Text

Summary of Mods,Minor Mods,Procedure Changes & Other Miscellaneous Changes Made at McGuire Nuclear Station,Per 10CFR50.59 for Jan-Dec 1997
	ML20236H732
Person / Time
Site:	Mcguire, McGuire
Issue date:	12/31/1997
From:	Barron H; DUKE POWER CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	NUDOCS 9807070324
	Download: ML20236H732 (89)
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' % Duke Power Company

/; 1 A Duas Ewy Company

, McGuire Nudear Station MG01VP 12700 Hagers Ferry Rd.

Huntersville, NC 28078-9340 H. B. Barron Vice President, McGuire l V04) 875-4800 omCE Nudear Generation Department 004) 875-4809 Mx June 30, 1998 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 I

Subject:

McGuire Nuclear Station Docket No. 50-369 and 50-370 Attached is a summary report per 10 CFR 50.59 (b) (2) of Nuclear Station Modifications, Minor Modifications, procedure changes and other miscellaneous changes made at l McGuire Nuclear Station under 10 CFR 50.59 for this I reporting period. This report is being submitted with the i Updated. Final Safety Analysis Report (UFSAR) per 10 CFR l 50.71e. No Unreviewed Safety Questions were identified during this reporting period.

Questions regarding this submittal should be directed to Kay Crane, McGu're Regulatory Compliance at (704) 875-4306.

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/- - H. B. Barron, ice President h McGuire Nuclear Station

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9 U. S. Nuclear Regulatory Commission June 30, 1998 Page 2 cc: Mr. Frank Rinaldi, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Mr. Luis Reyes, Regional Administrator U. S. Nuclear Regulatory Commission Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30323 Scott Shaeffer Senior Resident Inspector McGuire Nuclear Station l

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4 Procedure Changes TN/1/A/9710/00/01M TN/1/A/9710/00/03M TN/1/A/9710/00/02M TN/1/A/9710/00/04M These procedures were used to perform the removal and installation of the rerouted steam generator blowdown (BB) and the steam generator wet layup recirculation (BW) systems to accept the new Unit'l Replacement Steam Generators (RSGs). Most of the welding of the (BB and BW) piping was performed by the fabrication of piping assemblies, which is completed in the fab shop. Prefabbing the piping assemblies minimizes welding during the installation of these assemblies inside containment. .These assemblies were installed from the penetration at the crane wall to the (BB and BW) nozzles on the RSG. Implementation of this procedure does: not af fect plant operation in any way.

Welding was performed in accordance with the Corporate Welding Program and McGuire Welding Manual and meets the requirements for QA Condition 1. Procedural activities required to support the welding involve implementation of workplace procedure SGRP-WP-07, Foreign Material Exclusion and Cutting Tools. This procedure was used while the. steam generator and its related systems were out of service.

There are no Technical Specification requirements associated with the components being removed by the use of this procedure during no' mode. Therefore, the only equipment important to safety evaluated in the SAR which is affected by this procedure is the RSG. Use of the Corporate welding program ensures structural integrity of the RSG's.

Workplace procedure SGRP-WP-07-controls the introduction of unwanted materials into the RSGs that might otherwise cause damage to the RSG or other components during operation. ,

These precautions ensure that the consequences of a j malfunction of equipment important to safety evaluated in ;

the SAR-are not increased. There are no adverse effects on I station equipment due to these activities, these procedures ;

do not increase the probability or consequence of a malfunction of equipment important to' safety in the SAR.

The procedures do not create the possibility for-a malfunction-of equipment important to safety of a different )

type than any evaluated in the SAR. .The implementation of i

.the procedures did not increase the probability or

! consequence.of an. accident of a different type than j evaluated in the SAR. No safety limit, setpoint or operating parameters have changed. The fission product barriers are not degraded as a result of implementing these procedures. The margin of safety as defined in the basis of 1

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the Technical Specifications are not reduced. No USQ exists.

TN/1/A/9297/00/01C TN/1/A/9297/00/03C TN/1/A/9297/00/02C TN/1/A/9297/00/04C These procedures provide guidelines and requirements for the final shimming / gapping of the steam generator and reactor coolant pump lateral restraints in cavities lA,lB,1C and 1D.

Due to the removal and replacement of the steam generators, reshimming/ gapping of the lateral restraints was required in order to provide the correct preoperational conditions and assure the proper functioning of the NSSS lateral restraints. These procedures outline the necessary activities required to ensure that plant and personnel safety is maintained, and makes implementation personnel aware of any limits and precautions necessary to avoid safety hazards or undesirable plant interactions.

The activities controlled by these procedures are not accident initiators for accidents evaluated in the SAR.

Therefore, they cannot increase lthe probability of an evaluated accident. The activities controlled by these TNs provide for the re-establishment of the major NSSS component seismic supports prior to fuel load. Since the seismic stability of the NSSS will be reestablished prior to fuel load, no evaluated accident consequences are affected.

Minor adjustments to support gaps may be made after fuel load, as the unit heats up to Mode 3, however these adjustments serve to enhance the functioning of these restraints, and do not compromise the seismic stability of the NSSS. The activities controlled by the TN are for re-establishment of NSSS support gaps only. Thus no possibility of any accident type different than any evaluated or no increase in any malfunction probability different than any evaluated in the SAR is introduced. l Seismic stability of the NSSS is maintained for all modes of operation except No-Mode. None of the TNs activities affect i any system alignments or functions. These activities are !

being undertaken to maintain all margins of safety. No -

fission product barriers are degraded,-and no setpoint, design limit, or operating parameter'is changed. Therefore, the activities controlled by these TNs will not reduce the margin of safety as defined in the basis for any technical specification. No USQ exists.

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'OP/1/A/6100/03 OP/1/A/6100/SD-4 OP/2/A/6100/03 OP/2/A/6100/SD-4

i. OP/1/A/6100/SD-1 OP/1/A/6100/SD-7 OP/2/A/6100/SD-1 'OP/2/A/6100/SD-7 OP/1/A/6100/SD-2 AP/1/A/5500/10

.OP/2/A/6100/SD-2 AP/2/A/5500/10' CP/0/3/8100/45 These. procedure changes incorporated a revised limit on primary to secondary steam generator leakage.

.The operations procedures govern unit operation and cooldown

-operations along with the abnormal operating procedure'for identifying'and acting on a. steam generator leak. The chemistry' procedure provides. directions for primary to j

-secondary.leakrate1 calculation and response. These procedures direct. action for a steam generator tube leak based on the McGuire commitment'to initiate unit shutdown to mode 3 within'12. hours for an identified primary to secondary steam generator: leak of 50 gallons per day (gpd) or.to initiate. shutdown and be in mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months if the rate of change-of steam generator leakage is at a rate of 50 gpd/hr or more. This commitment is changed to initiate unit-shutdown and be in. mode 3 within 6-hours for a primary to secondary steam generator leak of- 100 gpd cnr to initiate a . '

l controlled shutdown as quickly as plant conditions. permit'if 1

Lthe rate of : leakage increase is 60 gpd/hr and continuing to {

increase. (Actual procedural setpoints may be set lower.)

LThis change in commitment is made by incorporating directions.for action based on this revised commitment in the; subject procedures and by.an information letter to the :

NRC. ;

' Rapidly. opening cracks are not dependent on stable leakrate l and experience shows-that the cracks lead to a shutdown based on rate of growth regardless of the stable leakrate chosen. Failure mechanisms which do not lead to' rapidly-opening cracks are safely bounded by a leakrate of 100 gpd.

Thistamount of leakage is'still well below the analyzed

'leakrate forJaccident analysis for McGuire and well within the industry recommended leakage allowables. No revisions I to the McGuire safety analysis or licensing calculations are

.necessary,to implement this change. The revised setpoint for. initiating action is enveloped by existing calculations.

TheLtime specified to bring the unit to mode 3 has been shortened from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for a stable leak above

-100 gpd and.has been changed to "as quickly as safe plant operation permits" from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for an accelerating 3

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l leakrate. This is more conservative than the current guidance and is more in line with industry guidance for response as provided in EPRI TR-104788. The leakrate is j monitored closely by graduated levels of action based on the leakage level and continuous monitoring by both N-16 1 monitors on each steam line and by monitoring condensate air ejector exhaust radiation levels. This monitoring ensures a ;

quick response to any sudden change in leakage. l The procedure changes to increase the allowable stable leakrate from 50 gpd to 100 gpd, the rate of increase from 50 gpd/hr to as much as 60 gpd/hr and continuing to increase l and to decrease.the time for responding to a steam generator tube leak do not result in an unreviewed safety question.

These changes do not require a change to the technical specifications. A letter dated October 1, 1996 was sent to the NRC since the previous setpoints were provided to the i NRC as part of the McGuire response to concerns with steam generator tube cracking. UFSAR section 9.2.8.3 was revised to reflect the new leakage setpoint. No USQ exists. I i

PT/0/A/4150/11B This is a new test procedure for measuring the reactivity worth of individual control and shutdown rod banks. This test procedure provides the necessary steps and instructions for use of the dynamic rod worth measurement technique for measuring the reactivity worth of individual control and i shutdown rod banks. The technique involves the insertion of a rod bank at a higher than normal stepping speed allowed, in a continuous motion, and without changing boron concentration. The neutron flux signals from the upper and lower sections of an excore detector, and the rod move signal (indicate when a rod bank starts or stops) will be recorded by the Advance Digital Reactivity Computer (ADRC).

Other selected parameters, provided from either the Operator Aid Computer (OAC) or the test patch panel such as Reactor Coolant (NC) system temperature and pressurizer level, will also be recorded by the ADRC.

A Westinghouse report (WCAP-13360-P-A, Westinghouse Dynamic Rod Worth Measurement Technique) describes this new method for determining the reactivity worth of a rod bank and provides supporting analysis for the technique used to determine rod worth. This Westinghouse report was submitted to the NRC for review and approval by a letter dated May 29, 1992. By a NRC letter dated January 5, 1996, the NRC staff found the report to be acceptable and issued a safety evaluation relating to the Westinghouse report. This new test procedure complies with the conditions and limitations 4

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delineated in the NRC safety evaluation, and thus is applicable to McGuire. No USQ exists.

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OP/1/A/6100/03 AP/1/A/5500/07 OP/2/A/6100/03 AP/2/A/5500/07 OP/1/A/6250/02 PT/1/A/4253/02B OP/2/A/6250/02 PT/2/A/4253/02B l

The intent of changes to these procedures is to allow for I the isolation of Auxiliary Feedwater (CA) nozzles and actions to take due to elevated CA nozzle temperatures, including re-establishing tempering flow. The Operator Aid Computer (OAC) will also be modified to include an alarm set point for computer points A0555, A0561, A0566 and A0567 (CA nozzles.) The alarm setpoint for these computer points will ;

be set at 500 degrees F. Subsequent to the procedure changes becoming effective a minor modification (MM) will be initiated. This MM will update appropriate design documents and drawings to reflect isolation of the tempering flow to the CA nozzles.

The functions of the tempering flow to the CA nozzles and the CA nozzle tempering flow isolation valves are to: 1) minimize the potential for possible void formation resulting in a damaging water hammer event; 2) reduce thermal stresses to the CA nozzle during normal operation and those resulting from auto-start of the CA system and; 3) mitigate the consequences of accidents and events by isolating feedwater to the steam generators and by isolating this containment flow path to the environment (containment isolation). The only safety related function provided by the tempering flow l and the CA nozzle tempering flow isolation valves is the containment and feedwater isolation function.

The elimination of the tempering flow will have no impact on minimizing the possibility of void formation within the feedwater systems (CF and CA) which could cause a destructive water hammer event. Tempering flow to the CA nozzles is one of several design features and operating j practices for accomplishing this function. The elimination !

of the tempering flow to the CA nozzles, in accordance with the procedure changes, will have no affect on other design features and operating practices. The remaining design features and operating practices will continue to minimize the potential for a water hammer event.

! The tempering flow to the CA nozzles serves as a coolant for the inner surface of the CA nozzles and also protects the CA nozzles against thermal shock during'CA auto-start. Based on review of the piping stress calculations for the CA 5

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o nozzles, there is no impact if there is no_ cooling of the CA nozzles by the tempering flow. The analyzed thermal operating modes for the CA nozzles already envelopes a no I tempering flow condition. The situation of thermal shock of the CA nozzles due to CA auto-start without tempering flow has been analyzed for a finite number of these thermal shocks. There are 20 CA auto-starts without tempering flow cycles allowed. The number of CA auto-starts without l

tempering flow cycles'is' monitored-within the Allowable Operating Transient Cycles (AOTC) program. Currently, there have been zero CA auto-starts without tempering flow event for both units. At the end of the current operating cycle for each unit, the steam generators will be replaced.

Accordingly, there is a sufficient number of cycles available such that the CA nozzles will not be adversely impacted due to CA auto-start without tempering flow events, g

The procedure changes eliminate tempering flow to the CA nozzles by closing the CA nozzle tempering flow isolation valves,.which places these valves in their safety related position, which is closed. This should enhance the performance and the overall reliability of these valves since the safety related function is accomplished without having-to rely on the closure of these valves. The safety related function is now accomplished by a passive component instead of by an active component.

The new Operator Aid Computer (OAC) alarm setpoint will warn the operators of approaching saturation temperature at the CA nozzles. This should orovide sufficient warning to the operator about the possibility of voiding within the CA nozzles. The new alarm set point to be implemented will have no affect on any structure, system or component required for the safe operation of McGuire. The OAC is only utilized for monitoring and trending of various plant parameters, it does not provide any control function of plant equipment.

To prevent the possibility of a damaging water hammer the CA system is utilized to dissipate the possible void. This is accomplished by operating the applicable motor driven CA pump.and slowly throttling open the CA flow control valve to allow CA flow to pass through the CA nozzle and into the steam generator. In simultaneous feeding of a steam generator during normal operations both the CA system and the CF system is not prohibited. The licensing basis does not explicitly restrict the operation of the CA and CF systems in this configuration during normal operation. CA flow into the steam generators during this evolution should have no impact on normal plant operations and should not i

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i result in any transient or event, including a reactor trip. j i The use of CA in this manner will not adversely affect any !

! safety related function of the CA system, including the l l automatic initiation of CA. In this operating l configuration, *he CA and the CF systems will be isolated from each other, the cross-connecting valves will be closed. l There are no unreviewed safety questions regarding the elimination of tempering flow to the CA nozzles in accordance with the procedure changes for closing the CA nozzle tempering flow isolation valves. These procedure l

change will not require any change to the Technical Specifications. No design bases or safety functions for the CA system is adversely affected. No USQ exists.

OP/1/A/6250/02 OP/2/A/6250/02 These changes involve revisions to enclosure 4.2 and 4.4 of the subject procedures. Enclosure 4.2 provides operating instruction for the manual operation of the l&2A and/or 1&2B motor driven auxiliary feedwater (MDAFW) pumps. Enclosure 4.4 provides operating instruction for the manual operation of the turbine driven auxiliary feedwater (TDAFW) pump. The

! step for monitoring the temperature upstream of the check valves by RTD is deleted and is replaced with a step to l l notify engineering in the event local monitoring of CA pump i l

discharge piping-indicates possible check valve leakage. l For both enclosures, the note regarding the starting of the l CA pumps when temperature exceeds 230 degrees F while aligned te tne upper storage tank or 250 degrees F while aligned to the CA storage tank is deleted. The final change

( revises steps by substituting the temperature limit with l L "following Engineering direction." ;

l These procedure changes eliminate the continuouu monitoring function and replaces it with local monitoring at the auxiliary feedwater pump discharge once a shift. In ;

addition, the RTD temperature indications will be monitored to determine any trends that may indicate the possibility of '

leakage past the check valves. The corrective actions to ,

cool the piping as specified within the procedures is not l changed. The.only change is the method of prompting the operator to implement the corrective actions. Instead of relying on the RTDs, the actions will be taken based on l guidance from engineering.

The monitoring of the temperature once a shift is consistent with' resolution of Generic Safety Issue 93 (Steam Binding of Auxiliary Feedwater Pumps) as documented within Generic 7

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Letter 88-03. The NRC concluded that various monitoring methods employed can be highly effective in preventing steam binding, if backleakage occurs and that the recommended monitoring actions of IE Bulletin 85-01 should be continued.

As documented within the Generic Letter, the NRC specified that procedures be maintained to monitor the fluid conditions within the AFW system each shift and to maintain I

procedures for recognizing steam binding and for restoring the auxiliary feedwater system to operable status, should steam binding occur. The changes still comply with regulatory actions specified by Generic Letter 88-03 and steam binding of the auxiliary feedwater pumps will not occur.

There are no unreviewed safety questions regarding the changes to the procedure for manually operating the AFW pumps. These procedure change will not require any changes to the Technical Specifications or revisions to any SAR documents. No design bases or safety functions for the CA system is adversely affected by this modification.

These changes are temporary in nature, a configuration very similar to that which existed before will be established with the completion of minor modifications MM-8456 and MM-8547. No USQ exists.

OP/1/A/6250/02 Calculations have determined that a vortex may form in the auxiliary feedwater condensate storage tank (CACST) as water is drawn from this source to the auxiliary feedwater pumps (CA pumps.) The condition is dependent on flow rate and tank level. This condition requires that the CACST be isolated from the CA pump suction prior to the tank being drawn down to a tank level where a vortex may form and allow air entrainment into the CA pumps. The procedure changes add an enclosure to the operating procedure for the auxiliary feedwater system to direct actions of a dedicated operator for isolating the condensate storage tank and other non safety sources of water at appropriate times during any event requiring emergency CA pump operation.

An additional concern has been identified regarding operation using suction from the upper storage tank after the tank has been drained into the CA suction piping and CA pump recirculation has been initiated. The possibility exists that operation in this configuration could allow air to be trapped in the suction piping and pushed to the CA pumps causing air binding. This concern is also addressed in the procedure changes. No USQ exists.

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3 TT/2/A/9295/00/01E This procedure establishes feedwater (CF), main steam (SM),

and feedwater bypass valve conditions which will allow flow

'into'the replacement steam generators gooseneck at a flow which assures that destructive waterhammer will not occur.

The procedure is used in mode 3 (full temperature and pressure with reactor shutdown). The procedure establishes a controlled pressure differential across the CF bypass

. valve and the valve position is determined which provides the appropriate flow rate.to the CF gooseneck. This data may then be used in operations procedure to recover from plant trips where the gooseneck is uncovered for extended periods due to low steam generator level.

The potential exists for a steam void to be created if the CF gooseneck in the replacement steam generators is uncovered for.an extended period of time with feed flow isolated. The startup of the CF system may then create a waterhammer as the steam void is collapsed. Babcock and Wilxox (BWI) has determined a range of feed flow rates that I will allow restart of;the CF system without damaging I waterhammer in these situations. This procedure establishes the conditions and valve positions which will provide a flow

. rate in the acceptable range. The procedure does not require the system to be operated outside normal design conditions at any. time. The control for the feed regulated bypass valve is placed in manual with a differential pressure of 50 psi across the valve. The bypass valve position is adjusted while ultrasonic equipment is used to closely measure flow. The valve position at which an I acceptable flow is reached per the BWI calculations is recorded. ,

I The pressure settings for the SM and CF systems are l maintained within the normal operating band for the systems.

