ML20210R431

Forwards Summary Rept of Mods,Minor Mods,Procedure Changes & Other Misc Changes Per 10CFR0.59

ML20210R431
Person / Time
Site:	Mcguire, McGuire
Issue date:	08/10/1999
From:	Barron H DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9908170107
Download: ML20210R431 (43)

Text

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. Q.- .g-V g Duke Energy Corporation j'Q "

. McGuire Nudear Station 12700 Hagers Ferry Road Huntersville. NC 28078-9340

11. h. Barron. (7N) 8754800 om Vice 1%ident IIM) EISA80N W August 10, 1999 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Subject:

McGuire Nuclear Station , '

Docket Nos.- 50-369, 370 Attached is a summary report per 10 CFR 50.59 (b) (2) of Nuclear Station Modifications, Minor Modifications, procedu"% changes and other miscellaneous changes made at McGuire Nuclear. Station under 10 CFR 50.59 for this reporting period. No Unreviewed Safety Questions were identified.

Any questions regarding this submittal should be directed to Kay Crane, McGuire Regulatory Compliance at (704) 875-4306.

/ML H.-B..Barron, Vice President McGuire Nuclear Station 170.30-TEE'7

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9908170107 990810 i PDR ADOCK 05000369 ,

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,U. S. Nuclear Regulatory Commission Document Control' Desk August-10, 1999 LPage'2-1 cc: ~Mr.;F. Rinaldi,.' Project Manager

. Office of Nuclear Reactor Regulation l Mail :Stop .14H225 Washington,.D.C. =20555.

M r..' ' L . A .' R e y e s , Regional' Administrator.

U. S.-Nuclear Regulatory Commission Atlanta Federal Center-

• 61 Forsyth Street - Suite 23T85

Atlanta, GA 30303 Mr. S. Schaeffer Senior Resident' Inspector McGuire Nuclear Station e'

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,U..S. Nuclear Regulatory Commission Docurhent Control Desk l

August 10,.1999 Page.3 bxc: - EC 0 50-ELL' I

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c' 4 Minor Modifications MGMM-10056 This modification installed a new metal roof enclosure on top of the existing corridor structure running from the Service Building across the Auxiliary Building roof to the Auxiliary Building office area stairwell. The new structure was designed with access doors to allow access to the cable tray on the existing roof. Louvers and vents were installed to allow natural circulation in the enclosed space and HVAC intake vents on the old roof to be extended to the new roof. The current roof access ladder was removed. The current roof parapets were removed to allow the new cover to attach to the existing corridor structural steel.

This modification provides a replacement roof structure over the existing roof and cable tray. The corridor does not perform a safety-related function and is not considered part of the tornado or missile protection function of the Auxiliary Building roof. The new roof enclosure does not impact the operation or function of any systems or components. No new missiles different than those already considered are I made credible. The design maintains accessibility to the cable tray in the area and I

does not degrade the HVAC provided for the corridor. The design was reviewed for '

fire protection concerns and no Appendix R or fire insurance concerns were noted.

The Auxiliary Building roof drainage as considered for flood protection is not impacted. No USQ exists.

MGMM-10092 The ice condenser maintenance crews leave % to % inches ofice on the ice  !

condenser floor to protect the surface of the wear slab and also provide an additional i upor barrier against moisture intrusion mto the foam concrete section of the floor.

This practice was documented on the ice condenser drawings and added to the ,

General Arrangement Drawmg.

The ice condenser is now operating in Unit 2 with some water entrapped in the foam concrete of the ice condenser floor. This modification updated the same drawing to  !

document the water intrusion in that unit. This is only a technical administrative I change to the drawing. The nuclear safety and operability considerations associated with water intrusion in the foam concrete was evaluated.

The following two notes were added to MCM 1201.17-681 Sheet 1: (1) Standard maintenance practice leaves approximately %" to %" ofice on wear slab surface to provide additional vapor barrier and surface protection. (2) Unit 2 Only- water intrusion into the foam concrete discovered July 1997.

The use of mechanical methods for the removal of ice from the ice condenser and the practice ofleaving a layer ofice on the ice condenser floor does not affect the ice condenser or other structures, systems or components or any safety functions. These are merely preferred methods ofice maintenance and involve no USQs.

The addition of a note to show that foam concrete water intrusion exists in Unit 2 on the ice condenser general arrangement drawing is performed as a technical  ;

administrative conection. It does not change the design of the ice condenser floor or i the operation or function of the ice condenser. No USQ exists.

MGMM-10177 This modification removed the Refueling Water System (FW) valve 2FW-9 and replaced the valve with a blind llange. 2FW-9 is a 8 inch drain valve which along with 5 other drains provides refueling cavity drainage during post accident safety system operation. The valve is locked open during operation and is closed during i

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• refueling operations when it is desired to fill the refueling cavity. A 2 inch drain line closed by valve 2FW-75 was added along with two 1 inch flush lines closed by valves 2FW-76 and 2FW-77. The current piping arrangement has been a continuing ALARA problem due to radioactive particle accumulation during refueling operations. The revised arrangement allows the drain to be flushed to reduce the dose problem.

Valve 2FW-9 is one of six drain valves for the refueling cavity. These valves are locked open during operation to provide a drain path to the containment sump from upper containment during post accident conditions. Water sources are containment spray and condensed reactor coolant. The containment analysis assumes these valves close to maximize lower containment pressure during initial blowdown for loss of coolant accident analysis. The longer term containment pressurization model assumes the valves are open to maximize ice condenser bypass flow. Removal of the 2FW-9 drain path makes the long term pressurization model more conservative since less bypass leakage will occur. Once water begins to accumulate in the cavity the drains are covered. The short term pressurization model will remain conservative since the valve was assumed to be closed for this analysis.

The new valves and piping will be qualified to Duke Class B criteria. The piping and valve material is stainless steel which is suitable for both normal and post accident environments. The added valves have no accident mitigation function and will remain normally closed and closed for refueling. The added piping is expected to be used only to flush and drain the 8 inch FW piping which has been a high radiation source due to radioactive particles settling in the drain in the past.

1 The sump is not significantly impacted by the removal of one drain. The analysis of l the water draining form upper to lower containment shows that equilibrium flow conditions will be established well before sump recirculation begins such that the sump is not impacted by this change. The passive failure of one of the remaining valves was considered. Equilibrium flows are still established prior to sump recirculation beginning and the additional water held in the refueling cavity is not i sufficient to impact sump operation. No USQ exists.

MGMM-10411 This modification added a safety related air reservoir for the control valves associated j with the nitrogen supplied operators for each of the feedwater isolation valves '

(CFIV). The assured air reservoir addition will increase the reliability of the CFIVs j in moving to their safety position (closed) at any time required. The control air will {

be provided from safety grade air reservoirs, near each isolation valve, located in the doghouses and supplied from the non-safety Instrument Air (VI) system.

The modification was implemented in two parts. The first part will provided a single assured control air reservoir for each CFIV. The second part ties in a second air tank for each CFIV which only acts to increase the air reservoir capacity.

The feedwater isolation valves are listed as containment isolation valves but do not receive a containment isolation signal. The feedwater isolation signals are provided as required by Table 3.3-3 of the technical specifications. The feedwater isolation valves are closed using a safety related nitrogen supply for the motive energy source.

This modification added safety related control air for operations of the feedwater isolation valve actuators. This control air source is separate from the nitrogen supply for closing valves and provides an assured source of control air to provide proper i operation of each CFIV to isolate on receiving a feedwater isolation signal. The provision of a safety source of air allows the valves to remain open on a loss of Nstrument air until a containment isolation signal is received or manual action is I ty f o close the valves.

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• Assured air supply reservoirs are dedicated to each CFIV and supplied from the VI system. The assured control air is isolated from the VI supply line by a safety related check valve such that a failure of the non-safety VI system does not result in failure of the assured control air for a CFIV. The non-safety VI system is taken from a 2 inch supply line feeding the main steam isolation valves in each doghouse. The piping control valves and isolation valves are leak tested to assure that they will maintain an adequate reserve of control air to ensure proper CFIV operation. A periodic visual monitoring of the local VI pressure and check of the doghouse for air leakage is performed to provide assurance that a local VI failure has not occurred which could allow the CFIV control air reservoir pressure to degrade unknown to plant operations. The added control air reservoirs are located in each doghouse near the associated CFIV and have been evaluated for seismic, pipe rupture and flooding concerns. A single failure of the safety components associated with the control air can only lead to failure of one CFIV. The failure of a CFIV is considered m accident analysis. A postulated control air failure has no impact on the failure analysis or consequence of a CFIV failure as previously considered. The plant equipment response to a loss of VI is unchanged by this modification. Operator actions for a long term loss of non-safety VI will now include manual isolation of the CFIVs if necessary. The response of the feedwater system to plant transients and feedwater isolation signals is the same as previously considered. No USQ exists.

MGMM-10454 This modification removed the Unit 1 Reciprocating Charging Pump (RCP) discharge accumulator pressure gauge (INVPG6500) and turbine from the wall outside the pump room back to the accumulator. The pressure within the Unit 1 RCP discharge accumulator has been difficult to maintain due to leakage from the pressure gauge tube littings. The original design did not include a pressure gauge mounted on the outside wall of the pump room. This modification restored the Unit 1 RCP discharge accumulator to its original design configuration, no pressure gauge,just a fitting to allow the accumulator to be charged when required. This modification will not affect the operation of the reactor coolant pump or the Chemical and Volume Control (NV) system. No USQ exists.

MGMM-10530 The Liquid Waste Recycle and Liquid Waste Monitor and Disposal (WUWM)

Systems Design Basis Document and FSAR Section 11.2.3.1 - UFSAR section 11.2.3.1 " Liquid Waste System Component Design" and Table 3-4 Mechanical System components was revised to delete the waste evaporator, recycle evaporator and waste drain tank as safety class equipment. These changes are consistent with Nuclear Station Modification MG-10892 which downgraded portions of the WU WM, Boron Recycle ( NB), and Solid Radwaste (WS) systems. No USQ exists.

MGMM-8704 This modification provides 2/2 logic for the main turbine trip on loss of autostop oil pressure which is currently provided by pressure switch 2LTPS5551 only. This trip function protects the turbine on loss of autostop oil pressure and is one of several trips which would trip the turbine on loss of turbine hydraulic pressure. The modification utilizes spare contacts on a relay activated by another pressure switch (2LTPS5550) to provide redundancy in sensing the low pressure. These switches are not safety I related but there are other safety related sensing devices and safety related equipment j used to provide the same results as the switches involved in this modification. The i addition of 2/2 logic on this trip circuit will decrease the possibility of single l component failures resulting in a turbine trip. The trip function is still provided from I multiple sources. The pressure switch is powered from a QA-1 source but is not safety related. The work for this modification is in 2ATC15a. All equipment being modified is existing equipment and is qualified for the application. No USQ exists.

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MGMM-9016 These modifications revised the orifice size on the reactor vessel head vents. The MGMM-9017 orifice is sized to assure that the vessel can be adequately vented during accident conditions to ensure natural circulation, but not allow leakage in excess of reactor coolant makeup flow, in the event of a pipe break or failed isolation capability. A train vent path is also used for letdown flow for Safe Shutdown Facility (SSF) operation. The orifice size is modified to assure that adequate letdown flow exists while continuing to meet oth :r design criteria.

The orifice sizing calculation has been revised to ensure that all functions of the vent system are met with adequate design margin. The vent orifice is sized to provide adequate vent flow, adequate flow for SSF events and limited flow for isolation failure evaluation. The revised orifice size ensures that adequate letdown flow will be available using this flow part in SSF events and that required venting capability and accident flow restriction capability is still met. The vent orifice was modified consistent with the revised calculations. No USQ exists.

MGMM-9056 This modification added instrument loops INSLP5060 and INSLP5070 to the list of instruments used for Post Accident Monitoring (PAM) as described in Section I of the UFSAR. The existing loops used for PAM indication (NS W/R instruments INSP5530 & 5540) have a range of-5 to 60 PSIG. It is useful to have narrow range PAM instrumentation available to more accurately monitor containment at lower pressures (3 PSIG). Existing control room indication will be utilized. The upgrade for loop 1NSLP5060 will required installation of a isolator card in Process Control System Cabinet 2 to separate a Operator Aid Computer (OAC) input from the remaining portions of the loop. A fault at the OAC could cause failure of the loop l receiver gauge or recorder. The control portion of the loop is not affected by this modification because the control and indication portions of the circuits are already separated by an isolator. There were no changes in the automatic controls for the Containment Spray System (NS). No USQ exists.

MGMM-9549 This change revised drawings for filter housing CUNO model CN-ID, manufacturer drawing number $1088 to ensure that each housing has a stand alone drawing and that each housing installed in the plant has a drawing, with plant application annotated. The following drawings were deleted: MCMM-1201.04-0025-002 through 007. CUNO Filter Assembly drawings MCMM-1201 Al-0022-005 and 007.

No USQ exists.

MGMM 9640 This modification added relief valves to the Safety Injection System (NI) test header and containment penetration piping. This piping is normally isolated from the safety i injection process piping and the Auxiliary Building piping and is subject to overpressurization due to thermal expansion of the contained fluids. The piping associated with these modifications is Class E since it is only used for testing and does not perform a safety function. Relief valve discharge is to containment and is limited by the valve to 20 gpm.

The NI system provides high pressure injection from safety injection pumps and accumulator tanks in the event of loss of reactor coolant system pressure. The test header is provided to allow testing of NI boundary valves between the reactor coolant system and the NI system for leakage. The test header provides no accident mitigation function or process function outside of testing and is not safety related.

The addition of relief valves to the test headers for Unit I and Unit 2 assure that this piping will not be overpressurized by thermal expansion of the fluid left in the line when the piping is isolated. The relief valves associated with the Emergency Core Cooling System (ECCS) are shown in UFSAR Table 6-132. The added valves are

similar to the valves shoe in the table for the nitrogen supply to the cold leg accumulators. These valves are also non-safety and normally isolated from the NI l process fluids. The valves and added piping are qualified for the environment and are supported as necessary to prevent the potential for damage to safety related ,

equipment. Failure of the valves may impair testing procedures but does not result in I failure of any safety function. Valve orientation and discharge from the valves is directed such that no safety related equipment function is impacted. No USQ exists.

