ML20059F252

Summary of Nuclear Station Mods,Minor Mods & Procedure Changes Made to Plant for Period of 920401-930401

ML20059F252
Person / Time
Site:	McGuire, Mcguire
Issue date:	10/25/1993
From:	Mcmeekin T DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9311040194
Download: ML20059F252 (51)

Text

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. 11 1 Duke 1%wr Company .

T C Al<stetxtN

- hicGuire Nuclear Generation Department Vice 14esident 12700 Hagers Terry Road (h1G01A) . . (704)Ri34800 '

flantersville, NC 28078 M85 (704)375-4809 l'Ax -

h DUKEPOWER October 25,1993 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555

Subject:

McGuire Nuclear Station Docket Nos. 50-369 and 50-370 Pursuant to 10 CFR 50.59, please find attached a summary of Nuclear Station Modifications, Minor -

Modifications, and Procedure Changes made to the McGuire Nuclear Station for the period of April 1,1992 to April 1,1993. ,

Questions or problems should be directed to Kay Crane, Regulatory Compliance at (704) 875-4306, ,

Very truly yours,

/

W T. C. McMeekin, Vice President McGuire Nuclear Station Attachment cc: Mr. Victor Nerses, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Mr. S. D. Ebneter, Regional Administrator U. S. Nuclear Regulatory Commission Region 11 101 Marietta Street, NW - Suite 2900 Atlanta, Georgia 30323 Mr. George Maxwell Senior Resident inspector -

' McGuire Nuclear Site 030072 o i 9311040194 931025 E  :

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'6 Duke Power Company McGuire Nuclear Station Summary of Nuclear Station Modifications Completed Under 10CFR50.59 MG-12324/00 and MG-22324/00

Description:

The purpose of these modifications are to prevent pumping potentially contaminated water from the-Refueling Water Storage Tank (FWST) trench, by disabling FWST Sump Pump A should excessive overflow from the FWST occur.

PIR 0-M88-0010 identified that the FWST overflow line capacity was insufficient to relieve the normal input flow to the FWST in the event of overfilling. This coul0 mult in failure of the FWST due to overpressurization.

MEVN-1316 and 1317 were written to correct this problem by modifying an existing vent line in the tank so that the vent line may be used to relieve the excess input flow. The liquid discharged from the vent' line will be directed to the FWST trench.

FWST Sump Pump A, located in the FWST trench, is used to pump rainwater to the yard. Routing the

' tank overflow to the trench introduces the possibility of pump contaminated water to the yard by the FWST Sump Pump.

These modifications will add an additional current alarm in the level loop (FWLT5341) to prevent automatic operation of the sump pump in the event of FWST overflow. A computer alarm will also be added on the Operator Aid Computer (OAC) in the Control Room to indicate an excess overflow condition. A manual reset will be provided for pump operation once the water has been analyzed.

The level indication circuitry is non-safety, and does not affect the ability of the FW system to perform its safety functions. Changes will be made on the non-safety control panel in the yard, which houses the electrical circuitry for the instrument loop. A nonefety enclosure to hose the reset circuitry will be located near the pump.

Safety Review and USQ Evaluation:

j The FWST is not an accident initiator. The added overflow protection circuitry and the modified control .

j panel are non-safety. Seismic qualification or mounting of the added equipment is unnecessary. An 1 Appendix R review was conducted, with no concerns identified. Therefore the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.-

The FWST is used as a source of borated water for refueling operations, and for mitigation of a loss-of-coolant accident, The overflow protection circuitry has no safety function, however, will be used to prevent the potential spread of contaminated water into the yard area of the plant. The added circuitry does not interface with the Refueling Water Storage Tank itself, or any associated safety-related equipment. The performance of plant safety functions will not be prevented, nor degraded by the modifications. The FWST -

will continue to perform the same functions. The FWST Sump Pump A will continue to perform its non-safety function of pumping rainwater out of the FWST trench, except in case of FWST overflow. When analysis of the water is desired prior to pumping the water out. No common failure modes are created.

Therefore the consecuences of an accident of malfunction or equipment important to safety evaluated in the SAR is not increased.

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l No new failure modes are created by these NSMs. If the added current alarm fails to operate simultaneous with a tank overflow condition , the potential will exist (due to MEVNs 1316 and 1317) for pumping mildly radioactive water into the plant yard. The water will be carried by the yard drain system to the waste ,

treatment system Effluent releases from the waste treatment system are monitored prior to offsite release to ensure regulatory limits are not exceeded. No accidents previously thought incredible are made credible by the NSMs. Thus the possibility of an accident or malfunction of equipment of a different type than evaluated in the SAR will not be created.

There are no changes of safety limits, setpoints, or plant parameters because of the modifications. The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSMs. Therefore, the margin of safety as defined in the basis for any Technical Specification is not decreased. No USO exists. ,

MG-12402/0

Description:

This NSM will replace certain instrumentation related to the DG fuel oil day tank. This instrumentation

• controls the status of the fuel oil transfer pump to maintain tank level between 43" and 51" and provide bl/lo alarms at 54" and 40". Also, the instrumentation provides a local alarm at the DG Control Panel, a

" Diesel Engine Trouble" alarm in the control room and computer alarm.

The existing level switches are manufactured by Barton and the replacement instrumentation will be Rosemount. The level switches are addressed in section 9.5.4, Fuel Oil System, of the FSAR. The replacement instrumentation is more reliable as a result of their design. Currently, each day tank (1 A and 1B) has two level switches. One of them is safety related, controlling operation of the fuel oil transfer pump between 43" AND 51" tank level. The other, non-essential, provides hi/lo alarms at 54" and 40" and other miscellaneous non-essential signals. The only safety related control function is the fuel oil transfer ,

pump on/off actuation. 7 The replacement instrumentation provides the same functions through a ditferent configuration. The safety related instrumentation also provides the non-essential current alarm utilizing an optical isolator to effect the required electrical separation. The new instrumentation has been environmentally qualified. The related tubing is seismically qualified. A new cable will be pulled for future use to facilitate a pressure gage replacement. No Appendix R concerns are identified.  ;

Safety Review and USQ Evaluation:

i The level instrumentation supports the DG providing fuel via the fuel oil transfer pump and as such is accident mitigation equipment. The instrumentation is not an accident initiator. The new instrumentation is more reliable and has no failures that result in different fuel oil transfer pump operation than were possible before. Therefore, the probability of accidents previously evaluated in the FSAR are not increased and no new accidents or malfunctions of equipment are created.  :

I The function of the instrumentation has not changed except that the replacement safety related instrumentation provides all functions previously provided by a safety and non-safety switch. The new instrumentation l'as a history of being more reliable and requires re-calibration lecs often. The safety functions have been preserved with replacement instrumentation of hjuivalent qualifica; ion and criteria.-  ?

Therefore, the probability of a malfunction of equipment important to safety evaluated in the FSAR has not increased. -

The accident mitigation capability of the DGs has not been affected by this NSM. No new failure modes have been introduced. Therefore, the consequences of accidents or malfunctions of equipment have not ]

increased.

Margin of safety is related to the confidence of the fission product barriers. There is no change to the fuel, cladding, RCS pressure boundary or containment. No assumptions in any accident analysis have been affected nor have any setpoints in the control of day tank level. Therefore, the margin of safety defined in the basis to the Technical Specifications is not reduced. No USO exists.

MG-52171

Description:

The 125 VDC Vital Instrumentation and Control Power System battery banks will be replaced with new batteries and battery racks, by a series of plant modifications. The modification numbers and corresponding battery designations are as follows:

NSM Battery Bank MG-52168 EVCA MG-52171 EVCD MG-52174 EVCC MG-52176 EVCB The existing batteries are being replaced because they are nearing the end of their useful life. The battery racks will be replaced because of deterioration and corrosion.

Required battery ratings and equipment capacities were verified. The instantaneous settings on incoming and tie circuit breakers will require change. Existing conductors are appropriately sized, therefore, will not require replacement. The replacer sent batteries are purchased to meet GA Condition 1 requirements with 10CFR50, Part 21 dedication prot ided by AT&T.

Replacement of each battery bank will require 3 weeks, and implementation will occur during normal plant operations. During the 3-week duration for each battery bank replacement, a temporary battery will be installed and connected as a backup for the battery being replaced. Initially, one of the new battery banks ~

will be used as the backup battery. After the first battery bank is replaced, the replaced battery will be used as the backup battery. The temporary battery bank will be installed in the Shared Equipment Room of the Service Building. Temporary cabling will be routed from the temporary battery bank to the Standby Battery Charger electrical bus, located in the battery room. The Standby Battery Charger and temporary battery bank will be connected in place of the battery bank being replaced and its associated battery charger. ,

Following each 3-week changeout period, the 125 VDC Vital Instrumentation and Control Power System will be returned to the normal configuration.

t Brief description of the 125 VDC Vitall&C Power System: -i The 125 VDC Vital Instrumentation and Control power System (system designation EPL) is divided into  !

'i four independent and physically separated load groups, each consisting of a battery bank, battery charger, and associated distribution equipment. The EPL system is a shared system, with each load group 1 supplying both Unit 1 and 2 loads. A single failure at any point does not disable more than one vital - l channel. i i

System design provides for manual connection to two distribution centers during periods of battery maintenance, or emergency. A spare battery charger is also available for use during equipment maintenance.  ;

I I

The EPL system is designed to meet GA Condition 1 requirements. The batteries and their related accessories are located in the Battery Room, located in the Auxiliary Building, which is a Category I structure. The Battery Room is divided into individual rooms for each battery bank.

The EPL system provides a source of reliable continuous power for safety-related control and instrumentation required for start-up, normal operation, and orderly shutdown of each unit. Following is a summary (not all inclusive) of the supplied loads:

Diesel Generator Load Sequencers 4 KV Switchgear Controls Engineered Safety Features Controls Auxiliary Safeguards Controls Process Protection System Post Accident Monitoring Recorders Nuclear instrumentation System Power 600 Volt Load Center Controls Reactor Coolant System Solenoid Valves Safety Review and USQ Evaluation:

Battery Replacement t The 125 VDC Vital Instrumentation and Control Power System is not an accident initiator, however, it serves as an accident mitigation system. The replacement batteries are purchased to meet OA Condition 1 requirements with 10CFR50, Part 21 dedication provided. The new batteries and racks will be seismically mounted. There is no change in cabling required for the new batteries. An Appendix R review was performed with no concerns identified. Taere is no change in the physical and electrical separaton provisions for the batteries. The performance of plant safety functions will not be degraded by the new batteries. Therefore, there is no significant increase in the probability or consequences of an accident previously evaluated.

There are no new or common failure modes created by the new batteries. The new batteries perform the r same function as the existing batteries. The existing batteries are nearing the end of their useful life, therefore, new batteries are expected to be more reliable than the existing batteries. The possibility of a new or different kind of accident is not created by the battery changeout.

A Technical Specification change is involved with the modifications to add a table with the new battery specifications. The Diesel Generator Sequencers are powered by batteries EVCA and EVCD. The Diesel' Generator Sequencers have the potential to impact the performance of all safety-related equipment that ,

rely on AC power to function (ECCS pumps, valves, etc.). Ultimately,3 of the 4 fission product barriers -

fuel, cladding, and containment - rely on the proper functioning of these components to preserve their integrity (The RCS pressure boundary is not included, as it is passively protected by the code safety valves, which have a safety limit setpoint of 2735 psig, per Technical Specifications). The new batteries '

are as capable of performing their accident mitigation function as the old batteries, and are fully qualified.

Thus the battery changeout does not involve a significant reduction in the margin of safety.

Temporary Battery Installation implementation of each battery replacement will require a 3-week duration of time. During the replacement period a safety-grade battery bank will be connected to substitute for the battery bank being >

replaced. The temporary battery will be installed in the Shared Equipment Room of the Service Building, because space is unavailable for locating the temporary battery in the battery room. The Service Building is not a Seismic Category 1 structure, therefore, the temporary battery mounting will not be seismically qualified. The 125 VDC Vital Instrumentation and Control System will be restored to the fully qualified

configuration following each 3-week battery replacement period. 'This condition has been discussed in a telecon with NRC staff personnel. Due to the relatively short timeframe for battery replacement, evaluation of the temporary battery installation will exclude seismic considerations.

The 125 VDC Vital Power System is not an accident initiator. However, it is an accident mitigation system.

during each battery replacement, the other 3 batteries and associated distribution equipment will remain in their normal configuration. The performance of their safety functions will not be degraded.

The ability to cross-tie the electrical buses for the 125 VDC Vital Batteries, by manual action, remains available as a backup option in the event that the temporary battery is tendered unavailable during the '

replacement periods. However, it should be noted that cross-tying the catteries after an accident, should the need arise due to failure of the temporary battery, is outside the bounds of the current accident  ;

analyses, unless it can be done witnin 11 seconds (the required time fr r the diesel generators to start and '

loads to be sequenced on). The ability to period this manual action !s stated in McGuire Nuclear Station ,

FSAR, Section 8.3.2.1.4.3. FSAR Section 8.3.2.1.4.2 states that each battery is sized to carry the continuous emergency loads and anticipated momentary loads of its own vital buses, and assume the loads of another battery in a backup capacity for the required one hour period. -Technical Specification 3.8.2.1 discusses the limitations for this configuration during normal operation. This cross-tie function can be performed without tying the opposite trains to the same bus. If the cross-tie was made prior to commencing battery changeout (with the Technical Specification operational limit extended for the duration of the changeout), all loads would be supplied by fully qualified batteries, and the configuration would allow the 11 second diesel generator start-up and sequencing time limit to be met.

The ambient temperature of the Shared Equipment Room will be periodically monitored by McGuire Operations personnel, to ensure the ambient temperature remains within battery specifications. The ventilation in the Shared Equipment Room is sufficient to prevent accumulation of excess hydrogen.

Therefore, with exclusion of seismic considerations, the temporary battery installation does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The temporary backup battery will be a new battery for the first replacement modification, and for each subsequent modification will be a replaced battery, in eitler case, the temporary battery and rack will be the same qualified equipment as normally used. There a e no new failure modes created for the batteries  ;

and associated distribution equipment not involved in flie particular changeout modification. With the ,

temporary battery connected, there are no new failure mo0es for the distribution equipment associated with the battery being replaced. The new failure mode fu the temporary battery installation (no seismic mounting, is considered insignificant), due to the shcit duration for which the temporary configuration will be used. Therefore, the possibility of a new or different kind of accident from any accident previously .

evaluated is not created.

There is a Technical Specification change associated with the temporary battery installation, to include the  ;

specifications of the new battery that will initially be used as the backup battery. As stated, the potential to impact fission product barriers exists, but excluding seismic considerations, no fission product barrier will be degraded. Thus, there is no significant reduction in the margin of safety.  ;

FSAR and Technical Specification revisions are required to include the specifications for the new batteries.

Since the modifications involve Technical Specification changes, the USO criteria for 10CFR50.59 are not l applicable to this evaluation. Instead the Technical Specification changes must be approved per i l

10CFR10.91 and 10CFR50.92 before the modifications are implemented.

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MG-22393/00

Description:

This modification upgrades the 2" Main Steam (SM) system drain lines and the 4" Main Feedwater (CF) system tempering lines, from the first block valves in the doghouse to the Turbine Building. The existing Class G piping will be analyzed and supported such that it meets Class F requirements.

This NSM is in response to PIRs 0-M91-0088 and 0-M91-0100. The PIRs were initiated to evaluate potential failure of the subject lines due to seismic loadings and tornado wind / missile loads.  ;

Safety Review and USQ Evaluation:

FSAR Section 15.1.5 addresses Steam System Piping Failure, and Section 15.2.8 addresses Feedwater System Pipe Break accidents, which involve the SM and CF systems. This NSM provides stress-analysis

• and design for upgrading the subject SM and CF system piping and supports from Class G to Class F.

This analysis and upgrade will satisfy the concerns of PIRs 0-M91-0088 and 0-M91-0100, which were initiated to evaluate potential failure of the subject lines due to seismic loadings and tornado wind / missile loads. The Class upgrade will increase the reliability of the subject lines for seismic and tornado events, precluding potential breaks of these lines in the vicinity of the DG air intakes. Therefore, the probability of an accident of malfunction of equipment important to safety previously evaluated in the FSAR is not increased.

These parts of the Main Steam and Main Feedwater systems are not used for accident mitigation. The CF Tempering Flow is important to safety as it protects the CA nozzles against thermal shock. The  ;

performance of plant safety functions are not degraded by the modification. The functions performed by the upgraded piping remain unchanged. There are no common failure modes created by this modification.

Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR is not increased.

The upgraded piping does not perform any new functions. There are no new failure modes created. No accidents previously considered incredible are made credible by these NSMs. The upgraded piping will be qualified for seismic and tornado wind loads. Thus the possibility of an accident or malfunction of equipment of a different type than evaluated in the SAR will not be created.

There are no changes of safety limits, setpoints, or plant parameters because of the modification. The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSM. Therefore the margin  ;

of safety as defined in the basis for any Technica! Specification is not decreased. No USO exists.

)

MG-12396 and MG-22396

Description:

These modifications are for the replacement of existing Diesel Generator grounding transformers and resistors with equivalent equipment having a higher thermal rating.