The CF_ regulating bypass valve'may normally be operated in manual lor automatic mode. The ultrasonic flow measuring :

equipment is not normally used and is set up only during this evolution. The flow measurement equipment on'.y provides data and does not effect operation of the system in any way. Normally CF flow is controlled to maintain steady temperature andfsteam generator level. During performance of this procedure, the control valve will be manually controlled to measure flow. Automatic isolation functions of the. valve.are not defeated. Temperature and level may vary in the steam generator being tested. Expected valve settings will be near the normal valve position and there should^be little change in temperature or steam generator i

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level. Administrative controls will be used to assure that conditions do not exceed any normal operating limits. The gooseneck will remain covered at all times and there will be no potential for water hammer. No USQ exists. l l

TT/2/A/9300/001 1 This procedure provides the steps to be taken in preparing for and operating the Containment Purge (VP) system during Unit 1, (2) heatup. This procedure is designed for use only !

during BOC-12 heatup with a one time technical specification i change which allows operation of the VP system during plant )

modes 4 and 3. The procedure provides the proper alignment '

and sequencing of controls, valve operations and fan starts to assure that all containment systems are in their approved alignments and no ESF actuation setpoints are reached. No USQ exists. I TT/2/A/9700/168 This is a temporary test procedure. The test method to be employed is based on the alternative test procedure j described in ASME PTC6-1996, " Performance Test Code 6 on i Steam Turbines." 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 testing is also provided. In support of test procedure TT/2/A/9700/168, a restricted change is 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 will require off-normal plant configurations during power operation, it will 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 t.st, the CA system will be fully operable and capable of performing its safety related function. The Engineered Safeguards and Reactor Trip systems and components will not be affected by this test procedure. Actions taken in accordance with the procedure to isolate the turbine cycle flows and systems will not involve the steam generator PORVs, the main steam safety valves, the main steam isolation valves, or the atmospheric 10 c______. _ _ _ _ - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

dump valves. These valves will remain operable, capable or performing their intended function. In the event of a sudden loss of load resulting in a turbine trip and a ,

reactor trip, the excess energy associated with this event l will be dissipated by the steam generator PORVs, the atmospheric dump valves and the main steam safety valves, if required. The main steam safety valves are safety related and have 100 percent relieving capacity. No USQ exists.

TT/1/A/9700/170 The replacement steam generators were tested to verify compliance with warranty requirements for moisture carryover i (MCO). The tests involved the injection of potassium I chloride (KCI) into the secondary plant upstream of the feedwater pumps at drain valve 1CM0333 and allowing the KCI to accumulate in the steam generators. The amount of MCO 1 was calculated by measuring the concentrations of potassium (K) in samples taken from various points in the secondary plant to the concentration in the steam generators.

The three primary activities which affect unit operation during the MCO test are isolation of steam generator blowdown bypassing the condensate polishing demineralizers (CPD), and injection of potassium chloride (KCI) tracer into the secondary system.

Prior to KCI injection, steam generator blowdown (BB) and the condensate polishing demineralizers (CPDs) were taken out of service to ensure stability of the KCI concentration 4 in the steam generators while sampling is being performed.

Following each test run, BB and the CPDs are returned to service to restore chemistry conditions to normal. Operation of the BB system is performed under the MCO procedure which is review by operations and chemistry. Existing unit procedures are used for operating the CPDs, and BB if invoked by the MCO test procedure,-which have been reviewed for unresolved safety questions. Two test runs were performed. Isolation of the steam generator blowdown and CPD systems does not place the unit in an alignment that is unsafe or render the unit less capable of handling any postulated accident in the SAR. Performance of the MCO test does not result in any increase in probability of a steam l generator tube rupture event of any other SAR postulated event.

The injection of KCI causes secondary chemistry to violate McGuire Chemistry Manual Action level 1 limits for a short

( period (expected to be less than 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months ) until restoration i of BB and the CPDs can restore secondary chemistry to normal 11

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i conditions. McGuire Chemistry Manual Section 3.2 allows l l operation at full power up to one week while in Action Level l 1 status. The violation of Action Level 1 limits is l performed in a controlled manner with full knowledge of j McGuire Chemistry who are integrally involved with the MCO l test. Violation of chemistry action Level 1 limits is not a ,

violation of requirements or specifications important to i safety that are contained in the Technical Specifications or !

the SAR, or a violation of steam generator design specifications. Performance of the MCO test does not re,9 ult in any increase in probability of a steam generator tube rupture event or any other SAR postulated event. There is no expected degradation of steam generator tube material or other secondary system materials as a result of the injection of KDI for the MCO test. No USQ exists.

TN/0/A/2483/00/01E TN/0/A/2484/00/01E These procedures control the installation of the modification for NSM MG-52483/00 and MG-52484/00. These modifications replaced existing 125 VDC Vital

' Instrumentation and control Power Battery EVCA and EVCB and their racks'. AT & T round cell batteries and battery racks was removed and replaced with GNB square cell battery and rack supplied by Nuclear Logistics Incorporated. These modifications also replaced the existing battery main fuse switch fuses with fuses that have a higher interrupt rating.

350MCM cables routed from battery EVCA and EVCB to

' distribution centers EVCA and EVCB were replaced with 500MCM.

Technien1 Specification Amendment Nos. 172 and 154 were approved to permit the use of a temporary battery to supply system loads during battery bank replacement. The affected channel was considered operable for up to 30 days using the temporary battery bank.

During battery replacement, the affected battery-and charger was de-energized and isolated from other plant systems. The access path for the battery during removal and installation through the QA-1 flood door separating the battery room area from the Service Building were reviewed and appropriate compensatory measures have been included in the procedure to ensure the VC and VA system requirements are met. These ;

measures also ensure that control is maintained to prevent flood or spread of fire. The battery replacement was i performed in accordance with industry safe work practices for this type of modification. Appropriate post-modification testing was performed to ensure proper 12

l t' installation of the replacement batteries. The margin of safety is not reduced by the implementation of these procedures. The installation-of the new batteries do not increase the probability or consequences of accidents previously evaluated within the SAR. No new or different i kind of accident or malfunction of equipment important to j safety is created as a result of replacing batteries in l accordance with these procedures During installation the EPL system will respond as needed in the event of an accident. No USQ exists.

TN/0/A/2488/00/01E i TN/0/A/2489/00/01E These procedures control the installation of the modification for NSM MG-52488/00 and MG-52489/00. These modifications replaced existing 125 VDC Vital L Instrumentation and Control Power Battery Chargers EVCA and

EVCB. The old chargers manufactured by C&D Charter Power i System were removed and replaced with a new chargers from Solidstate Controls Incorporated (SCI). 350MCM cables routed from battery charger EVCA'and EVCB to distribution center EVCA and EVCB was replaced with 500MCM. While battery EVCA and EBCB and battery charger EVCA and EVCB were being l replaced, a temporary battery bank was installed and connected to 125 VDC distribution center EVDA and EVDB. l Technical Specification Amendment Nos. 172 and 154 were approved to permit the use of a temporary battery to supply system loads during battery bank replacement. The affected j channel was considered operable for up to 30 days using the temporary battery bank.

During battery replacement, the affected battery and charger was de-energized and isolated from other plant systems. The access path for the battery during removal and installation through the QA-1 flood door separating the battery room area from the Service Building was reviewed and appropriate compensatory measures have been included in the procedure to ensure the VC and VA system requirements are met. These

-measures also ensure that control is maintained to prevent flood or spread of fire. The battery charger replacement was performed in accordance with industry safe work practices for this-type of modification. Appropriate post-modification testing was performed to assure proper installation of the replacement battery charger. The margin of f safety is not reduced by the implementation of this procedure. The installation of the new battery charger will not increase the probability or consequences of accidents previously evaluated within the SAR. No new or different l

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l-i L kind of accident or malfunction of equipment important to l safety is~ created as a result of replacing the battery.

During installation the EPL system will still respond as needed in the event of an accident.

TN/1/A/9610/00/01M TN/1/A/9610/03/03M TN/1/A/9610/00/02M TN/1/A/9610/00/04M These implementing procedures control the connection of approximately 40 feet of piping and 1 manual isolation valve L for the NM system. The change is due to a different nozzle location on the replacement steam generators. The new pipe ties into the existing NM upper sample piping about five feet away from the existing steam generator nozzle location.

In addition, the modification installs one new support for the.new piping and deletes or modifies a number of pipe supports on'the existing NM piping. The piping and manual valve removed for the fit up was discarded.

The steam generators and NM system will function as required and continue to meet all the necessary qualifications following these procedures. No other equipment is affected by the activities performed per these implementing procedures. For these reasons there is no increase in the probability of an accident evaluated in the SAR.

The NM system does not serve any accident mitigating functions, except for the containment isolation valves. No work will be performed on.these valves using these procedures. Foreign material exclusion practices will be used to ensure there is not adverse affect on these valves.

Part of the NM system is used for post accident monitoring for the station. The part of the system that is required for post accident monitoring will not be affected by the

' implementation of these procedures. The steam generators do serve accident mitigating functions, namely the removal of heat from the primary system by way of the steam generators and the CA system.. There is no adverse affect to these components by the implementation of these procedures. No other structure, system or component is affect by these l procedures. Therefore, the consequences of an accident l evaluated in the SAR will not be increased. l The work which is accomplished using these procedures does not introduce any new failure modes of the components and system involved. No temporary alignments were performed on plant systems by these procedures with the exception of isolating the system via a manual isolation valve. This isolation is controlled by station block tag out procedures.

f The isolation of the. system does not create the possibility l

14

of an accident of a different type than any evaluated in the SAR. These procedures do not create the possibility for an accident of a different type than any evaluated in the SAR.

The only structures, system and components affected by these ,

procedures are the steam generators and the NM system. l These procedures have no adverse affect on either the steam )

generators or the NM system. The probability or consequences of a malfunction of equipment evaluated in the j SAR is not increased. These pcocedures do not create the possibility of a malfunction of squipment of a different type than any evaluated in the SAR.

l No fission product barrier is. degraded. No setpoint, design limit, or operating parameter is changed. The activities !

prescribed by these procedures do not reduce the margin of i safety as defined in the basis for any Technical Specification. No USQ exists.

l l TN/1/A/9420/00/01M TN/1/A/9420/00/03M TN/1/A/9420/00/02M TN/1/A/9420/00/04M 1

l These temporary implementing procedures perform the removal and installation of the rerouted feedwater system to accept -

the new replacement steam generators. Most of the welding l of the CF piping was performed by the fabrication of piping assemblies, which was done in the fab shop. Prefabbing the piping assemblies minimizes welding during installation of these assemblies inside containment. The assemblies were installed from the penetration at the crane wall to the CF nozzle on the replacement steam generators. Implementation of these procedures does not affect plant operation in any way. Welding was performed in accordance with the Corporate Welding Program and McGuire Welding Manual and meets the l requirements for QA Condition 1. Procedural activities I required to support the welding involve implementation of workplace procedure SGRP-WP-07, Foreign Material Exclusion and Cutting Tools. These procedures will be used while the steam generator and its related systems are out of service.

There are no Technical Specification requirements associated with the components being removed by the use of this procedure during no mode. The only requirement important to safety evaluated in the SAR which is affected by this procedure is the replacement steam generators. Use of the Corporate Welding Program ensures structural integrity of j the replacement steam generators. Workplace procedure SGRP- !

WP-07, controls the introduction of unwanted materials into i the' replacement steam generators that might otherwise cause l damage to the replacement steam generators or other 15

l l

components during operation. These precautions ensure that the probability or consequences of a malfunction of l equipment important to safety evaluated in the SAR are not increased.

Since it has been determined that there is no adverse j l

effects on station equipment due to these activities, these '

procedures do not create the possibility for a malfunction of a different type than any evaluated in the SAR.

These procedures do not change any safety limit, setpoint, or operating parameter. The fission product barriers are not degradad a result of implementing this procedute.

Therefore, the margin of safety as defined in the basis of any Technical Specification will not be reduced.

The use of these procedures does not increase the probability or consequences of an accident previously address in the SAR, nor will the possibility of an accident be different than any already evaluated in the SAR be 3 created. No USQ exists.

TN/2/A/9285/00/01C This procedure outlines the operation of all cranes inside the region defined as the Reactor Building (RB) yard.

Conformance with the procedure ensures the continued safe operation of the plant including Unit 2 full power operation and spent fuel-pool cooling after Unit 2 is placed in no mode. The procedure serves to protect all nuclear safety related structures, systems and components within the RB yard in the event of an accident associated with a crane.

The Temporary Lifting Device (TLD) and the Outside Lift

. System (OLS) are excluded from the scope of the procedure.

The procedure controls the operation of non-safety related equipment. Although the cranes are not safety related equipment nor are they important to the safe shutdown of the plant,~it can be postulated that a crane accident.could damage nearby safety related systems, structures, or components.

As part of calculation MCC-1201.37-00-0022, safety related l structures placed at potential risk by the operation of j cranes in the RB yard were identified: the Fuel Building, I the Reactor Building, the Refueling Water Storage Tank (RWST) and the RWSTs supply trench.

This procedure controls the use of cranes in the RB yard such that the cranes cannot initiate an accident evaluated l

16

4

'in the SAR. Supporting calculations related to this procedure have determined that the activity does not have the possibility of initiating new or different types of accidents. Activities controlled by the procedure are such that equipment important to safety are not affected. No USQ exists. I PT/1/A/4200/018 PT/2/A/4200/018 The purpose of this new procedure is to check the leak rate of mechanical penetration bellows to see if there has been degradation since the last Integrated Leak Rate Test (ILRT).

The volume between the safety bellows and the test bellows will be pressurized to 3.9 psig and leak rate tested. If any leakage is detected, engineering will be contacted and a

' freon test may need to be performed to locate the leak and determine if the safety bellows is leaking. A 15 psig test ,

from inside containment may need to be performed. In the I freon test a detector is used both inside and outside containment to determine which bellows, safety (inside) or test (outside), is leaking. This procedure only involves ;

the initial 4 psig air test and the freon test (if l required). -This procedure is very similar to PT/1/A/4200/001G (Mechanical Penetration Leak Rate Test),

which is the:4 psig bellows test done to satisfy Technical Specification 4.6.1.2h, except that this procedure provides for freon testing. This test does not reduce any safety margins, nor does it increase the probability or consequences of an accident or malfunction. TR) USQ exists. ,

l TT/1/A/9100/513 l l

The purpose of this test is to record turbine speed versus time for a #1 turbine driven auxiliary feedwater (TD CA) pump start. This test is being performed prior to 1EOC13 to document the current condition of the #1 TD CA pump governor. This governor is being refurbished / replaced during the outage.

The auxiliary feedwater system is the assured, secondary l

side heat sink and must be operable to mitigate the consequences of UFSAR Chapter 15 accidents, 10CFR50 Appendix R fires, and a station black out scenario. Therefore, this system is safety significant.

The test will render the #1 TD CA pump inoperable . The #1 TD CA pump will be aligned to its normal quarterly performance test alignment. Several normally open valves will be closed, which makes the #1 TD Ca pump inoperable.

17

_ - _ _ _ =

I l

l l

However, the valves that isolate the #1 TD CA pump from the steam generators automatically open upon receipt of an auto

[ start signal. Also, the test flow path is. automatically isolated upon receipt of automatic start signal witout reactor operator action. Therefore, the #1 TD CA pump will-perform its' intended safety function during the performance

.of this test. NO'USQ exists.

1

[ TT/1/A/9100/511 !

l l The purpose of this procedure is to temporarily isolate tempering flow to steam generator 1C for the purpose of gathering data and information. This task is accomplished by closing 1CF155B. Once it is closed, data of system

. temperatures will be gathered. Once sufficient data is gathered, the valve will be reopened.

The auxiliary feedwater system (CA) is the assured, i

secondary side heat sink and must be operable to mitigate the consequences of UFSAR Chapter 15 accidents, 10CFR50 l Appendix R fires, and a station black out scenario.

Therefore, this system is safety significant.

A USQ evaluation has been performed for the permanent elimination of tempering flow. This test will only isolate tempering flow to a single steam generator for a short period of time. During the time that the tempering flow is isolated to one steam generator, tempering flow to the other three will increase. This increase in tempering flow to the other three steam' generators has no impact on the operation of the station. The procedural limits and precautions ensure system operability and are more restrictive than normal operating procedures. The isolation of tempering flow'for testing purposes is allowed per the UFSAR. No USQ

! exists.

l OP/1/A/6200/01 L OP/2/A/6200/01 l

These are the operating procedures for the chemical and l volume control systems (NV) for units 1 and 2. Enclosure q 4.16 for these procedures is the valve and power supply ,

checklist. The procedure revision is to modify the l checklist provided by Enclosure 4.16 to require that the valve position of 1,2 NV-132 and 1,2 NV-1026 is closed instead of open.

After each refueling outage and prior to entering mode 4, l Enclosure 4.16 to the NV operating procedures is performed. I I

18 l

This ensures the proper alignment of the NV system.

Maintaining 1,2NV-132 and 1,2NV-1026 closed will isolate flow through the boron meter. This will eliminate the diversion of-letdown flow through the boron meter. The flow of letdown fluid through this flow path is a potential source of reactor coolant leakage and a potential crud trap.

The isolation of this flow path should significantly reduce the likelihood of.this flow path as a source of reactor coolant (NC) leakage and significantly reduce the buildup of crud within the boron meter. The closure of these valves will have no impact on the operation and performance of the NV system. There will be no changes to the normal letdown flow by eliminating the diversion of letdown fluid through the boron meter. The leakage of letdown fluid through these valves should be insignificant and of no consequence to the safe operation of the plant. Valves 1,2 NV-1026 are % inch diaphragm valves which should provide acceptable isolation of letdown flow through the boron meter with minimal, if any, leakage. No.USQ exists.

EP/1/A/5000/ES-1.3 EP/2/A/5000/ES-1.3 These changes were made in response to McGuire PIP 0-M97-045. The changes streamline operator manual actions taken to realign the ECCS systems from injection from the refueling water storage tank to sump recirculation operation. The changes 1) move some caution statements to more appropriate location in the procedure when they are directly applicable, 2) delay actions to recover failed components until the realignment is complete for unfailed systems, and 3) . eliminate unnecessary re-verification of steps leading into these procedures. The modification of procedure steps in these revisions do not change the sequence of realignment and actions assumed in the evaluation of this evolution as described in the UFSAR.

These procedures were reviewed by the Duke Safety Analysis group to ensure that they met the requirements outlined in the UFSAR with regard to operator actions in transferring to cold leg recirculation from the injection mode to ensure that timely response is made and failures are properly I considered as required by the SAR. These changes were !

validated and tested to ensure proper actions were taken and proper results obtained. Information from this validation ,

and testing was then incorporated in the Safety' Analysis l Group evaluation of the procedure to verify acceptability.