MGMM-9829 This corrective modification abandoned in place several non-safety related, non QA Condition 1 Solid Radwaste System (WS) equipment and annunciators. In order to maintain the Blackboard philosophy for alarm conditions, it was necessary to issue this modification to delete several WS system annunciators. The ECST is not l

required to be used for hold-up of chemical concentrates for solidification; therefore, l the ECST heaters (which are described in the UFS AR), heater controls, associated l temperature instrumentation, and the high and low temperature alarms will be l j abandoned in place with power removed to the ECST heaters. This modification does j not affect the safety of the operation of the plant and the change to the SAR in minor. '

The change to the SAR will only document the abandonment of heaters in the ECST which are no longer being used per the original design of the component. The ECST is non-safety related and does not affect a safety related component. No USQ exists.

MGMM-9831 This modification abandoned in place and non-safety related, non QA Condition 1, electrical portion of the Waste Gas (WG) System flowmeters OWGFr5020 and OWGFr5030, and deleted their associated annunciator alarms. These flowmeters were designed to measure the purge flowrate of gas from each unit of Volume l Control Tanks (VCTs). Since initial operation of the WG system these flowmeters have proved to be unable to function within the process conditions of the purge header. In order to maintain the Blackboard philosophy for alarm conditions, it was necessary to issue this modification to delete the annunciators associated with these WG System flowmeters. This modification does not affect the safety of the operation of the plant and the change to the SAR is minor. The change to the SAR will only document the abandonment of the WG system VCT purge flowmeters, OWGFr5020 and OWGFr5030, which are no longer being used per the original design of the WG system. The electrical portion of these flowmeters, which is the only part being j abandoned, are non-safety related and do not affect any safety related components. 1 The piping pressure boundary for these turbine flow meters will remain. No USQ l exists.

MGTM-0057 This temporary modification was intended to provide operator assistance in controlling the water levels in the steam generators during the required modifications to the feedwater isolation valves. The vendor manual for the replacement steam >

generators specifies that the water level either remain above the top of the

" gooseneck" portion of the in'et nozzle when feedwater has been isolated or a slow ,

purge of the potential steam voids shall be conducted. l This intent of this modification is to improve the reliability of McGuire Unit I during the continued modification to the feedwater isolation valves by providing the capability of automatic level control of the steam generator water levels. By providing this control, the potential for forming steam voids in the gooseneck portion of the main feedwater nozzles into the steam generators has been eliminated This temporary modification sets a target level of 45% narrow range to satisfy this requirement and will provide the controlling level while power is limited to less than 25% reactor power. No USQ exits.

MGTM-Ol l4 This modification changed the controls for heater drain 2HWil5 from direct acting to

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reverse acting and documents that 2HW115 is a fail-open valve. Changing this controller to a reverse acting controller will make the system operate normally. If instrument air is lost to 2HWI15, water level in the 2F1 feedwater heater will increase. This will result in opening the condenser dump valve. If necessary, 2HW115 can be manually isolated and the feedwater heater string can be removed from service. This will only effect the efficiency of the secondary side plant operation. No USQ exists.

MGTM-0116 This temporary modification allows functional use of the Reactor Makeup Water Storage Tank (RMWST) pump 2A during the work required to change out the local pump control terminal box for RMWST Pumps 2A and 2B. The original designed function of the RMWST pumps will still be available in the manual mode. In manual mode, the operator will have direct control or makeup to the reactor coolant system to support normal operation of the unit. In an accident scenario, this system is automatically isolated from the safety systems. The RMWST makeup system is only required for normal operation of the reactor coolant system and normal makeup provisions can be in auto or manual.' In order to continue normal makeup to the RCS during the corrective maintenance on the local pump control box, it was necessary to prepare this temporary modification to allow use of the RMWST pump 2A in manual.

This does not change the provisions as specified in the SAR. This temporary modification does not affect the safety of the operation of the plant and no changes to the SAR are required. The RMWST pumps are non-safety related and do not affect any safety related components. No USQ exists.

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PROCEDURES IP/2/A/3007/17 'IT/2/A/9700/168 is a temporary test procedure. The test method employed is based

'IT/2/A/9700/168 on the a' hernative test procedure described in ASME PTC6-1996," Performance Test Code 6 on Steam Turbines". Tests that may be performed per this procedure are:

Unit at 100% power and in a normal alignment, Unit at 100% power and turbine cycle partially isolated, Unit at 100% power, turbine cycle partially isolated and turbine control valves wide open, and Unit at 100% power and in a normal alignment while varying TAVG. This procedure provides the necessary instructions and valve alignments for conducting the various tests to be performed. In general, the test method involves isolation of systems, equipment and flows associated with the turbine cycle. The procedure also provides instruction for the installation and removal of test instrumentation. Instructions for restoring systems and equipment following completion of the testing is also provided. In support of test procedure TT/2/A/9700/168, a Restricted Change was made to IP/2/A/3007/017 to add another criteria for when the power range detectors are to be adjusted.

Although this test procedure requires off-normal plant configurations during power operation, it does not result in configurations and situations that will compromise plant safety. Safety related and accident mitigation structures, systems or components are not impacted by this procedure. All throughout the test, the auxiliary feedwater system will be fully operable and capable of performing its safety related function. The Engineered Safeguards and Reactor Trip systems and components are not be affected by this test procedure. Actions taken in accordance with this procedure to isolate the turbine cycle flows and system do not involve the steam generator power operated relief valves (PORVs), the main steam safety valves (MSSVs) the main steam isolation valves (MSIVs) , or the atmospheric dump valves.

These valves remain operable and capable of performing their intended function. In the event of a sudden loss ofload resulting in a turbine trip and a reactor trip, the excess energy associated with this event will be dissipated by the steam generator PORVs, the atmospheric dump valves and the main steam safety valves. The main steam safety valves are safety related and have 100 percent relieving capacity. No USQ exists.

MP/0/AU700/102 This procedure puts into place compensatory measures needed to maintain the unit 1/2 ETA and unit 1/2 ETB Switchgear Rooms OPERABLE during modes 1 - 6 and No Mode. Doors along column line AA between the turbine building and switchgear rooms, are periodically blocked open to allow transport of materials and equipment into the auxiliary building. These doors act as a ventilation boundary for the switchgear room subsystems of the Control Area Ventilation System (VC).

Implementation of these compensatory measures will ensure that components within the switchgear room are not exposed to temperatures greater than their design qualification temperature. Overall performance of the components within the switchgear room will not be affected. Implementation of these compensatory measures does not change any actions described in the safety analysis report and will not alter any assumptions made in evaluating the radiological consequences of postulated accidents. This activity creates no new failure modes or operating characteristics for equipment or systems important to safety or plant operation. No USQ exists.

OP/0/A/6450/003 This procedure directs Operations personnel on how to operate the Auxiliary Building l Ventilation (VA) system properly during normal plant operation. The procedure was revised to include compensatory measures to ensure the opposite Unit's VA system filtered exhaust unit remains " Operable but Degraded' when the breaker for the l

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, affected Units filtered exhaust fan (FXF)is open for maintenance purposes. The i revision adds a compensatory action to place the opposite VA filtered exhaust unit in the filter mode and ensure both the A and B train filtered exhaust fans are running. l This will provide Operations personnel the option to either enter the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> action for both units, or invoke the compensatory measure to maintain the opposite units VA System fully operable. No USQ exists. l OP/0/A/6550/11 Due to the degradation of the Boraflex panels, a new set of burnup versus enrichment limits for Region I were established to ensure Ken remains less than or equal to 0.95. (

These new limits are more restrictive than the current limits specified in Tables 3.7.15-1 and 3.7.15-2 of the McGuire Nuclear Station Technical Specifications.

These new limits were implemented by changes to Procedures OP/0/A/6550/11.

Internal Transfer of Fuel Assemblies and PT/0/A/4150/37, Total Core Unloading.

The function of the spent fuel storage racks is to provide for safe storage of spent fuel assemblies in a flooded pool, while preventing criticality. To accomplish this safety 1 function, the Ken in the spent fuel pool, including all uncertainties, is to be less than or equal to 0.95, even if unborated water is used to fill the pool. The new limits ensure that K,n remains less than or equal to 0.95 with the anticipated degradation of the Boraflex panels and the spent fuel pool filled with unborated water. The I compensatory actions and procedure changes affect no design criteria or safety functions of any structure, system or component. No USQ exists.

OP/0/B/6400/OlM A restricted procedure change was added to the Recirculated Cooling Water j System (KR) operating procedure. The temporary change added a new enclosure to l

provide guidance on securing the "C" KR pump from service. The specific action is f the partial closure of the manual discharge isolation valve for the "C" KR pump, which will limit the amount of reverse flow that can occur when the pump is secured,in the event that the respective pump discharge check valve does not close promptly. The structures, systems or components (SSCs) affected or potentially affected by this activity are the "C" KR pump,1KR-6 (KR Pump C discharge check valve), and 1 KR-9 (recirculated cooling water pump C discharge isolation). These SSCs do not perform a safety function and no design basis events or accidents apply.

No unwanted system interaction will result from this activity. Based on the KR system design and the design of the potentially effected SCC's, the effects of this activity were found acceptable. No USQ exists.

OP/0/B/6400/004 A restricted procedure change was made to the Recirculated Cooling Water System (KR) Operating Procedure. This change lowers the KR system operating temperature at which additional KR heat exchangers are placed in-service, from 95 degrees F, to 90 degrees F. It revises the normal position for the 2A feedwater pump turbine (FDWM) oil cooler control valve bypass (2KR-158) from " closed" to " throttled" The specific change being incorporated is a change in the setpoint for placing additional KR heat exchangers in-service and throttling open the 2A FDWN oil cooler control valve bypass, which will provide additional cooling to the FDWN oil cooler, in response to higher than normal oil temperature. KR flow to the in-service oil cooler will be limited to approximately 200 GPM, to avoid tube damage due to erosion.

The structures, systems or components (SSCs) affected or potentially affected by this activity are the "2A" FDWN and its oil cooler control valve bypass,2KR-158. No design basis events or accidents apply. No unwanted system interaction will result from this activity. Based on the KR system design and the design of the potentially effected SCC's, the effects of this activity were found acceptable. No USQ exists.

OP/1/A/6100/003 This procedure was changed to support implementation of temporary modification

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, MGTM-0057. This temporary modification provided a control point for the steam generator level control system during the modification of the feedwater isolation valve actuators. The impact to the steam generator operation is that the inventory of water will be slightly larger during this limited time period. The base design is for the narrow range level to control from 39% at 0% reactor power up to 45% at 25%

reactor power. This change will set the control point of the water level to be a constant 45% narrow range.

The need for this temporary change is to prevent the potential formation of steam voids in the highest section of feedwater piping in the steam generators, which is the

" gooseneck" while the flow is isolated through the main nozzle. The setting of 45%

narrow range level to control the feedwater to the steam generators provides assurance that this section of piping will remain completely covered during the duration of this work. Operating for a limited duration at this slightly elevated water inventory does not compromise the design basis accident analysis below 25% reactor i

power. The inventory change does not change any probability of malfunctions as the control system adjustment is performed on non-safety circuits. Safety related equipment is not affected. No USQ exists.

OP/1/A/6100/010C These procedure changes revised the operator response to the annunciators for control OP/2/A/6100/010C rod bank lo limit and control rod bank to-lo-limit. These annunciator responses refer the operator to AP/l(2)/A/5500/013, Boron Dilution. This procedure would take the operator through the steps to begin boration to move the rods back above the insertion limits. The previous technical specifications required boration on loss of shutdown margin. The annunciator response assumed that the lo-lo limit alarm was received and that shutdown margin was lost. The Improved Technical Specifications shutdown margin specification (3.1.1) applies in mode 5,4,3, and 2 when Keff <l.0.

Improved Technical Specification 3.1.6 requires that when the rod insertion limits are exceeded then a shutdown margin calculation must be performed in one hour and rods returned above the insertion limit with two hours. These procedure changes update the annunciator response procedures to make them consistent with the Technical Specifications and the Technical Specification Bases. No USQ exists.

OP/1/A/6200/04A This procedure aligns a) Residual Heat Removal (ND) normal refueling water storage tank (FWST) suction to the Containment Spray (NS) pump and back to the (FWST) and b) the ND system to the FWST, through the NS system, for recirculation of water to remove radioactive contaminants. The water transferred to the FWST is cleaned by aligning the FWST to purification after the test. The alignment used in this procedure will result in one train of ND and the same train of NS being inoperable for a short period of time. The ND train will be inoperable due to the suction isolation valve being closed. The NS train will be inoperable during the procedure due to the manual alignment of NS to the FWST, and the closure of the normal FWST suction valve. NS pump operation (high velocity flush) during alignment to ND suction will have no adverse effects on the remaining operable ND train. During the ND heat exchanger flush, the ND and NS pumps will be racked to disconnect to prevent pump damage in the event of an auto start signal during the procedure. A deJicated operator will be available to manually close the NS suction isolation valve from the (Emergency Core Cooling System) ECCS sump (INS-1B or 1NS-18A) in the event of a safety injection signal and failure to remote valve control, to prevent the transfer of sump water to the FWST. No USQ exists. 1 OP/1/A/6800/011 The purpose of this procedure is to provide a controlled drain path through the i l

OP/2/A/6800/011 Groundwater Drainage (WZ) System in order for the Nuclear Service Water (RN) system to have outage maintenance performed. This activity is required to ensure the continued availability of this vital system during plant operations during the next operating cycle. The flow of RN into the WZ system is controlled by procedure steps

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, such that only one sump pump will be required to operate to move this flow out of the auxiliary building. Use of the WZ System for this purpose does not present any challenges to any safety equipment that is needed to protect the health and safety of the public by having an adverse affect on dose at the site boundary. Based on the limitations that are placed within the procedure in relationship to Auxiliary Building Flooding Accidents, no USQ exists.

OP/2/A/6100/0101 This modification raised the Hi alarm setpoint for 2W7.LS5%0 from 731' to 733'.

This in tu n raises the Hi-Hi setpoint from 741' to 743'. 2WZLS5060 monitors the groundwater level at the Unit 2 Reactor Building wall. The Reactor Buildings have been eva'uated for hydrostatic loads up tc grade (elevation 760'). Therefore, there is no concern with the loading on the Unit 2 Reactor Building wall. This well was removed form technical specifications but maintained in the Selected Licensee Commitments (SLC) in order to provide an alert to such events as a pipe leak / break or failure of the under drain grid. Pipes in the area that are not under pressure would be the Nuclear Service Water (RN) and Condenser Cooling Water (CCW) lines.

