ML20207F238

Forwards Copy of Catawba Nuclear Station Units 1 & 2 1998 10CFR50.59 Rept, for NRC Files

ML20207F238
Person / Time
Site:	Catawba
Issue date:	06/01/1999
From:	Gordon Peterson DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9906080129
Download: ML20207F238 (3)

Text

{{#Wiki_filter:.- !. Duke Energy Corporation j-j Catawba Nudear Station 4800 Concord Road York, SC 29745 Gary R. Peterwn (803) 831-4251 on7a Vice Pmident (803) 831-3426ux June 1, 1999

      'U. S. Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Duke Energy Corporation Catawba Nuclear Station Unit 1 and Unit 2 Docket Numbers 50-413 and 50-414 1998 10CFR50.59 Report On May 28,.1999 Catawba Nuclear Station was notified by Peter Tam of NRR that the 10CFR50.59 Annual Summary Report was not in the Docket File. Please accept the attached copy of the report for your files. Questions regarding this report should be directed to J. W.

      .Glenn at (803) 831-3051.

Sincer l . j G. R. Peterson 1 Attachment ,

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L L U.S. Nuclear Regulatory Commission June 1, 1999 Page 2 xc: w/o Attachment. L. A. Reyes U. S. Nuclear Regulatory Commission Regional Administrator, Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 l P. S. Tam L NRC Senior Project Manager (CNS) U. S. Nuclear Regulatory Commission Mail Stop O-8H12 Washington, DC 20555-0001 D. J. Roberts Senior Resident Inspector'{CNS) U. S. Nuclear Regulatory Commission Catawba Nuclear Site I u.

U.S.-Nuclear Regulatory Commission June 1, 1999 Page 3

        .bxc: w/o Attachment J.W.'Glenn NRIA File / ELL lRGC Licensing Group File:   801.01 G. D. Gilbert                  CN01RC K. E. Nicholson                CN01RC L. A. Keller                   EC050 J. E. Burchfield               ONO3RC M. T. Cash                     MG01RC Catawba Owners: NCMPA-1, SREC, PMPA, NCEMC Document Control File 801.01 1

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f[ l-t N Duke Energy Corporation I 8 bg - Catawba Nucicar Station 4800 Concord Road York, SC 29745 Gary R. Peterwn (803) 8314251 onn y;c, jy,,;s.,,, (803) 831-3426Mx April 1, 1999 U. S. Nuclear Regulatory Commission ATTENTION: Document Control Desk washington, DC 20555-0001 1

SUBJECT:

Duke Energy Corporation Catawba Nuclear Station Unit 1 and Unit 2 Docket Numbers 50-413 and 50-414 1998 10CFR50.59 Report Attached please find a report containing a brief description of changes, tests, and experiments, including a summary of the safety evaluation of each, for Catawba Nuclear Station Units 1 and 2 during 1990. This report is being submitted

    . per the provisions of 10CFR50.59 (b) (2) and 10CFR50.4.

Questions regarding this report should be directed to J. W. Glenn at (803) 831-3051. Si rely, G. R. Peterson l Attachment l l l- [ i n  : ? . l 0QY-{&l : 00(c& & Qf

y [ U.S. Nuclear' Regulatory-Commission April 1, 1999

    ' Page 2 l-xc:

L. A. Reyes U. S. Nuclear Regulatory Commission Regional Administrator, Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 l Atlanta, GA-30303 P. S. Tam NRC Senior Project Manager'(CNS) U. S. Nuclear Regulatory Commission Mail Stop O-8H12 Washington, DC 20555-0001 i D. J.' Roberts Senior Resident Inspector-(CNS) U. S.' Nuclear Regulatory Commission Catawba Nuclear Site l l t

f i [ ..

                                 ~

U.S. Nuclear Regulatory Commission

,1 - April 1, 1999 Page 3 bxc

J.W.' Glenn NRIA File / ELL RGC Licensing Group File: 801.01 l

            - bxc-w/o Attachment G. D.' Gilbert                  CN01RC K. E. Nicholson                 CN01RC
            ' L. A. Keller                   EC050 J. 8.'Burchfield                ONO3RC-M. T. Cash                      MG01RC Catawba Owners: .NCMPA-1, SREC,'PMPA, NCEMC-Document Control File 801.01 4

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i i I I Catawba Nuclear Station Units 1 and 2 1998 10CFR50.59 Report April 1, 1999 i This report consists of a summary of changes, tests, and experiments, including a summary of the safety evaluation of each, for Catawba Nuclear Station, Units 1 and 2, for 1998. The entries are organized by the type of activity being evaluated in the following order: Minor Modifications Pages 1- 98 Miscellaneous Items Pages 99-146 Nuclear Station Modifications Pages 147-164  : I Procedure Changes Pages 164-215 UFSAR Changes Pages 216-251 i

q r U.S. Noelear Regulatory Comunission April 1,1999 - Pase 1 of 247 p , l 322 ' _ Type: Minor Modification Unit: .1 I'

Title:

Minor Modification CE-03042, Provide a more reliable mechanical seal for Containment l i Spray Pump 1 A

Description:

The mechanical seal in the Unit'l Containment Spray pumps is not reliable and has a history ofleakage problems. Dese leakage problems have resulted in excessive l l contamination and extensive manpower requirements. John Crane Inc., the seal l . manufacturer, and Duke Power Engineering Groups have developed a new seal design that is easier to install and more reliable. De new design continues to utilize an 0-rir9 between the seal sleeve and the pump shaft to prevent leakage. , Evaluation: ne function or operability of the Containment Spray system will not be affected by providing the new mechanical seal.' he seal is designed to operate at the temperature and i pressure conditions experienced in this portion of the Containment Spray system. He  ! new seal will not affect the operation of the Containment Spray system Pump I A. The addition of the new seal will result in better sealing capability and more efficient seal installation. Therefore, this modification does not increase the probability or consequences of an equipment malfunction already evaluated in the UFSAR. He function of the Containment Spray system will not be affected by the modification; therefore, the modification does not increase the probability or consequences of an accident already evaluated in the UFSAR. For the same reasons, this modification does not create the possibility for an accident or equipment malfunction which is different than already evaluated in the UFSAR. His modification does not reduce the margin of safety as defined in any technical specification bases. There are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. No UFSAR changes are required.

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i f U.S. Nuclear Regulatory Comadssion April 1,1999 Page 2 of 247 l 323 Type: Minor Modification Unit: 1

Title:

Minor Modification CE4)3043, Provide a more reliable mechanical seal for Containment Spray Pump IB

Description:

The mechanical seat in the Unit 1 Containment Spray pumps is not reliable and has a , history ofleakage problems. These leakage problems have resulted in excessive l_ contamination and extensive manpower requirements. John Crane Inc., the seal

manufacturer, and Duke Power Engineering Groups have developed a new seal design l that is easier to install and more reliable. The new design continues to utilize an 0-ring l between the seal sleeve and the pump shaft to prevent leakage.

l l Evaluation: The function or operability of the Containment Spray system will not be affected by providing the new mechanical seat. The seal is designed to operate at the femperature and pressure conditions experienced in this portion of the Containment Spray system. The new seal will not affect the operation of the Containment Spray system Pump IB, The addition of the new seal will result in better scaling capability and more efficient seal l installation. Therefore, this modification does not increase the probability or l consequences of an equipment malfunction already evaluated in the UFSAR. The l function of the Containment Spray system will not be affected by the modification; i therefore, the modification does not increase the probability or consequences of an

accident already evaluated in the UFSAR. For the same reasons, this modification does not create the possibility for an accident or equipment malfunction which is different than

! already evaluated in the UFSAR. This modification does not reduce the margin of safety as defined in any technical specification bases. There are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. No UFSAR changes are required. l i 1 1 l

 ' U.S. Nuclear Regulatory Commission April 1,1999 Page 3 of 247 80     Type: Minor Modification                                               Unit: 2
            'Ihle: Minor Modification CE-04226 Replace Train B Fuel Po.sl Exhaust Air Flow Monitor Transmitter and add isolation valves and quick disconn. cts to instrument lines.

Description:

Minor Modification CE-4226 replaces the Train B Fuel Pool Exhaust Air Flow Monitor Transmitter and adds isolation valves and quick disconnects to instrument lines, it also adds isolation valves for the associated air flow pressure switch. His transmitter and its associated air flow pressure switch are not nuclear safey related. This instrumentation is provided downstream of the filter system to provide control room indication of exhaust flow rates. Evaluation: Since the instrumentation affected by this modification cannot increase the likelihood of a fuel handling accident, the probability of an accident evaluated in the UFSAR will not increase. Since filtration is provided by the Fuel Pool Ventilation System and Auxiliary Building Ventilation System, any failure as a result of the modification will rot increase the off-site radiological conrequ:r.es. Also, since the instrumentation is located downstream of the Fuel Poel Yeouiation System Filter Units, any radioactive material within the system will be filteied before reaching the instrumentation. Therefore, failure of the instrumentation will not restrict access to vital areas of the plant or impede actions that mitigate the consequences of an accident. As a result, the consequences of an accident evaluated in the UFSAR will not increase. Since no new credible failure modes or operating characteristics were identified during the above evaluations, the possibility for an accident of a different type than those already addressed in the UFSAR will not be created. The probability of a malfunction of equipment important to safety evaluated in the UFSAR will not increase as a result of the failure modes identified. Also, the consequences of a malfunction of equipment important to safety evaluated in the UFSAR will not increase. Since the failure modes identified are bounded by the seismic integrity of the proposed design and by the Fuel Pool Ventilation System and Auxiliary Building Ventilation System design, the possibility for an equipment malfunction of a different type than those evaluated in the UFSAR will not be created. This modification will have no effect on the ability of the system to restrict radioactive j material release to the environment. Herefore, the margin of safety defined in the

                    'Ichnical Specification Bases will not be reduced. There are no unreviewed safety questions associated with this modification. No Technical Specification changes are           j required. No UFSAR revisions are required.                                                    i s

4 L l t

U.S. Nuclear Regulatory Conunission April 1,1999 Page 4 of 247 4 6 Type: MinorModification Unit: 1

Title:

Minor Modification CE-04661, replace Train B Fuel Pool Exhaust Air Flow Monitor and add isolation valves and quick disconnects

Description:

Minor Modification CE-4661 replaces Train B Fuel Pool Exhaust Air Flow Monitor and adds isolation valves and quick disconnects. The modification replaces the Train B Fuel Handling Exhaust Air Flow Monitor transmitter, adds an isolation valve and quick disconnects, adds an instrument number for the transmitter and adds isolation valves for the associated air flow pressure switch which interfaces with the Fuel Pool Supply Fan. Evaluation: There is no unreviewed safety question as a result of this modification. The instrumentation involved in the modification is not nuclear safety related and cannot increase the likelihood of a fuel handling accident. No Technical Specification changes are required. UFSAR Figure 9-118 will be revised. { 7 Type: Minor Modification Unit: 1

Title:

Minor Modification CE-05016, Install a piping nipple on seal plate at hanger 1.R-BD-  ! 0032

Description:

Minor Modification CE-5016 Install a piping nipple on seal plate at hanger 1-R BD-0032, adds a 3/4 inch threaded nipple and pipe cap to the seal plate on the hanger. The cont.ection is required in order to pump insulation behind the seal plate to lower the temperature of the concrete at the hanger location. The hanger is located in the Auxiliary Building Steam Generator Blowdown Pipe Trench. Evaluation: There is no unreviewed safety question associated with this modification. The ability of the seal plate to perform its function will not be degraded. The addition of this nipple does not affect any assumptions or evaluations performed in the UFSAR. No Technical Specification changes are required. No UFSAR changes are required. l l (.-

U.S. Nuclear Regulatory Commission April 1,1999 Pate 5 of 247 8 Type: Minor Modification Unit: 0

Title:

Minor Modification CE-07353, Remove various fire extinguishers in the Service Building, Turbine Building, Auxiliary Building and Auxiliary Service Building

Description:

Minor Modification CE.7353, Remove various fire extinguishers in the Service Building, Turbine Building, Auxiliary Building and Auxiliary Service Building deletes approximately 150 fire extinguishers from the plant. The original number and location of fire extinguishers was based on compliance with NFPA 10-1978, " Installation, Maintenance and Use of Portable Fire Extinguishers" and on the assumption that any person was allowed to use the fire extinguisher. *ne policy now is for any employee who discovers a fire to call the site emergency number. Trained fire fighting personnel (fire brigade) would then respond to the fire. Based on OSHA Standard 29CFR1910.157, Subpart L, the extinguisher locations are based upon fire brigade need. Evaluation: The reduction in the number of fire extinguishers throughout the plant will not create an unreviewed safety question. The remaining fire extinguishers will provide adequate manual fire fighting equipment for use by the trained fire brigade. No technical specification changes are required. No UFSAR changes are required. 9 Type: Minor Modification Unit: 1

Title:

Minor Modification CE-07779 Abandon in place Pressurizer Heaters 31,59, and 60.

Description:

Minor Modification CE-7779, Abandon in place Pressurizer Heaters 31,59, and 60; abandons a part of one group of Pressurizer Heaters. Pressurizer Heater 60 has faulted to ground and it is judged impractical to repair it. Since Heater 31 and 59 are connected with Heater 60 in a delta electrical configuration all three heaters will be abandoned. This will result in a 69KW loss of capacity in Pressurizer Heater Group B. Adequate Heater Capacity remains to meet the Technical Specification minimum heater capacity requirement (347KW available,150 KW required). Evaluation: There is no unreviewed safety question associated with this modification. The Pressurizer Heaters and their control circuitry are not nuclear safety related. Per Chapter 15 of the UFSAR, non safety systems and equipment are only assumed to be operable in an accident analysis when the operation of the system would cause the analyzed transient to be more severe. Therefore the heaters are assummed to be unavailable if that would make the analyzed transient more severe. The effect of a 69KW reduction in heater capacity is bounded by these assumptior ; therefore, the probability or consequences of an accident would not be increased, t l l 1 i I

i U.S. Nucicar Regulatory Commission April 1,1999 Page 6 of 247 4 l 10 Type: Minor Modification Unit: I  ; l

Title:

Minor Modification CE-07900, Remove straightening sections from Auxiliary Building  ; Ventilation System air flow monitors i (

Description:

Minor Modification CE-7900, Remove straightening sections from Auxiliary Building l Ventilation System air flow monitors, removes elements from ventilation ducts that are caudng reduced airflow. The straightening sections are structurally shaped like a honeycomb with small openings that are being blocked with dirt and debris. After removal of the straightening section the air flow monitor will continue to perform its design function of sampling and monitoring air flow. Seismic integrity of the air flow monitor will not be adversely affected. Evaluation: Here is r.o unreviewed safety question associated with this modification. He air flow monitors are not required to function during an accident condition. The structural integrity of the air flow monitor will not be adversely affected. No changes to the Technical Specifications are required. No UFSAR changes are required. 11 Type: Minor Modification Unit: 1

Title:

Minor Modification CE-07901, Remove straightening sections from Auxiliary Building Ventilation System air flow monitors

Description:

Minor Modification CE-7901, Remove straightesang sections from Auxiliary Building Ventilation System air flow monitors, removes elements from ventilation ducts that are causing reduced airflow. The straightening sections are structurally shaped like a honeycomb with small openings that are being blocked with dirt and debris. After removal of the straightening section the air flow monitor will continue to perform its design function of sampling and monitoring air flow. Seismic integrity of the air flow monitor will not be adversely affected. I Evaluation: There is no unreviewed safety question associated with this modification. The air flow monitors are not required to function during an accident condition. He structural integrity l of the air flow monitor will not be adversely affected. No changes to the Technical Specifications are required. No UFSAR changes are required. l

  .U.S. Nuclear Regulatory ^==8==8=
 ' AprE 1,1999 Pase 7 of 247 f

~237l Type

MinorModification Unit 2 71tle: Minor Mod!5 cation CE-07944, Replace corroded Junction Box 2TBOX0018

Description:

Minor Modification CB-7944 will replace corroded Junction Box 2TBOX0018. lhis electricaljunction box is associated with the Safety Injection System which is designed to maintain core cooling during emergency conditions. The replacement of the box will be done while Unit 2 is in "No Mode" during the 2EOC9 refueling outage. The ability of the _ system to perform its function in normal plant operation will not be affected. Evaluation: The replacementjunction box is identical in form, fit, and function to the one it replaces. There is no unreviewed safety question associaH with this modification. No Tecimical Specification changes are required. No UFSAR changes are required. 228 Type: Minor Modification Unit: 2

Title:

Minor Modification CE-08223, Component Cooling Heat Exchanger 2B Tubesheet Coatings

Description:

Minor Modification CE-8223, Component Cooling Heat Exchanger 2B Tubesheet Coatings, addresses addition of epoxy coatings to tubesheets ar.d channel interior surfaces of the heat exchanger. The coating is designed as a protective coating for under water

                   . service conditions. Unprotected original carbon steel tubesheets of the Component Cooling Heat Exchangers have corrosion damage due to exposure to raw lake water used
                   ' for cooling. Corrosion damage has a potential to worsen if not corrected. This coating is approved for underwater use. The probability of the coating detaching after application was evaluated to be acceptably low.