UFSAR table 6-126 outlines major actions taken for the transfer to cold leg recirculation. This table is not 19

l

. 1 intended to detail each action taken during this evolution but defines significant actions and the sequence in which they are taken. The procedure changes made still incorporate all steps defined in table 6-126 and maintain the same sequence of operation. The changes do move other important actions, not detailed in the SAR table, to different points in the evolution and transfers l responsibility for some monitoring functions to different personnel.

l \

l These procedure changes do not result in a unreviewed safety I question. The sequence of events detailed in UFSAR Table 6-125 are not changed by these procedure revisions. No USQ exists. I l '

TT/1/A/9700/168 This is a temporary test procedure. The test method to be employed is based on the alternative test procedure described in ASME PTC6-1996, " Performance Test Code 6 on Steam Turbines." 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 testing is also provided. In support of test procedure TT/1/A/9700/168, a restricted change is made to IP/1/A/3007/017 to add another criteria for when the power range detectors are to be adjusted. j l

Although this test procedure will require off-normal plant configurations during power operation, it will 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. i l All throughout the test, the CA system will be fully operable and capable of performing its safety function. The

. Engineered Safeguards and Reactor Trip systems and components will not be affected by this procedure. Actions l taken in accordance with procedure to isolate the turbine cycle flows and systems will not involve the steam generator PORVs, the main steam safety valves, the main steam isolation valves, or the atmospheric dump valves. These valves will remain operable, capable of performing their intended function. In the event of a sudden loss of load 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 20

main steam safety valves, if required. The main steam safety valves are safety related and have 100 percent relieving capacity. No USQ exists.

TN/2/A/9203/00/02M This temporary procedure was written to expand replacement steam generator tubing into its tubesheet location. This activity will cause the as-built configuration of the tubes to match the configuration assumed by the controlling BWI design report for the RSG. Due to the fact that the activity is taking place on equipment not yet in service and well outside the perimeter of the operating plant, the activity has no impact on the operation of McGuire Unit 1 and Unit 2 reactors.

The change affects a component of a piece of safety related equipment. The component, the SG tubesheet to tubing interface, is not specifically mentioned in the UFSAR or any technical specification although the stress qualification of both the tubes and the tubesheet is addressed in Chapter 5 of the UFSAR. The equipment, the steam generator, is mentioned in both the UFSAR and Technical Specifications.

The change does not affeut tha ability of the steam generator to perform it9 required function. It does not increase the probability of a new or different type of malfunction not currently evalucted in the SAR. Based on the above, it is also seen that the probability or consequences of accidents evaluated in the SAR are not affected nor is the possibility of a different type of accident created. No safety margin defined in the basis of any technical specification is reduced. The strength of the tubesheet to tubing interface is not reduced by this activity. No USQ exists.

TN/1/A/9510/00/01M TN/1/A/9510/00/03M TN/2/A/9510/00/01M TN/2/A/9510/00/03M TN/1/A/9510/00/02M TN/1/A/9510/00/04M TN/2/A/9510/00/02M TN/2/A/9510/00/04M TN/1/A/9510/00/05M TN/2/A/9510/00/05M Nuclear station modification MG-19510 and 29510 are steam generator replacement modifications. The purpose of these procedures is to control the sequence of activities required to implement modifications MG-19510 and MG-29510. These modifications will remove Main Steam (SM) piping as required to facilitate removal of the old steam generators and installation of the replacement steam generators. Since the SM piping is prestressed in the cold condition, it must be 21

adequately restrained in order to prevent sudden or l

uncontrolled pipe movement when cut. These modifications, !

in conjunction with these procedures, establish a safe and effective method of securing the SM piping during steam generator replacement activities.

All work was performed in modes 5, 6 and no mode with the exception of an inservice leak test which was performed during mode 3. Technical Specifications 3/4.4.5 (Steam Generators), 3/4.6.3 (Containment Isolation Valves), 3/4.7.1 (Turbine Cycle), 3/4.7.1.4 (Main Steam Isolation Valves),

and 3/4. 7.8 (Snubbers) will not apply during plant modes 5,6 or no mode. Implementation of these temporary procedures 1 did not increase the probability of an accident described in i the SAR. No structure, system or component required to safely operate the plant is affected by the implementation of these procedures'. No type of accident other than any evaluated in the SAR was identified. These procedures do not directly or indirectly affect any operating equipment important to safety evaluated in the SAR. There is no j reduction in the margin of safety associated with any i technical specification. No fission product barriers are degraded, and no setpoint, design limit, or operating parameter is changed. No USQ exists.

l MP/0/A/7650/64 l This procedure was revised to include compensatory measures to ensure the Control Room Area Ventilation (VC) system remains operable for the duration of the use of the procedure. During this activity, fire penetrations were opened to allow cables to pass through the Control Room pressure boundary walls. Operability of the VC system is considered degraded while the penetrations are open since adequate pressurization can not be ensured. The procedure change adds a compensatory action to ensure the penetration is sealed within the first three minutes following an emergency core cooling system actuation.

The compensatory measure ensures a line of communication between maintenance personnel posted at the penetration and Operations personnel assigned to the Control Room is established prior to the start of the test. This line of communication is maintained the entire period in which the penetration is open per this procedure. Upon notification of an emergency core cooling system actuation, Operations personnel will notify the maintenance technician and the penetration will be sealed. No USQ exists.

22

IP/0/A/3050/13 This procedure change only helps ensure that the feedwater storage tank level instrumentation will operate as designed when called upon during design bases events. The installed design is adequate for the intended design function. No physical change to_the station is made by this procedure change. No USQ exists.

TN/0/A/2486/00/01E This implementation procedure controls the installation of modification MG-52486/00. This modification replaced existing 125 VDC Vital Instrumentation and Control Power Battery EVCD and its rack. AT&T round cell batteries and battery rack were removed and replaced with GNB square cell battery.and rack supplied by Nuclear Logistics Incorporated.

This modification also replaced the existing battery main fuse switch fuses that have higher interrupt rating. 350MCM cables routed from battery EVCD to distribution center EVDD were. replaced with 500MCM.

Technical specification amendment 172/154 was approved to permit the use of a temporary battery to supply system loads during battery bank replacement. The affected channel was considered operable for up to 30 days using the temporary battery bank.

During the battery replacement, the affected battery and charger were de-energized and isolated from other plant systems. The access path for the battery during removal and installation through the QA-1 flood door separating the i

battery room area from the Service Building was reviewed and appropriate prudent measures included in the procedure to ensure the VC and VA system requirements are met. These measures also ensure that control is maintained to prevent flood or spread of fire. The battery replacement was performed in accordance with industry safe work practices for this type of. modification. Appropriate post-modification testing will be performed to assure proper l installation of the replacement. batteries. The margin of l safety will not be reduced by the implementation-of this procedure. The installation of the new batteries in accordance with this procedure will not increase the probability of consequences of accidents previously evaluated within the SAR. No new or different kind of accident or malfunction of equipment important to safety is created as a result of replacing the batteries in accordance with this procedure. During installation the EPL system 23 j

1 1_____________ __ l

will still respond as needed in the event of an accident.

No USQ exists.

TN/1/A/9915/00/05M 1

This procedure is the implementation procedure that modifies and deletes: existing pipe hangers on the auxiliary feedwater piping in the " doghouses." The revision to this procedures

-adds detail to the sequence of work on the hangers affected by this procedure.

.The Auxiliary Feedwater System (CA) will function as required and continue to meet all the necessary qualifications. No other equipment is affected by the activities which will be performed per this implementation procedure. While the work in.this procedure does affect system which serve accident mitigating functions, namely the removal of heat from the primary system by way of the steam generators and the auxiliary feedwater system, there is no adverse affect to these components / system. No other structure, system or component is affected. No new failure modes are introduced. No temporary alignments will be

. performed on' plant systems by this procedure and no fission product barrier is degraded. No USQ exists.

PT/1/A/4350/19A PT/1/A/4350/19B' These procedures provide a method of testing Emergency Diesel Generator (EDG) 1A and 1B governor or voltage regulator maintenance which could affect EDG performance under emergency conditions. The diesel generators are tested by a step load greater than the largest single ESF loading. This load consists of injection motors on the IETB bus with a combined power rating of at least 915KW. The

. option is given to start either the RN and CA pumps or the RN, both KC and KF pumps simultaneously. These pump l combinations have been tested satisfactorily previously per approved' changes to this procedure. The EDG is also tested to verify that it can meet the technical specification load rejection acceptance criteria. This change incorporates previously approved changes and updates the procedure to the new procedure template format. This change also adds a verification that the governor and voltage regulator properly reset to 60 Hz and 4160 VAC.

The IETB switchgear is declared INOPERABLE during this testing and the length.of testing is well within on-half the technical specification allowed out of service time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Although declared inoperable, the switchgear and EDG 24

would function as designed if needed. The redundant Unit 1 EDG remains operable and able to provide the Unit 1 safety function if needed. This test does not change the design, function or methods of operation of the EDGs. No USQ exists.

PT/1/A/4350/064A This test has two sections which are performed to record data for use in validating a computer model (CYME) of 4160V essential pump motor and diesel generator performance when subjected to dynamic loading conditions. Initially, all motor loads on the 4160V essential bus being tested are shutdown. Section 12.1 has the Control Room Operator i

manually start and stop four 4160V motors (one at a time) while high speed recorders monitor electrical parameters on each motor. Section 12.2 has the Control Room Operator manually place the diesel generator (DG) on an isolated bus per normal operating procedure. With the diesel generator carrying the bus alone, the same four 4160V motors are manually started (one at a time) and are manually tripped (one at a time). All pumps are started and stopped per their respective operating procedures and use of normal flow paths.

All functional 4160V bus loads are powered from the 4160V bus not in test prior to beginning either section. In the unlikely event that the 4160V bus under test blacked out, that train would revert to the emergency mode of operation and the essential 4160V loads would sequence on. All components are operated under approved operating procedures and test conditions are well within their design capabilities. No USQ exists.

TN/1/A/9915/00/01M TN/1/A/9915/00/03M TN/1/A/9915/00/04M This procedures serve as the implementation procedures that install short radius elbows on the CA nozzle of the replacement steam generators. The CA piping fit up, the deletion of a number of pipe supports as well as the replacement of several snubbers with struts. Piping removal for the fit up will be discarded and replaced with new pipe.

The steam generators and CA system will function as required and continue to meet all the necessary qualifications following this procedure. No other equipment is affected by the activities performed per this procedure. While the work in this procedures does affect systems which serve accident 25

_ _ _ _ _ \

mitigating functions, namely the removal of heat from the primary system by way of the steam generators and the CA system, there is no adverse affect to these components / system. No other structure, system or component is affected. No fission product barrier is degraded. No USQ exists.

AP/1/A/5500/05 AP/2/A/5500/05 I These procedures provide guidance on control of the auxiliary feedwater system in the event the miniflow (recirc) valves do not operate for events other than loss of instrument air. This is an abnormal operating procedure to j provide guidance on manual control of the auxiliary feedwater system if normal control of the miniflow valves is not available. The auxiliary feedwater system is designed for operation during plant startup, a plant shutdown and emergency conditions where normal feedwater is not available. The auxiliary feedwater system can be started and controlled from either the control room or control panels local to the pumps.

These procedures provide the operator actions for manual control if the miniflou valves are not available due to loss of control power or indication. These actions are used in lieu of the normal manual control actions that are taken to control auxiliary feedwater flow to the steam generators.

The actions do not defeat any safety function or prevent the required operational features of the CA system from performing as required. No required automatic functions are defeated. The procedure assures that the auxiliary feedwater pumps are protected while performing the required feedwater supply function by assuring that minimum flow requirements of 100 gpm for the motor driven pumps and 200 gpm for the turbine driven pumps is met while the pumps are in operation.

l These procedures specifically delineate actions which have always been available to the operator. No USQ exists. l l

L TN/0/A/2491/00/01E ,

l l l

This procedure controls the installation of the modification for NSM-52491. This modification replaces existing 125 VDC vital instrumentation and control power battery charger EVCD. The old charger manufactured by C&D Charter Power System will be removed and replaced with a new charger from Solidstate_ Controls incorporated. 350MCM cables rerouted from battery charger EVCD to distribution center EVDD will i

26 l

i l

be replaced with 500McM. While battery EVCD and battery !

charger EVCD is being replaced a temporary battery bank will be installed and connected to 125 VDC distribution enter EVDD.

Technical specification amendment 172/154 was approved to

-permit the use of a temporary battery to supply system loads during battery bank replacement. The affected channel will be considered operable for up to 30 days using the temporary battery bank.

During the battery charger replacement, the affected battery and charger will be de-energized and isolated from other plant systems. The access path for the chargers during removal and installation through the QA-1 flood door separating the battery room area from the service building has been reviewed and appropriate prudent measures have been included in the procedure to ensure the VC and VA system requirements are met. These measures also ensure that control is maintained to prevent flood or spread of fire.

The battery charger replacement will be performed in

, accordance with industry safe work practices for this type

! of modification. Appropriate post-modification testing will be performed to assure proper installation of the

-replacement battery charger. The margin of safety will not L be reduced by the implementation of this procedure. The l installation of the new battery charger in accordance with l

this procedure will not increase the probability or consequences of accidents previously evaluated within the SAR. No new or different kind of accident or malfunction of equipment important to safety will be created. During

installation the EPL system will still respond as needed in L the event of an accident. No USQ exists.

TT/1/A/9815/00/04E l TT/2/A/9815/00/04E These procedures, Functional Tuning and Testing of the feedwater control system involves testing the operation and ,

control of steam generator regulating valves, bybass l regulating valves and feedwater pump speed controllers ,

during power ascension after steam generator replacement. l cThe tests consist of inducing steam generator level transients and verifying proper system response in various ;

plant alignments. Feedwater control transients are induced ;

I using test circuits installed for initial startup testing No USQ exists.

27 L - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _

r 9

l PT/0/A/4150/012 Revision 11 of this procedure, " Isothermal Temperature Coefficient of Reactivity Measurement" changes the analysis method from the endpoint method to the slope method. The Advance Digital Reactivity Computer (ADRC) is utilized to calculate Isothermal Temperature Coefficient (ITC). This change also restructures the acceptance criteria. The procedure will now specify a review criteria in addition to

'the acceptance criteria. This reissue will also reformat the procedure to be compatible with the SAROS format (document management software program) and will be modified to allow use of the ARDC for the ITC measurement. Enclosure 13.4 (Boron concentration Log) has been deleted since it is relocated to procedure PT/0/A/4150/021, Post Refueling Procedure for Criticality, Zero Power Physics and Power Escalation Testing.

The changes to this procedure will not result in plant equipment operated outside of their design specifications.

To perform this test in accordance with this procedure change, the plant will be in the same configuration as before and will be operated in the same manner as before.

The change in NC temperature will be accomplished as before, just the magnitude will be different. This change reduces the magnitude of the NC temperature change from at least 4 degrees F to at least 1.1 degrees F. The analysis of the data is to be performed by the ARDC and the method used is the slope method as defined in ANSI /ANS 19.6.1-1985. The change in analysis methods and reduction of the NC temperature change are not considered to be initiators of any previously evaluated accidents. The test criteria for review and acceptance is also revised. The changes will still ensure that unexpected results are appropriately evaluated to determine if the operating characteristics of the core are consistent with design predictions and that the core will be operated within the bounds of the safety analysis and technical specifications for McGuire. This test criteria will not cause any previously evaluated accidents. The other remaining change have no affect on plant equipment of how they are operated, the changes are editorial or administrative in nature. No USQ exists.

TP/0/A/1200/041 1 The Fire Protection System (RF) auxiliary building flow test

! will be run to assure the fire protection system in the i 1

auxiliary building meets performance standards and to gather )

28

j data for trending. No unusual or experimental conditions I are involved.

l f

l RF system water will be flowed through various piping configurations in the auxiliary building. The B main fire pump will be run throughout the test. Pressures and flows will be determined for each configuration. The first part of the test will flow water out hoses in the auxiliary building 716' level into the C WZ sump. The temporary hose will be adequately secured and the routing controlled to l prevent jeopardizing any equipment important to safety.

l Input to the sump will be limited to within sump capacity.

The other part of the test will circulate water around piping in the auxiliary building and back into the lake through permanent fire protection system nozzles used for other flow tests. No USQ exists.

TC/1/B/9400/029 The tests to be conducted under this procedure involve the addition of two pH control amines to the condensate polisher effluent line at the vendor supplied product concentrations.

The purpose of the tests is to verify the practicality of the injection method, and determine potential benefits with respect to system-concentration control, corrosion product transport and sludge deposition. The two chemicals addressed by the procedure are Dimethylamine (DMA), 2% by j weight and 3-methoxypropylamine (MPA), 40% by weight.

The use of these chemicals as described by this activity is in accordance with appropriate EPRI guidelines and consistent with established industry practices. Their i appropriateness for use in pH-and corrosion product l transport control has been documented by both field and laboratory data. They are compatible with secondary system materials of construction at established control concentrations and the vendor supplied concentrations are compatible with the proposed addition system components. No systems or components important to safety are impacted in a manner that would increase the probability or the consequences of malfunction. No USQ exists.

TT/2/A/9300/001 This procedure provides the steps to be taken in preparing for and operating the containment purge (CP) system during Unit.1, (2) heatup. This procedure is designed for use only during BOC-12 heatup with a one time technical specification change which allows operation of the VP system during plant modes 4 and 3. The procedure provides the proper alignment 29

and sequencing of controls, valve operations and fan starts to assure that all containment systems are in their approved alignments and no ESF actuation setpoints are reached. This procedure is of a temporary nature and only used to facilitate implemented changes being made to the facility.

l No USQ exists.

PT/0/A/4150/11B This is a new test procedure for measuring the reactivity worth of individual control and shutdown rod banks. This test procedure provides the necessary steps and instructions for use of the dynamic rod worth measurement technique for measuring the reactivity worth of individual control and shutdown rod banks. The technique involves the insertion of a rod bank at a higher than normal stepping speed, while not exceeding the maximum stepping speed allowed, in a continuous motion, and without changing boron concentration.

The neutron flux signals from the upper and lower sections of an excore detector, and the rod move signal (indicates when a rod bank starts or stops)will be recorded by the Advance Digital Reactivity Computer (ADRC). Other selected parameters, provided from either the Operator Aid Computer (OAC) or the test patch panel such as Reactor Coolant (;NC) system temperature (T avg or T cola) and pressurizer level, will also be recorded by the ADRC.

A Westinghouse report (WCAP-13360-P-A) describes this new method for determining the reactivity worth of a rod bank and provides supporting analysis of the technique used to determine rod worth. This Westinghouse report was submitted to the NRC for review and approval by a letter dated May 29, 1992. By a NRC letter dated January 5, 1996, the NRC staff found the report to be acceptable and issued a safety evaluation relating to the Westinghouse report. This new test procedure complies with the conditions and limitations delineated in the NRC safety evaluation, and thus is applicable to McGuire. No USQ exists.

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Nuclear Station Modifications l

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i NSM-12440 This modification replaces the control board display subsystem of the DPRI system with two independent computer systems and two CRT displays for each unit. The modification also installed new cables, connecting spare contacts associated with the reactor trip breakers to the new DPRI computer system. This will provide a means of using the reactor trip breakers as the starting point for i the rod drop timing test. Cables to the existing control l room annunciators are modified as well, so that either of the new DPRI computer systems being installed can send a i signal to the annunciators.

l The DPRI system is used to detect and display to the i operator the position of all control rods as they move up or j down in response to their respective rod drive mechanism controls. The DPRI system provides control rod position indication and alarms, including rod bottom indication, rod deviation alarm, and shutdown bank rod not fully withdrawn alarm. The DPRI system does not serve a nuclear safety related function. This system provides no protection or automatic control functions.