Postulated failure of the CCW pipe contributes 38 gpm to the grid flow and the RN failure contributes 666 gpm to the grid. Review the WZ sump pump starts indicates less than 30 gpm being contributed to the sumps from the under drain grid. Breaks form pressurized piping would manifest themselves by percolating up through the soil. This has not been observed. The well will still be capable of providing the SLC function with the new setpoints.

The current increase in groundwater level at this location has been noted and studied.

Engineering has determined that the increase is due to construction in the area and increase in operating level of Lake Norman. No USQ exis'.s. j OP/2/A/6100/SD-4 TO/2/A/9600/105 provides the method for bypassing the P/12 interlock and the method for use of additional condenser steam dump valves for unit cooldown while in procedure OP/2/A/6100/SD-4. The P/12 interlock is bypassed in the 7300 cabinets to disable the interlock when appropriate pressure and temperature conditions are met during unit cooldown in mode 4. The condenser steam dumps are controlled using the Steam Pressure Controller before and after the P-12 interlock is bypassed. This procedure reduces the amount of time the Residual Heat Removal (ND) system is needed to operate during unit cooldown and allows an acceptable cooldown rate using the condenser dump valves at lower temperatures.

The P-12 interlock provides automatic blocking of six of the nine condenser block valves when Reactor Coolant Temperature t-ave is below 553 *F and arms the steam dump system on high t-ave when the unit is in modes 1.2, or 3. This interlock prevents large overcooling transients when load rejection conditions occur. Technical .

Specification 3.3.2 requires that the interlock be operable during modes 1,2, and 3.

This interlock may be bypassed when the unit is in mode 4 since it is no longer required by technical specifications. There is insufficient energy in the secondary system at this point to cause an overcooling accident. The effectiveness of the three remaining condenser dump valves for unit cooldown decreases as Reactor Coolant j System temperature and pressure decreases. 'Ihis procedure provides for bypassing i the P-12 interlock during cooldown below 300 F. This makes use of the ND system for cooldown in this temperature range less necessary and should reduce dose rates from the ND system during plant outages. The ND system will be operable and ready i to respond to any failure of the steam dumps to operate as desired. No USQ exists.

OP/2/A/6800/012 This procedure is being performed to align and drain the 2B Nuclear Service Water (RN) System in preparation for normal outage maintenance to be conducted on specific equipment. The flow of RN water into the Groundwater Drainage (WZ)

System is controlled by procedure steps, such that only one sump pump will be

, required to operate to move this flow out of the Auxiliary Building. A prerequisite step requires a minimum of 5 WZ sump pumps to be operable during the draining evolution.

There are no USQ concerns based on the limitations that are placed within the procedure in relationship to Auxiliary Building Flooding Accidents. This conclusion was reached from an evaluation of the requirements for this outage maintenance activity and the continued operability of the Groundwater Drainage Sump System.

No USQ exists.

FT/0/N4150/009 As part of an inspection performed during a cycle initial criticality, an NRC inspection raised a concern over over-diluting the volume control tank as a result of diluting to the suction of the charging pump. Water would return to the volume control tank (VDC) via the mini-flow line. If diluting to critical, control rod insertion limits could be violated since dilution will continue to occur after the reactor make-up  ;

pumps are stopped. This issue was expanded to include dilution of the VCr during  !

the initial dilution procedure.

This procedure was revised to limit the dilution to 200 ppmb unless the mini-flow line flow path was changed to isolate the Residual Heat Removal (NV) pump mini-flow recirculation from the volume control tank (VCT) and return the flow to the ,

suction of the NV pump. This prevented the VCT from becoming over diluted. At the time of this procedure change,it was identified with the system in this configuration, an accident which resulted in the isolation of non-essential Component Cooling (KC) water to the Reactor Coolant (NC) pump seal water return heat exchanger would lead to overheating of the NV pumps. The procedure included contingency plans to return normal alignment within 45 minutes of receiving a safety injection signal.

Since the reissue of PT/0/N4150/009 where this alternate alignment was implemented, an additional concern has been raised as a result of operating experiences detailed SOER 97-01. When operating in the alternate alignment, gases may be stripped from solution by the NV recalculation orifices which would then be directed to the NV charging pump suction. NV charging pumps could be damaged unless the suction line was periodically vented.

As a result of these two significant issues, the concern of over-diluting the volume control tank during the initial dilution to the estimated critical boron was re-addressed. No USQ exists.

FT/0/A/4150/37 Due to the degradation of the Boraflex panels, a new set of burnup versus enrichment limits for Region 1 are established to ensure K,nremains less than or equal to 0.95.

These new limits are more restrictive than the current limits specified in Tables 3.7.15-1 and 3.7.15-2 of the McGuire Nuclear Station Technical Specifications.

These new limits are to be implemented by changes to Procedures OP/0/N6550/11 Internal Transfer of Fuel Assemblies and FT/0/N4150/37, Total Core Unloading The function of the spent fuel storage racks is to provide for safe storage of spent fuel assemblies in a flooded pool, while preventing criticality. To accomplish this safety

function, the K,n in the SFP, including all uncertainties, is to be less than or equal to l 0.95, even if unborated water is used to fill the pool. The new limits will continue to ensure that K,n remains less than or equal to 0.95 with the anticipated degradation of the Boraflex panels and the SFP filled with unborated water. The compensatory actions and procedure changes affect no design criteria or safety functions of any structure, system or component (SSC). There are no USQs involved with the new burnup versus enrichment limits.

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PT/0/A/4400/018 Selected Licensee Commitment (SLC) 16.9-1, Testing Requirements states that the ,

Fire Protection (RF) pumps are to be started at least once per 31 days and operated l for at least 15 minutes. There is no condition on how the pumps are to be started, i.e.,

manually or by automatic means. The proposed chr.nge will revise the UFSAR wording to be in-line with the SLC wording.

The change does not affect the Technical Specification, Selected Licensee Commitments (SLC), operation-related information , design bases, operation or functions of any structure, system or component (SSC). The change is only for noting that starting of the fire pumps for monthly testing by automatic means is not required.

The specific change is to the second sentence of the third paragraph of Section 9.5.1.4 and is proposed as follows: "In addition, each min fire pump is started monthly and operated for at least 15 minutes to demonstrate it is operable. No USQ exists.

PT/1/A/4206/15A These procedures were r: vised to enable differential pressure testing of Safety PT/1/B/4206/15B Injection System (NI) valves INI-121 A and INI-152B (Train A & B NI to Hot Legs) i during iEOCl2. This simply involves putting 1 A and iB NI Pump in mini-flow  !

recirculation to the FWST, measuring pressure across INI-121 A, and INI-152B then opening INI-121 A and INI-152B against this dp. When INI-121 A and INI-152B is open, flow is directed to Hot Legs. Pressurizer level is monitored during this test.

The period of time that INI-121 A iand INI-152B is open is minimized by procedure (it is expected to be very brief). l Three are no system alignments that are not already bounded by existing alignments in this procedure (FWST flow is already directed to the Reactor Coolant (NC) system in the full stroke of INI-101 (FWST to NI Check), and flow is already directed through the hot legs in the A and B Train Hot Leg Check Valve Stroke Verification). 1 This restricted change only sets up conditions so that INI-121 A and INI-152B can be opened against a differential pressure greater than 1090 psid (the maximum expected dp per MCC-1223.12-00-0017, Maximum Expected delta Ps of NI EMO Valves) and then closed. This stroke will occur three times in the procedure and VOTES data will be collected. Pressure upstream and downstream of INI-121 A and 1N1-152B will be recorded for each stroke.

This test verifies adequate valve operation in a Mode (Mode 5,6, or No-Mode) where typical emergency core cooling system accident alignments are not required.  !

Sufficient procedural control is provided to maintain acceptable reactor coolant j system cooldown limits and reactivity changes, and to further minimize potential variations in reactor c ' ,lant system water inventory. No accidents evaluated in the SAR will be more probable as a result of this change to the procedure. Initial pressurizer level is at 20% and the three periods when INI-121 A and INI-152B is open will be very brief, so that inventory addition to NC will be insignificant. NC system temperature is already controlled per the procedure.

This procedure revision does not alter any required system alignments or operational limits. This procedure continues to verify adequate performance of the Safety Injection System. The probability of an accident or malfunction is not increased, and the margin of safety as defined in Tech Specs is not reduced. No USQ exists.

l PT/1/A/4250/0NK This procedure was issued as a new procedure to test the mechanical overspeed trip PT/2/A/4250/0NK function of the Main Feedwater (CF) pump turbines. The test is performed in modes l

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4,5, or 6, when the CF system is not in service and when the Residual Heat Removal (ND) system provides core cooling. The turbine and feedwater pump are uncoupled, i and auxiliary steam is supplied to the CF pump turbine to accelerate the turbine to the l overspeed trip setpoint. Supervisory instruments are monitored during acceleration to ensure that bearing vibration and turbine speed do not exceed given limits. The CF pump turbines are not nuclear safety related components, and no safety related structures, systems or components (SSCs) are affected by this test while the turbine is uncoupled from the CF pump and while the Auxiliary Feedwater (CA) autostart signalis defeated. No USQ exists.

Irr/2/A/4206/10 The Residual Heat Removal injection Test (Acoustic / Magnetic Check Valve Test) procedure was changed to include: 1) use of magnetic data to " optional" for any valve as desired,2) clarifying what test equipment is to be used and associated steps for installation /use,3) modifying acceptance criteria to require analysis of the overall acoustic event, not just the " time of arrival technique", to document a full stroke occurrence,4) incorporating clarifications, format and editorial enhancements,5) adding acceptance criteria for full stroke of 2ND-8 and 23, and 6) changing section 12.2 to accommodate partial retests.

1 This test is performed in Modes 5,6, or No-Mode when emergency core cooling )

system accident alignments are not required. Sufficient procedural control is provided to maintain the Reactor Coolant (NC) system cooldown limits. The scope of this revision does not change test alignments. These changes are only 4 enhancements to enable valve retests, improve data evaluation to ensure the tested check valves opened, and clarify the role of magnetic testing. Injection flows to verify 2ND-8 and 23 are very conservative with respect to the Technical Specifications. The scope of the procecfure revision does not alter the mode of performance nor system alignments, so the possibility of an accident outside the events currently analyzed by the SAR is not created. The revised procedure continues to verify adequate performance of the Emergency Core Cooling System (ECCS) check valves, thereby assuring the margin of safety as defined by the Technical Specifications. The Technical Specifications state that the operability of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a loss of coolant accident (LOCA), and provides ling-term core cooling capability in the recirculation mode during the accident recovery period. This revision better ensures that emergency core cooling capability is present, by providing a better method to ensure these ECCS checks fully open. The margin of safety is not reduced by any of these enhancements.

This procedure revision does not increase the probability of an accident or malfunction, increase the consequences of an accident or malfunction, or create the possibility of any new type accidents or malfunctions. This revision only has enhancements to the existing procedure and does not introduce any new lineups.

Conservative acceptance criteria are applied to verify opening of 2ND-8 and 23. The margin of safety as defined in Technical Specifications is not reduced. No USQ exists.

Irr/2/A/4206/15A These procedure revisions to the 2A and 2B Safety Injection System (NI) Pump Head PT/2/A/2406/15B Curve Procedure included the following: 1) changes to show the Improved Technical Specifications (ITS) nomenclature (there is no change to the actual test or acceptance criteria from these changes), 2) changes providing the option of using process flow indication (flow indication for this test does not have to be instrumented), 3) pump vibrations rebaseline (there was no affect on acceptance criteria), 4) specifying that only the miniflow point must be plotted against the pump head curve (the only required hydraulic data is for the test (minillow) point: the full flow vibrations are needed for the Supplemental Test Program. but flow and dp are not needed). and 5)

a 'k 1

, providing for stroking the Hot Leg checks in loop pairs (B&C or A&D) rather than one loop at a time (there is no design or accident analysis required flow rate through individual Hot Legs.

This procedure revision does not increase the probability of an accident or malfunction, increase the consequences of an accident or malfunction, or create the possibility of any new type accidents of malfunctions. The margin of safety as defined in Technical Specifications is not reduced. No USQ exists.

RP/0/A/5700/025 This procedure was developed in respense a drill conducted on 12/11/96. This need for procedural guidance for Fire Brigade response when dealing with offsite agencies, use and operation of Fire Birgade radios, coordination of Fire Brigade and offsite agencies, and various other enhancements for our Fire Brigade responders was identified.

This procedure contains the details for the plant expectations for the fire Brigade, and will ensure consistent response during drills or actual events. No USQ exists.

TO/0/A/9600/102 The procedure was written to provide guidance on aligning 250 Auxiliary Power System battery IDP for IAE testing and provide contingencies for Loss of Offsite l Pour (LOOP) events on either unit. The procedure provides steps for removing l non-vital lighting loads for a period long enough to test and recharge the degraded battery. The systems and equipment affected by these activities are not associated with nuclear safety. There is no adverse impact to the nuclear safety of structures, systems or components (SSCs) due to this Temporary Operating procedure. No USQ exists.

TO/2/A/9600/105 This procedure change provides the method for bypass of the P-12 interlock and provides a method to use additional condenser steam dump valves for unit cooldown while in procedure OP/2/A/6100/SD-4. The P-12 interlock is bypassed in the 7300 cabinets to disable the interlock when appropiiate pressure and temperature conditions are met during unit cooldown in mode 4. The condenser steam dumps are controlled using the Steam Pressure Controller before and after the P-12 interlock is bypassed. This procedure reduces the amount of time the Residual Heat Removal (ND) system is needed to operate during unit cooldown and allows an acceptable l

cooldown rate using the condenser dump valves at lower temperatures.

The P-12 :nterlock provides automatic blockitig of six of the nine condenser block valves when reactor coolant temperature t-ave is beluw 553 T and arms the steam dump system on high t-ave when the unit is in modes 1,2, or 3. This interlock prevents large overcooling transients when load rejection conditions occur. Technical Specification 3.3.2 requires that the interlock be operable during modes 1,2, and 3.

