTR PENE 26 FAILED TO PRESSURIZE L12483 NR RY-2C51-K605GR 2D APRM UPSCALE NEUTRON TRIP LIGHT ENERGIZED L12498 RH 2E12-F048A TEST RHR BYPASS VALVE SPRING PACK l

L12508 RH 2E12-F090B LEAK-OFF PLUG LEAKING L12511 LC 2E32-F302E VALVE PACKING LEAK L12514 DC 2DC000 DIV I GRD DETECTOR ALARMING IN CONTROL ROOM l

L12517 ND LPR-2B21-R884B DIV II CHART DRIVE NOT WORKING L12521 MS 2TT060AA 24A SAMPLE LINE VALVE LEAKING l

L12524 MS 2HD049A 23A NORMAL DRAIN VALVE PACKING LEAK j

L12525 MS 2HD010F MSR DRAIN CHECK VALVE PACKING LEAK L12526 MS 20G37EA REPLACE PREHEATER DRAIN TRAP BYPASS LINE ELLOW l

L12528 MS 2HD059A REPAIR 26A HEATER NORMAL DRAIN CHECK VALVE L12529 MS 2B21-F396B STEAM SUPPLY FLOW VALVE PACKING LEAK L12530 MS 2N62-F313B PREHTR STEAM SUPPLY VALVE PACKING LEAK L12531 MS 2N62-F314A PREETR STEAM SUPPLY VALVE LEAKING L12532 MS 2N62-F315B PREHTR STEAM SUPPLY STOP VALVE LEAKING

]

L12534 LC 2E32-T009 DEPRESSIZING VALVE LEAKS L12539 MS 2LT-HD021 ROOT STOP VALVE PACKING LEAK L12540 MS 2AS057 AUX SAMPLE BOILER PACKING LEAK l

i I

q e'

--v---

g.

m-mae

+

.e-,

e v--i-c

e i

APPENDIX D REFUELING OUTAGE (L2R04)

SAFETY RELATED CORRECTIVE MAINTENANCE 3

WRNUM SYSTEM EPN DESCRIPTION L12541 MS 2HD149B 26B HEATER DRAIN LINE PACKING LEAK l

L12543 RH 2E12-F042A REPLACE VALVE MOTOR L12548 RD 2C11-F402A SDV VENT VALVE NO CLOSED INDICATION i

L12549 MS 2B21-F019 MSIV DRAIN ISOLATION VALVE PACKING LEAK j

L12667 VG 2PDI-VG021 MOIST SEPERATOR INDICATOR FAILED CALIBRATION l

L12684 DC 2EI-DC057 BATTERY VOLTMETER FOUND DEFECTIVE l

L12688 RI 2E51-F028 VAC PUMP DISCH ISOL' EXCEEDED ADMIN LIMIT L12700 MS TR-2B21-R816 REPAIR RECORDER

[

L12741 NB NB13-28075 REMOVE PSA-1/4 SIZE SNUBBER l

L12742 MS 2MS000 REMOVE HEATER BAY SNUBBERS i

L12762 MS MS01-2877S BOTTOM LOAD STUD NOT IN PLACE f

L12766 RI 2E51-F072 RCIC TEST TAP PIPE THREADS DAMAGED L12774 MS 2B21-F398B REINSULATE ROOT STOP VALVE i

L12779 RI 2E51-F004 UPSTREAM STOP VALVE MOVED FROM OPEN-CLSD

[

L12819 DG 2DG027 MOTOR COOLER INLET VALVE LEAKS

{

L12839 DG 2E22-5001 CLEAN EXPANSION TANK SIGHTGLASS

{

L12844 HP 2VD06C INSPECT HPCS SWGR BATTERY ROOM FAN WIRES

{

L12851 DG 2AP07E REPAIR AUX 243-1 CUBICLE 2B SLICE IN WIRE L12852 DG 2AP07E INSPECT 243-1 CUB 5F HPCS FUEL TANK FAN WIRES L12859 DG 2AP07E INSPECT 243 CUBICLE 5 BREAKER l