Evaluation: This modification does not change the function of the heat exchangers. The coating is a

proven system well suited for the application. No Unreviewed Safety Questions are l

! created as a result of this modification. No Technical Specification changes are required. l ~ No UFSAR revisions are required. l l l L l l L i

l , U.S. Nuclear Regulatory Con-da= Ion April 1,1999 . Page 5 of. 247 285l Type: MinorModification Unit: 2 "Iltle: Minor Modification CE-08355. Installation of torque collars on the Unit 2 Main Turbine for torsional vibration testing

Description:

Minor Modification CE-8355 installs torque collars on the Unit 2 Main Turbine for l torsional vibration testing. These items were installed to determine the natural frequencies of the Unit 2 main turbine rotors. These frequencies can be excited or affected by torsionally induced vibrations. The torque collars will be mounted on the turbine shaft at the number 4,6 and 8 bearings and will rotate with the shaft. A receiver

will be mounted on the bearing that will receive the signal from the torque collars and

[ transmit data to a computer via a cable that will be installed througf ^he sump wall for L collection of data. l Evaluation: Neither the Main Turbine nor the Main Turbine and Lubrication Oil Purification System will be affected in any way by the implementation of this modification. Timre are no l; unreviewed safety questions associated with this modification. No Technical j Specification changes are required. No UFSAR changes are required. 12 Type: Minor Modification - Unit: 2 l-

Title:

Minor Modification CE-08449, Convert Steam Generator Blowdown Pump 2B to shaft

packing in place of mechanical seals

Description:

Minor Modification CE-8449, Convert Steam Generator Blowdown Pump 2B to shaft packing in place of mechanical seals, changes the pump shaft sealing mechanism. This l pump is used to transfer condensed liquid in the blowdown tank to the blowdown l recovery heat exchangers. The pump has experienced frequent ar.d recurring mechanical seal leaks. This modification will change from mechanical seals to packing. Packing will provide greater radial stability for the pump shaft and will be less afftetM by system transients. An added benefit is that the seal water heat exchanger can be temoved. t - Evaluation: There is no unreviewed safety question associated with this modification. The pump is l' not nuclear safety related and is not required by the Technical Specifications. Pump packing is a proven technology which in this application will provide better reliability. All materials used in this modification meet the applicable design, material and construction standards. No changes to the Technical Specifications are required. A change is required for UFSAR Figure 10-31, I ( l b

~, U.S. Nuclear Regulatory Comunission Apdf 1,1999 Page 9 of 247 13 Type: Minor Modifkation Unit: 2 j J Iltle: Minor Modification CE-08558, Replace 2NWLT5020 from a Barton 386A to a - Rosemount ll53D

Description:

Minor Modification CB-8558, Replace 2NWLT5020 from a Barton 386A to a Rosemount 1153D, changes out transmitter 2NWLT5020 for a more reliable unit. The

  ,                     Containment Valve Injection Water System includes transmitter 2NWLT5020 The current Barton transmitter has a history of calibration problems. The transmitter will be replaced with a Rosemount unit in order to get & transmitter with the proper range.

2NWLT5020 monitors water level in a surge chamber and functions to open valve 2NW. 008A which allows the Nuclear Sevice Water System to supply seal water requirements of the Containment Valve injection Water System.

      - Evaluation: This modification will not change any parameters that affect the safe operation of the plant. The function of the transmitter is not changed by the modification. He replacement transmitter is considered equivalent in form, fit, and function to the old transmitter. There are no unreviewed safety questions involved widi this modification. No Technical Specification changes are required. UFSAR table 3-106 will require a revision.

14 ' Type: MinorModification Unit: 1

Title:

Minor Modification CE-08578, Replace INV 315, INV-320, INV-327, and INV 332 and delete valves INB-885, INB-887, INB-892, and INB 894,

Description:

Minor Modification CE-8578, Replace valves INV-315, INV 320, INV-327, and INV-332 and deletes valves INB-885, INB S87, ING-892, and INB-894; replaces several valves with a different type valve and deletes several other valves. The valves that are being replaced did not perform satisfactorily. The valves that are being deleted are no longer needed due to the characteristics of the replacement valves. Evaluation: The replacement valves will perfctm the same system function as the old valves. He differences between the old and replacement valves have been evaluated and determined i to be acceptable. There is no unreviewed safety question associated with this modification. No Technical Specifications changes are required. Changes to UFSAR Figure 9-91 will be required and a new figure will be added to the UFS AR. l l i I l l l- I f  ; b

U.S. Nuclear Regulatory Commission April 1,1999 Page 10 of 247 IS Type: Minor Modification Unit: 1

Title:

Minor Modification CE-08616, rewire Control Room Switch for INC33A Pressurizer PORV Block valve so that the close circuit bypasseslNCLIA310 and INCLLO350

Description:

Minor Modification CE-8616, rewire Control Room Switch for INC33A Pressurizer PORV Block valve so that the close circuit bypasses INCLIA310 and INCLLO350; changes control wiring for valve INC33A. There is a problem with the "close" circuitry of INC33A which has been evaluated to not adversely impact the Operability of the Pressurizer PORV or its block valve, INC33A. Due to a faulty interlock signal (probably caused by a block valve stem mounted limit switch contact in either valve INC31B or INC35B), INC33A can only be closed in " override". His is considered as an undesirable " work-around" for the operators as it may create an additional burden during transients and postulated accidents. Evaluation: The probability of an accident previously evaluated in the SAR will not be increased. He active safety function involved is maintaining the capability to close the block valve in response to a stuck open PORV following its actuation. The purpose of the interlock between the three block valves, while not an automatic safety function, is to serve as a minor obi tacle to the operator in closing more than one block valve, and hence, potentir.tly increasing the possibility that a Pressurizer Code Safety Valve would be challenged should a transient occur that would require PORV actuation. Tech Specs already allow the closure of up to 3 block valves if PORVs have excessive seat leakage, so it is apparent the NRC has reviewed and approved the potential for a Pressurizer Code Safety being challenged in this manner. Given the administrative awareness provided by maintaining any existing block valve closure in Technical Specification Action item Log, it isjudged that the benefit of this Minor Mod in assur~mg an uncomplicated operator response to a stuck open PORV more than offsets the potential for an operator to be unaware he has closed more than one block valve and, as a result, increase the potential for a stuck open Code Safety. Therefore there is no increase in the probability of an accident previously evaluated in the SAR. The consequences of an accident previously evaluated in the SAR will not be increased if an accident were to occur and the operator has inadvertently closed more than one PORY block valve as a direct result of this Minor Mod. All the Chapter 15 accidents are analyzed assuming the PORVs do not lift automatically, unless the consequences are more severe as a result of their lifting and lowering reactor coolant system pressure. Since these analyses are already performed assuming PORV action or inaction to maximize consequences, the consequences of accidents previously evaluated in the SAR will not be increased. The possibility of an accident different from one already evaluated in the SAR will not be created because the modification eliminates a known failure mode (INCO33A not closing on "close") and does not create any new failure modes. The circuit being jumpered performs an awareness function, and does not affect PORV automatic actuation, or block valve open or close functions except as intended by the operator. The probability of a malfunction of equipment important to safety will not be increased. The isolation of all three PORVs in the event of seat leakage as allowed by Tech Spec

l. U.S. Nuclear Regulatory Conunission Apdf 1,1999 Page 11 of 247 3/4.4.4 indicates that the NRC has reviewed and approved this configuration as an acceptable challenge to the potential for Pressurizer Code Safety valve lifting and possibly sticking open. Thus it apps that the administrative controls provided by the l Tech Spec, rather than the minor hardware obstacle provided by the interlock ensures operator awareness that the PORV automatic overpressure mitigative feature is blocked on more than one PORV. Additional procedural awareness of block valve position is specified in the Controlling Procedure for Unit Shutdown (Ref 5). Enclosure 4.2 of this

procedure specifies that PZR PORV blocks INC31B and 1NC33A be opened in l preparation for the operability of their associated PORVS. This step assures that the NC l System will not be cooled below 285 degrees F with either of these block valves

( inadvertently isolated. There are no unreviewed safety questions associated with this l; modification. No Technical Specification changes are required. No UFSAR changes are required. 67 - Type: Minor Modification Unit: 0 l

Title:

Minor Modification CE-08681 Revision to the Nuclear Service Water System Design Basis Specification

Description:

Minor Modification CE-8681, Revision to the Nuclear Service Water System Design Basis Specification, makes several changes to CNS-1574.RN-00-0001. Most of the ( changes were made to correct inaccurate or omitted information. These changes are consistent with the plant as described in the UFSAR. A significant change was made to j allow a second Nuclear Service Water Pump Motor Cooler to be supplied flow by l gagging open a failed closed automatic isolation valve. This guidance existed previously but did not consider the effects on the Nuclear Service Water System flow balance. The UFSAR previously stated that this flowpath was not estab!ished unless the pump was running. Since this flowpath may be established without the associated pump running, this statement was removed from the UFSAR. Evaluation: There are no unreviewed safety questions associated with this modification. Any single nuclear service water system pump is capable of providing sufficient flow for all required l loads even with a second nuclear service water system pump motor cooler being supplied. The probability of a malfunction of a nuclear service water system pump motor l is not increased since there is a testing program to ensure that motor insulation is not degraded. No Technical Specification changes are required. A change is required for UFSAR Section 9.2.1.2.3. l l l l l

U.S. Nuclear Regulatory Commission April 1,1999 Page 12 of 247 9 16 Type: Minor Modification Unit: 1

          'ntie: Minor Modification CE-08690, drill a one eighth inch hole in the discs of valves IND.

019 and IND-053

Description:

Minor Modification CE-8690, drill a one eigSth inch hole in the discs of valves IND-019 and IND-053, reheves pressure locking concerns on these valves. Valves IND-019 and IND-053 are the Residual Heat Removal Heat Exchanger Isolation valves for the 1 A and IB Heat Exchangers. Drilling these holes will alleviate pressure locking concerns that have been experienced on the equivalent Unit 2 valves. Evaluation: There are no unreviewed safety questions associated with this modification. The valves will still meet the applicable design, material and construction standards. Drilling the holes in the upstream side of the discs will not cause the residual heat removal system to be operated otr. side its design or testing limits. The valves are normally locked open, passive, mareal valves that are not required to perform a safety function during an accident. These valves provide pressure boundary integrity for operation of the residual heat removal system. Reducing the sealing capability from two seals to one will not inhibit the system from performing its intended safety function. No Technical Specification changes are required. Changes are required for UFSAR Figures 5-17 and 5-

                  ! 8, 246     Type: MinorModification                                               Unit: 2                      ,

Title:

Minor Modification CE-08691, Drill a one eight inch hole in the disc of valve 2ND019

Description:

Minor Modification CE-8691 allows for drilling a one eight inch hole in the disc of valve 2ND019. Valve 2ND019 is the Residual Heat Removal Heat Exchanger 2A isolation valve. Valve 2ND053, the equivalent B Train valve has pressure locked in the past. Drilling a 1/8 inch hole in the upstream side of the disc in valve 2ND019 will prevent pressure from becoming trapped in the valve's bonnet so that the valve will no longer be susceptible to pressure locking. No valve design parameters other than the item number will be affected by this modification. Evaluation: There are no unreviewed safety questions associated with this modification. This valve is a normally locked open, passive, manual valve that does not perform any safety function during an accident. The valve serves to provide pressure boundary integrity of the Residual Heat Removal System. Reducing the sealing capability from two seats to one will not keep the Residual Heat Removal System from performing its safety function. No Technical Specification changes are required. No UFSAR changes are required. l I i l i

7_

 ~

U.S. Nuclear Regulatory Chh ', 1 Apdl1,1999 . i Pase 13 of 247 68' Type: MinorModification Unit: 0

                 'Iltle: Minor Modification CE-08710, Setpoint change for instruments NDPG5040 and NDPG5050

Description:

Minor Modification CE-8710, Setpoint change for instruments NDPG5040 and NDPO5050, will prevent thermal and hydraulic cycling of the residual heat removal system minimum flow valves in unplanned low flow conditions. 'Ihe setpoints on the instruments control the opening and closing of the residual heat removal system miniflow valves (1,2ND25A and 1,2ND59B). This protects the residual heat removal pumps by assuring a flow path in low flow conditions and allows adequate ECCS igiection flow in LOCA events. > The current setpoints (1000 gallons per minute (gpm) to close and $33 L gpm to reopen) could cause cycling of the valves as described in a Westinghouse Technical Bulletin. The flow band between the setpoints is currently 467 gpm and the closing of the valves can set off a flow swing of 500-570 spm, thus starting the valves to cycle. This causes a " work-around" item for Operations. The new setpoints will provide a sufficient deadband to to avoid this problem. Evaluation: There is no unreviewed safety question is associated with this modification.- The residual heat removal system will operatejust as it did previously except that the cycling problem will be eliminated. There will be no significant change in flow during normal residual heat removal system operation. No changes to the Technical Specifications are required. Changes are required for UFSAR Sections 5.4.7.2.1,5.4.7.2.2 and 6.3.5.3. UFSAR Tables 5-31 and 6-91 also require a revision. 1 l J l l

m

      - U.S. Nuclear Regulatory Conumission
      - April 1,1999 Page 14 of 247.

P 299' Type: Minor Modification Unit: 1

Title:

Minor Modification CE-08748, Add flanged branch connection to an existing cooling tower makeup line f -.

Description:

Minor Modification CB-8748 adds a 16 inch flanged branch connection to an existing 30 inch header to the Unit ! Cooling Tower makeup line. The cooling tower makeup header is a 30 inch line. To allow for internal inspection of the 30 inch header during shutdowns, the 16 inch branch connection will be installed while the unit is at power. Prior to welding cf the 16 inch branch connection, the 30 inch header will be inspected by i ultrasonic testing (UT) to ensure the proper amount of pipe wall thickness exists. The 16 inch branch connection will include a slip on flange connection to allow access to the 30 inch header. The final tie in to the system will be achieved by cutting a hole in the 30 inch header through the 16 inch branch connection during the IEOC10 refueling outage, when the system is isolated and drained. All work will conform to the applicable codes and standards. l

          - Evaluation: There is no unreviewed safety question associated with this modification. The system involved is not nuclear safety related and serves no mitigating function for any accident analyzed in the UFSAR. The portion of the work which is performed on line could, in the worst case scenario, burn a small hole in a Low Pressure Service Water line. Such a hole would not have a siginifcant effect on the operation of the system. No Technical y                          Specification changes are required. No UFSAR changes are required.

i l l l i l I 1 l 1 l l l i

E L l I U.S. Nuclear Regulatory Cosamission April 1,1999 Pane 15 of 247 l l

• l 93 Type: Minor Modification' Unit: 0

          'ntie: Minor Modification CB 08772, Allow the option to reinove the one-hour fire barrier cable wrap on the power supply cable for the "B" Fire Pump.

Description:

Minor Modification CB-8772 allows the option to remove the one-nour fire barrier cable wrap on the power supply cable for the "B" Fire Pump of the Exterior Fire Protection System. The one hour fire barrier is no longer needed based on the findings of the

• Cost Beneficial Licensing Action" initiative. De intent of the " Cost Beneficial Licensing Action" (CBLA) initiative was to focus resources on features which are important to nuclear safety and reduce commitments and requirements which are marginal to nuclear safety, and thus are not cost beneficial. As part of the CBLA, the " Cost Beneficial Licensing Study for Fire Rated Barriers Marginal to Nuclear Safety" was performed.

This study identified committed fire barriers which do not contribute to the protection of nuclear safety equipment, nor provide protection of redundant safe shutdown equipment. This study concluded that the fire rated cable wrap for the Main Fire Pump B power cable can be deleted from the committed fire rated barrier program. This conclusion was based on the following:

1) %ere are no combustible materials below the Low Pressure Service Water intake Structure. Therefore, a fire beneath the Intale Structure in the area of the B Pump power cables is not a credible event.

2) The Main Fire Pumps are not required as part of the Post Fire Safe Shutdown program.

3) ne B Pump power cable fire barrier does not protect nuclear safety related equipment.

Evaluation: This modification allows the option for the removal of the "B" Main Fire Pump power cable fire barrier cable wrap. Damage to the Main Fire Pumps power cables is not a credible event based on the arrangement and location of these cables at the lew Pressure Service Water Intake Structure. Therefore, the allowance of the removal of the one hour  ! fire barrier cable wrap from the B Pump does not result in an Unreviewed Safety i Question. No Technical Specification changes are required. A change is required for l UFSAR Table 9-31. l l I l

g _ I-o l- U.S. Nuclear Regulatory Comunission

   .: April 1,1999 Page 16 of 247 '

I f 17 Type: MinorModification Unit: 0

               'I1tle: Minor Modification CE-08786, Minor Modification CE-8786, Modify the bolted access doors on the Control Room Air Handling Units

Description:

Minor Modification CE-8786, Modify the bolted access doors on the Control Room Air l Handling Units; makes provisions for easier repair of the access doors upon failure of

                       . bolting material. Periodically maintenance workers must remove the bolted access doors on the Control Room Air Handling Units to change the fihers and inspect the cooling coils. Some of the fasteners stripped during the last time this activity was performed. De current design has fasteners that attach the access doors directly to the ventilation unit which in places is only one quarter of an inch thick. His design does nor allow easy repair of stripped fasteners. His modification adds a thicker steel frame around the outside of the access area. His will provide for easier repair or replacement of damaged fasteners.