'Ihis interlock may be bypassed when the unit is in mode 4 since it is no longer I required by technical specifications. There is insufficient energy in the secondary system at this point to cause an overcooling accident. The d 'ectiveness of the three remaining condenser dump valves for unit cooldown decreaes as Reactor Coolant System temperature and pressure decrease. This procedure provides for bypassing the P-12 interlock during cooldown below 300T. This makes use of the ND system for cooldown in this temperature range less necessary and should reduce dose rates from the ND system during plant outages. The ND system will be operable and ready to respond to any failure of the steam dumps to operate as desired. No USQ exists.

7T/0/A/2493/00/IM Modification MG-52493 replaced the current air operated Nuclear Service Water

'lT/0/A/2493/00/2M (RN) flow control valves (IRN442 & 1RN445 for A train and IRN457 & 1RN460 )

for B train) with a single control valve for each train (AS&B). These valves will be l

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, actuated directly from the refrigerant via a sensing line that will be installed from the chiller condenser to the new valves. Valve IRN445 will maintain the Control Area chiller A, CRA-C-1 head pressure from between approximately 118 psig and 140 l psig. Valve 1RN460 will maintain the Control Area Chiller B, CRA-C/2 head pressure to the same pressure range.

These temporary test procedures will be used to verify the operation of the water regulating valve 1RN445and IRN460. The test includes leak testing of the existing i condenser isolation valve. The test will verify that the RN flow balance is not adversely impacted by the new valve installation. The flow versus valve pressure characteristics and stable operation of the valve will be verified.

These procedures align both the Unit I and Unit 2 RN systems per existing approved procedures. The procedures align RN 1 A and 1B train to all the Unit I essential loads and RN 2A and 2B train to all the Unit 2 essential loads. This ensures that while the RN 1 A and 2B train is being manipulated during testing, the RN system can meet its normal plant design basis cooling requirements. Valve IRN40A is closed to isolate train 1 A from the non essential header. Valve 2RN40A is closed to isolate train 2A from the non essential header. The remaining valve alignments provide Unit I flow balance conditions comparable to that required for a safe station shutdown normally or as the result of a postulated Loss of Coolant Accident (LOCA) ,

concurrent with the controlled shutdown on the opposite unit. These alignments do l not adversely affect the ability of the RN system to perform it s intended assured i cooling function. The testing does not prevent the manipulated valves from automatically repositioning to their safe position following a safety signal, as I previously described, and the manual valves from being repositioned by operations under emergency plant procedures, if during testing the refrigerant sensing line should leak when filled with nitrogen, valve 1RN445 and 1 RN460 will close. This would not impact the ability of the RN system to meet its other essential load flow requirements. If the refrigerant sensing line should leak whe:. filled with refrigerant, the sensing line would be isolated, evacuated, fixed and placed back into service. Chiller CRA-C-1 is not operable if the sensing line leaks. No USQ exists.

• lT/0/A/9100/90 This procedure "RF Jockey Pump, Pressurizer Tank and Associated Equipment Test" was run to verify equipment performance and to determine cause of pressurizer tank not filling when ajockey pump is running. No unusual or experimental conditions are involved.

The Fire Protection (RF) System will be aligned (normal alignment) in attempt to flow water from the jockey pump suction to the Pressurizer Tank. 'the only exception to the normal alignment is that the Station Air System (VS) header will be isolated to the RF pressurizer tank. Both jockey pumps will be used during the test.

Pump suction and discharge pressures and changes to tank level and pressure will be monitored throughout the test. This part of the test will verify pump performance.

The second part of the test is to check operation of the 3-way valve (lRF-7) which directs water either to the system or puts it in recirculation. This will utilize a bypass I line around 1RF-7 to open a flow path to the tank form the pumps. I i

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Other parts of the test check to ensure valves in the flow path are open and verify j operation of the instruments which monitor the tank level. No USQ exists.

l l TT/1/A/9100/0527 The purpose of this test is to determine the Unit ) gent fuel pool heat up rate with

o 0 l

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, Component Cooling (KC) to the Spent Fuel Pool Cooling (KF) heat exchanger secured.

I KC Flow to the operating KF Heat exchanger will be secured. Spent fuel pool l temperature will be recorded and the cooling will be restored to the KF heat i exchanger when the temperature reaches 100 degrees F or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the start of  !

the test. Form this information, a heatup rate for the spent fuel pool will be calculated.

No USQ exists.

'IT/2/A/9100/115 NSM MG-22515/00 replaced the discharge check valves for each unit 2 Auxiliary Tf/2/A/9100/1ISA Feedwater System (CA) pump with an automatic recirculation control (ARC) valve.

Tf/2/A/9100/Il5B These valves provide an assured minimum flow path for the CA pumps. The ARC )

valve operates as a combined check valve and globe valve with a globe valve port providing an assured flow path to prevent deadhead conditions on the pump. The check valve port is the main discharge flow path and is operated in the same manner ,

as the previous discharge check valve. The ARC valve is designed such that the l globe port closes as the check port opens assuring a continuous discharge path for the l pump and automatically eliminating recirculation flow as sufficient normal discharge  !

flow is established. The installation of this recirculation flow path will result in j removal from service of the current recirculation flow paths for the CA pumps. This  :

piping will now be used only for pump flow testing. The current recirculation valves (2CA20B,2CA27A, and 2CA32AB will have automatic controls removed and will only be operated manually. The instruments removed are: 2CAMV0200, 2CM AV0270, 2CAMV0320, 2CASV0200, 2CASV0201, 2CASV0202, 2CASV0270, 2CASV0271, 2CASV0320., 2CASV0321, 2 CAPS 5001, 2 CAPS 5011, 2 CAPS 5041, 2 CAPS 5G48, 2CALLO200, 2CALLO201, 2CALLO270, 2CALLO271, 2CALLO320, 2CALLO321, i

In addition, a continuous discharge flowpath will be added between each ARC valve and the feedwater pressure boundary valves to assure that this piping is not pressurized due to valve leakage. These flowpaths will have a flow control valve in i line to limit flow to 5 gpm at 1700 psid.

Post Modification Testing per these procedures will be performed on the revised system to ensure correct flow balance and operation of the discharge and recirculation j flowpaths. The operation of automatic functions on a system start signal will be verified using a simulated signal. This testing ensures that the revised system still responds as designed. Operation of all revised components will be tested and data j collected to allow future determination of system health.

This modification improves the reliability of the recirculation function for the CA i pumps. The current arrangement is subject to failure due to a loss of valve control instrument power. The revised flowpath is controlled by mechanical action of the valve such that either the normal discharge flowpath or the recirculation flowpath is ,

always available. The ARC valves are considered as reliable as the current discharge check valves and recirculation control valves for valve malfunction in other respects.

The use of the ARC valves along with the provision of a flow path to prevent pressurization due to valve leakage maintains all the design operation functions being provided by the current system design. The diversion of 5 gpm of CA supply to the feedwater system during operation does not significantly impact the ability of the CA system ta meet its design requirements. No new functions are added by the installation and use of the ARC valves. The testing of the system will still utilize the old recirculation flow path since the automatic action of the ARC valve prevents its being used for testing. The ARC valves and new recirculation meet the same QA standards as the associated piping and components. No USQ exists.

g , a-e TT/2/A/9100/524 The activity being evaluated is a temporary test procedure developed to provide direction to Operations w/r/t VL AHU alignments, to gather thermal performance data on Unit 2 VL AHU subcomponents. Compliance with Technical Specificaiton temperature limits for Lower Containment average temperature will be maintained.

No unwanted system interaction will result from this activity. Based on the VL system design and the design of the involved SCCs, the effects of this activity are judged to be acceptable. The change does not effect the licensing, design basis, operation of safety function of any structure, system or component (SSC). No USQ exists.

MP/1/An650/116 The purpose of these procedures is to provide guidance for operation of the polar MP/2/An650/116 crane in Unit 1/2 upper containment. The procedures are in support of McGuire's heavy load lift program, which is implemented in accordance with the regulatory criteria associated with NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." -

The following changes were made 1) Step 11.1.2 was modified to delete the need to write within the procedure whom in Radiation Protection was notified and a minor change regarding when Radiation Protection is to be notified,2) Enclosure 13.2 is modified to clarify the means by which lift supervisor and crane operator stay in contact with each other,3) commitment change to allow for the handling of heavy loads inside containment during Modes I and 2.

The first two changes identified do not affect the handling and control of heavy loads or the operation of the polar crane. The changes are more administrative in nature,in that they involve minor revisions in certain communication processes. In 1982, the use of the polar crane to move heavy loads in containment during Modes I through 4 j was not anticipated. The regulatory provisions, criteria and guidelines for the handling and control of heavy loads do not prohibit heavy load lifts in the containment during power operation. For the movement of heavy loads within the containment, one of the following three criteria need to be satisfied; 1) use of a single failure proof crane, or 2) rapid containment isolation, or 3) heavy load drop analysis performed. A calculation was performed to ensure that a drop of the pressure hatch j plug on the pressurizer enclosure roof or operating floor and a drop of the polar crane 1 load block on to the operating floor would not damage any equipment, components, i or systems necessary for safe shutdown. The methodology employed by this calculation is in accordance with the analytical process described in the December 20, 1980 NRC letter and NUREG-0612. Based on this calculation, the operating floor  :

and the pressurizer enclosure roof can withstand a drop of the pressurizer hatch plug of the polar crane load block without loss of function (including divider barrier integrity) and thus will protect safe shutdown equipment. No USQ exists.

l l

- . 1 e a j Nuclear Station Modifications MG-12482/00 These modificati ms replaced the 125 volt DC Diesel Generator Auxiliary Power MG-22482/00 System EDGA ai.d EDGB battery chargers for the respective units. The replacement chargers are similar in function to the existing components but are supplied by a different manufacturec and have some differences in specifications and operation.

The replacement of the existing chargers eliminates problems with obsolescence, high maintenance costs and parts problems. The replacement chargers will be located in the same place as the existing components.

The 125 VDC Diesel Generator Control Power System is designed as a unit system and is comprised of diesel generator 125 VDC batteries and battery chargers EDGA, EDGB which serve diesel generators A and B on each unit. Each battery and its respective charger are housed in a metal enclosed cabinet located in the unit diesel room. The input to the replacement chargers is similar to the existing components and no new cables are required. Output from the replacement chargers is similar also with the current limit setpoint adjusted such that all battery specifications will continue to be met.

The breaker system for the chargers will be modified to add a 35 amp DC Load q breaker, and change the existing 35 amp DC output breaker to 80 amps. The diagnostic and maintenance aspects of the new components are improved. A digital voltmeter is supplied as a part of the new charger system to improve diagnostics. I The charger startup sequence is modified from the existing components. These ,

components are not accident initiators for any accidents evaluated in the UFSAR. l The modifications do not change the function of the emergency diesel generator control power system or the manner in which the system operates during any accident conditions. The modifications only update the components and improve the i maintainability of the componen'.s. No new functions are added or deleted which could increase the probability of equipment malfunctions. The replacement components provide the same functions and meet all requirements to ensure the ability of the diesel system to mitigate the consequences of accidents evaluated in the SAR. No USQ exists.

MG-12503/00 These modifications replaced the present moving filter particulate monitors (RD I MG-22503/00 05) 1,2 EMF 35, and 1,2 EMF 38 with fixed filter particulate monitors (RD054). These modifications improve monitor reliability. The current monitors have a low and high range while the replacement monitors have only a single range. A bypass around the monitor was added to allow the downstream gaseous monitors to remain in service during maintenance on the particulate monitor.

The operation of the replacement monitors is similar to the existing monitors since both are Beta Scintillation detectors. The old detectors used a dual range monitor 1 i

setup with a range of 10 to 10 million counts per minute. The replacement monitor uses a single range monitor with a range of 10 to 10 million counts per minute. The wiring of the replacement monitor to the readout module was modified to account for a single range and the moving paper features deleted. The monitor signal processing is modified to allow interface with the existing RP-86A modules. The new monitors are as sensitive as the existing monitors for detection of Cesium - 137. The Containment Air Monitor (1,2 EMF-38) has been shown by analysis to be able to detect a reactor coolant leak of I gallon per minute in containment in one hour or less as required by Regulatory Guide 1.45. Monitor sensitivity is not degraded by this replacement.

1

a Bypass lines added around the particulate monitors are provided to facilitate maintenance. This will allow filter replacement without taking the downstream gaseous activity monitors out of service. The particulate radiation monitors are not accident initiators in any SAR analysis. The monitors will continue to perform the same function in a similar manner as the previous monitors. No accidents are made more likely by this replacement. The replacement will not degrade the ability of the operator to detect abnormal conditions in the plant and take appropriate action. The replacement will not impact any safety systems, structures or components and will not prevent any safety function from being performed. No USQ exists.

MG-12504/P1 These modifications replace the Unit I and 2 train A diesel generator load sequencer MG-22504/P1 timers (D/G LS). The current timers are Agastat pneumatic and Cutler-Hammer D87 on-delay timers. The timers were replaced because of aging and degrading timing tolerances. The replacement timers are NST Tempo 812 Series solid state timers which have been tested and demonstrate better operating characteristics. The replacement timers are qualified QA-1. The replacement timers will be located in the same cabinets as the existing timers.

The replacement timers perform the same function as the currently installed timers.

Heat loads have been reviewed to assure that these components will be in an appropriate environment and will not adversely impact other equipment in the cabinets. The mounting and seismic qualification of the replacement timers has been performed to assure that all QA requirements are satisfied. All changes are contained within the electrical cabinets such that no new Appendix R concerns are generated.

The modification will be implemented on a train basis by procedure. Each affected train is covered under an independent but related NSM. The modification is implemented by implementation procedures that control the work and are reviewed independent of this evaluation. The load sequence as shown in UFSAR Figure 8-1 will continue to be met as before. The replacement timers are expected to be more consistent in performance than the current components. The replacement timers are solid state electronic components whereas the old timers are electro-mechanical devices. The electronic devices have demonstrated greater reliability and consistency in performance.