LI2860 DG 2AP07E INSPECT 243 CUBICLE I BREAKER l

L12861 DG 2AP79E REPAIR 243-1 CUB 2D WIRE INSULATION j

L12915 RI FCK-2E51-R600 REPLACE RCIC FLOW CONTROLLER THUMBWHEEL l

L12928 NR RE-2C51-N001C SRM 2C INCORRECT INDICATION j

L12941 HP 2AP79E REPLACE MCC 243-1 BROKEN LUG l

L12949 NB 2B21-H402C MSIV INSTRUMENT STOP VALVE LEAKS l

L12950 NB 2B21-N400C ARI INSTRUMENT STOP VALVE LEAKS l

L12961 MS 2B21-MOVSV1 SEAT DRAIN VALVE BREAKER TRIPPING l

L12974 RD 2C11-D001-112 HCU 06-27 SCRAM VALVE RESTRICTED STEM TRAVEL l

L12989 LC 2E32-F007 OUTBOARD BLEED VALVE PACKING LEAK L13023 RI 2E51-F065 REPLACE TEST CHECK VALVE CONDUIT L13056 DC 2DC07E DIV 1 CELL $26 LOW SPECIFIC GRAVITY L13122 RH 2E12-F068A RHR OUTLET VALVE DIFFICULT TO OPERATE L13123 RD 2C11-D001-106 HCU 34-51 COIL INADVERTANTLY ENERGIZED i

L13134 RH 2E12-F334A SEAL COOLER OUTLET VALVE LEAKS L13136 MS 2B21-RSLLV2 STEAM TO MSR 2B BLOWN DIAPHRAM i

L13168 RP 2C71-S001A RPS M/G ATTEMPTED TO START WHEN BREAKER CLOSED L13172 DG 2VD06C BATTERY ROOM FAN FAILED TO TRIP BREAKER L13174 DG PS-2E22-N513 REPLACE DG LUBE OIL PRESSURE SWITCH L13264 DG 2DG048B COMPRESSOR DISCHARGE RELIEF VALVE LIFTS L13269 DG 2PI-DG098A CALIBRATE AIR SUPPLY PRESSURE INDICATOR L13292 DG PI-2E22-R512 BG LUBE OIL, UNABLE TO CALIBRATE GAUGE L13293 DG TS-2E22-N516 REPLACE TEMPERATURE SWITCH L13310 RD 2C11-D001-075 REPAIR RMCS SOLENOIDS LUG BROKEN L13332 RH 2E12-D300B DEBRIS FROM STRAINER IN WATERBOX L13393 PC 2LV96E ELECTRICAL PENETRATION LEAKAGE L13396 DG 2E22-P301B HPCS D/G PANEL INTERMITTENT GROUNDS L13446 RD PS-2C11-N138 ACCUM 58-39 PRESS SW FAILED L13453 HP 2E22-F015 REPLACE HPCS SUCTION VALVE ELECT TERMINATION

l

\\

l APPENDIX D l

REFUELING OUTAGE (L2R04)

