ML20100M136

Annual Rept for LaSalle County Nuclear Power Station for Period Jan-Dec 1995

ML20100M136
Person / Time
Site:	LaSalle
Issue date:	12/31/1995
From:	Ray D COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9603050224
Download: ML20100M136 (69)

Text

o .._ Qimmonwealth Edmn Company

' 12Nalle Generating $tation i *' 2Gli North 2 int Road I

Marseilles. IL 61.44 l>>757 -

Tel N W357M61 February 29,1996 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Enclosed for your information is the Annual Report for LaSalle County Nuclear ,

Power Station for the period January 1995, through December 1995.

D.J. Ray Station Manager i LaSalle C,ounty Station l DJR/mkl ,

I Enclosure cc: H. J. Miller, Regional Administrator - Region 111 P. G. Brochman, Senior Resident inspector - LaSalle  !

C. Matthews, IDNS Resident inspector, LaSalle lilinois Department of Nuclear Safety - Springfield, IL M. D. Lynch, LaSalle Project Manager, NRR D. L. Farrar, Nuclear Regulatory Services Manager, NORS GE Representative - LaSalle Regulatory Assurance Supervisor - LaSalle INPO Records Center Central File - LaSalle l

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9603050224 951231 y PDR ADOCK 05000373 /$ I R PDR m m,me, ,a, jP :' 1 '

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s TABLE OF CONTENTS I. Introduction II. Annual Reportable Documentation for Unit 1 and 2 A. 3 - ry of Operating Emperience f B. Unit Outages and Power Reductions C. Radiation Exposure l i

D. Failed Fuel Elements j E. Tests and Experiments not covered in the Safety Analysis Report F. Changes to Procedures Covered in the Safety Analysis Report ,

G. Sususary of Changes to the Facility Which are Described in the Safety Analysis Report E. Survey of Evaluation Results of Chlorine Shipments by Barge on the Illinois River i

I. Suasaary of Events Violating Technical Specification 3.4.5 Primary coolant Iodine Spiking Exceeding Allowable Limits l

Attachment A - Safety Related Maintenance Completed (Non-Outage  ;

Related) ,

1 Attachament B - Unit Sh'atdowns i Attachment C - Forood Reductions in Power l l

Attachment D - Radiation Exposure Appendix A - Critical Path Activities - L2R06 Refuel Outage Appendix B - Safety Related Corrective Maintenance- L2R06 ,

Refuel Outage i

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I. Introduction The Lasalle County Nuclear station is a two-Unit facility owned by Ccesmonwealth Edison Cossany and located near Marseilles, Illinois.

Each unit is a Boiling Water Reactor with a designed not electrical l output o* 1078 Megawatts. Waste heat is rejected to a anan-anada cooling pond using the Illinois river for make-up and blow-down. The architect-engineer was 5 argent and Lundy and the contractor was Cosmoonwealth Edison Company.

Unit one was issued operating license number NPF-11 on April 17, 1982. Initial criticality was achieved on June 21, 1982 and .

I ccannercial power operation casmssaced 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 ,

consorcial power operation cessmenced on October 19, 1984.

This report was compiled by Nichael J. Cialkowski, telephone number (815) 357-6761, extension 2056.

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II. Annual Reportable Documentation for Unit 1 and 2 ,

A. susunary of operating Esperience The summary of the operating experience has been reported monthly in LaSa11e's NRC Monthly Reports (Section II.A) dated ,

January 1995 through December 1995. For safety related j maintenance (non-outage related) perfonned during the period of  ;

January 1995 thru December 1995, see Attachment A.

3. Unit Outages and Power Reductions For unit outages, see Attachment B. For unit power reductions see Attachment C.

C. Radiation Esposure For the radiation exposure of LaSalle Station personnel for the 4

reporting period of January 1, 1995 to December 31, 1995, see Attachment D.  ;

D. Indications of Failed Fuel Elements During this reporting period, January 1, 1995 through December 31, 1995, there were no indications of failed fuel elements.

E. Tests and

• m riments not covered in the safety Analysis Report During this reporting period, January 1, 1995 through December 31, i 1995, there were no tests or experiments conducted that are not covered in the Final Safety Analysis Report.  ;

t F. Changes to Procedures Covered in the safety Ana*.ysis Report LAP-200-5 Revision 6, Transfer of Control Room Command Functions Between The Control Room Unit Supervisors, The Shift Engineer, And/or The Field Supervisor The procedure was revjsed to allow transfer of control room command functions between init Supervisors, Field Supervisor, and the Shift Engineer. This consisted of title changes to reflect the nomenclature describing the present description of the Operating organization at LaSalle Station. The Field Supervisor maintains the same qualification as the Shift Foreman.

The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question associated with this activity.

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O F. Changes to Procedures Covered in the Safety Analysis Report (continued)

LAP-240-7 Revision 6, Defeating Annunciators This procedure was revised to update the method of addressing control room nuisance alarms. It also incorporates the method of ,

time delaying annunciators from LaSa116 Administrative Procedure (LAP) LAP-240-6, Temporary Alterations, therefore, all annunciator problems are addressed by a single administrative procedure. I

, The 10CFR50.59 safety evaluation concluded that there is no unreviewed safety evaluation associated with the procedure revision.

LAP-300-29 Revision 2, Rigging and Lifting Program The procedure was revised to incorporate the contents of LaSalle Maintenance Procedure LMP-GM-09, Safe Rigging Practices.

LMP-GH-09 was subsequently deleted.

The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question associated with this activity.

LAP-820-10 Revision 10, " Periodic Procedure Review",

' and Changes to UFSAR Section 12.5.3 and Appendix B, and Addition of UFSAR Section 13.5.3.

LaSallo County Station previously committed to ANSI 16.7, 1972 through the UFSAR and Regulatory Guide 1.33. The station changed this commitment through the Commonwealth Edison Quality Assurance Ma nual Topical Report, dated 1993.

Changes to the above UFSAR and LAP-820-10 reflect compliance ,

with the Topical Report and document our approved deviation from Regulatory Guide 1.33 (as approved in the Topical Report of 1993) .

UFSAR Section 12.5.3 was revised to remove the procedure review requirement with information being added which directs that the review be performed per the controls stated in Section 13.5.

This review has now been incorporated into the review requirements in Section 13.5, which governs the review of all station  !

procedures.

Section 13.5.3 was added to control procedure review. This new section provides the foundation for the procedure review system.

It clearly states what ANS standard procedure reviews are -

conducted under, and goes further to explain that this standard does not apply elsehwere within the procedure change process.

This will assure that the station will remain within the established program guidelines as required by Regulatory Guide  ;

1.33 and the ANSI standards. l The 10CFR50.59 safety evaluation concluded that there were I nounreviewed safety question associated with this activity.

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O F. Changes to Procedures covered in the safety Analysis Report (continued)

LOA-DC-03 Revision 5, "48/24 VDC System Failure" The procedure revision identifies specific equipment lost due to a partial or complete loss of 24/48 VDC and other associated alarms with the loss of process radiation monitors. Revisions to references contained in the Technical Specifications to reflect loss of the 24/48 VDC system. which includes reference to the Offsite Dose Calculation Manual (ODCM), and reflects UFSAR sections 7.6. 7.7, 8.3, and 13.5.

A 10CFR50.59 safety evaluation concluded that no unreviewed safety question exists as a result of this modification.

LOP-WF-30, Revision 3, Generic Transfers and Draining / Filling Procedure This change provided direction for transferring chemical radwaste reprocessing (WZ) Collector Tanks to a vendor container located in the Radwaste Truck Bay. The change involved connection of a hose to a hydrolazing port in the WZ feed line located in the upper radwaste pipe tunnel. The transfer is implemented per procedure LOP-WF-30, and it provides a means to transfer a 1200 gallon sample to a vendor shipping container.

The 10CFR50.59 safety evaluation concluded that there was not an unreviewed safety question.

LOP-WX-06, Revision 9, Establishing a Waste Sludge Tank Transfer Loop The procedure outlines the steps necessary to establish a transfer loop for the waste sludge tank and the solid radwaste system. The revision incorporates allowance the running of the  !

loop without the recycle pumps, by backflushing to the bottom suction of the waste sludge tank. The backflushing with cycled condensate is aimed at eliminating pipe plugging, and to minimize the potential for elevated dose rates in the transfer pump area.

The safety evaluation concluded that there was no unreviewed i safety question.

LTP-100-6, Revision 2, Time Delay Relay Calibration Procedure This procedure was revised to reflect the addition of Reactor Core Isolation Cooling (RCIC) Low Reactor Water Level (L2) I Initiation Time Delay Relays. The relays 2821A-K710AX and 2B21A-K710CX were installed per Component Replacement C01-2-94-032 (DCP 9400117). Component Replacement C01-2-94-033 (DCP9400118) replaced relays 2B2LA-K710BX and 2B21A-K710DX.

i s-I F. ^ m e to Precedures covered in the Safety Analysis Empert (continued) l l

LTP-100-6, Revision 2, Time Delay Relay Calibration Procedure-(continued)

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UFSAR Section 7.4.1.2.3 is being revised to reflect that the time delay is now applicable to both units. The revision of the l UFSAR in effect at the time of the procedure revision reflects the j j time delay as being applicable to Unit 1 only.

The 10CFR 50.59 safety evaluation concluded that no unreviewed safety question exists for this procedure.

l LTP-1600-10, calculating Core Thermal Power, and Update to UFSAR Figure 1.2-1

' The procedure was revised to account for an additional 3 j megawatts thermal (MWth) in the heat balance methodology. 3 NNth

is added to the constant term for fixed heat losses from the reactor pressure vessel. This accounts for 8 gpm of seal purge flow to the reactor recirculation pumps that is routed from the i control rod drive (CRD) flow. During normal operation, 4 - 6 gpm j of seal purge flow exists, however conditions can exist where this

! flow increases to 7 gpm. 8 gpm per pump is incorporated as a bounding assumption.

UFSAR Figure 1.2-1 is being updated to reflect the revised heat balance methodology.

) The 10CFR 50.59 safety evaluation concluded that no unreviewed safety question exists for thie procedure. ,

I LAP-200-10 Revision 2, NRC Operator License Active Status

< Maintenance and Reactivitation.

i LAP-220-2 Revision 17, Unit Operators' Log.

' LAP-1600-2 Revision 50, Conduct of Operations.

LAP-200-1 Revision 30, Operating Department organization.

LAP-200-3 Revision 27, Shift Change.

LOS-AA-31 Revision 52, Shiftly Surveillance l

The above listed procedures were revised to reflect the enhancements made to the control room organization.

! This change was done to modify the organization of control room operators. The Center Desk and fourth Nuclear Station Operator (NSO) positions were replaced with Unit Assist NSos. Duties previously performed by the Center Desk NSO to support unit operations are performed by the Unti AJaist NSos.

UFSAR Section 7.8.1 was revised to reflect the change of the 4 Center Desk operator.

The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question associated with the UFSAR and procedure revisions.

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o F. Changes to Procedures Covered in the Safety Analysis Report (continued)

LOA Procedure Revisions associated with the Diesel Generators (DGs) e LOA ODG03J-3-4 Revision 3 includes a revised trip setpoint for the O DG neutral ground trip relay to reflect design change E01 94-960-D, DG 0 Neutral Ground Relay Replacement.

d LOA 1(2) DG03J 4-1 Revision 2 includes a revised trip setpoint for the LA DG underfrequency relay to reflect design change C01 94-007-B, Replacement of Underfrequency Relay for DG 1A.

LOA 1(2) DG03J-3-4 Revision 2 includes a revised trip setpoint for the LA DG neutral ground trip relay to reflect design change E01-1-94-960-A, DG LA Neutral Ground Relay Replacement.

LOA ODG03J 4-1 Revision 3 includes a revised trip setpoint for the O DG underfrequency trip relay to reflect design change C01 '

94-007-A, Replacement of Underfrequency Relay for 0 DG.

LOA 1(2) DG03J-3-5 Revision 2 includes a revised trip setpoint for the LA DG reverse power trip relay to reflect design change E01-1-94-960-B, DG LA Reverse Power Relay Replacement.

4 LOA ODG03J-3-5 Revision 3 includes a revised trip setpoint for

the O DG reverse power trip relay to reflect design change E01 '

94-960-c, DG-0 Reverse Power Relay Replacement.'

I The 10CFR50.59 safety evaluations concluded that there were no unreviewed safety questions associated with the procedure revisions.

LaSalle Special Procedure LLP-95-006 Revision 0, RHR System A/B Valves 1(2)E-F040A/B and 49 A/B Motor Operated Valve Dynamic Votes Test The procedure allowed for differential pressure (DP) VOTES testing as required by NRC Generic Letter 89-10 for RHR System A/B Valves 1(23)E-F040A/B and 49 A/B. Operation of the RHR system was conducted through existing operating procedures. The system was placed in the full flow test mode from suppression pool to  ;

suppression pool. Water was not injected in to the reactor vessel '

as part of this test. This procedure was considered as ' data i collection' only. The procedure was performed such that the operation of the subject valves did not impact safe plant operation.

The 10CFR50.59 safety evaluation concluded that there was not an unreviewed safety question associated with this design change.

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F. Changes to Procedures Covered in the Safety Analysis Report (continued)

LaSalle Special Procedure LLP-95-038, Installation and Removal of Temporary Connections for Unit 2 RHR Chemical Decon Temporary Heat Exchanger The procedure provided the control and approval method for installation of temporary connections for hoses for a temporary heat exchanger to support Unit 2 chemical decon. The small temporary heat exchanger was attached to the service water system using hoses on existing service water connections. The heat exchanger was part of the Vectra skid which was evaluated for seismic considerations prior to use. The system was returned to its original condition following completion of the decon.

The 10CFR50.59 safety evaluation concluded that there is no unreviewed safety evaluation associated with the procedure revision.

LaSalle Special Procedure LLP-95-046 Revision 0, Filling and Draining Reactor Recirculation (RR)

Loop 'A' with Jet Pump Plugs Installed.

The procedure provided guidance to drain and fill the Unit 2 A loop with the jet pump plugs installed. The plugs were installed to allow draining the RR system for naintenance activities on the 2B33-F067A valve. The plugs are designed to utilize head pressure with a mechanical clamp preventing leakage of coolant on to the drywell floor. A cover plate was installed whenever the valve bonnet was removed and no one was present.

The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question associated with the procedure.

LaSalle Special Procedure LLP-95-107, Operability Test of the new Agasat Time Delay Relay (2E31- KllB)

The test was performed to to verify the both the operability and the proper installation of a new Agasate Time Delay Relay (2E31- K11B) in the power sensing circuitry for Leak Detection.

The RCIC Relay Race was eliminated due to the replacement of the previous Agastat GPI relay with a Agastat TR Timing Relay for 1(2)E31-K11B.

The test only affected circuits in Division 2, and had no effect on instrumentation. The Division 2 circuits were restored to their original design condition after the testing was completed. Valves 2E51-F063 and 2E51-F076 were stroked closed during the testing, thereby rendering RCIC inoperable for the duration of this test.

The 10CFR50.59 safety evaluation concluded that no unreviewed safety questions exist.

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F. Changes to Procedures Covered an the safety Analysis Report j (continued) ,

LaSalle Special Procedure LLP-95-109, Operation with 7 Condensate Polishers On Line During Normal Power Operation  ;

The procedure was written to allow operation of a unit for a trial period with all seven condensate polishers are on line, as ,

opposed to the routine operacion consisting of 6 polishers, and 1 spare. During the trial period, determinations were to be made if there were any applicable improvements in reactor chemistry and/or other operation practices. -

The 10CFR50.59 safety evaluation concluded that there were no r unreviewed safety questions associated with the procedure.  !

I i LaSalle Special Procedure LLP-95-ll6, Operation of the Reactor Building Overhead Crane Within 15 Feet of the East or West Wall l This procedure provides for guidance for operating the Reactor Building Overhead Crane (ABOHC) in inaccessible regions of the  !

Refuel Floor. The procedure allows temporary removal of the trolley east and west limit swiuch arms to allow the crane access l to areas within 15 feet of the east or vsst walls. The limits are restored during closeout of the procedure.

The 10CTR 50.59 safety evaluation concluded that no unreviewed .

safety question exists for thie procedure.

LaSalle Special Test Procedure LST-95-007 Revision 0,

" Unit 2 Reactor Protection System Alternate Power Source Voltage Regulation Test."

The procedure provided instructions for testing the "Solatron" voltage regulator 2APA9E. This was done to verify that the regulator was capable of regulating its output voltage in accordance with manufacturer's operating characteristics.

The 10CFR50.59 safety evaluation ec7 eluded that there was no .

unreviewed safety question.

LaSalle Special Test Procedure LST-95-039 Revision 0,

" Unit 2 Reactor Protection System Alternate Power Source Voltage Recording During Reactor Recirculation Pump 2B33-C001A/B Start" The procedure provided instructions for the recording of voltsgee at the Reactor Protection System (RPS) alternate power source during Reactor Recirculation (RR) pump start. The output l of "Solatron" Voltage Regulator 2APA9E (the alternate RPS power supply) was electrically isciated from the RPS bus during this test.

The 10CTRa0.59 safety evaluation concluded that there was no unreviewed safety question. i i

f F. Changes to Procedures Covered in the Safety Analysis Report (continued)

• Lasalle special Test Procedure LST-95-055 Revision 0, Unit 2-RR FCV Oscillation Data Acquisition and Adjustment The 2A Reactor Recirculation (RR) Flow Control Valve (FCV) was observed to exhibit excessive oscillations while being moved in the open direction. The special test procedure was written to provide instructions for acquisition of data to determine the cause of the oscillations, as well as to provide guidance for system adjustments to correct or mitigate the problem.  ;

Temporary instruments were connected to the Hydraulic Power Unit (HPU) to deterudne cause of 2A RR FCV oscillations. Pressure transmitters were connected to high point vent connections in the ,

hydraulic system, and electrical connections were made to the i Hydraulic Power Unit (HPU), HPU control drawer, Flow control  ;

drawer, and Servo control drawer. Instruments were also connected to the 2B RR loop for comparison purposes. The gain settings on controllers before and after the velocity demand limiter were allowed to be adjusted, as required, based on the data gathered.

The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question associated with the special test procedure.

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1 . l S. Susmear ei' Chameos to the Facility Which .ze Dese,ibed in the

. ,et, 1 ,.i. e  ;

Modification M01-0-89-016D i The impressed current studies performed by Commonwealth ,

Edison's Corrosion Engineering Group indicate that the level of cathodic protection afforded by the distributed anode system was .

unsatisfactory due to system failures. The near-surface distributed anodes and wiring were readily distributed and faults ,

were difficult to locate and repair. To mitigate these problems, the new cathodic protection system utilizes deep-bed anodes for l the majority of the protective currant. l This modification completes design of deep bed anode cathodic f protection in the east area of the plant by installing two new -

deep anodes and by providing a local supply of D.C. current to distributed bed anode systems No. 1 and No. 2. These two near-surface zones protect underground piping from approximately the Lake Screen House to the CSC5 Cooling Pond water inlet chute and l are less subject to incidental damage. The remaining near-surface distributed bed anode systems 5 and 6 were disconnected and abandoned.

The safety evaluation concluded that there are no unreviewed safety questions associated with this modification.

Modification M01-1-91-008 i The 120 volt AC control circuits of dampers IV0037 and IVQO38 were modified by adding a limit switch intermediate open (IO) contact to the closing circuits. This allows closing the dampers on limit rather than on torque, to prevent unnecessary tripping of the thermal overload relays during the damper closure stroke.

1 The safety evaluation concluded that there was not an l unreviewed safety question. l Modification M01-1-93-008 This modification improved the reliability and operability of motor operated valves 1E12-F0425 and 1E12-F042C. Specific requirements relating to operability have been presented in NRC Generic Latter 89-10. This modification was selected as the preferred alternative because it has the lowest cost and provides the greatest direct benefit of improved valve reliability in normal operation.

This modification also insta11ec a new 3/C #6 ANG power cable for each of the subject valve operators. These cablos replaced 3/C #14 ANG power cables, which were abandoned in place. the new, larger cables provide increased voltage at the motor tenminals under all bus voltage conditions.

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G. Sussnary of Changes to the Facility Which are Described in the Safety Analysis Report-(continued)

Modification M01-1-93-008 (continued)

This modification also changed the overall actuator gear ratio from 42.50:1 to 88.40:1. The gear change was accomplished by replacing the motor pinion and wona shaft gears. The increased gear reduction increases the available operator torque and thrust.

UFSAR Tables 6.3-2 and 6.3-3 are being revised to reflect the valve stroke and system response time changes.

A 10CFR50.59 Safety Evaluation was performed, and concluded that no unreviewed safety question exists as a result of this modification.

Modification M01-1-94-001 The modification, installed a passive zinc injection skid to inject zine solution in to the Unit.1 Feedwater and Condensate Systems. This will' help ~.to mitigate the buildup of Cobalt 60 (Co-

60) levels in the recirculation system, thereby significantly reducing personnel dose rates, and maintaining exposure ALARA.

UFSAR Sections 10.4 and 12.2 are being revised to reflect the installation of the modification.

The Safety Evaluation concluded that there was not an unreviewed safety question associated with this design change.