The diesel load sequencer (EQB) system or its components are not considered to be accident initiators for any accident evaluated in the SAR. These modifications do not change the function of the system such that any accident evaluated the SAR becomes more likely to occur. No indication exists for an increase in probability of malfunction of equipment due to the use of these replacement components. The EQB system is part of the emergency power system which acts to mitigate the consequences of some accidents evaluated in the SAR. These modifications do not  !

change the function or operation of the system and do not adversely affect the ability j l of the emergency power systems to perform its intended function. No USQ exists. l l

MG-12505/00 These modifications replaced the Unit I and 2 feedwater isolation valves (CF-26,28, i MG-22505/00 30, & 35) hydraulic actuators with pneumatic actuators, which will be operated by an j assured nitrogen supply. Nitrogen will be supplied from the Bulk Nitrogen System  ;

l (GN) and will be stored in individual safety related tanks in the doghouse, j i

The replacement pneumatic actuator will be operated entirely by stored nitrogen, for l the open and close directions. No internal spring is used to close the valve. l Individual safety related nitrogen tanks are provided for each feedwater isolation l valve (FIV).

The repipement of the hydraulic actuators with the new, pneumatic actuators will not l

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/ .I increase the probability of failure of the feedwater isolation valves; in fact, the probability of failure of these valves will be reduced, by improving the reliability of these actuator systems.

There are no common mode failures for these valve actuators. Each actuator will j have an independent assured, safety related supply of nitrogen for valve actuation.

No degradation is imposed on any of the affected accident mitigation equipment. No j new failure modes are created. No assumptions made in any accident analysis or j basis to any technical specification are affected by these modifiations. No USQ l

exists.

MG-12511/00 This modification replaced motor operated butterfly valve iKC0050A in the j Component Cooling (KC) system. The modification is recommended based on the j increasing valve shaft loads that were beginning to effect the actuator's ability to j maintain sufficient operating margin.

1KC0050A is a butterfly valve that is used to isolate the train A supply from the nonessential auxiliary building compc,nent cooling water header. The isolation is made to allow a full flow of component cooling water to the train A residual heat removal heat exchanger.

The replacement valve is a corrosion resistant high performance butterfly valve. The valve will be a new stainless steel butterfly valve that is GL89-10 qualified, has field replaceable seats,is torque seated design, and is equipped with a H2BC Limitorque gearbox and Limitorque SMB-00-10 motor operator.

The Spent Fuel Cooling (KF) system is an accident mitigation system; it helps prevent criticality during fuel storage and handling. The function of the KF and KC systems will remain the same. The replacement procedure will isolate KC flow to the Letdown Heat Exchanger, Seal Water Heat Exchanger, and both KF Heat Exchangers, for less than the calculated time assured to be acceptable. No plant l safety limits, limiting safety setpoints, or design parameters are affected. No new failure modes are introduced. The probability of a malfunction of equipment I l important to safety is not increased. No USQ exists.

1 MG-12514/00 These modifications added two 10 inch Class B swing check valves (NS0161 and  !

MG-22514/00 NS0163) which were installed on the Unit I and 2 A and B Containment Spray (NS) l pump discharge piping. The check valves will be located inside the NS pump room and will be downstream of the NS pump discharge pressure instrument tap. In addition, this modifications added a 2-inch drain line to train A and B of the NS l system. The drain lines were be installed between the new check valves and the NS l pump discharge pressure instruments tap, in close proximity to each check valve.

Each drain line will also contain an isolation valve (NS0162 an.d4'NS0164). This modification will also replace or relocate the existing vent valves downstream of the  :

newly installed check valves.

l These modifications have no impact on the operation, function or design basis of the j l NS system. The NS system will be operated in the same manner as before. The 4 removal of thermal energy from the containment following a loss of coolant accident (LOCA) or main steam line break (MSLB) is unaffected. The additional equipment added per this modification will not impact the accident mitigation function of the NS system. The modification will not cause, directly or indirectly, a LOCA or MSLB.

All components associated with this modification will be QA-1. The reliability of the equipment to be installed will be equivalent to what is currently installed. The check valves installed will not introduce any new failure modes than what has been L

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previously considered. The overall reliability of the NS system should be enhanced since this modification will significantly reduce the likelihood of failure of a NS train due to water hammer. No USQ exists.

MG-22475/00 This modification replaced the Auxihary Feedwater (CA) system containment isolation valves and the valve operators. The existing CA containment isolation valves are four inch gate valves with Rotork operators for the motor driven pumps (MDP) and Limitorque operators for the turbine driven pumps (TDP). The replacement CA containment isolation valves and valve operators for both the MDP and the TDP will be four inch parallel slide gate valves with Rotork 30NAl operators.

There are no changes to the function of the valves or to the function of the CA system as a result of this modification. This modification will provide additional design margin for the operation of these valves (opening or closing of the valve) during design basis accidents and other events / transients. The qualification of these valves to close against design basis differential pressure requirements developed in accordance with Generic Letter 89-10 is documented within Calculation MCC-1223.42-00-0026. There will be no change to the QA classification, the new valves are Class B QA Condition 1. The acceptability of the design for this modification is documented within Calculation MCC-1223.42.00-0044. The 20 second closure time is still well within assumptions made within accident analysis regarding the isolation of this flow path. No USQ exists.

MG-22480/00 This modification added two disconnect switches to the 125 VDC Auxiliary 1&C Power System. A disconnect was installed between the KXB manual bypass switch and the KXB panelboard. Another disconnect was installed between 2KV manual bypass switch and the 30 KVA 2KU isolation transformer which feeds panelboard 2KU. The two disconnect switches are equipped with multiple lug connections at the incoming and outgoing locations of each disconnect. The equipment is located within the battery room area of the Auxiliary Building (Elevation 733) near column BB-59.

The 125 VDC and 240/120 VAC Auxiliary Control Power Systems are not safety related. These systems supply power to loads not related to the safe shutdown of the reactor, and as such, they are not designed as Class 2E systems. The disconnect boxes added are mounted QA-4 and the boxes watertight so that the equipment can not be damaged due to sprinkler actuation. The addition of the disconnect switches has no affect on the operating characteristics, performance or function of the 125 VDC and 240/120 VAC Auxiliary Control Power Systems.These disconnect switches will not introduce new failure modes for the 125 VDC and 240/120 VAC Auxiliary Control Power Systems. The disconnect switches will allow for a more reliable means of performing routine maintenance. No USQ exists.

MG-22495/00 This modification deleted the low steam line pressure signal as an input into Safety Injection (SI) actuation circuitry internal to the Solid State Protection System (SSPS).

The low steam line pressure signals are generated in the 7300 Process Control System. The signals are then sent to both A and B trains of the SSPS. The signals are sent from here to four 2/3 Universal logic boards in the logic bay of the SSPS.

The output of these cards generates main steam isolation and safety signals. This modification removed the wiring between the logic cards and the S!/ reactor trip circuits. This will delete safety injection on low steam line pressure. The steam line isolation signal will remain. Control board labels were modified to remove reference to this safety injection function also.

l McGuire submitted a Technical Specification change by application dated October 6, (

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e i 1997 and supplemented by letter dated August 24,1998. The NRC issued approved license amendment 182/164 by letter dated September 22,1998. The amendments deleted all references to the steamline low pressure safety injection function. The impact of the change on accident analysis and plant response to an accident are considered in the NRC review and approval of this license amendment. No USQ exists.

MG-22507/00 This modification replaced the protective relaying for the Unit 2 500 kv buslines ; hat runs from the station to the McGuire Switchyard. All four channels ofline sensing and ground fault detection relays were replaced. The modification also replaced the communications portion of the buslines which involved both supervisory and transfer trip functions. This modification was implemented to replace and update old equipment which is near the end ofits service life and difficult to maintain.

The modification does not change the function of the busline Protective Relaying System. The equipment is updated and will operate faster than the current system to isolate the station from the busline in the event of a busline fault, it will not change the criteria for separation from the busline.

The Offsite Power System is the preferred source of power for station operation but is not safety related. Loss of offsite power is considered a design basis event and the station is designed and operated such that the effects of this event can be safely mitigated. The replacement of the busline protective relays does not increase the likelihood ofloss of offsite power or cause the station to remain connected to the offsite grid when abnormal conditions justify separation from the system. The criteria for operation of the Eusline protective relays will remain unchanged. The new equipment increases the reliability and speed of the system in responding to problems and will be easier to maintain due to the ready availability of parts. The replacement components meet the criteria of the Maintenance Rule in that they increase reliability and reduce the risk of maintenance preventable failures.

No USQ exists.

MG-22515/00 This modification replaced the discharge check valves for each unit 2 Auxiliary Feedwater System (CA) pump with an Automatic Recirculation Control (ARC) valve.

These valves will provide an assured minimum flow path for the CA pumps. The ARC valve operates as a combined check valve and globe valve with the globe valve port providing an assured flow path to prevent deadhead conditions on the pump.

The check valve port is the main discharge flow path and is operated in the same manner as the previous discharge check valve. The ARC valve is designed such that the globe port closes as tne check port opens assuring a continucus discharge path for the pump and automatically climinating recirculation flow as sufficient normal discharge flow is established. The installation of this recirculation flow path will result in removal from service of the current recirculation flow paths for the CA pumps. This piping will now be used only for pump flow testing. The current recirculation valves (2CA20B,2CA27A, and 2CA32AB) will have automatic controls removed and will only be operated manually. The instruments removed are:

2 CAM V0200, 2CAMV0270, 2 CAM V320, 2CASV0200, 2CASV0201, 2CASV0202, 2CASV0270, 2CASV0271, 2CASV0320, 2CASV0321, 2 CAPS 5001, 2 CAPS 501 1, 2 CAPS 5041, 2 CAPS 5048, 2CALLO200, 2CALLO201, 2CALLO270, 2CALLO271, 2CALLO320,2CALLO321.

In addition, a continuous discharge flowpath was added between each ARC valve and the feedwater pressure boundary valves to assure that this piping is not pressurized due to valve leakage. These flowpaths will have a flow control valve in line to limit flow to 5 gpm at 1700 psid. ,

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Post Modification testing will be performed on the revised system to ensure correct How balance and operation of the discharge and recirculation flowpaths. The operation of automatic functions on a system start signal will be verified using a simulated signal; This testing ensures that the revised system still responds as ,

designed. Operation of all revised components will be tested and data collected to I allow future determination of system health.

This modification improves the reliability of the recirculation function for the CA pumps. The current arrangement is subject to failure due to a loss of valve control instrument power. The revised flowpath is controlled by mechanical action of the valve such that either the normal discharge Dowpath or the recirculation Dowpath is j always available. The ARC valves are considered as reliable as the current discharge l check valves and recirculation control valves for valve malfunction in other respects.

l The use of the ARC valves along with the provision of a flow path to prevent I

pressurization due to valve leakage maintains all the design operation functions being l' provided by the current system design. The diversion of 5 gpm of CA supply to the feedwater system during operation does not significantly impact the ability of the CA system to meet its design requirements. No new functions are added by the installation and use of the ARC valves. The testing of the system will still utilize the I old recirculation flow path since the automatic action of the ARC valve prevents its being used for testing. The ARC valves and new recirculation meet the same QA standards as the associated piping and components. No USQ exists.

MG-42481/P6 This modification primarily involves security changes associated with the Dry Cask q Storage (ISFSI) project. The scope includes: i 1

• Procurement and installation ofintrusion detection equipment for expansion of j Protected Area (PA) Boundary (ISFSI site) i e PA lighting design procurement and installation oflighting equipment for i expansion of PA fence.

Closed Circuit Television (CCTV) surveillance installation e Electrical Service for ISFSI site e Spent Fuel Cask Seal Integrity Pressure Monitoring Instrumentation System design and installation.

• CAS/SAS Video switching & PICS system modifications.

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= Perimeter fence and equipment grounding, j e Security Computer 1/O points modification and addition.

• Security Fence layout and Specifications.

• Surveying / layout service coordination.

. Vehicle Barrier review for modification of PA Boundary.

• Mounting base design and installation for security intrusion detection and security lighting equipment.

= ISFSI Crash Gate specification, procurement and installation

• Anti-tunneling device installation The work is summarized to be the installation ofintrusion detection equipment, modification of the PA boundary to encompass the ISFSI site and Installation and tie-j in of the Spent Fuel Cask Seal Integrity Pressure Monitoring Instrumentation System.

The PA boundary and Security system is not described in the SAR. The Security program is reviewed under 10CFR73 and is not subject to 50.59 review. The parts of the modification which are considered separate from the security plan only involve commercial power, site survey work and installation of a non-safety system for pressure detection of dry storage canisters. No safety-related interface is involved in 1

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4 changes outside of the security-related part of the modification. No USQ exists.

MG-42502 This modification provides a new meteorological tower located west of the Energy Explorium and on the north side of the discharge canal. The new tower will be nearly 200 fee: high and will be designed to withstand 90 mph winds. The associated instrumentation in the tower and in the control room will be updated as appropriate to ensure reliability and continued compliance with regulations and licensing commitments. An 8'X16' concrete instrument shelter will be constructed near the tower. Underground electrical and telephone service will be provided for the location and nearby electrical lines will be put underground.

The tower will have an instrument elevator to facilitate calibration and maintenance of equipment. The tower will be grounded to protect it from lightning. Protection from birds is incorporated into the tower design. Freeze Protection is provided as needed for the instrumentation.

This tower is designed to be in compliance with Regulatory Guide 1.23, and instrumentation will meet Regulatory Guide 1.97 requirements. The new tower and meteorological instruments will continue to meet all requirements for providing meteorological information for routine and accident radioactive effluent dose assessment, and other local meteorological effects. No USQ exists.

MG-52493/00 This modification replaced the current air operated Nuclear Service Water (RN) system flow control valves (1RN442 & IRN445 for A Train and IRN457 & IRN460 for B Train) with a single control valve for each train (A&B) that will continue to modulate after an engineered safety feature (ESP) actuation. These valves will be actuated directly from the refrigerant via a sensing line that will be installed from the chiller condenser to the new valves. The new valves will not require any electrical power or additional controls. All electrical power and controls for the current four flow control valves will be deleted, as well instruments and computer points associated with these valves.