SAFETY RELATED CORRECTIVE MAINTENANCE WRNUM SYSTEM EPN DESCRIPTION L13456 LP FS-2E21-H004 LPCS MIN FLOW SWITCH OUT OF TOLERANCE L13471 MS MS38-28115 SNUBBER LOCKED RIGID L13472 RH 2E12-D300A DETERMINATE RHR WS STRAINER GEAR MOTOR L13479 AP 2AP06E REPLACE OUTPUT BREAKER TSC L13486 PC 2H13-P611 TIGHTEN RX VESSEL ISOLATION RELAY SCREW L13487 PC 2H13-P609 TIGHTEN LOGIC ALARM RELAY L13488 PC 2H13-P609 TIGHTEN LOW COND VACUUM RELAY L13491 DG 2DG08CB AIR COMP MOTOR BREAKER TRIPS ON THERMALS L13511 RD 2C11-D001-070 REPLACE HCU 02-19 FITTING L13512 RD 2C11-D001-073 REPLACE HCU 22-15 FITTING L13513 RD 2C11-D001-147 REPLACE HCU 58-27 FITTING L13525 NR RY-2C51-K605V LPRM 48-49B UPSCALE ALARM WILL NOT RESET L13535 RD 2C11-000 HCU 54-27 DRAIN LINE OBSTRUCTIONS L13536 RD 2C11-000 REPLACE HCU 42-19 DRAIN VALVE CONNECTOR L13594 DG 2AP07E REPLACE SWGR 243 CUBICLE 3 PLUNGER CLIP L13597 RH PS-2B22-N413A RX INJECTION LINE INST STOP VALVE LEAK L13626 DG 2AP07E ADJUST DG OUTPUT BREAKER PLUNGER WASHERS L13653 RD 2C11-D001-112 HCU 38-47 NITROGEN BLOCK VALVE STEM LEAK L13669 RH 2E12-F011A VALVE PACKING LEAK L13670 RH 2012-F027A STOP VALVE LEAK L13676 RH 2E12-F042B DETERMINATE RHR ISOLATION VALVE ACTUATOR L13688 RH II-2E12-C002C RHR PUMP AMMETER FAILED CALIBRATION L13694 RD 2C11-D001-007 HCU 26-51 UNION LEAKING AIR L13695 MS 2B21-F068 VALVE LEAK L13696 MS 2B21-F069 VALVE LEAK L13697 MS 2B21-F069 VALVE WILL NOT TORgUE WHEN FULL CLOSED LI3700 RD PS-2C11-N067 ACCUMULATOR PRESS SWITCH FAILED TO OPEN L13704 MS 2B21-F068 DETERM/RETERM LIMITORQUE FOR VALVE REPAIRS L13724 RH 2E12-F041B REPAIR TESTABLE CHECK VALVE CONDUIT L13725 RH 2E12-F092B REPAIR LPCI VALVE CONDUIT L13743 AP 2AP37E MCC 231A-3 CUBICLE TRIPS L13759 RH 2TS-VY004 INSPECT LPCS/RCIC PUMP COOLER FAN INTERLOCK L13767 RD 2C11-D001-052 HCU 10-27 CRD VENT DRAIN VALVE LEAKING L13768 RD 2C11-D001-051 HCU 14-27 CRD VENT DRAIN VALVE LEAKING L13783 MS 2B21-F513B MSR STEAM STOP VALVE WIRE HEAT DAMAGED L13802 NB 2H13-P644 REFLACE CRACKED COVER ON AGASTAT RELAYS L13810 MS 2B21-F067C REPLACE MSL DRAIN VALVE CONDUIT L13833 RH 2E12-F004A RER PUMP SUCTION BREAKER TRIPPED L13815 RD PS-2C11-N116 HCU 46-55 NITROGEN LEAKS L13820 RD 2C11-D001-052 REPLACE HCU 10-27 SCRAM VALVE DIAPHRAGM L13824 RD 2C11-D001-146 HCU ALARM UP WITH 10008 PRESSURE L13825 AP 2AP000 REALLIGN SAT FEED CONTACT L13833 NB TC-2B21-H030J2 THERMOCOUPLE READS OPEN L13838 AP 2AP000 REPAIR 241X-Y CONTROL SLIDE L13839 AP 2AP20E MCC 235Y-3 FAILED TO AUTO TRIP L13854 RD

' 2C11-D167-11F REPLACE HCU 34-19 VALVE O-RING L13857 RD 2C11-D001-066 HCU 18-19 ACCUMULATOR LEAKS L13906 AP 2AP20E FEED BREAKER WOULD NOT CYCLE L13911 RD 2C11-D001-113 HCU 10-27 CHARGING WATER VALVE LEAKS a

. ~

I I

l l

APPENDIX D REFUELING OUTAGE (L2R04) i SAFETY RELATED CORRECTIVE MAINTEN ANCE WRNUM SYSTEM EPN DESCRIPTION L13926 RD 2C11-D001-177 HCU 42-11 WATER IN INSTRUMENT BLOCK L13927 RI 2E51-F019 SEAL-IN CONTACT WORKS INTERMITTANTLY L13958 RH 2E12-C300B REPAIR 235Y COMPT 203C BREAKER CABLES L13964 RH 2E12-F009X-R REPLACE EMERGENCY FEED CONTACTOR l

L13965 RH 2E12-F009 -F REPLACE NORMAL FEED CONTACTOR l

L13969 RD 2IA186 REPAIR HEADER PRESSURE REGULATOR L14042 MS 2B21-F514B DETERM AND RETERM LIMITORQUE r

L14111 RI 2E51-r360 TRIP / THROTTLE VALVE NO INDICATION L14253 RI 2E51-F063 RCIC ISOLATION VALVE FAILED LLRT l

L14254 RI 2E51-F063 REPLACE VALVE L14260 AP 2AP90E INSPECT AUX TPANSFORMER L14303 RI 2E51-F065 INJECTION CHECK VALVE FAILED LLRT L14314 RD 2C11-000 HCU 50-43 NO POSITION INDICATION L14333 MS 2B21-MOVCA4 VALVE TO TORQUE OUT WHEN 1/2 CLOSED L14375 NH 2H13-P601 NO ALARM WHEN SRV WAS OPENED I