The existing grounding transformers were sized based on grounding calculations that did not include  :

consideration of the 4.16 kV system cable and load capacitance. - As a result, the thermal rating of the existing equipment is insufficient to operate continuously with the DG supplying its load during a 4.16 kV system ground fault. ,

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' The inadequately rated equipment was found during a Self-initiated Technical Audit of McGuire's 4.16 kV Essential Auxiliary Power System (EPC). .This modification is a result of the finding.

The existing grounding transformers and resistors are safety-related (Class 2E). The replacement equipment will also be Class 2E. OA Condition 1 applies to the equipment.

The existing DG grounding equipment is located in the DG rooms. The replacement equipment will be re-located within the DG rooms. An Appendix R Review was conducted with no concems identified.

Settings for the overvoltage relays were reviewed, and will be revised to match the characteristics of the replacement equipment. Ground faults on the 4.16 kV system are detected by these relays, initiating control room and computer alarms. The DG breaker is also tripped if the DG is paralleled with the normal or standby power source (test alignment) at the time of the fault. The DG breaker is not tripped when the DG is aligned with the essential bus at the time of a fault.

Heat load calculations for the DG rooms were reviewed and updated. A seismic review was conducted for the anchorage of new equipment.

Safety Review and USQ Evaluation:

There are no FSAR Chapter 15 accidents or events initiated by the emergency diesel generators or associated equipment (EPC system). Thus, the EOC system is not an accident initiator. The replacement DG grounding equipment is qualified in accordance with OA Condition 1 criteria. Heat load calculations for the DG rooms were reviewed and updated. A seismic review was conducted for replacement equipment anchorage. The replacement equipment is expected to be more reliable than the existing equipment due to higher thermal rating capacity, which envelopes the heating effect of the higher calculated fault currents. An Appendix R review was conducted, with no concems identified. Therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased.

The emergency diesel generators supply emergency power for equipment used for safe plant shutdown and accident mitigation in the event of a Loss of Offsite Power (LOOP). The replacement equipment performs the same functions as igjting equipment, with only a change in thermal rating capacity.

The safety functions performed by the EPC system are not degraded by these modifications. Each DG and associated equipment are separated from and are independent of the other DGs. The replacement equipment for each DG will be located within the DG rooms for the corresponding DG, as is the existing equipment. There are no common failure modes created by the modifications. Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR is not increased.

There are no new functions added by the modifications. No new failure modes are created. No accidents previously considered incredible are made credible by these NSMs. The replacement equipment is expected to be more reliabs for the required application than the existing equipment. Thus the possibility of an accident or malfunction of equipment of a different type than evaluated in the SAR will not be created.

There are no changes of safety limits or plant parameters because of the modifications. The overvoltage relay setting change will not cause a DG trip off-line during emergency use. The fission product barriers (RCS boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any  :

accident analysis are affected by the NSMs. Therefore, the margin of safety as defined in the basis for any Technical Specification is not decreased. No USO exists.

MG-12196/00 and MG-22196/00

Description:

These NSMs will add selector switches on Electrical Boards 1EB6 and 2EB6 (one switch per electrical i

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board) in the Control Room to block the initiation of the high speed printing mode on the existing Esterline-Angus (EA) recorders, at operator discretion.

• The EA recorders are used for monitoring Unit Main Power System voltage and current.. They are typical chart recorders but with an added feature of a "high speed printing mode". When sensors in the recorders detect a transient, the EA recorder shifts to the high speed mode for better print resolution. When the transient has cleared, the sensors reset the EA recorder to the normal low speed operation.

The NSMs will add a cutoff switch to the control circuit for each recorder to allow the operators to block initiation of the high speed printing at their discretion. This will save paper and reduce unnecessary wear of the paper feed motors in case of sensor failure.

• A control switch with an indicating light will be added on 1 EB6 and one also on 2EB6 near each recorder.

The Electrical Boards are part of the Main Control Boards in the Control Room. . The added switches will ,

be seismically mounted (OA Condition 4). Only minor rewiring internal to the EA recorders and from the added switches to the recorders is necessary. No cabling is required. The control switch and wiring will be installed in accordance with separation criteria applicable within OA Condition i electrical panels.

Safety Review and USQ Evaluation:

The Electrical Boards are part of the Main Control Boards. The Main Control Boards are designed as a OA Condition 1 structure. The added selector switches are the EA recorders perform no safety functions, ,

and are not accident initiators. Seismic mounting of the added non-safety switch is required (OA Condition 4), because the Main Control Boards are OA Condition 1. A seismic review of the switch mounting and effect of the added weight on the Main Control Boards was performed, with no concerns identified. An Appendix R review was performed and no concerns were identified. Minor rewiring within the EA recorders and Electrical Boards 1(2)EB6 is required. The selector switch and Electrica: Board wiring will be installed in accordance with separation criteria applicable within OA Condition 1 electrical panels. Therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.

The EA recorders are not an accident mitigator. The EA recorders and added selector switches perform no safety functions. The function of the EA recorders is not changed by this modification. Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR is not increased.

The new operating characteristic of the EA Recorder has no significant effect on any SSCs evaluated in the FSAR of addressed in the Technical Specifications. No accidents previously thought incredible are made credible by these NSMs. Thus the possibility of an accident or malfunction of equipment of a different type than evaluated in the SAR will not be created.

There are no changes of safety limits, setpoints, or plant parameters because of this modification. The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by this NSM. Therefore, the ,

margin of safety as defined in the basis for any Technical Specification is not increased. No USQ exists.-

MG-12406

Description:

The Condensate Air Ejector Radiation Monitor (1 EMF 33)is designed to provide early indications of steam generator tube feaks. In its present configuration,1 EMF 33 may not be capable of performing this function is all cases. Due to the location of the sample tap, the CSAR off-gas is significantly diluted with the gland steam exhaust, resulting in lower sensitivity readings for the CSAE off-gas. In order to adequately monitor

m J the CSAE off-gas for early indications of tube leaks, this modification will relocate the sample tap from  ;

downstream of the gland stream exhaust to a location between the CSAE header and the bland stream

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exhaust, thus removing the dilution factor caused by the gland stream exhaust. The gland seal air ejector" exhaust is monitored by the unit vent monitor; the reading from the unit vent monitor is used to show l regulatory compliance. The CSAE off-gas is also monitored during accident conditions from the unit vent .

monitors.1 EMF 33 is not used for verifying regulatory compliance. In addition, the 1 EMF 33 skid will be l moved from its current location to a location just below the CSAE header. This will result in a need for ,

a new power supply'and floor loading analysis.

Safety Review and USQ Evaluation:

1 1 EMF 33 is classified as a non-OA condition Process Radiation Monitor. Therefore, there are no electrical i

, separation requirements imposed on the monitors or the control room readouts.' The current configuration of the rnonitor skid includes a 600 volt sample pump. The 600 volt power to this pump is classified as '

associated and requires appropriate electrical separation. The samplo pump is currently powered from essential power bus 1EMXC. However, as part of this mod, the 600 volt sample pump will be replaced by a 120 VAC pump. The new power supply for this sample pump will be from the 120 VAC battery-  ;

. backed Auxiliary Control Power Bus KXB. While the pump's power supply will no longer be supplied.by A

an essential bus, the KXB bus supplies a battery backup in case of a LOOP. There are no licensing' commitments requiring the sample pump to have essential power. . Based on Duke Power's responses to  :

Regulatory Guide 1.97,1 EMF 33 is used only for normal operational periods, not for post-accident 1 monitoring. Therefore, essential power is not required. The sample pump originally received essential' l power to ensure a reliable source of 600 volt power. With the new 120 VAC pump, reliable power can be ,

supplied with a battery-backed power supply. However, this modification will require FSAR Figure 8.1.2-1 Table 11.4.2-2 to be revised to reflect the change from essential power to battery-backed power, The 1 EMF 33 skid weighs approximately 3200 lbs. A floor loading analysis has been performed to ensure ,

that the proposed new location on the mezzanine level of the Turbine Building is structurally sound and -

can handle the additional load. The detector skid will be anchored to the floor to ensure that it can't be  ;

moved by accidents. >

The CSAE Radiation Monitor is not an accident initiator. In addition, it does not function as an accident rnitigation system. The purpose of the monitor is to provide early indications of steam generator tube .

leaks. The relocation of the sample tap and detector will not have any adverse impact on the operation .

of the CSAE system. The relocation will result in 1 EMF 33 being more reliable as an early detection system for tube leaks. Therefore, the consequences of accidents previously evaluated in the FSAR are '

not increased.

1 The relocation of the sample tap and the detector skid will ensure that the CSAE Radiation M.onitor-functions as designed. The new sample tap location will provide air with no dilution from gland steam exhaust, thus ensuring that the detector readings will be undiluted.1 in addition, the location of the sample -

tap, at the CSAR header, provides a better real time indication of potential tube leaks. No new failure -

modes are created by the sample tap and detector. skid relocation. .Therefore, no new accidents or ^  ;

malfunctions of equipment are created. .

The relocation of the sample tap and detector skid will not have any adverse impact on the CSAE system. 3 Similarly, the same equipment is being used since the detector is actually being relocated from its current :

location to the Turbine Building location. The only change is the 600 volt sample pump being replaced y by the 120 VAC battery-backed sample pump. This pump will be reliable as the original 600 volt pump and easier to maintain. Therefore, the probability of malfunction of equipment important to safety evaluated in the FSAR is not increased.  ;

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The 1 EMF 33 system is designed to provide early indications of tube leaks. This modification will result i in a more sensitive and reliable system that can provide better data based on undiluted samples. With

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the new configuration and sample tap location, Operations personnel should have earlier indications of tube leaks. Therefore, the probability of an accident previously evaluated in the FSAR is not increased.

No currently postulated failures are affected by this modification. In addition, no new failure modes are created; therefore, there is no increase in the consequences of a malfunction'of equipment important to safety evaluated in the FSAR.

The modifications do not reduce the capability of any systems to function as assumed in the Chapter 15 L accident analyses. No fission product barriers are degraded by these modifications. No safety limits, setpoints, or limiting safety systems settings are affected by the modifications. Therefore, the margin of safety as defined in the basis to the Technical Specifications is not redu.ced. No USO exists..

MG-12374 and MG-22374

Description:

These modifications are for replacing two of the VO system cooling units that serve the Transducer Rooms <

in the Turbine Building. The modifications are expected to decrease maintenance, and increase reliability and performance of the VO system. The existing units are water-cooled by the Conventional Service Water (RL) system. The new cooling units will be air-cooled. All associated 1" RL piping and valves will be removed. The new cooling units will be installed in approximately the same location as the existing units.

Safety Review and USQ Evaluation:

The VO system is not an accident initiator in any accident analyses. The cooling units are non-safety related. They are located in the Turbine Building, a non-seismic structure. There are no seismic or electrical separation requirements applicable to the system. An Appendix R review was performed, with no concerns identified. The modifications are expected to increase VO system reliability, and therefore, improve transducer room cooling. The overall reliability margin (as affected by ambient temperature) for secondary side instrumentation in the transducer room should be improved as a result of the modifications.

The VO system cooling units are not used to mitigate accidents. The system will continue to perform the same functions as before the modification. The cooling units do not perform any safety functions, nor - ,

interact with other equipment so that the performance of a safety function is degraded. There are no new ,

system functions added by the modifications,'and no new failure modes created. No accidents previously considered incredible are made credible by the NSMs. The new cooling units are expected to be more reliable than the existing units. There are no changes of safety limits, setpoints, or plant parameters because of the modifications. The fission product barriers (RCS boundary, containment, fuel pellets, and cladding) are not degraded. No assumption made in any accident analysis are affected by the NSMs.

No USQ exists.

MG-52362

Description:

This modification replaces a computer room chiller due to corrosion-induced failure. The replacement is an air-cooled chiller. The computer room HVAC system has redundant chillers,"A" and "B"(not Safety Trains A and B). The new chiller, which includes the evaporator, compressors, expansion valves and <

control panel, will be mounted in the same location as the existing chiller, The new chiller is smaller than the old chiller in both size and weight. One new roof penetration is required for refrigerant lines. The added lines will be class H, non-QA, non-seismic. RN system piping, which provides cooling water to '

the existing chiller, will be cut and capped. Chilled water (YC) piping, from the evaporator to the condensers, will be mod _ified to accommodate the new chiller A new transformer and cabling are required to furnish power for the new chiller. A non-essential power source will be used.  ;

O

s Safety Review and USQ Evaluation:

There is no OA condition applicable to the equipment involved in the modification. The computer room HVAC system maintains room temperature within temperature specifications for computer-type equipment.

The chillers will be powered from segrate load centers to maintain power supply diversity. Redundant chillers and condensers are employed for computer room cooling because a high degree of reliability is desirable for computer-type equipment. The replacement chiller is expected to be more reliable than the '

existing chiller. Since the Computer Room is located in the Service Building, a non-seismically qualified structure, there is no applicable seismic criteria. Electrical separation criteria is not applicable. An Appendix R Review was performed, with no concerns identified. The computer room HVAC system performs no accident mitigation functions nor plant safety functions. - The new chiller will perform the same ,

function as the existing chiller. The performance of plant safety functions are not degraded. No common failure modes are created. No new system functions are added or failure modes created by the chiller i replacement. No accidents previously considered incredible are made credible by these NSMs. There are no changes of safety limits, setpoints, or plant parameters because of the modifications. The fission product barriers (RCS boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSMs. No USO exists.

MG-12375 and MG-22375

Description:

These modifications involve replacing three Service Building and Warehouse (VO) System HVAC units.

One unit serves the Unit 1 Turbine Building Battery Room, another serves the Unit 2 Turbine Building Battery Room, and the third unit serves the Central Alarm Station (CAS). The existing units are water-cooled by the Conventional Servicc Water (RL) System. The HVAC units are significantly degraded due to condenser fouling. The new units will be air-cooled. The HVAC units for each battery room will be replaced with two 50% units. The RL supply line will be capped at the main header. Most of the 2" RL piping and valves associated with the HVAC units will be removed. The new cooling units will be installed in the same location as the existing units. Some concrete pad, ductwork and wall opening changes will be necessary.

Safety Review and USQ Evaluation:

The VO system is not an accident initiator in accident analyses. The system is non-safety related. There are no seismic or electrical separation requirements applicabl3 to the system. An Appendix R review was performed, with no concerns identified. The modifications are expected to increase the reliability and performance of the HVAC supply to the Turbine Building Battery Rooms and the Central Alarm Station.

Temperature margins for equipment in these areas should be improved. The Turbine Building battery rooms contain batteries 1 DP and 2DP. These batteries supply non-essential unit DC loads such as motors and backup fighting. The Central Alarm Station contains security alarm equipment, which is not required for reactor shutdown. The VO system HVAC units are not used to mitigate accidents. The VO system will continue to perform the same functions as before the modification. The VO system HVAC units are ~

non-safety and do not perform any safety functions, nor interact with other equipment so that the performance of a safety function is degraded. There are no new system functions added by the modifications, and no new failure modes created. No accidents previously considered incredible are made credible by these NSMs. There are no changes of safety limits, setpoints, or plant parameters because of the modifications. The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSMs.

No USO exists.

l MG-12317 and MG-22317

Description:

These modifications will install main bearing temperature sensors in the McGuire emergency diesels and connect the sensors to the diesel monitoring and self-diagnostic system. The currently installed bearing caps will be removed from the diesel, drilled, tapped, fitted with RTDs for detecting bearing temperature, then replaced in the diesel. A crankcase penetration for each RTD will be drilled at the upper right comer of each of the diesel right bank side crankcase covers. Each penetration will be tapped to allow a hex

• nipple to be threaded into the penetration to provide support for the RTD termination heads. The RTD leadwires will be sealed within the hex nipples with RTV silicone rubber sealant to maintain the crankcase pressure boundary. The new equipment will permit monitoring and trending of bearing temperatures to allow early problem detection, which may prevent major engine damage.

Safety Review and USQ Evaluation:

The ernergency diesels are not addressed as an accident initiator in any accident analyses. The added RTD sensors and associated cabling will serve no safety-related function. All added equipment will be-installed to meet OA condition 4 (seismic mounting) requirements. The seismic anchoring calculation for the diesel monitoring system was reviewed. Cable separation criteria has been applied to the cable routing. An Appendix R review was performed with no concerns identified. Emergency Diesel reliability is expected to be increased through earlier detection of bearing problems. The performance of emergency diesel safety functions will not be degraded by the modification. Each emergency diesel is completely separate from the others. There are no common failure modes created which could render more than one J emergency diesel inoperable. Addition of temperature sensors represents a new function added to the emergency diesels. A stress analysis was performed to ensure that a new failure mode was not introduced by modification of the bearing caps. The use of RTV sealant, for RTD leadwire penetrations into the crankcase, was also reviewed and judged acceptable. The new electray has been routed to avoid obstructing the fire suppression halon nozzles in the DG rooms. There are no changes of safety limits, setpoints, or plant parameters because of the modifications. The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSMs. No USO exists.

MG-12293 and MG-22293 i

Description:

These modifications change out the in-containment signal cables for radiation monitors EMF 51 A and EMF 51B to coaxial type, and route the cables in sealed conduit. The cables will be routed through new triax feed-throughs to be used with the existing Conax containment penetrations. The D. G.'O'Brien containment penetrations, presently used for signal and power cables, will only be used for EMF 51 A and EMF 51B high voltage connections. The changes are being made to resolve potential concems over the degradation of signal cables for the Containment High Range Radiation Monitors during Design Basis Accident conditions. The modifications will allow the post accident readings from the radiation monitors to remain within the accuracy recommended by Regulatory Guide 1.97, Revision 2.