This modification also installs cabling between the DPRI .

control board display subsystem and some spare contacts associated with the reactor trip breakers. These spare contacts are located within the reactor trip breaker switchgear cubicles. The spare contacts are non-safety related. Malfunctions or failures within the replacement DPRI subsystem will not impact the performance of the reactor trip breakers.

The DRPI system does not serve a nuclear safety related function. The system provides no protection or automatic control functions. The system is used only to detect and display the position of all control rods as they move up or down in response to their respective rod drive mechanism controls. This system does not initiate any accident nor is it used to mitigate the consequences of an accident. No USQ exists.

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NSM-12473 NSM-22473 These modifications move Diesel Generator Lube Oil System (LD) safety related pressure switches 1/2LDPS5120, 1/2LDPS5123, 1/2LDPSS130, 1/2LDPS5133 and non-safety pressure transmitters 1/2LDPT5121, 1/2LDPT5131. These instruments are moved to improve response time for the trip switches. Along with this relocation, non-safety pressure switches 1/2LDPSS121 and 1/2LDPS5131 and non-safety pressure transmitters 1/2LDPT5120, 1/2LDPT5130 and pressure gauges 1/2LDPT5120, 1/2LDPG5130 were deleted. The pressure gauges were replaced by electronic indicators 1/2LDP5120 and 1/2LDP5130 which receive their signal from 1/2LDPT5121 and 1/2LDPT5131. The electronic indicators supply the signal to annunciator point 1/2AD19/C3 and the associated computer point. This replaces the function of 1/2LDPS5121 and 1/2LDPSS131. The overspeed dump valve and associated tubing l and components which were abandoned in place by a previous change was removed.

Safety Related Pressure Switches 1/2LDPSS122 and 1/2LDPS5132 are also deleted and their function down rated to non-safety related. These switches provided input to the before and after lube oil pump on low lube oil pressure. This function improves the lube oil function but is not. required to operate for the diesel to start and operate properly. This signal will now be provided by non-safety related pressure transmitters 1/2LDPT5121 and 1/2LDPT5131 and current alarm module. Electrical separation is provided between this non-safety signal and the lube oil pump circuit by a safety relay.

The relocation of the instrumentation results in shorter impulse time and less instrumentation on the impulse lines.

This should improve instrument response and lessen the chance of a false low lube oil pressure signal stopping the diesel. No instruments necessary for safe reliable operation of the diesel are deleted. No USQ exists. ;

i NSM-12475 This modification replaces the Asxiliary Feedwater (CA) ,

system containment isolation valves and the valve operators. l 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 30NA1 operators.

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There will be no changes to the function of the valves or to the function of the CA system as a result of this modification. This modification provides additional design margin for the operation of these valves (opening or closing or 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 will be 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.

NSM-12477 This modification adds supplementary passive neutron sensors in the reactor cavity annulus between the reactor vessel wall and the biological shield in order to maintain an adequate neutron dosimetry program as required by 10CFR50, Appendix H. The new passive sensors provide more accurate data on fast neutron fluence through the reactor vessel wall,. including the fast neutron exposure distribution as it varies azimuthally and axially. The data will be used to assess reactor vessel condition, especially in the belt line weld-region, to ensure the integrity of the reactor vessel is maintained for the design life of the vessel, and to provide a. sound basis for plant life extension. The sensors

.are. intended-to replace ~the' reactor vessel. surveillance capsules currently in place.

The modification does not increase the possibility of an accident. evaluated in the UFSAR. The modification installed a' set of passive neutron detectors in order to better l characterize neutron fluence through the reactor vessel to better. determine reactor. vessel embrittlement and integrity concerns. The system is passive and the only potential !

interaction with any existing structure, system, or component is seismic, and the mass of the system is small (approximately 25 pounds for all system components), is ;

. securely attached as not to present a seismic interaction concern.

The modification does not increase the consequences of an

! accident evaluated in the UFSAR. This modification does not

. interact with any structure, system or component during any t

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postulated accident sequence. The location of the sensors in the rector cavity annulus is such that they cannot cause any damage to equipment or instruments necessary for accident mitigation unless the sensor string comes loose and migrates to the containment sump. The location and mounting of the hardware does not present a post-LOCA hazard.

This modification will not create the possibility for an accident of a different type than any evaluated in the UFSAR. The sensors are passive and will not interact with any existing SSC during plant operations. The passive nature of the sensor string prevents any accident scenarios from being created.

The modification does not increase the probability of a malfunction of equipment important to safety or of a different type than evaluated in the UFSAR. The sensor string does not interact with any plant equipment which is important to safety, thus the passive nature of the sensor string ensures there is no equipment interaction which may increase the consequences of a malfunction of equipment important to safety or a malfunction different than previously evaluated in the UFSAR.

The modification does not reduce the margin of safety defined in any technical specification. No current margins may be negatively impacted by the addition of the sensors.

No USQ exists.

NSM-22496 This modification provides a temperature controlled area inside the missile shield surrounding the Refueling Water Storage Tank (RWST). This enclosure will replace the enclosure currently in place. The new enclosure will be provided with redundant heaters with temperature monitoring electrical outlets, telephone, and lighting for instrument protection and use during instrument calibration and testing. The enclosure will be built to QA-4 standards consistent with the missile shield. Level and temperature instruments for the FWS will be relocated inside the area.

The current narrow range level transmitters will be replaced with wide range level transmitters. Barriers are provided between the instrument lines to provide separation between redundant instrument loops. The instrument loops are provided with heat tracing outside the temperature controlled area. l 1

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l The redundant heaters in the enclosure and the heat tracing will act to remove freeze protection concerns. Safety power is provided for all safety components NSM-19020 This modification involves the installation and modification of the platforms and the modification of other items as necessary to ensure sufficient access to the replacement steam generators. A platform will be provided at the top of each steam generator. The platform will provide access to l the secondary access manway and access for ISI weld inspections. A lifting device will also be provided in each steam generator cavity. The device will be located above the steam generator top platform and be used for removal of the secondary access manway cover. Platforms used to access the secondary manways on the original steam generators were removed. No USQ exists.

NSM-19050 NSM-29050 These modifications involve the removal of items that will interfere with the removal of the original steam generators and the installation of the replacement steam generators.

The majority of the items removed by the modifications will be reinstalled and therefor effectively do not modify the plant. This work includes removal and reinstallation of the platforms above each steam generator enclosure, ladders and handrails on the steam generator cavity divider walls that interfere with the placement of the knuckleboom cranes. WI piping and its associated support restraints under the platform at El 824 in each steam generator enclosure, a humidity monitor, supports and cabling in the area of the B cavity, two cables in the B cavity enclosure, and a hydrogen sniffer, tubing and tubing supports in the B cavity.

These modifications remove (and not reinstall) the steam generator wet layup recirculation system (BW) instrument tubing that is no longer in use, a makeup demineralized water system (YN) line on the C cavity along the steam generator enclosure plat in upper containment as well as a l VI compressor, aftercooler, dryer, associated piping and pipe supports in the D cavity at the EL 824 platform that have been isolated by a previous minor modification.

Modification 19050 will also modify the lower platform in the B/C cavity to accommodate the manway cover handling device (unit 1 only), removing old members and installing 5

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l new members of the platform, reactor coolant (NC) tubing

rerouting and strut. modification,.two VR duct supports, each I

one located on a divider wall.

The changes meet all the necessary designErequirements and do not adversely affect any structures, systems and components therefore no new failures modes have been created and accident initiation is unaffected. The probability of an accident previously evaluated in the UFSAR is not increased nor is the possibility of an accident of a different. type created. Accident mitigation is unaffected since all mitigation equipment will continue to meet i applicable requirements. The consequences of an accident or malfunction of. equipment important to safety evaluated in the SAR is not increased. There are no changes to any safety limits, setpoints, or operating parameters. No fission product barriers are degraded. TheLmargin of safety in the. basis for any technical specification is not decreased. No USQ exists.

MSN-19140

-This work package provides temporary services inside

-containment during the~ steam generator replacement outage.

These' services include welding, cutting, gases, compressed air,'HVAC, and communications. Temporary power will be provided for outside the containment needs. With the exception of HVAC, these services will be provided for the replacement outage and will be removed at the'end of the outage. The containment purge vent system (VP) and VL systems will be used for ventilation during the replacement

outage.

The installation of temporary power / services for the steam generator replacement outage will not permanently change the facility. Piping'for Argon, Argon / Helium and compressed air will be run along and attached to the doghouse and containment walls. Installation of anchor bolts will not decrease.the effectiveness of these seismic category 1 structures. Credit is taken for the VP filters in the fuel handling accident inside containment (UFSAR Chapter 15).

The VP filters will be replaced if necessary and demonstrated operable per Technical Specification 4.9.4.2 prior.to reloading the core. Gas bottles located in the temporary services supply center will be tied down.

Therefore, the potential for missiles from compressed gas bottles is considered negligible. The increased fire loading in containment due to the temporary welding gases is covered by use of site directive MSD-462, Hot Work Authorization.

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No equipmant important to safety will be adversely affected by this. work activity. The effects of accidents evaluated in the-UFSAR will not be altered. No new failure modes are created nor is the margin of safety reduced. No USQ' exists.

NSM-19203 L This modification involves activities on-the replacement steam generators such'as modifying the nozzles, eddy. current testing,. ultrasonic testing, match marking, fit-up and electropolishing. These activities can be best l

' characterized as'* final prep" activities..

The modifying of the reactor' coolant (NC), main steam (SM), i feedwater (CF), auxiliary feedwater (CA), steam generator blowdown-(BB) and steam generator wet layup recirculation (BW) system nozzles and the instrument taps will alter the J' nozzle geometry.to facili*, ate welding operations and proper fit'up to-the system pining. This process is limited to removing material from the nozzles. Following this operation, the nozz1ce and the safe ends will meet al necessary. qualification requirements. .In addition, the elecropolishing~ process will also not adversely affect any of the replacement steam generators qualifications since.

.this process is limited to removal of small nick and cuts

! left on the steam generator surfaces left over from manufacturing.

The replacement steam' generators are the only components affected.by.this modification. Accident mitigation is i

~ unaffected.' There are no changes to any safety limits, i setpoints or' operating parameters. No fission product barriers are degraded. .No USQ exists.

NSM-19205 i

This modification provides~the information needed to move I l

L the old steam generators from their current location in

! containment to the retired steam generator storage facility and bring the replacement steam generators from the onsite

. manufacturing facility and place-them in their positions in l1 L

the containment building. The installation of the bolts l -that attachLthe steam generator to its support columns,

' including the appropriate shim engineering to achieve the final set criteria is provided by this package. This' modification also provides the computer model used for the L first cut onl interference removal for movement of the steam l'

generators. Items'such as load testing the haul route for movement,of the steam generators to the support columns, 7

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7 installation of lifting trunnions and temporary supports and removal of shipping restraints are provided by this package.

This package will also install the manway covers on the old l

steam generators and paint them prior to movement to the retirement facility. No USQ exists.

l NSM-19210 NSM-19210 addresses the removal and reinstallation of three I

major Reactor Building structural' steel components per steam generator. They are the steam generator enclosure dome, enclosure wall, and main steam (SM) rupture restraint.

These items were designed as bolted construction for potential removal; however, they are dimensionally very large and heavy components which require careful handling and transport. Removal and reinstallation of these components will require the temporary relocation or modification of identified structures and systems to eliminate interferences. The shear pins at the corners of the enclosure walls will be eliminated. The gasketed interface along the vertical sections of the enclosure walls will be replaced by a seal welded leak chase. No USQ exists.

NSM-19230 NSM-29230 These modifications involve removal of the existing metal reflective-(mirror) insulation from the steam generators, connected piping and piping supports and replacing it with blanket insulation.

The components used in these modifications are non-QA. All the components have been designed for seismic loadings. All the material used in this work is compatible with the ;

containment environment (ir.cluding fire concerns) , piping and piping support materials (including piping corrosion concerns) and will not affect the hydrogen evolution and heat sink calculations. The performance of the blanket insulation from an insulation perspective was compared to that~of the mirror insulation and heat load calculation or equipment qualification. Stress analysis and support / restraint design for the steam generators as well as other components that will be insulated per these i modifications is unaffected since the weight of the

-insulation was factored into the seismic analysis for the components and piping. The issue of containment sump I blockage by insulation debris following a design basis event was reviewed. This review concluded that emergency core 8

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-C cooling (ECCS) pump operation will not be adversely affected.

The~use of-blanket insulation does not adversely affect the

. stress analysis of the insulated components <nr piping nor does it' adversely affect their material integrity. In-

. addition,. emergency core cooling (ECCS) equipment operation has been shown by calculation not to be adversely affected.

Since the' insulating ability of-blanket insulation is at least equivalent to that of the existing mirror insulation containment heat removal equipment challenges are unaffected. The probability of a malfunction of equipment

'important to safety previously evaluated in'the SAR is not

' increased nor is the possibility of a malfunction of equipment of a different type than evaluated in-the SAR crer.ted.

Structures, systems and components are not been adversely affected including the pressure boundary integrity of the steam generator and process piping (material compatibly and not corrosion concerns). There are no new failure modes for the blanket insulation that serve as design basis accident precursors and accident initiation. The probability of an accident previously evaluated in the SAR is not increased nor is the possibility of an accident of a different type than evaluated in the SAR created.

Accident mitigation is unaffected by these modifications.

Containment heat removal equipment and the pressure boundary integrity of the affected piping and steam generators will continue to meet.all the applicable requirements. The consequences of an accident or malfunction of equipment important to safety evaluated in the SAR is not increased.

No USQ exists.

NSM-19250 This' modification involves installation of a redesigned upper lateral support (ULS) for the replacement steam l generators. . The redesigned ULS'uses the existing A-frames but alter the band that encircles the steam generators below the transition cone and replace the snubbers. The new band design will bolt together four pieces to completely encircle the steam generators.

The ULSs are QA1 components which support the steam generators to assist'in ensuring pressure boundary integrity. After the redesign the ULS will continue to be L QA condition 1. The material used in the USL work is L !

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. l compatible with the containment environment and will not affect the hydrogen evolution and heat sink calculations.

The modification of the ULS does not have any adverse affect on structures, systems or components since the design conditions are met. The probability of a malfunction of equipment important to safety previously evaluated in the SAR is not increased nor is the possibility of a malfunction i of equipment of a different type than evaluated in the SAR.

l Since the redesigned ULS will support the steam generator as 1 l

per design requirements and all other structures, systems {

and components have not been adversely affected there are i not new failure modes created and accident initiation is unaffected. The probability of an accident previously evaluated in the SAR is not increased nor is the possibility of an accident different than evaluated in the SAR created.

Accident mitigation is unaffected by this modification since all accident mitigating equipment will continue to meet all the applicable requirements. Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR is not increased.

There are no changes to any of the safety limits, setpoints or operating parameters as a result of this modification.

No fission product barriers are degraded. The margin of safety is not decreased. No USQ exists.

NSM-19260 This modification provides the special handling equipment that is currently not available at McGuire which will be needed to move the steam generators. This special equipment will meet a minimum design standard of ANSI N45.2.15. The qualification of the polar crane girders will be per planned ,

engineered lift guidelines of ANSI B30.2. !

The c .ponents supplied are: the support trestle, J-frame, 1 idler cart, utility cart, temporary lifting device, Allied Crane, power cart, transporters, outside lift system, test weights, prime movers and the Broderson crane. The components which will be installed inside the Reactor Building will be used during the steam generator replacement j outage only, they will not represent permanent modifications i to the plant. The installation of the foundation for the Outside Lift System (OLS) and the support trestle (exterior of the Reactor and Fuel Building) will occur prior to the outage and remain in place following the outage. The OLS structure is temporary and has connections to its' 10 I

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foundation and the fuel building but will be removed l following the outage. l l

The installation of all equipment has been evaluated to l ensure that there will be not unacceptable impact on any i accident initiators. Handling of the steam generators, and l

other heavy equipment will be procedurally controlled, to l ensure that the unit is in an evaluated condition when heavy l lifts occur. Heavy lifting conditions have been evaluated, l and it has been determined that the drop of a heavy load, up l to and including a steam generator will not have any effect :

on spent fuel pool cooling or the operating unit. Station I structures have been evaluated for the loads which will be I imposed by this modification, both inside the plant and due to the outside Lift System and have been determined to be acceptable. Buried piping and cables will not be adversely 1 impacted by this modification, they have been evaluated for j possible equipment drops and the loads imposed by the equipment which will be installed and operated during the l replacement outage. For the above reasons it has been I determined that this modification will not increase the l probability of an accident evaluated in the SAR. Similarly, {

since there will be no unacceptable impacts on station !

structures or systems it can be concluded that the possibility for an accident of a different type than any I evaluated in the SAR is not created.

l The Reactor Building portion of this modification will be implemented and removed during no mode. Evaluations have been performed for containment at this time, and is has been determined that no equipment used to mitigate the consequences of an accident will be adversely affected. The dose consequences of any load drop, both inside and outside containment have been evaluated and determined to be within l acceptable limits. The conclusion is that the consequences of an accident evaluated in the SAR will not be increased.

It can be further concluded that the consequences of a ;

l malfunction of equipment will not be increased, due to the fact that the unit will be in no mode when the Reactor Building portion of this modification is implemented, and to the precedural controls which will be in place to control l the movement of heavy items.

All containment structures are qualified for the loads which will be imposed during this modification. No adverse impact on the site ground water drainage or adverse affects result i from the removal of sells for the previously abandoned cathodic protection system. 'An evaluation has been l performed to ensure that there will be no unacceptable l interactions with unit 1 or 2 equipment due to installation 11

( or use of.theLequipment supplied by this modification. Also l the power requirements are being met such that no adverse effect to safety related power can occur as a result of this l modification, either by installation, possible handling l

accidents, or adverse weather conditions. No USQ exists.

l NSM-19270 The replacement steam generators for units 1 and 2 were delivered by railroad and off-loaded to the on Site Manufacturing Facility (OSM). The replacement steam generators were then transported by multi-axle rubber tire transporter over land from this facility to the respective Reactor Building. This transporter will be pulled / pushed by a special motorized tractor (prime mover) loaded with counterweight that enables increased traction. The same equipment will also' transport the old steam generators from the Reactor Building to the Retired Steam Generator Storage Facility (RSGSF). The route traveled by the transport vehicle and prime mover have been reviewed to assure that all surface and buried items in the proximity of the route are qualified to support the high loads imposed by the transporter and prime mover, and without causing damage to plant systems, or components. Items that cannot be qualified in their existing state will be modified as necessary to accommodate passage of the steam generators..

This modification does not reflect a permanent change to the facility. No'USQ exists.

NSM-19285 This modification provides for additional lifting devices to augment the polar crane during the steam generator replacement outage. One ringer crane, located just outside of the reactor building, two knuckle boom cranes, located atop of the steam generator cavity divider walls, and a bridge crane located on the Outside Lift System will serve to meet those needs. In the case of the ringer crane grading and crushed stone backfilling will occur outside of the equipment hatch to support the concrete slab and the crane. Everything associated with the ringer crane and bridge crane will be removed following the steam generator replacement outage. This modification does not represent a permanent change to the plant. No USQ exists.

l l NSM-19310 This modification providea for the Reactor Coolant System as-built measurement and elevation, temporary support,

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l reactor coolant pipe cutting, pipe end decon, pipe plug shielding, weld prep / beveling and narrow grove welding that is needed for steam generator replacement. Temporary supports for the hot leg, cross-over leg, and reactor coolant pumps will be used to restrain the reactor coolant system in its "as-found" condition. Non-Destructive examination will be performed on the completed weld. A lead test per ASME Code Case N-416-1 will then be performed in mode 3 to verify the leak integrity of the weld. The hot leg and cross-over leg rupture restraints will be disabled by this modification as previously approved by the NRC. No USQ exists.