L14403 MS 2B21-F013N LEAK IN SOLENOID L

L14420 HP 2E22-F007 REPAIR CHECK VALVE BONNET LEAK f

L14431 RH 2E12-F050B ADJUST RHR VALVE PACKING l

t L14444 RD XY-2C11-K908 REPAIR CRD PRESS CONTROLLER STABILIZING VALVE L14445 RD 2C11-0001-117 SCRAM SOLENOID PILOT VALVE LEAKING L14461 NB LT-2B21-N406B INVESTIGATE VALVE LEAKAGE L14467 RD 2C11-D097-127 HCU 42-55 OUTLET SCRAM VALVE PACKING LEAK L14468 RD 2C11-D049-127 HCU 22-27 OUTLET SCRAM VALVE PACKING LEAK L14469 RD 2C11-D088-127 HCU 18-07 OUTLET SCRAM VALVE PACKING LEAK L14470 RD 2C11-D084-127 HCU 14-07 OUTLET SCRAM VALVE PACKING LEAK L14471 RD 2C11-D021-127 HCU 18-43 OUTLET SCRAM VALVE PACKING LEAK l

L14487 RH 2VYO7C RER WS FUMP FAN TRIPPING ON OVERLOAD L14507 HP 2E22-C003 REPLACE MCC 243-1 CUBICLE 2C HANDLE L14538 RD 2C11-000 REPLACE SCRAM SOL ELECTRICAL BOX COVERS L14555 MS MS33-2801 REPLACE PIPE CLAMP SPACER & SPACER BOLT L14558 MS MSO4-2655C UNPIN CONSTANT PIPE SUPPORT L14575 HR RY-2C51-K605CB LPRM LOCAL DONNSCALE LIGHT L14589 MS 2B21-MOVCA5 MSR CROSS AROUND MOTOR LOOSE L14624 RH 2E12-F041C LPCI TESTABLE CHECK VALVE NO INDICATION L14634 RD 2C11-000 HCU 30-35 NITROGEN LEAKS

{

L34727 RI SE-2E51-N908 RCIC GOVENOR SPEED PICKUP SEPARATED l

L14838 DC 2DC16E BATTERY CHARGER PUTTING OUT LOW VOLTAGE L14858 MS 2B21-F001 RX HEAD VENT WILL NOT CLOSE FULLY L14864 MS 2B21-F067D ADJUST DRAIN VALVE ADJUST CLOSE LIMIT SWITCHES l

L14865 MS 2B21-F067A ADJUST DRAIN VALVE ADJUST CLOSE LIMIT SWITCHES

}

L14866 MS 2B21-F067B ADJUST DRAIN VALVE ADJUST CLOSE LIMIT SWITCHES L14867 MS 2B21-F067C ADJUST DRAIN VALVE ADJUST CLOSE LIMIT SWITCHES L14971 RI 2E51-F068 RESET CLOSE TORQUE SWITCH BYPASS L73C60 RI 2E51-F360 RCIC TURB TRIP & THROTTLE VALVE STEM LEAYT L87333 MS 2B21-F514B SCAV STEAM STOP VLV YOKE SEPARATING L87699 MS 2B21-RSHLV-2 MSR 2B SSR STM HI LOAD BROKEN ROD L89193 LC PDT-2E32-N054 MSIV LEAKAGE CNTL IB DP XMTR LEAKS L94613 RH 2E12-F072B RHR PMP DISCH DRAIN STOP VLV LEAKS j

L97545 RD 2C11-D001-062 REPLACE HCU 02-23 ACCUMULATOR l

l I

m

_y

_..n,.