Modification M01-1-94-008 Agastat type 7022AB time delay relays were installed in the low suction pressure trip circuitry for pumps 1FC01PA and IFC01PB, and functionally tested by a temporary systam change. They were added to avoid spurious trips. The time delay relays are set at 2 seconds, which doesn't cause excessive cavitation of the pumps. A low suction pressure trip will still occur at the present setpoint, however, the low suction pressure must be sustained for a 2 seconds. This modification and modification M01-2-94-011 will make the change permanent, thereby clearing Temporary System Change (TSC) No. 2-306-93.

A 10CFR50.59 safety evaluation for this Modifiestion concluded that no unreviewed safety questions exist.

s. Summasy of Chameos to the Facility Which are Described in the Safety Analysis Report-(oontinued)

M01-2-87-007-3 2E12-F05QA M01-2-87-087-4 2E12-F0508 2E22-F005 M01-2-07-087-7 M01-2-87-007-8 2E21-F006 M01-2-87-087-9 2E12-F041A l These modifications replaced each Eccs Testable check Valve's packing with fewer rings of packing combined with a carbon spacer, .

to reduce valve shaft frictions removed of the leakoff line to accomodate the newer packing; and replaced limit switches with smaller and lower torque microswitches. The valves previously encountered rotational resistance during low flow testing ,

conditions, which could prevent the valves from closing during this testing. These changes are intended to minimize the rotational resistance. UFSAR Figure 7.3-7 is being revised to reflect these changes.

The 10CFR50.59 safety evaluation concluded that there is no unreviewed safety question associated with these design changes.

t Modification M01-2-89-011-A f This modification replaced, for each Turbine Driven Reactor.

Feed Pump (TDRFP), the speed control system with a state-of art, highly reliable and fault-tolerant micro-processor control system. '

The replacement system is manufactured by Lovejoy Control corporation (LCC). The previous design was obsolete, it couldn't adequately maintain turbine speed and feedwater flow and required ]

, extensive maintenance.

f The redundancy provided in the new system does not allow any single electronics failure to cause a loss of or increase in the turbine speed. The design also provides stable operation at relatively unloaded or zero speeds to pumping speed, pumping speed range regulation for precise but stable control and prevents most accidental overspeeds. The system requires less overall maintenance as there are fewer moving parts.

UFSAR Sections 7.7.4.2.2 (page 7.7-39) and 7.7.4.3 (page 7.7-

, 40) are being revised to reflect the installation on the design change.

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The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question associated with the design change.

Modification M01-2-91-001 The modification provided overpressure protection by installing a relief valve on chiller units 2VP04AA and 2VPO4AB (station

, heating system), and 2VP16A (service water system). The relief valves are set to lift at a pressure of 150 psig. j l

The 10cFR50.59 safety evaluation concluded that there was no l unreviewed safety question associated with this activity, i l

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S.. Susumsy of chanses to the rasility Which are Described in the-Safety Analysis Report-(eentinued)

Modification M01-2-92-001 -

This modification converted the Primary Containment Ventilation ,

system (VP) dampers serving the Control Rod Drive (CRD) and Reactor Recirculation (RR) Pump areas from control daspers to balancing daspers. This was accomplished by removing the damper ,

actuators and limit switches from the dampers. j The dampers modified were the 2VP01Y, 02Y, 08Y and 09Y. These  ;

despers control the amount of air supplied to the RR Pump area and the CRD area under the vessel, and operate in pairs with one  ;

damper opening to the RR Pump area and the other to the CRD area.

The dampers that serve the CRD area are closed during normal i operating conditions and open only if the CRD area temperature exceeds 150 degrees F. Maximum temperatures are limited to less than 185 degrees F. This modification was implemented due to  ;

repeated failures of both the damper actuators and the limit 4 switches. While trouble with ITT actuators is not uncommon, access to the dampers and limit switches is limited to outages. Repeated maintenance required by the dampers is undesirable due to the high dose rate in the under vessel area.

In 1991 temperature data was collected, per procedure LST '

013, to determine the actual temperature in the CRD area during normal operating condition. Data was collected prior to and  ;

during a-full core SCRAM. It was postulated that the maximum heat load would occur during and after a SCRAM. Collected temperature i data confirmed that the maximum temperature was well below the )

allowable temperature of 185 F. Based on this data, it was decided 1 that the damper actuators and associated control and indication devices were not needed to insure that CRD temperatures remained below 185 F.

UFSAR Table 3.11-13 is being be revised to reflect this modification. A 10 CFR 50.59 Safety Evaluation concluded that there was no unreviewed safety questions exist due to the design change.

! Modification M01-2-92-006 i

' The modification added a full flow recirculation line for CRD j pumps 2C11-C001 A and B, added test taps for each CRD pump

discharge line, and replaced the CRD pump discharge stop-check valves 2C11-F393A and B with globe type valves.

The full flow recirculation lines will be used during operating periods of reduced flow requirements, such as a refuel outage.

The line provides a means of maintaining the pump recommended

minimum flow during reduced CRD system flow conditions. Lack of adequate flow has resulted in a' considerable amount of pump wear,

- damage, and degradation in pump performance. The recirculation line is shut down and isolated during reactor power operation.

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G. 3 - ry of Changes to the Facility Which are Described in the Safety Analysis Report-(oontinued)

Modification M01-2-92-006 (continued)

UFSAR Section 4.6.1.1.2.4.2.1 is being revised to reflect the Unit 2 modification. The safety evaluation concluded that there was not an unreviewed safety question.

Modifications M01-2-93-001, M01-2-93-003, and M01-2-93-005 These modifications replaced the ECCS and RCIC water leg pump discharge lift check and globe stop check valves with swing check and globe stop valves. Some of the replacement valves are repositioned to minimize piping stress and ensure accessibility.

The valves were replaced to reduce the amount of required maintenance and improve the reliability of the valves. The lift check valve has proven to be susceptible to binding due to the small clearances within the valve. The replacement swing check valve, with its simple design and larger clearances, is considered a more reliable valve.

The required functions of the removed valves will are performed by the replacement valves. The number of check valves in each system is decreased from two to one, since only one check valve is needed to perform the backflow isolation function. The globe stop valve will provide the isolation necessary for maintenance.

UFSAR Section 6.3.2.2.5 is being revised to reflect the modifications.

The safety evaluation concluded that there was not an unreviewed safety question.

Modification M01-2-93-008 The modification inproved the reliability and operability of MOVs 2E12-F042B and 2E12-F042C. Specific requirements relating to operability were presented in NRC Generic Letter 89-10. Several alternatives were considered to resolve the operability concerns.

This modification was selected as the preferred alternative because it has the lowest cost and provides the greatest direct benefit of improved valve reliability in normal operation.

This modification changed the subject valves' overall actuator gear ratio from 42.50:1 to 88.40:1. This was accomplished by replacing the moter pinion and worm shaf t gears with Limitorque parts. The increased gear reduction increases the available operator torque and thrust.

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S. Susumary of Chaneos to the Facility Which are Described in the Sai'ety Analysis Report-(oontinued) 4 Modification M01-2-93-008 (continued)

The gear changes nearly double the valves' stroke times. The new calculated stroke times are about 37 seconds in each direction. These stroke times are approximate and may vary due to  !

actual valve stroke and motor speed. These stroke times meet the Technical Specification Table 3.3-3 stroke time requirement of 40 ,

seconds or less.

A 10CFR50.59 safety concluded that no unreviewed safety ,

question exists as a result of this modification.

Modification M01-2-93-009 i 7his modification improves the reliability and operability of mo',or operated valves 2E12-F042A and 2E21-F005. Specific re,quirements relating to operability have been presented in NRC Generic Letter 89-10. Several alternatives were considered to resolve the operability concerns. This modification was selected as the preferred alternative because it has the lowest cost and provides the greatest direct benefit of improved valve reliability in normal operation.

The overall actuator gear ratio (QAR) for 2E12-F042A was changed from 42.50:1 to 88.40:1. The gear change was accomplished  !

by replacing the motor pinion and worm shaft gears. The increased gear reduction increases the available operator torque and thrust.

i The Limitorque operator for 2E21-F005 was changed from an SMB-1 with a 40 ft-lb motor and QAR of 42.50:1 to an SMB-2 with an 80 ft-lb motor and an QAR of 72.01:1.

The gear changes increase the valves' stroke times 4

significantly. The new calculated stroke times are about 37 seconds in each direction for 2E12-F042A and about 30 seconds in each direction for 2E21-F005. These stroke times are approximate and may vary due to actual valve stroke and motor speed. These stroke times are within the Unit 2 Technical Specification stroke -

I 2

time requirements for these valves.

A 10CFR50.59 safety evaluation concluded that no unreviewed safety question exists as a result of this modification.

4 Modification M01-2-93-011 This modification improved the reliability and operability of motor operated valve 2E22-F004.- Specific requirements relating to operability have been presented in NRC Generic Letter 89-10. This modificatien was selected as the preferred alternative because it has the lowest cost and provides the greatest direct benefit of ,

improved valve reliability in normal operation. t

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S. 3 - ey of chamens to the Facility which are Described in the Safety Analysis Report-(continued)

Modification M01-2-93-011 (continued)

The previous gearing was not self-locking, and required a motor brake to ensure that the operator and valve remain in the desired position when the operator is de-energized. The new gearing is self-locking and does not require a motor brake.

This modification also changed the overall actuator gear ratio from 48.45:1 to 92.12:1. The gear change was accomplished by replacing the motor pinion and woom shaft gears.

UFSAR Tables 6.3-2 and 6.3-3, and Section 7.3.1.2 are being revised to reflect the modification. A 10CFR50.59 Safety Evaluation was performed, and concluded that no unreviewed safety question exists as a result of this modification.

Modification M01-2-93-012 This modification improves the reliability of and decreases the ,

stroke time of motor operated valve 2E12-F024B. Reliability was )'

improved by changing the closing circuit control logic so that the closing circuit can't be re-energized after the initial torque switch trip.

The control circuit revision eliminates the valve " hammering" that sometimes occurs when this valve is closed in the presence of an isolation signal. This situation occurs when the operator gearing relaxes following valve seating, allowing the torque q switch "close" contacts to re-close, and re-energizing the closing

• circuit. This situation occurs only when an isolation signal is present or if the handswitch is manually held in the "close" position after the valve has seated.

No unreviewed safety question exists as a result of this modification.

Modification M01-2-93-013

+ This modification improved the reliability and operability of l MOV 2E12-F024A. Specific requirements relating to operability i have been presented in NRC Generic Letter 89-10. Several j alternatives were considered to resolve the operability concerns. l This modification was selected as the preferred alternative. l The modification installed of a friction element (high pressure i drop) trim in the 2E12-F024A valve. The trim is intended to prevent severe cavitation when the valve is used to throttle flow through the "A" RHR full flow test loop. The cavitation results

- from the large pressure drop that occurs across the valve seat i

under such flow conditions.

1 The 10CFR50.59 safety evaluation concluded that there was no

, unreviewed safety question associated with the modification.

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G. S - =y of Changes to the Facility which are Described in the Safety Analysis Report-(oontinued)

Modification M01-2-93-014 The purpose of this modification is to improve the reliability and operability of motor operated valve 2E12-F021. Specific requirements relating to operability have been presented in NRC Generic Letter 89-10. Several alternatives were considered to resolve the operability concerns. This modification was selected  ;

as the preferred alternative.

The modification installed of a friction element (high pressure drop) trim in the 2E12-F021 valve. This type of anti-cavitation ,

trim is used when normal flow is over the seat of the valve. The trim is intended to prevent severe cavitation when the valve is used to throttle flow through the "C" RHR full flow test loop.

The cavitation results from the large pressure drop that occurs across the valve seat under such flow conditions.

The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question associated with the modification.

Modification M01-2-93-016 The modification brought the Unit 2 System Auxiliary Transformer (SAT) 242 protective relaying package to the level ,

required by the Commonwealth Edison Company standards. Overcurrent <

relay additions CO-2 and CO-7 were added to accomplish this.

CO-2 relay, ground differential schemes The low voltage (LV) windings of the SAT is resistance 1 grounded, which limits the amount of ground current available for i phase to ground faults. The differential and sudden pressure relays may be insensitive for phase to ground faults in the transformer LV windings or leads. Therefore, a transformer LV lead differential scheme, using high speed overcurrent relays, were needed as primary protection for these faults.

CO-7 relay, transformer low voltage breaker back-up:

This time overcurrent relay provides back-up protection in the l event that the SAT LV breaker fails to interrupt for a LV bus multiphase fault. It also provides back-up protection for transformer internal or LV lead multiphase faults.

The 10CFR50.59 safety evaluation concluded that no unreviewed i safety questions exist. l l

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S. 3 - 4 of changes to the Facility which are Described in the I

Safety nnmlysis Report-(oontinued)

Modification M01-2-93-019 The modification brought the Unit 2 Unit Auxiliary Transformer <

(UAT) 241 protective relaying package to the level required by l Commonwealth Edison Company standards. Overcurrent relay additions CO-2, CO-7, and ITH were added to accomplish this.

CO-2 relay, ground differential schemes The low voltage (LV) 1 windings of the UAT is resistance grounded, which limits the I amount of ground current available for phase to ground faults.

The differential and sudden pressure relays may be insensitive for j phase:to ground faults in the transformer LV windings or leads.

Therefbre, a transformer LV lead differential scheme, using high i speed overcurrent relays, were needed as primary protection for these faults. 1

CO-7 relay, transformer low voltage breaker back-up

This time overcurrent relay provides back-up protection in the event that i the UAT LV breaker fails to interrupt for a LV bus multiphase fault. It also provides back-up protection for transformer internal or Lv lead multiphase faults.

ITH relay, high speed overcurrent back-up protection: The current transformers used for the differential relays may saturate l for heavy multiple internal faults. To ensure high speed  ;

operation for severe saturation, solid state instantaneous  ;

overcurrent relays are provided as back-up protection. The ITH relay addition provides back-up protection to adtigate transformer damage due to heavy internal faults. ,

i i The 10CFR50.59 safety evaluation concluded that no unreviewed safety questions exist. l l

Modification M01-2-93-027

! The modification installed a passive zine injection skid to inject zine solution in to the Unit 2 feedwater system. This will help to mitigate the buildup of Cobalt 60 (Co-60) levels in the  !

recirculation systen, thereby significantly reducing personnel dose rates, and maintaining exposure ALARA. ]

UFSAR Section 3.11.1.1 is being revised to reflect the l installation of the modification.

The Safety Evaluation concluded that there was not an unreviewed safety question associated with this design change.

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8. Su-=*y of N--s to the Facility which are Described in *As ,

safety Analysis soport-(oontinued) l Modification M01-2-94-002 This modification improves the reliability and operability of motor operated valve 2E12-r053A. Specific requirements relating to operability were presented in NRC Generic Letter 89-10.

Several alternatives were considered to resolve the operability concerns. This modification was selected as the preferred alternative because it has the lowest cost and provides the l greatest benefit of improved valve reliability.

This valve was equipped with a LLaitorque SMB-3 motor operator i

~

a having a 60 foot-pound, 3600 rpm (nominal) motor and an OAR of 61.5:1. The operator's overall ratio (OAR) was changed to 57.4:1.

This was accomplished by replacing the motor pinion and worm shaft pinion gears. The motor was also replaced with a 100 foot-pound, 3600 rpm unit with the required upgrade of the power cable also being performed.

The gear change slightly reduces the valve stroke time. The original stroke time was slightly less than the Technical Specification limit of 29 seconds. The new calculated stroke time is about 27 seconds in each direction. This stroke time is approximate and may vary due to actual valve stroke and motor speed. The intent of the gear change is to provide additional margin between the actual stroke time and the required stroke time.

I The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question.

Modification M01-2-94-003 This modification improved the reliability and operability of motor operated valve 2E12-F053B. Specific requirements relating to operability have been presented in NRC Generic Letter 89-10.

Several alternatives were considered to resolve the operability concerns. This modification was selected as the preferred alternative because it has the lowest cost and provides the greatest benefit of improved valve reliability.

This valve is equipped with a Limitorque SMB-3 motor operator having a 60 foot-pound, 3600 rpm (nominal) motor and an overall ratio (OAR) of 61.5:1. The operator's OAR was changed to 57.4:1, by replacing the motor pinion and worm shaft pinion gears. The motor was also replaced with a 100 foot-pound, 3600 rpm unit.

These changes, in addition to other benefits, increase the operator's gearing capacity.

A 10CFR50.59 safety evaluation concluded that no unreviewed safety question exists as a result of this modification.

i

e S. Summazy of chameos to the Famility Which are Described in the Safety Analysis Report-(oontinued)

Modification M01-2-94-007 i This modification addresses NRC Generic Letter 89-10 and General Electric Nuclear Services Information Letter (SIL) #377 concerns related to the Reactor Core Isolation Cooling (RCIC) steam supply stop Valve, 2E51-F045.

This modification replaced the Rockwell globe valve with an Anchor / Darling globe valve having a special contour plug design. >

j This was done to improve the startup performance of the RCIC system. l The modification also increased the valve's overall actuator gear ratio (OAR) from 34.96:1 to 51.80:1. The increased gear reduction will increase the available operator torque and thrust, a concern of the Generic Letter.

i i The new 2E51-F045 valve will reduce turbine speed overshoot on

• RCIC startup and it will have more capacity to perform its l function, thus making the RCIC system more reliable.

The safety evaluation concluded that there was not an  !

, unreviewed safety question associated with this design change.

M01-1-94-006 Unit 1 M01-2-94-008 Unit 2 i The modifications revised subsystem A of both units Rod Worth i Minimizer (RNM) so that new control rod position information is stored in RWM A memory and displayed on the CRT at control room panels 1H13-P603 and 2H13-P603 after each scan of the Rod Position Information System (RPIS) during a scram. Also, unknown position codes generated by rods which have not activated any of the reed switches on the rod position probe, such as those which have inserted past the full-in reed switches, will not erase previously

stored valid position data.

These design changes were installed to resolve a problem with rod position indication following a scram. Following a scram, position indication has been lost in several instances for a few i control rods. A preliminary investigation determined that this 4 problem was caused by the control rod inserting past the full-in position indication reed switch. Following the installation of these changes, the last position indication reed switch that is picked up by the control rod is displayed by RWM A during a scram.

A 10CFR50.59 safety evaluation concluded that no unreviewed safety questions exist.

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S. Sesmary of Chameos to the Famility Which are Deecribed in the i Safety Analysis Report-(oontiamed) l i

Exempt Change E01-0-93-955' The design change installed an interlock between the Reactor Building Crane Radiation Monitor and the Reactor Building Crane Auxiliary Hoist raise control circuit such that a high radiation 1 alarm stops upward motion of the Auxiliary Hoist. This interlock  !

already existed in the Main Hoist circuit, the change used spare contacts on the existing relays.

The bases for this interlock is to stop upward motion of the crane to mitigate the inadvertent lifting of radioactive material from shielding casks or water to prevent excessively high radiation fields on the Refuel Floor that could contribute to exceeding 10CFR20 exposure limits and violating Regulatory Guide 8.8 (ALARA) requirements.

UFSAR 5ection 12.3.4 and the original specification, J-2532, state that this interlock already exists, however, the interlock ,

was inadvertently omitted from wiring and schematic diagram drawings during initial construction, and was not installed. This change brings the Auxiliary Holst controls into compliance with the UFSAR.

The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question associated with this activity.

Exempt Change E01-0-94-956 LaSalle Station requires the addition of 16 sand filled 55-

{ gallon drums at various locations in the plant, as security barriers. 15 of these drums are polyethylene (plastic) and i i positioned on Elevations 710' and 731' within the Turbine ,

Building. These drums will be free standing, as they are located j within the non-seismic Turbine Building and do not require ,

restraints. i l The one remaining drum is made of steel, and is located on  :

Elevation 710' within the Unit 2 Reactor Building. It is i constrained by chaining to a structural concrete column to prevent if from falling over, since it is in a seismic area. ,

i The following Fire Zones are affected: Units 1 & 2, Elevation l 710 Turbine Building, Fire Zone SC11; Unit 1, Elevation 731 1 Turbine Building, Fire Zone 5B5; Unit 2, Elevation 731 Turbine i

Building, Fire Zone 5B6; and Unit 2, Elevation 710 Reactor l Building, Fire Zone 3G. This zone remains unaffected due to the l use of a steel drum.

Appendix H, Section H.3 and Table H.3-2 of the UFSAR are revised due to increases in the total combustible loadings as a result of using polyethylene drtsm.

A 10CFR50.59 safety evaluation concluded that there was no ,

unreviewed safety question associated with this DCP. l l

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t G. Sussnary of changes to the Facility Which are Described in the safety Analysis Report-(continued)

Exempt Change E01-1-93-916 Temporary System Change (TSC)# l-141-93 was completed in January 1993. This TSC installed a seal-in contact in the closing circuit for valve 1B21-RSCV1. This TSC was initiated in order to eliminate the " hammering effect" the valve was experiencing in the closing direction. The " hammering effect" was occurring in closing the valve, as a result of the torque switch contact relaxation when the valve logic receives a continuous close signal. The TSC has corrected the problem, this design change makes this solution permanent.