Safety Review and USQ Evaluation: ,

The radiation monitors and containment penetrations are not accident initiators in any accident analyses.

There are no control functions performed by the radiation monitors or containment penetrations. The new '

conduit will be mounted per QA Condition 4 requirements. An Appendix R Review was conducted with no concems identified. The new cabling, conduit, and penetration feed-throughs are OA Condition 1 qualified (which includes environmental qualification for the containment environment). The radiation monitors and containment penetrations will continue to pectorm the same functions. The equipment J

i

.l l

i involved in these modifications will not interact with other equipment in any ways such that the performance of any plant safety functions are degraded. Applicable safety train separation criteria will be followed for cable installation. No common failure modes are created. No new failure modes are created. i The post-accident radiation monitor readings are expected to be more accurate after the modifications,  ;

providing the plant operators with more reliable post-accident in-containment radiation readings. No accidents previously considered incredible are made credible by these NSMs. No technical specifications or plant parameters are affected by this modification. There are no changes of safety limits, setpoints, or plant parameters because of the modifications. The fission product barriers (RCS boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSMs. No USO exists.

MG-52405 This modification will remove the four existing Instrument Air (VI) system air dryers and associated afterfilters , and install three new dryers and filter packages. The replacement will be made to meet air quality standards defined by ISA-S7.3-1975, and resolve a concern by the NRC in Generic Letter 88-14.

The existing dryers are refrigerant type, the new dryers will be desiccant-type. The VI system connection to the Station Air (VS) system will be installed downstream of the new dryers. The new dryers will be sized to supply air simultaneously to the VI and VS systems. Vi piping at the dryers will be modified to accommodate the new dryer packages. The replacement will be made on one dryer at a time, without interrupting the VI supply. The inlet valves for each dryer will be replaced with new isolation valves. Test vents will be installed for easier VI air quality testing.

Safety Review and USQ Evaluation:

The VI and VS systems are not accident initiators in any accident analyses. The VI dryers do not perform accident mitigation functions or other plant safety functions. The VS system has no safety classification.

The VI equipment and piping involved with the modification has no safety classification. Safety class equipment requiring compressed air from VI is provided with an air reserve sufficient to perform its safety function should the VI system fail. There are no seismic or separation criteria applicable to the equipment or cabling involved in the modification. An Appendix R review was conducted with no concerns identified.

The expected improved air quality should increase the overall reliability of instrumentation that uses air supplied by VI and VS. The modification will not degrade the performance of any plant safety functions.

The modification does not change any VI or VS system functions. There are no common failure modes created by the modification. No new functions are added by the modification. The new dryers will perform the same functions as the existing dryers, except using a different process. No accidents previously considered incredible are made credible by this NSM. There are no changes of safety limits, setpoints, or plant parameters because of the modification. The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSM. No USO exists.

MG-52407

Description:

This NSM is for modification of the intrusion detection system at the cargo access portal (CAP) exterior to the warehouse. The modification resolves an NRC audit finding and will allow elimination of compensatory measures. The warehouse loading dock will be extended to provide an even elevation across the front of the warehouse. The dock leveler will be relocated. Several existing microwave units will be removed, and several relocated. An active infrared link will be added across the loading dock area.-

Underground barriers will be added beneath the infrared beam.

Safety Review and USQ Evaluation:

1 1

I The security system has no associated QA condition. The station security plan will require revision. The security sy:: tem is not evaluated as an accident initiator in any accident analysis. The modification affects only non-safety related equipment and structures. There are no seismic, separation, or Appendix R ,

requirements associated with the equipment or structures involved in the modification. The security _

system is not an accident mitigator. The modification does not affect the plant structure or any equipment inside the plant. The performance of plant safety functions will not be degraded by this modification.

There are no common failure modes created by the modification. The active infrared link is a new security function added by the modification. This will serve to increase the security system reliability for the warehouse cargo portal area. There are no new failure modes created by the modification. No accidents previously considered incredible are made credible by this NSM. There are no changes of safety limits, setpoints, or plant parameters because of the modification. The fission product barriers (RCS prassure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSM. No USO exists.

MG-12364 and MG-22364

Description:

These modifications will down-grade the Diesel Generator tuming gear for each of the 4 DGs from OA-1 to QA-4. The DG tuming gear (also called "barring gear") are not required for emergency operation of the DGs. The turning gear are engaged manually, and operated only during DG maintenance. These modifications will eliminate the need to order QA-1 replacement parts for the DG turning gear. Shunt trip devices will be installed at the breakers for each turning gear motor to trip the safety related Dower feeding the motors during a design basis accident (on SS Signal). The breakers, shunt trip devices and control cables will be safety related. ' The power cables to the tuming gear motors will be declassified to non-safety related, and will be painted black to designate no safety classification.

Safety Review and USQ Evaluation:

The DG systems and DG turning gear are not accident initiators in any accident analyses. Limit switches, in each turning gear, prevent the corresponding DG from starting when the tuming gear is engaged.

These limit switches will remain OA-1. Declassified cabling will conform to applicable separation requirements. An Appendix R review was performed with no concerns identified. A seismic review was not required because the additional weight from added fuse and terminal blocks was within approved weight limits for the affected cabinets. The DGs provide emergency power for accident mitigation equipment. However, the DG tuming gear are used only during DG maintenance, and do not perform any safety functions, nor interact with other equipment in a way that causes the performance of any safety functions to be degraded. The DG and DG tuming gear will continue to perform the same functions. The 4 DGs will remain fully independent and separated. This modification creates no common or new failure modes. The DG tuming gear will perform the same functions as prior to the modifications. Breaker coordination was reviewed to insure that any electrical faults associated with the reclassified turning gear motors will not affect the safety related power source. No accidents previously considered incredible are made credible by these NSMs. There are no changes of safety limits, setpoints, or plant parameters because of the modifications. The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by .

the NSMs. No USO exists.

MG-52378 <

Description:

Two water-cooled air conditioning (AC) units, that serve the Secondary Alarm Station (SAS) and personnel Access Portal (PAP), will be replaced with roof-mounted air-cooled units. The existing units have been subject to cooling water line and condenser fouling. The modifications are expected to i j

q i

decrease required maintenance, while increasing cooling system performance and reliability. The existing Conventional Service Water (RL) line which supplies the AC units will be capped at the main header. No longer needed piping, valves, and fitting willbe removed. The existing AC units are located in the Service Building. The replacement units will be located on the roof of the Service Building. - Existing ductwork will be modified as required to accommodate the new units. A new return air duct will be added. An existing fresh air intake will be closed, because the new AC units incorporate fresh air intakes. New roof penetrations will be added for refrigerant tubing. A roof curb will be provided for each unit. Supporting steel will be provided. The new AC units were sized based on the original capacity of the existing units.

The power supply to the units will be upgraded due to increased power requirements. A new thermostat control will be added for one AC unit. Existing thermostats will be u;ed for the other unit.

Safety Review and USQ Evaluation:

The Service Building HVAC (VW) and the Conventional Service Water (RL) systems are not accident initiators in any accident analyses. No seismic or separation criteria are applicable to the systems. The existing and new units are non-safety related. An Appendix R review was performed, with no concerns identified. The roof-mounted equipment was evaluated for any potential to become a tornado missile, with no concern identified. The RL and VW systems are not used for accident mitigation, aad do not perform any plant safety functions. The performance of plant safety functions will not be degraded by the modifications. There are no common failure modes created. No new failure modes are created. The Service Building roo! loading was review and loading calculations were undated. No accidents previously considered incredible are made credible by this NSM. There are no changes of safety limits, setpoints, or plant parameters because of the modification. The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSM. No USO exists.

MG-12289, MG-22289 and MG-52289

Description:

These NSMs install test connections on the WZ (Groundwater Drainage) system, sumps A, B. and C, to facilitate pump flow tests and check valve leakage tests. The system is safety related as it performs the function of relieving hydrostatic pressure from the Auxiliary and Reactor buildings which are Category 1 -

structures. Each sump (of the 3 existing) has two 100% capacity pumps aligned to separate essential  ;

power sources (trains A and B). There are no electrical changes involved with these NSMs. The '

mechanical scope includes necessary piping and valves to accomplish the tests. The piping being added ,

is class C (safety related and seismic) out to an isolation valve at which a class change occurs to class ,

G (non-safety). This design preserves the integrity of the existing lines and offers seismic protection to the functionally safety related portions of the system.

Safety Review and USQ Evaluation:

The WZ system is used in control of and mitigation of internal flooding events that may occur in the Auxiliary Building. Flooding events are design events and are not accidents. Mitigation of design events does not necessarily require single failure considerations unless the flood logically exists when making design basis event combination assumptions such as earthquakes and accidents. In these cases, the scenario would be considered a design basis event, not a design event (flood). The WZ system is redundant and safety related and can evacuate all flood sources to which it can be subjected within design basis assumptions. Mitigation of flooding events will not be affected by these NSMs. The NSMs modify the existing WZ system to facilitate pump flow and check valve leakage tests. All applicable cliteria have been followed and the design creates no new failure modes. The seismic integrity of the system is preserved with class C piping connecting to the existing piping and locked closed isolation valves at the class C to G boundary. No USO exists.

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MG-12301

Description:

This modification will install additional pressure monitoring instruments, which will monitor wide range annulus pressure for the annulus ventilation system. These instruments will trip the exhausting fan should '

annulus pressure exceed a low-low setpoint (-7.0" w.g.) and the dampers are in the exhaust alignment.

Without these new instruments, should the existing instruments fail in such a way that the system exhausts continuously, then annulus pressure could reach a vacuum as low as 22" w.g., which could cause deformation of the WP and VW ductwork. Two computer alarms are also added (Low and Low-Low Annulus Pressure). The installed instrumentation will be QA Condition 1 and seismically mounted.

Safety Review and USQ Evaluation:

No operating parameters are changed. The function of the VE system will remain the same. no new failure modes are introduced. No USO exists.

MG-12302 and MG-12305

Description:

These modifications change the main steam safety valves on the 1 A and 1D main steam headers from a butt-welded intet design to a ilanged inlet and outlet design. New valve bodies and internals, with flanged inlet and outlet connections, have been procured. Additionally, this modification will rotate the valves 15 degrees to the vertical position, to prevent excessive seat leakage.

Safety Review and USQ Evaluation:

The main steam safety valves are OA Condition 1 and nuclear safety related. The valves are built to ASME Boiler and Pressure Vessel Code, Section 111, Class 2. The modifications have received written relief, from the NRC, of the post installation hydrostatic testing as required by Section XI of the ASME Boiler and Pressure Vessel Code. The relevant FSAR Chapter 15 accidents are loss of externalload, main steam line break, loss of feeclwater, loss of offsite power, loss of condenser vacuum, inadvertent closure  ;

of MSIVs, turbine trip, inadvertent opening of a steam generator relief or safety valve, and steam generator tube failure. The new main steam safety valves are constructed to the same design, material, and construction standards as the present valves. The valves will be tested prior to installation. The valves lift settings, orifice sizes and flow capacities are unchanged. No USO exists.

MG-12319 and MG-22319

Description:

Presently there is no way to test the diesel generator room sump pumps (WN system) or the associated check valves located downstream of the pumps. These modifications will create a mear':, Dy which the WN sumps pumps and check valves can be tested in accordance with ASME section N subsection lWP and GL 89-04. New test connections will be installed on the WN system to facilita% pump flow tests and perform check valve leakage tests. Parts of the system are safety related as e performs the function of flood protection for the DG in case of a pipe rupture.

The new piping, valves, and instrumentation will be Class C upstrera of and including the throttle valve and Class F downstream, except for the butterfly valve located E.i the common discharge header. This valve will be Class G and located in Class G piping the TurKne Building. The reason this is acceptable is because failure of this valve or this Class G piping will '.ot degrade the WN system nor prevent it from performing it's safety function. The diesel generate sump will require modifications to accept piping i

~

support restraints.

Safety Review and USQ Evaluation:

The new test lines will be used for testing purposes only and the only impact on normal operation will be the head losses associated with the butterfly valve and tee installed in the existing common discharge line.

The design preserves the integrity of the existing lines and offers seismic protection to the functionally safety related portions of the system. These modifications will facilitate testing to help ensure system readiness. The added components will not affect the systems' ability to function as designed. No new failure modes are created because the piping and components added are of consistent design classification and no control changes are being made. No accidents previously thought incredible are made credible by these NSMs. No USO exists.

MG-12344 and MG-22344

Description:

In 1978, the Diesel Generator Sump Pump System (WN) was downgraded from Class C to Class F because the system could not be hydro tested in accordance with ASME Section Ill. The reclassification was justified, in part, because the worst case flood scenario was thought to be an RN (Nuclear Service Water System) pipe rupture which could not affect both Diesel Generator rooms. During the resolution of PIR 2-M88-0185 Rev. 2, a flood source outside the DG rooms was discovered. An analysis of a Turbine Building flood indicated that the leakage through the wall penetrations would continue into both the DG rooms until the Turbine Building flood was stopped. This would lead to potential flooding of both DG rooms from a single fault if the WN system is inoperable. Portions of the WN system will be reclassified to Class C for protection against this scenario.

Safety Review and USQ Evaluation:

The WN System is designed to remove leakage and waste from drains in the Diesel Generator rooms and to protect the DGs from flooding due to a major pipe rupture. The modification will upgrade the WN system which will increase confidence in the capability of the system to perform as designed, including the capability to handle the TB flooding (and newly identified subsequent leakage into the DG rooms) coincident with loss of offsite power. No USO exists.

MG-12400 and MG-22400

Description:

These NSMs install an annulus door monitoring system to monitor the open and closed status of the annulus doors. The NSMs are intended to preclude future occurrences of annulus doors being left open which impact operability of the annulus ventilation system. The modifications also ensure the doors remain as fire barriers. Limit switches will be installed in the door frames and provide indication after the door has been unlatched for a pre-set time.  :

1 Spare security cables will be used to wire both Unit 1 and Unit 2 timit switches to the Unit 1 MG Set Room. .l A single new panel, called the annulus door monitoring cabinet or TB-670, will be seismically mounted )

in the room to house the reflash modules and associated circuitry for both units. Cables will be required l to tie in the security cabinets to the new annulus door monitoring cabinet. There will be no affect on any .j security system. No control changes will be made. An annunciator in the control room will indicate that  !

a door has been open for a pre-set time but will not indicate the particular door; the particular door will  !

be identified at the annulus door monitoring cabinet. ,

Safety Review and USQ Evaluation:

i l

3 No control changes are being made to the VE system. No new failure modes are created. The NSMs make it easier to maintain the pressure boundary aspects of the annulus satisfying the assumptions on which the VE system was designed and on which the Technical Specifications are based. No bypass leak paths are created that would allow any unfiltered air to escape the annulus. The VE filters and flowrates are unaffected by these NSMs. The VE system functions will not be affected and the post accident performance will be unchanged.

An Appendix R evaluation has been performed with no concerns identified. The new cabinet in the Unit ,

1 MG Set room is seismically mounted to preclude any adverse interactions with other SSCs should an earthquake occur. The limit switches added to the doors will not create any additional leak paths. All applicable design criteria have been preserved. No USO exists.

MG-12410 and MG-22410

Description:

These modifications are expected to improve the reliability of the Steam Generator Feedwater Control Valves (FCV), CF17,20,23 and 32. These valves, located in the Turbine Building, are modulated by a three-element feedwater control system using feedwater flow, steam generator water level, and steam flow as parameters. Presently, two solenoid valves (Train A and Train B) in series close the respective FCV by a signal from either Train A or Train B. These NSM's will add redundant solenoid valves of the same train in parallel. The new solenoid valves will be Valcor Model V70900-65 valves. Although this change will now require two solenoid valves to close each FCV instead of one, the FCV closure still meets single failure analysis, and the potential for inadvertent FCV closure due to a solenoid valve spuriously changing position is reduced.

Additionally, all of the existing solenoid valves will be replaced with the more reliable model, Va!cor Model V70900-65, under Minor Modification MM-5127 for Unit 1 and under MM-5128 for Unit 2.

Safety Review and USQ Evaluation:

The FCV's require air to open and they fail closed upon loss of air or loss of power to the solenoid valve circuit. In the original control scheme, a single circuit or solenoid coil failure causes the associated FCV to fail closed with results in a unit trip due to low steam generator level. In the new control scheme, if one solenoid valve falls the other solenoid valves keep the control pressure to the FCV. Also, the solenoid valves associated with a single FCV and of the same train are placed on separate circuits to improve electrical reliability. Consideration has also been given to Main Feedwater flow isolation on high doghouse water level. Control valves CF20 and 23 are connected to the feedwater lines that extend through the "

interior doghouse while CF17 and 32 are associated with the lines that run through the exterior doghouse.

Doghouse high water level is provided as a trip for each FCV; therefore, the valves are paired electrically according to the piping connections. The new solenoid valves are considered more reliable than the existing solenoid valves. The new tubing, although not safety related,is comparable to the present design.

The power supply and breaker arrangement does not introduce any new or common failure modes. The new solenoid arrangement is functionally as capable as before of closing the FCV's under all required .

conditions. No USO exists.