Evaluation: De modification will be leak tight so there will be no effect on air flow to the control room. He operation of the control room air handling system will not be affected by the modification. No changes to the Technical Specifications are required. No changes to the UFSAR are required. I i l l l

FE ~ U.S. Nuclear Regulatory Commission April 1,1999 Page 17 of 247 69 Type: Minor Modification Unit: 1

Title:

Minor Modification CF 08801, Replace valves IWL-814, IWL816, IWL 818, IWL820

Description:

Minor Modification CE-8801, Replace valves IWi 814, IWL816, IWL 818, IWL820, l changes the valve type used for these applications. Existing Y-Type lift check valves l IWIe 814,816,818, and 820 which serve as the Containment Floor and Equipment Sump Pump Discharge Checks, will be replaced with swing check valve Item number 9J-l 368. The new check valve will provide more reliable seat leakage performance in a

debris laden sump system. The full bore flow path and free swinging disc will allow l small particle debris to freely flow through reducing the possibly of gross leakage from a l stuck open lift check. His enhancement modification will reduce the frequency of outage l maintenance required for cleaning these checks and the possibility of losing sump level l control due to recire flow through a failed check valve. This modification is a part of the i solution to improve system reliability along with the efforts to better control the amount and type of debris entering the sump.

l l Evaluation: The Liquid Waste Recycle system provides a means to collect, monitor, and treat waste [ water.The containment floor and equipment sump section of the Liquid Waste Recycle l system is not nuclear safety related and consists of 2 sumps,4 sump pumps, and 4 l discharge check valves all feeding into a common header. Failure of one of the 4 check valves can lead to a operational problem supporting outage maintenance (equipment isolation / draining) but is not safety significant. The valve change out is a functionally equivalent component replacement. All pertinent valve specific parameters have been reviewed by engineering. Design pressure / temperature, flow coefficient, constmetion materials, and valve weight were evaluated, ne significant valve parameters that have been affected are a slightly increased flow coefficient and a very minor weight reduction and center of gravity shift. There are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. A change is required for for UFSAR Figure Il-15. l l 1 I i

                                                                                                                )

U.S. Nuclear Regulatory Comunission Apdf 1,1999 Page 18 of 247 18 Type: Minor Modification Unit: 1

Title:

MinorModification CE-088%, Replace valve 1RLl47 with a CMV 219 valve

Description:

Minor Modification CE-3806, Replace valve IRLl47 with a CMV-219 valve, changes the type valve used for valve tag number IRLl47. Valve IRLl47 was changed from a one inch globe valve to a three quarter inch ball valve. All applicable design parameters were reviewed and it was determined that the new valve is an equivalent replacement for the previously installed valve. Evaluation: This modification involves a component replacement / improvement. De modification will result in enhanced system operability since the new valve is expected to provide better service. The accidents evaluated in the UFSAR do not include this valve in any design basis, therefore there is no increased probability of an accident evaluated in the UFSAR. The new valve will serve the same function as the previously installed valve. No changes are required to the Technical Specifications. A change is required for UFSAR Figure 9-68. 19 Type: Minor Modification Unit: 1

Title:

Minor Modification CE-08809, Replace valve IRLil6 with a CMV-219 valve

Description:

Minor Modification CE-8809 Replace valve 1RL116 with a CMV-219 valve, changes the type valve used for valve tag number IRLil6. Valve IRLil6 was changed from a one inch globe valve to a three quarter inch ball valve. All applicable design parameters were reviewed and it was determined that the new valve is an equivalent replacement for the previously installed valve. Evaluation: This modification involves a component replacement / improvement. The modification will result in enhanced system operability since the new valve is expected to provide better service. The accidents evaluated in the UFSAR do not include this valve in any design basis, therefore there is no increased probability of an accident evaluated in the ' UFSAR. The new valve will serve the same function as the previously installed valve No changes are required to the Technical Specifications. A change is required for UFSAR  ; Figure 9-68. i l k.

                                                                                                                 )

c L o (' U.S. Nuclerr Regulatory Cocamission l April 1,1999 Pase 19 of 247 f 20 ' Type: Minor Modification Unit: 1

Title:

Minor Modification CE-08815 Remove Oil Piping from the Unit i Reactor Coolant Pump Motors l

Description:

Minor Modification CE-8815. Remove Oil Piping from the Unit 1 Reactor Coolant Pump, j deletes unused oil fill and drain piping. On all four reactor coolant pump motors piping between vendor supplied valves on the upper bearing fill and drain line and lower bearing l fill and drain line and flexible metal hose connections will be deleted. The flex hose will

                     . be deleted and the remaining piping will be capped with blind flanges. Screwed plugs will

! . be used to cap the ver. dor supplini valves. Since the piping to be deleted is supported l ' with steel members attached to the motor, this support steel will also be removed. Evaluation: The oil piping and associated support steel is being removed to facilitate the removal and replacement of the reactor coolant pump motors. The deleted piping was not used for filling the oil reservoirs. Drainage of the oil will be performed by alternate means. The new configuration has been analyzed and found to be accepable with reference to piping stress analysis and support restraint steel. Tuere is no unreviewed safety question as a result of this modification. No Technical Specification changes are required. Changes are required for the UFSAR (flow diagrams). ! 70 Type: Minor Modification Unit: 1

Title:

Minor Modification CE-08851, Replace valves ICA-037

Description:

Minor Modification CE-8851, Replace valve ICA-037, changes the valve type used for j this application. Valve ICA-037, Auxiliary Feedwater Motor Driven Pump Discharge Flow to Steam Generator ID Check Valve, was replaced with a swing check valve. Evaluation; his modification is considered a valve replacement for maintenance and reliability. The original valve cannot be repaired efficiently. The new valve will serve the same function as the original valve and it is functionally equivalent to the previously installed valve. Dere are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. A revision will be required for the UFSAR (flow diagram). l

   \

p E U.S. Nuclear Regulatory Commission Apdf 1,1999

 . Page 20 of 247 282    Type: Minor Modification                                               Unit: 2

Title:

Minor Modification CE-08872, Remove torque collars on the Unit 2 Main Turbine used for torsional vibration monitoring

Description:

Minor Modification CE-8872 removes torque collars on the Unit 2 Main Turbine used for torsional vibration monitoring. He purpose for the torque collars was to determine the natural frequencies of the Unit 2 Main Turbine Rotors. The frequencies have been determined and these torque collars which were installed per a previous modification (CE-8355) can be removed. nis modification will rcturn the turbine to its original state. Evaluation: Neither the Main Turbine nor the Main Turbine and Lubrication Oil Purification System - will be affected in any way by the implementation of this modification. There are no unreviewed safety questions associated with this modification. No Technical l Specification changes are required. No UFSAR changes are required. 278 Type: Minor Modification Unit: 2

Title:

Minor Modification CE 088?,5, Replace Main Feedwater System Regulating Valve Bypass Piping

Description:

Minor Modification CE-8895 addresses replacing Main Feedwater System Regulating l Valve Bypass Piping with a material that is resis: ant to flow accelerated corrosion. The new piping material will be A335/P22. The bypass valves will not be affected by this l modification and they will remain in their present configuration. All dimensions will remain unchanged. Support / restraint design will not be affected. The material properties  : l of the P22 material are similar to those of the A106/B material that is being replaced. The j modification was reviewed from a stress analysis perspective and found to be acceptable. This modification is only a piping replacement. Evaluation: There are no unreviewed safety questions associated with this modification. All piping i involved is not nuclear safety related. The piping is assumed to fail in a seismic event. His modification will not change that assumption. No Technical Specification changes are required. A UFSAR change is required because a piping flow draw'mg was revised.

                                                                                                               )
 . U.S. Nuclear Regulatory Conunission April 1,1999 Page 21 of 247 f

21 Type: Minor Modification. Unit: 1

Title:

Minor Modification CE-088%, Replace valve IGP-010 with a new valve l

Description:

Minor Modification CE-8896, Replace valve IGP-010 with a new valve, replaces the currently installed valve with a functionally equivalent valve. The currently installed I valve is a one-half inch vendor supplied gate valve. The replacement valve is a one-inch  ! ball valve. The application that the valve is used in is not nuclear safety related. He replacement valve was evaluated by Engineering and found to be an acceptable replacement. - l Evaluatbn: The valve is used in the Carbon Dioxide Generator Purge System. The application is not nuclear safety related. This valve is not involved in any analyzed accident. There are no Unreviewed Safety Questions as a result of this modification. No Technical Specification Changes are required. Changes are required for the UFSAR (flow diagram revision). 1 22 Type: MinorModification Unit: 1

Title:

Minor Modification CE-08907, Replace valve IllW-254

Description:

Minor Modification CE-8907 Replace valve IllW-254, replaces valve IllW-254 with a functionallyequivalent valve. .The currently installed valve is a one-inch Y-type globe valve used in a drain application. The replacement valve is a one-inch bellows sealed gate valve. Evaluation: The replacement valve has been evaluated to be a functionally equivalent replacement for the currently installed valve. The valve is used in the Ileater Drain System. There are no Unreviewed Safety Questions as a result of this modification. No Technical Specification  ; Changes are required. Changes are required for the UFSAR (flow diagrr.m revision). 1

b U.S. Nuclear Regulatory Commission April 1,1999 Page 22 of 247 23 Type: Minor Modification Unit: 0

Title:

Minor Modification CE-08978, Change to the Diesel Generator Cooling Water Heat Exchanger heat capacity test acceptance criteria stats

Description:

Minor Modification CE-8978, Change to the Diesel Generator Cooling Water Heat

                 - Exchanger heat capacity test acceptance criteria sheets, rnodifies test acceptance criteria.

Dese changes are considered editorial. He test frequencies are being deleted from the Test Acceptance Criteria Sheets because the frequency of the test is controlled by the Work Management System and the Service Water System Program Manual. A requirement to install multiple shell side outlet temperature instruments is being deleted from the test acceptance criteria sheets for the I A, IB, and 2A Diesel Generator Cooling Water Heat Exchangers. The requirement actually applies only to the 2B Diesel Generator Cooling Water Heat Exchanger. Evaluation: There are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. No UFSAR changes are required. l

U.S. Nuclear Regulatory Conunission Apdf 1,1999 Pane 23 of 247 71 Type: MinorModification Unit: 2

                        'Iltle: Minor Modification CE-09001, Replace Check Valves 2WL818 and 2WL820

Description:

Minor Modification CE-9001, Replace Check Valves 2WL818 and 2Wb820 with a functionally equivalent valve. Existing Y Type lift check valves 2WU818 and 2WL820 which serve as the Containment Floor and Equipment Sump Pump Discharge Checks, will be replaced with swing check valves. De new check valves will provide more reliable seat leakage performance in a debris laden sump system. The full bore flow path and free sw'mging disc will allow small particle debris to freely flow through reducing the possibility of gross leakage from an unseated lift eneck. His modification will reduce the maintenance required for cleaning these valves and reduce the possibility oflosing sump level control due to recirculation flow through a failed check valve. This modification will improve system reliability and help to better control the amount and type of debris entering the sump. Evaluation: This valve change out is a functionally equivalent component replacement. Valve specific parameters have been reviewed by engineering. The following paramters were considered in this review: design pressure, temperature, flow coefficient, construction materials, and weight. The significant valve parameters that have been affected are a slightly increased flow coefficient and a very minor weight increase and center of gravity shift. Systems engineering and stress analysis have approved these minor changes . No operational parameters, procedures, or indications will be affected by this valve replacement, therefore there is no unreviewed safety question associated with this modification. No Technical Specification changes are required. No UFSAR revisions are required. The valve will be installed using existing maintenance procedures including a pre-installation seat leakage test and the appropriate controls to ensure proper installation. Following installation, an audible inspection for disc tapping / flutter under flow conditions and a visual leakage inspection at full system pressure and temperature will verify system integrity and valve operation. This modification was previously performed in similar Unit 2 applications with satisfactory results.

U.S. Nuclear Regulatory r'w-t-AprG 1,1999 Pame 24 of 247 24'- Type: Minor Modification Unit: 2.

               ' 'Iltle: Minor Modification CE-09064, Change Unit 2 Steam Generator Blowdown Pump
                         . packing to mechanical seals

Description:

Minor Modification CE-9064, changes the Unit 2 Steam Generator Blowdown Pump packing to mechanical seals, and provides an option to use a new type of nwchanical seal in all Steam Generator Blovnlown Pumps. Two 100% capacity non-safety related Steam Generator Blowdown Pumps are provided per unit. He purpose of these pumps is to transfer water that has been separated in the blowdown tank to the blowdown recovery heat exchangers. The Steam Generator Blowdown Pumps are not nuclear safety related. In an effort to improve pump reliability, prior modifications CE-8261 and CE-8449 were performed to change Pumps 2A and 2B, respectively, from mechanical seals to packing. His change has had limited success and frequent packing adjustments have been required to maintain leakage at acceptable levels. At the same time, improvements in pump venting prior to starting, seal flush pipe arrangement, and careful alignment checks during re-assembly have resulted in increased mechanical seal reliability on the Unit 1 Pumps. In an effort to restore reliability to the Unit 2 Steam Generator Blowdown Pumps, shaft packing is being replaced by mechanical seals. This modification changes the pumps back to their original configuration. In addition, a new single stationary seal (self-aligning stationary seal design) will be used to improve seal reliability during Steam Generator Blowdown System operating transients. 1 Evaluation: Hese pumps are not required by any Technical Specification and are isolated in the Stean) Generator Tube Rupture analysis presented in UFSAR Chapter 15. There are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. A revision will be required for the UFSAR Figure 10-31. I l I i l l l

i U.S. Nuclear Regulatory Commission April 1,1999 Page 25 of 247 l l 25 Type: MinorModification Unit: 2 Utle: Minor Modification CE-09073 and associated Variation Notice VN-9073A

Description:

Minor Modification CE-9073 and associated Variation Notice VN-9073A made a change to the Annulus Ventilation System and then removed the change. During performance of an 18-month surveillance for the Annulus Ventilation System, the airflow rate for Annulus Ventilation Filter Unit 2A (AVFU-2A) was found to be below the acceptance criteria. The filters within the filter unit had been replaced during the most recent Unit 2 refueling  ! outage and the system subsequently met retest criteria including the airflow rate criteria. Several factors contributed to the low flow condition. 'Diese factors were: (1) the pressure drop across the carbon filter bed had increased, (2) the pressure drop across the prefilter/ moisture eliminator had increased, and (3) a manual volume damper (MVD) in the supply header ductwork had failed closed. In order to increase the Annulus Ventilation System flow rate, several changes were considered. These changes included the removal of the downstream HEPA filters and the removal of the existing restrictor plates in the reiurn header ductwork. The downstream HEPA filters were removed per . minor modification CNCE-9073 and the restrictor plates were removed under an l associated work order. After the MVD was identified as the root cause of the low flow condition, a decision was made to reinstall the downstream HEPA filters in order to keep the configuration of AVFU-2A consistent with all the other carbon filters at the plant. Reinstallation of the HEPA filters was done per Variation Notice VN-9073A. It was decided not to reinstall the restrictor plates since they were 'not necessary and had already , been removed on the B-Train Header. When the failed MVD was re-opened and returned ) to its design position (full open) the Annulus Ventilation System flow increased to the l nominal flow range. l Evaluation: Removal of the ductwork restrictor plates will not adversly affect the ability of the Annulus Ventilation System to perform its design function. There are no unreviewed l safety questions as a result of this modification. No Technical Specification clamges are required. No UFSAR changes are required. I i l l' l l

U.S. Nuclear Regulatory Commission April 1,1999 Page 26 of 247 26 Type: MinorModification Unit: 0

Title:

Minor Modification CE-09119. Recycle Evaporator Concentrates Filter Fine Filtration

Description:

Minor Modification CE-9119, Recycle Evaporator Concentrates Filter Fine Filtration, allows use of smaller micron rated filter cartridges in the Recycle Evaporator Concentrates Filter. In addition to the 25 micron filter elements currently reflected in site documents,1.0,0.45, and 0.1 micron cartridges (manufactured by Pall) and 3.0,1.0, and 0.45 micron filters (manufactured by BWNT) will also be approved for use in the concentrate filters. The high temperature alarm setpoint for 0NBEM5750 will be reduced to 175 degrees F. to accomodate the lower design temeperature off the Pall filters. Evaluation: No unreviewed safety question is created as a result of this modification. The new filters meet or exceed the the design parameters specified in the UFSAR. The setpoint change for 0NBEM5750 is in the conservative direction. No Technical Specification changes are required. Changes are required for UFSAR Section 9.3.5.2.1.6 and Table 9-24. 27 Type: Minor Modification Unit: 1 Titic: Minor Modification CE-09154, Replace valves 1KC-463 and iKC-466 with new valves

Description:

Minor Modification CE-9154 replaces valves IKC-463 and IKC-466 with new valves that are acceptable replacements. Valves IKC-463 and IKC-466 are component cooling system valves. The valves currently installed are gate valves. Both valves have had significant leakage problems. The valves will be replaced with new buttedly valves. Evaluation: The replacement valves have been evaluated by Engineering to be acceptable replacements for the currently installed valve. The two valves do not have any function in an accident and are no considered accident initiators. No unreviewed safety question is created as a result of this modification. No Technical Specification changes are required. A UFSAR change is required for the system flow diagram (Figure 9-41). l

g-O

  ~ U.S. Nuclear Regulatory Co--s a-
  ' Apdf 1,1999' Page 27 of 247 i

28 - Type: Minor Modification Unit: I

              'ntle: Minor Modification CE-09193, Remove Oil Piping from the Unit i Reactor Coolant Pump Motor Bearings

Description:

Minor Modification CE-9193, Remove Oil Piping from the Unit 1 Reactor Coolant Pump

                     - Motor Bearings, deletes unused oil fill and drain piping. On all four reactor coolant pump motors piping between vendor supplied valves on the upper bearing fill and drain line and lower bearing fill and drain line and flexible nr'.al hose connections will be deleted. The flex hose will be deleted and the remaining pip ng will be capped with blind flanges.