The new safety related control system will. allow for the automatic modulation of the RN flow control valves, even during an accident or transient that result in an ESF actuation. The replacement RN flow control valves. IRN445 (Train A) and 1RN460 (Train B) will be pilot operated valves and will be located on the outlet of the respective RN/YC chiller condenser. The new valves and the RN piping modification needed to install the valves will not have an impact on the hydrau!ic characteristics for the RN system. The flow through these valves during normal operations as well as during an accident will be essentially the same as that provided by the current valve configuration and control system. This modification not impact the ability of the RN system to provide necessary flow to both essential as well as non-essential loads. The replacement valves will be QA-1, environmentally and seismically qualified. The overall intent of the mod is to enhance the performance of the YC ,

chillers during an accident. This modification revises how the RN flow control I valves will operate. Instead of failing full open, following an ESF actuation, the l valves will continue to be automatically modulated based on the chiller refrigerant pressure. This will reduce the likelihood of a train of YC tripping due to low refrigerant pressure. No USQ exists.

MG-52494/00 This modification replaced the existing Residual Heat Removal (ND)/ Containment Spray (NS) sump level instrumentation with a float-type level instrument mounted above the sump on the 695 floor level. The existing ND/NS sump level instrument l will be abandoned in place. A new QA Condition-1 float-type level transmitter and transmitter receiver will be mounted on the top of the sump close to the existing level i transmitter so that the existing cables can he utilized. The range of the new level i

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instrumentation will be from 0 to 84 inches. l The new QA Condition-1 level transmitter and transmitter receiver to be installed will comply with all regulatory criteria and requirements associated with post-LOCA monitoring function for the ND/NS sump. The new instrument will be QA-1 and seismically qualified. The operating range for the transmitter to be installed will be fromjust above the sump to a level of 7 feet, for monitoring level up to the safety-related pump motors for ND and NS. This difference in operating range will not adversely impact the function of the instrument loop to alert operators of possible Emergency Core Cooling System (ECCS) leakage which could affect operation of i

safety related pumps. No USQ exists. l MG-52498 This modification installed diesel powered air compressors as an additional supply of Instrument Air (VI) and Station Air (VS) when needed. The compressors will be located in an enclosure on the east side of the Unit 2 turbine building. The enclosure is an additional structure to the existing plant and will be designed and constructed to house the diesels and support equipment. The enclosure will provide appropriate environmental controls for inaintaining diesel reliability and for storage of diesel fuel.

The new diesel compressors are non safety and are tied in to the existing VI system and VS system in non safety related portions of the systems. The new compressors j will also have a 120 volt power supply from the plant non safety power supply to provide crankcase heating and battery charger loads. The heaters and battery chargers improve reliability but are not required.

1 The compressors will automatically start and supply to the VI system on decreasing i VI system pressure, failure of the VI Sequencer panel, and loss of Recirculated l

Cooling Water (KR) system supply to the existing VI Compressors D E.F. The '

compressors have a capacity of approximately 2400 cfm at 100 psig and will have pressure regulators to maintain operation within normal Instrument Air (VI) System requirements. The new piping and compressors will be separated from the existing VI system by a check valve to prevent system blowdown due to pipe leak.

Connection of the new compressors to the VS system will be controlled by a manual valve.

The new compressors do not add any new functions to the VI or VS systems. The I

compressors enhance the reliability of the systems and will provide additional capacity for the VS system when needed. The new compressors and related structures, instrumentation and piping are not safety related and are not considered in response to any accidents evaluated in the SAR. The additional compressors should enhance the reliability of the VI system and improve the ability of the systems to function in accident response. No USQ exists.

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MG-52506 This modification replaced all of the Waste Gas (WG) Recombiner gas analyzers I with a more reliable model and restored the automatic capability for the Hydrogen Recombiner reagent O2 flow control. The modification also revised the QA condition associated with the Auto Gas Analyzers and associated piping. The Auto Gas l Analyzers and associated piping will be downgraded from Duke Class C (ANS Safety ]

Class 3) to Duke Class E (ANS Safety Class NNS). j i

The currently installed analyzers are obsolete and replacement parts are difficult to procure. This modification replaced the current installed analyzers with a more reliable model and will also restore the automatic capability for the Hydrogen Recombiner reagent O2 and H 2analyzers on the gas analyzer racks. The new (

analyzers have an overall accuracy of 1% span and can be setup to monitor any j j concentration range within 0-100%. The new analyzers are more reliable by virtue of  !

continuous pressure compensation, moisture tolerant cell design, and internal l

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capability to zero out helium contributions to measure hydrogen concentration. This

. modification will not result in any changes to the function, design bases or operation of the analyzers. The physical changes associated with this modification will not revise how the analyzers will be utilized. There will be no changes associated with the information that would be provided, the replacement monitors will still provide l the hydrogen and oxygen concentrations within the WG system. The alarm functions {

associated with these monitors will not be affected by this modification. The existing analyzer setpoints will be incorporated within the new analyzer, which have improved accuracy capabilities relative to the old analyzers. The setpoints will continue to provide adequate systtm monitoring. The change in safety system classification does not effect the probability of an accident since the WG system is not an accident initiator for any analyzed accidents The downgrade of th: gas analyzers will not result in an overall increase in the likelihood of a malfunction of equipment important to safety. This is considered to be an equivalent replacement.

No USQ exists.

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, MISCELLANEOUS CHANGES SLC Change - 16.1 l- The requirement to use concurrent meteorology for gaseous effluent dose pathway 16.2 calculations for the Annual Effluent Release Report comes from draft NUREG-0472 which was never issued by the NRC in final form. NUREG-0133 provides the basis for NRC effluent technical specification requirements. NUREG-0133 does not require that gaseous effluent dose pathway calculations using dispersion factors and deposition factors concurrent with the time of the release be performed if the conditions for applying annual average meteorological data set forth in the NUREG are met. SLC 16.11-16.2 was revised to read. "A five-year average of representative onsite meteorological data shall be used in the gaseous effluent dose pathway calculations. Dispersion factors (x/Qs) and deposition factors (D/Qs) shall be generated using the computer code XOQDOQ (N" REG /CR-2919) which implements NRC Regulatory Guide 1.111. The meteorological conditions concurrent with the time of release shall be reviewed annually to determine if the five-year average values should be revised." NUREG-0133 clearly allows the use of historical annual average meteorological data for calculating gaseous effluent dose pathway calculations for both long-term and short-term releases. No USQ exists.

Evaluation of software This change involves an upgrade to software used for offline core power distribution upgrade to monitoring. The current software, COMET 01,is being replaced by COMET 02. This COMET 02 SDQA- software, or the workstation it runs on, is not part of any structure, system or component as 70096-COM.R00' described in the SAR. The SSCs indirectly involved are the Movable Incore Detector System, the excore power range detectors, and the Reactor Protective System (OPAT &

OTAT trip functions).

COMET 02 is an improved version of COMET 01. The fundamental methodologies employed are unchanged. The new software incorporated various improvements and enhancements predominantly editorial in nature, with one exception. The capability was added to apply a burnup-dependent penalty factor to certain Improved Technical Specification power distribution surveillance calculations. Currently the code uses a single penalty for an entire cycle. Recently submitted proposed Technical Specification amendments for implementation of Westinghouse Fuel as described in Topical Report DPC-NE-2009/DPC-NE-2009P address the use and methodology of the burnup-dependent penalty.

COMETO2 was certified per Duke Power's directive for software verification and verified to yield the same results as COMETOl, excepting the new modifications. The modifications were verified and, as applicable, are in compliance with the Improved Technical Specifications or submitted revisions. COMETO2 is, therefore, considered equivalent to COMETOl.

This change involves no material changes to the plant. The COMET software and resident workstation are not part of any SSC important to safety and do not directly affect any SSCs.

The three systems indirectly associated, the Moveable Incore Detector System, the excore power range detectors, and the Reactor Protective System, are all unaffected by this change.

The only safety significant function performed by, or involved with, this system is to generate data to evaluate Improved Technical Specifications (ITSs) 3.2.1,3.2.2 and 3.2.4, and to periodically calibrate the power range AFD indications, as required per Surveillance Requirement 3.3.1.6 (Table 3.3.1.1). The new software is functionally equivalent to the replaced software and yields the same analytical results. Assurance of the fuel integrity limits associated with the referenced ITSs and with the OPAT and OTAT reactor trips (AFD parameter input) are not compromised. This change does not impact any plant parameters, safety limits or setpoints that potentially affect the fission product barriers. No USQ exists.

SLC 16.11-7 Selected Licensee Commitment (SLC) 16.11-7 did not provide guidance for containment

1 r, t Kadioactive Gaseous atmosphere release when EMF-39L, Containment Noble Gas Activity Monitor is inoperable.

Effluent' Monitoring item 4 on SLC Table 16.11-5 is for the Containment Purge System. Item 4 lists EMF-39L Instrumentation and as the noble gas activity monitor for this system. The Action Statement associated with 16.15-3.3.3.1, EMF-39L for the Containment Purge System requires the immediate suspension of Radiation Monitoring PURGING or VENTING of radioactive effluents via this pathway. SLC Table 17.11-5 for Plant Operation related to the Containment Air Release and Addition System was revised to allow for 3

containment air release if either EMF-39L or EMF-36L is inoperable, or up to 30 days 1 provided unit vent grab samples are taken at lease once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are l analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if both EMF-39L and EMF-36 are inoperable.

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SLC 16.15-3.3.3.1, Table 3.3-6, Action 27 is also revised to refer it to SLC Table 16.11-5 for the requirements on the Containment Air Release and Addition System. No USQ exists.

MCC 1553.26 Calculation MCC 1553.26-00-0266 documents the safety review and unresolved safety 0266 question analysis for the text changes to sections 4.2.3.3.1,4.2.4, and Table 4.4 of the UFSAR. The safety review and USQ were completed and it was determined that no USQs exists.

The UFSAR changes documented in calculation MCC 1553.26-00-0266 describe, in a comparable level of detail, the design bases of the reactivity control components as currently described in the UFSAR. They also document small design differences of the replacement reactor control components as compared to the original components. These changes do not affect the design of these components but merely clarify their design bases for the reactivity control components currently used in the reactor. The current reactivity control component designs have been previously evaluated and shown in other sources (50.59 and functionality calculations) as having met the appropriate design criteria. The capability of these components to perform their safety related function is not adversely impacted by the proposed changes to the UFSAR. None of the changes affect the function or design of the reactivity components or their capability to perform their safety related function. No USQ exists.

ITS 3.7.11 Bases, Aux Technical Specification 3.7.11 Bases imply that an Auxiliary Building Filtered Ventilation Bldg Filtered Exhaust System (ABFVES) consists of exhaust fans l A and IB, or exhaust f ans 2A and 2B.

Ventilation Exhaust This implication seems to be inconsistent with the other Technical Specifications (TS)

System where it is normally required that two trains of the system are operable, i.e. train A and train ]

B. Based on this implication both ABFVES systems would have to be declared inoperable and Condition B of Technical Specification 3.7.1I would have to be entered for both Units when a Diesel Generator (DG) that is aligned to these fans becomes inoperable because this DG supplies one fan from Unit I and one fan from Unit 2. Condition B requires one ABFVES to be returned to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Technical Specification 3.8.1, AC Power Sources, requires the inoperable diesel generator to be returned to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. However, because of the ABFVES implication described above, the DG would need to be returned to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to avoid both Units shutdown as required by TS 3.7.11 Condition C if Condition B is not met.

Technical Specification Bases 3.7.11 LCO was changed to specify that an ABFVES is considered OPERABLE when its associated two fans (1 A and 1B, or 2A and 2B, or I A and 2A, or IB and 2B, or any two fan combination provided that SR 3.7.11.4 has been performed for that combination prior to taking credit for that combination), filters, ductwork, valves, and dampers are operable. Technical Specification Bases Surveillance Requirement 3.7.11.4 was revised to specify that this surveillance requirement is required to be performed for each fan combination (I A and IB 2A and 2B,l A and 2A, IB and 2B) described in the LCO Bases. With this proposal, an 'noperable diesel generator that is aligned to these fans would make only one ABFVES inoperable and Condition A of ITS 3.7.11 would be entered which required the inoperab:e ABFVES to be returned to operable status within 7 days. In this case, the inoperable DG ,vould have to be returned to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. No USQ was identified with this TS Bases change.

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,COLR changes The changes addressed by this evaluation are the result of the implementation of the MlCl3 and M2Cl2 Improved Technical Specifications (ITS) at McGuire. The majority of the changes involve wording and nomenclature changes that were made to improve clarity and to maintain consistency between ITS and the Core Operating Limits Report (COLR). Because of technical specification numbering and ordering changes in the ITS, like kind changes were made in the COLR. The changes made to the MICl3 and M2Cl2 COLRs do not result in a change to the intent, interpretation or understanding of the technical content contained in these reports. All limits present in current COLRs and Technical Specifications were preserved in the creation of the M1Cl3 and M2Cl2 ITS COLRs. No USQ exists. ,

J SLC 17.13-2 The activity involved changes to the McGuire Selected Licensee Commitments Manual Tech Review and (Chapter 16 of the UFSAR). The proposed changes included changing " Incidents reportable  !

Control pursuant to station Technical Specification and all violations of Technical Specifications" in SLC 17.13-3

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SLC 16.13-2 Applicability e to " Incidents reportable pursuant to 10 CFR 50.73 requirements Plant Operations and all violations of Technical Specifications", adding " Facility Operating License and Review Committee Technical Specification Bases (that are attached with Technical Specification changes)" to SLC 17.13-2 Applicability c and SLC 16.13-3 Applicability a.4 to clarify that the Plant Operations Review Committee (PORC) shall review these items, clarifying the meaning of "the effectiveness of corrective actions" in SLC 16.13-3 Applicability b, and deleting the phrase "and a proposal for discretionary enforcement" in SLC 16.13 3 Applicability a.2 to clarify that the PORC shall review all JCOs. No USQs exist.

WO 98023351. This activity involves drilling three 2 inch holes through the wall around the Refueling Task 5, Core Drill of Water Storage Tank (RWST)in accordance with Drawings MC-1385-17 and MC-1385-18.

FWST Wall and immediately after each hole is completed, a Gedney plug is to be installed. The plug will be l Compensatory Action inserted from the inside of the wall. A compensatory action is also identified that requires I the above activity not be performed if weather conditions exist in which the formation of a tornado is possible.