Duke Power Company McGuire Nuclear Station Summary of Minor Modifications Completed Under 10CFR50.59 i

MM-3219

Description:

The purpose of this exempt variation notice is to allow the " repair" of defective Unit 2 Steam Generator tubes by either " plugging" or " sleeving." Due to indications detected in previous Eddy Current Data, evaluation and/or testing will be performed to determine the condition of the primary to secondary head exchanger tubes. From the results of evaluation and/or testing, all tube defects, or in some cases, indications will be repaired by the above indicated processes. All field work will be performed using approved station and/or vendor procedures. Babcock and Wilcox personnel will perform the required maintenance activities.

Safety Review and USQ Evaluation:

A detailed LOCA analysis has been performed by Westinghouse. Babcock and Wilcox has performed a LOCA Analysis for insertion of their fuel in the McGuire Units per Report BAW-10174, Rev.1, dated November 1990 and titled " MARK-BW Reload LOCA Analysis for the Catawba and McGuire Units." In addition, the Design Engineering Safety Analysis Group has performed a LOCA Analysis for the limiting case of a Feedwater Line Break which was submitted with the McGuire 2 Cycle 8 Reload Report and subsequently approved in November 1991. This LOCA Analysis was included in the McGuire Cycle 8 Reload Report, by reference.

These LOCA analyses support an " equivalent plugging" limit of 10% (467 tubes) per Steam Generator -

this is a MAXIMUM number. It is not expected to exceed the 10% limit during the repair of these Steam Generators. As long as " equivalent plugging" does not exceed this limit, then the FSAR and Technical Specifications will not be affected in any significant manner due to the implementation of this exempt variation notice. The possibility, probability, and consequences or an accident previously or not previously evaluated in the FSAR will not be increased due to this repair operation. There is potential for tube plugging /sieeving to affect Reactor Coolant System flowrate which in turn could affect allowable reactor power level The margin of safety as defined in any Technical Specification bases will not be decreased due to the implementation of this MEVN. The components affected are QA Condition 1 and nuclear safety related.

MM-3217

Description:

The purpose of this minor modification is to review the Low Temperature Overpressure Protection (LTOP) setpoint. The existing setpoint is 395 +0/-30 psig. Due to current temperature and reactor vessel material considerations, as well as instrument accuracies, it is desirable to lower this setpoint. Therefore, the setpoint is being lowered to 385 +0/-20 psig.

Safety Review and USO Evaluation:

Engineering has evaluated the current material condition of the reactor vessel in conjunction with anticipated temperature variations in the NC system at shutdown. Based on this evaluation, the calculations performed and documented in Calculation File MCC 1223.93-00-0005 verify that the new

I setpoint provides the desired margin should a LTOP actuation occur.

k Discussions with the Safety Analysis Group indicate that credit is taken for the PORV's only in the case l

of a SG tube rupture event. Should this event occur, the new setpoint provides a level of protection that is equal to or greater than the existing setpoint.

Reduction of the setpoint does not adversely affect the level of protection as it is conservative with respect to the current analysis. The margin of safety as defined in the bases to Technical Specifications will not be reduced as a result of this modification. Technical Specification 3.4.9.3 requires a PORV setpoint of less than or equal to 400 psig. Reduction of the setpoint from 395 to 385 psig is conservative with respect to this limit.

MM-3317

Description:

This minor modification installs a new high side pressure tap / root valve for 1 RNPT5320 due to indications of clogging on the existing tap / root valve. Installation of a new root valve will ensure a reliable means of measuring and monitoring the differential pressure (DP) across the RN side of the *1 A' KC HX. The new tap will be in roughly the equivalent location relative to the HX and will therefore not have an effect on the measured DP.

Safety Review and USQ Evaluation:

This OA-1 MM will not have any effect on the RN system Design Basis or Technical Specifications. HX l DP is not a Tech. Spec. surveillance, but is monitored continually by the Operator Aid Computer via output from 1RNPT5320. With a clogged root valve, this monitoring capability is unreliable, and the operability of the HX indeterminant.

The MM will therefore not increase the possibility, probability or consequences of any previously analyzed events. Nor will it create any previously unanalyzed accident scenarios. The DP limit specified in the KD HX Test Acceptance Criteria will not be effected by this MM nor will the margin of safety as defined in any Technical Specification be reduced as a result of implementing this MM.

MM-3425 and MM-3426

Description:

NRC Generic Letter 89-10 instructs nuclear power stations to develop a program for the testing, inspection and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much -

greater than that previously performed by Duke Power Company nuclear stations. As required by Generic Letter 89-10, Duke has developed a comprehencive program plan that describes the actions we will accomplish in order to comply the Generic Letter. This minor modification provides for the diagnostic testing for MOV 1ND0068 and 1ND0067 and constitutes part of the actions necessary for compliance to GL 89-10.

The actual changes to 1ND0068 and 1ND0067 involve re-setting the open and close torque switches so

l. that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.189-00-0003 and is provided by controlled document MCM 4 1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision l

I l

i of Duke Power Specification DPS-1205.19-00-0002 which establishes the parameters and criteria used t to determine the minimum required and maximum allowed thrust levels for 1ND0068 and 1ND0067.

The diagnostic test system used to facilitate thrust testing for 1ND0068 and 1ND0067 associated with the diagnostic test system used in testing 1ND0068 and 1ND0067 have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at' design differential pressure and system pressure without exceeding the limitations of the operator or valve components.

Safety Review and USQ Evaluation: ,

The MOV affected by MM-3426 is in the Decay heat Removal System. The function of 1ND0068 is to provide a recirculation flow path to protect the ND pumps from low flow conditions. The safety position of 1ND00S8 is open during low flow conditions to protect the ND pumps and close when the ND pumps ,

are injecting to maximize the injection flow rate.1ND0068 and 1ND0067 are active valves which are #

normally closed. Re-setting the open and close torque switches will not affect open and closure times of 1ND0068. The existing stress analysis of the piping associated with 1ND0068 and 1ND0067 will not be affected by re-setting the open and close torque switches. Since this MM will ensure that 1ND0068 and 1ND0067 will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not ,

increased. No USO exists.

MM-3259 and MM-3260

Description:

Duke Power Company has committed to the NRC to evaluate all valve electric motor operators (EMO) as per Generic Letter 89-10.

Safety Review and USQ Evaluation:

This evaluation is being performed to help ensure that the actuators will perform their intended function under design differential pressure in the event of an accident. ,

Revisions to the 89-10 set up sheets are editorial in nature. No additional work will be required. All listed actuators have been previously set up and are within the new thrust windows.

No new failure modes will be introduced as a result of this modification. Additionally, the probability, -

possibility, or consequences will not be increased by this modification. No FSAR or Technical Specifications are affected. No USO exists.

MM-3322 t

Description:

The purpose of this minor modification is to allow the replacement and setpoint change for instruments 2RNPS5001 (RN pump strainer 2A outlet pressure) and 2RNPS5011 (RN pump strainer 2B outlet - ,

pressure).

Safety Review and USQ Evaluation:

The "as-built" setpoint for these pressure switches is 2 psig decreasing. Under normal operating conditions with 3 or 4 RN pumps in service, it is typical to experience the annunciator alarm from these switches cycling in and out. With the alarm being received under norrnal conditions, it is considered as i

a nuisance alarm by the Operations Group. The new setpoint for these pressure switches will be 0 psig -

' decreasing. This change should prevent the recurrence of low pump suction pressure alarms under normal operating conditions. The existing pressure switch is manufactured by United Electric as model

1. J6D-148. It has a process range of 0-40 psig, it has an inadequate process range for this application.

The new pressure switch to be installed under this MM is also manufactured by United Electric as model

1. J6-134. This model has a process range of 30" HG vacuum to 20 psig. This instrument has an excellent process range for the intended application. The design or function of the RN system will not be compromised in any manner due to the replacement and setpoint change for these two pump suction pressure switches.

The FSAR and Technical Specifications will not be affected in any significant manner due to the implementation of this MM. The possibility, probability, or consequences of an accident not previously evaluated in the FSAR will not be increased by the installation and setpoint change of these instruments.

The margin of safety as defined in the bases to any Technical Specification will not be decreased due to this MM. The components affected by this MM are nuclear safety related and OA Condition 1. The new  :

pressure switch to be installed has undergone a commercial grade evaluation for this application. With a setpoint of 0 psig, the pumps still have approximately 5 psig NPSH margin under accident conditions. >

No USQ exists.

MM-5200 and MM-5201

Description:

The setpoint for the Auxiliary Feedwater (CA) suction pressure switches listed below will be changed to 3 psig.

ICAPS 5002 Motor-driven CA pump 1 A 1 CAPS 5350 Motor-driven CA pump 1 A 1 CAPS 5360 Motor-driven CA pump 1B 1 CAPS 5012 Motor-driven CA pump 1B 2 CAPS 5360 Motor-driven CA pump 2B 2 CAPS 5012 Motor-driven CA pump 2B 1 CAPS 5042 Turbine-driven CA pump 1 1 CAPS 5370 Turbine-driven CA pump 1 1 CAPS 5390 Turbine-driven CA pump 1 1 CAPS 5381 Turbine-driven CA pump 1 2 CAPS 5042 Turbine-driven CA pump 2 2 CAPS 5370 Turbine-driven CA pump 2 ,

2 CAPS 5390 Turbine-driven CA pump 2 2 CAPS 5381 Turbine-driven CA pump 2 1 CAPS 5044 Standby Shutdown System 2 CAPS 5044 Standby Shutdown System 1 CAPS 5380 Standby Shutdown System 2 CAPS 5380 Standby Shutdown System

, The setpoint for the CA suction pressure switches listed below will be changed to 4 psig. The zero reference elevration for these switches is at a lower elevation than for the other switches.

2 CAPS 500 : Motor-driven CA pump 2A 2 CAPS 5350 Motor-driven CA pump 2A The CA suction pressure switches initiate automatic swapover to RN on decreasing CA pump suction j pressure for their respective CA pumps.

, + . . - . . . - . ,

This change increases the suction pressure at which the CA suction swaps to the Nuclear Service Water ,

(RN) Assured Supply upon decreasing suction pressure. The new pressure switch settings facilitate RN swapover above a minimum static water elevation of 730'+10" for Unit 1 and 731'+7-3/8" for Unit 2. The setting change increases the probability that swapover will occur before it is no longer feasible to obtain water from the normal suction sources (upper surge tank, CA condensate storage tank, or the condenser hotwell).

The CA system is safety-related, and is used for normal plant shutdown, accident mitigation, and can be used during startup. The RN assured supply to CA system is safety-related.

Safety Review and USQ Evaluation:

The CA system is not an accident initiator addressed in any FSAR Chapter 15 accident analyses. The reliability of the RN system assured supply to the CA system is increased by the setpoint change.

Therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased.

The CA system is an accident mitigator for any emergency or normal shutdown operation that requires decay heat or reactor coolant heat dissipation. The performance of this safety function is not degraded '

by the setpoint change. There is no change of system functions and no common failure modes are created between CA system trains. Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR is not increased.

The modification adds no new CA or RN system functions. No new failure modes are created. No accidents previously considered incredible are made credible by these modifications. Thus the possibility of an accident or malfunction of equipment of a different type than evaluated in the SAR will not be created.

Technical Specification Table 3.3-4 gives the allowable and trip setpoints for CA suction pressure. The trip setpoint is greater than or equal to 2 psig. The change of pressure switch setpoints to 3 psig or 4 psig is in the conservative direction. The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are adversely affected by the modifications. Therefore, the margin of safety as defined in the basis for any Technical Specification is not decreased. No USO exists. i MM-3285

..1

Description:

Revise appropriate documents to reflect new stroke time requirements. Increase stroke times for 1 & 2 NI0009 and N10010 from 10 seconds to 11 seconds, for 1 & 2 NIO121, N10162,1N10152, from 10 seconds to 20 seconds.  !

Due to the new actuator setup procedure that is now being used, some stroke times have been found to be slightly longer than in the past. On some valves that were near the end of their stroke time limits -

previously, we now find times exceeding their old requirements. The new procedure; however, gives a  ;

more accurate indication of true valve position and is now incorporated in the Rotork manual. This minor ]

modification is a result of PlR 1-M91-0191. These changes are editorial in nature and will require no field j work. 1 Safety Review and USO Evaluation:

Design performed an evaluation on the changing of time response to the N10009 and N10010 under PIR i

1 -M91-0191. A review of the FSAR resulted in the determination that no credit was assumed for flow to be delivered through the valves of concern until the valves fully stroked opened. In reality, flow will exist throughout the stroke. The integrated flowrate through the valve during this period exceeds the integrated flow deficit caused by the increased stroke time. Therefore, adequate margin exists in the assumptions utilized in the FSAR Chapter 15 transient and accident analyses to offset the impact of a slightly increased stroke time. This same reasoning is also applied for the other valves in question. Refer to design document, CAL No. SRC-MC2-SA-86-025-0, for a more detailed evaluation of these stroke times.

The probability of possibility of an accident is not increased by this modification to increase the valves' stroke time. The safety margins as dercribed in the FSAR will not be compromised. No USO exists.

MM-3534

Description:

The expansion joint between the Interior Doghouse and the Reactor Building is filled with a self expanding cork material. The cork was installed during the construction of the concrete floors to provide an expansion joint which would allow the two structures to move independent of one another. The El. 750 ft. floor of the Interior Doghouse serves as a pressure and flood boundary. The design pressure in the Doghouse is 9 psig (steam line break) and the design flood water elevation from a feedwater line break '

is 5.2 feet above the El. 750 Ft. floor.

in the south end of Unit 2 Interior Doghouse, a 10 to 15 foot strip of cork has slipped about 6 inches downward out of the El. 750 Ft. floor joint. the El. 750 Ft. slab is 3 feet thick, therefore there is still about 30 inches of cork remaining in the joint. The area where the cork has slipped down is a low area in the floor. The standing water has infiltrated the cork over time and caused it to lose some of its' elasticity.

This MM will replace the degraded cork with a foam which will serve as the pressure and flood barrier, if during the work and inherent closer inspection other degraded joint areas are observed, additional repairs may be made at the discretion of the appropriate personnel.

The top 10 inches of damaged cork will be removed from the joint and replaced with foam. In order to qualify the foam, pressure tests were conducted on mock-ups of the joint using Dow Corning 3-6548 Silicone RTV foam and heavy foam at a depth of 10 inches. The lab tests indicated that both foams could withstand the 9 psi pressure after being aged for 40 years. The replacement foam will be installed OA 1 per the foam installation specification; the installation specification follows vendor instructions and guidelines. The Silicon RTV foam and the heavy foam are easily compressed and will prevent any foreign rigid material from entering the joint. The Auxiliary Building and the Reactor Building masses will easily compress the foam material during a seismic event and therefore maintain their independent movements.

The repairs will be made while on line creating a potential concem during installation that il some of the cork falls out during the removal of the top 10 inches of the cork, there may be a temporary breach in the pressure and flood boundary. Repair of any particular joint area will be completed in approximately one day. There are no fire protection requirements and the joint location precludes it from susceptibility to interior and exterior design basis missiles. No electrical panels or equipment are located directly below the Unit 2 joint. Additionally, a small breach in the barrier for a short duration will not cause any HVAC concerns. No dose or security concerns were identified.

Safety Review and USQ Evaluation:

There is no increase in the probability of an accident evaluated in the SAR because the joint under repair )

is not an initiator of these accidents, nor is it involved in any accident sequence. Design considerations . .!

of the new foam barrier as discussed above are expected to permit the new joint to perform it's design functions of flood and pressure barrier, expansion material, and floor separation at least as well as the old cork joint. Any temporary breach of the barrier will be short term, and easily noticeable as would be any I

accident conditions affecting the area. This should eliminate potential flooding concerns without any adverse affects. Based on this, there is no increase in the consequences of an accident or a malfunction of equipment important to safety evaluated in the SAR. Likewise, and since there is no safety related equipment or electrical panels located directly below any potentially breached barrier, there is no increase in the probability of a malfunction of equipment important to safety evaluated in the SAR. The barrier should function as before. No new credible failure modes were identified either for the MM or during it's installation. Therefore, the possibility for an accident of a different type or a malfunction of a different type ,

than evaluated in the SAR is not created.

No safety limit, setpoint, or operating parameter will be changed by thin modification. Therefore, the margin of safety as defined in the basis of the Technical Specifications will not be reduced. No USO ,

exists.  ;

MEVN-2123

Description:

Duke has committed to the NRC to test all safety related valve actuators. This test is being conducted to assure that the actuators will perform their intended safety functions under design differential pressure in ,

the event of an accident.

P Currently, the EMO list does not include thrust valves required to set the actuator for valve 2KC56. This MEVN will add the needed information.

Safety Review and USQ Evaluation:

This MEVN will not degrade any system important to safety. Thrust valves have been calculated based on design differential pressure and packing force; which is obtained from the valve manufacturer. The total thrust valve is the combination of the two forces.

This MEVN provides information by which valve 2KC56 actuator may be set up and verified capable of proper operation during a design basis accident scenario, thus increasing safety of the plant and public.

No USO exists.

MEVN-2721

Description:

This MEVN will revise the VE Design Bases Specification and Test Acceptance Criteria sheets to incorporate a new test method for the Unit 1 VE carbon filter trains.