, Screwed plugs will be used to cap the vendor supplied valves. Since the piping to be deleted is supported with steel members attached to the motor, this support steel will also be removed. Evaluation: The oil piping and associated support steel is being removed to facilitate the removal and replacement of the reactor coolant pump motors. The deleted piping was not used for filling the oil reservoirs. Drainage of the oil will be performed by alternate means. The new configuration has been analyzed and found to be accepable with reference to piping stress analysis and support restraint steel. There is no unreviewed safety question as a result of this modification. No Technical Specification changes are required. Changes are required for UFSAR Figure 5-6.

29 Type: MinorModification Unit: 1

Title:

Minor Modification CFA9231. Redistribute Airflow in Lower Containment

Description:

Minor Modification CE-9231, Redistribute Airflow in LowerContainment, modifies the . air distribution pattern in IAwer Centainment. During the Ush 1 Steam Generator Replacement Outage new insulation was installed on the Steam W jenerators and associated

                      - piping which reduced temperatures within the Steam Generator cavities. Hot piping near the crane wall openings was also relocated. This hot piping was rear the temperature monitoring location for Technical Specification 3.6.1.5. These changes resulted in reduced Lower Containment temperatures during Mode I (Power Operation). During
                     ' Unit 1 Cycle 10 !ower than expected temperatures were noted in Lower Containment Quadrant D. It was determined that this was caused by cold air flowing directly on the Quadaant D temperature sensors. His modification redistributes supply airflow throughout Iower Containment and blocks the airflow that was flowing directly onto the Quadrant D temperature sensors.
      - Evaluation: No unreviewed safety question is created by the modification. The modification will l                       ensure that temperature sensors are sensing the actual lower Ccontainment temperature.

No Technical Specification changes are required. No UFSAR changes are required. L l' i l L .. J

U.S. Nuclear Regulatory Commission Apdf 1,1999 Page 28 of 247 - 31 Type: Minor Modification Unit: 1

Title:

Minor Modification CE-09289, Leak repair on the first root valve for 1CFLT5551

Description:

Minor Modification CE-9289, Leak repair on the first root valve for instrument ICFLT5551, allows a leak repair option for this root valve. This valve presently has body to bonnet leakage which is to be repaired by use of a sealant material which will be injected into the body of the valve. The valve is a three quarter inch T-type globe valve. Other alternatives for repair of the valve were considered and this was determined to be the best method for repair. This repair ruethod is procedurally controlled, reviewed by Engineering and is a normal maintenance activity. There will be no effect on the Condensate Feedwater System to which this valve is attached. Evaluation: 'Ibs repair will not affect the function of the valve. The valve will be tested after the leak repair to ensure it is still functional. No unreviewed safety question is created by the modification. No Technical Specification Changes are required. No changes to the UFSAR are required. 32 Type: Minor Modification Unit: 1

Title:

Minor Modification CE-09290, modify Reactor Coolant Loop IC low pressure flow transmitter tap connection for INCFr5070 l

Description:

Minor Modification CE-9290, modifies the Reactor Coolant Loop IC low pressure flow transmitter tap connection for instrument INCFT5070. This modification will replace the existing Metal Bellows Corporation flex hose with a stress loop made of one halfinch stainless steel tubing. The existing flex hose is leaking and a replacement flex hose is not available. Evaluation: No unreviewed safety question is created by this modification. The stress loop is an equivalent to the flexible hose. No Technical Specification changes are required. No UFSAR changes are required. 9 l l

\  ! U.S. Nuclear Regulatory Co==I= ton

     - April 1,1999 Pase 29 of 247-9 33    Type: Minor Modification                                                  Unit: 1 -
                 'I1de: Minor Modification 'CE-09292, modify Reactor Coolant loop 1C low pressure flow transmitter tap connection for INCFT5070

Description:

Minor Modification CE-9292, modifies the Reactor Coolant Imop IC low pressure flow transmitter tap connection for INCFT5070. This modification will replace a stress loop made of use half inch stainless steel installed per modification CE-9290 with a flexible i

      ,                  . hose. A flex hose was originally used in this application.

Evaluation: No unreviewed safety question is created by this modification. A flexible hose was originally ustxt in the application. 'Ihis modification will return the installation to its original configuration. No Technical Specification changes are required. No UFSAR changes are required. J

         '34    Type: Minor Modification                               -

Unit: 1

Title:

Minor Modification CE-09305, Modify the circuit for solenoid valve 1MISV5233 to prevent spurious closure

Description:

Minor Modification CB-9305, Modify the circuit for solenoid valve IMISV5233 to

                        ~ prevent spurious closure, will prevent a problem which is being caused by a faulty reed

switch.' Spurious closure of IMISV5233 renders radiation monitors IEMF38, IEMF39, .

and IEMF 40 inoperable. The circuitry for this valve will be modified to change the seal ' in circuitry for the valve. Evaluation: "Ihe modification will not change any parameters that affect the safe operation of the . L plant. These wiring changes will not prevent the valve from performing its safety function

                        ' which is to close upon receipt of an St signal for containment isolation. No unreviewed safety question is created by the modification. No Technical Specification changes are required. No UFSAR changes are required.

( I i  ! i 1

p  ! l I< = U.S. Nuclear Regulatory Commission

     ' April 1,1999 '                                                                                              l j-    < Page 30 of 247 9

30 Type: Minor Modification Unit: 0

                'Iltlei Minor Modification CE-09316 and CE-9317, leave Fuel Pool Ventilation System Train Motor operated isolation dampers in the open position at all times

Description:

Minor Modification CE-9316 and CE-9317. Ieave Fuel Pool Vent 9ation System Train Motor operated isolation dampers in the open position at all times, . Slows these dampers to remain open. An option is added to allow the back draft damper counterweight to be tack welded to the damper shaft. The outside air damper control interlock with the supply and exhaust fans will be changed. 'Ihis change will allow the er.haust fan to maintain outside air ventilation even if the supply fan is not operable. 1 Evaluation: No unreviewed safety question is created by these modificat'.ons. The postulated accidents for the Spent Fuel Pool analyzed in the UFSAR are not affected by this modification. No Technical Specification changes are required. A revison to UFSAR Figure 9-118 will be required. 35 Type: Minor Modification Unit: 0 i

Title:

Minor Modification CE.09320, Revise Reactor Coolant System and Chemical and l Volume Control System Design Basis Documents and revise UFSAR description of boron addition to shutdown margin i i

Description:

Minor Madification CE-9320, Revise Reactor Coolant System and Chemical and Volume Control System Design Basis Documents and revise UFSAR description of boron addition to shutdown margin, revises various editorial and technical discrepancies. These documents describe boration to cold shutdown margin prior to initiation of reactor j coolant system cooldown. The actual practice is to verify shutdown margin at the target cooldown temperature. Evaluation: There is no unreviewed safety question associated with this modification. No technical specification changes are required. Changes are required for UFSAR Section 9.3.4.2.4.3. l-l 1 I L

n fL .' U.S. Nuclear Regulatory Conunission l' Apdf 1,1999 Page 31 of 247

i. r L 66~ Type: MinorModification. Unit: 1 l-

                  'I1tle: Minor Modification CE-09326, Replace valves IWi-313 and IWl 314 with new valves l.

Description:

Minor Modification CF-9326, Replace valves !Wi-313 and IWi-314 with new valves, !- will change these valves for a different type valve. The currently installed valves are plug l valves. They are located in a trench and are difficult to operate. %e valves are being replaced with a three inch diaphragm valve. Evaluation: %ese valves are normally closed and do not serve n ' safety function. The difference between the new valves and the previously installed valves have been evaluated and found to be acceptable. There are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. A change is required for a UFSAR Chapter 11 Figure (flow diagram). 92 Type: MinorModification Unit: 1

Title:

Minor Modification CE-09327, Perform NRC Generic letter 89-10 Testing on various Motor Operated Valves and replace actuators for valves IKC0003A and IKC018B.

Description:

Minor Modification CF 9327 performs NRC Generic letter 89-10 Testing on various Motor Operated Valves and replaces actuators for valves IKC003A and IKC018B with slower actuators. NRC Generic letters 89-10 and 96-05 require a higher level of operability determination, maintenance, and surveillance of critical motor operated valves (MOVs). The following valves are affected by this Minor Modification: IFWO27A, IFWOSSB, IKC003A, iKC018B, IKC040B, I RN0028, IVX001 A, and IVX002B. Meeting the requirements of Generic Letters 89-10 and 96-05 will require revisions to CNM.1205.00-1997 001 " Torque Switch Setting Sheets", which serves as the source for valve testing and set-up data. New thrust or torque set-up windows are being established , to increase each MOV's margin for operation. In addition, new actuators are being installed on iKC003A and 1KC018B. A new spring pack is also being installed in the actuator for 1RN002B. These modifications are also being performed to increase each l MOV's margin for operation. i Evaluation: There is no change in the operation of the valves or associated systems due to this modification. %e valves will function when called upon just as they did prior to this modification. No new failure modes have been created as a result of this modification.

                          %is modification does not change the fit, form, and function of the MOVs involved.

There is no Unreviewed Safety Question associated with this modification. UFSAR changes are required for Figure 9-35,(piping flow drawing). No Technical Specification changes are required. I- I I ( j

i I l U.S. Nuclear Regulatory Commission April 1,1993 Page 32 of 2C

  .72     Type: Minor Modification                                             Unit: 2

Title:

Minor Modification CE-09329, Perform NRC Generic Letter 89-10 Motor Operated Valve thrust testing of various Motor Operated Valves -

Description:

Minor Modification CE-9329, Perform NRC Generic letter 89-10 Motor Operated i Valve thrust testing of various Motor Operated VtJves, will test various Motor Operated l Valves and replace the actuator on valve 2KC003A, and the motors on valves 2N1076A, l 2N1088B,2 Nil 83B. NRC Generic letters 89-10 and 96-05 require a higher level of l operability determination, maintenance, and surveillance of critical motor operated l valves. Ap;wximately 69 valves are affected by this minor modification. Meeting the requirements of these generic letters will require revisions to CNM-1205.00-1997 001,

                  " Torque Switch Setting Sheets," which serves as the source for valve testing and set-up data. New thrust or torque set-up windows are being established to increase each Motor Operated Valve's margin for operation. A new actuator will be installed for valve 1                  2KC003A and new motors will be installed on valves 2NIO76A,2NIO88B,2 Nil 83B. A l                  new spring pack will be installed in the actuator for KC429B. These modificcions are being performed to increase each Motor Operated Valves margin for operation.

! Evaluation: There is no change in the operation of the valves or associated systems due to this modification. The valves will function when requiredjust as they did prior to the modification. No new failure modes have been introduced.The modification does not change the form, fit, or function of the valves. There are no unreviewed safety questions as a result of this modification. No Technical Specification changes are required. No l UFSAR changes are required. l l

o i U.S. Nuclear Regulatoay Co==d= ton April 1,1999 Page 33 af 247. 36 Type: MinorModification Unit: 1

            'Hele: Minor Modification CE-09349, Containment Ventilation Chilled Water System Supply Temperature Range Adjustment

Description:

Minor Modification CE-9349, Containment Ventilation Chilled Water System Supply Temperature Range Adjustment, changes the setpoint of the Containment Ventilation Chilled Water System Chillers. The modification will change the current setpoint of 45 degrees Fahrenheit to a range of 42 degrees to 48 degrees. During the Steam Generator Replacement Outage, new insulation was installed on the Steam Generators and

                  . associated piping which reduced temperatures within the Steam Generator cavities. Hot piping near crane wall openings (the location of temperature monitoring instrumentation for Technical Specification 3.6.1.5) was also relocated. These changes resulted in reduced temperature in the Unit i Lower Containment during power operation. The Containment Ventilation Chilled Water System is a major factor in controlling the average lower containment air temperature. He system operates with an average chilled water supply temperature of between 42 and 48 degrees F. De temperature control function of the Lower Containment Ventilation Units was originally designed for a nuclear service water sytem supply temperature of 88 degrees F., thus creating additional problems in controlling Lower Containment temperatrure. A wider setpoint range will allow for more flexibility in controlling the containment average temperature.

Evaluation: He Containment Ventilation and Chilled Water Systems will continue to maintain an  !

                  . acceptable temperature within the Upper and fewer Containment areas. %e                   !

environmental qualification of equipment in these areas will not be adversely affected. l

                    %e Containment peak pressure and temperature analyses will not be adversely affected.

Operation of the reactor coolant pumps will not be adversely affected. There are no Unreviewed Safety Questions associated with this modification. No Technical Specification changes are required. No UFSAR changes are required. j l i l l

y l l !: ' U.S. Nuclear Regulatory Conendssion

. April 1,1999

     ' Page 34 of 247.-

37 Type: MinorModification Unit: 0

Title:

Minor Modification CB-09350, Containment Ventilation Chilled Water System Supply Temperature Range Adjustment

Description:

Minor Modification CE-9349, Containment Ventilation Chilled Water System Supply

l. - Temperature Range Adjustment, changes the setpoint of the Containment Ventilation .

Chilled Water System Chillers. The modification will change the current setpoint of 45 j i degrees Fahrenheit to a range of 42 degrees to 48 degrees. During the Steam Generator l  ! Replacement Outage, new insulation was installed on the Steam Generators and I associated piping which reduced temperatures within the Steam Generator cavities. Hot - piping near crane wall openings (the location of temperature monitoring instrumentation for Technical Specification 3.6.1.5) was also relocated. These changes resulted in reduced temperature in the Unit i 14wer Containment during power operation. De Containment Ventilation Chilled Water System is a major factor in controlling the average lower containment air temperature. The system operates with an average chilled

                          . water supply temperature of between 42 and 48 degrees F. The temperature control function of the lower Containment Ventilation Units was originally designed for a n'uclear service water sytem supply temperature of 88 degrees F., thus creating additional problems in controlling lewer Containment temperatture. A wider setpoint range will allow for more flexibility in controlling the containment average temperature.
         ~ Evaluation: The Containment Ventilation and Chilled Water Systems will continue to maintain an                I acceptable temperature within the Upper and Lower Containment areas. The                      i environmental qualification of equipment in these areas will not be adversely affected.      ;

The Containment peak pressure and temperature analyses will not be adversely affected. i Operation of the reactor coolant pumps will not be adversely affected. There are no Unreviewed Safety Questions associated with this modification. No Technical Specification changes are required. No UFSAR changes are required. l 214 Type: MinorModification Unit: 1 ! 'i

Title:

Minor Modification CE-09367 I

Description:

Minor Modification CE-9367 removes a straightening section from Air Flow Monitor 1 ABSAFMD-1 in the Auxiliary Building Ventilation System supply ducting. He  : straightening section is structurally shaped like a honeycomb with small openings that are being blocked with dirt and debris. After removal of the straightening section the air  ; ' . flow monitor will continue to perform its design function. - After removal of the straightening section the seismic integrity of the air flow monitor will be maintained as a l~ ducting enclosure for the passage of air both during normal and accident conditions. i Evaluation: The Auxiliary Building Ventilation System will continue to serve its safety function wkh the straightening section removed. De removal has 1,een evaluated from a seismic design perspective and found to be acceptable. No unreviewed safety questions are created by

                         &is modification. No Technical Specification changes are required. No UFSAR changes are required.

l

U.S. Nuclear Regulatory Commission - Apdf 1,1999 Page 35 of 247 38 Type: MinorModification Unit: 0

Title:

Minor Modification CE 09411. Revise the equipment database and flow diagrams to correct locking valve designations 1

Description:

Minor Modification CE.9411, Revise the equipment database and flow diagrams to l correct locking valve designations, revises valve locking information for several valves. Evaluation: There is no Unreviewed Safety Question associated with this modification. The changes

• are editorial in nature. No Technical Specification changes are required. UFSAR ,

changes are required for various UFSAR Figures (flow drawings). ) 1 39 Type: Minor Modification ' Unit: 0  ;

Title:

Minor Modification CE-09416, Revise Liquid Radwaste System Design Basis Documents and UPSAR description of heat tracing power supply.

Description:

Minor Modification CE-9416, Revise Liquid Radwaste System Design Basis Documents i and UFSAR description of heat tracing power supply, clarifies that heat tracing is not supplied by the Blackout Power System. Power supplies for heat tracing on the liquid radwaste and solid radwaste systems have been reviewed . Redundant heat tracing is provided. The primary and backup heat tracing are powered from Motor Control Centers IMXK and SMXB respectively. Neither Motor Control Center is fed by an emergency, blackout, or battery backed power source. Evaluation: This system is not used for mitigation of any accident described in the UFSAR. There is no Unreviewed Safety Question associated with this modification. No Technical Specification changes are required. UFSAR changes are required for Sections 11.2.1, 11.2.2.3 and 11.2.2.7.3. I l f-L:

U.S. Nuclear Regulatory Commission April 1,1999 l Page 36 of 247 l 40 Type: Minor Modification Unit: 0

Title:

Minor Modification CE-09417. Revise the equipment database and flow diagrams to correct locking valve designations and add locks to valves 1(2)NM393 and IWL019.