The design change implements the design of TSC # l-141-93 with a minor revision. Wired in parallel with the seal-in contact is an IO contact from the limit switch of the valve. The Io contact maintains the throttling capability for the valve in the close direction. When the valve is closing from a full open position, the Io contact is closed with the valve withdrawing from the backseat and will remain so until just prior to the valve full closed position. Just prior to full closed position, the Io contact will open and the torque switch contact will operate to stop the valve. With relaxation of the torque switch contact the valve will not close any further because the seal-in contact is open inhibiting any " hammering effect" to the valve.

A 10CFR50.59 safety evaluation concluded that no unreviewed safety question exists for this design change.

Exempt change E01-2-93-902B This design change revised the 4160/480V transformer taps for the ESF Division 1 unit substations, 235X and 235Y, to boost the secondary voltage by 2.5%. The worl scope consisted of changing the high voltage connections to decrease the primary to secondary transformer turns ratio. Revising the transformer taps boosts the voltages on the 480V buses by approximately 2.5% and, thus, will substantially reduce the number of electrical components with insufficient terminal voltages at the degraded voltage setpoint.

This design change does not change the function, operation, or design basis of the auxiliary power system. In addition, the increase in operating voltage will not decrease the reliability of this system.

This design change does not alter the function of the affected system, result in any unreviewed safety questions, or require a change to the Technical Specification or FSAR. The 10CFR50.59 safety evaluation concluded that no unreviewed safety questions

- exist for this design change.

9 G. Sumanary of Changes to the Facility which are Described in the Safety Analysis Report-(oontinued)

Exempt Change E01-2-93-910 LVDTs (Equipment Part Numbers (EPNs) 2B21-N575A through V) function to give position indication in the control room for Safety Relief Valves EPH # 2B21-F013A through V. For the ease of maintenance, electrical quick disconnects were installed at both the junction box and LVDT ends of the cable for each LVDT.

A 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question associated with this design change.

Exempt Change E01-2-93-937 Motor Operated Valve (MOV) 2HG009 is a normally closed containment isolation valve for the hydrogen reconbiner system.

The exempt change replaced the 2 foot-pound motor with a 5 foot ~

pound motor, operating at the same nominal speed, to improve valve operability. The safety evaluation concluded that there was not an

. unreviewed safety question.

Exempt Change E01-2-93-957A Exempt Change E01-2-93-957D Exempt Change E01-2-93-957B Exempt Change E01-2-93-957E These Exempt Changes drilled 1/4" vent holes in the reactor-side disc of each of the in the 2E12-F042A/B, 2E21-F005, and 2E22-F004 valves. The vent hole eliminates the possibility of the valves becoming pressure bound while in the closed position. This work was proposed in response to various industry and regulatory notices.

The 10CFCR50.59 safety evaluations concluded that there were no unreviewed safety questions associated with these design changes.

Exempt Change E01-2-94-802B This design change revised the logic for the diesel generator 2B reverse power circuit to interlocx a DG-2B output breaker auxiliary contact. This interlock prevents unnecessary tripping of the DG when the generator is not connected to the bus. In this mode reverse power protection is not required. When the breaker is closed, the function of the reverse power prctective circuit remains unchanged. This change also replaced the previous reverse power relay with a GE type GGP-53C, model no. 12GGP53C1A, directional power relay. The previoas GE type GGP-53B relay is obsolete. The new relay performs the same function, and meets or exceeds the requirements of the relay it replaced.

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9. S - ey of changes to the Facility Which are Described in the Safety Analysis Report-(oontinued) 1 Exempt Change E01-2-94-802B (continued)

UFSAR Table 8.3-2 is revised to list the new interlock for the reverse power relay and to change the relay type. However, instead of revising this table to identify the type of the new relay, the proposed change consists of replacing the relay type ,

with a note that refers the reader to the Q-List to obtain the .

relay manufacturer and model number.

A 10CFR50.59 Safety Evaluation concluded that no unreviewed safety question exist.

Exempt Change E01-2-94-803  ;

This-exempt change installed a 1" drain line approximately 3-1/2 feet long with two isolation valves to the Reactor Core Isolation Cooling (RCIC) System Turbine Exhaust Pot 2RIO2B. It also installed 3/4-inch pressure class, socketweld, break flanges in the 3/4" drain piping located immediately downstream of the RCIC System Turbine Exhaust Drain Pot 2RIO2B. A new support was welded to existing support RIO2-2811X.

A 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question.

Exempt Change E01-2-94-809A Exempt Change E01-2-94-809C Exempt Change E01-2-94-809B Exempt Change E01-2-94-809D These modifications replaced the Unit 2 outboard Main Steam Isolation Valve (MSIV) ASCO pilot solenoid valves (ASCO l NP8323A20V) with Valcor Model V70900-87 solenoid valves. This

}' replacement was needed because the ASCO pilot solenoid valves were nearing the end of their qualified service lives. They were replaced with Valcor solenoid valves because the ASCO NP8323A20V pilot solenoid valve is obsolete and is no longer being manufactured by ASCO.

The ASCO NP8323A20V pilot solenoid valve has been attributed to be the cause of failures at LaSalle, Grand Gulf, and Perry. The failure at LaSalle occurred on 12-17-87 and resulted in failure of MSIV 1B21-F028C to close during routine testing. The failure occurred because the ASCO pilot solenoid valve failed to change position when de-energized. ASCO has discontinued manufacturing the NP8323A20V solenoid valve. The Valcor Model V70900-87 was specifically designed to specifications from the R. A. Hiller ,

Company for use as an air pilot control valve for MSIV applications in boiling water reactor plants.

The 10CFR50.59 safety evaluation concluded that no unresolved safety questions exist as a result of these design changes.

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C G. S - =y of chances to the Facility Which are Described in the safety Analysis Report-(oontinued)

Exempt Change E01-2-94-934A Exempt Change E01-2-94-934C Exengt Change E01-2-94-934B Exempt Change E01-2-94-934D These Exempt Changes replaced the Limitorque operators on ,

valves 2E12-F016A/B and 2E12-F017A/B with SMB-1-25 operators. The larger operators are required to increase the available valve operator thrust. This requirement was the result of motor ,

operator design reviews performed in response to NRC Generic Letter 89-10. The yokes on the valves were replaced to support the additional load. The valves' molded case circuit breakers (McCBs) and thermal overload relays were also accordingly replaced.

The net effect of the operator change was to increase the '

valves' stroke time from approximately 75 seconds to approximately 95 seconds. These new stroke times are outside of the standard operating time range for a motor operated gate valve.

A note is being added to UFSAR Table 6.2-21 specifying the '

approximate operating times for the Unit 2 RHR 16A/B and 17A/B valves.

A 10CFR50.59 Safety Evaluation concluded that no unreviewed safety questions exist as a result of these design changes.

3 Exempt Change E01-2-94-938 The design change removed air flow elements and installed nozzles in the Primary Containment KVAC System (VP) duct supplying the head area. This change also eliminated the need to seal the ,

manway access hatches to the drywell head area. The purpose of these changes is to reduce the temperature in the upper head area.

The temperature in this area has resulted in the upset / shutdown reactor water level indication condensing pot and instrument leg boiling off during shutdown when reactor pressure is below about 10 psig. As a result the indicated reactor water level was  ;

actually much lower than indicated.

The low velocity of supply air entering the head area causes stratification, as a result of the size change in the discharge duct fitting increasing from 18" to 24", as it penetrates the head. This results in a 50% reduction in the air velocity. The  ;

nozzles that were added increase the velocity of the air entering the head by a factor of 4.

The increase in VP system pressure due to the nozzles will be offset by the removal of the air flow elements and the manway access hatch covers. The net effect on the VP system air flow rate will be so small as to be immeasurable. There will be no effect on fan horsepower or performance.

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G. Susanary of Changes to the Facility Which are Described in the safety Analysis Report-(continued)

Exempt Change E01-2-94-938 (continued)

UFSAR Section 6.2.1.2 is being revised to reflect this design change. This section addresses the differential pressure across the refueling bulkhead plate following a line break in the head or in the general area of the drywell. The current analysis in the in the UFSAR assumes that the manway access hatch covers in the bulkhead are in place. A new analysis of the addition of the nozzles on the VP supply ductwork to the head area and removal of the manway access hatches (during power operation) will not cause the differential pressure across the refueling bulkhead plate to exceed existing UFSAR limits. A note is being added to the UFSAR to identify the changes made as part of this Exempt Change, and the note will indicate that the original analysis bounds these changes.

A 10 CFR 50.59 Safety Evaluation concluded that no unreviewed safety question exists due the design change.

Exempt Change E01-2-94-939-E This modification improves the reliability and operability of motor operated reactor water cleanup (RWCU) return isolation valve 2G33-F040.

The overall actuator gear ratio (OAR) for 2G33-F040 was changed rom 46.80:1 to 82.00:1. This change increases the available valve actuator thrust to assure that the valve can be repositioned under all design bases conditions. This change increased the valve's 4 stroke time from approximately 21 seconds to approximately 39 l seconds. l I

UFSAR Table 6.2-21 is revised to reflect the new stroke time of I the valve. A 10CFR50.59 safety evaluation concluded that no ]

unreviewed safety question exists as a result of this 1 modification.  !

l Exempt Change E01-2-94-946B Circuit breaker type Klockner-Moeller N2MH6-160/ZM6-63/800 for RPS MG Set B in MCC 236X-2, compartment B3 was replaced with circuit breaker type Klockner-Moeller N2MH6-160/ZM6-100/1200, which has higher thermal and magnetic trip ranges. The circuit breaker was replaced because it was tripping during the MG set start. The replacement circuit breaker allows its thermal and magnetic trips to be set at a higher value and their new settings will prevent nuisance tripping of the circuit breaker during MG set start.

The safety evaluation concluded that there was no unreviewed safety question.

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i G. Susumary of changes to the Facility Which are Described in the Safety Analysis Report-(continued)

Exempt Change E01-2-94-968G This design change increased the thrust that the actuator delivers to the stem of valve 2E12-F049B by replacing the 2ft-lbf motor with a 5 ft-lbf motor. The new motor operates at the same nominal speed as the old retor (1800 RPM).

NRC Generic Letter 89-10 requires that nuclear plant licensees evaluate the ability of certain motor operated valves to be repositioned when subjected to design basis conditions of flow and differential pressure. The results of the evaluation for valve 2E12-F049B indicate that additional actuator thrust is required to ensure the valve's reliability under design basis flow and differential pressure conditions.

The safety evaluation concluded that there was not an unreviewed safety question associated with this design change.

Exempt Change E01-2-94-987 This Exempt Change replaced the flex-wedge gate valve for MOV 2B21-F016 with a double disc parallel slide gate valve manufactured by the Anchor / Darling Valve Company. The valve was replaced because the hardfacing on the valve seats was almost completely removed by extensive machining. The machining was required to refinish the seats so that local leak rate test (LLRT) acceptance criteria could be met. This valve has frequently failed to meet its LLRT acceptance criteria in the past. The new valve is expected to provide better LLRT performance, while requiring less maintenance than the previous valve.

A vent hole was put in the new valve's reactor side disc. The intent of the vent hole is to eliminate the potential for thermal binding or pressure locking of the valve discs. No provisions were made for this in the valve's procurement specification. With the vent hole installed, the valve will still seal effectively because the vented side of the valve is always exposed to a higher pressure. The higher pressure on the back side of the non-vented valve disc will force the disc into it's seat and provide flow shut-off.

The valve's live loaded spring packages were also changed under the scope of this design change, at the request of the Mechanical Maintenance Department. They have more confidence in the leak tightness capability of the Garlock Set over the live loaded spring packs.

C G. Susanary of Changes to the Facility Which are Described in the Safety Analysis Report-(continued)

Exempt Change E01-2-9500001 In the previous design, the SSPVs supported one end of the scram air header. This design change installed separate supports to secure the scram air header for 14 banks of hydraulic control units (HCUs). Previously installed copper tubing was replaced with flexible tubing. This eliminated the situation where SSPVs supported one end of the scram air header. This was a change to the air lines providing the source of air from to and from the SSPVs, not the source of the air. The function of the SSPVs n'nd the HCUs was not affected by this design change. UFSAR Appendix '

H, Fire Hazards Analysis, is being revised to reflect the change in combustible loading resulting from the change.

The 10CFR50.59 safety evaluation concluded that there was no i unreviewed question.

Exempt Change E01-2-9500120 The minor plant change added a test tap on line 2FC11C-10",

located between primary containment isolation valves 2FC086 and 2FC 115. The test tap was added toallow the station to obtain consistent local leakage rate by only including these two valves in the test boundary. The location of the previously used test tap resulted in several valves being included in the test boundary.

UFSAR Volume XI Piping & Instrumentation Diagram (P&ID) drawing M-144, sheet 1, is revised. Table 6.2-21 is revised to reflect the new containment valve arrangement for the Unit 2 FC System Valves 2FC086 and 2FC115, including the test tap. Figure 6.2-31 (sheet 10c) is revised to reflect the new Containment Valve Arrangement " Detail AD". UFSAR section 6.2.4.2.4 is revised to add a brief description of the containment valve arrangement for the Reactor Well Bulkhead drain piping valves.

The 10CFR 50.59 safety evaluation concluded that there is no unreviewed safety question associated with the design change.

Exempt change E01-2-9500158 The exempt change removed valve 2E51-F091 (RHR Steam Condensing Mode Warming Valve) and associated 1" piping up to a 10" header.

The electrical power and controls associated with Valve 2E51-F091 were also removed. The one inch line and the one inch valve (2E51-F091) provided a bypass flow path around 10" valve 2E51-064 to serve as the steam warming line prior to initiation of the RHR steam condensing mode and is not required for any operating mode.

The reason of this change was that the valve 2E51-F091 did not reseat during testing. The steam condensing line was capped near its connections to process header lines 2RIOLA-10" and 2RI41A-10".

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S. s======y of Ams to the Facility Which are Described in the ,

safety Analysis memort-(oontinued)

Exempt Change E01-2-9500158 (continued) ,

The RHR Steam Condensing has been eliminated as a mode of operation of RHR system, therefore, removal of warming line will not affect plant operation. The 10" header can still be used for ,

alternate decay heat removal operation. i LaSalle Updated Final Safety Analysis Report (UFSAR) Section 1.2.2.3.4, and Tables 6.2-21 (Sheet 1) were revised to reflect the design change, and incorporated in new Table 6.2-28 (Sheet 2).

The 10CFR50.59 safety evaluation concluded that there is no unreviewed safety question associated with this design change.

Minor Plant Change P01-1-92-524 The minor change removed the spring and disk from check valve 1DG036. The valve is downstream of the Unit 1 LPCS pump motor cooler cooling water line. The valve body was left installed in the piping system. The design change was in response to valve corrosion concerns. The disk and spring were replaced during LlR04, due to' corrosion discovered during an inspection. The permanent removal of the valve spring and disk removes the funtional capability of 1DG036 to perform as a check valve. The design change eliminates future corrosion concerns associated with '

the valve. An engineering review determined that the cheke valve is not required in the piping system.

The 10CFR50.59 safety evaluation concluded that no unreviewed safety questions exist.

Minor Plant Change P01-1-93-501 The minor change allows the 1B Diesel Generator (DG) to be shutdown using an Energize-to-shutdown solenoid configuration, as opposed to the previous Energize-to-run configuration. The change also revises the shutdown control logic of the 1B DG which

! previously allowed the DG to be shutdown with either the Manual Control Switch /Pushbutton or Emergency Stop Pushbutton, whether a LOCA signal is present or not. The revised control logic allows the 1B DG to be shutdown with the Manual Control Switch /Pushbutton only when a LOCA signal is not present and would leave the '

Dmergency Stop Pushbutton as currently configured. This change makes the 1B DG consistent with the logic currently configured on DG 0, lA, and 2A. 1 The 10CFR50.59 safety evaluation concluded that no unreviewed safety questions exist.

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l I S. Basmary of channes to the Facility Which are Described in the  !

Safety Analysis Report-(oontinued) i i

Minor Plant Change P01-2-92-506 i This design change replaced the motor operator on valve 2E51-F013, in order to increase the thrust available to operate the j

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valve. The minor plant change addresses concerns raised by NRC _j Generic Letter 89-10 and related documents. The Limitorque SMB l 60 operator replaces the previous SMB-00-25 operator.

l 1

An addendum to this design change consisted of drilling a 1/4" '

vent hole in the valve's reactor side disc. The vent hole is I intended to eliminate the potential for pressure binding.

Pressure binding may occur in the event that the valve is closed while the bonnet is full of relatively cool water and is  :

subsequently heated.  !

s The 10CFCR50.59 safety evaluations concluded that there were no l

]

unreviewed safety questions associated with this design. I i  :

i i Minor Plant Change P01-2-92-523 l

The minor change removed the spring and disk from check valve i 2DG036. The valve is downstream of the Unit 2 LPCS pump motor  !

, cooler cooling water line. The valve body was left installed in  ;

the piping system. The design change was in response to valve l:

corrosion concerns. The disk and spring were replaced during

L2R04, due to corrosion discovered during an inspection. The  !

! permanent removal of the valve spring and disk removes the l functional capability of 2DG036 to perform as a check valve. The {

i design change eliminates future corrosion concerns associated with l the valve. An engineering review determined that the check valve j i is not required in the piping system. l 1

The 10CFR50.59 safety evaluation concluded that no unreviewed i

safety questions exist. l l i 4

[ Component Replacements C01-2-94-032 and C01-2-94-033 l These design changes replaced relays 2B2LA-K710AX and 2B21A- l' K710CX, and 2B2LA-K710BX and 2B21A-K710DX, respectively. The l- Agastat model TR-14B3A replacement time delay relays that replaced j the Agastat model EGPB control relays were purchased as commercial j grade and upgraded for safety related use. j i These design changes do not change the function, operation, or >

i design basis of the Auxiliary Power System, RCIC System, or MSIV l Leakage Control System as described in the Technical i Specifications or UFSAR. ,

i

=- UFSAR Section 7.4.1.2.3 is being revised to reflect that the l time delay is now applicable to both units. The current revision [

of the UFSAR reflects the time delay as being applicable to Unit 1  :

4, only. l i

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G. Sussnary of Ams to the Facility Which are Described in the safety Analysis Report-(oontinued)

Setpoint Changes 501-1-92-024, 501-1-92-025 and S01-2-92-030 The design changes lowered the setpoint for these parameters to more realistic values to incorporate NRR approved Technical Specification amendments 105(Unit 1) and 91(Unit 2). The previous setpoints were found to be higher than needed for verifying that the systems were full of water. For the HPCS and RCIC systems, the setpoints resulted in unnecessary activation of the alarms when associated water leg pumps are aligned to take suction from the suppression pool.

These changes did not affect the UFSAR since neither the functions nor descriptions of the ECCS and RCIC discharge line

" keep filled" alarm instrumentation channels as described in that document were changed.

The 10CTR50.59 safety evaluations concluded that there were no unreviewed safety questions associated with these design changes.

Setpoint Change 301-1-94-036 UFSAR Section 11.2.2 was revised to delete reference to the radwaste domineralizer efflvent conductivity. This allows future demineralizer effluent conductivity monitors setpoint changes to be made without having to revise the UFSAR. Changes are required to UFSAR pages 11.2-9 and 11.2-10.

Setpoint Changes 501-1-94-038 and S01-2-94-037 These setpoint changes increased the calibration trip setpoint for Reactor Core Isolation Core turbine exhaust pressure switches 1E51-N009A and B and 2E51-N009A and B, respectively, from 25.0 psig to 43.7 psig plus 2.3 psig head correction, increasing.

These setpoint changes did not require the replacement of the pressure switches since the new setpoints are within their adjustable range. These changes also did not involve any hardware, wiring, or cable changes.

The purpose of these setpoint changes is to provide sufficient margin between the calibration setpoint and the postulated RCIC turbine backpressure of 24.5 psig at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 15 minutes following a Station Blackout (SBO) to account for all known instrument errors. Another purpose is to increase the availability of RCIC following a LOCA by increasing the setpoint as close as possible to the upper analytical limit which was determined to be 50.0 psig.

UFSAR Table 7.4-1, and Sections 15.9.3 and 15.9.4 are being revised to reflect design changes. The 10CFR50.59 safety evaluation concluded that no unreviewed safety questions exist.

G. Summanry of changes to the Faoility which are Described in the Safety Analysis Report-(oontinued)

Setpoint Changes S01-1-9400446 and 301-2-9400447 The lowering of the interlock allows the RR pumps to be upshifted or downshifted at a lower power. This may result in a slight increase in time in which the RR pumps are in fast speed and the FCV in a reduced position (below 70 Mlb/hr) . While this is not a cavitation concern, this is a known point for increased RR pump vibrations.

By allowing RR pump speed changes at lower feedwater flows, the margin from the region of instability can be increased during normal plant evolutions. Since RR pump speed changes can be performed at a lower Flow control Line (FCL), the overall power change during the evolution can be reduced. Thus, the transient on the plant, (feedwater control, heater response, etc.) is reduced.

• UFSAR Sections G.2.1, G.2.3, G.2.5, G.3.3, G.5.2 and G.7 were revised to reflect the setpoint changes. The safety evaluations concluded that there were no unreviewed safety questions.