NSM-19420/P2 1 This modification is part of the steam generator replacement project. In order to eliminate the problems associated with the preheater design, the replacement steam generator will incorporate a feedring design. This design raises the main feedwater elevation approximately 33 feet and will require that the main feedwater piping be rerouted when the steam l generators are replaced. Normal startup through the CA nozzle, reverse purge and the tempering flow modes of operation will be eliminated by disconnecting selected valve operators and locking closed valves. Specific pressure taps j and thermocouple will no longer be needed and will be deleted.

The rerouted CF piping will meet present criteria including NRC criteria to minimize water hammer. Piping supports will be added and deleted to the CF system piping to maintain the seismic qualification of this piping. Appropriate calculations were updated to account for the additional heat load from the rerouted CF piping. An Appendix R review has been performed as well as a pipe rupture interaction and jet impingement analyses. The piping material meets Appendix I of Section III of the ASME Code. The materials used are consistent with the containment environment and other adjacent materials.

The rerouted instrumentation tubing will maintain its i current design criteria and qualifications. l The 16" swing check valve closest to the steam generator feedwater nozzle will be eliminated by this modification.

The potential impacts of this deletion were examined and issue was identified that would require the existence of the check valve.

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l The rerouted CF piping will remain the Class B pressure

! boundary. The CF system will continue to function to provide feedwater to the secondary side of the steam generator. A number of components are no longer needed and will be deleted including one check valve per steam generator and a number of pressure taps and thermocouple.

One instrument loop in cavity 1B will be rerouted and continue to function as before. The probability of a malfunction of equipment important to safety previously evaluated in the SAR is not increased nor is the possibility of a malfunction of equipment of a different type than evaluated in the SAR created.

Rupture of a feedwater line is discussed in UFSAR Section 15.2.8. The break is postulated to occur between the CF check valve and the steam generator nozzle. With the deletion of the check valve closest to the steam generator ,

the feedwater line break is assumed to occur between the i remaining check valve and the nozzle at the terminal end.

Therefore, the break location is unchanged. Thus, the probability of an accident previously evaluated in the SAR is not increased nor is the possibility of an accident of a different type created.

The function of a number of containment isolation valves is being changes as a result of this modification. These valves will be changed from motor operated valves receiving a containment isolation signal to manual, normally closed valves needing no isolation signal. All other accident l mitigating equipment is unaffected. Therefore, the !

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

There are no changes to any of the safety limits, setpoints, or operating parameters as a result of this modification.

l No fission product barriers are degraded. The margin of safety as defined in the basis for any technical specification is not decreased. This evaluation takes credit for the in-depth review of the deletion of the CF check valve. No USQ exists. l I

NSM-19510 This modification provides for the cutting, removal and l replacement of a portion of the main steam process and guard j pipe necessary for steam generator replacement. They will be re-welded and inspected in accordance with NSD-400, Corporate Welding Manual. The SM piping will be subject to a leakage test per ASME Code Case N416-1. The SM piping at 14 i

the nozzle of the steam generators is not considered a target for other high energy line breaks. Since the SM piping welds will be equivalent to the original welds there is not increase in the probability of an accident or .

malfunction of equipment important to safety evaluated in the UFSAR.

The replacement steam generators are approximately 80 tons heavier at operating. conditions with a slightly higher center of. gravity. This change would result in higher seismic loads on the SM. piping and supports. The ISM methodology was used for reanalysis and the piping and I supports were requalified or modified for the new loads.

l The effect of a main steam line break at the steam generator

outlet nozzle was considered in Section 6.2.1.2 of the

! UFSAR.. The guard. pipe enclosing the main steam line acts in

! conjunction with a crush pipe energy absorber device and-load transfer device to limit the break opening area. The current steam generator subcompartment analysis described in UFSAE section 6.2.1.2 remains valid. Other Chapter 15 accidents are unaffected. Therefore, there is not increase i in the consequences of any accident or malfunction of l equipment important to safety previously evaluated in the UFSAR.

This modification does.not result in any degradation or adverse. impact to any structure, system or component important to safety. No new failure modes were' introduced l as a result of the modification. All component functional capabilities were maintained. No new functional L requirements were added as a result of this modification.

h Therefore, the possibility for an accident or for a malfunction of a different type than any evaluated in the UFSAR has not been created.

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~No safety limit, setpoint, or operating parameter will be changed as a result of this modification. The fission product barriers are not being degraded as a result of this modification. Therefore, the margin of safety as defined in the basis of the Technical Specifications will not be reduced. No USQ exists.

NSM-19710 This modification involves the reconfiguration of the BB and I BW system piping in the area of the steam generators to accommodate the replacement steam generators. Existing BB piping at the steam generators will transition from 2" OD of the pipe to the 3" OD of,the replacement steam generator ;

blowdown nozzles. This transition will occur at each of the l 15 !

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8 nozzles (2 per steam generator.) Additionally, the l replacement steam generators have a BW tap which the original steam generators did not have. This tap in service during wet lay-up to ensure sufficient recirculation I the area of the tube bundle. Piping to this nozzle will be a 2" OD line and come off the BB system.

There are no adverse affects on structures, systems or components. There are no new failure modes introduced.

Accident mitigation is unaffected since the BB/BW system will function as it did prior to the modification including containment isolation functions. There are no changes to safety limits, setpoints, or operating parameters. No fission product barrier is degraded. No USQ exists.

NSM-19810 This modification provides the instrumentation tubing for the wide range and narrow range steam generator level transmitters for the steam generator replacement project.

The instrument tubing for the replacement steam generators l is designed to meet the same criteria as the current tubing.

The tubing is Duke class B. A walkdown was performed to '

identify any pipe rupture interactions that would impact the instrument tubing. Three interactions were identified. The affected instrument tubing will be protected with jet l

barriers. The reference legs for the narrow range instrumentation will be insulated to guard against l measurement errors due to the effects of a high energy line i

break. Testing will be performed to ensure that the instruments will function properly and will meet all design l criteria and performance standards, including response time.

l These tests will be performed as part of modification NSM-l 19815. This instrument tubing is not an accident initiator; it serves an accident mitigation function. The instrument tubing will continue to meet current requirements, and will

! therefore not increase the probability of an accident evaluated in the SAR. Since no new failures are created, this modification will not create the possibility for an accident of a different type than any evaluated in the SAR.

The instrument tubing continues to meet the required criteria to allow the wide and narrow range instrument to function as designed. No setpoints are changes due to this modification, however the technical specification change submitted September 30, 1994 will change setpoints for instrumentation which this tubing supports. The instruments served by the rerouted tubing provide an accident mitigation j function in that the narrow range instrumentation provides l l

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the low-low and high-high steam generator ESFAS functions previously identified. Since this modification will continue to allow the narrow and wide range level instrumentation to function properly, the consequences of an accident and malfunction of equipment important to safety evaluated in the SAR will not be increased.

This modification will allow the wide and narrow range instrumentation to function as required for the replacement steam generators. The tubing will meet the same design i' asis requirements as the original instrument tubing. The replacement root valves will meet the same design basis requirements as the original root valves. No new failure modes are created. For this reason, the modification will not increase the probability or a malfunction of equipment important to safety evaluated in the SAR. Since the equipment will operate in the same manner as it does currently, the possibility for a malfunction of a different type is not created.

This modification does not change any safety limit, setpoint, or operating parameter. Setpoint change related to these instruments were procedurally handled by the technical specification change. The fission product barriers are not i degraded. There is no degradation to containment imposed.

The margin of safety as defined in the technical specifications is not reduced. No USQ exists. ,

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l NSM-19815 NSM-29815 These modifications replace all the narrow range steam

! generator level transmitters (QA1) while replacing the QA1 l narrow range level and recalibrating the wide range transmitters. The existing loose parts monitor and new cabling will be installed on the replacement steam generators. Feedwater control, Tner for rod control and the steam dumps will be modified to reflect the new steam generator water level and Tag, respectively. Steam generator lo level alarm and the lo-lo level setpoint will be revised. The setpoint will change to a constant, non-nuclear power dependent, level and alarm to be consistent with the new setpoint. The hi-hi steam generator water level will also be change due to the changed span. The OAC programming for a " watchdog" route OPDT and OTDT will change as a result of the revised Tng. The pressurizer level control program will have the program setting changed. OPDT and O?DT setpoints for a reactor trip will change.

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There are no negative impacts on any structure, system or l component which are used to mitigate the consequences of an accident. These modifications make the setpoint changes needed to operated the replacement steam generators in accordance with the technical specification changes and the safety analysis assumptions made for the replacement steam generators. No fission product barrier is degraded. No USQ exists, j

NSM-19915 This modification replaces the auxiliary feedwater elbow j attached to the auxiliary feedwater steam generator nozzle. l This modification is the result of both the steam generator i replacement and of a study that recommended, based on i thermal fatigue, that the CA pipe elbows attached to the steam generator flow nozzles, be replaced. This thermal 1 fatigue is the result of CA tempering flow to the steam generators. Replacement of the elbows will renew the elbow fatigue ~1ife, and tempering flow through these nozzles will l no longer be used. Due to the new steam generator design, the distance from the center of the steam generator to the end of the Ca nozzle must be increased by 9/16 inch.

Shorter radius elbows will be used to compensate for the i longer nozzles.

The temperature element on each ca line going into the steam generators (located on the pipe transition piece transitioning the pipe from schedule 160 to schedule 80) )

will be removed and replaced on the new transition piece. I New thermowells will be purchased for these temperature elements. In addition, a seismic accelerograph on the top of the elbow going into the D steam generator will be removed and replaced in the same relative location.

There are no changes to any of the safety limits, setpoints or operating parameters as a result of this modification.

No fission product barrier will be degraded. The margin of safety as defined in the basis for any technical specification will not be decreased. No USQ exists.

NSM-22333 l This NSM provides a continuous drain for the loop seals on the pressurizer safety valves and replaces the valve internals with new flex-i-disc internals designed to seal on steam. Also, new tied expansion joints were installed on the discharge piping. New insulation for the piping from l the pressurizer nozzle to each of the safety valves was also l installed. The new insulation is qualified for use in 18

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containment and is compatible with the piping material.

Eliminating the water in~the loop seals requires re-analysis of the discharge piping. The loads on the supports will be significantly reduced because the water hammer loads will be

. eliminated. This will permanently resolve concerns with NUREG-0737.

'The safety valves are accident initiators as-addressed in Chapter 15, section 15.6.1. Modifying the internals to seal on steam will not make them more likely to open at design setpoint or experience drift of setpoint under any conditions. Therefore, the probability of accidents previously evaluated in the UFSAR is not increased.

The accident mitigation function of Reactor Coolant System pressure relief is not affected by this modification. The technical specification value setpoint is unchanged at 2485

+/- 1% psig. Therefore, the consequences of accidents

' previously evaluated in the UFSAR are not increased.

. Eliminating the water hammer loads in the discharge piping will make that.part of the system more reliable. Also, the industry concerns related to setpoint shift of safety. valves will be resolved by eliminating the water in the loop seals.

All of the added components are Class 1 and are consistent with the Reactor Coolant System. The existing instrument nozzle used for pressurizer level has not been degraded by the integration:of functions including'a drain path to the pressurizer. Therefore, the probability of a malfunction of equipment different than previously evaluated is not

. created.

No new pipe rupture locations are created by this modification. The: safety valves function is unaffected.

Some Class 1 piping is being added to the drain lines consistent wit the specification for Class 1 piping and components. The manual valves that were normally closed to prevent draining of the loop seals are being removed to

- eliminate the possibility of preventing draining of the water in the new seal on steam design. Therefore, no new malfunctions are created.

Since no new failures are created, the consequences of malfunctions of equipment previously evaluated in the UFSAR are not increased and no new accidents are created.

The margin of safety defined in the basis to the Tech specs is related.t the' confidence in the fission product barriers.

L The' pressurizer safety valves have a safety limit of 2750 psia to ensure the reactor coolant pressure boundary 19

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integrity is not compromised. Since the function of the safety valves is unchanged, the margin of safety between the safety limit and the unknown failure pressure of the RCS is

! not reduced. Therefore, the margin of safety defined in the basis to the Tech Specs-is not reduced. No USQ exists.

NSM-22412 This modification replaces the Honeywell 4400 system presently in use for McGuire Unit 2.. The system is outdated, based on 1970's technology, and has hardware and software maintenance limitations.

The replacement OAC will have many improvements over the existing Honeywell system. New software applications will provide many capabilities beyond the existing OAC such as rending and reports. Analog points coupled with the setpoint review effort will have a unique set of HIHI, HI, Lo and Lo-Lo limits for each mode of operation. Alarming will now have an " alarm destination and alarm priority." Also,

" alarming by mode" will be used tied to eight modes of operation. The six modes defined by technical specifications (modes 1-6) plus additional modes of zero and reduced inventory as defined by NSD-403, " Shutdown Risk Management."

The technical specifications were reviewed for applicability. While many tech spec surveillance requirement are fulfilled via the OAC, there is no tech spec which stipulates limiting conditions of operation requirements for.the OAC. Therefore, no tech spec changes are required.

It should be noted that normal operations without the OAC available may be excessively burdensome in performing certain tech spec surveillance. In those cases, plant shutdown would follow as prescribed by applicable tech specs.

l l The OAC is not uniquely required by any emergency procedures. For any references to the OAC which exist in these procedures, there is an alternate qualified means of obtaining the same information. The " Loss of OAC" procedure was consulted to make this determination. No USQ exists.

'NSM-52384 This modification provides a low level radioactive waste interim storage facility between the pole barn and the low level pump facility and adjacent to the land farm. The 20

L facility will consist of a concrete pad capable of supporting sea-land containers loaded with low level waste containers and the storage modules (i.e. tractor trailers, l cranes, etc.) All wastes will be in a form suitable for I disposal prior to arriving at the storage facility. The storage area will require a concrete slab, approximately 9,600 square feet, surrounded by appropriate security fencing, safety lights, and access road.

The LLRW is-a passive structure and will not be connected to or housed in any safety related structure. There are not impacts on the operation of any systems or structures required for safely operation of the facility in an accident or normal mode. No USQ exists.

NSM-52432/P1 This modification installed redundant badging servers and will install two badging workstations at McGuire. The badging servers will incorporate redundant hardware and connectivity to the Duke network. Two cabinets that will house the components of the badging servers will be located in the Operator Air computer (OAC) room. Each badging server will be powered from its own dedicated power source l from different security panel boards. One server will be I powered from panel board SKSS and the other one will be powered from 1KS. The cabinets in which the new badging equipment is located is powered from lighting panel board LS3A. A badging workstation consists of the following equipment; camera, badging printer, badging encoder, report printer, hand geometry reader, and a computer workstation.

One of the badging workstations will be located in what is now the badge issue / storage area in the south Personal j Access Portal (PAP) and the other badging workstation will ;

be in room MOC-ll86 of Building 7422 (McGuire Office Complex.) The badging servers will be connected to the Duke Network via two routers. A new eight conductor fiber optical cable will be installed. All data connections will be made with fiber optic conductors / cables that meet i existing Duke Telecommunications standards for data '

I transmission.

l The proposed modification only involves plant security related issues. The modification does not alter any plant security related structures, systems or components. This q modification only provides new hardware to support the i personnel access authorization program. The new badging i authorization requirements for nuclear power plants. The data and information that will be contained within the new l badging system will be protected in compliance with the 21 l

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O regulatory requirements of 10CFR Part 73. This modification will not adversely impact the current personnel access authorization program. The authorization of personnel that should not be granted unrestricted access to the protected areas or vital areas of the plant will not be increased by this modification. Subsequent to the modification,-the personnel access authorization program will continue to provide high. assurance that individuals granted unescorted access are trustworthy and reliable, and do not constitute an' unreasonable risk to the health and safety of the public including a potential to commit radiological sabotage. The electrical power for the new badging system will be provided from three different non-safety related panel boards, (SKSS, 1KS and LS3A). The new badging system will also increase the head load in the OAC room. The OAC room is services by the non-safety portion of the VC/YC system and the increase heat' load will have no adverse impact on the performance of the VC/YC system. No USQ exists.

NSM-52465 This modification provides a dose rate monitoring system (DRMS) for the auxiliary building. The system consists of about 240 small dose rate detectors located at points determined by the Radiation Protection Group which feed gamma does rate data through junction boxes to area monitoring equipment and then to the Radiation Protection SCADA computer and the local area network. The detectors are hard wired to the data collection system. The system also will be capable of using wireless remote monitoring units which se radio transmission between the detector and the junction box. The system will allow RP to continuously remotely monitor dose rate at selected locations in the auxiliary building. This data will be used to better plan work for ALARA purposes-and to better understand the effects of plant evolutions on dose rates. Modification 52438 installed about 40 detectors in the auxiliary building to demonstrate the benefits of the system.

The installed dose rate monitoring ystem will have no impact on nuclear safety. The systis will draw power from non-safety station power and will b t.nted such that the cables maintain station separation s A isolation criteria.

The system does not interact with oc.: 9r station systems except that the data from the syster is made available on the station local area network. The detectors are lightweight and do not post a threat to any safety equipment due to weight load or seismic interaction. The cabling to the detectors will be field routed in-a controlled manner to preclude a single failure of the cable violating separa~ ion 22

criteria for.other system cables. Junction boxes and DRMS cabinets will be mounted QA-4 and cables other than the cable to the individual detector will be permanently routed in accordance with applicable station procedures. The system does not interface with any stations systems and does not interfere with any safety functions. Radio transmissions from the wireless remote detectors do not interfere with safety signals or security transmissions. No USQ exists.

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Miscellaneous Changes UFSAR Change - Fuel Offload During Refueling The practice at McGuire is to disassemble the reactor core and move all fuel assemblies to the spent fuel pool during refueling outages. McGuire UFSAR sections 9.1.3.1 and 9.1.4.1 were clarified to more clearly reflect this ;

practice. j l

The spent fuel cooling system is designed to maintain acceptable pool temperatures at all times when fuel is stored in the spent fuel pool. The safety function of the spent fuel cooling system is to ensure that spent fuel stored in the pool is cooled and remains covered with water during all storage conditions. The system is analyzed for the most adverse conditions of cooling water temperature and decay heat load to assure that all storage conditions are bounded. The components are QA-1 and active components are located and powered such that no single active failure will cause loss of cooling from both trains. The fuel storage conditions specifically considered in this review occur i during refueling operation when the full core is offloaded j into the spent fuel pool. A cooling period of 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months is I assumed prior to discharge into the pool. Technical I Specification 3/4.9.3 requires that no fuel movement occur j prior to 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after reactor shutdown. Fuel handling j process and procedures utilized for unloading the core I assure that the core offload is not complete until at least 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months after shutdown and that spent fuel pool ,

temperatures remain with the design limits. Decay heat I loads are conservatively calculated using the BTP 9-2 or ANSI /ANS 5.1 methodology assuming a full pool of discharged assemblies. Although the Oconee fuel stored at McGuire {

combined with previously discharged McGuire fuel has decay 1 heat-loads lower than calculated for discharges from 18 month fuel cycles, the pool is conservatively considered to i be filled with only 18 month cycle McGuire discharges. The pool is qualified for pool temperatures up to boiling for postulated loss of cooling events. The pool cooling system is capable of preventing boiling for any single active failure. Poo1' temperatures are shown by analysis to remain within acceptable limits during single train operation for !

storage. conditions and heat loads other than full core discharge. Single failure is not assumed during full core discharge. Full core discharge occurs for a short period during refueling activities. If loss of a cooling loop caused temperature in the pool to rise above acceptable 1

temperature limits, loss of spent fuel cooling procedures would be used to respond. Heat up of the pool would be within the limits shown in UFSAR Table 9-6. Therefore, sufficient time will be available to restore cooling within the bounds of the recovery operations assumed for loss of all forced cooling. No USQ exists.