Description:

Minor Modification CE-9417, Revises the equipment database and flow diagrems to correct locking valve designations and add locks to valves 1(2)NM393 and IWL019. The ANSI B31.1 code requires that there be no intervening stop valves between the relief device and the piping which it is intended to protect. This includes relief valve discharge piping. To comply with this requirement, the intervening valves will be locked open to prevent an inadvertent mispositioning. Valves 1(2)NM393 are in piping which provides over-pressure protection for the Post Accident Liquid Sampling System Panel. Valve 1WL109 is in piping which provides a relief path / valve stem leakoff to the waste evaporator feed tank sump. Evaluation: This modification does not change the plant as described in the UFSAR. The valves are currently in the correct position and this modification will ensue that they are not inadvertently mispositioned. There are no Unreviewed Safety Questions associated with this modification. No Technical Specification changes are required. Revisions to UFSAR Flow Drawings will be required. 4

g p l l !. 1 I U.S. Nuclear Regulatory Co==le= Ion

      - April 1,1999 --

Page 37 of 247 l f L -. 41 Type: Minor Modification Unit: 0 Iltle: Minor Modification CE-09428, Revise Nuclear Service Water System controlled l documents i

Description:

Minor Modification CE-9428, Revise Nuclear Service Water System controlled documents, made several changes to Nuclear Service Water System design documents. l The modification revised instrument functional requirements on data sheets for Nuclear Service Water Pump pit level instrument loops 1/2RN7400,1/2RN7370, ORN7390 and l ! ' OkN7360.' Design calculations have shown that the instruments have a total loop ! ancertainty of +/- 1.833 feet and that a total uncertainty of +/- 2.1 feet can be tolerated for these loops based on the current setpoints utilized in station procedures. Instrument

  -                        Functional Requirements on data sheets for Nuclear Service Water to Containment Spray heat exchanger flow instrument loops 1/2RN5800 and 1/2RN5850 are being revised.

Design calculation CNC 1223.24-00-0012 determined that the normal operating temperature for Nuclear Service Water discharge from the Containment Spray heat l exchangers increased from 120 degrees F to 125 degrees F due to changes in the l Technical Specification limit for the Standby Nuclear Service Water Pond temperature [ that were performed previously. In addition, two editorial changes are being made. Valve

Design Parameters (Generic letter 89-10) data sheet for Nuclear Service Water system valves 1(2)RN144A and 1(2)RN225B are being revised to show the Normal Valve position as closed. Also, the reference to the Lower Containment Ventilation Units and the Upper Containment Ventilation Units as safety related in the Nuclear Service Water Design Basis Document is being deleted since those items are not safety related. j

\ \ Evaluation: The Nuclear Service Water system is the ultimate heat sink for various Nuclear Safety Related heat loads during normal operation and design basis events. The Nuclear Service Water system supports Emergency Core Heat Removal operation by providing cooling to the Component Cooling System via the Component Cooling heat exchangers and also to the Diesel Generators via the Diesel Generator Engine Jacket Water Cooler system heat exchangers. Other Nuclear Safety Related loads include the Containment Spray heat exchanger and the Control Room Chilled Water System Chiller Condenser. The Nuclear Service Water system also provides assured makeup to the Component Cooling system. The Spent Fuel Pool, Auxiliary Feedwater supply and the Containment Seal Water Injection system. His modification does not increase the probability of an ac:ident evaluated in the SAR. He Nuclear Service Water System and the Standby Nuclea: Service Water Pond are not identified as accident initiators, but rather support safety j related systems. less of Nuclear Service Water or the Standby Nuclear Service Water Pond could result in a loss of Component Cooling. His could r:sult in reactor coolant I pump seal LOCA via a loss of charging flow (from a charging pump motor cooler failure) and a Reactor Coolant pump thermal barrier heat exchanger failure. However, this i modification does not increase the probability of a loss of Nuclear Service Water or the Standby Nuclear Service Water Pond. The inventory of the Standby Nuclear Service Water Pond and the Nuclear Service Water f,ystem are unaffected as is the water temperature assumed for the Standby Nuclear Service Water Pond and the Nuclear Service Water System in design calculations. , l This modification does not increase the probability of a malfunction of equipment I important to safety evaluated in the SAR. Since this modification does not increase the i l l L

g . I U.S. Nuclear Regulatory Convaission l April 1,1999 ' l Page 38 of 247 1 l l , l water temperature of the Nuclear Service Water system, the cooling needs of the Safety Related equipment supplied by Nuclear Service Water are maintained. In doing so, the qualification of this equipment is unaffected. The swapover of the Nuclear Service Water { l system to the Standby Nuclear Service Water Pond has been evaluated and the increased i error associated with the swapover setpoint does not result in a degradation of the Nuclear Service Water pump performance or Nuclear Service Water valve actuation. l This modification does not create the possibility of an accident or malfunction of a different type than those evaluated in the SAR. There are no new failures postulated that would result in a different type of accident or equipment failure for this modification.  ; The performance of the Nuclear Service Water pumps is assured by the evaluation of Net I l Positive Suction Head , submergence requirements and valve performance. The normal  ! temperature exiting the Containment Spray heat exchanger is consistent with assumed design parameters. This modification does not increase the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR. Design basis operation of the Nuclear Service Water system is unaffected by this modification. l This modification does not reduce the margin of safety as defined in the basis for any Technical Specification. The Nuclear Service Water system design basis is unchanged by this modification. Evaluation of the Nuclear Service Water pump pit level transient I associated with the pond swap of the Nuclear Service Water system verified that it will { respond as required by the assumptions for all design basis events. The thermal inputs to j the ultimate heat sink are unchanged by this modification. There are no unreviewed 1 safety questions as a result of this modification . No Technical Specification changes are j required. No UFSAR changes are required. [ 1 I 1 i l l-l'

k l' L l l l l f.- U.S. Nuclear Regulatory Conunission - l J April 1,1999 ' Pase 39 of 247 43 Type: MinorModification Unit: 2

Title:

Minor Modification CE-09452, Replace selected Nuclear Service Water System Vent Valves with ball valves and new piping

Description:

Minor Modification CE-9452, Replace selected Nuclear Service Water System Vent - Valves with ball valves and new piping, makes changes in the Nuclear Service Water System to discourage several degradation mechanisms. The Nuclear Service Water _ System small bore piping is extremely susceptible to fouling, clogging, and through-wall pitting. His modification will change out a number of vent valves and their associated piping in the Nuclear Service Water System. These sections will be replaced with new l carbon steel piping from the header to the vent valve and new ball valves. By using a ball l I valve, the piping can be cleaned out if it becomes clogged with silt or corrosion debris.

                     ' his modification will enhance Nuclear Service Water system maintenance operations.

l The configuration of the Nuclear Service Water system will remain unchanged. ( Replacement materials will meet the same ASME or ANSI requirements as the original parts. The only changes to the system flow diagrams are to show a ball valve instcad of a  ; globe valve, increased pipe diameter, if applicable, and to reference the stainless steel j material under the appropriate design parameter. ) i I l Evaluation: There are no Unreviewed Safety Questions associated with this modification. The . changes do not affect the ability of the system to meet its design function or increase the  ! possibility of accidents or equipment malfunctions. No Technical Specification changes - l are required. UFSAR Figure 9-23 (piping flow drawing) will be revised. l 44 _ Type: Minor Modification Unit: 2

Title:

Minor Modification CE-09455, Delete valves 2RN911 and 2RN912 and associated  ; piping from the Nuclear Service Water System l

Description:

Minor Modification CE-9455, Delete valves 2RN911 and 2RN912 and associated piping from the Nuclear Service Water System, removes items that are no longer used. De affected piping was associated with chlorine additions which were previously used to counteract biological fouling. Deleting these items will eliminate any future maintenance

                     ; problems such as through leaks.

Evaluation: The piping is nct referenced in the UFSAR and does not serve a nuclear safety related function. No Technical Specifications changes are required. A revision is required for UFSAR Figure 9-23 (piping flow drawing) h t

U.S. Nuclear Regulatory Comunission

  ' April 1,1999
 ' Page 40 of 247 76 ' Type: Minor Modification '                                                Unit: 2-

Title:

Minor Modification CE-09474 Hotwell Pump 2A and 2B Castridge Type Mechanical

                     ' Seal,

Description:

Minor Modification CE-9474 Hotwell Pump 2A and 2B Castridge Type Mechanical Seal, installs a cartridge seal in place of the component seal on Hotwell Pumps 2A and 2B, De replacement cartridge seal is a tandem configuratiion rather than a double configuration, but also has an inboard and outboard seal to prevent air in-leakage. In addition, the cartridge design allows for replacement without stuffing box removal. The cartndge seal is a more recent techonology and will provide equal or improved seal reliability. Evaluation: The only components involved in this modification are the Hotwell Pumps. Rese Pumps are not nuclear safety related. The replacement seal will provide equal or improved performance. There is no unreviewed safety question associated with this mod. No Technical Specification changes are required. Changes are required for UFSAR Flow Diagrams. 262 Type: Minor Modification Unit: 0

Title:

Minor Modification CE-09492, Changes to the Component Cooling System Design Basis Document

Description:

' Minor Modification CE-9492 makes the following changes to the Component Cooling System Design Basis Document.1. Adds a reference to the design calculations for flexible hoses that attach to Component Cooling System Piping to the Reactor Coolant Pump Oil Coolers 2. clarifes how the Component Cooling System piping is designed to withstand a Reactor Coolant Pump thermal barrier failure. 3. Deletes data in one section about a High Pressure Select Relay because the information is already present in another section. 4. Changes the recommended action statements for valves KC056A, KC057A, KC081 B, and KC082B to agree with a Technical Specification Interpretation. 5. Corrects a mistake which stated that valve KC053B receives no automatic signals when in fact the valve does receive an automatic signal. 6. Clarifies that a traia of Component Cooling is not inoperable when only one train of the assured makeup water is out of service. 7. states that the only exception for valves KC002A, KC 018B, KC230A and KC 228B being open and incapable of closing is when the Component Cooling System is placed in cross train alignment during heat exchanger maintenance. Evaluation: The changes are all considered editorial in nature. There are no unreviewed safety questions involved with this modification. No Technical Specification changes are l required. No UFSAR changes are required. I 1 i

U.S. Nuclear R;. '"ry Conunission Apdf 1,1999 - Page 41 of 247 73 Type: Minor Modification Unit: 0

           'I1de: Minor Modification CE-09494, Install locking devices on Auxiliary Feedwater System bonnet vent valves Descripden: Minor Modification CF 9494, Install locking devices on Auxiliary Feedwater System             '

bonnet vent valves, installs locking devices on 16 valves of the Auxiliary Feedwater System. nc valves are 1(2)CA281,1(2)CA282,1(2)CA283,1(2)CA284,1(2)CA285, 1(2)CA286,1(2)CA287,' 1(2)CA288. Dese are bonnet vent valves which were installed on the ' Auxiliary Feedwater Pump discharge to Steam Generator isolation valves. Installation of a locking device on these one half inch T-type Kerotest valves will ensure I that the bonnet tap valves remain in an open position. %is will prevent the possibility of these valves being mispositioned. Evaluation: There are no Unreviewed Safety Questions associated with this modification. Adding locks to the bonnet vent valves associated with the Auxiliary Feedwater Pump isolation valves is an activity which is not addressed in any manner in the UFSAR and is not a significant plant change which would require inclusion in the UFSAR. The modification will result in enhanced system reliability by ensuring that pressure locking will not occur on the larger isolation valves. No Technical Specification changes are required. A change to UFSAR Figure 10-34 (piping flow drawing)is required. l l l 1 I L1

y i I U.S. Nuclear Regulatory Co===Indon Apsil 1,1999 Pase 42 of 247 74 Type: Minor Modification Unit: 1 i .

Title:

Minor Modification CE-09561. Replace valve INM144 with a new valve. 1 . . . .

l. .

Description:

Minor Modification CE 9561, Replace valve INM144 with a new valve, replaces the currently installed valve with a new valve to improve system performance. *Ihe Nuclear Sampling System provides repn:sentative samples for laboratory analyses which are used

                            ' for guidance in the operation of various primary and secondary systems. A sampic room is provided for each Unit. Sample lines originating within the Containment are provided with remote, motor-operated containment isolation valves, located both inside and outside l                              the Containment, which are closed automatically by a Containment isolation signal in the event.of a LOCA. In addition, a manual valve is located close to each sample source, and manually operated valves for flow control and isolation are located in the sample room.

Valve INM144 is one of these isolation valves located in the sample room. l This Minor Modification replaces valve INM144 with a new valve. 'Ihe presently installed valve is a one quarter inch metering globe valve which regularly experiences clogging due to debris / resins at the systera sample point. In order to eliminate this f problem the replacement valve will be a ball valve design with a larger orifice. !- The new ball valve will have larger flow coefficient than that the currently installed valve. l Past experience has shown that this increase in flow coefficient will not cause a personnel f safety hazard with respect to the manageability of the Waste Evaporator Feed Tank (WEFT) sample flow. Troubleshooting efforts have shown that with the valve removed,

the sample flow which originates at the discharge of the WEFT pumps was easily j managed and was not a safety concern.

l Evaluation: Replacement of valve INM144 is a maintenance activity. This valve is a sample point isolation valve and is not specifically addressed in the UFSAR. Replacing this valve with a more suitable valve for this application will not increase the probability of occurrence of any accidents evaluated in the UFSAR. No Technical Specification changes are required. A revision is required for UFSAR Figure 9-81 (piping flow drawing) l t

                                                                                                                            ]

l l l I i 1

h , I L U.S. Nuclear Regulatosy Commission

       " April 1,1999 Page 43 of 247 -

75 Type: Minor Modification Unit: 1

    .i           1 ' Hale:' Minor Modification CE-09562. Replace valve IRN299 with a new valve
          - Desenption: Minor Modification CE-9562 Replace valve IRN299 with a new valve, changes the curiently installed valve to one more suitable for the application. De currently installed valve is a one inch Y-Type globe valve. It will be replaced with a one inch ball valve.The valves function is to provide a drain of nuclear service water to the Auxiliary Building Vent Unit I A. ,

Evaluation: This modification replaces a valve with a functionally equivalent item. He replacement valve will provide improved system reliability. The difference between the previously j installed valve and the replacement valve were evaluated and determined to be acceptable. Dere is no unreviewed safety question associated with this modification. No Technical Specification changes are required. A revision is required for UFSAR Figure 9-28 (piping flow drawing). 94 Type: MinorModification Unit: 0

Title:

Minor Modification CE-09584, Remove the Control Room Floor as a commited fire boundary

Description:

Minor Modification CE-9584 removes the control room floor as a commited fire

boundary. He control room floor serves as a boundary between the control room ard the Unit I and Unit 2 Cable Spread Rooms. Both of these areas are a part of the " Control Complex". Additionally, the dedicated safe shutdown train for both the control coom and the cable spread rooms is the Standby Shutdown System. A fire in either the control room or the cable spread rooms would result in reliance on the Standby Shutdown System to bring the plant to a hot standby condition. Therefore removing the control room floor from the NRC committed fire barriers is consistent with the criteria adopted in Selected Licensee Committment 16.9.5 Evaluation: No unreviewed safety questions are introduced as a result of this modification. No Technical Specification changes are required. UFSAR Table 9-31 will be revised.

l '- l I i o I

y o I U.S. Nuclear Regulatory Co== dada = April 1,1999 l _ Page 44 of 247 l l l-l 172 Type: MinorModification Unit: 2

Title:

Minor Modification CE-09601

Description:

Minor Modification CE-9601 installs a test connection with a manual ball valve upstream

                     ' of valve 2KC230A. The test connection is for a temporary pressure transducer which is required to meet differential pressure testing reqirements for valve 2KC230A. These testing requirements are specified by the Joint Owners Group to meet Generic Letter 96-05.' Current drains and vents in the Component Cooling System are not logistically -

compliant according to the standards set forth by the Joint Owners Group. nis modification will install a one inch half coupling, associated piping and ball valve identified as 2KCE26, ite.n number 02D4701, which will be capped during non-test periods. Evaluation: No new failure modes are created by the installation of this modification. This installation will not change the operation of any sytem, structure, or component addressed in the UFSAR There are no unreviewed safety questions associated with this modification. No UFS AR changes are required. No Technical Specification changes are required. 77 Type: Minor Modification Unit: 2

Title:

Minor Modification CE-09624 Modify circuit for 2MISV5231 to prevent improper actuation due to a foaly reed switch Descripden: Minor Modification CE-9624 will rewire the seal-in relay associated with solenoid valve 2MISV5231 to prevent improper actuation. The reed switch which energizes the scal-in relay, has been prone to failure. The wiring changes to be implemented will place the j seal-in relay in parallel with the solenoid valve coil. This will prevent improper valve actuation due to a reed switch failure. The reed switch will still be used to provide open indication on the pushbutton on the Main Control Board. Should the reed switch fail in the new arrangement, the open indication on the Main Control Board will be lost. However, open valve position can be verified by ensuring that a loss-of sample-flow condition for radiation monitors 2 EMF 38,2EMP39, and 2 EMF 40 is not present. Evaluation: nc snodification will not change any parameters that affect the safe operation of tim plant.These wiring changes will not prevent the valve from performing its safety function which is to close upon receipt of an St signal for containment isolation. No unreviewed safety question is created by the modification. No Technical Specification changes are required. No UFSAR changes are required. l t i: 1 I

g L U.S. Nuclear Regulatory Commission L  : /.pril 1,1999 - !. Pane 45 of 247 - l >

  ~

! r

        '267f ' Type: Minor Modification                                                   Unit: 1.