Safety Review and USQ Evaluation:

The VE carbon filter trains are OA-1, nuclear safety related equipment. The new carbon testing method tests to a more restrictive standard that is normally used for ventilation systems that don't utilize heaters for humidity reduction. In addition, the required heat dissipation for the heaters is being changed from 43 kW+/- 6.4 kW to 43 kW + 6.4 / - 17.5 kW to compensate for potential undervoltage conditions that were identified in PIR 0-M90-0122. Although this allows for less humidity reduction by the heaters, the ability of the filter train to remove radioisotopes in a design bases accident has not been degraded because the new filter testing technique takes no credit for the heaters. The Technical Specification changes required for this new testing have previously been submitted to and approved by the NRC under Amendment 109. No physical changes have been made to the Unit 1 VE system under this MEVN. No b

. l nuclear safety related structures, systems, or components are adversely affected by this modification. No USO exists.

MEVN-1316 & MEVN-1317 l

Description:

Currently associated with the FWST are two vents and one overflow. The purpose of this Exempt Change is to use one of the vents as an additional overflow. This will prevent the FWST from reaching a level at which structural damage to the tank can occur.

Safety Review and USQ Evaluation:

The new " overflow" will still be open to the atmosphere but instead of being considered a vent, it will be considered an overflow, in order to accomplish this the current vent will have to be modified. The FWST is aligned to the safety injection pumps and the RHR pumps during normal operation and is part of the ECCS. This modification will not affect the ability of the FWST to perform its safety function. Because of this, neither the probability nor the consequences of an accident situation addressed in the FSAR will be increased. No new accident situation will be created because neither the structural integrity nor the ability of the FWST to perform its design function will be degraded. There will be no increase in the probability, possibility, or consequences of a malfunction of equipment important to safety because no other piece of equipment except the FWST vent line will be worked on and the functionability of the FWST after this modification has already been addressed. No Technical Specification or FSAR sections will be affected. No USO exists.

MEVN-2196

Description:

The purpose of this exempt variation notice is to allow the adjustment of the closing thrust on the valve actuators 1YC278,1YC30A,1YC38A and 1YC398.

t Safety Review and USQ Evaluation:

This adjustment will enhance the affected valves ability to correctly function under their design conditions.

No wiring modifications will be performed on these valves, only torque switch adjustments will be made.

The normal torque switch bypass adjustment for clearing the valve seat by the valve gate will be checked for proper operation during the opening cycle.

The FSAR and Technical Specifications will not be affected in any significant manner due to the implementation of this exempt variation notice. Essentially, this modification willimprove the valves ability to function as designed. All components affected are OA Condition 1 and nuclear safety related. The ,

possibility, probability, or consequences of an accident not previously evaluated in the FSAR will not be increased by the performance of this modification. No new accident scenarios are created. The margin of safety as defined in the bases to any Technical Specification will not be reduced.

MEVN Valve MEVN-3158 2CF153 MEVN-3032 2FW1 MEVN-3166 2CF155 MEVN-3056 2N176 MEVN-2859 1CA86

4 4

MEVN-3101 2NV1047 MEVN-2860 1CA116 '

MEVN-2931 1NV245 MEVN-2921 1NV7 MEVN-2933 1NV842

Description:

Duke Power Company has committed to the NRC to evaluate all valve Electric Motor Operators (EMO) as per Generic Letter 89-10. This evaluation is being performed to help ensure that the actuators will perform their intended function under design differential pressure in the event of an accident.

Safety Review and USQ Evaluation:

The existing EMO Setup List does not contain the complete information necessary to support this evaluation for these valves. These MEVNs provide the necessary setup information, developed under DPS-1205.19-00-0002 (Guideline for Performing Motor Operated Valve Reviews and Calculations), in response to Generic Letter 89-10.

This information will be used to setup and verify the valve operator is capable of proper operation during a design basis accident. No new failure mode will be introduced as a result of this modification.

Additionally, the probability, possibility, or consequences of a malfunction will not be increased by this modification. No FSAR or Technical Specifications are affected. No USO exists.

MEVN-3107 and MENV-3108

Description:

The purpose of these MEVNs is to install splash guards around the openings of ND pump coupilng cavities to prevent the pumps from splashing water to the pump rooms. This change will be implemented for both j trains of ND for Units 1 ar.d 2.

- Safety Review and USQ Evaluation:

ND pumps have the tendency to leak through their seals and splash water into the pump rooms. The leakage occurs when the pumps are first started af ter being inactive for an extended period of time. These pumps are equipped with Durametallic mechanical seals. Mechanical seals used in pumps, as a general rule, tend to leak when the pumps are put in use af ter an extended outage period. The splash guards will contain the leakage within the coupling cavity of the pumps, and will significantly reduce the potential' contamination of the pump rooms. The splash guard will also protect the rooms from contamination, if i the seal leakage occurs as the result of mechanical seal failure during normal pump operation.

The splash guards will not affect operation of the pumps. They will not introduce any loose part which may interfere with the pump's rotating components. The splash guards will each be bolted to the pump and motor flanges, from the four corners, using angle pieces designed for this purpose. Installing the splash guards requires drilling eight holes on the pump and eight holes on the motor flanges. The holes, 1/4" in diameter and 3/C deep, are needed for mounting the splash guards to the flanges. Drilling the holes will not affect integrity of the pump and motor flanges which are 2.0 and 1.0 inches in thickness respectively.

Continuous operation of mechanical seals requires dissipation of heat that is generated by the seal's internal components. This heat is generated by frictional contacts between rotating parts and stationary components of the seals, and is dissipated to atmosphere through the openings in the pump coupling

cavity. The splash guards are designed to have adequate venting capability to dissipate this heat. They are installed in four pieces, with each piece covering an opening between the motor support blocks. There will be an opening of about 1" X 8" on each side of each splash guard between the splash guard and the motor support block, which air can flow through to provide the necessary venting. In addition to the venting path on the sides, a 1" wide by 14" long opening on top side of each piece has been cut-out to provide additional venting. This opening and the openings on the sides will provide adequate venting to dissipate the heat generated by the mechanical seals.

ND system is an accident mitigating system, and is not a contributor to the cause of any accidents listed in chapter 15 of the FSAR. The addition of the splash guards will not initiate an accident or degrade safety of ND system or any of its components, implementation of this modification will not affect ND pump safety limits, and ND system will not be operated outside any of its design lim.its. An evaluation of the pumps concluded that no safety related equipment could be damaged if the splash guards fell. Therefore, it was acceptable for the mod to be designed as non safety related and non seismic. The probability of an accident previously evaluated in the FSAR is not increased, and margin of safety as defined in the bases to the Technical Specifications will not be reduced. No USO exists.

MEVN-3109 and MEVN-3110

Description:

The purpose of these MEVNs is to install splash guards around the openings of NS pump coupling cavities to prevent the pumps from splashing water to the pump rooms. The change will be implemented for both trains of NS for Units 1 and 2.

Safety Fleview and USQ Evaluation:

NS pumps have the tenderny to leak through their seats and splash water into the pump rooms. The leakage occurs when the puraps are first started after being inactive for an extended period of time. These pumps are equipped with Durametallic mechanical seals. Mechanical seals used in pumps, as a general .

rule, tend to leak when the pumps are put in use after an extended outage period. The splash guards will ,

contain the leakage within the coupling cavity of the pumps, and will significantly reduce the potential ~ r contamination of the pump tooms. The splash guard will also protect the rooms from contamination, if the seal leakage occurs as the result of mechanical seal failure during normal pump operation.

The splash guards will not affect operation of the pumps. They will not introduce any loose part which may interfere with the pump's rotating components. The splash guards will each be bolted to the pump and motor flanges, from the four corners, using angle pieces designed for this purpose. Installing the splash guards requires drilling eight holes on the pump and eight holes on the motor flanges. The holes, 1/4"in diamnter and 3/4" deep, are needed for mounting the splash guards to the flanges. Drilling the holes will not affect integrity of the pump and motor flanges which are 2.0 and 1.0 inches in thickness respectively.

Continuous operation of mechanical seals requires dissipation of heat that is generated by the seal's  !

internal components. This heat is generated by frictional contacts between rotating parts and stationary j components of the seals, and is dissipated to atmosphere through the openings. In the pump coupling cavity. The splash guards are designed to have adequate venting capability to dissipate this heat. They .  ;

are installed in four pieces, with each piece covering an opening between the motor support blocks. There .  !

will be an opening of about 1" X 8" on each side of each splash guard between the splash guard and the motor support block, which air can flow through to provide the necessary venting. In addition to the venting path on the sides, a 1" wide by 14" long opening on top side of each piece has been cut-out to provide additional venting. This opening and the openings on the sides will provide adequate venting to l dissipate the heat generated by the mechanical seals.

4

1 3

NS system is an accident mitigating system, and is not a contributor to the cause of any accidents listed in chapter 15 of the FSAR. The addition of the splash guards will not initiate an accident or degrade safety of ND system or any of its components. Implementation of this modification will not affect ND pump safety limits, and ND system will not be operated outside any of its design limits. An evaluation of the pumps concluded Inat no safety related equipment could be damaged if the splash guards fell. Therefore, it was acceptable for the mod to be designed as non safely related and non seismic. The probability of an accident previously evaluated in the FSAR is not increased, and margin of safety as defined in the bascs to the Technical Specifications will not be reduced. No USO exists.

MEVN-2811

Description:

The purpose of this change is to allcw wiring modifications to test lamp VX54 and test switch VX31 circuits. The "as-built" field wiring for these devices will not allow the devices or circuits to operate in their intended function. Contacts 3-4 and 5-6 of test switch VX31 do not control their designed test functions.

To resolve this problem, two conductors in the HVAC Contrni Panet will be re-located. This change will ensure normal and intended use of the affected switch and test lamp.

Safety Review and USQ Evaluation:

Ai no time did the "as-built" condition of these circuits affect the safety related function of Containment '3 Air Return Fan 1 A. The FSAR and Technical Specifications will not be affected in any significant manner due to the implementation of this change. The possibility, probability, and consequences of an accident not previously evaluated in the FSAR will not be increased due to this wiring modification. The circuits affected are considered QA Condition 1 and nuclear safety related. This change will ensure the intended function of the circuits and components is achieved. The margin of safety as defined in the bases to any Technical Specifications will not be decreased in any manner due to the re-location of the two affected VX System conductors.

MEVN-3197

Description:

The purpose of this exempt variation notice is to allow the installation of four new high point vents placed I at specific locations on the Ca System suction piping. The new vent valves are 3/4" globe valves. The valve tag numbers are 1CA215 (located near the suction of the Turbine-Driven Pump), ICA216 (located on the 18 Train assured makeup line between 1CA116B and 1 RN162B),1 CA217 (located near the suction of the 1 A Motor-Driven Pump), and 1CA218 (located near the suction of the 1B Motor-Driven Pump).

The installation of the new vent valves will not compromise the design or function of the CA System in any manner. It will provide improved system venting capability following periodic or corrective maintenance.

Safety Review and USQ Evaluation:

The FSAR and Technical ')ecifications will not be affected in any significant manner due to the implementation of this exempt variation notice. The possibility, probability, or consequences of an accident -

not previously evaluated in the FSAR will not be increased by the installation of these high point vents.

The margin of safety as defined in the bases to any Technical Specification will not be decreased in any way due to the installation of these valves. The system affected by this MEVN is OA Condition 1 and nuclear safety related. All material and components used for this exempt change are compatible for use on the Auxiliary Feedwater System.

MEVN-3203 l

9

Description:

MEVN-3202 will add 3/4 inch stainless steel tubing from an existing RN vent closed by 1RN1060 to the Ground Water (WZ) sump where the tubing will be connected to a 1 mch stainless steel pipe to be mounted on the side of the sump. Valve 1RN1006 will be added in the piping mounted on the WZ sump and will be sei to throttle flow from the vent line. The 3/4 inch tubing will field routed and supported using the instrument tube routing criteria for McGuire and thus will be seismically qualified. The attachment to

• the RN vent line will be made according to the requirements of the instrument tube routing specification to ensure that the existing RN vent and process piping remain qualified. The added valve 1RN1066 and ,

1 inch piping associated with this valve will be seismically mounted and qualified to Duke class F criteria. .t Valve 1 RN1060 will normally be operated in the open position while new valve 1 RN1066 will be operated in a throttled position to limit water flow from the RN system into the WZ sump. The total flow into the sump from RN vents will be limited to ensure that no flooding will occur as a result of this venting during ,

any anticipated accident or loss of sump pump. No USO exists.

Safety Review and USO Evaluation:

The .RN system is not an accident initiator for accidents evaluated in the SAR other than plant flooding.

The added tubing is routed to seismic Duke Class F criteria and the r ew piping supported to Duke Class F criteria to ensure that no pipe faFure will occur and cause ficoding Therefore, there is no incrsre in the probability of an accident as evaluated in the FSAR. The venting of air and water through the vent during operation will not impair the RN system from supplying required flow at the necessary pressure to any equipment or system. Therefore, the consequences of any accident evaluated in the SAR are not increased.

The added piping does not change the use or function of RN or other plant systems. The operating parameters of the affected systems arid components are not changed by this modification. There: ore, no _

possibility of an accident of a different type than evaluated in the SAR is created by this modification. The _ ,

vent line is used to remove pockets of air from RN water svhich is available for use as supply to the Auxiliary Feedwater (CA) pumps during accidents or events requiring use of the Safe Shutdown Facility.

This modification decreases the probability of malfunction of the pumps should use of this suction path be needed. Therefore, there is no increase in the probability of a malfunction of equipment important to safety as evaluated in the SAR.

This modification does not affect normal operation of safety related equiprnent, and helps ensure availability of a suction source for safety related CA pumps during accidents or other transients requiring CA pump operation. Therefore, there is no increase in the consequences of a malfunction of equipment important to safety as evaluated in the SAR. Since RN system operation and other component operation is unaffected by this modification, there is no possibility for a malfunction of a different type than evaluated in the SAR.

The margin of safety defined in the basis to the Technical Specifications is related to the confidence in the fission product barriers. The RN system provides assured cooling water to accident mitigation equipment.

This modification does not impair or degenerate the ability of the RN system perform any safety function but serves to improve the quality of an available water supply for CA pump suction to help mitigate the consequences of an accident. Therefore, the modification does not reduce the margin of safety defined  !

in the basis to the Technical Specifications. No USO exists. t MEVN-2798

Description:

)

l

. +

I The purpose of this exempt variation notice is to allow the movement of Auxiliary Building Ventilation System differential pressure switches from the north side of column NN-54 which is located inside the Unit 1 VA Filtered Exhaust Package to the south side of column NN-54 which is outside of the Filter Package.

No change to the design, function, or setpoint for the differential pressure switches will be performed under this exempt variation notice. This change is an enhancement to future periodic and/or corrective maintenance on these instruments. The change involves remounting the DP instruments, extension of the high and low sioe impulse tubing of the DP instruments through the filter package wall, and re-routing the existing instrument cables. No new cables are required. All existing filter package cable penetrations -

will be positively sealed.

i Safety Review and USQ Evaluation:

The FSAR and Technical Specifications will not be affected in any significant manner due to the implementation of this exempt change. The possibility, probability, or consequences of an accident not previously evaluated in the FSAR will not be increased in any significant manner due to the movement of these differential pressure switches. The margin of safety as defined in the bases to any Technical Specification will not be decreased due to the implementation of this exempt change. Two of the three instruments affected are OA Condition 1, but none of the affected differential pressure switches serve a safety function. The system is not inoperable due to the loss of any of these instruments. No USO exists.

MEVN-1296

Description:

Duke has committed to the NRC to test all safety related actuators t y U2EOC7. This test is being conducted to assure that the actuators will perform their intended function under design differential ,

pressure in the event of an accident.

Safety Review and USQ Evaluation:

Adding thrust values will not degrade any system important to safety. Thrust values have been calculated based on design differential pressure, and packing force. The latter value is obtained from the valve manufacturer. The total thrust valve is a combination of the two forces. No functiona! changes will be made to any system important to safety. No USO exists.

r MEVN-2196 a

Description:

1 The purpose of this exempt variation notice is to allow the adjustment of the closing thrust on the valve actuators for 1 YC27B,1 YC30A,1 YC38A, and 1 YC39B. This adjustment will enhance the affected valves ,

ability to correctly function under their design conditions. No wiring modifications will be performed on these valves, only torque switch adjustments will be made. The normal torque switch bypass adjustment for clearing the valve seat by the valve gate will be checked for proper operation during the opening cycle.

Safety Review and USQ Evaluation:

The FSAR and Technical specifications will not be affected in any significant manner due to the implementation of this exempt variation notice. Essentially, this modification willimprove the valves ability  ;

to function as designed. All components affected are OA Condition 1 and nuclear safety related. the j possibility, probability, or consequences of an accident not previously evaluated in the FSAR will not be  !

increased by the performance of this modification. No new accident scenarios are created. The margin of safety as defined in the bases to any Technical Specification will not be reduced. No USO exists. l l

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4 ME-3207

Description:

The purpose of this minor modification is to allow the installation of two new high point vents placed at specific locations on the CA system suction piping. The first new high point vent will be installed on the 2B train assured makeup line adjacent to check valve 2CA166 located in the Turbine-Driven Pump Room.