The FWST provides a source of borated water for the Emergency Core Cooling System (ECCS) to mitigate the consequences of any UFSAR Chapter 15 accident, for normal reactivity control and for filling the refueling cavity during refueling. The lower portion of the FWST is protected from tornado missiles by a missile wall to assure sufficient borated water is available for maintaining the plant in a stable condition in the event of a main steam line rupture due to tornado generated missiles. The wall is a Category I structure that is fourteen feet high and twenty four inches thick. In the event that a design basis tornado missile strikes the wall, the missile will not strike the tank. The wall was also designed to contain water in the event the FWST is ruptured, the wall is also a water barrier.

The missile protection function of the wall will be degraded by the activity associated with this Work Order. The holes could allow a tornado missile to strike the FWST below the 14 I foot level. In the event the 1 inch steel rod where to pass through the 2 inch hole and strike the FWS T, the impact will not rupture the tank. The design criteria for the wall to serve as a water barrier is also impacted. Until the Gedney plug is installed into the hole, a opening will exist to allow water to escape from the FWST. The plug will be inserted immediately after each hole is done. As such, the time that exists in which there would be a path to allow water to escape from the FWST in the event of a rupture in the lower portion of the tank is very short (on the order of minutes). The core drill of the FWST wall will only be performed during weather conditions in which the formation of a tornado is not possible. As such, the supporting function of the FWST wall (missile barrier and water barrier) is not required. No USQ exists.

SLC 16.10.1 Current Technical Specifications do not limit the number cf steam generator power operated relief valves ( PORVs) that may be out of service. New Selected Licensee Commitment 16.10.1 will address this concern and provide Remedial Action commensurate with the number ofinoperable steam generator PORVs. Improved Standardized Technical

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, Speci0 cation 3.7.4 provides guidance in this area. Selected Licensee Commitment 16.10.1 was deleted upon implementation of Improved Technical Specification 3.7.4. No USQ exists.

TS 3.6.3 Bases Change This activity involved an editorial change to Technical Speci0 cation Bases 3.6.3, Containment Isolation Valves, to correct a referencing problem. The Bases was revised to indicate that the containment isolation valve data is located in UFSAR Table 6-111,6-112, and 6-113 instead of 6.2.4-1 The Technical Specification Bases change does not result in a change to the intent or interpretation of the technical content. The change is editorial and does not involve any changes to the operation, design bases, function of any structure, system or component, or the interpretation of the technical content. No USQ exists.

Calculation DPC- This evaluation was performed to determine if a USQ exists when the current methodology 1553.05-00-0153 is applied to a fuel design that differs from those previously benchmarked and documented in topical report DPC-NE-1004A, Reference 1. For the Duke Power Westinghouse design nuclear plants, reference I is considered applicable for Westinghouse OFA, Standard, and FCRF Mark B&W (similar to Westinghouse STD) fuels. The November 1992 SER to this topical stipulated that "the application of CASMO-3 and SIMULATE-3P to fuel designs that differ significantly from those included in the topical data base should be supported by additional code validation to ensure that the DPC-NE-1004 methodology and uncertainties apply " The new fuel type is Westinghouse Performance Plus fuel with integral fuel burnable absorber (IFBA). This is a thin coating of ZrB2 applied directly to fuel pellets of selected fuel rods. This had not previously been modeled at Duke. Supporting calculations documented the performance of a benchmark analysis and a calculation of the uncertainty factors for Performance Plus fuel with IFB A. From these calculations, it was determined that IFBA fuel can be acceptably modeled using the methodology of DPC-ME-1004A. No USQ exists.

1 COLR As part of Duke's efforts to maintain updated and accurate topical reports, a review of DPC-DPC-NE-2011PA NE-201 IPA, " Nuclear Design Methodology for Core Operating Limits or Westinghouse DPC-1553.05-00-0145 Reactors", March 1990, was conducted to identify any changes that have been made since the original approval (reference.7). Three changes, summarized below, were identified that had not previously been addressed by a 10CFR50.59 cvaluation, NRC SER, or approved Technical Specification change.

Update the methods for generating xenon transients in order to analyze more severe axial offsets in the maneuvering analysis. Section 2.3 Remove the model bias term in the square-root sum of the squares term of the SC factor in i the CFM calculations. Section 4.5 l I

Update the control rod positions used to calculate design FQ values required for Technical Specification surveillance of FQ to include the use of expected operating rod positions.

Section 6.1.

No USQ exists.

COLR The upper limits of the McGuire cycle 12 Core Operating Limits Report, Sections 3.8.1 U1 Cycle 12 (boron concentration in the Cold Leg Accumulators) and 3.2.1 and 3.9.1 (boron concentration in the Refueling Water Storage Tank) will be increased from 2675 PPM to 2875 PPM. There is no adverse effect to either (1) post accident shutdown margin, (2) time following a LOCA to transfer to hot let or (3) post LOCA containment sump pH. No USQ exists.

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,MICl210CFR50.59 A safety evaluation was performed for the Unit 1, cycle 12 core reload. This evaluation Reload Safety Eval addressed issues affected by the new core design, and does not address any other plant changes made concurrent with the refueling outage. A reduction in the time to initiate hot l leg recirculation following a loss of coolant accident (LOCA) was evaluated in order to l increase the maximum refueling water storage tank (FWST) and cold leg accumulator i

(CLA) boron concentrations to 2875 ppmB in the MICl2 Core Operating Limits Report (COLR). The MlCl2 REDSAR serves as the safety review for the USQ evaluation. No USQ exists.

l Evaluation of Topical A 50.59 cvaluation was performed to determine the presence or absence of any unreviewed l Report DPC-NE-3002 safety questions due to revisions in the UFSAR transient analysis methodology of topical Rev 2a report DPC-NE-3002 (Rev 2a). The results of the transient analyses performed using the revised methodology show that the acceptance criteria for all affected accidents continue to be met. No unreviewed safety questions associated with the revisions to topical report DPC-NE-3002 in Rev 2a were identified.

Update to DPC-NE. As part of Duke's effort to maintain updated and accurate topical reports, a review of DPC-2010PA NF-2010A was conducted to identify any changes that have been made since the original approval in June 1985 (Ref 1). Two changes were identified that had not previously been addressed in a 10CFR50.59 or approved by the NRC. The changes evaluated in this 10CFR50.59 relate to the calculation of core shutdown margin and to the calculation of maximum ejected control rod worth methodologies described in DPC-NE-1003A. The changes being evaluated are simply refinements of the original methodologies.

In the case of the shutdown margin calculation, Duke has added generic penalties to the calculation of core power defect and to the rod insertion allowance to account for the possibility of operation at off-nominal conditions within permitted operating conditions.

The topical reserves the possibility of specifically calculating shutdown margin under unusual conditions, such as when extended operation at off-nominal conditions occurs (e.g.

Axial Offset Anomaly).

For the ejected md worth analysis, the topical originally stated the " Single rods in control banks D,C, and B are removed in subsequent cases and the worth of the effected rod is j calculated...", however the Bank B rods and the group 1 rods of bank C (eighth core  !

locations B-10 & B-08 respectively) have proven to be of such low worth relative to the other locations (H-08, F-10 & D-12) that it is not necessary to analyze these locations.

Also, bank B is at most very lightly inserted, even at zero power conditions, which further reduces the available worth at B-10. No USQ exists.

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, UFSAR Changes I

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, EPG and EPL This UFSAR change made editorial and clarifying changes which correct l Systems discrepancies between the as-built condition of the 120 V AC Vital Instrumentation and Control Power (EPG) and 125 V DC Vital Instrumentation and Control Power (EPL) systems and the UFSAR descriptions. No changes were made to the as-designed, as-built equipment, functions, or operation of the EPG and EPL systems.

The probability, consequences, or possibility of an accident of a different type than evaluated in the SAR are not increased or created. These changes do not create the possibility for equipment malfunctions nor reduce the margin of safety as defined in the Technical Specifications. No USQ exists.

Section 9.5.3.4 This revision clarifies that 1.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> lights, not 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> lights are used in the Standby  ;

Shutdown Facility (SSF). Appendix R typically requires 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> self contained battery lights for hot standby. This change clarifies the current installation and approved by the NRC (reference letter from NRC dated 4/16/1984 to H.B. Tucker, 3 subject: Amendment 31 to Facility Operating License, NPF-9, Amendment 12 to Facility Operating License NPF-17). No existing plant equipment is modified. This is a documentation change only and does not affect the actual configuration of the '

plant. No USQ exists.

Chapter 7.6.7 This change involves updating the description ofinaccurate technical information to clarify Chapter 7.6.7. The changes bring the UFSAR into agreement wilt the  ;

approved design and licensing documents. The Spent Fuel Cooling Syste is described in Chapter 7.6.7. 'I he changes or corrections are limited to tF ent that.

1) They will not affect the function, design bases or operation (as shown. (he operating procedures) of any structure, system or component. 2) They not affect any operation-related information. 3) They do not affect the Technical Specifications. 4) The changes are supported by high level design documents and/or operation-related documents. j Changes were made to Sections 7.6.7 through 7.6.7.2 to the description of" Spent j Fuel Cooling System" Section 7.6.7.1.10(3) was clarified to read,"One level switch is provided in the fuel pool, for the high and the low level alarms" Section 7.6.7.2(3) was revised to read, " Components of the Spent Fuel Cooling System instrumentation are of a quality consistent with 10CFR50.65," Maintenance Rule Program." A change was made to Section 7.6.7.2(8) to correct an incorrect reference number. The statement was revised to read, 'There are no bypasses of protective functions, as defined in Section 4.12 of IEEE 279-1971, associated with this instrumentation".

l No USQ exists.

Section 9.3 This change involves correcting inaccurate technical information or adding or j making changes to verbiage to enhance or clarify information and/or concepts I contained in UFSAR Chapter 9. Corrections or changes are limited to those that: 1) have no affect on the operation, design bases, or function of any structure, system or component,2) are supported by either high level design or lower level ( i.e., i supporting) design and/or operation-related documents, and 3) do not affect the Technical Specifications.

Section 9.3.6: Various editorial changes were made throughout Section 9.3.6, including corrections to referenced figure numbers and deletion of section page numbers.

Section 9.3.6.1: Clarifies that the Boron Recycle System (NB) system is designed for l

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two core burnup dilution operations and reactor coolant pump seal leakoff flow in (

. addition to the criteria stated in system collection requirements.

Section 9.3.6.2: Mechanical seal cooling water heat exchanger outlet temperature was changed to read less than 110 'F to agree with the concentrates pump trip setpoint of l 110 *F increasing. The description of redundant heating for the recycle evaporator was revised to read, " Redundant electrical heat tracing is provided on the recycle evaporator".

Section 9.3.6.5: The description of the pressure measuring devices was changed to agree with the plant configuration as shown in the system flow diagrams.

Table 9-34: Recycle evaporator condensate demineralizer design flow was changed ,

from 35 gpm to 18 gpm (min) to 75 gpm (max) based on the demineralizer drawings. J The reactor makeup water storage tank operating temperature was changed from ll5*F to 110 F based on the inlet flow temperature setpoint. The mechanical seal cooling water pump design temperature and head at design flow were changed from 100 psig to 150 psig and from 140 ft to 200 ft, respectively, based on pump data sheets provided in the Boron Recycle System file. Mechanical seal cooling water heat exchanger outlet temperature was changed to read less than i10"F to agree with the concentrates pump trip setpoint of 110 F increasing.

Correcting inaccurate technical information or deleting, adding or making changes to verbiage that enhances or clarifies information and/or concepts contained in the UFSAR does not result in a USQ. The corrections or changes do not involve any changes to the operation, design basis, or function of any structure, system or component (SSC). No safety or licensing issues are involved. No USQ exists.

Section 3.6.2.2 Section 3.6.2.2 provides design criteria for postulated pipe breaks in all other systems except the Reactor Coolant (NC) system. One of the design criterion was revised as i follows: "A pipe break in a non-seismic system (Duke System Piping Class l D,E,G,H) cannot result in damage to an essential system necessary for the mitigation of the postulated pipe break."

i This change only clarified the intent of the design criteria. The chanFe is consistent with the overall intent of GDC-4. Specifically, the criterion states that the damage '

can not be to essential systems and equipment that are required to mitigate the consequences of the postulated pipe break, in the event a non-seismic pipe ruptures, the consequences of that event can not result in the loss of systems and equipment that would be needed to mitigate the break in the non-seismic pipe.

The change will not affect the design, fabrication, or testing of non-seismic piping nor will it affect how the piping would be installed or how the system containing the non-seismic piping would be operated. No USQ exists.

Containment Spray These changes concerning the Containment Spray System are a result of the UFSAR System accuracy review project and do not change the current operation, design bases, or function of any structure, system or component. As such, they do not constitute any physical changes to the current design, configuration or operation of the plant. The changes were made to reflect the current plant design, configuration and operating practices within the UFSAR. The changes are supported by current plant design documents, calculations and operating procedures and do not affect the Technical Specifications.

Section 6.5 and associated tables and figures: Various editorial changes were made throughout the section to add punctuation, correct spelling or correct grammatical

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errors. These changes do not alter the intent of the UFSAR and meet the

, requirements of a non-technical editorial change.

Section 6.5.2:

The explanation of the assumed sump temperature was revised to reDect current expected sump temperature values reDected in calculation MCC-1552.08-00-0277 and UFSAR Figure 6-11. The reference to UFSAR Figure 6-35 was corrected to Figure 6-11. The calculated net positive suction head (NPSH) was revised from 24.9 ft. to 25.7 ft. based upon the results of calculation MCC-122312-00-0010. The required NPSH for the containment spray pumps was revised from 18 ft. to 24 ft. to reDect the assumption that the calculated actual NPSH be compared with required NPSH at nmout now.

Spray Nozzles: The description of the fluid path through the containment spray system during recirculation was revised to more accurately describe the flow path.

Material Compatibility: A statement was added to include the exception allowing 20,000 square feet of unqualified coatings within containment.

Containment Sump: The description of the containment sump water sources and quantity of water available was revised to reflect more conservative values.

Section 6.5.4: Test and inspection criteria for the Containment Spray (NS), Residual Heat Removal (ND) Component Cooling (KC), and Nuclear Service Water systems were revised to include reference to Technical Specifications and the Pump and Valve Inservice Testing Program.

Section 6.5.5: The description of temperature instrumentation was revised to show the containment spray heat exchanger outlet temperature is indicated and not recorded. The description of the function of the Containment Pressure Control system was enhanced.