Title:

Minor Modification CE-09635 Replace valve 1RNC42 vent with ball valve and new l piping \. .

Description:

Minor Modification CE-9635 replaces valve IRNC42 vent with ball valve and new piping. The Nuclear Service Water small bore piping is very susceptible to fouling,-

                                                                                                                        -)'

clogging, and through wall pitting. This modification will change out a specific valve (IRNC42), and piping to discourage these degradation mechanisms. The section will be replaced with new carbon steel piping from the header to the vent valve and a new ball valve. By using a ball valve, the piping can be cleaned out if it becomes clogged with silt or corro= ion debris. 'Ihis modification will enhance nuclear service water system maintenance operations. Evaluation: There are no unreviewed safety queastions associated with this modification. The changes proposed do not affect the ability of the system to meet its intended design function or increase the the possibility of an accident or equipment malfunction. No 3 Technical Specification changes are required. A Nuclear Service Water System flow I diagram will be included as a new figure in Chapter 9 of tie UFSAR. 81 Type: Minor Modification Unit: 'l

                  'Iltle: Minor Modification CE-09641, revise locking requirements for Valves INV240 and INV221

Description:

Minor Modification CE-9641 revises locking requirements for Valves 1NV240 "Borie l Acid to Centrifugal Charging Pump Suction Manual Supply" and INV221. SOER 97-01 l required utilities to evaluate potential loss of high pressure injection and charging pump capability from gas intrusion. Valves INV240 and INV221 were identified as potential  ;

                          ' gas intrusuion pathways. These valves are normally closed, infrequently used pathways,        j and special attention is warranted when they are used. This modification involves a           :

physical change to add a lock to valve INV240 and to designate a new locking code to l identify the reason for locking valves INV240 and 1NV221. Evaluation: There is no unreviewed safety question associated with this modification. No safety analysis inputs are affected by these changes. The flow paths affected by locking these valves closed are not utilized in accident analyses. No Technical Specification Changes I are required. No UFSAR margins are reduced. A revision is required for UFSAR flow drawings. I l-L

P l U.S. Nuclear Regulatory Co==latan l April 1,1999 l Pase 46 of 247.. 82 Type: MinorModification Unit: 2

               'I1tle: Minor Modification CE-09644, revise locking requirements for Valves 2NV240 and 2NV221

Description:

Minor Modification CE-9644, revises locking requirements for Valves 2NV240 " Boric Acid to Centrifugal Charging Pump Suction Manual Supply" and 2NV221. SOER 97-01 required utilities to evaluate potential loss of high pressure injection and charging pump ( capability from gas intrusion. Valves 2NV240 and 2NV221 were identified as potential gas intrusuion pathways. Dese valves are normally closed, infrequently used pathways, i and special attention is warranted when they are used. This modification involves a physical change to add a lock to valve 2NV240 and to designate a new locking code to identify the reason for locking valves 2NV240 and 2NV221. l Evaluation: %ere is no unreviewed safety question associated with this modification. No safety analysis inputs are affected by these changes. The flow paths affected by locking these valves closed are not utilized in accident analyses. No Technical Specification Changes are required. No UFSAR margins are reduced. A revision is required for UFSAR flow drawings. 78 Type: MinorModification Unit: 0

Title:

Minor Modification CE-09646, Remove Auxiliary Building Roof Areas from committed fire boundaries

Description:

Minor Modification CE-9646 Remove Auxiliary Building Roof Areas from committed fire boundaries, changes committed fire boundaries. His modification removes the 1- Auxiliary Building roof areas from the scope of committed fire barriers. A significant fire hazard does not exist in the Containment Purge Ventilation System or the Auxiliary Building Chilled Water System Pump Room on the Auxiliary Building Roof. He exterior walls and roof of the Auxiliary Building do not need to be maintained as l committed fire barriers unless a significant fire hazard exists. These areas do not create a I credible fire exposure to the Auxiliary Building Roof. Additionally, the dedicated safe shutdown train for the Auxiliary Building common area at elevation 594'is the Standby Shutdown System (SSS). A fire in either the Containment Purge Ventilation System Supply Room or in the Auxiliary Building Chilled Water System Pump Building that resulted in an exposure to the 594' elevation of Auxiliary Building would not affect the safe shutdown capability of the plant. l Evaluation: There is no unreviewed safety question associated with this minor modification. No technical specification changes are required. No changes are required for the UFSAR.

U.S. Nuclear Regulatory Conunission

 ' April 1,1999 Page 47 of 247 253      Type: Minor Modification                                            Unit: 0

Title:

Minor Modification CE-09660 and CE-09661

Description:

Minor Modifications CE-9660 and CE-9661 will show that the Chemical and Volume Control System positive Displacement Pump breakers are in the disconnect position. Also, the suction valve (NV-470) and the discharge valve. (NV-307 and NV-475) are closed. His modification isolates the positive disp!acement pump from normal alignment and allows the removal of an

• Operable but Degraded" issue relating to the positive displacement pumps.

d Evalua .lon: There are no unreviewed safety questions associated with this change. No safety analysis inputs are affected by these modifications. He positive displacement pumps are not required for accident mitigation. No Technical Specification changes are required. Changes will be required for UFSAR Sections 9.3.4.2.1,9.3.4.2.3.1, and 9.3.4.3.4, Figure 8-1 Figure 8-19 and Figure 9-91. - 86 Type: Minor Modification Unit: 1 Titic: Minor Modification CE-09668, Replacement of Refueling Water Storage Tank Level Transmitters

Description:

Minor Modification CE-9668, Replaces Refueling Water Storage Tank Level Transmitters for Unit 1. Four transmitters will be replaced. The function of the transmitters will not be changed. The replacement transmitters are designed to function while submerged whereas the previously installed transmitters were not. Submergence of the transmitters is possible if the Refueling Water Storage Tank is ruptured. The old transmitters were not qualified for submerger.ce but were assumed to function for the time duration required. Evaluation: There is no unreviewed safety question associated with this modification. The replacement transmitters will perform tie same function as the previously installed transmitters. The ability to withstand submergence will enhance this instrumentation. No Technical Specification changes are required. A change is required for UFSAR Section 6.3.5.4 to enhance the description of the Refueling Water Storage Tank level instrumentation and to correct wording about annunciator logic. i a o

Y [~ U.S. Nreelear Regulatory Connaission i

  .. April 1,1999 6 Page 48 of 247 J l

I

    -215      Type: Minor Modification                                                Unit: 0

Title:

Minor Modification CE-09681, Ice Condenser Intermediate Deck Door Enhancements

Description:

Minor Modification CE-9681 makes revisions to the Ice Condenser Intermediate Deck i- Door drawings, clarifies material descriptions and specifications for future procurement, revises QA condition on specific items and revises the Intermediate Deck Door Frame to Beam joint configuration. These revisions will facilitate the use of an Intermediate Deck Door Mainipulation and Corrective Maintenance procedue which will formalize guidance '

                       . originally contained in a Westinghouse Construction Procedure. The clarification of Material Specifications along with the QA Condition will ensure proper materials are utilized in this application.

Ev.iluation: No unreviewed safety questions are created by this modification. No Technical Specification changes are required. No UFSAR changes are required. 252 Type: MinorModification Unit: 0

Title:

Minor Modification CE-09714, Change to show damper 1(2)ABF D-1,2,8,9 as administratively open

Description:

This modification will revise the Auxiliary Building Ventilation flow diagram, vendor control drawings, and the equipment data listing. Rese design drawings will be revised to show dampers 1(2)ABF D-1,2,8, and 9 remain in their failed position (closed) at all times. De dampers are not locked in the closed position. Evalualion: This modification does not affect the ability of the Auxiliary Building Ventilation System to meet its Technical Specification requirements and UFS AR perfo,mance standards. Dese dampers are not required to function during an accident condition. No Technical Specification changes are required. A change is required for UFSAR Figure 9-123 (Piping flow drawing) . j 115 Type: MinorModification Unit: 2 i

              'Htle: Minor Modification CE-09729 Replacement of 2FWLT5000

Description:

Minor Modification CE-9729 replaces 2FWLT5000 which is the Channel 1 Refueling Water Storage Tank level transmitter. The previously installed transmitter was damaged by lightning. j q Evaluation: His modification does not involve an unreviewed safety question. The new transmitter is functionally equivalent to the one that was installed previously. The function of the transmiiter will not be changed. No Technical Specification changes are required. No j

                      . UFSAR changes are required.

1 l L  ! i

p i N h U.S. Nuclear Regulatory Commission April 1,1999 l Pane 49 of 247 i 238 Type: Minor Modification - Unit: 0

            'Ilde: Minor Modification CE-09736, Identification of bolting material for Fuel Pool j                     Ventilation System Filtered Exhaust Fans

Description:

Minor Modification CE-9736 will identify motor hold down bolts on design drawings for

                   . the Fuel Building Ventilation System filtered exhaust fans (FPXF-1 A1, FPXF-1 A2, FPXF-1B1. FPXF-1B2, FPXF-2A1, FPXF-2A2, FPXF-2B1. FPXF-2B2). This -

modification will allow for the use of the current motor hold down bolts (SAE Grade 2) or a stronger bolt (ASTM A325). In addition motor hold down bolts will have lock washers specified and the bolts will be tightened securely. Revision 1 of CE-9736 identifies torque values for motor hold down bolts on the Fuel Pool Ventilation System Filtered Exhaust Fans.

Evaluation

The Fuel Building Ventilation System is designed to ensure that all radioactive material released from an irradiated fuel assembly damaged during a postulated fuel handling accident will be filtered through the HEPA filters and activated carbon adsorber prior to l dischange to the atmosphere. 'the modification allows an option to use stronger bolting

! material and allows the use of lock washers. This will have no effect on the ability of the l system to perform its function. Postulated accidents involving the Spent Fuel Storage l Building evaluated in the UFSAR consist of a fuel handling accident (dropping a spent i fuel assembly) and dropping of a weir gate onto fuel assemblies in the spent fuel pool. This modification does not affect the probability or consequences of either of these accidents. In additon, the Fuel Pool Ventilation System is not an accident initiator.There

                   . are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. No UFSAR changes are required.

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U.S. Nedear Regulatory Ch=I= Apsil 1,1999 Page 50 of 247 f 116 - Type: Minor Modification ' Unit: 1 Male: Minor Modification CE49745 .

                     . I6. ', ."- : Minor Modification CE-9745 revises piping flow drawings to depict valves ICM6,15, 24,310,359,360 as normally closed. Valves ICM310,359,360 are the main condenser neck expansionjoint seal trough makeup valves. Maintaining the condenser neck seal troughs has resulted in corrosion of turbine extraction piping due to overflow of the trough. Per the manufacturer of the main c=ha , the seal troughs are optional therefore er.aintaining seal trough makeup is optional. The life of the c=ha- neck
                                       . expansion joint is not increased or reduced by seal water makeup. In addition it is desireable to isolate the makeup domineralized water supply to the hotwell pump seals during normal system operation in order to prevent aerated water makeup to the condensate system. The makeup demineralized water supply is not needed during normal operation since a supply on the hotwell pump discharge is provided.
                                                                                                                                              ~

Evaluation: These changes do not affect any systems or components credited in accident mitigation and the probability of failure of the condensate system components is not increased so the probability of an acc' dent evaluated in the UFSAR is not increased. This modification does not involve an unreviewed safety question. No Technical Specification changes are required. No UFSAR changes are required. 117 Type: MinorModification Unit: 2 Mile: Minor Modification CE-09746

Description:

Minor Modification CE-9746 revises piping flow drawings to depict valves 2CM6,15, 24,310,359,360 as normally closed. Valves 2CM310,359,360 are the main condenser neck expansion joint seal trough makeup valves. Maintaining the condenser neck seal troughs has resuhed in corrosion of turbine extraction piping due to overflow of the trough. Per the manufacturer of the main condenser, the seal troughs are optional therefore maintaining seal trough makeup is optional. The life of the condenser neck expansion joint is not increased or reduced by seal water makeup. In addition it is

                                       ' desireable to isolate the makeup demineralized water supply to the hotwell pump seals during normal system operation in order to prevent aerated water makeup to the condensate system. The makeup domineralized water supply is not needed during normal operation since a supply on the hotwell pump discharge is provided.

Evaluation: These changes do not affect any systems or components credited in accident mitigation and the probability of failure of the condensate system components is not increased so the probability of an acci6nt evaluated in the UFSAR is not increased. This modification

                                      ' does not involve an unreviewed safg question. No Technical Specification changes are required. No UPSAR changes are recoired.

1 .: .- i. .

r.m . U.S. Nuclear Regulatory Comadssion April 1,1999 Page 51 of 247 f 164 Type: Minor Modification Urdt: 1

Title:

Minor Modification CE-09763 Remove references to Integrated Leak Rate Test Compressor motor from one line drawings

Description:

Minor Modification CE-9763, removes references to Integrated leak Rate Test Compressor motor from one line drawings. This compressor was planned to be connected to the station blackout power system. The compressor was never installed and the current Integrated Izak Rate Test Plan calls for using a diesel powered compressor with aftercoolers powered by the temporary power system. Here is no field work involved with this modification. Station documents will be revised to incorporate this change. Evaluation: No unreviewed safety questions are created as a result of this modification. No Technical Specification changes are required. A revision is required for UFSAR Figures 8-1 and 8-21, 171 Type: Minor Modification Unit: 1

Title:

Minor Modification CE-09787

Description:

Minor Modification CE-9787 revises design drawings to allow the repair of damaged portions of the ice condenser including the upper blanket panels sealing tape, retainers, and vent curtain seals, thereby improving their material condition. Evaluation: There is no unreviewed safety question associated with this modification. The ice condenser is not an accident initiator. No safety analysis inputs are affected by these changes. No Technical Specification changes are required. A revision to UFSAR Section 6.7.10.2 and Figure 6-167 is required.

U.S. Nuclear Regulatomy Co===da=laa Apeil 1,1999 Page 52 of 247 163 Type: Minor Modification Unit: 0 litle: Minor Modification CE-09791, Ice Basket Top Ring Connection Option

Description:

Minor Modification CE-9791 allows for an optional connection at the Ice Basket Top Ring Connection. The ice condenser basket top ring (rim) originally was connected to the ' sheet metal that forms the ice basket with 12 self tapping screws. The screws were fastened with the head on the outside of the ice basket. The basket metal fits on the outside with the ring on the inside. The primary purpose of'the ring is to stiffen the top of the assembly for the original configuration and to provide an attachment for the cable cruciform suspension system str.r plate J bolts in the modified configuration mentioned next. In the original configuration the ice is supported by internally attached cruciforms. A number of the ice baskets have been modified so that the ice is supported from above by a suspension system (Note that this modification was evaluated and implemented in the past and is not part of the minor modification being evaluated here). Because of vibration during various work activities in the ice condenser, omission in the original installation, or other reasons, a number of the self tapping screws are missing. Westinghouse has been contacted and has stated in a letter dated August 21,1998, (serial DPC 98-037) that for

                                                                                            - the ice baskets as originally designed (internally attached cruciforms) a minimum of three
                                                                                            . screws are necessary, two approximatelf 180 degrees apart, and a third screw somewhere in betwen the other two. Westinghouse has further stated that for the ice baskets that have had their cruciforms modified, a minimum of eight screws need to be provided at any of the Iwelve screw locations. The stipulations from Westinghouse were based on the c

original screw sizes and the original screw material. It is noted that the top rings as described above are located at the top of the nominal 48 foot high ice baskets. l In addition access limitations prevent inserting screws with the screw heads on the outside. Under certain conditions for future work the screws will need to be inserted with j the screw heads on the inside. I This minor modification allows field work to take place in the future and provides documentation of an existing situation. Based on the stipulations provided by Westinghouse this minor modification allows for the connection of the stiffening rim (ring) at the top of the ice basket to have screwed connections as follows: (1) For the ice baskets as originally designed (internally attached cruciforms) a minimum of three screws are necessary, two approximately 180 degrees apart, and a third screw somewhere in between the other two.