.. e 4

y

.b

J APPENDIX D i

REFUELING OUTAGE (L2R04)

SAFETY RELATED CORRECTIVE MAINTENANCE f

f WRNUM SYSTEM EPN DESCRIPTION i

L97549 RD 2C11-D001-039 REPLACE HCU 02-35 ACCUMULATOR L97553 RD 2C11-D001_047 REPLACE HCU 30-27 ACCUMULATOR L97554 RD 2C11-D001-185 REPLACE HCU 34-07 ACCUMULATOR L97572 FD 2C11_D001-033 HCU 26-35, MULTIPLE WATER ALARMS L97719 AP 2AP07E BUS 243 CUB 3 CRACKED TRIP & CONTROL FUSE L98067 RH 2DC13E DIST PNL 212Y CB2 UNACCEPTABLE BOLT CONNS L98319 RH 2E12_F024B RER FULL FLOW TEST WILL NOT STAY IN HANDWHEEL L98421 RH 2E12-F027B RER CONT SPRAY VLV RPLC MISSING PLATE L98667 AP 2AP07E BUS 243 CUB 1 SAT BKR INSTALL /RMV SHIMS TO ALIGN L98668 AP 2AP07E BUS 243 CUB 5, TUSE HOLDER UC FD 243-1 RPLC STAT L98669 AP 2AP07E BUS 243, CUBES 1,3,4,5 FUSEHOLDERS RPLC PULLOUT L98687 AP 2AP07E BUS 243 CUB 1 SBM SW SAT BKR RPLC SBM SW L98688 AP 2AP07E EUS 243 ENCLOSURE BOT CNTR BOLT HOLE IS STRIPPED l

L98689 AP 2AP07E BUS 243 CUB 1 DOOR SAT BKR MAKE ADJUSTMENT L98690 AP 2AP07E BUS 243 CUB 1 SAT BKR RPLC/RPR/ADJ POSITIVE MECH l

L98692 AP 2AP07E BUS 243 BUS SUPPORT RPLC DAMAGED FIBERGLASS l

L98709 HG 2HG002A INLET O/B VLV RPLC TORQUE SW LIM PLATE L98827 RI 2E51 F031 RCIC PP S.P.

SUCT VLV LIMITER PLATE L99035 RD 2C11-D001-120 CRD HCU 34-43 DIR CNTRL VLV REPL L99068 DC 2DC11E 125 VDC DISTRIBUTION 211Y CIRCUIT BREAKER 21 l

L99095 RD 2F23--E006 RE CONDUIT DAMAGED JUST UNDER CAROUSEL

[

L99171 MS 2B21_RSSV2 MSR 2B SSR STM SOURCE VLV TORQUE SW j

L99176 NB NB13-2807S PIPE CLAMP FOR SUPPORT NB13-28075 GOUGED L99232 RD 21A39A SCRAM AIR HDR PIPING ELBOW IN 21A39A HAS SM DENT i

L99478 MS 21A050A 2B21-F028B O/B MSIV AIR ISOL GLOBE VALVE L99481 MS 21A000 PILOT AIR ISOL VLV TO D MSIV O/B LEAKS L99500 RD 2C11-D001-115 HCU 22-43 BALL CK VLV LEAKS BY L99501 RD 2C11-D001-126 HCU 54-31 SEAT LEAKS BY f

L99502 RD 2C11-D001-126 HCU 38-51 SCRAM VALVE SEAT LEAKAGE L99521 RD 2C11-D001-113 HCU 38-07 WATER STOP VLV INOP L99534 DC 2DC01E 250V BATT CELL 8101 RMV'D, REPLACE L99553 RD 2CIl-D001-048 CRD MECHANISM LOCATION 26-27 HAS BAD SEALS L99554 RD 2C11-D001-134 CRD MECHANISM LOCATION 46-35 HAS BAD SEALS L99622 RI 2E51-F063 SEAL LEAK OFF REACHED ALARM PT j

L99655 LC 2E32-F001E B LINE I/B MSIV-LCS UPSTEM BLEED VALVE L99700 RD 2C11-D001-113 CRD HCU CHARGING WTR VLV EXTREMELY

. ATTACHMENT B II.B UNIT SHUTDOWNS i

(UNIT 2) l DATE:_220.415 GENERATOR OFF-LINE: 103.4 OUTAGE TYPE:.f.ntced (L2 fill (YYMMDD)

(Hours)

REASON: Manual reactor scram due to pressure oscillations induced by spurious opening and closing of the turbine bypass valves.

CRITICAL ACTIVITY PATH:

Performance of a Main Condenser tube leak test.

Return of the Condensate and Circulating Water systems to operation.

l Investigation of the Electro Hydraulic Control system pressure oscillations.