USQ Evaluation of the BWU-Z Critical Heat Flux Correlation Application at McGuire and Catawba This evaluation documented the substitution of the NRC approved Z form of the BWU critical heat flux correlation (CHF) for DNB analyses of Mark-BW fuel at McGuire and Catawba. The BWU-Z correlation defines a 20% DNB performance margin increase that is inherent in the Mark-BW Zircalloy Mixing Vane grid design. The NRC reviewed and approved this correlation and the associated statistical core design DNB limit as Appendix C to topical DPC-NE-2005.

The increased DNB performance is implemented by increasing the reference radial peak pin power in the McGuire/ Catawba DNB analyses from the current value of 150 to a value of 1.60. This required updating the Technical Specification bases for McGuire and Catawba. Additionally, two approved topical reports, DPC-NE-2004 and DPC-NE-2007, require updating to reflect the new correlation and the increased peak pin.

The changes were evaluated for their impact on the existing license bases. These impacts include the modifications to cycle specific core design criteria and the effect on safety related components, equipment, instrumentation, and analyses. No USQ exists.

UFSAR Change - Table 5-12 An editorial correction to Table 5-12 which lists reactor coolant pressure boundary materials for Class 1 and 2 auxiliary components, is required in order to update it with as-built information on materials for Motor Operated Valves (MOVs), Air Operated Valves (AOVs) and to correct the misuse of material type for material grade.

Nuclear safety related valves at McGuire are designed and manufactured in accordance with Section III of ASME Code.

Accordingly, pressure boundary materials for Class 1 and 2 components are governed by Section III of the ASME code which mandate acceptable materials. Section 5.2.3.2 of the UFSAR adds requirements that the base materials used in principal pressure retaining applications exposed to reactor 2

4 coolant are austenitic stainless steel, nickel-chromium-iron alloy, or martensitic stainless steel. Since all reactor coolant pressure boundary valves installed.at

- McGuire are built to the appropriate sections of ASME Section II and the materials meet the requirements of Section 5.2.3.2, the materials added to Table 5-12 by this revision do not change the design bases of the plant. The intent of Table 5-12 of the UFSAR is not to control the allowable materials for installation in pressure boundary i L applications, but rather to document the current materials in use at McGuire.

t l The materials added to Table'5-12 as-built the information to reflect what is currently installed in the. plant. In i summary, reactor coolant pressure boundary valves are designed and manufactured to ASME Section III, which is the l

design code for McGuire, and the materials meet the

' requirements of Section 5.2.3.2; therefore, the design bases and safety analysis of the plant are not affected by updating the table with current material applications. No l

USQ exists.

l UFSAR Change - Table 1-6, Variable D-29 l

l The UFSAR includes a listing of the Regulatory Guide 1.97, L . Rev. 2 variables. Component Cooling Water Temperature to ESF System is listed as Variable D-29, which is Regulatory-Guide 1.97 recommended category 2 instrumentation. The

range of this instrumentation is stated as32-200 degrees F.

McGuire's temperature indications for this function are

provided by 1&2KCLP5570 and 5580, which monitor the outlet _

of each train of the Component Cooling water heat exchangers. .The " existing design" for this variable states that the range for the read-out is O degrees F to 367 degrees F. Nuclear Station Modification 1/22119 replaced this. instrumentation. Thermocouple were originally installed and the modifications replaced them with dual-element RTDs. The OAC range was changed to 50 to 150 i ' degrees F to monitor a more narrow temperature span and L improve the accuracy. This improved accuracy was necessary l

to-correctly perform the heat balance testing on the component cooling heat exchangers.

-The UFSAR table listing Regulatory Guide 1.97, Rev. 2 variables was-revised to update the information related to parameter component water temperature to ESF system.

Specifically, the indicated range for Component Cooling heat

. exchanger outlet temperature is different than the range listed in Table 1-6. The installed-instrumentation is acceptable for-.this application. No USQ exists. j I

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UFSAR Change - Table 1-6, Variable D-20 This change revised Table 1-6 Variable D-20 " Auxiliary Feedwater Flow" which describes McGuire's Regulatory Guide 1.97 Post Accident Monitoring Review, Revision 2. Presently the UFSAR table for this variable addresses only the four QA1 loops that monitor auxiliary feedwater flow. The

. technical specification states that there are two channels required per generator. This change added a' statement to show that the intent of Technical Specification 3.3.3.6 is met. The change included the existing non-QA loops that also monitor this parameter. These loops also provide OAC indication. The transmitters are Rosemount 3051C smart transmitters. The receiver gages are Hays-Republic VSA's andoare located on the auxiliary feedwater motor driven pump panels. These loops share the same process connections as.

that of the QAl loops, therefore they monitor the same changes in'the process. Auxiliary feedwater flow indication is considered a category 2 type D variable by Regulatory Guide 1.97 and therefore needs less stringent qualification requirements-than category 1 type A,B or C variables. Since the transmitters share thessame process connections as those of the safety loops, they have been evaluated and approved from a pressure boundary integrity aspect. The impulse lines for the non-safety loops have been fabricated using QA1 materials and are seismically supported. The instrument loops are powered from the 240/120 VAC auxiliary control power system which is a highly reliable battery backed buss with both a normal and alternate supply.

This revision to the UFSAR does not add the potential for any accident or create any additional consequences of an accident other than those previously documented in the SAR.

The' revision.provides clear documentation of the instrumentation installed to meet the intent of the:

technical specifications and that this instrumentation is capable of providing post accident monitoring capability. No USQ exists.

UFSAR Change - Section 9.3.4.2 Changes specified in this section clarify operational alignments and characteristics of the Chemical and Volume Control System (NV). In general the changes are provided to reflect correct system operation and alignment. All of

'these changes reflect current operational practices and alignments and are reflected in various operations l procedures.

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Most of these changes deal with portions of the NV system l

which are not used to mitigate design basis accidents. For example the letdown header, MC filters, demineralizers and boronometer are all contained in portions of NV which are isolated in accidents and serve no accident mitigation

function. The change to this section involving portions of the system which are safety related, involves clarification

[. that the return flow from the seal water heat exchanger is normally to the VCT with an alternate alignment possible directly to the NV pumps suction.

The original UFSAR wording / description indicated that return flow from the seal water HX would normally be routed directly to the NV pumps suction and that an alternate path was a:railable for returning this flow to the volume control tank. In reality, the preferred operating modes is to return this flow to the volume control tank where it is hydrogen entrained (via spray nozzles on each input line that mix the returning flow with the H2 gas in the gas space), combined with letdown flow and finally returned to I the NV pumps suction. This alignment to the VCT is preferred because it entrains H2 back into the solution for transport to the core which aids in controlling H2 created during I radiolytic dissolution of water. I t

Secondly, return to the VCT is preferred because it simplifies system response to a safety injection signal. In that event, VCT outlet isolation valves receive an automatic signal to close. This isolates the VCT from the NV pumps suction as suction is transferred to the FWST. Seal water HX return alignment which is routed directly to the NV pumps suction, would bypass these isolation valves. Because NCP

1 seal return flow is also isolated during a safety l injection, no safety significant consequences would result !

from this alignment (e.g. consequential dilution of injection water); however, small amounts of diverted flow >

from piggy-back operation could reach the VCT via backseated check valve NV-143 under this alignment. NV-143 is checked every refueling outage against gross diversion of flow, per the McGuire In-service Testing Program and ASME Section XI requirements. This testing ensures that the potential alignment of the seal water HX directly to the NV pumps suction (and bypassing NV-141A and NV-152B) is acceptable.

Based on this discussion, the flow path of returning seal water HX flow to the volume control tank is preferred and I

currently used. The alternate return path directly to the NV pumps suction is allowed but of lower operational preference. No USQ exists.

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i UFSAR Change - Section 6.3.2.2.1, 7.4.1.6, 7.4.1.6.1.2.2 and j Table 6-125 These UFSAR revisions correct information to match the design bases plant configuration and to improve clarity.

The proposed changes do not constitute a physical change to the plant, procedures nor methods of operation. The l outlined UFSAR revisions constitute administrative corrections, correct obvious errors and/or enhance clarity.

The changes regarding valve interlocks is supported by existing design documentation (electrical elementaries) and/or station procedures.

The proposed changes do not increase the probability of occurrence, nor the consequences of an event (accident of equipment malfunction) previously analyzed by the SAR.

Similarly, the proposed changes do not create the possibility of a new event (accident or equipment malfunction) not previously evaluated by the SAR. There is no affect on any existing technical specifications. No USQ exists.

Technical Specification Bases Change The bases discussion for Technical Specification 3/4.8.2 was revised to clarify that the 10-day waiting restriction only applies to the-AT&T high specific gravity (HSG) batteries and that the square low specific gravity (LSG) batteries are not subject to the 10-day restriction. This paragraph in the bases is-also revised to indicate what when all of the HSG batteries installed in the 125 VDC Vital I&C system are replaced with square LSG batteries, the 10-day waiting restriction expires. A paragraph is added regarding.the modified performance discharge test. This paragraph states that a modified performance discharge test can be performed in lieu of the performance test specified by specification 4.8.2.1.2.e and also states that the modified performance discharge test is a combination of the performance discharge test and the service test that results in a more conservative surveillance test. In addition to the above changes, this section of the bases was revised to include the battery parameters for the LSG batteries.

These changes to the bases have no affect on fission product

. barriers nor the operating characteristics of the batteries.

The function, performance and operation of the batteries will not be impacted. The battery capacity is not affected.

No design criteria or safety function of any structure, system or component is affected by the change. No USQ exists.

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UFSAR Change - Tables 6-111,112, 113 These changes were made to reflect the actual configuration for NM steam generator, hot leg and pressurizer sampling line containment penetration overpressurization protection.

The UFSAR described the containment overpressure protection for penetrations M235, M309, M335, M338, M340 and M341 as relief valves between the inside and outside containment isolation valves. The actual configuration is spring loaded check valves between the inside and outside isolation valves, with different valve numbers; the relief valves were replaced by check valves by modifications.

There are no safety issues identified with this change. The 10CFR 50.59 evaluations performed by under the modifications. These prior safety evaluations determined that no safety systems would be degraded because functionally the system will remain the same because the check valves will perform the same pressure relieving capacity as the relief valves that they replaced. No USQ exists.

UFSAR Change - Section 9.1 through 9.1.2.4, Section 9.1.5 These changes are revisions based on license amendments 159/141 for unit 1 and 2 spent fuel pool to increase the maximum allowable enrichment of fuel stored in the pools.

Other changes are editorial in nature.

The revisione, are based on approved license change and do not represent a change to the submittals presented to the NRC. By approval, these license change became part of the UFSAR. The change maintains consistency and improve the accuracy of the UFSAR. No USQ exists.

UFSAR Change - Section 1.8.19 and 5.5.15.3 The UFSAR section on response to Three Mile Island concerns (section 1.8.19) and the reactor coolant vent system design evaluation (section 5.5.15.3) include discussions of the design flowrate of the head vent system assuming 100%

' hydrogen is being vented. The supporting calculation was recently found to be in error and the actual flowrate is less than currently reported. The head vent flowrate is controlled by an orifice plate in the 1" line, which is unchanged. The calculated flowrate for 100% hydrogen at normal operating pressure and temperature is approximately 14 cfm. This will allow the entire reactor head to be vented of gas in approximately 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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This UFSAR update does not make a physical change to the station nor the head vent system. NO USQ' exists.

UFSAR Change - Section 5.2.7 and 15.6.3 The UFSAR section of reactor coolant system leakage detection methods (section 5.2.7.1) and steam generator tube rupture accident description (15.6.3.1) include discussion of the steam generator blowdown recycle demineralized effluent radiation monitor and the condenser air ejector radiation monitor for reactor coolant leakage to the main steam and feedwater systems. The main steam line N-16

- monitors, added by station modifications are included in the

' discussion of each of these sections. The main steam line N-16 monitors augment the control. room's ability to detect

and correctly determine which steam generator has a primary Eto' secondary leak. Descriptive information about the main steam line N-16 was added to Table 5-30.

No physical change to the reactor coolant system leakage detection system or the radiological monitoring systems are made. These updates are consistent with the current technical specification and safety analysis. No USQ exists.

UFSAR Change - Table 9-1

-This change reflects the actual fuel pool cooling system component configuration. The UFSAR described an installed fuel pool sump pump in Table 9-1, Spent Fuel Pool Cooling System Component Design Data. This pump, however does not i exist; a portable submersible pump is used for pumping down ;

various' portions of the. fuel pool. !

.There are no safety issues identified with this change. The installed sump pump provides for pumping down various portions of the fuel pool, and.is described as a cast iron pump. Current practice is to use a portable submersible

- pump made of stainless. steel, brought to the fuel pool only l when necessary to drain down a portion of the pool. The

. pump is only used in areas of the pool where no fuel is ;

directly stored, so it.would'not be' operated over the fuel -

The use of a stainless steel pump involves material that is )

more corrosion resistant to boron' solution in the fuel pool than cast iron. No USQ exists.

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UFSAR change - Sectica 5.2.1.10 This change revises the affected UFSAR section to accurately i reflect actual operating practices. The actual operating practice maintains licensed basis requirements regarding shutdown margin while maintaining some flexibility not reflected in the current wording. The affection section is regarding reactor coolant system pressure boundary and stress analysis. Reactivity control and shutdown margin was not the intended focus of this section.

This change reflects the method by which the operators maintain compliance with shutdown margin during plant cooldowns. The required shutdown margin as assumed in the SAR for all. operating modes is ur af fected by. this change.

By preservation of shutdown margin tbo integrity of the fission product barriers are e.esured under all operating and accident conditions.

This change does not alter.the method by which the operator is manipulating any plant structures, systems or components as compared to the current licensing basis. This change is not operating any plant structure, system or control outside of current licensing basis assumptions. No USQ exists.

UFSAR change - Section 15.6.3 This revision deletes an assumption that was made to !

ca);.u' ate the activity release and offsite dose for the I poo diated steam generator tube rupture event. The steam i generator blowdown rate of 50 gpm was deleted. This change i does not result in a change in how the facility is operated ;

nor how the facility mitigates the consequences of an ;

accident. There are no changes to any structure, system or ;

component. No USQ exists. ;

UFSAR Change - Section 8 1

This change primarily makes editorial changes to the identify changes to Duke Energy's transmission system. Two of the stability curves have been updated based on the '

addition of the Antioch Tie Station. This change has no

. impact on safety related equipment, does not increase the consequences of a malfunction. The margin of safety 1 described in the Technical Specifications is not reduced.

No USQ exists.

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l UFSAR Change - SNACORE Software Upgrade This change is associated with a software upgrade; no safety

. significant material changes are involved. The offline software used to calculate the core power distribution and to verify the related Technical Specifications, SNACORE 3.2, has been upgraded and consolidated into two codes: comet 01 and burn 01. The new software incorporates several enhancements, relative to human factors, but the basic methodologies are functionally equivalent. The new codes are certified QAl per NSD-800 and yield the same results are the previous software. This software is not part of a SSC important to safety and does not directly affect any SSC.

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.

No USQ exists.

UFSAR Change - Section 9.1.3.2.6.2 This change reflects actual fuel pool cooling system instrumentation configuration for the purification loop flow 1 alarm. The description is the UFSAR is that high or low flow through the loop is alarmed through the computer. The actual configuration is that only low flow is alarmed.

There are no safety issues identified with this change. The purification loop is manually aligned with flow controlled with a manual throttle valve. No automatic actions could change the flow which would necessitate a high flow alarm.

Low flow is alarmed, which would be indicative of a loss of pumped flow through the loop, or from low flow due to filter or demineralized clogging in the loop. Flow indication is also provided locally by a gage. High flow is alarmed for the heat exchanger flow in the cooling subsystem, which would be indicative of hiah flow in the entire system, which would include the purification subsystem.

There are no design basts accidents'regarding the KF system or the KF purification subsystem described in the UFSAR.

There is a loss of forced cooling by the KF cooling subsystem described in the SAR; however, the activity has no

affect on the cooling' subsystem, nor'could it have any affect.on the purification subsystem that would increase the probability of a loss of the cooling subsystem. The loss of the cooling subsystem could be from either a loss of one of the KF cooling pumps or from a loss of cooling water to the KF heat exchangers, neither of which are affected by a change in the purification loop flow alarm indication.

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4 The loss of forced cooling would be indicated by low heat exchanger or demineralized low flow alarm, and ultimately, a spent fuel pool temperature alarm. Not having a

' demineralized high flow alarm would not affect the consequences of an accident. A different type of accident is not possible since the indicators do not provide any control functions, and cannot initiate an accident of a different type. Having nu demineralized high flow alarm would not affect operator response to a loss of forced cooling. The possibility of any new malfunctions are not

- created. No USQ exists.

UFSAR Change - Section 9.4.5.2 This UFSAR section contains factual errors that state that both upper and lower containment minimum air temperature are maintained at 60 degrees F with heaters. This statement is incorrect since there are no heaters in the upper of lower containments for the purpose of maintaining containme c air temperature. Neither the VU or the VL system utilizes heating coils in any of the air handling units. There is no design documentation specifying a heating coil requirement and operating experience has shown that the heating coils are not necessary. The minimum temperature criteria has been satisfied due solely to equipment loads within the building.

i

'Neither the VU or VL systems are safety related or accident )

initiators. There are no affects on any assumptions based !

on containment temperature in accident analysis. Removing j references to heaters in both upper and lower containment l l could not initiated an event of any type nor impact the ;

environmental qualifications of equipment inside l containment. Ambient temperature ranges in upper and lower containment will not change from Technical Specification and UFSAR requirements. The margin of safety is not reduced.

No USQ exists.

DPC-1167.01-00-0001 This evaluation involves the changing of coating materials in containment. The new coating materials have been tested

- to the requirements of ANSI N101-2 and comply with Regulatory Guide 1.54. The new coating materials have been qualified over the existing Mobil /Valspar coating systems and as a new coating system for-radiation exposure,

' pressure, temperature, and water chemistry exposure during a l design basis. accident. The coating materials will provide l the same level of corrosion protection and will function as designed. No new failure modes or operating characteristics l

11

are created. The function of equipment in containment is not affected. The new coatings do not affect any safety limits, setpoints or safety system parameters. No USQ exists.