                                                                                                                                                                                                                           )

(2) For the ice baskets that have had their cruciforms modified (to be supported from above) a minimum of eight screws need to be provided at any of the twelve screw locations. (3) Screws may be driven so that the head is on the inside rather than the outside. Evaluation: There are no unreviewed safety questions associated with this modification. The ice condenser performs no function under normal operating conditions. During a P : sign Basis Accident steam and air are directed from lour containment through the ice condenser primarily to condense steam and reduce pressure. The steam and air go through passages between the ice baskets and melt ice.

r U.S. Nuclear Regulatory Commission April 1,1999 Page 53 of 247 I For the original design (with internally attached cruciforms) the top ring serves only to 1 stiffen the top of the ice basket. During a Design Basis Accident with the force of steam  ; and air from below the self tapping screws serve to insure the top ring remains in place. i Based on determination by Westinghouse the minimum three self tapping screws are  ; sufficient for this purpose. 1 For the ice baskets that have had their cruciforms modified (to be supported from above) the self tapping screws also have a role in helping to resist upward ice movement by helping to restrain the star plate at the top of the ice basket. Based on determination by Westinghouse eight of the self tapping screws are required for this purpose. The self tapping screws were originally designed with the screw head on the outside of the basket (going through the basket and into the top ring (which is located completely within the metal forming the ice basket). It is not significant whether the screw head is located on the inside or the outside. Westinghouse has agreed that the screw may be located with its orientation in either direction with the screw head either outside the ice basket or inside the top ring. In section 6.7.4.1 of the CNS UFSAR it is currently stated that the ice baskets are designed to minimize any external protrusions which would interfere with lifting, weighing, removal, and insertion. The fact that the minor modification being evaluated here allows for the self tapping screws to be driven from the inside out means that there I will be a small protrusion (probably slightly farther out than the screw head for the design l with the screw head on the outside). All of the small protrusions created with the ahernate detail (screw oriented in the opposite direction) will be at the top of the ice basket. The top of the ice basket is clear of the lattice frames and any interferences; therefore no interferences will be created that would interfere with lifting, weighing, l removal, and insertion. UFSAR Section 6.7.4.1 will be revised to explain and document that the slight protrusion created meets the intent of the stated minimization of protrusions. The ice condenser is physically separated from any areas of the containment where an accident could occur or be caused to occur. The components and structures involved in this minor modification serve no function during the normal operation of the plant and i provide no normal operating support to plant functions or activities. Because of separation and lack of any relationship to normal plant operation there is no increase in probability of an accident because of this minor modificatinn. No Technical Specification 4 changes are required. UFSAR Section 6.7.4.1 will be revised 1 l l l l l l 1 I I l

 ,A.

r  ; i U.S. Nuclear Regulatory Conunission April 1,1999 Page 54 of 247 213 Type: Minor Modification . Unit: 2

Title:

Minor Modification CE-09867 Modify circuit for 2MISV5233 to prevent improper actuation due to faulty reed switch

Description:

Minor Modification CE-9867 will rewire the seal in relay associated with 2MISV5233 to prevent improper actuation due to a faulty reed switch. The reed switch which energizes the seal in relay has been prone to failure. The wiring changes to be implemented will place the seal in relay in parallel with the solenoid valve coil. This will prevent improper valve actuation due to a reed switch failure. The reed switch will still be used to provide an open indication on the pushbutton to the main control board. If the reed switch fails in the new arrangement, the open indication on the Main Control Board will be lost. However, open valve position can be verified by ensuring that a loss of sample flow condition for radiation Monitor 2EMP38,39, and 40 is not present. Evaluation: This modification will not change any operating parameters that will affect the safe operation of the plant. No changes to temperatures, pressures, setpoints, or other parameters that could affect reactivity will occur. The wiring changes associated with the modification will not prevent the valve from going to its fail-safe position (closed) as required upon receipt of an St signal for Containment Isolation. The seal in relay will be energized directly by the Main Control Board pushbutton rather than by the solenoid valve reed switch. There are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. No UFSAR changes are required. 217 Type: Minor Modification Unit: 2

Title:

Minor Modification CE-09872, replace valve 2WL450A and 2WIA51B with an item Number 09J-625 valve

Description:

Minor Modification CE-9872, replaces valve 2WL450A and 2WL451B with new Item Number 09J-625 and 9J-634 valves. These valves are currently item number 06J-588. l The currently installed valves are carbon steel three quarter inch and one inch Y-Type motor operated globe valves. The valves have seat damage. The valve bodies are being replaced with stainless steel one inch Y-type globe valves which are the only replacement bodies available. Reducing inserts will be used and hanger clamps will be sized as required for the necessary piping changes. These two containment isolation valves will continue to serve the same function. i Evaluation: There are no unreviewed safety questions associated with this activity. The replacement of the valve bodies with equivalent components is considered a maintenance activity . The replacement valves will serve the same function as the valves being replaced, therefore no new accidents or increased accident probabilities are created. No Technical Specification changes are required. UFSAR Figure 11-18. Table 3-104 and Table 6-77 j will be revised, i I l ! l ! 4 i L j

g e U.S. Nuclear R,- '"- y Commission April 1,1999 - Page 55 of 247 280 Type: Minor Modification Unit: 0

               'I1tle: Minor Modification CE-09984, Revise the Containment Penetration Valve Injection Water System Design Basis Document and add Surge Chamber Surveillance Requirements Descriptient Minor Modification CB-9984 revises the Containment Penetration Valve Injection Water System Design Basis Document and adds Surge Chamber Surveillance Requirements.

These changes resolve editorial and technical discrepancies . De Design Basis Document change will account for the requirements for the Surge Chamber level Surveillance. The actual practice to verify the level with the surveillance item will remain in the applicable procedures. De Technical Specification reference number will be removed from Procedure PT/1/A/4600/003A Enclosure 13.5 Item 7 and IT/2/A/4600/003A Enclosure 13.4 Item 7. Evaluation: Procedures IT/1/A/4600/003A and PT/2/A/4600/003A, Monthly Surveillance Items, list in Enclosure 13.5 for Unit I and Enclosure 13.4 for Unit 2, a miscellaneous monthly surveillance item on the Containment Penetration Valve Injection Water System (labeled as Technical Specification reference 4.6.6.1) that is not a Current Technical Specification or Improved Technical Specification requirement, but is a requirement for operability per Engineering. The surveilance involves verifying that the surge chamber level is greater than 23% The basis for this verification is to ensure that no gas is introduced in the system. Therefore, this level check should be retained in the procedure. Background information about this which is presently in a Technical Specification interpretation will be deleted with the implementation of the Improved Technical Specifications. The information will be relocated to the System Design Basis document . The procedure reference to a Technical Specification requirement will be deleted. This change does not directly affect any plant equipment. The change involves relocating requirements. There are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. No UFSAR changes are required. l , l i 1 a k

UJ. Nuclear Regulatory Commission - Apdf 1,1999 Pase 56 of 247 281 Type: Minor Modification Unit: 1

           *Iltle: Minor Modification CE-09998, Replace Valve INV282 with a new valve (Item DMV-1079):

Description:

Minor Modification CE-9998 replaces valve INV282 with a new valve. The valve that is presently installed is a stainless steel three quarter inch Y-type globe valve with a belleville disc assembly and seat damage. De valve will be replaced with a stainless steel one inch bellows sealed globe valve. Reducing inserts will be used for the piping size changes. The new valve will perform the same function as the old valve. Evaluation: There are no unreviewed safety questions associated with this modification. Replacement of the valve is considered a maintenance activity (component replacement / improvement). He change will result in improved system availability. ne new valve has been evaluated l to be a suitable replacement. Since the new valve will serve the same function as the old one, no new accidents or increased accident probabilities are created. No Technical Specification changes are required. UFSAR Figure 9-96 (a piping flow diagram) will be revised.

302 Type: Minor Modification -. Unit: 0

Title:

Minor Modification CE-10059 Revise Diesel Generator Lube Oil Sump Tank Drawings to show as-built tank size

Description:

Minor Modification CE-10059 makes drawing changes to show the correct as built size of the Diesel Generator Lube Oil Sump Tanks. During implementation of the Improved Technical Specifications, it was noted that the actual size of the vendor supplied diesel l generator lube oil sump tanks is 579 gallons versus 700 gallons. The vendor supplied tank drawings listed the tank as 700 gallons and this value was used in the development of the UFSAR. This change is considered editorial since the volume of oil required to support emergency diesel operability is not being changed nor is the normal operating system volume being changed. No operating system parameters are being affected by this change. Evaluation: Dere are no unreviewed safety questions associated with this UFSAR change. The i-change is editorial. No Technical Specification changes are required. A change is required for UFSAR Section 9.5.7.2.1. l l 1

V: 5

         ' U.S. Nuclear R;. -Wj c'.---a a..
         ' Apdf 1,1999 E       ,
         -- Page 57 of 247 l

t. 305L Type: Minor Modification Unit: 2 litle: Minor Modification CE 10062, Revise flow diagram to show bonnet vent isolation valve

                            - 2NI488 as normally ' locked closed" l   1

Description:

Minor Modification CE-10062 revises flow diagram CN-2562 1.3 to show bonnet vent isolation valve 2NI488 as normally " locked closed". Valve 2 Nil 85A has a seat leak which will not be repaired because of high cost. This valve was previously modified with a bonnet vent line due to pressure locking / thermal binding concerns. Bonnet vent isolation valve 2NI488 is shown on Flow Diagram CN-2562-1.3 as normally locked in the open position. Because 2 NIX 85A has seat damage and leaks, there is no need for 2NI488 to be open. Operations must issue a restricted procedure change after every Unit 2 outage to control the position of 2N1488 as locked closed. These restricted procedure changes are (and will be) necessary since there are no plans to repair 2NI185A.

                           ~ Flow diagram CN-25621.3 will be revised to show valve 2NI488 as normally locked closed. Once the flow diagram has been revised, Operations will no longer be required to issue restricted procedure changes and what has become the " normal" position for 2N1488 will be consistent with what is shown on the system flow diagram.

Evaluation: Concerns raised by NRC Generic I.etter 95-07 caused Duke Power to analyze certain valves to determine the likelihood of experiencing pressure locking / thermal binding L (PLTB). The response to this generic letter required the plant to install bonnet vent lines on many gate valves. 2 nil 85A was one of the gate valves modified. The vent line for 2 nil 85A contains isolation valve 2NI488. Currently the flow diagram shows this valve as normally locked open to e asure that 2 nil 85 A will not be susceptible to PLTB. 1 Valve 2 nil 85A has a seat leak and will not be repaired due to the large cost of I performing the work. As it ng r.s the seat leak exists, PLTB is not a concern for 2 nil 85A l because there is no mechaniser to trap pressure in the valve's bonnet area. Any pressure build up in the bonnet wot id. be bled to the system via the damaged valve seat. In this manner, seat leakage performs the same function as an open bonnet vent isolation valve. At refueling, Operations must verify whether or not seat leakage is occurring past 2 nil 85A prior to returning the system to service. This leakage test must continue to be performed after this modification to ensure that the seat leak is present and that 2N!488 may remain closed for the next cycle. Changing the normal position designation for 2NI488 from locked open to locked closed effectively returns the system configuration to what it was prior to the bonnet vent modification on 2 Nil 85A. This configuration has been previously analyzed and deemed acceptable. There are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. No UFSAR changes are required. i 4

                                                                                                                         )

4 l l

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                     -t p      U.S. Nedear Regulatory Con-d-la=

Apdf 1,1999 Pase 58 of 247

                                                                                            +

45 Type: MinorModification Unit: 1

             . Iltle: Minor Modification CE-60316, Disconnect power supply and instrument controls for the Incore Instrument Tunnel Booster Fans and remove switches and indication from the Main Control Boards per NUREO 0700 Desedption: Minor Modification CE-60316 disconnects the power supply and instrument controls for the Incore Instrument Tunnel Booster Fans and removes on/off switches and indication from the Main Control Boards per NURI:0 0700. The Incore Instrument hanel Booster -

Fans IITBF-I A and IITBF-1B are located in lower containment as a part of the - Containment Ventilation System. he primary purpose of the Incore Instrument Tunnel Booster Fans was to ensure that adequate air, from the lower Containment Ventilation Units (LCVUs), flowed through the Incore Instrument Tunnel to cool the lower portion of the reactor vessel. It was determined that the LCVUs leve sufficient capacity to supply air flow through the Incore Instrument Tunnel to provide adequate air cooling for the reactor vessel. Therefore, the Incore Instrument Tunnel Booster Fans are no longer needed. Continued speration of these fans presents a maintenance and cost concern. he Incore Instrument i onnel Booster Fans have been out of service (electrically isolated) for several years. -

                       . The power supply and instrumentation controls for the Incore Instrument Tunnel Booster Fans will be disconnected and ON/OFF switches and indication will be removed from the Main Control Boards.

Incore Instrument Tunnel Booster Fans IIMF-IA and IB are identified in UFSAR Chapter 8 Table 8-6 and Chapter 16 (Selected Licensee Commitments) Table 16.8-IA. The Incore Instrument Tunnel Booster Fans (IITBFs) are not addressed in the Technical l - Specifications. he IITBFs do not serve a safety-related function and have no QA I Condition assigned. Since the IITBFs are located in lower containment, this modification will be implemented during refueling outage IEOC10. No equipment or components, other than switches and indication from Main Control Board IMC3 (approximately 2.15 ounces each), are being removed per modification CE-60316, thus no seismic concerns are created. Due to the fact that the IITBFs have been out of service for so long, no safety concerns exists as to the effect the elimination of the IITBFs would , have on any other system or component to perform its designed function. Based on the i above discussion, there are no new f%e modes created by this modification. 1 Evaluation: The Incore Instrument Tunnel Boosta Fans do not serve a nuclear safety related function.  ! No seismic concerns are created. The Incore Instrument Tunnel Booster Fans were not used for any phase of either power generation or conversion or transmission, normal shutdown cooling, fuel handling, or radwaste treatment. No component used for any of i the above activities associated with normal plant operation is affected by this j

modification. The loads on power supplies for the affected circuits are not increased. The l modification will not increase the probability of an accident evaluated in the UFSAR. No l

! new failure modes are created. Compliance with Appendix R criteria is not degraded. l Therefore, this modification will not increase the probability of a malfunction of l l equipment important to safety evaluated in the UFSAR. No Unreviewed Safety Questions I (' are created by this modification. No Technical Specification changes are required. A revision is required for UFSAR Table 8-6 and Table 16.8-1 A , i

n . # U.S. Nuclear Regulatory Conumission Apdf 1,1999 Pane 59 of 247 tc 110- Type: MinorModification Unit: 1

                         'Iltle: Minor Modification CE-60~t76, Disconnect power from heaters DB-H-1 A and DB-H-1B.

Description:

Minor Modification CE-60376 disconnects power from heaters DB-H-1 A and DB-H-1B and abandons the heaters and controls in place. There have been past problems with failure of the Diesel Building Ventilation System duct heaters. The heaters were originally provided to ensure that the minimum Diesel Building temperature was maintained above 60 degrees F. A review of engineering calculation CNC-1211.00 . 0013 and temperature trending data from the Diesel Rooms when the heaters were out of service for extended periods during the winter indicated that the heaters were not required to maintain the minimum room temperature. The heaters are not nuclear safety related and were designed to operate during normal plant operation with the diesels not running. The modification will disconnect power cables from DB-H-I A and DB-H-1B. The heaters and associated controls will be abandoned in place. The power cables will be spared out for future use if needed. i_ Evaluation: This modification will not impact the operation of the Diesel Building Ventilation System or the ability to maintain the required temperature in the Diesel Rooms. The heaters are not accident initiators and serve no accident mitigation function. No unreviewed safety question is created by this modification. No Technical Specification changes are required. p A revision is required for UFSAR Section 9.4.4. 111 Type: Minor Modification Unit: 2

Title:

Minor Modification CE-60377, Disconnect power from heaters DB-H-2A and DB-H-2B.

Description:

Minor Modification CE-60376 disconnects power from heaters DB-H-2A and DB-H-2B and abandons the heaters and controls in place. There have been past problems with failure of the Diesel Building Ventilation System duct heaters. The heaters were originally provided to ensure that the minimum Diesel Building tempa.rature was [ maintained above 60 degrees F. A review of engineering calculation CNC-1211.00 l- g' 0013 and temperature trending data from the Diesel Rooms wien the heaters were out of l< service for extended periods during the winter indicated that the heaters were not required to maintain the minimum room temperature. The heaters are not nuclear safety related and were designed to operate during normal plant operation with the diesels not running. l The modification will disconnect power cables from DB-H-2A and DB-H-2B. He heaters and associated controls will be abandoned in place. The power cables will be spared out for future use if needed. Evaluation: His modification will not impact the operation of the Diesel Building Ventilation System or the ability to maintain the required temperature in the Diesel Rooms. The heaters are not accident initiators and serve no accident mitigation function. No unreviewed safety question is created by this modification. No Technical Specification changes are required. A revision is required for UFSAR Section 9.4.4. l

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Il r-U.S. Nuclear Regulatory Conunission April 1,1999 Page 60 of 247 t 216 Type: Minor Modification Unit: 2

Title:

Minor Modification CF-61024, Disable and Abandon in Place the Containment Purge Ventilation System Humidistats (2VPMT5670 and 2VPMT5680) along with humidity sensors (2VPME5670 and 2VPME5680), and delete associated subcomponents Descriptiont Minor Modification CE-61024, Disable and Abandon in Place the Containment Purge Ventilation System Humidistats and Related Subcomponents, will disable and abandon in place the Containment Purge Ventilation System Humidistats, along with humidity sensors and delete the associated timing relays, optical isolators, and computer points. Evaluation: Abandoning the humidistats and humidity sensors will not affect the existing safety injection or high radiaton containment isolation signals to the containment purge isolation valves. Spurious trip concerns of the humidistats will be eliminated without introducing any challenges to other safety systems. No new failures are introduced. There are no unreviewed safety questions associated with this change. No change to the Technical Specifications is required. Section 9.4.53 of the UFSAR requires a revision. I 5

r U.S. Nuclear Regulatory Cr * %n Apdf 1,1999 j Page 61 of 247 89 Type: MinorModification Unit: 0

Title:

Minor Modification CE-61026, Abandon in place several Control Room Ventilation System Instrumentation Loops

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Description:

Minor Modification CE-61026 will abandon in place several Control Room Ventilation ) System Instrumentation Loops. Air flow monitoring instrumentation loops 1,2 VCLPS280,5290,5300,5310, and 5320 are not being used by Operations or any other station group. Some of the associated transmitters in these loops will not calibrate within the allowable range. This modification will abandon in place instruments 1,2 VCFE5280,  ! 5290,5300,5310, and 5320 located in the Control Room Ventilation System ductwork. Instruments 1,2 VCP5280,5290,5300,5310, and 5320 will be removed from the Main Control Board and the associated cables will be voided and the circuitry deleted. The necessary repairs will be made to the Main Control Boards and new nameplates will be installed to accommodate the remaining control board meters. Evaluation: The Control Room Ventilation System is designed to maintain appropriate temperature, cleanliness, and pressurization of the areas it serves. Operability of the system ensures temperature limits for equipment and instrumentation are maintained within acceptable limits, as well as to maintain a habitable environment for control room personnel during i and after any credible accident conditions. This modification will affect nuclear safety related components (Main Control Board and safety-related duct). Flow-rate indication in the Control Room downstream of the filter units will be maintained. Removal of the control board meters will meet human facters requirements. The abandoned equipment will not affect system operation in any way. This modification does not affect the ) frequency of any UFSAR evaluated accident. Off-site and on-site radiological doses are not affected. Therefore the probability or consequences of an accident evaluated in the UFSAR will not increase. No new credible failure modes or operating characteristics will result from this modification, Therefore, the possibility for an accident of a different type than those evaluated in the UFSAR will not be created. Since the performance of the sy *em will not be affected and challenges to safety systems will not result, the peability or consequences of an equipment malfunction important to safety evaluated in the UFSAR will not increase. The consequences of equipment malfunctions of a different type than those evaluated in the UFSAR will not increase since no new credible failure modes have been identified. The operability of the system is not affected by the , modification. Therefore, the margin of safety as defin.d in the Technical Specification j basis will not be reduced. This modification does not involve an unreviewed safety  ; question. No Technical Specification changes are required. Changes are required for  ; UFSAR Figures 9-108,9-109,9-110,9-111 (piping flow drawings). j I

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     - U.S. Nuclear Regulatory.Connaission April 1,1999 Pane 62 of 247 83    Type: Minor Modification '                                              Unit: 2                          !