Setpoint Change 501-1-92-026 Setpoint change S01-2-92-032 Setpoint change 501-1-92-027 Setpoint Change S01-2-92-033 Setpoint Change 301-1-92-028 Setpoint Change 301-2-92-034 Setpoint Change 301-1-92-029 Setpoint Change S01-2-92-035 Setpoint Change 301-2-92-031 These design changes lowered the setpoint for the ECCS and RCIC ,

Discharge Line " Keep Filled" Pressure Alarm parameters to more realistic values to incorporate NRR approved Technical Specification amendments 105 (unit 1) and 91 (Unit 2).

The previous setpoints were found to be higher than needed for verifying that the systems were full of water. For the LPCI RCIC, HPCS, LPCS, systems, the setpoints resulted in unnecessary activation of the alarms when associated water leg pumps are aligned to take suction from the suppression pool.

These changes did not affect the UFSAR since neither the functions nor descriptions of the ECCS and RCIC discharge line

" keep filled" alarm instrumentation channels as described in that a document were changed.

The 10CFR50.59 safety evaluations concluded that there were no unreviewed safety questions associated with these design changes.

3 i 4

s N 1 S. Sususasy of Chameos to the Facility Which are Described La the  !

Safety Analysis Report-(oontinued)  !

Temporary System Change (TSC) 1-0033-95 l l'

The Unit 1 A Reactor Recirculation (RR) Flow Control Valve-(FCV)' actuator drain alarm was up solid. This situation was a  ;

nuisance, and masked other valid alarms. The TSC bypassed the ,

' control room alaon allowing other hydraulic power unit alane signals to annunciate in the control room.

Actuator drainage was verified by system engineers weekly

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l rounds and by periodically performing LOS-RR-SR3, Reactor Recire l FCV Actuator Leakoff Line Flow Rate Test.

• The safety evaluation concluded that there was not an unreviewed safety question.  ;

Temporary System Change (TSC) 2-0017-95 The TSC consisted of adding test equipment to monitor  !'

! Suppression Pool temperature element loop 2TE-CM057A-2. This was i- done to determine the cause of erratic output from element 2.

Non-safety related and non-EQ test equipment was connected to e safety related and EQ components. System electrical segregation

! was not affected because an isolation transformer was used to {

. power the test equipment.  !

1 The safety evaluation concluded that there was no unreviewed l safety question associated with the TSC.

l Temporary System Change (TSC) 2-0062-95 L i 1 This Temporary System Change allowed fire detection zones 2-16,

' 2-16P, 2-30, 2-31, 2-32, 2-33 to be taken OOS using switch jumpers

- which are operated by using keys controlled by the Operating l

]- department. Detectors in above areas were taken OOS whenever i welding, cutting or grinding was performed in these areas to l' prevent false alarms at the fire protection panel 2FPO4JA. The

position of these switches was controlled administrative 1y via the i equipment out of service procedure LAP-900-4, and a fire impairment was written in accordance with LAP-900-16 whenever a detection zone was taken OOS. The fire impairment specified and established required fire watches and Out of Services for affected fire zones.

The operating modes of subject detection zones were controlled administratively via procedure LAP-900-4 (Equipment Out Of Service) and LAP-900-16 (Fire Protection Impairment).

The safety evaluation concluded that there was no unreviewed safety question associated with this change.

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s. sussnary of ^ r:s to the Facility Which are Described in the Safety Analysis Report-(continued)

Temporary Alteration (TALT) 2-186-S5 I i

This Temporary Alteration was initiated to bypass the alarms when they were up solid. This was due to the alarms masking other i alarms. The TALT was accomplished by installing a jumper across l flow switch 2B33-N097 ("A" alarm) and 2833-N908 ("B" alarm). The jumpers were installed as needed whenever an alarm was.up solid.

This TALT allows other alarm signals to annunciate in the contrcl -

room.

Actuator drainage, which occurs slowly over time, was monitored by system engineer and operating personnel. The unit would be shutdown before this degradation affected control of the FCV. ,

Degradation resulting in unreliable FCV control would require i that the unit be shut down, to allow the actuator and switch to be repaired.

1 The 10CFR50.59 safety evaluation concluded that there were no unreviewed safety questions associated with the TALT. p LaSalle County Station Units 1 & 2 Reclassification of components Associated With Residual Heat Removal Steam Condensing Mode of Operation LaSalle County Station has eliminated the Steam condensing Mode of Operation of the Residual Heat Removal (RHR) system. As a result of this elimination, several components previously operated

• only for the Steam condensing Mode have now become passive in their safety function or non-safety related. ,

t Passive valves do not require quarterly stroke time testing in accordance with the American Society of Mechanical Engineers (ASME)Section XI In-Service Testing (IST) Program, and components no longer important to safety can be removed from the 10CFR50.49 j Environmental Qualification (EQ) Program, thereby, removing stringent procurement and maintenance requirements.

By formally changing the Safety Classification of these components, the cost of procurement and maintenance for these components can be dramatically reduced, resulting in significant savings to Commonwealth Edison Company. Associated RHR Steam condensing Mode Motor Operated Valves (MOV) were removed from the i Generic Letter (GL) 89-10 MOV Testing Program.  ;

UFSAR Sections 5.2, 3.4, and Appendix L is being revised to e reflect the reclassification of components. L i

The safety evaluation concluded that there was no unreviewed.

safety question associated with the reclassification, j i

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G. Summary of Changes to the Facility which are Described in the Nafety Analysis Report-(oontinued)

UFSAR Revision to Reflect the New Edition of 10CFR20 and the Facility Improvement Project (FIP)

UFSAR Chapter 12, Radiation Protection, and Appendix I of the UFSAR are being revised to incorporate the current revision of 10CTR20, and the Facility Improvement Project. The update of Chapter 12 also reflects the present structure of both the Radiation Protection and Chemistry Departments.

The safety evaluation concluded that there was no unreviewed safety question associated with this change to the UFSAR.

Revision to UFSAR Figure 5.4-3, Reactor Core Isolation Cooling System Process Diagram UFSAR Figure 5.4-3 Revision 9 - April 1993 improperly shows the E51-F064 valve in series with the reactor core isolation system (RI, RCIC) turbine and fails to show the E51-F008 valve. The change will make this diagram agree with the controlled station drawings and the actual configuration.

This change will correct this UFSAR figure to properly match the as designed, as built configuration of this system. No physical plant changes are occurring. No procedures or analyses are changed, since the existing controlled P & ID's, other UFSAR sections and drawings accurately depict the system configuration.

The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question associated with this activity.

Revision to UFSAR Section 5.4.8.2 and Figure 5.4-7 Delete Table III aua V (a list of operating parameters) and the referenced notes fre- JFSAR Figure 5.4-7. Replace with a note to Figure 5.4-7, "For Reactor Water Cleanup System operating values, see current station procedures.". This revision permits changes to the air and water flow rates during a RWCU filter backwash or precoat, without the administrative burden associated with updating the UFSAR. In section 5.4.8.2 (System Description), the reference to Table 5.4-2 is delete, since there is no Table 5.4-2.

The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question associated with this activity.

O a

G. Susseary of Changes to t's,g Faellity which are Described in the Safety Analysis Repor'.9 g inued)

UFSAR tr . e for fixed air sampling-panels and sampling areas.

UFSAR Section 12.3.4.4 and Table 12.3-16 is revised to reflect actual plant configuration of continuous air monitors (CAMS) sample points and locations.

The safety evaluation concluded that there was no unreviewed safety question.

Replacerant of Valve lE51-F357 in Reactor Core Isolation Cooli'ng (RCIC) Steam Drain lint with Blind Flanges and Section of Piping.

Drain line 1RIO9A-2" from the RCIC Steam Supplf line (line no. .

1MS06A-10") contained manual valve lE51-F357. This valve was normally closed, and indicated as locked closed on the piping and instrumentation drawing. This change removed this valve from the drain line, and installed a short section of pipe with blind ,

flanges welded at each end in its place.

UFSAR Table 6.2-21, " Summary of Lines Penetrating the Primary containment", was revised to reflect this activity.

The safety evaluation concluded that there was no unreviewed

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safety question associated with the change.

4 UFSAR update to Section 9.3.5 to reflect valve setpoint change The standby liquid control (SBLC) Pump discharge relief valves, 1(2)C41-F029A/B, are tested each refuel outage to meet Technical Specification Surveillance requirements for the SBLC system. The valves were previously required to be set and verified at less than or equal to 1400 psig and to verify that the relief valves do not actuate during recirculation to the test tank.

An investigation determined an inconsistency between the Vendor setpoint tolerance to that which was previously specified in the station procedure for testing and adjusting the relief valve setpoint.

< A setpoint which is 75 psi above the pressure that is required for SBLC Pump injection was the required criteria. SBLC is required to supply 41.2 gpm at 1220 psig to meet its functional requirements. General Electric recommended that the setpoint be greater than 1295 psig.

The Vendor (Crosby) stated that it was acceptable to adjust the valve relief setting within the range of 1260 psig to 1499 psig with the current spring in each relief valve.

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G. S====*y of Changes to the Facility Which are Described in the safety Analysis Report-(oontinued) 6 UFSAR update to Section 9.3.5 to reflect valve setpoint change (continued)

The relief valve setpoint was changed to 1340 i 40 psig, which ,

provides margin to the upper and lower limits when testing the  !

relief valves.

UFSAR Section 9.3.5 was revised to reflect the SBLC pump i discharge relief valves 1(2)C41-F029A/B setpoints being changed '

from 1400 psig to 1340 i 40 psig.

The 10CFR50.59 safety evaluation concluded that there was no  ;

unreviewed safety question.

UFSAR update to reflect the deletion of the Rod Sequence Control System (RSCS)

The RSCS is no longer in Technical Specifications as a result of the Rod Worth Minimizer (RWM) Low Power Setpoint (LPSP) being lowered to 10% rated core thermal power. UFSAR Section 7.2.1 is revised to reflect the deletion of the reference to RSCS.

The 10CFR50.59 safety evaluation concluded that there is no unreviewed safety evaluation associated with the procedure revision.

Non-Fire Rated Penetrations in the Fire Barrier Separating Fire Zones SB2 and 3K. ,

The existing fire rated penetrations in the boundary wall between fire zones 5B2 and 3K at elevation 687 feet-0 inch to the floor slab of elevation 710 feet-6 inch, and from column lir.es 18.7 to 21 are reclassified as non-fire rated penetrations. -

The boundary wall which separates fire zones SB2 and 3K at elevation 687 feet - 0 inch to the floor slab of elevation 710 feet - 6 inch does not separate fire zones containing redundant  ;

safe shutdown equipment.

The Safe Shutdown Analysis (UFSAR Appendix H, Section H.4) demonstrates that for a fire in any single plant fire area /ztene, at least one metht;d exists to achieve and maintain a safe shatdown condition independent of that fire area / zone.

f The safety evaluation concluded that there was not an unreviewed safety question.

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r S. Susmeary of changes to the Facility Which are Described in the safety Analysis Report-(oontinued)

Evaluation of Turbine Building Flooding An Operability Assessment and Safety Evaluation was prepared in response to a flooding concern associated with the potential failure of circulating water and service water piping, which could result in a gravity fed flood from the perched cooling lake, installed outside of the flood protected areas as detailed in the SAR.

Calculations determined that piping with no isolation valves in the flood protected zones had no postulated leakage cracks and based on bounding flood rates, sufficient time existed for manual actions to isolate any of the identified lines using valves installed outside of the flood affected area. Therefore, it was concluded that the identified piping installed outside of the flood protected areas does not result in an operability concern.

The safety evaluation concluded that there are no unreviewed safety questions involving the current piping configuration, based on the analytical qualification of the major piping to show that a moderate energy line break is not a credible event, and, even if a break was postulated, a sufficient time frame exists to take actions to isolate the affected piping from the lake prior to the flood challenging any equipment important to safety.

MCPR/EOOS Bases Change, Revision of the Administrative Technical Requirements to reflect the U2C7 Reload and the Implementation of ARTS and EOOS Revisions to the UFSAR.

Technical Specifications Bases section 3/4.2.3 (Minimum Critical Power Ratio) was changed to delete specific information regarding LaSalle's analyzed EOOS combinations. Instead, the Core Operating Limits Reports, which contain the needed information is referenced.

The Administrative Technical Requirements were revised to incorporate the power and flow dependent ARTS thermal limits and reference a new addition to LaSa11e's EOOS analyses.

Sections 4.4.1.3, 7.3, 7.6, 15.A.5 and 15.A.6 of the UFSAR were revised to describe the new ARTS power and flow dependent thermal limits and the addi+. ion to LaSalle's EOOS analyses described above.

TS Bases Section 3/4.2.3 was changed to update the Bases with regards to currently allowed EOOs. The Administrative Technical Requirements (ARTS) were revised due to approval of the ARTS amendment by the USNRC and due to L2C7 startup. The UFSAR was changed to reflect the ARTS amendment and the new EOOS combinations allowed.

The safety evaluation concluded that there was not an unreviewed safety question associated with these activities.

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e G. Sumanary of Changes to the Facility Which are Described in the Safety Analysis Report-(continued)

Relocation of Technical Specifications Table 3.6.3-1,

" Primary Containment Isolation Valves" to the UFSAR Technical Specification Amendments 102 and 87 (for Unit 1 and 2, respectively), deleted Table 3.6.3-1, " Primary Containment Isolation Valves". Specification 3/4.6.3 was also changed by the amendments to refer to the required components (valves) by description instead of by use of a list of valves. The Unit 1 and 2 tables were added to Administrative Technical Requirements (ATRs) to provide a ready reference of the components / devices that are covered by Specification 3/4.6.3. ATR Table 3.6.3-1 (Unit 1) and Table 3.6.3-2 (Unit 2) have minor format / editorial changes, which do not affect the content of the tables. Additional valves are added from UFSAR Table 6.2-21 to nake the list of components required to be operable per Technical Specification 3/4.6.3 complete. The new ATR tables are also combined and added to the UFSAR as Table 6.2-28.

The safety evaluation concluded that there was no unreviewed safety question.

Drywell Floor Drain and Equipment Drain Sumps - Operability Assessment An assessment was performed in response to discovery of the cover plates for the drywell floor drain and equipment drain sump in the pedestal area being unfastened. The design drawings indicate that these plates should be bolted in place and the joints sealed with caulking material.

A calculation performed to evaluate the response of the unfastened plates determined that these plates will not significantly move during the occurrence of any SAR accident scenario. The unbolted condition was found to be acceptable.

A 10CFR50.59 Safety Evaluation concluded that there are no unreviewed safety questions involving the sump cover plate configuration.

Update of the functional requirements of the alternate rod insertion (ARI) system in the UFSAR The time allowed for the start of the insertion of all control rods after receipt of an initiation signal was increased from 15 seconds to 35 seconds; the maximum time delay between the receipt of an initiation signal and the time when all control rods reach the full-in position was increased from 25 seconds to 45 seconds.

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b G. Sussnary of Chances to the Facility Which are Described in the

• safety Analysis Report-(oontinued) ,

Update of the functional requirements of the alternate rod insertion (ARI) system in the UFSAR l (continued)

The time requirements for the ARI system are based upon the  ;

time needed to fill the scram discharge volume. The original times  :

proposed for the start of motion (15 seconds) and the time to full insertion (25 seconds) are very conservative. Further analysis has determined that the time for all control rods to start inserting should be 35 seconds, while the time for all control rods to be fully inserted should be 45 seconds.

When ARI was installed, a generic evaluation was performed L which assumed the volume of the scram discharge piping would be the same for all BWRs. LaSalle's scram discharge volume is larger than that assumed in the evaluation. Therefore, the maximum t allowable time delay between receipt of an initiation signal and  !

when all control rods begin to insert is to be changed. Also, the maximum allowable time between the initiation of ARI and all control rods reaching their full-in position is changed.

LaSalle's UFSAR previously stated that for the ARI function, the control rods must start to insert within 15 seconds from the receipt of the initiation signal. It further stated that all control rods must be fully inserted within 25 seconds from the receipt of the initiation signal. These values are changed to 35 seconds and 45 seconds, respectively. UFSAR Section 7.6.5 is revised to reflect the updated functional requirements of ARI. 3 This change is based upon a realistic analysis of the time  ;

required to fill the scram discharge volume. Interaction between the ARI system and the control rod drive system will remain unchanged. The ARI system will continue to provide a means of ndtigating an Anticipated Transient Without Scram (ATWS) .

The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question associated with this change.

Bypassed Reactor Water Cleanup (RWCU) Filter Inlet Temperature Element, 2TE-G33-N007  !

The RWCU Filter Inlet Temperature Element was removed for repairs. The RWCU Filter Inlet Temperature Hi-Hi function was bypassed because of this. An operability assessment was performed to allow for start up of Unit 2 with this function bypassed.

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G. Sumunary of Changes to the Facility Which are Described in the safety Analysis Report-(oontinued)

Bypassed Reactor Water Cleanup (RWCU) Filter Inlet Temperature Element, 2TE-G33-N007 (continued)

Element 2TE-G33-N007 measures the temperature of the water entering the filter demineralizers. The RWCU filter inlet temperature can be monitored by the main control room operator from temperature indicator 2G33-R607, located on panel 2H13-P602.

This instrument provides indication only. As a compensatory action, a caution card was placed on annunciator vindow 2H13-P602, A303, "RWCU Filter Inlet Temp Hi-Hi" during the bypassing of the element.

The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question.

Reactor Vessel Bottom Head Drain Line Effects on LOCA Analyses An evaluation was performed in response to discovering that per operating procedures, the reactor vessel bottom head drain valve G33-F101, is normally open, while the UFSAR shows the G33-F101 valve as being closed during routine operations.

This evaluation consisted of determining if this placed the station in an unanalyzed condition. The evaluation concluded that this situation was acceptable, based on the water loss through the bottom head drain line with valves G33-F101 and G33-F103 open has minimal impact, and is bounded by analysis.

UFSAR Figure 5.4-6, sheets 1 and 2 are being revised to document the operating mode for motor operated valve G33-F101 and the fluid flow rates through the associated bottom head drain line.

UFSAR Figure 6.2-1 is being revised to incorporate the reactor bottom head drain line configuration.

A 10CFR50.59 safety evaluation concluded that there is no unreviewed safety question associated with this issue.

UFSAR Revisions per impending Diesel Generator Design Changes Component Replacement C01-2-94-006-E will replace the Basler type PRC-130 differential relay with a Westinghouse type SA-1 differential relay in Diesel Generator (DG) 2A.

Component Replacements CR 94-006A (DG-2A), CR 94-007A (DG-0),

and CR 94-007B (DG-1A) will replace the underfrequency trip relays 1481-DG007, 1481-DG014, and 2481-DG014 in the DG-0, lA, and 2A shutdown control circuits.

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G. Susanary of Changes to the Far.llity Which are Described in the safety Analysis Report-(co,ap nued)

UFSAR Resisions per impending Diesel Gentrator Design Changes (continued)

Component Replacement No. C01-2-94-006D will replace the Diesel Generator 2A Overcurrent Basler type PRC-110-108 relays (which are obsolete) withABB/ Westinghouse type CO-6, model no. 264C898A05 relays.

Exempt Change E01-2-94-060E will replace the Diesel Generator 2A Neutral Ground Relay, Basler type PRV-120 overvoltage relay (which is obsolete), with GE type IAV, model no. 12IAV51DLA.

Exempt Change E01-2-94-960F will replace the Diesel Generator 2A Basler type PRP-110 Reverse Power Relay (which is obsolete),

with a GE type ICW, model no. 12ICW51A2A, directional power relay.

UFSAR Table 8.3-2 specifies the type of differential relay used for DG-2A. However, instead of revising this table to identify the type of replacement relays to be installed, the table will be revised by replacing the relay type with a note that refers the reader to the Q-List to obtain the relay manufacturer and model number.

The 10CFR50.59 safety evaluations concluded that there were no unreviewed safety questions.

Continued operations with the high pressure steam supply not available to one or both Turbine Driven Reactor Feed Pumps (TDRFP)

Continued operations with the high pressure steam supply not available to either TDRFP due to the unavailability of the TDRFP high pressure steam supply was evaluated. This was due to operational considerations which may require that this supply be isolated.

While unavailability of the high pressure steam supply limits the maximum flow available of the TDRFP, the TDRFP can meet all design requirements, other than being able to support plant startup.

The unavailability of this source does not affect adjacent equipment. The unavailability of this source doesn't affect the control system. The logic inputs for the steam supplies are parallel such that either must be available and the steam supply motor operated valves (MOVs) do not have logic inputs to feedwater control.

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G. Summaary of changes to the Facility Which are Described in the safety Analysis Report-(oontinued)

Continued operations with the high pressure steam supply not available to one or both Turbine Driven Reactor Feed Pumps (TDRFP)

(continued)

As discussed in UFSAR Section 10.3, this source is not required for the feedwater system to meet its design basis, except for the option of being able to start up on a TDRFP. The system as described in UFSAR sections 10.3 and 10.4 has the high pressure steam supply available for operation of a TDRFP prior to the main turbine being place on line. There are no procedural actions are affected by the unavailability of this steam source, since the use of the TDRFP high pressure steam supply for plant startup is not addressed in station operating procedures.