VN-22121A VN-22121A is to delete reactor coolant loop purge flow l

transmitter (NMDT5260), volume control tank purge flow transmitter (NMFT5110), and the signal comparator (located in NM sampling room valve operating panel) receiving inputs from the aforementioned flow transmitters. These devices serve a leak detection on the sample line from the reactor coolant system. If the signal from NMFT5260 is greater than the signal from NMFT5110 (indication of sample line leak),

the comparator sends a signal to the containment isolation valve NM26B to close.

The lead detection equipment has never functioned properly and resulted in containment isolation valve 2NM26B undergoing frequent and unnecessary cycling. As a result of the cycling, this leak detection equipment was taken out of service. This automatic leak detection service is not required or mandated; therefore, this equipment will be deleted. This modification permanently removes the signal comparator, two flow transmitters, and associated wiring.

The deletion of this malfunctioning equipment will remove automatic leak detection capabilities; however, Chemistry personnel use procedure that provide directions for manually operating the nuclear sampling system which includes containment isolation valve 2NM26B.

The nuclear sampling system still performs its design purpose to provide a means of collecting liquid and gas samples for laboratory analysis from various systems inside and outside of containment in a safe and shielded location.

No USQ exists.

UFSAR Change - Section 9.2.4.2 The changes to this section are editorial in nature and do not affect the function or operation of any structure, system or component. The changes are a clarification of the terminology used in UFSAR Tables 9-22 and 9-23 and the description of Table 9-22. The component cooling system (KC) will continue to operate in the same manner. No USQ exists.

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Inclusion of K(BU) in LCO Monitoring of Fq MkBW LOCA limits are a function of axial height and burnup.

These effects are accounted for by applying multipliers to the base LOCA limit of 2.32 to account for axial (K(Z)) and burnup (K(BU)) dependent effects. The burnup dependency is included in the calculation of monitor factors used in the power distribution surveillance monitoring specified in the Technical Specifications; however, the explicit monitoring of this burnup dependence does not exist for LCO monitoring.

This 50.59 justifies an update to the Core Operating Limits Report (COLR), which is referenced by the Technical Specifications, to include burnup dependent effects in the LCO monitoring of Fq.

A limit of Fq is established to ensure that emergency core cooling acceptance criteria are not exceeded during a loss of coolant accident (LOCA). For MkBW fuel the LOCA Fq limit is equal to 2.32 from 0 to 46.354 GWD/MTU and is then reduced linearly from 2.32 at 46.354 GWD/MTU to 1.91 at 60.000 GWD/MTU. Measured power distribution results are compared to the Fq limit to ensure that initial conditions of the LOCA analysis are met. Reduction of the Fq limit input for LCO monitoring does not change key parameters assumed in any accident evaluations, nor does this change affect the response of systems, structures, and components important to safety. The effects of this change on power distribution monitoring of the core are negligible, since this change does not affect the minimum margin to the Fq limit calculated for the core. High burnup assemblies are not limiting, therefore, the margin of safety is not reduced. No USQ exists.

Potential for Water Hammer in NS Heat Exchangers The NS system is aligned to the refueling water storage tanks during the injection phase of an accident. After depletion of the RWST, the NS pumps are secured and the suction is realigned to the ECCS sump. Due to a lack of a check valve between the NS heat exchangers and the ECCS sump, a potential for a water reversal to the ECCS sump exists while the NS pumps are secured. Check valves downstream of the NS heat exchangers will prevent water from continuing to flow to the sump; however, water column separation in the heat exchangers is possible due to the elevation difference between the sump and the heat exchanger high point. The water column separation (low pressure steam void) can lead to a water hammer within the heat exchangers when the NS pumps are restarted. This can occur any time the NS pumps are stopped and restarted while aligned to the 13

ECCS sump. Based on a review of the water hammer potential and analytical evaluation of the several possible

! mechanisms, it is concluded that the NS system is not degraded by potential heat exchanger voids during swapover.

The results of the water hammer analysis have shown that the l water hammer will remain localized to the heat exchanger tubes or the shell. The pressure wave and resulting stresses-in the heat exchanger components were calculated and found to be well within the ASME code allowable limits.

t The presence of a void in the NS heat exchangers is only

possible in the ECCS sump recirculation mode of operation.

l No systems, structures or components are degraded. No

! piping or equipment code allowable stresses are exceeded.

l The NS system is not considered an accident initiator in l analyzed accidents. No USQ exists.

l UFSAR Change - Section 7.6.7.1.10 This change. reflects the actual fuel pool cooling system instrumentation for heat exchanger temperature and pump discharge pressure. The UFSAR description states that one temperature element is provided at the outlet of each fuel pool heat exchanger for the local alarm and indication. The actual configuration is a temperature test connection at the inlet and outlet of each heat exchanger for local i indication, and one pressure detector at the discharge of i each fuel pool pump for local indication. I l

There are no safety issues identified with this change. The heat exchanger temperature test points are used to check the temperature at each heat exchanger inlet and outlet. This

. is to allow for monitoring of heat exchanger performance l over long periods of time. The pump discharge pressure l instruments provide local indication of each pump's discharge pressure. KF system parameters are adequately monitored by.other system indications. Indication of loss of pressure bv the pumps would be determined by other indicators, such as low heat exchanger flow alarms, low demineralized flow alarms and ultimately, a high spent fuel ,

pool temperature alarm. No USQ exists. i UFSAR Change - Section 9.5.1.4 j l

Operations identified a discrepancy between station j procedures and the UFSAR in regard to testing of water spray (deluge) systems. The resolution of this discrepancy was to revise the UFSAR to agree with current practices. A review of correspondence determined that the testing frequency had been previously. discussed with the insurer (NEIL) and l

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4 approval was granted to change from " annual" to "during refueling (maximum 18 months)." The design and operation of the water spray systems has not changed, only the testing frequency.

The change to the UFSAR to reflect this testing during refueling does not involve any USQs. The approved fire protection program is maintained and the changes do not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. No USQ exists.

UFSAR Change - Sections 6.2.2.1, 6.2.2.2, 6.2.2.6 and 6.2.2.15 The UFSAR needs to be updated to incorporate the Operating l 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 l the UFSAR is greater than most of the systems and provides l information with respect to the reaction of the ice

i. condenser floor to wall panel and floor defrosts. Operating l Experience has shown that these assumptions are no longer l valid.

l Related to the OE, McGuire never used floor defrosts with j wall panel defrosts as described in the UFSAR. McGuire maintenance practices rely on mechanical methods for ice removal from ice baskets and flow passages. Defrosts, when ;

l used, have been localized and limited in time as to prevent '

l creating excess water-in the ice condenser. A %" to %" l layer of ice 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 UFSAR describes 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 of ice from the ice condenser and_the practice of leaving a layer of ice on the. ice condenser floor does not affect any ice condenser or_other SSC's or any safety functions. These are merely preferred methods of ice maintenance. No USQ exists.

UFSAR Change - Table 7-14 The UFSAR includes a listing of control room indicators and/or recorders available to the operator to monitor 15 l

L______-_____

t significant unit parameters during normal operation.

Auxiliary feedwater flow is listed as a parameter under the feedwater and steam system section. There are four QA condition 1 auxiliary feedwater flow indication loops, one per steam generator, monitoring flow from all auxiliary feedwater pumps to each steam generator. Each flow indication loop is scaled for 0 to 600 gpm. This flow range is adequate to monitor auxiliary feedwater flow to each steam generator in that design flow to each generator is between 225 and 450 gpm. These instrument loops are required per Regulatory Guide 1.97, Rev. 2, and their existing design is addressed in UFSAR Table 1-6, " Regulatory Guide 1.97, Rev. 2 Review."

UFSAR Table 7-14 incorrectly states the. indicated flow range for auxiliary feedwater flow as 50 to 800 gpm. This UFSAR change corrects this range to O to 600 gpm.

This UFSAR change is editorial and is made to reflect the current as-built configuration of the Auxiliary Feedwater System. The revision does not make a physical change to the plant. No USQ exists.

UFSAR Change - Section 7.6.9.1.2.1 This change will reflect the actual control room temperature instrumentation for the FWST water temperature. The current description in the UFSAR is that instrumentation is provided in triplicate to allow the operator accurate indication of the FWST water temperature. The actual configuration is a single temperature indicator (gage) with three temperature elements feeding computer alarms.

There are no safety issues identified with this change. The single visual gage temperature indication is used in conjunction with the computer points for high, low, low and emergency low temperature, fed by a two-out-of-three logic from the three temperature elements. Therefore the operator has alternate means of determining increasing or decreasing temperatures in the FWST. No USQ exists.

Compensatory Measures for Operability Evaluation of PIP 0-M97-3286 and procedure changes to OP/0/A/6550/11 and PT/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 Keff remains less than or equal to 0.95. These l new limits are more restrictive than the current limits specified in Tables 3.9-1 and 3.9-3 of the McGuire Technical 16

Specifications. These new limits are to be implemented by changes to procedures OP/0/A/6550/ll, 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 function, the Keff 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 will continue to ensure that Keff 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.

No design criteria or safety functions of any SSC is affected by the compensatory actions and procedure changes.

No USQ exists.

UFSAR Chapter 16, Selected Licensee Commitments 16.9-8 (MM-9799)

This modification raised the Hi alarm setpoint for 2WZLS5060 from 731' to 733'. This in turn 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 evaluated for hydrostatic loads up to grade (elevation 760'). Therefore, there is no concern with the loading on the Unit 2 reactor building wall. This well was removed from Technical Specifications but maintained in the SLC in order to provide an alert to such events as a pipe leak / break or failure of the underdrain grid. Pipes in the area that are not under pressure would be the RN and 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 of WZ sump pump starts indicates less than 30 gpm being contributed to the sumps from the underdrain grid. Breaks from 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. The increase has been attributed to construction in the area and increase in operating level of Lake Norman. No USQ exists.

UFSAR Section 2.2.1 UFSAR Section 2.2.1 is being revised to clarify aircraft flight paths and to make editorial revisions to the civilian airport information. The plant was evaluated in SER 2.2 and 17

it was concluded that aircraft activity at Charlotte Douglas Airport would not adversely affect safe operation of McGuire. UFSAR Section 2.2.3 states that McGuire is not in a FAA established flight path (Military or Federal airway).

UFSAR Section 3.5.1.4 states that aircraft are not credible missiles due to flight patterns close to the station.

Aircraft in class B airspace are controlled by the ATC but there are no restrictions on the ATC from directing aircraft over the station. Air traffic over the site from 3,600 feet mean sea level'(MSL) to 10,000 feet MLS is class B airspace and is regulated by air traffic control. Normal flight patterns in the vacinity of the site, for aircraft going into and out of Charlotte airport, would be at or above 3,600 feet MSL for arrivals and at or above 5,000 feet MSL for departures. Aircraft activity over the site was evaluated in 1994 for the IPEEE study, which used guidance from the Standard Review Plan, NUREG-0800. This report found that the probability of a plane crash is 1.3 x 10-8 for annual aircraft operations of 531,415, which is less than 1 x 10-7. As outlined in the standard review plan. These UFSAR changes are editorial in nature. No USQ exists.

UFSAR Change - Section 5.2.2.3, Tables 5-10 and 5-47 The UFSAR on Overpressure Protection is revised to update the discussion of the Low Temperature Overpressure

! Protection (LTOP) System. Changes to the section include adding reference to the approved use of ASME code case N-514, deletion of Technical Specification figures 3.4-4 and 3.4-5 which were eliminated following approval of unit specific Technical Specifications. Correction of the Unit 2 LTOP enable temperature and correction of the PORV stroke l time. No physical changes to the station were made by this update. No USQ exists.

l UFSAR Changes - Accuracy Review Project The activity being evaluated involves correcting inaccurate technical information or deleting, adding or making changes !

to verbiage to enhance or clarify information and/or concepts contained in the Duke Power McGuire Units 1 and 2 1 Updated Final Safety Analysis Report (UFSAR). Corrections ;

or changes are limited to those that; l) 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., supporting) design and/or !

operation-related documents, 3) do not affect any operation-related information and 4) do not affect the Technical Specifications for Duke Power McGuire Nuclear Units 1 and l

. 2.

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I The corrections or changes do not involve any changes to the l 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 t

involved by the corrections or changes.

l Correcting inaccurate technical information or deleting, adding or making changes to verbiage that enhances or j clarifies information and/or concepts contained in the Duke Power McGuire Nuclear Units 1 and 2 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. Also, no changes to the Duke Power McGuire Units 1 and 2 Technical Specifications are required. Therefore, corrections or changes can be implemented without prior NRC approval. No USQ exists.

l- .UFSAR Change - Sections 5.1 and 9.1.4.2.1 These UFSAR changes address reactor disassembly and assembly. The majority of the changes rearrange existing descriptions in the correct sequence of events and added additional information which is explained in~ detail. Section 5.1, was changed to address the issue of lifting the reactor hcad without concurrently flooding the refueling canal. In l Section 9.1.5.2.1, the general description of reactor i

disassembly was out-of-order and several activities omitted that were of equal importance. Other clarifications were i made including clarification of the description of the reactor vessel level indication system (RVLIS). These L changes to the UFSAR improve the accuracy of descriptions of l the actual events and sequence of reactor disassembly.

Activities for reactor disassembly for refueling have a low l probability to impact loss of core cooling, cause fuel l damage, or impact reactivity. Equipment associated with

disassembly does not interface with or depend on any safety SSC. There are no non-conservative or unanalyzed conditions that impact reactivity management. No USQ exists, l

UFSAR Change - Sections 7.6.8 - 7.6.8.2.2 and t

9.1.4 - 9.1.4.5 The technical changes to these sections can be summarized as follows:

I

1) Clarification that some fuel handling equipment interlocks j are not " redundant." In each of these cases it was determined that the interlock still met the single failure l l

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criterion and therefore provides an adequate level of protection to the fuel and equipment.

2) Inclusion of manipulator crane " slow zones" and underload features in the interlock discussions. Inclusion of these features have always been a features of the cranes-but were not discussed in the UFSAR. These features add additional levels of protection to the fuel and equipment and were not previously accounted for in the UFSAR.

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3) Deletion of statement that the cranes have features that prevent raising fuel assemblies to within 10 feet of the water level. The cranes do'have features that prevent raising the fuel assemblies to above a safe height, although this height may not always provide a full 10 feet of water adequate shielding is provided.

Although is was determined that some of the equipment interlocks previously identified as " redundant" were not

" redundant," in each case the interlocks meet single failure criteria and therefore provide the same level or protection.

Inclusion of " slow zone" and underload interlocks provide additional level of protection against fuel _ handling accidents not previously discussed in the UFSAR. The crane features ensure adequate shielding of fuel assemblies, personnel protection, and are not related to Chapter 15 accidents. No USQ exists.

UFSAR Changes - Nuclear Service Water Sections Section 9.2.2 of the UFSAR discusses the original design function and operating parameters / requirements for the i Nuclear Service' Water system and ultimate heat sink.

l Although the information contained is reasonably accurate, it does not reflect revisions incorporated into important RN I system calculations. These calculations provide the analysis and methodology for determining heat loads, system

. flow rate, and standby nuclear service water pond thermal performance.

The function and design of the Nuclear Service Water System or Standby Nuclear Service Water-Pond are not changed. The activities discussed only refer to system operation and predicted. performance under bounding conditions. The revisions detail system response under worst case design l accident conditions. The postulated worst case accident is ;

~a large break LOCA on one unit with a controlled shutdown of ;

the alternate unit simultaneous with a loss of off-site power on both units. The UFSAR revision details the expected system response with regards to heat input, flow 20 1

rate requirements, and SNSWP thermal performance. Nuclear Safety related equipment operation is not affected in any manner. All system accident flow rate changes are in accordance with approved NRC analysis methodology. No USQ exists.

I UFSAR Change - Section 6.2 This revision adds the minimum calculated hydrogen skimmer system flows required to maintain hydrogen concentration below 4%. The minimum flows were not previously included.

These calculated minimum flows are supported by an approved calculation. The appropriate references to the minimum flow l have also changed.

l

! The normal flows were originally shown on the flow diagram which was replaced by the summary flow diagram. These flows were incorrect because they added the total of the single fan flows to get a two fan flow. Since portions of the system duct is shared, the total flow cannot be added. This was corrected to show single fan flows and the data was adjusted to take into account typical flows based on field measured fan performance. All flows exceeded the minimum j required. The data was collected using approved test procedures.

The hydrogen mitigation system is an accident mitigation system, versus an accident initiator system. The changes only reflect previously evaluated data which was determined to be acceptable. -No new failure modes are introduced. No structures, systems or components are changed. No operating parameters are changed. All equipment important to safety is operated within design parameters. No USQ exists.

UFSAR change - section 3.9.2.7 UFSAR inconsistencies were identified concerning organizational structure that is no longer reflective of the current corporate organization. Specifically, Construction, Design Engineering, Nuclear Production, QA and CMD-N Mechanical Technical Support organizations are stated to be involved in monitoring, review, inspection, and approval of various aspects of field routed piping and tubing. These organizations no longer exist by the same name within the company, and it is inappropriate to refer to them in describing the program of field routed lines as a process that is currently used or would be used on an ongoing basis. j The wording in this section is changed to remove l organization references. A statement is added to the !

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section to describe the current and ongoing method of control of field routed lines by design specifications. A reference to frequent Engineering visits to the site is changed to apply to engineering walkdowns, and various monitoring previously described by organization was revised to reflect functional monitoring that is on-going. These revisions are clarifying in nature. The corrections are based on previous evaluated plant changes which did not get incorporated. No physical changes to the plant are made.

No USQ exists.

Editorial /non-technical changes to the COLR The conclusion of this 10CFR 50.59 evaluation is that administrative and non-technical changes to Topical Reports and the COLR (Core Operating Limits Report) can be performed without prior NRC approval and do not constitute any unreviewed safety question provided that the intent, interpretation or understanding of the technical content is not altered.

The types of changes addressed in this evaluation do not involve the modification of equipment important to safety, or impact the function, design bases performance or operation of safety related equipment and components. No USQ exists.

Revision la of DPC-NE-1003-A As part of Duke's effort to maintain updated and accurate topical reports, a review of DPC-NE-1003A was conducted to identify any changes that'have been made since the original approval in December, 1986. Only one change was identified that had not previously been addressed under 10CFR 50.59.

The change evaluated is related to the iteration technique used to predict the critical height in the rod swap methodology. The change evaluated is simply a replacement of a manual iteration technique, which had a subjective convergence criteria, with a computer automated iteration technique with a very tight mathematical convergence

. criteria. Determining the critical heights in the rod swap calculations simply involves iterating on the reference bank position using a core simulator until the Keff matches the

.Keff for the reference bank only inserted (base Keff).

Given'the same convergence criteria, the iteration technique used to determine predicted critical heights is independent of the final answer.

This change does not involve the modification of equipment important to safety, or impact the function, design bases 22

O performance or operation of safety related equipment and i components. The SER issued for DPC-NE-1103A does not '

address, nor apply restrictions on the iteration technique used to calculate predicted critical heights. NO USQ exists.

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Minor Modifications NGNN-7067, 7068 i These modifications installed an overflow line on the 1A, 1C, 2A, and 2C condenser boot seal trough. The line from each condenser was field routed to the nearest floor drain with a shutoff valve at (within easy operator reach) the floor drain.