Table 6-139: The containment spray pump available NPSH was changed from 23 ft to 25.7 ft and the shutoff head was changed from 480 ft to 450 ft. The pump motor voltage was changed from 4160 to 4000 and the rpm was changed from 1780 to 1790.

The containment spray heat exchanger shell side flow was changed to 3800 for Unit I A, iB and Unit 2A, and 3400 for Unit 2B.

Table 6-141: The number of spray nozzles was revised to more accurately reflect as build conditions in the plant and the description of the comments relatmg to clogging malfunction was enhanced.

Figure 6-194: Added AB-RB-CV building wall penetrations for ND system auxiliary l spray and made editorial changes to reference plant drawing numbers. i Figure 6-196: Made an editorial change to delete dimension lines for dimensions that have been removed. l i

No USQ exists. l Chapter 6,9 These changes involve correcting system summary Dow diagram figures. Changes to the figures either correct original errors that existed in the initial revision of the figures or update the figures to incorporate the latest changes in plant configuration as

! shown in the system flow diagrams. All changes have been previously evaluated and I approved and are reDected in the latest revision of the system flow diagrams. These changes are a result of the UFSAR accuracy review project. As such, they do not l

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'r s e constitute a change to the current design, configuration or operation of the plant.

Rather, these changes are being made to reflect the current plant design, configuration and operation within the UFSAR.

The corrections or chanFes do not involve any changes to the operation, design basis or function of any structure, system or component (SSC). No safety or licensing issues are involved and no revisions to regulatory commitments are involved by the corrections or changes.

Specific Changes:

Figure 9-31: The drawing legend was revised and flow diagram references were corrected to reflect current drawing numbers. Diesel Generator Engine Starting Air System (VG) air dryers and Nuclear Service Water (RN) check valves 891 and 892 were deleted to reflect current system configuration. VG after cooler equipment numbers were corrected in Essential Equipment List.

Figure 9-57: The drawing legend was revised and flow diagram references were corrected to reflect current drawing numbers. Component Cooling (KC) heat exchanger outlet temperature instrument and Unit 2 system isolation valve was added and incorrect or missing equipment labels wem revised to show current plant configuration.

Figure 9-65: The drawing legend was revised and flow diagram references were corrected to reflect current drawing numbers. Valve symbols were reissued to agree with drawing legend. Building boundaries and critical temperature, level and flow instrumentation was added.

Figure 9-77: The flow diagram references were corrected to reflect current drawing numbers. Sample pumps were deleted and unwatering pump discharge piping was revised to reflect current plant configuration.

Figure 9-79: The flow diagram references were corrected to reflect current drawing numbers. The reciprocating air compressor after cooler was revised to show Conventional Low Pressure Service Water (RL) system heat exchanger shell side flow.

Figure 9-96: The flow diagram references were corrected to reflect current drawing numbers and discharge of valve 1012C was changed for chlorine contact chamber to the Liquid Radwaste (WL) system.

Figure 9-110: The flow diagram references were corrected to reflect current drawing numbers. Boron Recycle System valve number INB196 was correct to read INB198.

Figure 9-124: The flow diagram references were corrected to reflect current drawing numbers. The chlorination pump and educator were added and associated valves and piping were revised to reflect current plant configuration.

Correcting inaccurate information in system summary flow diagram figures contained in the UFSAR does not result in a USQ.

Auxiliary These changes to the UFSAR concerning the Auxiliary Feedwater System do not Feedwater change the current operation, design bases, or function of any structure, system or component. The changes are editorial in nature or are supported by plant design documents. These corrections or changes are a result of the UFSAR accuracy review project. They do not constitute any physical change to the current design, l

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l configuration or operation of tne plant. The chanFes were made to reflect the current

. plant design, configuration and operating practices.

Specific Changes:

A new section, Section 10.4.10 was created to specifically address the Auxiliary Feedwater System. Information concerning the Auxiliary Feedwater System was removed from UFSAR Section 10.4.7 and was placed in Section 10.4.10. This change is considered a formatting change only.

Section 10.4.10.2 (formerly Section 10.4.7.2.2): The description of the travel stop settings on the steam generator flow control valves was revised to delete specific  ;

reference to flow rates since these values vary significantly with system pressures. .

Auxiliary Feedwater System valve numbers were de-unitized because the system  !

description is applicable to both units. Information concerning the motor driven auxiliary feedwater pumps was added to the section. A reference to Table 10-10 for '

design information on the turbine driven auxiliary feedwater pump was added to the section. The term " automatic control mode" was replaced with the term " standby j readiness" to make the description consistent with current operating procedures. )

Table 10-9: The pump rated RPM was changed to agree with vendor design j information.

l Table 10-13: Table was revised to add valve CA18B to the Nuclear Service Water (RN) supply valve control, local control of manual block of automatic auxiliary feedwater pump start logic was removed, suction auto switchover to RN defeat was removed, low turbine stop valve steam pressure alarm was noted as being a computer alarm, heading of " Controls" was revised to " Indicators", condensate storage tank level indication was noted as being computer indication, and local motor running lights was added.

Figure 10-47: Piping and valve alignment was revised to reflect current plant configuration.

Correcting inaccurate technicalinformation concerning the Auxiliary Feedwater Systemdoes not result in a USQ or any licensing issues. The corrections or changes l do not affect the licensing or design bases nor will they affect the operation or safety )

function of any SSC. No USQ exists.

Condenser Cooling These changes concerning the Condenser Cooling Water System made to the UFS AR Water System do not change the current operation, design bases, for function of any structure, system or component. The changes are supported by current plant design documents.

These corrections or changes are a result of the UFSAR accuracy review project. The changes do not constitute any physical change to the current design, configuration or operation of the plat. Rather, the changes are being made to reflect the current plant design, configuration and operating practices within the UFSAR. The changes are supported by current plant design documents.

Specific Corrections:

Sections 10.4.5.1 & 10.4.5.2: Reference to the " Conventional Service Water System" was revised to read " Conventional Low Pressure Service Water System" The reference to Figure 10-29 was removed. Description of the flow level intake pumps and associate piping was changed to indicate that the Unit 2 Ivv level intake pumps were abandoned in place.

Section 10.4.5.4: System inspection requirements were reworded to state normal

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O system performance monitoring would indicate deterioration of system components

, and initiate corrective actions.

Figure 10.4.5.4: System inspection requirements were reworded to state normal system performance monitoring would indicate deterioration of system components and indicate that the Unit 2 low level intake pumps were abandoned in place.

Section 10.4.5.4: System inspection requirements were reworded to state normal system performance monitoring would indicate deterioration of system components and initiate corrective actions.

Figure 10-29: Valves RV18, RV23, and RV28 were deleted since these valves were removed from the plant. Valve positions were modified as required to indicate normal valve position during plant operation. Safeguard signal was revised to indicate valve position and add reference to the drawing legend.

Correcting inaccurate technical information concerning the Condenser Cooling Water System contained in the UFSAR does not result in a USQ or any licensing issues.

The corrections or changes do not affect the licensing or design bases nor will they affect the operation or safety function of any SSC No USQ exists.

Section 6.2.6 and Section 6.2.6 is updated to reflect changes in analysis of Hydrogen Production and Table 6-117 Accumulation in containment. The UFSAR is updated to more clearly state the quantity of Zirconium considered in the analysis and to simplify the description of corrodable metal inventory in containment. The detailed inventory values contained in Table 6-117 are removed.

i The acceptance criteria for Hydrogen Production and Accumulation is not changed by l these UFSAR changes. The description of the amount of zirconium considered in the i analysis is changed to clearly reflect that the active core volume is considered in determining the amount of hydrogen produced by the zirconium- water reaction.

Table 6-117 is revised to delete the weight and surface area of Aluminum and Zine associated with various components in containment. The acceptance criteria for these elements is based on the overall quantity and surface area which is not changed. The inventory amounts associated with various components change as plant modifications are implemented such that Table 6-117 requires very frequent revision to remain accurate. The supporting analysis is not updated with station changes since bounding )

values are assumed and qualified for Aluminum and Zine. No USQ exists.

Section 9.2.5 These changes do not constitute any change to the current design, configuration or i operation of the plant. Rather, these changes are being made to reflect the current  !

plant design and configuration within the UFSAR.

Changes:

Section 9.2.5.2 & Section 9.2.5.3: Editorial changes were made to delete reference to section pages numbers, correct referenced section numbers, and to indent enumerated component descriptions under the heading " Component Description" Section 9.2.5.5: An editorial change was made to delete reference to section page numbers. The statement "An overflow line to the spent fuel pool is available to handle overflow" was chanFed to read "An overflow line to the spent fuel pool and the fueling water storage tank roof vent are available to handle overflow". It had been determined that the overflow line to the spent fuel pool was insufficient to provide adequate overflow capabilities. Subsequently, the refueling water storage tank roof vent was modified to provide additional overflow capabilities.

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1 Table 9-25: The usable volume of the refueling water storage tank was corrected to 350,000 gallons. The design pressure and design liow of the refueling water recirculation pumps was changed to 100 gpm and 35 ft.

No USQ exists.

Chapter 9 The activity being evaluated involves updating UFSAR Chapter v.1.3 through 9.1.3.4, Table 9-1 and Figure 9-13 by making changes to description of inaccurate technical or editorial information to clarify the intent and concepts contained in the UFSAR. The Spent Fuel Cooling and Purification System (KF) is described in this chapter The changes bring UFS AR into agreement with the approved design and licensing documents. All the technical changes being evaluated have been reviewed by NRC.

The changes will not affect the functions, design basis or operation of any structure, system or component and any operation-related information. The changes are supported by either high level or low level design and/or operation related documents.

No USQ exists.

Section 9.1 - 9.1.2.4 The activity being evaluated involves updating and making changes to description of l inaccurate technical or editorial inforraation to clarify the intent and concepts

!- contained in the UFSAR. The changes bring UFSAR into agreement with the approved design and licensing documents. All the technical changes being evaluated have been reviewed by NRC in previous submittals on reracking, fuel enrichment changes or spent fuel pool storage Technical Specification changes.

These changes will not affect any fission product barriers previously analyzed and documented in UFSAR. No changes to station design or operation are associated with this update. No USQ exists.

Chapter 9 The activity being evalur.:cd involves correcting inaccurate technical information or making changes to verbiage to enhance or clarify information contained in the UFSAR. Corrections or changes are limited to those that; (1) are supported by either high level desgn or lower level (i.e. supporting) design and/or operation-related documents, (2) have no affect on the operation, design bases, or function of any structure, system or component, (3) do not affect any operation-related information and (4) do not affect any Technical Specifications. Corrections or changes do not involve any safety related or licensing issues and do not initiate or revise any regulatory commitments. 1 The changes to Chapter 9. Sections 9.5.4,9.5.5,9.5.5,9.5.7,9.5.8,9.5.9 and 9.5.11 of the UFSAR does not result in a USQ or any licensing issues. The corrections or changes do not affect the Units licensing or design bases nor will the operation or ,

safety function of any structure, system or component be affected. No USQ exists. l Chapter 9 - The activity being evaluated involves correcting inaccurate technical information or making changes to verbiage to enhance or clarify information contained in the UFSAR. Corrections or changes are limited to those that; (1) are supported by either high level design or lower level (i.e. supporting) design and/or operation-related documents. (2) have no affect on the operation, design bases, or function of any structure, system or component. (3) do not affect any operation-related information and (4) do not affect any Technical Specifications. Corrections or changes do not involve any safety related or licenring issues and do not initiate or revise any regulatory commitments.

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i The corrections or changes do not involve any changes to the operation, design basis or function of any safety related structure, system or component (SSC). No safety or licensing issues are involved and no revisions to regulatory commitments are l

involved by the corrections or changes.

The specific corrections to Chapter 9, Section 9.2.3.2 of the UFSAR addressed by this evaluation are 1) Editorial change to correct reference to figures,2) technical change concerning the supply of water to the Drinking Water System. The Drinking Water System is currently supplied from the Charlotte Municipal Utilities Department,3) minor technical change to describe flow to the main turbine lube oil coolers.

No USQ exists. j Section 12.1.3 This change revised the wording to reflect current radioactive material storage locations and provide for temporary storage of radioactive materials. The change involves adding areas for storage outside the Reactor Buildings and Auxiliary

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Building. ,

1 1

This change does not involve any changes to the operation, design basis, or function

)

of any structure, system or component. The change does not involve any safety or licensing issues. No USQ exists.

Ice Condenser This change incorporated the Operating Experience (OE) at Sequoyah Nuclear Station and in Unit 2 at McGuire Nuclear Station with respect to the ice condenser floor. The level of detail in the Ice Condenser Section of the UFSAR is greater than <

most of the systems and provides information with respect to the reaction of the ice l condenser floor to wall panel and floor defrosts. OE has shown that these I assumptions are no longer valid. j Related to the Operating Experience, McGuire never used floor defrosts with wall panel defrost as described in the UFSAR. McGuire maintenance practices rely on mechanical methods for ice removal from ice baskets and flow passages. Defrosts, l when used, have been localized and limited in time as to prevent creating excess  ;

water in the ice condenser. A %" to %" layer ofice is left on the floor at the end of a refueling outage to protect the floor from tooling damage during maintenance and provide additional thermal barrier to prevent any water from reaching the floor.

Since the UFS AR described to a significant level of detail the intended maintenance of the ice condenser the correct description of the McGuire maintenance practices are incorporated to maintain the same level of detail and provide accurate information.

The use of mechanical methods for the removal ofice from the ice condenser and the practice ofleaving a layer ofice on the ice condenser floor does not affect any ice condenser or other SSCs or any safety functions. These are merely preferred methods ofice maintenance and involve no USQs.

Section 15.6.5.3 This change is to update section 15.6.5.3 to correct the description of the Control Room Dose Calculations and assumptions to agree with other SAR documents and other sections of the UFSAR. Assumptions and parameters are modified as follows.

Assumption 5 regarding ECCS leakage is modified to agree with Table 6-130 of the UFSAR.

Assrmption 7 is modified to only credit 1% annulus volume to agree with Calculation MCC.? 227.00-00-0048.

Assumption 14 is deleted since credit is allowed for redundant hydrogen mitigation

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e equipment.

The Control Room Pressurization rate is corrected to agree with UFSAR Table 15-34 which was updated in 1994.

No USQ exists.

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