The safety evaluation concluded that there were no unreviewed safety quesitons associated with the evaluation.

LaSalle Station Dry Active Waste (DAW)

Storage Facility for Low Level DAW The evaluation was performed to evaluate the storage of DAW in an on-site warehouse building. The issues of container strength and durability, pallet strength and durability, direct and skyshine radiation levels at the nearest restricted area boundary and the nearest off-site receptor, fire protection, and accident scenarios were evaluated to determine the acceptability of an on-site DAW storage facility.

The evaluation concluded that an unreviewed safety does not exist.

Special Operating order 95-29, Provision for Locking the Position of the Reactor Recirculation (RR) Flow Control Valves (FCV)

The operating order provided instructions for dealing with the possibility of the 2A RR FCV oscillating while moving in the open direction, such as during an increase in reactor power. The order allowed for locking the position of the FCV if valve oscillations were to occur.

Locking the FCV position gives operators a means to prevent reactor power spikes which could be caused by FCV position instability. Other instructions in the order are controlled in accordance with station operating procedures.

The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question.

l l

G. Susanary of Changes to the Facility Which are Described in the safety Analysis Report-(oontinued) ,

Unit 1 Reactor Recirculation (RR) delta temperature (dT) Resistance Temperature Detector (RTD) failure on 9/27/95, RTD element #2 of Reactor Recirculation (RR) pump 1B suction temperature instrument 1B33-N028B was found to have failed. The element is part of the RR delta temperature downshift logic, and in its failed condition would prevent the IB 10.1 dT logic from tripping both RR pumps to slow speed. The circuit is arranged such that both channels of the B RR suction dT logic must sense a 10.1 F dT to cause a downshift. The failed RTD does not feed the dT indicator in the Auxiliary Electric Equipment Room (AEER), therefore the B RR dT indication remains valid. The redundant trip system based on the A RR suction temperature remains available to downshift both RR pumps if the dT reaches 10.1 F. Based on this condition. The following recommendations ,

were made The unit may be placed in power operation with the pumps in high speed without restrictions; The circuit should be lef t in the normal, unbypassed condition; and, once every two weeks the proper function of the 1B33-N028A element #2 RTD which does not feed the dT indication be verified by measuring the resistance of the RTD.

The 10CFR50.59 safety evaluation concluded that there was no unreviewed safety question associated with either the RTD failure or power operation of the unit with the pumps in high speed without restrictions.

Revision to UFSAR Section 7.8 UFSAR Section 7.8 is revised to reflect an update to the Safety Paramater Display System (SPDS) shown in Figure 7.8-1, SPDS Primary Display. This is due the installation of a new process computer graphics system. There is also revision to the text of Section 7.8.2.2 associated with this. The 10CFR50.59 safety evaluation concluded that there is no unreviewed safety question associated with the change.

Changes to UFSAR Sections 12.5.3, 13.5, and Appendix B to reflect Revision 0 of the LaSalle Procedures and Surveillance Writers Guide UFSAR Sections 12.5, 13.5, Tables 13.5-1, 13.5-2, and 13.5-4, and Appendix B are revised to reflect the publishing of the Writers Guide.

The 10CFR50.59 safety evaluation concluded that there is no unreviewed safety question associated with this activity.

4

y e

o I. Survey of Evaluation Results of Chlorine Shipments by Barge on the Illinois River This survey was not required to be performed for 1995, the survey was last completed and is included with the NRC Annual Report for the year of 1994.

I. Susanary of Events Violating Technical Specification 3.4.5 Primary coolant Iodine Spiking Exceeding Allowable Timits During this reporting period, January 1, 1995 through December 31, 1995, there were no violations of Technical Specfication 3.4.5, i Primary Coolant Iodine Spikes Exceeding Allowable Limits.

f J

l 4

ATTAcadENT A

• SAFETY-RELATED MAINTENANCE CCMPLETED (NON-GUTAGE MLATED) i N0sa 350 0 STS EgpIPMENT DESCRIPTION

]

950038533-01 AF 1AF000 INSFECT ERA WCEOUT BELAIS.

950038535-01 AF 1AF000 INSPECT EEA LOCEDUT arr.avs, 950038538-01 AF 1AF000 INSPECT EEA McEOUT RELAYS.

4 950038539-01 AF 1AF000 INSPECT EEA MCIOUT RELAIS.

950038540-01 AF 1AF000 INSPECT EEA McEOUT mar.avs, 930046982-05 AF 1AF04E-3 INSTALL BAMANA JACES.

i 930047382-01 AF 1AP20E-203C INSPECT mar m a.

I 950112872-01 AF 1AP21E-3035 MPLACE DEFECTIVE 3088-9 UNIT.

950063509-01 AF 1AP21E-303D REPLACE m u m m ,

f 950063513-01 AF 1AP22E-404A REPLACE m m m, i j 940050090-01 AF 1AF75E-C6 MPLACE OVERMAD MSET PUSE BUTTON.

1 950117499-01 AF 1AP76E-D5 RCIC MkTER LEG POMP TRIPS maa m m.

$ 950010769-02 AF 1AF785-D5 PERPCsef MCC CUBICLE INSFECTION.

! 950096900-01 AF 1AF785-F1 'A' FEASE FAILED TRIP TEST.

! 950059602-01 AF 1AF81E-34 EREAIER FAILED TRIP TEST.

j 950032658-01 AP 2AF000 REPAIR EOLE IN NON-SEG BUS DUCT.

950009095-01 AP 2AF045 INSTALL BANAMA JACES.

1 950026239-01 AP 2APO4E-01 REPAIR mmm ARC CEUTE.

) 940080309-02 AP 2AF04E-2 INSPECT ITE CIRCUIT maa m m8 l j 920046905-04 AF 2AF04E-2-C REPLACE TIMERS AND TEST. l 1

940000411-02 AF 2AP04E-4 INSPECT ITE CIRCUIT anw m an, j 950099566-01 AP 2AP04E-7 REPLACE PthiP 'TSC' SWITCE.

i 950026655-01 AP 2AF00E-2 REPLACE DEFECTIVE LUG NOUND.

! 950032397-02 AP 2AP19E-102A INSPECT LOW VOLTAGg unwnwyne, l 940060582-01 AP 2AP20E REPLACE BREAKER.

940060583-01 AF 2AP205 REPLACE BREARER.

i 950071362-01 AP 2AP71E-C5 BREAIER TRIPS T" "St8 1- 940087589-02 AP 2AF72E-AA4 PERSO8bi MCC CUBICLE INSFECTION.

950019654-01 AF 2AF755-A1 REPLACE CONTACTOR.

l 950013821-01 AP 2AF76E-E2 OVERLOAD NOULD NOT MSET.

950032397-04 AP 2AP76E-E3 INSPECT LOW VOLTAGE anwawwna, l

950032397-03 AP 2AP76E-E4 INSPECT LOW VOLTAGE maam m a.

f 950118149-01 AF 2AP78E-Al TRIP TEST anwa m .

l 940087569-02 AP 2APS3E-C4 PERMM0f MCC CUBICLE INSPECTION. l i 950081511-01 321 1PS-321N5600 REPLACE PRESSURE SWITCE. ,

l 940061618-03 321 2521F028A REPLACE ASCO VALVES WITE VALCOR. l 940061619-03 321 2521F0205 REPLACE ASCO VALVES WITE VALCOR. l 940061621-03 521 2B21F028D REPLACE ASCO VALVES WITE VALCOR.

l 950040842-01 321 2321F571 INSPECT CEECE VALVE.

940092472-01 521 2LIS-321N702D POTENTICadETER IS INTERMITTENT.

950098369-01 321 2LR-521R8845 MPLACE DISPLAY, MISSING SEmdENTS.

950055138-01 321 2LR-521RSO4B RECORDER PEN DOES NOT OPERATE.

950001823-01 521 2PS-321N560E REPLACE PRESSURE SWITCE.

950033931-01 521 2PS-321N561D LOW PRESSURE ALA308 IS UP. i i 950096741-01 333 lyt-333N024A TRANSMITTER SPIKED DONNSCALE. I 4

950058858-01 B33 2EY-533E610C FLOW CONVERTER FAILED M N.  ;

j 950011966-01 533 2FI-333E6065 MO METER RESPONSE.

i 940079684-01 Of 1 AIR-Cas012 MPLACE DISPLAY.  :

i 950057432-01 CBE 1 AIR-CtdO10 INDICATION IS ERRATIC. l 950092407-01 Cat 1LY-OdO30A CCMPUTER ALAIOf CCbdES IN EARLY. l 950062869-01 Os 1FI-Cas056 CCNTROLLER READS EIGEER TEAM NOIDEAL.

j 950106502-01 Cat 1RE-C3dO17 MPAIR CABLES AND COIDEECTORS. l l 950027653-01 Of 1RIT-C3dO11 NO INDICATION. '

! 950066796-01 CM 1TR-C3dO38A PEN IS INOPERABLE. i 950075401-01 C38 1TR-Cbt038B PEN IS STICKING. l 950100642-01 Of 1TR-CbdO303 PEN 2 IS READING EIGE. '

I

. i

. .. - _ - - - . - - - . . - . _ - - . . . - _ . . . ._-. - ~ . . . - -- -- -

ATTACMENT A 3 ,.

SAsNTY-RELATED MkINTEMANCE OMPLETED

. (NON-OUTAGE RELkTED)

NORE reg # SYS EgvIsMENT DESCRIPTIcN 950104216-01 Chi 2 AIR-C30018 REPLACE DISPLAY.

5 950106217-01 (38 2 AIR-C30047 REPLACE DISPLhr.

) 950106214-01 Cat 2RR-t30017 REPLACE DISPLhY.

950004546-01 Cat 2TI-C30037 INDICATION OUT OF TOLERANCE.

i 950077075-01 C38 2TR-CB0037A **'ve** RED PEN IS STICEING.

j 95009492t-01 tai 2TR-C3003SA **rve** PEN DOES NOT RE57005.

950107801-01 C11 1C11D001011-111 REPLACE 111 VALVE ON ECU 10-51.

. 950009939-01 C11 1C11D001115-SRI-A REPLACE TEST SWITCE A & B.

950106000-01 C11 1C11D001117 REPLACE TEE 117 AMD 110 SSW's CSE ECU'S 34-39,34-55, 22-15 AND 02-39.

? 940079614-01 C11 1C11000 REPLACE TEE DIAPERA NS IN TEE 117 AMD f 118 VALVES.

i 9500f,7779-01 C11 1C11000 REPLACE TEE 117 AMD 118 SSW's CM ECU'S 14-23,46-39,50-15 AMD 10-47.

950057770-01 C11 1C11000 REPLACE TER 117 AMD 118 SS W'S ON ECU'S 10-31,42-31,34-15 AND 26-47.

i 950057702-01 C11 1C11000 REPLACE TER 117 AND 118 SSPV'S ON I ECU'S 22-23,30-39,42-15 AMD 18-47.

950057780-01 C11 1C11000 REPLACE TEE 117 AMD 118 SSPV'S ON i ECU'S 25-15,34-47,46-23 AMD 14-39.

l 950057783-01 C11 1C11000 REPLACE TEE 117 AMD 118 SSPV's ON ECU'S 42-47,18-15,38-23 AMD 22-39.

{

950057743-01 C11 1C11000 REPLhCE THE 117 AMD 110 SSPV'S ON ECU'S 22-07,30-55,02-31 AMD 58-31.

940061246-01 C11 2C11D001002 REPLACE ECU 22-59 M TER ACCtadDLkTOR.

. 940051258-01 C11 2C11D001005 REPLACE ECU 22-55 nTER ACCtadDLkTOR.

I 940061247-01 C11 2C11D001021 REPLACE ECU 18-43 n TER ACCtadDLATOR.

! 940062012-01 C11 2C11D001044 AIR LEAK AT CEARGING NIPPLE. 1 i 940061254-01 C11 2C11D001060 REPLACE ECU 10-23 NETER ACCtadDLATOR.

940061253-02 C11 2C11D001079 REPLACE ECU 26-11 NkTER ACCtadDLkTOR.

940051257-01 C11 2C11D001092 REPLACE ECU 1G-03 NETER ACCtadDLkTOR.

l 940061243-01 C11 2C11D001112 REPLACE ECU 30-47 NATER ACCtadULkTOR.

l 940061242-01 C11 2C11D001127 REPLACE ECU 42-39 NATER ACCtBdDLkTOR.

I 940061239-01 C11 2C11D001131 REPLhCE ECU 50-35 NkTER ACCtBdULATOR.

940073370-01 C22 2PIS-C22N7015 ARNORMAL READINGS, REPLACE RELAY.

i

! 950049502-02 C41 1C41F004A REPAIR AMPEENOL C0001ECTOR.

950036414-02 C41 1C41F0295 RELIEF VALVE LIFTS AT EARLY 12208.

! 950067471-01 C41 1C41000 REEUILD RELIEF VALVE.

! 950063960-01 C41 2C41F004A-AMPE IASS OF ALARM.

, 950022697-01 C51 1C51000 CEECK COMMECTORS FOR LOOSE PIMS. ,

j 940076713-01 C51 1RR-C51R602 DIGITAL DISPLhr MISSING SE MENTS. l j 950068707-01 C51 1RR-C51R603A DIGITAL DISPLAY MISSING SE M ENTS.

{ 950100645-01 C51 1RR-C51R603C INACCURATE RECORDER INDICATION.

950070693-01 C51 1RY-C51K002D EETREME NOISE LEVEL.

l 950022332-01 C51 1RY-C51K601A IBM RESET SWITCE BROEEN.

950097243-05 C51 1RY-C51K501A REPIACE TRANSISTORS IN PREREGUIATORS.

950097243-06 C51 1RY-C51E6015 REPLACE TRANSISTORS IN PREREGULATORS.

l 950097243-07 C51 1RY-C51K601C REPLhCE TRANSISTORS IN PREREGULATORS.

950097243-00 C51 1RY-C511601D REPIACE TRANSISTORS IN PREREGULkTORS.

950097243-09 C51 1RY-C511601E REPIACE TRANSISTORS IN PREREGULATORS.

940086082-01 C51 1RY-C51E601F REPLACE IIti F VOLTAGE PREREGULATOR.

950097243-10 C51 1RY-C51K601F REPLACE TRANSISTCRS IN PREREGUZATORS.

! 950097243-11 C51 1RY-C51K6018 REPLhCE TRANSISTORS IN PREREGUIATORS.

950004226-01 C51 1RY-C511601E ERRATIC YIELDING AIAReds.

950097243-12 C51 1RY-C51K6015 REPLACE TRANSISTORS IN PREREGUIATORS.

I i

, l ATTACMENT A SArETr-REIATED MuMTENANCE COMPLETED l

< MON-OUTAGE REIArED:

MQ3tK 313Q $ SYS EQUIPMENT DESCRIPTION 2 940092263-01 C51 1RY-C51K605CS IMACCURATE INDICATION.

910048577-01 C51 1RY-C51K605EM LP388 FAILED DOWNSCALE.

] 950009081-01 C51 1RY-C51K505 W REPLACE K18 RELAYS ON APERE'S.

950015555-01 C51 1RY-C511605GM CAUSING 200 BICCES TO EtiCS.

950009001-02 C51 1RY-C51K605W REPLACE K18 REIAYS ON AP3ti'S.

950009081-03 C51 1RY-C51K605GP REPIACE K18 RELAYS ON AP3BI'S.

950114867-01 C51 12Y-C51K605GR SPURIOUS AP388 UPSCALE 3EEUTRON TRIP.

950009081-04 ';51

. 1RY-C51K605GR REPLACE K18 REIAYS ON APEGi'S.

950058917-01 C51 1RY-C51K605GS REPLACE K18 REIAYS ON AP38d'S.

950009001-Of C51 1RY-C51K605GT REPLACE K18 RELAYS ON AP38E's.

a 940078168-0A C51 2C51000 EIGE NOISE LEVELS ON 83088/I3045.

l 940062006-01 C51 2RR-C51R603A DIGITAL DISPLAY MISSING SEWENTS.

l 950031481-0t C51 2RR-C51R603C DIGITAL DISPIAY MISSING SEmdENTS.

950007552-0d C51 2RR-C51R603D ELECORDER FAILED TO GO FULL SCALE.

I 950003608-01 C51 2RY-C51K600CA REPLACE CEASSIS COIDEECTOR.

! 940061879-01 C51 2RY-C51K601A REPLACE IBM 'A' PREREGULATOR.

' 1388 METER, INACCURATE INDICATION.

950056699-01 C51 2RY-C51K601E 950054823-01 C51 2RY-C51K605BF POWER SUPPLY IS READING IDN.

950073660-01 C51 2RY-C51K605BF SPURIOUS II ALA388 UP.

940089885-01 C51 2RY-C51K605W REPLACE K-18 REIAYS.

j 940089885-02 C51 2RY-C51K605GN REPLACE K-18 REIAYS.

i 940089885-03 C51 2RY-C51K605GP REPIACE K-18 RELAYS.

' 950075737-01 C51 2RY-C51K605GR APRM 'D' INDICATOR IS ACTING UP.

950032162-01 C51 2RY-C51K605GR COMPUTER POINT FAILED TO RESPOND.

940089885-04 C51 2RY-C51K605GR REPIACE K-18 REIAYS.

950072265-01 C51 2RY-C51K605GS 'E' AP308 CAUSED SPURIOUS BALF-SCRAM.

940089885-05 C51 2RY-C51K605GS REPLACE K-18 REIAYS.

950007295-01 C51 2RY-C51K605GT REPLACE QUAD TRIP CARD.

i 940089885-06 C51 2RY-C51K505GT REPLACE K-18 REIAYS.

I- 940062489-02 C61 1EY-C611002 CALIBRATION.

l 950033194-01 C71 1C71AK010D INSPECT LIMIT POR CAUSE OF 1/2 SCRAM.

950037621-01 C71 1C71AK010D REIAY CONTACTS FAILED.

, 930043774-01 C71 1C71000 INSTALL BANANA JACKS.

I 940059473-01 C71 2C71000 INSTALL BANANA JACKS.

l 950012257-01 DC 1DC001E ADJUST BATTERY ACID CONCENTRATION.

l 940057641-01 DC 1DC06E DEVICES FAILED TO TRIP.

l 950014491-01 DC 1EI-DC055 INDICATION READS IAN.

950018720-01 DC 2DC03E REPIACE BATTERY CNNm ALARM BOARD.

l 940060810-01 DC 2DC16E REPLACE BATTERY CAPACITOR.

l 950116676-01 DC 2DC18E CELL ICV BELON TECE SPEC LIMITS.

950092691-01 DG ODG01K CYLINDER #20 EIGE FIRING PRESSURE.

940062049-01 DG ODG01K REPIACE CYLINDER TEST VALVE.

940078453-01 DG OPS-DG047 REPIACE CRANKCASE PRESSURE SWITCE.

i 950062401-01 DG OTS-DG041 TEMPERATURE SWITCE FAILED TO ACTUATE.

950112187-01 DG 1BS-DG5027 REPIACE VOLTAGE ADJUST POT.

940060075-01 DG 1RS-DG015 INSPECT CONTROL SWITCE.

! 950065758-01 DG 2DG01K INVESTIGATE LOAD FLUCTUATION.

! 950033138-01 DG 2DG01P CLEAM SAND DUST.

i 950004305-01 DG 2DG015-36 INSPECT AND REPAIR CABLE.

i 950013570-03 DG 2DG02JA CALIBRATE REIAYS AND METERS.

, 950055770-01 DG 2PS-D00423 D826 ALA3Gi COMES IN EARLY.

950104145-01 DR 1DR000 REPLACE NUT AND ROD.

950106202-01 DR 2DR000 REPIACE NUT AND ROD.

950306494-01 DR 2DR000 REPLACE NUTS AND 2005.

950050086-01 D10 ORE-D18N511 SEGT LON RANGE ERRATIC READINGS.

950048633-01 Die ORE-D18M513 SEGT WR W EIGE RANGE ERRATIC RKADINGS.

ATTACBENT A SAFETY-RELETED MkINTENANCE CCBIPLETED

~

(NON-OUTh W RELETED)

Nose REQ $ STS EQUISMENT DESCRIPTIM 94000s324-01 D10 ORE-D1SN537 TSaeERATURE ELEMENT FAILED UP SCALE.

940062079-01 D10 005-D1SR516 DIGITAL DISPLAY FAILS.

950116781-01 D10 1RI-D1SN009A REPAIR LOOP COMPONENTS.

950000295-01 Die 1RIY-D18E7513 READS EIGE.

950034109-01 D10 1RIY-D18E751D 'D' CEANNEL SPIEED RIGE.

950114911-01 D18 2RII-D18E610A MONITUR READING EIGE.

950070695-01 D18 2RIY-D10K7513 NO INDICATION LIGBT.

950034200-01 312 1E12C003 REBUIIA NhTER LES PtadP.

940064800-01 E12 1E12C300A MEGGER/ CURRENT READINGS ON MOTORS.

940064885-01 E12 1E12C3003 MEGGER/ CURRENT READINGS ON MOTORS.

950061744-01 E12 1PDS-E12N010AA DIAPERAW FAILED.