This new 3/4" globe valve will be tagged as 2CA215. The second new high point vent will be installed on the 2A train assured makeup line adjacent to isolation valve 2CA15A located in the Turbine-Driven Pump room. The installation of the new vents will not compromise the design or function of the Auxiliary Feedwater System in any manner. The second nuA J/4" globe valve will be tagged as 2CA214. The general purpose of these new vents is to remove any air entrapped in the suction line to prevent the possibility of pump malfunction.

Safety Review and USQ Evaluation:

The possibility, probability, or consequences of an accident not previously evaluated in the FSAR will not be increased due to the installation of these new system vents. The FSAR and Technical Specifications will not be affected in any significant manner due to the implementation of this tr.inor modication. The margin of safety as defined in the bases to any Technical Specification wiii not be decreased due to this system addition. The system affected by this minor modification is OA Condition 1 and nuclear safety related. All material used for this minor modification is compatible for use on the Auxiliary Feedwater System.

ME-3112 The purpose of this exempt variation notice is to allow the setpoints of VA differential pressure switches 2VAPS9410 and 2VAPS9411 to be revised from 0.2 INWD increasing to 0.2 INWD increasing. This minor modification should enhance the operation of these differential pressure switches. The existing setpoint of 0.2 INWD does not allow for reliable switch reset action since the deadband of this type of Solon pressure switch is approximately 0.17 INWD. The current differential pressure created by the manual volume damper located in the bypass line around the Unit 2 filtered Exhaust Package is approximately 1.0 INWD. With this level of differential pressure, the switches should have not problem with sensing the new instrument setpoint or allowing proper reset action.

Safety Review and USQ Evaluation:

This exempt variation notice will not compromise the design or function of the affected instrument loops in any manner. This change should enhance their operation. Although receiving 1E electrical power, these switches do not perform any safety related function. Their output is for indication only. the FSAR and Technical Specifications will not be affected in any significant manner due to the implementation of this exempt variation notice. The possibility, probability, or consequences of an accident not previously evaluated in the FSAR will not be increased in any significant manner due to the setpoint revision of these differential pressure switches. The margin of safety as defined in the bases to any Technical Specification will not be decreased due to the implementation of this MEVN. The VA System is not inoperable during .

the implementation of this change. No USO exists.

ME-3203 This modification will replace the 2/16" washer of the sleeve anchor of the EVCA battery rack base plate with a 1/4" Unistrut plate. The seismic performance of the base plate will be improved by this modification.

Safety Review and USQ Evaluation:

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The function of the EVCA battery is not affected by this change. No new failure modes are created. The EVCA battery is used for accident mitigation and it not an accident initiator. Therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased. Replacement of the washer to a thicker one does not adversely affect the function of the EVAC battery but improves the seismic performance of the battery rack. The consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased. This modification will not affect any safety limits, setpoints, or core parameters so the margin of safety as defined in the bases to any technical specification will not be reduced. No USO exists.

ME-3124 The purpose of this exempt variation notice is to allow the setpoint revision for differential pressure switches associated with the NS and ND pump-motor air handling units. In addition, the low pressure impulse line (tubing) for instruments 1VAPS9040,1VAPS9050.,1VAPS9070 will be revised such that they connect with the low ptessure impulse lines associated with instruments 1VAPS5080,1VAPS5090, ,

1VAPSS100, and 1VAPS5110 respectively. The differential pressure sensed by all or the above listed instruments will be across the air handling unit fan instead of the entire air handling unit. This setup provides a larger and more reliable differential pressure signal for monitoring by these instruments. None of the above listed instruments provide a safety related function. Four of the instruments provide control board flow indication while the other four provide flow indication to the unit operator aid computer.

Safety Review and USQ Evaluation:

The FSAR and Technical Specifications will not be affected in any significant manner due to the implementation of this exempt variation notice. The possibility, probability, or consequences of an accident will not be increased in any significant manner due to the implementation of this exempt change. Four ,

of the instruments are considered OA Condition 1 only due to being supplied by 1E power. The primary function of all affected instruments is indication only -- not control. The margin of safety as defined in the bases to any Technical Specifiation will not be decreased in any manner due to the setpoint and impulse line revision for these affected instruments.

An indirect function of differential pressure switches IVAPS9040,9050,9060 and 9070 is possible to detect a obstructed air handling unit coil condition. This will not be possible with the new impulse line reuting which monitors the f an DP only. This is not considered a significant problem since other methods are available for detecting AHU problems such as the NS and ND pump-motor bearing and stator temperatures, compartment temperature monitoring, and an effective preventive maintenance program on these air handling units. The current PM program for these AHU's consists of a specific component and overall unit inspection on a six month cycle. This PM cannot be modified without an engineering analysis currently performed by the Component Engineering Group.

MEVN-2722 and MEVN-2725 Description.

l The Diesel Generator Fuel Oil System (FD) storage tank vent lines are not protected from tornado-  ;

generated missiles. The vent lines are the only part of these tanks that are above ground. Damage to the two vent lines (one per tank), which are located only 17 feet apart on each unit, could prevent fuel oil from being pumped to the DGs from the storage tank in the unlikely event that the lines are crimped shut.

In addition to damaging the tank vent lines, the same tornado could damage the plant switchyard, cause a Loss of Offsite Power (LOOP), thus necessitating diesel start. The initial supply of fuel oil to the diese9 will be exhausted from the day tank in approximately 30 minutes with the day tank at its Tech. Spe:.

minimum volume. The vent piping will be modified by adding a breakaway flange, with screens to preveat foreign matter from clogging the vent line.

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Safety Review and USQ Evaluation:

One of the safety-related design bases of the FD system is to supply fuel di to the diesel engines that  ;

drive the DGs. The system allows the diesels to operate for a period of several days without refueling.

The vent lines are class C, carbon steel, atmospheric pressure, Nuclear Safety Class 3, QA condition 1, seismic category I.

All fuel oil is delivered by tanker truck and stored in the 50,000 gallon FD storage tanks. Oil is then -

transferred by the fuel oil transfer pump to the 275 ga!!on day tank, which, when full, allows approximately a one hour supply of fuel to be stored near the diesel.11 a diesel's day tank is not full, the fuel oil transfer pump is activated by a level switch on the tank to fill it to its proper level.

Per Technical Specification 3.8.1.1, in order to maintain operability of each diesel, the following are required:

1) a separate day tank (containing at least 120 gallons of fuel)

2) a separate Fuel Storage System, i.e. tank, (containing at least 28,000 gallons of fuel)

The diesels are required to start automatically for a LOOP and/or a Safety injection signal.

Tornado-generated missiles are evaluated in FSAR section 3.5.2.8. Safety-related structures, systems, and components are required to be protected from the effects of these missiles. Regulatory Guide 1.117, although not adopted for McGuire, states that the physical separation of redundant components, e.g. the tank vent lines, is generally not considered acceptable by itself for protection against tornado effects. In addition, the joint occurrence of low-probability events such as an earthquake with a tornado is sufficiently small to exclude from consideration.

The FD system is an accident mitigation system, and is not a contributor to the cause of any accidents listed in FSAR Chapter 15. Vents are constructed to OA 1 standards; no SSCs will bt dejraded. The modification will upgrade the vent from piping class F to piping class C. The pressure equalization function of the vent lines is maintained. Therefore, the probability of an accident er malfunction of equipment _important to safety previously evaluated in the FSAR are not increased.

The operating characteristics of the FD system will not be changed by the MEVNs. The FD system will still be capable of adequately supplying fuel oil to the diesels under all conditions. The vent lines will be ,

operable after a tornado missile strike. The vents will be inspected af ter a tornado to ensure that they are not obstructed. No new credible failure modes are created. Therefore, these actions will not increase the consequences of an accident or a malfunction of equipment important to safety.

i In addition, the possibility of an accident not already evaluated in the FSAR, or the possibility for a

' malfunction of a different type is not created, since no new failures are created and no previously incredible accidents are made credible by the MEVNs. No foreign object could permanently clog the vent, should it be sheared off.

These conditions of operability do not degrade any fission product barriers. These actions do not affect i any plant safety limits, setpoints, or design parameters. No changes to the Technical Specifications are required. Therefore, the margin of safety as defined in the bases of the Technical Specifications will not be reduced. No USO exists.  ;

a MEVN-2937

Description:

I The purpose of this change is to install a remote leak rate test system for Containment Purge (VP) inside and outside containment isolation valvesi i

There are a total of nine containment penetrations for the VP system, five for purge air supply and four *

- for purge air exhaust. These penetrations are in the upper compartment and lower compartment.

Installation of the remote leak rate test system is proposed for each of the VP containment penetrations.  ;

Safety Review and USQ Evaluation:

The purpose of the Containment Purge System is to reduce radioactivity in containment by supplying fresh .

air which, in turn, is exhausted through cleanup filters and unit vent to the atmosphere. This system t performs no safety function during a LOCA and is not Nuclear Safety Related, however, the isolation valves and connection piece between the isolation valves are Duke class B, OA condi%n 1. This modification will be installed as OA condition 1 since the instrument tubing for the remote leak rate test >

will be tied-in to the existing penetration pressure test isolation valves located on the piping between the containment isolation valves.

The existing penetration pressure test isolation valves will be placed in the OPEN position and secured.

Instrument tubing will be routed from the existing pressure test isolation valves to the new pressure test isolation valves. The new pressure test isolation valves will be normally closed, and will only be used for containment penetration testing. The instrument tubing will be 3/8" stainless steel and will be routed along with reactor building concrete wall. Expansion loops will be installea on all tubing routes to minimize tubing stresses caused by expansion or eontraction. All tubing will be installed in accordance with existing installation specifications for class 6S? .nstrumentation. The new pressure test isolation valves will be 1/2 inch Dragon valves. These valvec are compatible for use in liquid and gaseous systems, and have demonstrated excellent characteristics against leakage through the valve seat as well as the valve stem.

Implementation of this exempt change will facilitate maintenance work on the VP system, and it will both degrade safety of the system. Probability and consequences of an accident previously evaluated in the FSAR will not increase, and the possibility of an accident different than any already evaluated in the FSAR will not be created. The margin of safety as defined in the bases to any Technical Specification will not be reduced. No USO exists.

ME-3204

Description:

This modification will replace the 3/16" washer of the sleeve &nchor of the EVCC battery rack base plant with a 1/4" Unistrut plate. The seismic performance of the base plate will be improved by this modification.

Safety Review and USQ Evaluation:

The function of the EVCC battery is not affected by this change. No new failure modes are created. The EVCC battery is used for accident mitigation and as such is not an accident initiator. Therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the -

FSAR will not be increased. Replacement of the washer to a thicker one does not adversely affect the function of the EVCC battery, but improves the seismic performance of the battery rack. Therefore, the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.

This modification will not affect any safety limits, setpoints, or core parameters so the margin of safety as defined in the bases to any Technical Specification will not be reduced. NO USO e'xists.

MM-3594, MM-3604, MM-3606 b

Description:

These minor modifications are to install a portable waste processing unit consisting of two 50 cubic foot demineralizes with an associated booster pump and both a bag filter skid and a dual bag filter skid. This i addition along with two 25 cubic foot portable demineralizes currently being used will allow bypass of existing domineralizes and associated filters when desired. The additional waste processing equipment <

will be located in the waste processing building truck bay area and connected to the existing waste system -

through valves 1WL-1054, -1055, -1058 and -1059. Since the waste processing equipment is portable, it is connected to the existing piping using flexible hose which is qualified for the service conditions by the vendor. All components associated with the waste processing units such as valves and instrumentation  ;

are provided as part of the unit package by the vendor (Chem-Nuclear). Only MM-3605 is to install the portable unit. MM-3594 is written to modify building in floor load analysis to qualify the building to carry the additional loads associated witn the unit and MM-3604 details the installation of a 600v power supply  :

to the unit.

Safety Review and USQ Evaluation:

The waste processing system is not safety related and neither creates an accident analyzed in the licensing basis for the plant nor acts to mitigate the consequences of any analyzed accident. This .

equipment neither acts to prevent an accident nor is relied on for accident mitigation. Failure of the 1 equipment such that the process fluid is released may however result in release of radioactive fluids. The  !

equipment is located in a seismically qualified, curbed area with a capacity to contain a spill of 25000 gallons of liquid. Additionally, a sump pump is located in the spill centainment to transfer spilled fluid to ,

a seismically qualified tank and thus contain any credible sp!!! The Auxiliary Floor Drain Tank and Auxiliary Waste Evaporator Feed Tanks which may be used to feed the demineralizes are limited to 25000 gallons capacity by administrative controls. The building structure is qualified to carry all loads identified with the addition of this equipment. The mode of operation will therefore be similar to the current operation of the portable demineralizes in the area. The added power supply to the unit is not safety related and does not impact safety power to any equipment. No safety equipment is located in the area  !

with the waste processing unit or the spill containment. Therefore, the failure of this equipment will not impact the ability of any safety equipment to perform its function. The equipment is located and operated ,

to meet the intent of Regulatory Guide 1.143 and ALARA principles for prevention of spills.

The waste processing equipment is not considered in any SAR evaluated accident. The failure of this -

equipment will be contained by seismic structures and by sump pump systems in the drainage area of the equipment, such that no otisite release would be expected from this equipment. Therefore, there is no increase in the probability of an accident evaluated in the SAR. The equipment is not located with safety -

related equipment and does not directly interact with safety systems. Therefore, there is no impact on the consequences of any accident evaluated in the SAR as a result of these modifications. The added portable demineralizes are similar in function and operation to current equipment located in the same area.

Thus, no new accident or equipment malfunction, not previously considered, is created by this addition.

Since the equipment is not located near safety related equipment and does not interact with safety systems -

during operation, there is no increase in the probability of malfunction of safety equipment or the increase in the consequences of safety equipment malfunction as evaluated in the SAR. The margin of safety defined in the Technical Specifications is related to the confidence in the fission product barriers. The portable demineralizes do not perform any function as part of the fission product barriers. Therefore, this modification does not reduce the margin of safety defined in the basis to the Technical Specifications. No USQ exists.

MEVN-3156  !

Description:

This MEVN has been written to allow the plugging of degraded tubes in the 2A and 28 NS heat i

.e .

exchangers. As a result of eddy current testing, two (2) tubes will be plugged in the 2A NS HX and 144 tubes will be plugged in the 2B NS HX.

Safety Review and USQ Evaluation:

Design Engineering has performed a tube plugging analysis and determined that tube plugging is acceptable in these HX's up to a maximum of 194 tubes. This analysis is documented in. Calculation MCC-1223.13-00-0016.

Because of the large number of degraded tubes in the 1B NS HX, portions of three tubes have been removed from the Outlet Pass for failure analysis. These three locations will be plugged at the tube sheet with a welded tapered plug.

Since the number of tubes to be plugged does not exceed the maximum number allowed, the above referenced calculation ensures that the required heat transfer capability of the NS heat exchangers will be maintained. The possibility, probability, or consequences of an accident not previously evaluated in the FSAR will not be increased due to this plugging operation. The margin of safety as defined in any Technical Specification bases will not be decreased due to the implementation of this modification. No FSAR or Technical Specification sections are affected. The NS heat exchangers are QA Condition 1 and nuclear safety related. NO USO exists.

MEVN-3195

Description:

The purpose of this exempt variation notice is to allow the " repair" of defective Unit 1 steam generator tubes by either ." plugging" or " sleeving". Due to indications detected in previous eddy current data,.

evaluation and/or testing will be performed to determine the condition of the primary to secondary heat exchanger tubes. From the results of evaluation and/or testing, all tube defects, or in some cases, indications will be repaired by the above indicated processes. All field work will be performed using approved station and/or vendor procedures. Babcock and Wilcox personnel will perform the required maintenance activities.

Safety Review and USQ Evaluation:

A detailed LOCA analysis has been performed by Westinghouse. Babcock and Wilcox has performed a LOCA analysis for insertion of their fuel in the McGuire Units per Report BAW-10174,' Rev.1 dated November 1990 and titled " MARK-BW Reload LOCA Analysis for the Catawba and McGuire Units." In addition, the Design Engineering Safety Analysis Group has performed a LOCA Analysis for the limiting case of a Feedwater Line Break which was submitted with the McGuire 1 Cycle 8 Reload Report and subsequently approved in November 1991.

These LOCA Analyses support an " equivalent plugging" limit of 10% (467 tubes) per steam generator -

this is a maximum number.' It is not expected to exceed the 10% limit during the repair of these steam generators. As long as " equivalent plugging" does not exceed this limit, then the FSAR and Technical Specifications will not be affected in any significant manner due to the implementation of this exempt -

variation notice. The possibility, probability, and consequences of an accident previously or not previously evaluated in the FSAR will not be increased due to this repair operation. There is potential for tube plugging / sleeving to affect Reactor Coolant System flowrate which in turn could affect allowable reactor power level. The margin of safety as defined in any Technical Specification bases will not be decreased due to the implementation of this MEVN. The components affected are QA Condition 1 and nuclear safety related. No USO exists.

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MM-3417

Description:

NRC Generic Letter 89-10 instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, ,

and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power nuclear stations. Duke has developed a  ;

comprehensive program plan that describes the actions that Duke nuclear stations will accomplish in order to comply with the Generic Letter. This MM provides for the diagnostic testing for MOV 1LD0113 and  ;

constitutes part of the actions necessary for compliance to Generic Letter 89-10.

The actual changes to 1LD0113 involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the steam nut to thrust, to fully open "

and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by engineering  ;

calculation MCC-1205.19-00-0003 and is provided by controlled document MCM 1205.19-00-0039-001.