CORRECTIVE ACTIONS (DVR/LER5 if applicable):

DVR8 1-2-92-048 LERR 92-004-00 t

RADIOACTIVITY RELEASE / EXPOSURE OVER 10% ALLOWABLE VALUES: None.

l l

l SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED HEIE711_ SIS EPN DESCRIPTION L15032 DC 2DC17E Division II Charger amps / voltage oscillating L15162 RD 2C11-0000 Scram valve delay indication L15184 RD 2C11-117 Rebuild scram solenoid for rod 38-11 L15301 PC 2FS-CM902 Flow switch failed to alarm l

DATE:._91032D__

GENERATOR OFF-LINE: 1.6 __

OUTAGE TYPE:

Scheduled (L211011 (YYMMDD)

(Hours)

REASON: Main Turbine overspeed trip test.

CRITICAL ACTIVITY PATH:

Completion of the overspeed trip test.

CORRECTIVE ACTIONS (DVR/LERE if applicable): None.

RADIOACTIVITY HELEASE/ EXPOSURE OVER 10% ALLOWABLE VALUES: None.

1 SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED: None.

l

i ATTACHMENT B II.B UNIT SHUTDONNS (UNIT 2)

DATE: 9208.27 GENERATOR OFF-LINE: 116.8 OUTAGE TYPE: _Igreed (L2F18)

(YYMMDD)

(Hours)

~

REASON: Automatic reactor scram due to a main turbine trip. The turbine trip was caused by a spurious signal from the main turbine thrust bearing wear detector.

i CRITICAL ACTIVITY PATH:

Investigation and correction of the Turbine Driven Reactor Feed Pumps.

Testing of the Reactor Core Isolation Cooling system piping.

Testing of all eighteen Safety Relief Valves.

CORRECTIVE ACTIONS (DVR/LERS if applicable):

DVRS 1-2-92-067 LER# 92-012-00 RADIOACTIVITY RELEASE / EXPOSURE OVER 10% ALLOWABLE VALUES: None SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED:

WFJiUli_ SXS EPN DESCRIP_IION L17110 DC 2PA08J 25 VDC ground on the Division I detector L17389 NR 2C51-K601E Erratic indication L11395 NB 2H13-P601 SRV full open alarm not functioning L17396 NB 2B21-N575B SRV false position indication i

L17397 NB 2B21-N575A SRV false position indication L17451 RI 2E51-F066 Repair indication hinge pin L17460 RI 2E51-F065 Adjust position indication cams L17516 NB 2B21-N575N Inspect position indication transmitter L17517 NB 2B21-N575M Inspect position indication transmitter L17519 NB 2B21-N575K Inspect position indication transmitter I

L17557 NB 2B21-D004A Repair damaged insulation on condensing pot L17558 NB 2B21-D004B Repair damaged insulation on condensing pot L17559 NB 2B21-D004C Repair damaged insulation on condensing pot L17560 NB 2B21-D004D Repair damaged insulation on condensing pot 1

i I

l

i ATTACHMENT B II.B UNIT SHUTDOWNS (UNIT 2) 1 DATE:_221116 GENERATOR OFF-LINE: 67.0 OUTAGE TYPE:

Forced (L2F19)

(YYMMDD)

(Hours)

REASON: Automatic reactor scram due to a loss of the station service / instrument air system.

CRITICAL ACTIVITY PATH: Repair a flange leak on the

'B' moisture separator reheater.

CORRECTIVE ACTIONS (DVR/LERI if applicable):

DVR5 1-2-92-082 LERI 92-016-00 RADIOACTIVITY RELEASE / EXPOSURE OVER 10% ALLOWABLE VALUES: None SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED:

WPRUIL SIS EPN DESCRIPTION L14658 NH 2C51-K600CA Erratic indication L17682 RD 2C11-D001-022 Replace the withdrawal supply valve l

s l

1

ATTACID4ENT C

=

II.B FORCED REDUCTIGZS IN POWER g

GREATER THAN 20% IN DESIGN POWER LEVEL (UNIT 2)

DATE:

920515 OPERATION AT REDUCED POWER:

24.0 (YYMMDD)

(Hours)

REASON: Reduced power level due to high levels in the 21 and 22 Feedwater heaters when the Motor Driven Reactor Feedpump was placed in service.

CRITICAL ACTIVITY PATH: None CORRECTIVE ACTIONS (DVR/LERI if applicable): None RADIOACTIVITY RELEASE / EXPOSURE OVER 10% ALLOWABLE VALUES: None l

SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED: None i

l 1

1 I