950040363-01 E12 1PS-E12N413A INSTRIBGENT STOP VALVE LEAKS.

950017820-01 E12 2E12C002A ADJUST mamwn ALIgggggNT.

950025994-01 E12 2E12C003 MEGGER/ CURRENT READINGS ON MOTORS.

940080684-01 E12 2E12C300A LOOSE GROUND STRAF AT TEE MOTOR BASE.

950068055-01 E12 2E12C300A MEGGER/ CURRENT READINGS ON MOTORS.

950068057-01 E12 2E12C3005 MEGGER/ CURRENT READINGS ON MOTORS.

950033163-01 E12 2E12C300C CLEAN SAND DUST.

950065613-01 E12 2E12C300C MEGGER/ CURRENT READINGS ON MOTORS.

950033169-01 'E12 2E12C300D CLEAN SAND DUST.

950065612-01 E12 2E12C300D MEGGER/ CURRENT READINGS ON MOTORS.

930047047-01 E12 2E12F0063 VALVE BAS PACEING LEAE.

940057847-02 E12 2E12F016A REPLACE MOTOR.

940057033-02 E12 2E12F0165 REPLkCE MOTOR.

940057839-02 E12 2E12F0175 REPLACE MOTOR.

940060100-01 E12 2E12F042A MEGGER/ CURRENT READINGS ON MOTORS.

940075401-01 E21 1E21C002 MOISE CCMING FROM PUMP.

950032540-02 E21 1E21F001 PERIORM STEM LUBE.

940085301-01 E22 1E22F362A INSPECT CBECE VALVE.

940062051-01 E22 1E228001 REPLACE TEST VALVES.

950111775-01 E22 1ES-E2255027 REPLACE THE VOLTAGE ADJUST POT.

950061945-01 E22 2E22C003 GEAR BOE EAS OIL LEAE.

950068425-01 E22 2E22C003 GEAR BOX EAS OIL LEAE.

! 950064065-01 E22 2E22C003 REPLACE PUMP.

! 950104001-01 E22 2E22C003 CHANGE BREATHER TO ELIMINATE CIL LEAEAGE.

! 950075886-01 E22 2E22F362A CCMPRESSOR RELIEF VALVE LIFTING.

) 950044371-01 E22 2E22F362B CHECE VALVE IS LEAKING BY.

i 950036734-01 E22 2E22F375A PACKING LEAE.

! 950075344-01 E22 2E22F3005 ISOLATION VALVE LEAKS.

! 930046741-01 E22 2E22P3013 LOW ALhEM IS UP WEEN D/G Is RUNNING.

950111361-01 E22 2E228001 VARS FLUCTUATING DURING RUN.

950057788-01 E22 2E228001 INSPECT TEE 23 DG POR TRIP.

j .950050109-01 E22 2E228001 REPAIR VAR RELhTED Pomma.

i 950111361-02 E22 2ES-E22BS027 VARS FLUCTUATING DURING RUN.

950031791-01 E31 1PDS-E31N013ak SUPPLY FLOW SWITCE DIAPERAW BLOWN.

940004060-01 E31 1TS-E31N6015 RILEY MODULE IS CHATTERING.

! 950119671-01 E31 1TS-E31N601J INDICATOR TRIPPED, TEMPERATURE NORMAL.

j 950111390-01 E31 1TS-E31N601J RILEY MODULE CHATTERS.

l 950112049-01 E31 2PDS-E31N013RA REPLACE SWITCE DIAPERAM.

! 950104361-01 E31 2TE-E31N004A TIndPERATURE ALARMING WITE NORMAL TEMP.

i 940089867-02 E31 2TE-E31N0295 PERIURM EQ INSPECTIONS.

950097114-01 E31 2TS-E31N601K RESISTER LEAD BROEEN.

950011942-01 E32 1E32F002J DUAL INDICATION WEEN CIDSED.

950064432-01 E32 1FI-E32R653E INDICATION IIGE.

950059342-01 E32 13T-E32N053A REPAIR RING LUGS, 950057412-01 E32 1FT-E32N053E LINE FIDW PEGGED EIGE.

- . ~ - . . - . . - - - - - . - - . - . - ~ -----_-

ATTACMENT A

SAFETY-REIJ.TED MkINTIMANCE CCMPLETED (NCat-OUTAGE RELATED)

NonE REQ # SYS EQUIsMENT DESCRIPTICII 950092263-01 E32 1PI-E32R661A REPLACE METER, BROKEN.

950106504-01 E32 2FT-E32N053J FAILED SURVEILIANCE TEST.

950009233-01 532 2PS-E32N651A REPAIR BROKEN LUG.

I 950044365-01 E32 2PY-332AK008E REPAIR WIRING.

950080580-01 E51 1E51C002 REPAIR FLEE CONDUIT.

J 950025591-01 Eli 1E51C003 MEGGER/ CURRENT READINGS cN teoTORS.

l' 950118483-01 E51 1E51C003 PtB8P TRIPS m m

• UPON START.

950008699-01 E51 1E51F360 REPLACE SPLIT COUPLING.

940072539-01 E51 1E51000 RESET TO TEE CORRECT POSITION.

I 950044294-01 E51 2E51C001 TURRIME TRIP NOULD NOT RESET. '

REPAIR SMkLL OIL LEAX.

_ 940079303-01 E51 2E51C003 j 950065165-01 E51 2E51C003 VIRRATICM INCREASING.

950059795-01 E51 2E51F019 REPLACE m a m 74 RELAY.

, 950052189-01 E51 2E51F022 GROUND DETECTOR AIA398 WEEN CYCLED.

j 950013545-03 FC 1FCA000 10 YR PIPING INSPECTION.

j 950013545-04 FC 1FCB000 10 YR PIPING INSPECTICIt.

j 950037023-02 FC 1FC017 VALVE WILL NOT GO FULL CIASED.

4 940058871-01 FC 1FC0445 IWTERNAL T N N , REPAIR.

! 940058870-01 NC 1FC0455 INTERNAL LEAEAGE, REPAIR.

940061986-01 FC 1FE-FC037 INSPECT FLON ELEMENT.

l CEECE ANTI-ROTATION CLAMP, COLLAR.

950041906-01 FC 23C017 950068859-01 FC 2FC03PA MEGGER/ CURRENT READINGS ON MOTORS.

l 940074030-01 FC 2FC03PA EIGE VIBRATION READINGS.

950033154-01 FC 2FC03PB CLEAN SAND DUST.

! 930047088-01 FC 2FC03PB EIGE VIERATICII IN B00 TOR.

940061419-01 FC 2FC044A INSPECT CEECE VALVE.

940061355-02 FP 2FP000 REMOVE THEIedo-IAG MATERIAL.

I. 950019655-01 EG 1BG006A TW N af* TRIPPED WITE OPEN SItatAL.

950007966-01 EG 1E0009 PERFOpti EQ INSPECTION CII ACTUATORS.

l 950032673-02 EG 150009 PERICBei STEM LUBE.

940059980-02 EG 250018 PERF0508 EQ INSPECTICIf CM ACTUATOR.

950114627-01 E13 1513P644 REPLACE BLOWN FUSE.

l 950044793-01 E13 2E13P601 REPLACE MISSING COVERS.

j 95C059047-01 513 2B13P509 REPIACE WIRING.

j 950021719-01 E13 2513P611 REPIACE TERMINAL BLOCK.

i 950071146-01 IS 2IS000 DOOR DOES NOT SEAL.

930046195-01 LP 1LP000 REPLACE POMP ARGdETER. ,

f 950025820-01 LV 1LV93E PENETRATION E-26 READS LON.

' 940079202-01 NR 2NR000 'A' RBW SPIRED UPSCALE.

!' 950037118-01 FC 2PCM111 REPIACE DOOR SEAL.

! 950001114-01 PL OPLD2J PtB8P WILL NOT RUN IN AUTO OR MANUAL.

3 950104030-01 PL IPL76J 'E2' REAGENT GAS BOTTLE LEAKING.

l 950105681-01 PL IPL77J 'E2' SAMPLE READING IIGE.

950068717-01 PL 1PL77J OSCILLATES IRCM 0-54.

l 940061207-01 PL 2PL76J REPLkCE MISSING SCREWS / COVER CII PANEL.

940051205-01 PL 2PL77J INSPECT ASSEMELY.

950056371-01 RD 1RD000 ROD 30-39 DRIFTED IN DURING TESTING.

, 950013385-01 RE 1RE000 PERF0308 FIDW TEST.

I 950007075-01 RI 1RE13-1131E CIAMP BOLT I40SE.

! 940006111-01 RE 2REA000 10 YR BYDRO INSPECTICII.

940086111-02 RE 2RE3000 10 YR EYDRO INSPECTION.

950013545-05 SC 1SCA000 10 YR PIPING INSPECTICIS.

l 950013545-06 BC 1SCB000 10 YR PIPING INSPECTIcet.

950075805-01 Vc Orm-vC028 RECORDER NOT FUNCTIceIING PROPERLY.

950010245-01 VC OvC01SA RELAY CONTACTS DIRTY.

A ATTACB8ENT A

' ~

EAPETY-MLATED MAINTENANCE CChiPLETED (NON-QUTAGE RELkTED) a i

Nose reg # SYS EQUIsNENT DESCRIPTICM i

950027296-01 VC OVC05CB CEg g am uv3 AND CONTACTORS.

950076573-01 VC OVC05YB DAMPER FAILED CIASED.

950056970-01 VC 0VC05YB DAMPER FAILED CIDSED.

j 950019453-02 VC OVC24YB DAMPER FAILED CLOSED.

920049006-01 VC OvC43Y INSTALL ACCESS DOOR IN DUCTNORE.

920049007-01 VC 0VC45Y INSTALL ACCESS DOOR.

! 940060653-01 VC OEY-VC125A ALMess FAIL TO RESET.

i 950005646-01 VC OEY-VC1258 DETECTOR IS INDICATING UPSCALE.

} 940060653-02 VC OEY-VC185A ALARMS FAIL TO RESET.

] 950054500-01 VE OFZ-VE067 INDICATOR OUT OF CALIBRATICE.

950068056-01 VE OTI-VE044 TSMPERATURE INDICATION READING EIGE.

! 950069479-01 VE OTS-VE109 CCBiPRESSOR TRIPPED ON EI TEMPERATURE.

} 950100184-01 VE OVE01AA REPLACE BOTE DOOR GASEETS.

i 930047337-01 VE OVE03CA EECESSIVE VIBRATION.

i 950045529-02 VE OVE03CA FAN TRIPPED NEILE STARTING.

950070746-01 VE OVE04CA COMPRESSOR TRIPPED ON EI TEMPERATURE.

l 950027290-01 VI OVE04CA CEECE MLAYS AND CONTACTORS.

950107207-01 VE OVE04CA CEECE RELAYS.

! 950060071-01 VE OVE04CA TRIPPED MULTIPLE TIMES ON OVERLOAD.

! 950061404-02 VE OVE04CA TRIPPED ON IIGE OIL TEMPERATURE.

950020079-01 VE OVE04CA CCMPRESSOR NOT OPERATING PROPERLY.

950027297-01 VE OVE04CE CEECE RELAYS AND CONTACTORS.

950036286-01 VE OVE09YB REPLACE PUSE.

I 950111839-01 VG 1PC-VG003 REPLACE CAPACITORS IN FION CONTROLLER.

950113419-01 VG 1PC-VG003 REPLACE CAPACITORS IN FLOW CONTROLLER.

950025990-01 VG 1FE-VG003 REPIACE DAMPER ACTUATOR.

950017234-01 VG 1PDS-VG021 ALMei CAME IN EARLY.

940062070-01 VG 1TT-VG012 REPAIR POWER LEAD.

l 940060241-01 VG IVG001 PERIORM SPRINGPACE TEST.

! 940060242-01 VG 1VG003 PERICSei SPRINGPACE TEST.

950030908-01 VG IVG003 VALVE SEOWS DUAL INDICATION.

930045040-01 VG 1ES-VG004A REPLACE CICSED E2ID LIMIT SWITCE.

930045039-01 VG 1ES-VG0043 REPLACE OPEN END LIMIT SWITCE.

940090400-01 VG 2FS-VG036 EIGE FION ALARM INTERMITENT.

l 950056028-01 VG 2FE-v0003 DAMPER FAILED CIASED.

I 950016902-01 VG 2VG01C REPIACE EFA REIAY.

l 920044926-01 VG 2VG02Y REPLACE RASE BOLTS.

i 940060767-01 VG 2ES-VG004A REPIACE LIMIT SWITCE.

i 940060768-01 VG 2ES-VG004B REPLACE LIMIT SWITCE.

940062114-01 VQ 1VQO31 REBUILD ACTUATOR.

940059509-02 VR 2VR04YA REPIACE DAMPER SPRINGS.

940059507-02 VR 2VR04YB REPLACE DAMPER SPRINGS.

940059508-02 VR 2VR05YA REPLACE EEISTING SPRINGS.

) 950022516-01 VI 1VE07Y DAMPER IS LEAEING OIL.

950058063-01 VY 1 TIC-VY024 ALMei UP, TEMPERATURE - ?..

950054867-01 VY 2 TIC-VY024 PUMP CUBICLE TEMPERATURE IIGE ALMef.

l 950016300-01 VY 2VYO4C INSPECT WIRING TO PILOT LIGHT.

950104061-01 EE 0EE000 YOEOGANA RATTERY REPLACEMENT.  !

950104057-01 EE OEE000 YOEOGANA RECORDER'S TIME CEANGE.

3

_~ - -. - - - -_ _ --. --. _ ._. - _. ..

, . ATTACBENT D

- II.D UNIT SEUTDOWN3 (UNIT 1)

DATE: 950209 GENERATOR OFF-LINE: 140.3 OUTAGE TYPE: Forood (T.1F29)

~(YYMOD) (Bours)

REASON: Manual Reactor scram to repair the 1E51F357 valve.

$ CRITICAL ACTIVITY PATE: Repair of the 1E51F357 valve.

CORRECTIVE ACTIONS (FIF/LER# if applicable): None.

! RADIOACTIVITY RELEASE /EEPOSURE OVER 10% ALLONABLE VALUES: None.

SAFETY RELATED CORRECTIVE MAINTENANCE CCadPLETED:

l WRNUM SYS EPN Description i 940061467-01 N M -C51N002E IV PLOTS Issa =Ea SEONS RESISTIVE.

950010998-01 C51 1RY-C51R601A DOWNSCALE IS ERRATIC.

950011034-01 VG 1FC-VG003 CYCLES RECESSIVELY WEEN VG IS RUNNING.

f DATE: 950611 GENERA'tCR OFF-LINE: 158.2 OUTAGE TYPE: Forced (L1F30)

(YDedDD) (Sours) f

REASON

Manual Reactor scream to replace the Main Steam Isolation valve solenoid valves.

CRITICAL ACTIVITY PATE: Replacement of solenoid valves.

CORRECTIVE ACTIONS (PIF/LER# if applicable): None.

RADIOACTIVITY RELEASE /EEPOSURE OFER 10% ALLONABLE VALUES: None.

SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED:

f 4

WRNUu SYS EPN Description E00197ss-01 N M -C51N002A A IRu OUTPUT uRTER IS ERRATIC.

950039092-01 C71 1ES-C71N006D REPLACE MSV #2 LIMIT SWITCE. l

, 950039093-01 C71 1ES-C71N0065 REPLACE MSV #1 LIMIT SWITCE.

950039894-01 C71 1ES-C71N006F REPLACE MSV #3 LIMIT SWITCE.

950039895-01 C71 1ES-C71N0065 REPLACE MSV #4 LIMIT SWITCE.

950042440-01 C51 1RY-C51R601E INSE 'E' FREQUENTLY SPIRES EIGE.

, 950051999-01 B21 1821F02BA 'A' MSIV REOPENED WITE SWITCE IN THE CLOSED POSITION.

950052000-01 B21 1821F02SB REPLACE TEE ASCO PILOT SOLENOIDS.

950052001-01 B21 1521F02SC REPLACE TEE ASCO PILOT SOLENOIDS.

, 950052002-01 321 1B21F02SD REPLACE TEE ASCO PILOT SOLENOIDS.

950052009-01 C51 1RE-c51N002E ERRATIC INDICATION AND 1/2 SCRAM.

950052029-01 PL 1PL77J O2 MONITOR STUCR AT 17.5%, ACTUAL IS 21%.

1 1

I

e ATTACESNT D

• II.D UNIT SEUTDONNJ (UNIT 1)

DATE: 950816 reuwanTOR OFF-LINE: 130.4 OUTAGE TYPE: Forood (L1F31)

Tff53E) (Bours)

REASON: Automatic Reacter scram from Main Steam Isolation Valve closure due to Main Steam Tunnel high tesperature which was caused by the loss of the Reacter Building ventilation system.

CRITICAL ACTIVITY PATE: Troubleshooting and repair of the EPA logic card.

CORRECTIVE ACTIONS (FIF/LERS if applicable): LERS95-014.

RADIOACTIVITY RELEASE / EXPOSURE OVER 10% ALLONABLE VALUES: None.

SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED:

WRNtSi SYS EPN Description N E7T W 18003A REPLACE TRIP CARD.

940000675-01 C71 1C718003B REPLACE TRIP CARD.

940008675-02 C71 1C7180035 CALIBRATION INDICATES BAD TRANSFORMER.

940080676-01 C71 1C713003D REPLACE TRIP CARD.

940000606-01 C71 1C713003C REPLACE TRIP CARD.

950054043-01 C51 1RY-C51R601E 'E' IBM ERRATIC INDICATION.

950054043-02 C51 1RE-C51N002E REPLACE IRN "E" DETECTOR.

950056549-01 C51 1RY-C51R601A ERRATIC OUTPUT CAUSING REIJY CEATTER.

950056549-02 C51 1RE-C51N002A REPLACE I3ei "A" DETECTOR.

950071755-01 PL 1PL76J OKYGEN CONCENTRATION IS UPSCALE EIGE.

950072292-01 E31 1TS-E31N604C ALA3ed RELAY CRATTERS.

950072362-01 C71 1C715003D RESOLDER CONNECTIONS ON LOGIC BOARD.

DATE: 950924 GENERATOR OFF-LINE: 102.1 OUTAGE TYPE: Forood (LIF32)

TYf55E) (Bours)

REASON: Manual Reactor scram due to loss of the '15' Turbine Driven l Reactor Feed Pump during surveillance testing.

! CRITICAL ACTIVITY PATE: Inspection and troublesooting of the 81B' Turbine Driven Reactor Feed Pusp.

f CORRECTIVE ACTIONS (FIF/LERS if applicable): LER# 95-016.

1 RADIOACTIVITY RELEASE / EXPOSURE OVER 10% ALLONABLE VALUES: None.

SAFETY RELATED CORRECTIVE MAINTENANCE CCSGPLETED:

WRNUM SYS EPN Description

, N E3Y H 33r060s INSPECT AND BODIFY THE LVDT LINKAGE AND VALVE INSULATION.

950092399-01 B33 1533F060A INSPECT AND BODIFY THE LVDT LINRAGE AND VALVE INSULATION.

J i

4 1

1

~

l . AT N D II.D UNIT SEUTDOIINS

. (UNIT 2) 5 DATE

950219 MNERATOR OFF-LINE: 2713.0 ccTAGE TYPE: Refuel (L2R06)

(rzandoD) (sours) l 1 mEAScN

Refuel estage. ,

! I i

CRITICAL ACTIVITY PATE: See Z - 'i= A.

coSasCTIVE ACTIcNS (FIF/IER$ if applicable): None.

RADronCTIVITr RELEASE /EEPOSURE OVER lot ALIAu1BI2 VALUES: None.

1

• SAFETr RELATED CoSSECTIVE ndAINTENANCE CCB8PLETED: See 5. "4w B.

j *

DATE: 950612 eEastmATok oFF-LINE: 1.5 ocTAaB TrPE: Scheduled (L2M10)

} (YYbSSD) (Bours) l ESASON: temia Tushine everspeed test.

I

! CRITICAL ACTIVITr PATE: Completion of everspeed testing. l l CORRECTIVE ACTICIls (PIF/LERG if applicable): Nene.

RADIQACTIVITE RELEASE /EEPOSURE OVER 10% ALIANEABLE VALUES: None.

SAFETr RELATED COR33CTIVE ndAINTENANCE CCSEPLETED: None.

i 1.

DATE: 950916 GENERATOR OFF-LINE: 189.5 OUTAGE TrPE: Scheduled (L2F23)  !

{ '

{ (YItemD) (Bours)

?

REASONt Idanual Reactor scram for main +====a= of the '2A' Roaster Rooirculation Flow Control Valve.

t l

CRITICAL ACTIVITY PATE: Repair of the Flow Control Valve.

l CORRECTIVE ACTICIls (PIF/LER$ if applicable):

i RADIQACTIVITr RELEASE /EEPOSURE OVER 10% ALICIEABLE VALUES: None. ,

SAFETr RELATED CORRECTIVE tikINTENARICE CC38PLETED:

l

} NRHtBi SYS EPN Description

] 930043461-05 E32 2PT-E32N051J ISOLATION VALVE LEAES BY.

J 930044523-04 E32 2PT-E32N051A VALVE LEAKING BY.