This engineering calculation was performed in accordance with the latest revision of Duke Power Specification DPS-1205.19-00-0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels for 1LD0113.

The diagnostic test system used to facilitate thrust testing for 1 LD0113 associated with the diagnostic test system used in testing 1LD0113 have been included in the engineering calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.

Safety Review and USQ Evaluation: ,

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The MOV affected by this mod is in the Diesel Generator Lube Oil System. The function of 1LD0113 is to provide a bypass of the full flow lobe oil filter on high differential pressure. The safety position of 1LD0113 is open and it's normal position is closed. Re-setting the open and close torque switches will not affect open and closure times of 1LD0113. The existing stress analysis of the piping associated with 1LD0113 will not be affected by re-setting the open and close torque switches. Since this MM will ensure >

that 1 LD0113 will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USO exists. I ME-3125

Description:

The purpose of this change is to allow the setpoint revision for differential pressure switches associated with the NS and ND pump-motor air handling units. In addition, the low pressure impulse line (tubing) for instruments 2VAPS9040,9050,9060 and 9070 will be revised such that they connect with the low pressure impulse lines associated with instruments 2VAPS5080,5090,5100 and 5110 respectively. The differential pressure sensed by all of the above listed instruments will be across the air handling unit fan instead of the entire air handling unit. This setup provides a larger and more reliable differential pressure .

signal for monitoring by these instruments. None of the above listed instruments provide a safety related function; Four of the instruments provide control board flow indication while the other four provide flow indication to the unit operator aid computer.

Safety Review and USO Evaluation: >

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The FSAR and Technical Specifications will not be affected in any significant manner due to the implementation of this exempt variation notice. The possibility, probability, or consequences of an accident will not be increased in any significant amount due to the implementation of this change. Four of the instruments are considered QA condition 1 only due to being supplied by 1 E power. The primary function of all affected instruments is indication only - not control. The margin of safety as defined in the bases to any Technical specification will not be decreased in any manner due to the setpoint and impulse line revision for these affected instruments.

An indirect function of differential pressure switches 2VAPS9040,9050,9060 and 9070 is possible to-detect a obstructed air handling unit coil condition. This will not be possible with the new impulse line routing which monitors the fan DP only. This is not considered a significant problem since other methods are available for detecting AHU problems such as the NS and ND pump-motor bearing and stator temperatures, compartment temperature monitoring, and an effective preventive maintenance program on these air handling units. The current PM program for these AHUs consists of a specific component and overall unit inspection on a six month cycle. This PM cannot be modified without an engineering analysis currently performed by the Component Engineering Group. No USO exists.

ME-3136

Description:

This modification will provide a discharge flow path for the WN system that will be independent of the WV system. Therefore, the DG sump pumps can run simultaneously with the \W pump without deadheading.

Safety Review and USQ Evaluation:

This modification does not adversely impact the WN system function with respect to flood protection in the DG rooms.

Since there is no adverse on the safety function of any system, the probability and consequences of an accident or equipment malfunction previously evaluated in the FSAR is not increased. The possibility of an accident or equipment malfunction important to safety not previously evaluated in the FSAR is not created. No USQ exists.

MM-3245 Desenption:

Performance of maintenance on the Auxiliary Boilers requires removal of weather tight metal covers and concrete hatch covers for access to the boilers. The concrete hatch covers pose a potential crane stability problem when they are handled with the station mobile crane. This minor modification documents the permanent removal of the concrete hatch covers.

Safety Review and USQ Evaluation:

The concrete hatch covers were originally required by Nuclear Mutual Limited (insurers) Standards for fire protection purposes. The standards have been revised and in this case no longer apply to electric boilers.

Therefore, the concrete hatch covers are not required.

The FSAR and the Technical Specifications are not affected. This MM does not involve any structures, systems, or components important to safety and, therefore, will not degrade the effectiveness of any equipment important to safety due to any design basis event. No USO exists.

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4 MM-3287 and MM-3288

Description:

These mms will replace Fuel Transfer isolation Valve 1(2) KF122 with a new knife gate valve, and add / modify the support restraints for the new valve. The purpose of the modifications are to eliminate steam and drive system galling problems which caused the valve to fail. Valve teplacement requires modification of two lateral supports on the transfer canal liner, and the addition of two spring hangers supported from the operating floor. The fuel transfer isolation valve is non-OA, however the supports are OA Condition 4 to ensure seismic interaction of the valve and the fuel transfer tube does not create excessive loads on the transfer tube. Installation requires removal of the existing limit switch and connection of the vendor supplied Namco EA-170 limit switch for open indication. The existing interlock wit the fuel transfer system is not changed.

Safety Review and USQ Evaluation:

The fuel transfer tube connects the refueling canal (inside the reactor containment) and the spent fuel pool (outside containment). The fuel transfer tube is closed on the refueling canal side by a blind flange at all times except during refueling operation. The fuel transfer isolation valve located on the spent fuel pool side of the fuel transfer tube serves to isolate the fuel transfer tube during refueling to allow draining of the tube, and provide a low pressure (temporary) containment isolation during refueling modes. The fuel transfer valve is normally open to provide a flow path of water from the spent fuel pool to the fuel transfer tube for suction to the standby makeup water pump for SSF.

The Fuel Transfer Isolation Valve (KF122) is not considered a containment isolation valve, neither does it have a OA1 function. Its function is to isolate the fuel pool. Valve leakage requirements have not changed. Reference to the valve being a containment isolation valve will be removed from the FSAR.

The reason for this is that the McGuire FSAR currently reflects in part the original design of the KF system -

requiring valve KF122 to be locked closed. However, when the Standby Shutdown Facility (SSF) was fully implemented and declared operable for compliance with 10CFR50 Appendix R valve KF122 was then' required to be locked open during plant operation in order to provide a suction source of borated water for the Standby Makeup Pump to be used as an attemate Reactor Coolant Pump seal injection system.

Since the Standby Makeup Pump takes suction from the spent fuel pool via the fuel transfer tube, for Standby Shutdown System (SSS) operation, this valve must remain locked open in Modes 1-4.

Environmental conditions and material requirements have been considered. The fuel transfer isolation .

valve is non-seismic. Failure of the valve will not cause fuel in the spent fuel pool to be uncovered due .

to elevation and building structural arrangement.. The fuel transfer tube to which the isolation valve is attached is part of the containment boundary and is OA1. The deadweight and seismic loads transferred to the end of the fuel transfer tube are less than the original design; therefore, the original design basis and associated calculation are considered bounding. No USO exists.

MM-3306

Description:

NSM MG-22154/00, in part, added Diesel Generator lube oil filters 2A and 2B into the DG engine tube oli -

system (LD). The new filters, located upstream of their respective DG lube oil intake strainers in the DG building, are located in Duke class C piping but were purchased ASME section Vllt with the vendor '

accepting Part 21. This is acceptable for a OA condition 1 component. The filters were also required to be seismically qualified and anchored.

A shrinkage type grout was used to anchor these components instead of a non-shrink grout.f This MM

. will add additional anchoring support for the filter including additional concrete anchor bolts. ' OA1

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materials are used and the MM will not adversely affect either the OA1 qualification or the seismic

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'i qualification of the filters. ,

Safety Review and USQ Evaluation:

The new filters are qualified for use in this application. They meet seismic, code, and QA condition 1 requirements. The LD system is designed to supply lubricating oil to the diesel engine, its bearings, crankshaft;and other friction surfaces. All design requirements have been met. No USO exists.

MM-3312, MM-3313 and MM-3314

Description:

It has been documented in VIL-W 92-09 that there is a potential for inadvertent isolation of pressurizer power operated relief valves (PORV). This would be the result of a valve stem failure causing the PORV block valve to remain closed, thus inadvertently isolating the PORV. The valve stem failure is attributed to a combination of age-related embrittlement and Stress Corrosion Cracking (SCC). These mms will replace the valve stems for PROV block valves 1NC31B,1NC33A, and 1NC35B. The new stems will be made of SB 637 (inconel718). This material was recommended by the valve manufacturer, Borg Warner Industrial Products (BW/IP Intemational, Inc.), af ter viewing the block valve design requirement including Reactor Coolant (NC) System chemistry and temperature. The new stems are expected to improve block valve integrity and reliability. No USO exists.

Safety Review and USQ Evaluation:

The block valves will no be less reliable and are available for isolation of the PORV, and otherwise will perform as before. From an accident and transient mitigation standpoint, the mms are expected to increase the likelihood that the PORVs and associated block valves will be available for performing safety related functions. No new failure modes were identified. No USO exists.

MM-3523

Description:

This MM will replace eight of the snubbers in the Diesel Generator Engine Cooling Water System (KD) with adjustable rigid struts manufactured by Anchor / Darling. The snubbers being replaced are those expected to have small movements (less than 1/8 inch). The replacement is a drop-in (pin to pin) replacement.

The replacement will also allow deletion of 3 spring supports.

Safety Review and USQ Evaluation:

The purpose of this modification is to reduce snubber maintenance as required by the Technical  ;

Specifications, Section 3/4.7.8. The subject snubbers are located on Duke class C lines. The same analysis methodology as before is used. 1 The piping system and support / restraints replaced were done so in consideration of loads, load combinations, and allowable stress criteria in conjunction with applicable codes as discussed in the FSAR.

The replacement of the snubbers with the adjustable rigid struts will not create any new line break 'i locations or adversely affect previously evaluated break locations. No redundancy or separation criteria are violated by this MM. Since the system has been re-analyzed and found acceptable, no new failure modes are manifested as a result of this modification. No USO exists.

MEVN-5058, MEVN-5066

Description:

p-These modifications will remove the VG Control Tank, located in the control air supply line for the 2A and 2B Emergency Diesel Generators. The tanks are to be removed due to maintenance considerations. The tanks will be replaced by 3/4" Class C carbon steel tubing.

Safety Review and USQ Evaluation:

The Starting Air System (VG) supplies compressed air to the diesel generator controls and instrumentation as the engine starts and continuously while the engine is running. Sufficient control air pressure to keep the diesel running shall be maintained. The ability of the system to meet this requirement is verified periodically by test. The control air tanks were part of the vendor package for the diesel generators, but served no functional purpose except providing .7 cubic feet of reserve control air capacity. Performance tests and Technical Specification tests have been performed which show that the diesel generator performance, without the tanks, remains acceptable. No accidents or vents are initiated by the emergency diesel generators of the VG system. The QA condition 1 tubing added to replace the control air tanks meets the material requirements for the application. Seismic qualification for the tubing is provided. There are no electrical changes involved with the modifications. No Appendix R or separation criteria is applicable. The emergency diesel generators supply emergency power, required by equipment used for accident mitigation, in the event of loss of offside power (LOOP). The modification involves no functional changes to the emergency diesel generators and associated VG system. No common failure modes are created by this modification. This modification adds no new functions to the emergency diesel generator of VG system. No new failure modes are created. No accidents previously considered incredible are made credible by this modification. Successful operation of the diesel generator is essential in preserving the Technical Specification safety limits and all other accident analysis assumptions that involve a LOOP.

Since diesel generator tests show acceptable performance, the fission product barriers (RS pressure boundary, containment, fuel pellets, and cladding) will not be degraded. No assumptions made in any accident analysis are affected by the modification. No USO exists.

MM-5130

Description:

This modification will update design drawings for the Auxiliary Feedwater (CA) Pump Control Panels, to show fuse and terminal block information that was not shown on the original drawings. The existing fuses within the CA panels will be changed out to FNO-type to be consistent with present fusing standards, and fuse blocks will be changed out as necessary. The drawing discrepancies were identified during a panel inspection, mandated by a previous NRC commitment to inspect all panels for proper fusing.

Safety Review and USQ Evaluation:

There are no accidents or events initiated by the CA system. The CA Pump control Panels are OA condition 1. The fuses and fuse block meet GA condition 1 and 4 (seismic mounting) specifications. An Appendix R Review was completed with no concerns identified. The FNO-type fuses to be substituted were sized according to established fusing criteria. The CA system is used for accident mitigation, and t therefore, performs plant safety functions. The fusing equipment will perform the same functions as prior to the modification. No common failure modes are created by the fuse changeout to FNO-type. The fuse changeout does not degrade the safety functions of the CA system. No new functions are added and no new failure modes are created by the modification. No accidents previously considered incredible are made credible by the modification. The FNO-type fuses are at least as reliable as the existing fuses.

There are no changes of safety limits, setpoints, or plant parameters because of the modification. The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are.not -

degraded. No assumption madein any accident analysis are affected by the modification. No USO exists.

MM-5125

Description:

This modification updates the electrical load lists for load centers KXA,'1MXA, and 1EMHA, as follows, to refiect as-built conditions. No physical changes are required in the plant.

Pane! Affected Load Chance System i 1EMXA Valve ICA-15A Change listed horsepower Aux. Feedwater from 6.38 to .38 1MXA Valve 1CF-26 Change listed horsepower Main Feedwater from 1 to 2 i 1MXA Valve 1CF-35 Change listed horsepower Main Feedwater from 1 to 2 KXA Unit 1 VP Sys. Correct load description Containment Purge  :

Safety Review and USQ Evaluation:

Panels KXA and 1MXA are non-essential. EMXA is safety-related. The CA and VP systems are not accident initiators in any accident analyses. " Loss of Normal Feedwater Flow"is an applicable for the CF system. The CF system performs no accident mitigation functions. The CA and VP systems do perform accident mitigation functions. The electrical load list will be updated to correct the horsepower listed for ,

two CF and one CA system valve motors. The actual motor horsepower will not be changed. The load list update involves no changes in electrical loading on the load centers. Seismic, separation, or Appendiv.

R criteria are not applicable, because no pN hl plant changes are involved. No common failure modes - ,

are created. The load list update involves no change in VP system electrical loads, it corrects the listing of the load distribution point description. The performance of CA and VP system accideat mitigation i functions are not degraded by these editorial changes. There are no new functions added by the load list update. No new failure modes are created. No accidents previously considered incredible are made credible by the editorial revisions. There are no changes of safety limits, setr%nts, or plant parameters e because of the electrical load list update. The fission product barriers (RCS boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in ar') accident analysis are affected by -

the changes. No USO exists.

MEVN-5175

Description:

l This modifications modifies the mounung configuration for the door panels on the DG battery charger ']

cabinets (EPO system), located h kne DG rooms. The modification will allow easier access to the cabinets i for routine weekly maintenance on the DG batteries. The modification is applicable to Units 1 and 2, Train l A and B DG battery charger cabinets. Each existing door panel (two per charger cabinets) is attached  ;

with 18 screws spcced B" apart around the door perimeter. This type of mounting was used because the l door panels are required for the seismic integrity of the charger cabinets. Instead of continued reliance ;j on the door panels, the proposed modification adds cross-bracing within the cabinets to provide the i required seismic integrity. The door panels will be modified and re-installed to serve as covers for the I access opening. - The DG battery chargers are OA Condition 1. Material that meets OA condition 1 ,

requirements will be used for the cabinet bracing. Material that meets OA 4 requirements (or better) will be used to fasten the cabinet doors.

Safety Review and USQ Evaluation:

The EPO system is not evaluated as an accident initiator in any accident analyses. The modification was reviewed and approved to meet seismic requirements. OA condition 1 requirements will be met for the cabinet bracing. OA Condition 4 requirements will be met by the door fasteners. The modification involves no electrical changes. No electrical separation or Appendix R requirements are applicable. The diesel generators perform the plant safety function of providing power for plant safety equipment in case of loss of offsite power. The DG batteries supply power for DG starting. Seismic integnty of the battery  ;

charger cabinets is maintained by the proposed modification, therefore, the performance of plant safety ,

functions are not degraded. There will be no change in EPO system functions, and no failure modes are  !

created in common with the other DG units / trains. There are no new functions or failure modes created by the proposed modification. No accidents previously considered incredible are made credible by the modification There are no changes of safety limits, setpoints, or plant parameters because of the modification. The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the modification. No USO exists.

MM-3287 and MM-3288

Description:

These mms will replace Fuel Transfer Isolation Valve 1(2) KF122 with a new knife gate valve, and add / modify the support restraints for the new valve. The purpose of the modifications are to eliminate stem and drive system galling problems which caused the valve to fail. Valve replacement requires modification of two lateral supports on the transfer canal liner, and the addition of two spring hangers supported from the operating floor. The fuel transfer isolation valve is non-OA, however the supports are OA Condition 4 to ensure seismic interaction of the valve and the fuel transfer tube does not create excessive loads on the transfer tube. Installation requires removal of the existing limit switch and connection of the vendor supplied Namco EA-170 timit switch for open indication. The existing interlock with the fuel transfer system is not changed.

Safety Review and USO Evaluation:

The fuel transfer tube connects the refueling canal and the spent fuel pool. The fuel transfer tube is closed on the refueling canal side by a blind flange at all times except during refueling operation. The fuel transfer isolation valve located on the spent fuel pool side of the fuel transfer tube serves to isolate the fuel transfer tube during refueling to allow draining of the tube, and provide a low pressure (temporary)  !

containment isolation during refueling modes. The fuel transfer valve is normally open to provide a flow path of water from the spent fuel pool to the fuel transfer tube for suction to the standby makeup water [

pump for SSF.