These modifications do not increase the probability or )

consequences of an accident or malfunction of equipment !

important to safety evaluated in the SAR. .The modifications l l

correct a housekeeping problem and possible personnel safety l concern due to the spillage of water form the boot seal i troughs. The piping and boot seal troughs are non safety related and have no effect on the probability of any accident evaluated in the SAR.

The modifications have no effect on the operation of the l turbines, condenser, condensate system,-or any equipment l important to safety. The only purpose of the modifications is

! to provide an orderly means of removal of excess water from the L boot seal troughs during certain operating conditions. The I level or presence of water in the boot seal troughs has no l effect on any. equipment important to safety. No functional l . changes to the plant equipment / systems operation are made. No l USQ exists.

MGMM-7096,7125,7254,7255 l- Sight glasses were added to unit 1 and 2 'A" and "C" reactor l coolant pump motor lower bearing cooler drain piping and l Reactor Building Component Cooling System (KC) drain header to I determine drain valve seat leakage'and input to KC Drain Tank from the Reactor Building. These sight glasses were installed into the non safety related portion of the Component Cooling l

System. These modifications do not increase the probability of an accident evaluated in the SAR. No USQ exists.

t MGMN-7492 This Modification makes the jumper installed in cabinet TB-1672 permanent. Temporary mod number 6475 placed a jumper between 1

C_ __ __ _

1

)

l terminals B-12 and B-14 in cabinet TB-1672 to bypass the function of pressure switch 2VGPS5430 which was installed under NSM MG-22279/P2. The pressure switch controlled solenoid valve 2VGVA0093.

'There are no new safety concerns introduced for the Diesel Generator Engine Starting Air (VG) or Instrument Air (VI) systems:since VG will only be aligned to VI following the manual positioning of 2VG95 and 2VG96 under abnormal procedure AP/2/A/5500/22. Monitoring of Diesel Generator Engine Starting Air System pressure is adequate - time to ensure the systems can be isolated if needed. The pressure switch will be left in place and does not pose a safety concern. No USQ exists.

MGNM-7493 This modification makes the jumper installed in cabinet TB-1673 permanent. Temporary mod number 6478 placed a jumper between terminals B-13 and B-14 in cabinet TB-1673 to bypass the function of pressure switch 2VGPS5440 which was installed under NSM MG-22279/P2. The pressure switch controlled solenoid valve 2VGVA0094.

There are no new safety concerns introduced for the Diesel Generator Engine Starting Air (VG) or Instrument Air (VI) systems since VG will only be aligned to VI following the manual positioning of 2VG97 and 2VG98 under abnormal procedure AP/2/A/5500/22. Monitoring of VG pressure is adequate - time to ensure the systems can be isolated if needed. The pressure switch will be left in place and does not pose a safety l concern. No USQ exists.

l MGNM-7563 This modification replaces valve 1 nil 75 with a new valve. This

- valve has been evaluated and the applicable NI (Safety L Injection) system calculations have been reviewed and reconciled for the changes. No new failures are introduced or operating characteristics ~of the ECCS (Emergency Core Cooling System) are changed. The control logic is unaffected, and there are no appendix R concerns. Therefore, there are no new malfunctions or accidents of a different type than evaluated in the SAR. No USQ exists.

MGNM-7750,7819 i

These modifications replace or remove Carbon Steel piping which !

supplies and discharges Conventional Low Pressure Service Water ;

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(RL) to Service Building Air Handling Units (SB AHUs) SB-7, SB-8, SB-12, and SB-13. This piping has become plugged or flow has become severely restricted. The piping to be replaced includes all piping 2.5 inches in diameter or less and the 4 inch supply header downstream of valve RL-55 if possible. The replacement piping will be stainless steel pipe which meets the service requirements for class G or class H. In addition a new 4 inch header with new valve 2RL-307 will be added from the 6 i inch RL header near valve RL-196 to provide an alternate supply !

to the SB AHUs and to take advantage of cooler water from this !

part of the system. Drain lines will be changed from % inch piping to 2 inch piping to improve the ability to clean this l system. MGMM-7750 will be implemented first. This i modification installs the new 4 inch header and valve RL-307 with a blind flange downstream of the new valve. It also installs a blind flange downstream of 1RL-55 and a blind flange with new valves 1RC891 and 1RC892 in the RL return to the Condenser Circulating Water System (RC) via valve 1RN1078 in the RN (Nuclear Service Water) system. This modification thus isolates the piping to be replaced in MGMM-7819. The blind flanges may be removed when MGMM-7819 is implemented.

l The system is not safety related and does not perform any functions important to safety. The design function and capacity of the RL System is not changed by this addition. The RL system will serve the same functions in the same manner as before. No USQ exists.

MGMM-7807,7808 Instrumentation for monitoring the heater bed temperature of the Electric Hydrogen Recombiners required frequent display tube replacement. These tubes and the instrument itself are obsolete. A functionally equivalent device was selected for installation. The new device is to be located in the same panel using existing sensors and cabling. The instrumentation is not used for operation of the Recombiner and will continue to be considered non-nuclear safety related. No USQ exists.

MGMM-7852 This modification adds equalization lines to the bonnet of valves 1ND0058A, 1NIO136B and 1NI0183B (ND- Residual heat

' Removal System, NI- Safety Injection System). These valves are split-wedged or flex-wedge gate valves and have the potential to trap high pressure water and air between the disks. The trapped water and air are postulated under certain conditions to cause the disks to seat so tightly that the valve operator would be unable to open the valve when necessary. The equalization line consists of a % inch piping line tapped and I

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welded into the valve body and welded to a % inch class B

! packless valve (1ND-105 for valve 1ND-0058A, 1NI858 for 1NI-136B and valve 1NI-859 for 1NI-0183B) which is fully qualified for the service conditions of this application. Equalization valves 1ND-105, 1NI-858 and 1NI-859 will be normally open.

These manual valves are instrument valves manufactured by Dragon. The installation is such that they allow depressurization of the volume in the isolation valve bonnet but should prevent high flow to the bonnet area from the Residual Heat Removal (ND) or Chemical and Volume Control (NV) system. The downstream side of the equalization valve is %

inch piping which is routed back to the piping system. The equalization line is thus configured to maintain the pressure in the bonnet at a pressure no more than the operating pressure in the piping when the system is in its normal alignment. The motor operators for the valves are set to properly open and close the valves for these conditions. The addition of this equalization piping and valve has negligible impact on the 8 l and 12 inch ND and NI valves and they remain fully qualified.

Neither the containment isolation function nor the ability to isolate ND flow from the NV and NI system is degraded by this modification. The valve stroke time and interlocks are not adversely affected by this modification. This modification L

ensures that differential pressures across the valve disk are always low enough to allow the valve operator to open the valve as required. No USQ exists.

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L This modification adds equalization lines from the bonnet of valves 1NS0001B, 18A, 1FW0027A, 1NV0221A and 1NV0222B (NS-Containment Spray System, FW-Refueling Water System, NV-Chemical & Volume Control System), to the downstream side of the valve. These valves are flex-wedge or splitwedge gate valves and have the potential to trap high pressure water and air between the disks. The trapped water and air are postulated under certain conditions to cause the disks to seat l so tightly that the valve operator would be unable to open the l

! valve when necessary. The equalization line consists of a %

l inch piping which is routed back to the valve on the downstream L

side (toward the pump). The equalization'line is thus E configured to maintain the pressure in the bonnet at a pressure l no more than the operating pressure in the pump suction piping.

E The equalization valves act to prevent overpressurization'of the bonnet in normal operating alignments. No USQ exists.

MGMM-7868 This modification adds equalization lines from the bonnet of valves 1NS0038B (NS- Containment Spray System) and 43A to the l

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upstream side of each valve. These valves are flex-wedge gate L valves and have the potential to trap high pressure water and j air between'the disks. .The trapped water and air-are l postulated under certain conditions to cause the disks to seat L so tightly that the valve operator would be unable to open the valve'when necessary. .The equalization line consists of a %

inch piping line tapped and welded into the valve body and l welded to a % inch class B packless valve which'is fully qualified for the service condition'of this application. The equalization' valves will be normally open. The downstream side I l of the equalization valve is normally open. The' downstream I side of the equalization valve is % inch piping which is routed

! back to the piping.toward the pump. The equalization line is thus configured to maintain the-pressure in the bonnet at a pressure no more than the operating pressure in the NS or ND l (ND Residual Heat Removal System) pump discharge piping. No USQ exists.

NGMM-7893 This modification will replace an existing door on the control L room HVAC common. plenum with a new door that will be fabricated l in accordance with MGMM-7893. The new door will provide a l better sealing mechanism than the existing door. The main design changes to improve the sealing mechanism for this door are 1) Improvements to the hinges for the door; 2) Use of four. bolts (at each corner of the door); and 3) Improvements to the door gasket.

The existing door utilizes a standard industrial door hinge, while the new door will utilize a slotted door hinge. This will allow the new door to move back and forth, perpendicular to the door frame before opening and closing of the new door.

l At each corner of the door, a bolt and nut arrangement is

! provided. During closing of the new door, the four bolts will L be hand tightened. A new door gasket with a larger sealing surface (approximately % inches) will be provided.

l This MM does not affect the design function or operation of the L common plenum access door. The replacement door will contirue

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l l to provide access to.the common plenum ductwork for such things

! as change out of the air filters. The purpose of this MM is to l . enhance the sealing. mechanism for the door. No other ;

structures, systems or components will be affecced by this modification. No USQ exists.

MGNN-8033 This modification blocks channel 3 directional ground overcurrent relay on switchyard buslines 1A and 1B. Channel 3 I i

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relay trip logic will be removed from the switchyard protective relaying circuits. There are a total of four relay channels.

Therefore, with one channel blocked, the scheme will change from 2 out of 4 to 2 out of 3 logic.

The switchyard protective relaying subsystem still performs its

! design purpose. The switchyard protective relaying subsystem provides equipment protection and power system stability by promptly and reliably removing from service any element of the system when the element experiences a fault or subjects the system to an abnormal condition that may prove detrimental to the effective operation of the power system. No USQ exists.

MGMM-8037 Exterior Fire Protection System valve 1RY-14 is a 125 lb class, cast iron, manually operated (handwheel), globe valve, used to isolate the main fire pump header from the test header. The nozzles in the test header are also isolated by sindividual isolation valves; so 1 RY-14 serves as double isolation for the nozzles, and allows maintenance of the downstream valves. The existing valve is a Walworth globe valve which is not UL listed. It has a very large leak through the seat, and cannot be used to isolate, or even restrict flow. No parts are available to repair it; nor is an exact replacement valve available.

The modification will replace the existing valve with a Stockham brand gate valve. The new valve is UL listed, is also a 125 lb class, cast iron, manually operated (handwheel) valve, and was designed and built to the same standard as the existing valve, ANSI B 16.10. No USQ exists.

MGMM-8070,8071 These modifications are to change the tap settings on the main step-up (MSU) transformers to correspond to increased system voltages on the 525 kv and 230 kv grids. MGMM-8070 covers changes to the 230 kv grid and MSU transformers for McGuire Unit 1 while MGMM-8071 covers changing MSU transformer 2B for the 525 kv grid changes tied to McGuire Unit 2. Changes to MSU 2A are done separately from 2B since the associated bus line must be out of service for the change. The change to MSU transformer 2A is done under MGMM-8243.

The main step-up transformers step up the 24 kv power generated by the main generator of the unit to the grid voltage. For unit 1 the transformers step up from 24 kv to 230 kv while on unit 2 the step up is from 24 kv to 525 kv. The voltages stated change within a controlled range depending on system 6

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! demand and transmission control. Duke power is implementing a l plan to more closely control grid voltage and to control voltage on the 525 kv and 230 kv grids at a 2.5% higher normal voltage. The 230 kv and 525 kv grids are the source of offsite power to the station. During periods when a units generator is not aligned to the system, the offsite system supplies power to l the station through the MSU transformers to the 6.9 kv bus.

In order to provide more stable voltages, Duke Power has decided to raise the normal voltage on the 230 kv and 525 kv l grids and implement a scheduled of variations in the voltages generated by the stations on the grid to better meet demand schedules. The variation in voltages is within the tolerances of all McGuire station equipment important to safety. The change in MSU transformer tap settings will match station voltages more closely to optimum voltages for equipment while operation with tap settings unchanged could result in equipment using voltages slightly higher than optimum but within accepta' ole limits.

Revised tap settings have been. analyzed to ensure that proper voltages are supplied to all safety related buses for all i acceptable grid conditions and bus loadings. Minimum required i grid voltages have been recalculated and reviewed with System )

Planning based on the revised tap settings. There are no

, concerns associated with being able to meet the new minimum voltages. No USQ exists. i MGMM-8186 The Waste Gas system compressors skid A has a cracked pipe at the bottom of the separator. It is expected that the pipe will have some wall thinning as the reason for the crack. Since the separator must be removed to gain access, there is an opportunity to replace many lines off the separator with erosion / corrosion resistant stainless steel pipe. The skid has been recently reclassified as QA Condition 2. The pipe stress analysis team has reviewed the replacements with no changes needed due to the low design temperature of 140 degrees F. No USQ exists.

l-MGMM-8262 l

l This modification replaces valve 2NI-180 (NI- Safety Injection System) with a new valve (Item 05J-001). This valve has been evaluated and the applicable NI system calculations have been reviewed and reconciled for the changes. No new failures are introduced or operating characteristics c' che ECCS changed.

The control logic is unaffected, and there are no Appendix R concerns. No USQ exists.

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o MGMM-8289 This modification replaces valves 1NV457 and 1NV458 with gate valves. (NV- Chemical & Volume Control System)

Two letdown orifices are arranged in parallel to reduce the pressure of the letdown stream to a value compatible with the letdown heat exchanger design. One of the orifices is sized to pass normal letdown flow; the other can pass less than the normal letdown flow. A third letdown path is provided via a control valve. Any combination of the orifices and control valves can be utilized in order to increase letdown flow, such as during reactor heatup operations and maximum purification.

This arrangement also provides a full capacity for control of letdown flow. The two orifices are placed in and taken out of service by remote manual operation of their respective isolation valves NV457 and NV458. The control valve is also controlled by remote manual operation. No USQ exists.

MGMM-8446 This change affects the main turbine electro-hydraulic (LH) system oil reservoir low / lockout level switch and main fluid pumps. The LH system, main fluid pumps and level switch are not nuclear safety related and not described in the technical specifications. The oil reservoir low lockout level trip function for the main fluid pumps will be defeated. This change incorporates a secondary system reliability improvement.

The change will prevent inadvertent main turbine trips due to false oil reservoir low lockout signals and eliminate an operator challenge to bypass the lockout function in order to prevent a turbine trip. No USQ exists.

MGMM-8449 The reactor coolant pump motor stator is a non-safety related sub-component of the Reactor Coolant System (NC) pump motor.

The minor modification involves rewinding the 2B NC pump motor stator to an approved industrial standard but without the VPI (vacuum pressure impregnation) process. The VPI process ensures that voids are minimized in the winding and that it is wound more densely, thereby resulting in a mechanically stronger component. The VPI process adds considerable time to the rewind evolution. Therefore, due to system electrical requirements (peak summer load), the rewind without the VPI process was chosen as the most prudent evolution to undertake.

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The stator will be electrically equivalent (but not as mechanically strong) as the original stator prior to rewinding The lesser mechanical-strength will result in a shorter life expectancy for the stator. Due to this, the stator will require rewinding after one cycle unless electrical testing and a visual inspection indicate that the stator can be used for an additional cycle without rewinding. The rewound stator has 100 percent of its end turns tied to the stator surge ring to minimize potential vibration degradation in the future.

Preinsulated, hardened coils will be used for the rewinding.

The original stator coils were designed with a NEMA Class F insulation system. There is no change in the NEMA insulation classification for the rewound stator as compared to the original stator. The NEMA insulation classifications based on expected temperature environments in which the equipment will be operated. No USQ exists.

NGMM-8546, 8547 These minor modifications replace the existing eight RTDs (Minco S44PA43Z48/FG180) with a Minco sS328PAZ^T36UB RTD.

These minor modifications will also raise the Operator Aid Computer (OAC) alarm set points from 200 degrees F to 230 degrees F. Finally, these minor modifications revise the

-design temperature limits for a portion of the Auxiliary Feedwater System (CA) pump discharge piping. The design temperature limit for this piping will be raised from 240 degrees F to 275 degrees F.

The changes associated with this MM enhances the performance of j the monitoring system to ensure early warning of the potential 1 for steam binding of the CA pumps. The. regulatory commitments )

that were made in response to Generic Letter 88-03 will now be met as a result of the changes provided by this MM. This MM only' returns the tempe ature monitoring system'that was installed'to its-original purpose. The changes provided by this MM will not affect how the temperature monitoring systen operates. The procedural actions to be taken when an OAC alarm is received, as committed to in response to Generic Letter 88-03, is unaffected by this MM.

l This MM does not impact the design basis, function, or method L of operation of the CA system. The new alarm set point is within the design limits for the piping upstream of the check valves. The new RTDs will provide a more accurate and conservative reading of the temperature of the piping upstream of the check valves. The changes ensure that the temperature l

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monitoring system will provide early warning of the potential for steam binding of the CA pumps, as committed to in response to Generic letter 88-03. Accordingly, this MM will not affect the design, function or operation of any structure, system or component that is necessary to operate McGuire in accordance with the SAR. No USQ exists.

.- MGNM-8559 l

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This minor modification relocates 2NV0392 (NV- Chemical Volume

& Control System) to the downstream side of orifice assembly 2NVFE7800. The flex control cable to'2NV0392 will be replaced or lengthened. The instrument lines to 2NVFE7800 will be shortened when the orifice is moved. No changes are made to change any license commitments. No USQ exists.

MGMK-9046 This modification will correct containment vessel drawings and specifications to reflect as-built configuration of the vertical stiffeners, and the material used for replacement containment vessel plates. FSAR is being revised to reflect as-built configuration and material specification-changes for containment vessel plates described herein. This MM involves y no USQs or safety concerns. No USQ exists.

MGMM-9303 This modification (Component Cooling (KC) System Valve TAC j sheets) revises existing KC system valve Test Acceptance j Criteria (TAC) sheets, deletes TAC sheets for KC system valves no longer in the Inservice Testing (IST), the KC DBD, MCS-1573.KC-00-0001, will be revised to incorporate the changes to the TAC sheets. These documents describe the design criteria, function, operability requirements and testing acceptance criteria for valves in the In Service Test (IST) program. No changes to the current design, operability, testing methods or acceptance values are made by this minor modification. No USQ exists.

MM-4184, 4185 l: These modifications delete the use of the containment sump discharge flow indication for monitoring unidentified leakage inside' containment. This system was determined to be unsuitable for this application due to the errors associated with the measurements. TS 3/4.4.6 requires that this system or the containment floor and equipment sump level system be 10 l

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( operable to detect reactor coolant system leaks during l operation. The technical specifications will be reviewed to only consider the sump level for meeting this part of the specification. This evaluation assumes the change to tech

! specs is approved. Design Basis Document MCS-1565.WL-00-0001 will be revised to reflect these changes. Instrument

, components associated with the WL5720 (EL- Liquid Radwaste L System) instrument loops will be removed.

Since the function this system performs is provided by several alternate means and the system is not required to be used if the other specified measurement systems are operable, it is acceptable to remove this system. The system is not safety related and is not used for any control function. It is not considered in any accident analysis or accident mitigating function. No USQ exists.

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