940058620-03 E32 2E32N051N VALVE LEAKING BY.

950053301-01 E51 2E51F066 NO INDICATICII.

I 950055503-01 B21 23215932A DUAL VALVE INDICATICII AT NULL POIRER.

950069089-01 533 2333F060A DISASSEBERLE/ REPAIR AS REQUIRED.

j 950069009-04 333 2333F060A INSTALL IARGER FEED BACE LEVER PINCE BOLT.

l 950069009-05 333 2333F060A LOOSEN & RETIGRTEN PACKING.

j 950077175-01 B33 2333F067A INSPECT AllTI-ROTATICII DEVICES VALVE.

950077175-02 333 2333F0675 INSPECT ANTI-ROTATICII DEVICES VALVE.

950090461-01 E12 2E12F041A PACKING 12AE.

950090461-02 E12 2E125941A REPLACE ACTUATOR.

1 950090461-03 E12 2E12F041A INSPECT ACTUATCEL RETURN SPRING.

950092327-01 333 2B33r0605 INSPECT /tdODIFK VALVE INSUIATION.

e

. ATTACBSNT C

• II.D PORCED REDUCTIONS IN POWER GREATER TRAN 20% IN DESIM POWER LEVEL (UNIT 1)

DATE: 950705 OPERATIcet AT REDUCED POWER: 7.0 (YT5555T (E) ewm em; yover reduction for scram time testing of control rods due to replacement of scram solenoid pilot valves.

CRITICAL ACTIVITY PATE: Cogletion of testing.

CORRECTIVE ACTIONS (FIF/LERG if applicable): None.

RADICACTIVITE RELEASE / EXPOSURE OVER 10% ALLONABLE VALUES: None.

SAPETY RELATED CORRECTIVE MkINTENANG C(EdPLETED:

MORR REQ # SYS EQUIPMENT DESCRIPTION 950057109-01 M 1C11000 REPLACE SSPV'S ON ECU 14-55.

950057110-01 C11 1C11000 REPLACE SSPV's ON BCU 46-07.

950057111-01 C11 1C11000 REPLACE SSPV's ON BCU 10-15.

950057112-01 C11 1C11000 REPLACE SSPV's Ott BCU 50-47.

DATE: 950717 OPERATION AT REDUCED PONER: 7.0 (YT55EBT (Bours)

REASON: Power reduction for scram time testing of control rod due to replacement of scram solenoid pilot valves.

CRITICAL ACTIVITE PATE: Completion of testing.

CORRECTIVE ACTIONS (PIF/LER# if applicable): None.

RADI0 ACTIVITY RELEASE / EXPOSURE OVER 10% ALLONAELE VALUES: None.

SAFETY RELATED CORRSCTIVE MAINTENANCE COMPLETED:

MORK REQ # SYS EQUIPMENT DESCRIPTION 950055593-01 M 1C11000 REPLACE SSPV'S ON ECU 30-39.

DATE: 950718 OPERATICN AT REDUCED PONER: 6.0 I

(TT M D) (Hours)

REASON: Power reduction for scram time testing of control rods due to replacement of scram solenoid pilot valves.

CRITICAL ACTIVITE PATE: Completion of testing.

l I CORRECTIVE ACTIONS (PIF/LER# if applicable): None.

RADI0 ACTIVITY RELEASE / EXPOSURE OVER 10% ALLONAELE VALUES: None.

SAFETY RELATED CORRECTIVE MkINTENANCE COMPLETED:

MORR REQ # SYS EQUIPMENT DESCRIPTION 950057754-01 M 1C11000 REPLACE SSPV'S ON BCU'S 06-23, i 54-39, 34-31 AND 26-31. ,

i j

i

. - - - - - - - - - _ _ . - --- + - ,

I e ATTACEMENT C

'

• II.B FORCED REDUCTIONS IN POWER

{ GREATER TRAN 20% IN DESIGN PONER LEVEL (UNIT 1)

DATE: 950719 OPERATION AT REDUCED POWER: 6.0 (YTismbT (nours)

REASON: Power reduction for scram time testing of control rods due to replacement of scram solenoid pilot valves.

CRITICAL ACTIVITY PATE: Completion of testing.

CORRECTIVE ACTIONS (PIF/LER$ if applicable): None.

RADIOACTIVITY RELEASE / EXPOSURE OVER 10% ALLOWABLE VALUES: None.

SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED:

WORR REQ # SYS EQUIPMENT DESCRIPTION 950057785-01 M 1C11000 REPLACE SSPV's ON BCU'S 30-23, 38-07, 22-55, 10-31 AND 50-31.

l DATE: 950722 OPERATION AT REDUCED POWER: 7.0 (fDe@b) (Hours)

REASON: Power reduction for scram time testing of control rod.

CRITICAL ACTIVITY PATE: Completion of testing.

CORRECTIVE ACTIONS (PIF/LER# if applicable): None.

RADIOACTIVITY RELEASE / EXPOSURE OVER 10% ALLONABLE VALUES: None.

SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED:

WORK REQ f SYS EQUIPMENT DESCRIPTION 950062853-01 M 1C11000 REPLACE PROBE HLTE CARD FOR CONTROL ROD 22-55.

DATE: 951027 OPERATION AT REDUCED POWER: 7.5 l (YY5535BT (sour =>

REASON: Power reduction for scram time testing of control rod due to replacement of scram solenoid pilot valves.

1 CRITICAL ACTIVITY PATE: Completion of testing.

CORRECTIVE ACTIONS (PIF/LERf if applicable): None.

RADIOACTIVITY RELEASE / EXPOSURE OVER 10% ALLOWABLE VALUES: None. f l

SAFETY RELATED CORRECTIVE MAINTENANCE COMPLETED: l WORK REQ # SYS EQUIPMENT DESCRIPTION 950105037-01 M 1C11000 REPLACE SSPV'S ON BCU 42-39.

1 1

ATTACBdENT C II.D PORCED REDUCTIONS IN POWER

, GREATER TRAN 204 IN DESI M POWER LEVEL (UNIT 1)

DATE: 951029 OPEaATION AT REDUCED PouBR: 0.5 (Y13eEID) (Bours)

REASON: Power reduction for scram ties testing of oontrol rods due to scram solenoid pilot valve concerns.

CRITICAL ACTIVITY PATE: Completion of testing.

CORRECTIVE ACTIONS (PIF/LERG if applicable): None.

RADIQACTIVITY RELEASE /EEPOSURE OVER lot ALLONABLE VALUES: None.

SAFETT RELATED CORRECTIVE RELINTENANCE COMPLETED: None.

DATE: 951100 OPERATICM AT REDUCED POWER: 4.5 (yInedDD) (Bours)

REASON: Power reduction for scram ti- testing of control rods due to replacement of sorsa solenoid pilot valves. ,

CRITICAL ACTIVITY PATE: Completion of testing. T l

CORRECTIVE ACTIONS (PIF/LER$ if applicable): None.

RADIQACTIVITg amtmamW/EEPOSURE OVER 100 af.Ymm2 VALUES: None.

SAPETE RELATED CORRECTIVE BGLINTENANCE CCRIPLETED:

WORE REQ f SYS EQUIPMENT DESCRIPTION 950106007-01 C11 1C11000 REPLACE SSPV'S ON ECU's 06-47, 10-23, 30-47 AND 46-15.

DATE: 951217 OPERATION AT REDUCED POWER: 30.5 (rYnedDD) (Bours)

REASON: Soram tim testing of control rod 50-23 due to Eydraulio Control Unit t

water accumulator replacement.

i l CRITICAL ACTIVITY PATE: Completion of testing.

1 l CORRECTIVE ACTICels (PIF/LERf if applicable): None.

i RADIQACTIVITY RELEASE / EXPOSURE OVER 10% ALIDHARLE VALUES: None.

I SAPETE RELATED CORRECTIVE 3GLINTENANCE COMPLETED:

NORE REQ $ SYS EQUIPMENT DESCRIPTION 940086544-01 C11 1C11001156 REPLACE THE MATER ACCUMULATOR ON l ECU 50-23.  ;

i a

f l

i

• ATTACEMENT C

• II.D FORCED REDUCTIONS IN POWER GREATER TRAN 20% IN DESIGN POWER LEVEL (UNIT 2) i I

DATE: 951221 OPERATION AT REDUCED POWER: 8.0 (YT55EST (sours)

REASON: Power reduction for maintenance on the '2A' and '2B' Feedwater Beater Drain Pump.

CRITICAL ACTIVITI PATE: Completion of maintenance.

CORRECTIVE ACTIONS (FIF/LER# if applicable): None.

RADIQACTIVITY RELEASE / EXPOSURE OVER 10% ALLONARLE VALUES: None.

SAPETI RELATED CORRECTIVE MAINTENANCE CCS4PLETED: None.

l l

l l

?

I I

l l

1

r ATTACIDENT D

,' RADIATION EXPOSt7RE REGULATORY OUIDE 1.16 REPORT NLMBER OF PDSONNEL Ale PEkSON REM BY WORE AND JOB FUNCTION FOR 1995

. .. ... .. . MLMBER OF PER SONNEL . . .. ... ..... GUANTITY OF PERSON-REM ....

STAflow CDNTRACTORS/ UTILITY TOTAL STATION CONTRACTORS / UTILITY TOTAL WORE AND JOB FUNCTION EMPLOYEES OTHERS EMPLOYEES PER$0NS EMPLOYEE 5 07 6 5 EMPLOYEES PERSON REN REACTOR OPERATI0ers AND SURVEILLANCE ENGINEERING 76 5 0 8.275 1.438 0.000 NEALTH PHYSICS 43 7 123 17.090 1.677 1.100 MAINTENANCE 34 18 1 15.982 3.384 0.057 OPERAfl0Ns 132 115 0 34.418 5.730 0.000 SUPERVISORY 100 64 0 6.819 1.648 0.000 385 209 124 718 82.584 13.877 1.157 97.618 ROUTINE MAINTENANCE ENGICERING 56 91 0 6.083 25.475 0.000 NEALTN PNTSICS 36 29 147 14.154 7.041 1.313 MAINTENANCE 230 647 21 108.542 123.029 1.089 OPERAfl0Ns 18 0 0 4.718 0.001 0.000 SUPERVISORY 148 69 0 10.153 1. 774 0.000 488 835 168 1491 143.650 157.321 2.402 303.373 INSERVICE INSPECTICW ENGINEERING 8 39 0 0.858 11.022 0.000 NEALTH PHYSICS 1 21 3 0.238 5.224 0.028 MAINTENANCE O 242 0 0.094 46.016 0.000 OPERATIONS 0 0 0 0.042 0.000 0.000 SUPERVis0RY 6 24 0 0.388 0.615 0.000 15 326 3 344 1.620 62.876 0.028 64.524 1 SPECIAL MAINTENANCE l ENGINEERING & 0 0 0.475 0.003 0.000 NEALTN PHYSICS 1 0 1 0.453 0.017 0.006 MAINTENANCE 2 38 2 1.023 7.262 0.124 OPERATION $ 0 0 0 0.121 0.000 0.000 SUPERVISORY 4 6 0 0.257 0.164 0.000 12 45 3 60 2.329 7.645 0.130 9.904 ,

REFUELING . l ENGINEERING 1 57 0 0.110 15.934 0.000 NEALTH PHYSICS 4 2 13 1.591 0.389 0.114 MAINTENANCE 3 42 0 1.242 8.079 0.000 OPE RATIONS 7 0 0 1.8 72 0.000 0.000 SUPERVISORY 18 3 0 1.214 0.080 0.000 34 104 13 149 0.046 h.4al 0.114 30.623 WAETE PROCESSING ENGINEERlWG 0 0 0 0.026 0.000 0.000 NEALTH PHYSICS 1 0 4 0.360 C.000 0.034 MAINTENANCE 3 10 0 1.267 1.917 0.000 OPERAfl0NS 1 31 0 0.380 1.546 0.000 SUPERVISORY 2 4 0 0.113 0.114 0.000 7 45 4 56 2.146 3.577 0.034 5.757

s ATTACENE:NT D RADIATIORI EXP08URE REQULATORY CUIDE 1.16 REPORT IRmRER OF PERSONNEL Aae PERSON REM BY %dORK AND JOB FUNCTION FOR 1995

- ..- QUANTITY OF PERSON REM --

. . . . . . . . . IRAIRE R OF PER SCWNE L * * . . . .

STATION CONTRACTORS / UTILITY TOTAL STATION CONTRACTORS / UTILITY TOTAL WORK AaC JOB FUNCTION DIPLOTEES OTHERS EMPLOTEES PERSONS EWLOYEES OTHERS EMPLOTEES PERSON REM ,

......o.0................

TOTAL ST JCB FUNCTIDW ENGlWEERInG 146 192 0 338 15.827 53.872 0.000 69.699 NEALTN PsTSICS 86 59 290 435 33.886 14.348 2.595 50.829 ,

MAINTEmAact 271 997 25 1293 128.149 189.686 1.270 319.105 OPERATIns 159 146 0 305 41.552 7.277 0.000 48.829 suPERvisonY 277 170 0 447 18.943 4.394 0.000 23.337

• ORAac TOTAL = 939 1564 315 2818 238.357 269.577 3.865 511.799 l

l l

l l

l l

l i

l I

4 APPENDIX A CRITICAL PATE ACTIVITIES REFUEL OUTAGE (L2R06) d

'B' DIESEL ENERATOR REIAY CALIBRATIONS AaB TRIP TESTS.

330WE REACTOR EEAD.

SOUP 2 Aam 4 INBOARD ISOLATICII FUNCTIONAL TEST.

I PtBEF DOIRf INNER GRATING.

. EmdOVE EXCITER REATER SENSCR.

3 INSPECTION OF TER INBOARD MSIV ACCthi CEECE VALVES.

j F1400 IIDGER GRATING.

1 A,5 ABID C MSIV'S ACTUATOR PISTott LEAR CHECK.

INSPECT ALL EEACTOR STUDS DURING DETENSIONING.

j START 'A' SUPPRESSICN POOL CLEANUP PtBEP.

5 MAIN STEAM PRE-48AINTENANCE LEAR RATE TESTS.

j EEACTOR RECIRCULATION FlatP IAGIC TEST.

4' REACTOR RECIRCULATICE PtB8P ATWS TRIP RELAY LOGIC TEST.

LEAIAGE OONTROL SYSTEM MAINTENANCE.

INSTALL TRI-NUCLEAR FILTER EQUI 3 MENT.

i DRAIN ETROGEN SEAL OIL.

! CAv1Tr rIooDUP.

i INSTALL CATTLE CEUTE.

j EIGE PRESSURE CORE SPRAY LEAR RATE TESTING.

{ 2E22-F004 AND 2E22-F005 LEAR BATE TEST.

FILL REACTOR VESSEL NITE SUPPRESSICII POOL CLEANUP.

ESTABLISE NATER CLARITY.

CD/CB SYSTEM BOUNDRI OUT OF SERVIG, SYSTEM DRAINED.

l EIGE P3333073 CORE SPRAY SYSTEM MAINTSMANCE.

d DISASSEMBLE /3EASSWRf2 2E22-F004 VALVE FOR EQ INSPECTICIt.

REMOVE FUEL POOL GRTE.

j FUEL POOL COOLING ESTABLISEED AS ALTERNATE SEUTDOIDI COOLING.

l UNLOAD RX CORE.

RESIDUAL BEAT REMOVAL ' A' Ih0P LEAR RATE TESTING.

c REPLACE TIMER 3EIAYS Cet TR232A, TR2325, TR234A AND TR2343.

DIV I 3ER RELAY LOGIC FUNCTIONAL TEST.

j EQ INSPECTION FOR 2E12-F064A.

SECUES B RESIDUAL EEAT REMOVAL IN SEUTDOWN COOLING.

INSERVICE INSPECTI0tt OF 2E12-P000 AND 2E12-F009.

l SEUTDONN COOLING SYSTEM LEAR RATE TESTING.

. 2E12-F005, 2E12-F000 AND 2E12-F009 LEAK RATE TEST.

! DRAIM CCBe000f SEUTDONN COOLING EEADER.  ;

i 2D10-N009A,B,C AND D REACTOR BUILDING CALIBRATION.

A,B,C AND D REACTOR BUZIDING EXEAUST PLENthi CALIBRATION.

4 VOTES TEST FOR 2E22-F004 AND 2E21-F005.

DYNAMIC VOTES TEST CII 2E12-F064A. I 1

SYSTE38 AUZILIARY TRANSIO30tER MkINTEMANCE.

2AP91E REVISE TRANSF038 DER COOLING.

'C' LOOP RESIDUAL IFAT REMOVAL SUCTIOtt DECON. ,

'A' AND 'B' EER SYSTEM DECCIt.

'C' 140P RESIDUAL IEAT REMOVAL VENTED.

2D18-N015A,B,C AND D IUEL POOL VENT CALIBRATION.

EYDRO TEST CCIORECTICIts 005 'B' RER SYSTEM.

Rett 0VE DECOtt FLANGE AT 2RE27bt, 2E12-D311A, 2E12-F384A. j 4

i

_ _ . _ . . . _ . . . _ _ . _ _ . . . _ _ _ . . . _ ____ _.._. _ _ . _ . _ _ _ . _ _ _ _ _ _ . . ~ _

e 4 e i

APPENDIE A CRITICAL PATE ACTIVITIES REFUEL OUTAGE (L2R06) 3 A,B,C ARED D REACTOR BUIIDING FUEL POOL EXEADST CALIBRATICIt.

RESIDUAL EEAT RZ200 VAL ' A' IAOP MAINTENANCE.

4 PERNCSBf STATICII AUX TRANSNCEedER CIRCUIT CEESS.

2DC14E REPLACE POST SEAL ASSEMEMES, j CEEMICAL DECON NCR REACT 00L RECIRCUIATION PIPING.

l STATION AUX TRANSF058 DER mauw ADDITIONS.

j VERIFI TELIPPING F5Kei NEN STATICII AUX TRANSFogedER RELAYS.

1 ndOTOR OPERATED VALVEg amwarren OUT OF SERVIG. '

PERF00Si TELANSFOgedER TESTING.

3 2DC14E BATTERY 125 VDC BATTERY INSPECTICII.

REACTOR BUIIDING IVAC AVAILABLE, CLEAR 005 CII FANS.

j SYSTERE AUX TRANSFOIndER 243 TESTING.

INSTALL NUEL POOL GATES FCEL REACTOR RECIRC DROQIf DRAIN DOWN.

1 EQ INSPECTION OF VALVE 2E12-F042A.

OIL CIRCUIT mo m en 1-6 RELAY CALIBRATICII. l

< RESIDUAL EEAT REMOVAL 'B' IAOP ndAINTEIEAN G . 1

, DIV 1 REACTOOL VESSEL ZON LEVEL ROSE 3dOUNT TRANSIEITTERS. l

! REACTOOL LEVEL ECCS TRANSMITTER EERO SEIFT CALIBRATICIt. i 2E21-C001 RELAY CALIBRATICII. l l

2C11-C001A RELAY CALIBRATICIt.  !

I 2E12-C002A RELAY CALIBRATICII.

! DIVISICII 1 AND 2 DC ndETER CALIBRATIONS.

} BUS 241Y TELIP TESTING.

l BUS 235X AND 235Y TRIP TESTING.

i CRANGE GEAR SET AND 300TCOL NOR 2E12-F042A, i

EQ INSPECTICII OF VALVE 2E12-F0245.

DIVISICIt 1 AND 2 125V m w m CAPACITY TEST.

l DIVISI0It 1 AND 2 RATTERY PERFQEbdANG DISCEARGE TEST.

DIVISICII 1 BATTERY LON/EIGE VOLTAGE CEARGE.

CORE SEROUD INSPECTION AFTER DECOII.

POST MAINTENANG TESTING OF VALVE 2E22-F005.

l 2E22-F012, F014, F015 AND F023 LEAR RATE TEST. l POST MAINTE3 TANG LEAR RATE TESTS CII EPCS SYSTEM. l j FILL AND VENT THE RIGE PRESSURE CORE SPRAY SYSTEM. i

' l INSE3LVIG INSPECTICII OF EPCS SYSTEM VALVES.

I STEM LUBES & GREASE SAMPLES OF EPCS VALVES.

EIGE PRESSURE CORE SPRAY RELAY LOGIC TESTING.

I DYNAMIC VOTES TEST cet 2E12-F0243. )

CLEAR BOUIEDRY CUT OF SERVI G Ott B RER DISCEARGE. l l DIV 2 RER REIAY IDGIC FUNCTIOBULL TEST.

j LOAD FUEL.

CORE VERIFICATICIt.

! SINGLE ROD SURCRITICAL CEE G.

RILL ndAIN STEAM LINES.

REASSERGELE REACTOR VESSEL.

j INSTALL FUEL POOL GATE, POST REIAAD. )

l DEMATER IMMER CAVITY AND RETWEEN IVEL POOL GATES.

REACTOR LEVEL AT 210 TO 215.

REle0VE CATTLE CEUTE. I DYIRAMIC VOTES TEST 015 2E12-F0425. l l

l l I

+

4 .