The Fuel Transfer Isolation Valve is not considered a containment isolation valve, neither does it have a OA1 function. Its function is to isolate the fuel pool. Valve leakage requirements hav6 not changed.

Reference to the valve being a containment isolation valve will be removed from the FSAR. The reason for this is that the McGuire FSAR currently reflects in part the original design of the KF System requiring valve KF122 to be locked closed. However, when the Standby Shutdown Facility was fully implemented and declared operable for compliance with 10CFR50 Appendix R, valve KF122 was then required to be locked open during plant operation in order to provide a suction source of borated water for the Standby Makeup Pump to be used as an alternato Reactor Coolant Pump sealinjection system. Since the Standby Makeup Pump takes suction from the spent fuel pool via the fuel transfer tube, for Stanoby Shutdown System (SSS) operation, this valve must remain locked open in Modes 1-4. Environmental conditions and material requirements have been considered. The fuel transfer isolation valve is non-seismic. Failure of the valve will not cause fuel in the spent fuel pool to be uncovered due to elevation and building structural arrangement. The fuel transfer tube to which the isolation valve is attached is part of the containment boundary and is OA1. The deadweight and seismic loads transferred to the end of the fuel transfer tube are less than the original design; therefore, the original design basis and associated calculation are r -+- -

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i considered bounding. No USO exists.

MM-3306

Description:

NSM MG-22154/00 in part, added DG Lube Oil Filters 2A and 2B into the DG Engine Lube Oil System (LD). The new filters, located upstream of their respective DG lube oil intake strainers in the DG building, are located in Duke Class C piping but were purchased ASME Section Vllt with the vendor accepting Part

21. This is acceptable for a OA Condition 1 component. The filters were also required to be seismically qualified and anchored.

A shrinkage type grout was used to anchor these components instead of a non-shrink grout. This Minor Modification (MM) will add additional anchoring support for the filter including additional concrete anchor ,

bolts. QA1 materials are used and the MM will not adversely affect either the OA1 qualification or the seismic qualification of the filters.

Safety Review and USQ Evaluation:

The new filters are qualified for use in this application. They meet seismic, code, and QA Condition 1 requirements. The LD System is designed to supply lubricating oil to the diesel engine, its bearings, crankshaft, and other friction surfaces. All design requirements have been met. No USO exists.

MM-3312, MM-3313, MM-3314 *

Description:

It has been documented in VIL-W 92-09 that there is a potential for inadvertent isolation of pressurizer power operated relief valves. This would be the result of a valve stem failure causing the PORV block valve to remain closed, thus inadvertently isolating the PORV. The valve stem failure is attributed to a combination of age-related embrittlement and Stress Corrosion Cracking. The mms w!il replace the valve steams for PROV block valves 1NC31B, INC33A, and 1NC358. The new stems will be made of SB 637 (inconel 718). This material was recommended by the valve manufacturer Borg Warner industrial products (BW/IP International, Inc,), after reviewing the block valve design requirements including Reactor Coolant (NC) System chemistry and temperatures. The new stems are expected to improve block valve integrity and reliability.

Safety Review and USQ Evaluation:

The block valves will be na less reliable and are available for isolation of the PORV, and otherwise will -1 perform as before. Form an accident and transient mitigation standpoint, the mms are expected to increase the likelihood that the PORVs and associated block valve will be available for performing safety ,

I related functions. No new failure modes were identified. No USO exists.

MM-3523

Description:

This MM will replace eight snubbers in the Diesel Generator Engine Cooling Water System with adjustable rigid struts manufactured by Anchor / Darling. The snubbers being replaced are those expected to have small movements (less than 1/8 inch). The replacement is a drop-in (pin to pin) replacement. The rerfacement will also allow deletion of 3 spring supports. <

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. Safety Review and USQ Evaluation: i The purpose of this modification is to reduce snubber maintenance as required by the Technical 1 Specifications, Section 3/4.7.8. The subject snubbers are located on Duke Class C lines. The same .

analysis methodology as before is used.

The piping system and support / restraints replaced were done so in consideration of loads, load 'l combinations, and allowable stress criteria in conjunction with applicable codes as discussed in the FSARc  ;

.The replacement of the snubbers with the adjustable rigid struts will'not create any new line break. l locations or adversely affect previously evaluated break locationse No redundancy or separation criteria .

are violated by this MM. Since the system has been re-analyzed and found acceptable, no new failure g modes are manifested as a result of this modification. No USQ exists.

MM-5520 j

Description:

i The setpoint for ten Unit 1 and eight Unit 2 Auxiliary Feedwater (CA) suction pressure switches will be y change to 3 psig. The setpoint for two Unit 2 CA suction pressure switches will be changed to 4 psig (the J zero reference elevation for these switches is at a lower elevation than for the other switches). The CA ll suction pressure switches initiate automatic swapover to RN on decreasing CA pump suction pressur.e for  ;

their respective CA pumps. The setpoint change increases the suction pressure at which the CA suction -!

swaps to the Nuclear Service Water (RN) Assured Supply upon decreasing suction pressure. The new _

pressure switch settings facilitate RN swapover above a minimum static water elevation of 730' + 10" for - y Unit 1 and 731' + 7-3/8" for Unit 2. The setting change increases the probability that swapover will occur ,

~

a before it is no longer feasible to obtain water from the normal suction' sources (upper surge tank, CA- -l condensate storage tank, or the condenser hotwell).  ;

Safety Review and USQ Evaluation: ,

The CA system is safety-related, and is used for normal plant shutdown, accident mitigation, and can be' .(

used during startup. The RN Assured Supply to CA system is safety-related.- The CA system is not an  !

I accident initiator addressed in any accident analyses. .The reliability of the RN system Assured Supply to the CA system is increased by the setpoint change. The CA system is an accident mitigator for any 's emergency or normal shutdown operation that requires decay heat or reactor coolant heat dissipation. l The performance of this safety function is not degraded by the setpoint change. . There is no change of 4 system functions and no common failure modes are created between CA' system trains. The modification adds no new CA or RN system functions. No new failure modes are created. No accidents previously i considered incredible are made credible by these modifications. The Technical Specifications give the ' )

allowable and trip setpoints for CA Suction Pressure. The allowable setpoint is greater than 1 psig, and 1 the trip setpoint is greater than or equal to 2 psig. The change of pressure switch setpoints to 3 psig or 4 psig is in the conservative directiori. The fission product barriers are not degraded. No assumptions made in any accident analysis are adversely affected by the modificstions. No USO exists.' .

i 1

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.- ~ .-- - - -

,s Duke Power Company McGuire Nuclear Station Summary of Procedure / Miscellaneous Changes Under 10CFR50.59 MP/0/A/7700/81 The purpose of this procedure is to allow tube plate hole sealing by tapered weld plug for Westinghouse Model D2, D3 and D6 Steam Generators. This procedure would be utilized when normal steam generator ,

tube plugging methods are not possible, such as when tubes are removed from the tube plate region. Use of this procedure will be controlled by Nuclear Station Modification or Exempt Variation Notice These documents will provide information and detail on current licensing basis tube plugging limits.

Safety Review and USQ Evaluation:

The plugging of SG tubes has an impact on the Emergency Core Cooling System accident analysis as presented in Chapter 15 of the FSAR. Plugging also has an affect on important plant performance characteristics. Reactor Coolant System flowrate is significantly alterer by tube plugging operations. As ,

long as steam generator tube plugging is held within existing licensin[, basis limits then performance of this procedure will have no affect on the FSAR or Technical Specifica' ans.

This maintenance procedure modified QA Condition 1 and nuclear safety related equipment. The possibility, probability, and consequences of an accident not previously evaluated in the FSAR will not be significantly increased due to the performance of this procedure. No Technical Specification changes are mandated by the implementation of this procedure. No USO exists.

TN/2/A/9700/062 The purpose of this procedure is to provide the necessary guidance in preventing personnel injury and inoperability of the WN system during implementation of exempt change 3136.

Safety Review and USQ Evaluation:

Requiring valves 1WC-199 and 1WC-200 to be positioned in the open position prior to installation will prevent the flow path of the WN pumps from becoming blocked. The final position of the valves will be verified by Operations to be in accordance with the valve checklist of the corresponding OP procedure.

This procedure does not adversely impact the WN system function with respect to flood protection the DG rooms because the sump pumps are functional and a flow path out of the DG room exists at all times during implementation of the modification. There is no adverse impact on the safety function of any other system in the DG room. No USO exists.

PT/0/B/9100/394 This procedure provides for periodic blowdown and inspection of the Instrument Air (VI) system at random points within the system. Its purpose is to satisfy a conditional operability requirement of PIR 0-M90-0205. This PIR was generated after it was determined that the VI dryers are unable to lower system dewpoints below the acceptance criteria of PT/0/B/4453/04, VI Air Quality - Dewpoint Measurement. One i of the conditions of operability specified by the PlR was to periodically blowdown and inspect the VI system l to ensure that excessive amounts of moisture are not collecting inside of system piping or components. I Safety Review and USQ Evaluation:

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This test provides for controlled and short blowdowns of the VI system for the purposes of determining if l

, any excessive amounts of moisture are collecting in the system. Personnel performing this test are instructed to monitor VI system pressure when performing blowdowns. They are also cautioned against opening blowdown valves so far that major system pressure fluctuations occur. If a blowdown valve was opened too f ar, the VI compressors would have the capacity to maintain Vi system pressure above the low .i pressure alarm setpoints. In the unlikely event the compressors were unable to maintain system pressure, the system low pressure alarms in the control room would alert plant personnel and the blowdowns would be terminated.

Based on the above, the performance of this test will not increase the probability or consequences of an accident previously evaluated in the FSAR. In addition, this test will not create the possibility of an accident occurring which has not been p,eviously evaluated in the FSAR. The probability or consequences of FSAR evaluated equipment malfunctions will not be increased. Also, the possibility of unevaluated equipment malfunctions occurring will not be created. A review of Technical Specifications revealed no margins of safety which will be affected by implementation of this test. No USO exists.

TN/0/A/2176/00/AE1

This procedure is to provide instruction for implementing NSM MG-52176

Replacement of the existing

' 125 VDC Vital Instrumentation and Control Power Battery EVCB and its rack with new battery and battery rack.

Safety Review and USQ Evaluation:

This modification will be performed during normal plant operation. A temporary battery has been installed 1 per temporary modification WR#890305 and will be connected to EVDB bus by Operation procedure during the EVCB battery replacement. The 50.59 for the temporary mod and NSM MG-52176 is referenced for this TNs safety discussion.

TN/0/A/2171/00/AE1 The purpose of this implementation procedure is to control and facilitate the correct installation replacing  ;

the 125 VDC Vital Battery EVCD which was performed under NSM MG-52171. )

Safety Review and USQ Evaluation:

The safety evaluation for this implementation procedure is discussed under modification MG-52171.

McGuire 1 Cycle 8 Monitoring Factor Generation

Description:

The center-line fuel melt (CFM) and 100% FP LOCA and DNB Monitoring Factors used in the power distribution monitoring software for McGuire 1 Cycle 8 were replaced with new monitoring factors.

Monitoring factors are used to define allowable power distribution surveillance limits to ensure that fuel failures will not occur, or are within acceptable limits during postulated transients as the result of DNB, CFM, or a LOCA. Monitoring Factors for McGuire 1 Cycle 8 were developed in accordance with the methodology described in " Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, March 1990." These f actors are input to the core power distribution monitoring software and are used in the calculations performed to satisfy the core power distribution monitoring requirements of Technical Specifications 3/4.2.2 and 3/4.2.3. The purpose of the McGuire 1 Cycle 8 modification was to remove some of the excessive conservatisms used in the generation of the original monitoring factors to reduce unnecessary restrictions in core operation. The generation of these monitoring factors is documented in "McGuire 1 Cycle 8 Monitor Factors, Dec.1992." This modification does not affect any

s 'l other structures, systems or components. ,

Safety Review and USO Evaluation:

Monitoring Factors are considered safety related since they are an integral part of the power distribution monitoring software which is used to determine adjustments, if any, to the axial flux difference limits, the OTDT setpoints, the power range neutron high-flux trip setpoints, and the allowable thermal power. The modification performed to the McGuire 1 Cycle 8 monitoring factors replaced the existing CFM, and 100% ,

FP LOCA and DNB monitoring f actors. The generation of these new monitoring factors was performed i using the approved methodology described in " Nuclear Design Methodology for Core Operating Limits of  ;

Westinghouse Reactors, March 1990." This modification does not increase the probability of accidents previously evaluated, or reduce the margin to safety as defined in the bases of Technical Specifications.

In addition, the software used to perform core power distribution monitoring calcolmions is unaffected by this modification and performs power distribution monitoring consistent with current Technical Specifications. No USO exists.

Unit 1 Cycle 8 Monitoring Factor Generation - EOC Data at 222 SWD The modification performed generated new McGuire 1 Cycle 8 monitoring factors at EOC conditions for input into the core power distribution monitoring software, MONITOR. New monitoring factors were developed to establish center-line fuel melt (CFM), LOCA and DNB surveillance limits based on an ARO parked position for control and shutdown banks of 222 SWD. The original set of monitoring factors for the McGuire 1 Cycle 8 core were developed with the ARO parked position set at 222 SWD for BOC and '

MOC conditions, and 226 SWD for EOC conditions. The BOC and MOC monitoring factors previously developed are unaffected by this modification.

Monitoring factors are used to define allowable power distribution surveillance limits to ensure that DNB, CFM or LOCA induced fuel failure will not occur, or are within acceptable limits during postulated transients. McGuire 1 Cycle 8 monitoring factors were developed in accordance with the methodology described in " Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, March 1990" and are input to the core power distribution monitoring software to assure compliance with the requirements of Technical Specifications 3/4.2.2 and 3/4.2.3.

The two sets of ARO parked positions (222 SWD and 226 SWD) used in the generation of monitoring factors for the cycle 8 core were the result of an effort to reduce control rod wear. The ARO position of the control and shutdown banks are periodically moved to reduce wear at any one position on the rods.

The ARO parted position of 222 SWD represents a case where the rods are inserted slightly into the active region of the core. This condition results in a reactivity penalty if the rods are not repositioned outside the active fuel region prior to cycle shutdown. Therefore, in order to offset this reactivity penalty, both the control and shutdown rods for the cycle 8 core are to be repositioned from 222 SWD to 226 SWD at approximately 30 days before shutdown.

The ARO parked position chosen for the developemnt of the original set of EOC monitoring factors was based on the ARO parked position which yielded the most limiting peaking. This parked position was 226 SWD. However, a consequence of this assumption, in part because monitoring factors are interpolated as a function of burnup, is that certain core locations were unnecessarily penalized. This penalty was the ,

result of a axial power distribution shift produced from the repositioning of the control and shutdown banks l from 222 to 226 SWD.

The generation of new EOC monitoring factors based on an ARO parked poisition of 222 SWD removed the unnecessary peaking penalties produced from the 226 SWD EOC factors. This is because the axial power distribution used for the MOC and EOC factors, and the measured axial power distribution (prior to the control rod repositioning) are now all based on a consistent ARO parked position.

.J*

-O Safety Review and' USQ Evaluation:

Monitoring Factors are considered safety related since they are an integral part of the power distribution monitoring sof tware which is used to determine adjustments to axial flux difference limits, OTDT setpoints and power range neutron high-flux trip setpoints when an out-of-limit condition is determined by the monitoring software. This software is also used to set allowable operating power levels depending on the i type and severity of the out-of-limits condition.

The generation of new EOC McGuire 1 Cycle 8 CFM, and 100% FP LOCA and 100% FP DNB monitoring f actors with control and shutdown rods parked at 222 SWD was performed in order to remove unnecessary peaking penalties being imposed by the EOC monitoring factors generated at 226 SWD.~ The generation of these factors was performed in accordance with the NRC approved methodology described in " Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, March 1990."

The original monitoring factors at BOC and MOC were developed at 222 SWD and the EOC monitoring factors at 226 SWD. The ARO parked position for the McGuire 1 Cycle 8 core is 222 SWD until t approximately 30 days before EOC, when both the control and shutdown banks are repositioned to 226 SWD.

Monitoring factors define allowable peaking factors such that LOCA, transient DNB and center-line fuel melt limits would not be exceeded during postulated condition il transients. The ARO parked position chosen for the development of the EOC monitorir,g factors was chosen to produce the most limiting. .;

peaking (ie. the most restrictive monitoring f actors.) For the McGuire 1 Cycle 8 core, this position was 226 SWD. However, a consequence of this assumption was that the axial power shape assumed in the generation of the monitoring factors was not consbtent with the axialshape of the measured power distribution prior to rod repositioning. While this was a conservative assumption, it resulted in the unnecessary penalization of various regions of the core during the routine Technical Specification surveillance performed as part of Technical Specifications 3/4.2.2 and 3/4.2.3. In order to remove this penalty, new monitoring factors were developed with rods parked at 222 SWD. These factors are only valid for power distribution surveillance performed prior to rod repositioning. The new monitoring factors do not increase the probability of accidents previously evaluated or reduce the margin of safety as defined in the bases of Technical Specifications. They only provide a more accurate assessment of the actual surveillance margin available to LOCA, DNB and CFM surveillance limits relative to the EOC factors previously generated. The software used to perform core power distribution monitoring calculation is unaffected by this modification and performs power distribution monitoring consistent with the current Technical Specifications. Therefore, it can be concluded that no USO exists. .,

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