6 APPENDIE A CRITICAL PATE ACTIVITIES i REFUEL OUTA M (L2R06) l s DYNAMIC VOTES TEST W 2E12-F042C.

DIY 2 BCCS INJECTICE VALVES OPERABLE.

4 DYNAMIC VOTES TEST W 2E12-F042A.

REINSTALL DRYMELL EEAD.

DRYMELL EEAD LEAR RATE TEST.

, REACTOR MAThR LEVEL ABOVE MAIN STEAM t m a. I PRESSURIEE TO 40.5 PSIG NCE INTEGRATED LEAR RATE TEST.

DEPSSSURIEE Pa m mv CONTAIIBGENT TO 2.5 PSIG.

' VENT SUPPRESSICK POOL TO A'DADSPEERIC PRESSURE.

, DEPRESSURIEE PRIMARY CONTAI3BIENT TO 1.5 PSIG.

DRYMELL TO SUPPRESSION POOL RYPASS LEAK TEST.

yminaamY CONTAIIBGENT DEPRESSURIEE TO ATMOSPEERIC PRESSURE.

l l INSPECT Pa m *v CONTAIIBdENT.

! IANIER SSACTOR VESSEL LEVEL NOR EYDRO.

j REFILL FEEDWATER SYSTEM.

STARTUP INST 3ttBIENT MITROGEN SYSTShi.

1 STARTUP REACTOR MkTER CLEANUP SYSTEM.

i IIGE PRESSURE EECESS FLOW CEECK VALVE OPERABILITY TEST.

~

REACTOR P3tESSURE VESSEL EYDRO.

SCRAM TIME SELECTED CONT 3tOL ROD DRIVES.

PERIUsef NEEELY NR FUNCTIONAL TEST'S.

- MODE SMITCE TO STARTUP.

i EtOD NORTE MINIMIEER OPERABILITY CHECK.

PULL 31008 TO CRITICAL.

SEUTDONN MARGIN TEST.

! REACTOR CRITICAL.

! EEATUP NCR RCIC TESTING.

RCIC TURBINE OVERSPEED AT 150 PSI.

i ADJUST RCIC OVERSPEED TRIP SETTING.

' SIEPERICSM RCIC TURRIME OVERSPEED TRIP.

!. RCIC PtB87 OPE 3tABILITY TEST.

! INSTRUMENT RACE CEECE AT 250 PSI.

l START UP MOTOR DRIVEN REACTOR FEEDPt2dP (300 PSI).

! EEAT UP TO 600 PSIG.

500 PSI ECCS PERMISSIVES CLEAR.

1 REACTOR PatESSURF >600 PSIG ON STARTUP.

. INSTILIBGENT RACE CEECE AT *150 PSI.

PULL 2005 UNTIL kYPASSES OPEN.

i CEECK 1388 TO APESi OVERLAP.

l REACTOR P3LESSURE AT 920 PSI.

BADDE SWITCE TO "RUN".

DYNAMIC VOTES TEST OF 2E51-P013, F019, F022 AND F045.

! ROLL HAIN TUstBINE.

SYNCE GSMERATOR TO THE GRID.

i n

I

• APPENDIE D

. REFUELING OUTAM (L2R06)

• sAPITr RELATED CORRECTIVE nsLINTENANCE NoRK reg i Sr8 30DIPREENT DESCRIPTION 950044021-01 37% 2PS-321N5608 "S" SRV ACClBi ION PRESS ALA3Si UP.

950031448-01 512 2RE40-2921X ADAPTER ASSEMBLY PSA-3 BENT, REPLACE.

950033931-01 221 2PS-321N561D LON PRESSURE ALAsti UP AFTER CYCLING.

940062012-01 C11 2C11D001044 LEAK AT CEARSING NIPPIE ABB INST VALVE.

950040707-01 RP 2C718003F APPARENT TRIP DURING Alm TRANSFER.  ;

950046606-01 C51 2RE-C51E600BA B SISi IS SPIKING CAUSING B50RT PERIOD ALARhis. J 950020114-01 521 2PS-321N4133 REPLACE 3/C RER INJECTICE VALVE INTem w r, 950027918-01 VP 2VP02CB mawarwm STAYED CIASED MIEN C/S TAEEN TO TRIP.

950043052-01 E31 2TDS-E31N6153 DIY 2 GRP 1 ISOLATICII DDE TO A BROKEN LEAD.

930048140-01 E22 2E22P3013 REPAIR CR REPLACE 3 DETER AS NEGSSARY.

950013168-01 E22 2E22F342 CIE G VALVE FAILED TO DRAIN.

950033138-01 DG 2D001P CLEAN AND INSPECT.

950033169-01 E12 2E12C300D CLEAN AND INSPECT.

950033163-01 E12 2512C300C CLEAN AND INSPECT.

950033154-01 FC 2PCO3PB CLEAN AND INSPECT.

950032162-01 C51 2RI-C51E605GR COMPUTER POINT B600 DID NOT RESPOND.

930046458-01 DG 20001K CateIECTION TO ENGINE MAS LEAKING.

95CO28507-01 E13 2E13P642 CCIITACTS ON BET 1Y NEED TO BE CLEANED.

950016246-01 E31 2PDS-E31N009A CONT % CTS UNSTABLE DURING SURVEILLANG.

940059105-01 E51 2E51F063 CURRENT TORQUE SWITCE SETTING TOO IDN.

940061655-02 E21 2E21C001 DISASSEbdBLE AND CLEAN LPCS DIOTOR COOLER.

940061421-01 E51 2E51F002 DISASSEMBLE AND INSPECT CEE G VALVE.

940061422-01 E51 2E51F084 DISASSEMBLE AND INSPECT CEE G VALVE.

950038600-01 D18 2D18000 DIScotelECT CABLE C0008ECTOR POR INSPECTION.

930047650-01 DG 2DG01E DISCHARGE PIPING LEAKING AT UNIOtt.

950020676-01 521 2PS-321N023BA REPLACE PRESSURE SWITCE.

930040337-01 521 2B21N4025 PACKING t.wara m ,

J 930047328-01 321 2321F473 CEECK VALVE BAD TO BE tiANUALLY RESET.

! 930047333-01 B33 2333F313A CLEAN ABID INSPECT CEEG VALVE.

l 930047324-01 521 2B21F3283 CEECK VALVE EAD TO BE MANUALLY RESET.

930047325-01 533 2533F311C CEECK VALVE BAD TO BE MAllUALLY RESET.

l 930047326-01 B21 2321F328A CEECK VALVE EAD TO BE MANUALLY RESET.

i 930047332-01 B33 2333F307C CLEAN AND INSPECT CEECK YALVE.

l 930047329-01 533 2333F315A CEECK VALVE BAD TO BE MANUALLY RESET.

950028675-01 521 2PS-321N023AA REPLACE PRESSURE SWITCE.

950011966-01 533 2FI-333K606B CLEALI E-7 CARD.

I 950022629-01 AP 2AP75E SIDE OF DACC CCBIPT E6 EAS REEN mann, d

950033075-01 DG 2DG01K WMenenCY STOP BUTT 005 FAILED TO TRIP DIESEL.

950047509-01 EG 25G193 PIPING LINE 2EG195 IS BENT.

930047912-01 VP 2VP053A FAILED LEAK RATE TEST.

) 950040842-01 521 2B21F571 VALVE DIRTY.

l 950029709-01 C11 2C11F4230 FAILED EIGE PRESSURE SEAT r.warme yggy, 950029785-01 E32 2E32F001A FAILED LEAK RATE TEST.

! 950029505-01 421 2521F022A FAILED I.EAK RATE TEST. l

. 950029506-01 321 2821F020A FAILED LEAK RATE TEST. ,

950016994-01 VQ 2VQO27 FAILED LEAK RATE TEST.

l 950033244-01 E12 2E12F0413 FAILED TO CIASE DURING SURVEILLANG TESTING. >

l 950042075-01 R21 2B21F0325 FAILED TO CYCLE DURING SURVEILLANCE TESTING.

l 9500h D67-01 C11 2C11D001019-107 ECU QUICK DISColetECT FAILED.

950020968-01 IT 2ET03EE EEAT TRAG IS NOT ARRANGED CORRECTLY.

950050402-01 E51 2551F063 Bor TORQUE BonnetTT RIOTS. ,

950020506-01 E22 2E22C003 INCREASED VIRRATICII, REPIACE PtBEP BEARINGS.

i 950038601-01 AP 2AP19E-102A INSPECT ABID ADJUST RACKING ndECIANTEM.

930047327-01 533 2333F301B INSPECT, CLEAN AND REPAIR CIECK '. '7E.

l 930047331-01 B21 2B21F570 INSPECT, CLEAN AltD REPAIR CEECK V uYE. i 950030399-01 E31 2TDS-E31N614C INSTRUMENT FAILURE.

940061726-01 513 2513P620 REPAIR INSULATED WIRE LUG.

J"

.- A,PMDIE o

. REFcELING ouTha*. (L2206)  ;

• SAFErr REIATED C003ECTIVE R$kINTENANCE  !

NomK Rag # SYS RooIstenu ossenPT CII l

950045656-01 C51 2EL-C51A5000lD 1888 D ERRATIC OPERATION AND INDICATXCK.

950031481-01 C51 2RR-C512603C Z388 E DIGIThL DISPLAY ttISSING SEmdENTS.

950027950-01 B21 2321F013A JUNCTION SCE IS DAREAGED.  ;

950004615-01 E12 2E12F065A GAGGING DEVICE NOT umums.

950028966-01 BIS 2bdS000 LINE EAS ERECDED BEYOND ACCEPTABLE LIntITS.

940070160-01 C51 2C51000 EIGE NOISE LEVELS CIE SEBAS ARG 13385.

940057292-01 DC 2DC17E M&T WEEN C/S TAREN TO EQUALIEE.

950044678-01 IIR 2513P603 LP388 II ALh308 IS UP.  ;

940050672-01 E32 2332F001N tikINSTEAtt LIIIES IRAK BY.

950022739-01 G33 2PDIS-G33N025 REPLACE DEICRO SWITCE #2. l 940058676-01 E32 2E32F002N tikINSTEAM LINES LEAK BY.

950031471-01 C11 2C11F402A NO LIGIET INDICATICIt.

940079588-01 E12 2E12F0903 NO OPEN INDICATICIf IN CCIITROL 20008.

950045723-01 C51 2RY-C51K6015 310 RESPONSE FRCat 1308 3 CN RANGE 5.  ;

940060399-01 DS 20001K OIL SIGETOLASS CII LUBE OIL COOLER LEAKING.

950000216-01 B33 2ESO-533F020 CIII LIBEIT SWITCE FAILING.

950030653-01 333 2333F3435 OPEN INDICATICII NOT WMEMG. ,

950040481-01 E12 2512C300C 03T30ARD BEARING IS BOTTER TLAN IICOBdkL.  ;

950030374-01 AP 2AP715-C2 REPLACE OVERIA3AD ARID BREAEER. l 950036734-01 E22 2E22F375A PACKING LEAK. l 950041290-01 E12 2E12F008 PACKING LEAK. i 950043150-01 IN 2IN114 PACKING LEAK. l 940059501-01 E51 2E51F046 PACKING LEAK.

950040080-01 B21 2821F067D PACKING LEAK, 950040719-01 321 2R21F067A PACKING LEAK.

940060105-01 G33 2G33F040 PACKING LEAK.

940060969-03 DG 2D0060A PERICIti INSPECTICIf.

950040854-01 321 2321F026A PIPE CAP AT VALVE IS LEAKING.

950040054-02 521 2321F025A PIPE CAP AT VALVE IS LEAKING.

950022849-01 C71 2PS-C713f002C P3 ESSURE TEST SWITCE.

f 950034809-01 E12 2312C0023 PEREP SEAL COOLER LINE LEAKING.

950040645-01 IN 2IN038 REGULATOR LEAKING BY SEAT.

l 950033569-01 DG' 2DG03J RELkY FAILED CALIRFATICOf.

! 950014886-01 E13 2E13P632 MLAY IS WORKING INTEst(ITENTLY.

l 950025878-01 C71 2C71AE16C MLAY TIMING IS ERRATIC.

950039180-01 321 2321EK1C RELkY FAILED TO tihEE UP WEEN TRIPPED.

i 940061317-01 LV 2Lv000 REMOVE PROTRUDING CARLES FRCat PENETRATICIt.

j 940061019-03 E12 2E12D3005 RE300VE END PLATE AND INSPECT.

l 950032754-01 E12 2E12C3005 INSPECT CUT 50ARD PtadP BEARING.

950022348-01 RE 2R5000 reb 8DVE PIPE CLAbdP FOR SUPPORT B504-28688.

i 940061226-01 C11 2C11D001088 RE30DVE/ REPLACE CCIITROL RCD DRIVE 10-07.

, 9500.M530-01 AP 2AP025-0 REPAIR BREAEER GUIDE.

! 950026239-01 AP 2APO45-01 REPAIR CRACE3D ABC CEUTE.

l 950032658-01 AP 2AP915 REPAIR EOLE NOUND IN NCII-SEG BUS DUCT.

. 940081131-02 C51 2C51J003 REPAIR TURING F3Gi DRYMELL MALL 20 IIIDEEER.

940061222-01 C11 2C11D001145 REbdDVE/REPIACE CCIETROL RCD DRIVE 34-31.

940061223-01 C11 2C11D001155 RE30DVE/MPLACE CCIITROL RCD DRIVE 54-23.

i' 940061228-01 C11 2C11D001181 reb 80VE/ REPLACE CCIETROL RCD DRIVE 30-03.

l 940061225-01 C11 2C11D001133 RaneDVE/ REPLACE CCIETROL 200 DRIVE 50-35.

950027920-01 AP 2AP755-D3 REPLACE DENECTIVE manren.

} 950026655-01 AP 2AP08E-2 MPLACE DEFECTIVE LUG.

[ 950018720-01 DC 2DC03E REPLACE ALA388 BOARD CIf 250V BATTERY CIARGER.

950033547-01 C71 2C718300D REPLACE TEE WGIC CARD.

i~ 950017190-01 321 2321F028A REPLACE TEE 3-WhY == === VALVE.

! 950020166-01 C71 2C718003A BEPLACE "A" RPS BAS SET EPA M GIC CARD.

! 950023756-01 ,AP 2EI-AP077 REPLACE VOLTAGE RELhY.

940058358-01 E31 2PDS-E31N008A REPLACE RI FMIf SWITCE 3 VALVE NANIFOLDS.

i

. APPENDIE D

• • REFUELING OUTAes (L2206)

SarETY RELATED CORRECTIVE BOLINTENANCE

! NORK REQ f SYS EQUIBMENT DESCRIPTIM 940058358-02 E31 2PDS-E31N009A REPLACE II FLOW SWITCE 3 VALVE MhMIFOIDS.

940058358-03 E31 2PDS-E31N010A REPIACE II FIAN SWITCE 3 VALVE MANINOLDS.

940058358-04 E31 2PDS-E31N011A REPLACE EI FIDW SWITCE 3 VALVE MhKINCIDS.

940058358-05 E31 2PDS-E31N0005 REPLACE EI FIAN SWITCE 3 VALVE MANIFOLDS.

! 940058358-06 E31 2PDS-E31N0095 REPIACE EI FIDW SWITCE 3 VALVE MhMIFOLDS.

940058358-07 E31 2PDS-E31N010B REPLACE EI FIAN SWITCE 3 VALVE MANINOLDS.

940058358-08 E31 2PDS-E31N0113 REPLACE RI FIAN SWITCE 3 VALVE MANINDIDS.

940058358-09 E31 2PDS-E31N000C REPIACE II FIDW SWITCE 3 VALVE MANIFOLDS.

940058358-10 E31 2PDS-E31N009C REPLACE II FLOW SWITCE 3 VALVE MANINOLDS.

. 940058358-11 E31 2PDS-R31N010C REPLACE II FIDW SWITCE 3 VALVE MhWIFOLDS.

940058358-12 E31 2PDS-E31N011C REPLACE II FLOW SWITCE 3 VALVE MANINDIAS.

940058358-13 E31 2PDS-E31N008D REPLACE II FIDW SWITCE 3 VALVE MANIFOIDS.

j 940058350-14 E31 2PDS-E31N009D REPIACE BI FIAN SWITCE 3 VALVE MhWINCIDS.

940058358-15 E31 2PDS-E31N010D REPLACE II FLOW SWITCE 3 VALVE MhMINOLDS.

, 940058358-15 E31 2PDS-E31N011D REPIACE II FIDW SWITCE 3 VALVE MhMIFOLDS.

l 940056944-01 E13 2513P628 RELkY DID NOT rent IEE DURING TESTING.

I 940058078-02 333 2333F060A REPLACE ACTUATOR, INCORRECTLY MhCEINED.

950000973-01 521 2LT-521N407A ROSEMOUNT TRANSMITTER EERO SEIFT OF >.50.

950028347-01 RE 2RE024 FAILED IOCAL LEAK RATE TEST. I 1 930046719-01 AP 2AP07E SECNna*Y CONTACTS ARE BAD.

950033389-01 321 2EY-5211752 TEST POWER SUPPLY.

950033390-01 C22 2JY-C22K601B TEST POWER SUPPLY.

4 930047211-01 C11 2C11D045115 DEGRADATION DURING SURVEILIANCE TESTING.

930046621-01 E22 2E22C001 SMALL LEAK ON MIST SIDE OF PUMP.

l 950039407-01 521 2B21F032A SOLENOID BLOWING AIR CONTINOUSLY.

l 940061960-01 W 2TE-3057A TEMPERATURE INDICATION FAILED.

950027315-01 C11 2C11D001128-112 VALVE FAILED TO 90 IULL CLOSED.

l 950027316-01 C11 2C11D001113-112 VALVE FAILED TO GO IULL CLOSED.

950027317-01 C11 2C11D001142-112 VALVE FAILED TO 90 FULL CICSED.

930042804-01 V9 2VQ026 ACTUATOR REPr.arenT.

950019431-01 521 2PS-321N0155 SWITCE FAILED RESPONSE TIME TEST.

. 950017191-01 E31 2PDS-E31N000A SWITCE Nh8 ERRATIC DURING SURVEILLANCE.

i 950017192-01 E31 2PDS-E31N0085 SWITCE MAS ERRATIC DURING SURVEILLhMCE.

j 950043250-01 IN 2IN115 SYSTEM LEAKS.

950024117-01 C22 2JY-C221601A TEST DIV I POWER CONVERTERS.

950022091-01 VR 2ES-VR002AE RELAY FAILED TO ACTUATE.

950030654-01 E32 2E32F007 TROUBLESBOOT AND REPAIR FOR EIGE AMPS.

950033950-01 533 2333F3455 TROUBLESBOOT AND REPAIR. ,

950046294-01 E51 2E51C001 CONTROL ROOM INITIATED TRIP FAILED TO RESET. l 950034208-01 EP 2EP02-2801C UNPIN PIPE SUPPORT.

950041299-01 E12 2E12F009 VALVE PACKING LEAK.

950040848-01 E32 2E32F001N VALVE PACKING LEAK.

950043051-01 E51 2E51F091 VALVE FAILS TO CYCLE.

950030184-01 G33 2G33F101 VALVE PACKING LEAK.

930047112-01 E12 2E12F092C VALVE SEAT T21Ekm.

950002729-01 521 2321N413D-ES VALVE traram , J 950050285-01 E51 2LSE-E51N010 VALVE FAILED TO OPEN ON EIGE LEVEL.

930047336-01 B33 2333F342A VALVE FAILED TO INDICATE OPEN.

930046196-01 DG 2DG004 VALVE SEAT AND PACKING TMhram.

940061644-01 E12 2E12F3325 VALVE OPERATOR IS NOISY AND EARD TO OPERATE.

950037085-01 E32 2E32F302E VALVE PACKING Tsaratar, 950039231-01 C11 2C11D009127 VALVE PACKING traram ,

940058961-05 E12 2E12F017A VALVE BODY TO BotBET FLANGE LEAK.

930047223-01 C11 2C11D129101 VALVE raaram, 950050082-01 C11 2C11D001001-120 DIRECTIcetAL CONTROL VALVE T21***.

940073299-12 521 2321F4183 REFURB LIMITORQUE ACTUATOR.

940061172-02 333 2333F023A REPLACE VALVE INTS3 GULLS.

/ ' APPENDIE D o REFmfTMG OUTAGE (L2206)

O SAFETE RELATED CORRECTIVE MAINTENANCE I

i Noms reg i srs EQUIsamurr DESCRIPTION ,

I 950016184-01 E12 2E12F041C VALVE REMAINED OPEN AFTER CYCLING.  !

l 950014855-02 E51 2E51F020 FAILED LOCAL IZAK RATE TEST. <

950014858-01 E51 2E51F040 FAILED IACAL IZAK RATE TEST. l 950047657-01 E51 2E517063 VALVE LEAKING. I 950004712-02 IN 2IN017 VALVE PACKING LEAK. l 950016991-01 VQ 2VQO26 FAILED ICCAL LEAK RATE TEST.

930047564-01 AP 2AP099A TARGETS FAILED TO DROP CN AN UNDERVOLTAGE.

950025559-01 RE 2RE03-2889E CLAMP IS NOT MAKING BUFFICIENT CONTACT.

l I

1 l

f b

l l

I i

l 1

j