Title:

Minor Modification CE-61118, revise Auxiliary Building Ventilation Design

Description:

Minor Modificadon CE-61118 revises Auxiliary Building Ventilation Design to prevent the Auxiliary Building Filter Units from allowing filtered exhausted air to bypass the 1 Filter Units. A potential design deficiency exists in which any single failure that would

                       , prevent the Auxiliary Building Ventilation System from aligning to the filter mode during a design basis accident would appear to place the plant in a condition which is outside the current Dose Analysis. It has not been determined if the dose analysis will be affected by a failure of an Auxiliary Building Ventilation System filter unit to align to its filter mode.

However, to be conservative and to assure proper operation of the system, the Auxiliary Building Ventilation System filter units are considered " Operable but Degraded". He condition for operability is that the Auxiliary Building Ventilation System is aligned to the filter mode. The design basis of the Auxiliary Building Ventilation System is to provide a negative

                       . pressure within the Emergency Core Cooling System (ECCS) pamp rooms so that any radioactive iodines in the pump rooms can be filtered prior to release to the atmosphere.

The Auxiliary Building Ventilation System filter units are the component that provides the filtering mechanism for the system. Normally, the Auxiliary Building Ventilation System is aligned to bypass the filters. However, during a design basis accident significant amounts of radioactive iodines could esist in the pump rooms. These radioactive iodines will be drawn into the Auxiliary Building Ventilation System and should be filtered before being discharged. Although controls exist to place the Auxiliary Building Ventilation System in the fiker mode, certain failures could occur that would prevent the Auxiliary Building Ventilation System from aligning to the filter mode. These failures have not been assumed in the CNS dose analysis and, therefore,'an unanalyzed condition exists. Although it has not been determined whether the dose analysis values would actually be exceeded, it was decided to secure the Auxiliary Building Ventilation System in the filter j mode. This will assure that radioactive iodines from the ECCS pump rooms are filtered prior to being released and assures that the plant is in an analyzed condition. Temporary Station Modifications were used to align the Auxiliary Building Ventilation System in the filter mode and ensure that no failure could cause the system to re-align to j the bypass mode. This was achieved by removing the instrument air supply to the J dampers that are used to align the Auxiliary Building Ventilation System. This Modification will revise the Auxiliary Building Ventilation System design to prevent the l Auxiliary Building Filter Units (ABFU-2A and ABFU-2B) from allowing filtered exhaust p air to bypass the filter units by permanently aligning the Auxiliary Building Ventilation System to the filter mode.- p Evaluation: The Auxiliary Building Ventilation System filtered exhaust subsystem will be secured in

the accident alignment. ThP % the alignment necessary to allow the system to perform its

!.' design basis function and this alignment has already been evaluated in the SAR. The only other system to be impacted by the permanent alignment of the Auxiliary Building Ventilation System filtered exhaust subsystem is the Instrument Air System. Instrunwnt , Air System changes (closing isolation valves and disconnecting lines from the Auxiliary I l 1 l I a J

G i l U.S. Nuclear Regulatory Ceaunission l: Apsil 1,1999 Pane 63 of 247 Building Ventilation System dampers) are not significant. No electrical isolations or deletions will occur per this modification. Thus, the modification will not increase the [ probability of an accident evaluated in the SAR. De ability to manually or automatically i put the Auxiliary Building Ventilation System Filter Units in the bypass mode will be

eliminated. No new failure modes are created. Derefore, this modification will not

                    ' create the possibility of an accident of a different type than those evaluated in the SAR.

The permanent alignment of the Auxiliary Building Ventilation System filtered exhaust ! subsystem in its accident mode does not produce significant challenges to the Auxiliary l Building Ventilation System filter carbon beds. No equipment or devices are being removed per this modification Therefore, no adverse effect on the seismic qualification of the interfacing safety related components exists. Thus, compliance with j environmental, seismic, and Appendix R criteria is not degraded. Therefore, the ~ modification will not increase the probability of a malfunction of equipment important to safety evaluated in the SAR. The modification eliminates the concern of certain failures . occurring that would prevent the Auxiliary Building Ventilation System from aligning to the filter mode, without introducing any other failure modes. Therefore the modification l will not increase the consequences of a malfunction of equipment important to safety l evaluated in the SAR. Likewise, the modification will not create the possibility of a

j. malfunction of a different type than those evaluated in the SAR. Neither any fission L product barrier not any source term calculation is affected. Therefore the modification l will not increase the consequences of an accident evaluated in the SAR. It will not reduce

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the margin of safety as defined in the basis for any Technical Specification, nor does it change any setpoint design limit, or operating parameters. The modification does not l involve an unreviewed safety question. No Technical Specification changes are required. l Changes are required for UFSAR Section 9.4.3.2.3. l \ I i 79 Type: Minor Modification Unit: 0 j

Title:

Minor Modification CE41128, Abandon in place the smoke detectors associated with l Unit I and Unit 2 Control Room Area Air Handling Unit

l.

Description:

Minor Modification CE41128, Abandon in place the smoke detectors associated with Unit I and Unit 2 Control Room Area Air Handling Unit, abandons smoke detectors that  ; i have proven to be unreliable.nc Unit I and Unit 2 Control Room Area Air Handling j l Units are nuclear safety related items that are designed to cool the Control Room Area j l during all plant conditions. Since these unreliable smoke detectors could trip the Air Handling Units without an alarm alerting the Control Room, it was decided to remove the ! smoke detector interlocks. Evaluation: The probability or consequences of an accident evaluated in the UFSAR will not increase i as a result of this modification. The UFSAR references the outside air intake smoke detectors only and they will not be affected by this modification. The are no unreviewed

                    . safety questions associated with this modification. No Technical Specification changes     )

are required. A change is required for UFSAR Figure 9 109 (piping flow drawing). ti

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 ! U.S. Nuclear Regulatory Commission April 1,1999 Page 64 of 247                                                                                            ,
                                                                                                             )

46 Type: Minor Modification Unit: 1

Title:

Minor Modification CE-61137, Provide supplemental chilled water to the Fuel Building . Ventilation System Air Handling Unit Cooling Coil

Description:

Minor Modification CF 61137, Provide supplemental chilled water to the Fuel Building Ventilation Sys em Air Handling Unit Cooling Coil, provides cooling for the Fuel Building during Refueling Outages. This modification provides supplemental chilled I water from the Containment Chilled Water System to the Fuel Building Ventilation System air handling unit cooling coil. Existing Nuclear Service Water piping will be modified so that it can be disconnected from the Fuel Pool Supply Unit. Either Containment Chilled Water System water or Nuclear Service Water System water can be used to supply the Fuel Pool Supply Unit Cooling Coll. The modification includes installation of polyethlene piping and making two pipe penetrations through the Fuel Building exterior wall. This modification will provide a safer working environment during summer outages. . Evaluation: This modification does not prevent any existing system or component from completing its design function in support of plant operation. The modification does not involve any nuclear safety related systems or equipment. There are no unreviewed safety questions as a result of this modification. No Technical Specification changes are required. No UFSAR changes are required. l i 1 I f L

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          .       tU.S. Nuclear Regulatory Commission ,

AprH 1,1999 - Pase 65 of 247 - 114- : Type: Minor Modification ' Unit: 1 11 tic: Minor Modification CE-61258,- Add Vent Valves for Residual Heat Removal System and

                      ,           ' Containment Spray System Piping 3

Desedption: Minor Modification CE-61258, Adds Vent Valves for Residual Heat Removal System and Containnwnt Spray System Piping. Vent lines are provided on the Residual Heat Removal System side of valves INS 18A and INSIB. The vent lines willinclude valves

                                   ~ IND130 (Upstream of valve INS 18A) and IND131 (upstream of valve INSIB). These vent lines are needed because there is a lack of appropriate vents for venting the Residual Heat Removal System Pumps and the Containment Spray System Pumps suction piping.

Currently the operations group must perform a complex alignment that includes removing a flange on a flush line and manually cracking open motor operated valves INS 18A or INSIB. Evaluation: De new vent valves (IND130 and IND131) are one inch 45 degree inclined bellows seal

                                  . globe valves. These valves are stainless steel nuclear safety related valves with design i                            conditions of 2550 psig at 650 degrees F and will meet the design temperature and .

pressure requirements of the interfacing system. De associated vent piping will also meet the design requirements. The new globe valves and piping are suitable for use in tim Residual Heat Removr.1 System and the Containment Spray System as vent valves. The addition of the weight of these components has been evaluated for impact to the stress - analysis and it has been determined that support / restraint modifications are not required. Adding these vent valves will not affect the operation or function of these systems during any phase of of normal or accident mitigation operation. These vent valves simply

                                  . provide a more efficient and convenient method for venting the Residual Heat Removal
                                  . System and Containment Spray System Pumps suction piping. The affected portion of these systems will continue to function as described in the U13AP. Emergency Core Cooling System operation will not be affected by the addition of these vent lines. There
 ..                                 is no unreviewed safety question as a result of this modification. No Technical Specification changes are required. Changes are required for UFSAR Figure 5-17 and 5-18 (piping flow drawings).

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r:  ; U.S. Nuclear Regulatory Commission April 1,1999 ' l Page 66 of 247 257 Type: Minor Modification Unit: 2 11tle: Minor Modification CF-61292, Remove internals from check valves 2CA171 and 2CA172

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Description:

Minor Modification CE-61292 removes the internals (Disc assembly and associated hardware) from check valves 2CA171 and 2CA172. After this modification is completed the valves will no longer require laservice testing and they will be removed from the IST Program. Valves 2CA171 and 2CA172 are in the Nuclear Service Water System supply piping to the Auxiliary Feedwater system pumps. The valves are no longer needed due to the recent installation of check valves 2CA291 and 2CA292 which were install per minor modification CE-61240 to ensure train separation of the Auxiliary Feedwater System , Motor Driven pumps supply from the Nuclear Service Water System. Evaluation: The new check valves will perform the functions previously performed by valves 2CA171 and 2CA172. These functions include preventing gross diversion of flow from one Nuclear ServP e Water System Train Header through a failed header on the opposite Nuclear Serw Water System Train and providing secondary side isolation during an SSS event. Therefore valves 2CA171 and 2CA172 are no longer needed and their { j l internals can be removed without loss of system function. 'Ihe Auxiliary Feedwater j System and its assurred source of Nuclear Service Water will still serve their normal operation and accident mitigation function. 'Ihere are no Unreviewed Safety Questions associated with this modification. No Technical Specification changes are required. No j UFSAR changes are required. i i

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 . U.S. Nuclear Regulatory Comunission April 1,1999 Page 67 of 247

90 . Type: Minor Modification' Unit: 1 l

Title:

Minor Modification CE-61307 Add an access hole in the Unit 1 Component Cooling System Sump Cover.

Description:

Minor Modification CE-61307 adds an access hole in the Unit 1 Component Cooling System Sump Cover for drainage of Component Cooling System components into the Component Cooling System Sump. This modification provides access to the Component i Cooling System Sump on the 522 foot elevation. 'Ihe access hole is needed to provide a means of draining Component Cooling System piping without using 55 gallon storage

                     . drums. Component Cooling System water cannot currently be sent to the radwaste system -

via floor drains due to the chemicals in the water which cause damage to the demineralizers used in the radwaste system. Also the Component Cooling System Sump cover detail will be changed to non nuclear safety related which is consistent with the associated pumps and tanks. l Evaluation: The Component Cculing System serves as an int:rmediate. system between the Reactor l . Coolant System and the Nuclear Service Water System. Each unit has one Component Cooling System Drain Sump with two sump pumps, each located at the 10 west point of the system. Most of the Component Cooling drains are piped to the drain sump and then pumped to the Component Cooling Surge Tank. This minimizes makeup and waste treatment problems associated with the chemically treated Component Cooling System Water. The Component Cooling System Sump does not serve a safety related function. There are no unreviewed safety questions associated with this modification . No technical l specification changes are required. UFSAR Figure 9-40 will be revised (piping flow drawing). l l L I l l i l l i i l l

U.S. Nuclear Regulatory Com Apdl I,1999 Par,e 68 of 247 91 Type: Minor Modification . Unit: 2

Title:

Minor Modification CE41308, Add an access hole in the Unit 2 Component Cooling System Sump Cover.

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Description:

Minor Modification CE41308 adds an access hole in the Unit 2 Component Cooling System Sump Cover for drainage of Component Cooling System Components into the Component Cooling System Sump. His modification provides access to the Component Cooling System Sump on the 522 foot elevation. De access hole is needed to provide a - i; means of draining Component Cooling System piping without using 55 gallon storage

                        ' drums. Component Cooling System water cannot currently be sent to the radwaste system via floor drains due to the chemicals in the water which cause damage to the demineralizers used in the radwaste system. Also the Component Cooling System Sump cover detail will be changed to non nuclear safety related which is consistent with the associated pumps and tanks.

Evaluation: The Component Cooling System serves as an intermediate system between the Reactor Coolant System and the Nuclear Service Water System. Each unit has one Component i Cooling System Drain Sump with two sump pumps, each located at the lowest point of the system. Most of the Component Cooling drains are piped to the drain sump and tien pumped to the Component Cooling Surge Tank. This minimizes makeup and waste treatment problems associated with the chemically treated Component Cooling System water. The Component Cooling System Sump does not serve a safety related function. There are no unreviewed safety questions associated with this modification . No technical specification changes are required. No changes to the UFSAR are required. I l l l l l I

r; , U.S. Nuclear Regulatory Comndssion - April 1,1999 Page 69 of 247 112 Type: Minor Modification Unit: 1

Title:

Minor Modification CE-61309, Provide loop Seal around the Component Cooling Drain Header Valve 1KC432

Description:

Minor Modification CE-61309 Provides a Loop Seal around the Component Cooling Drain Header Valve 1KC432. This is being done to alleviate problems resulting from leaking Component Cooling relief valves on the reactor coolant drain header. During an outage, the Component Cooling drain header inside containment is normally left open , such that any plarmed drainage activities or unplanned / unknown leakage can go to the Component Cooling drain sump outside containment. This drain header is allowed to l remain open as an exception to containment closure except while fuel alterations are in progress. During this time, inputs to the Component Cooling drain header can fill this piping until it is water solid. When the drain piping fills to the point where water enters , i the discharge of the Component Cooling relief valves inside containment, these valves will leak out of the cap screws and other threaded connections. The new loop seal will be used with the existing Component Cooling drain header to provide a path from inside containment to the Component Cooling Drain Sump. This will allow the drain header to remain open during all outage modes including core alterations. Evaluation: There are no unreviewed safety questions associated with this modification. This modification will not prevent any existing system or component from completing its design function. No Technical Specification changes are required. Changes are required for UFSAR Sections 9.2.2.6 and 9.2.2.6.1. l l 1 l 1

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              . U.S. Nuclear Regniatory Conunission

April 1,1999 l" Page 70 of 247 113 Type

MinorModification Unit: 2.

l l *Iltle: Minor Modification CE-61310, Provide Imop Seal around the Component Cooling Drain l Header Valve 2KC432 l

Description:

Minor Modification CB-61310 Provides a loop Seal art und the Congonent Cooling Drain Hender Valve 2KC432. This is being done to allevlste problems resulting from leaking Q==wnt Cooling relief valves on the reactor coalant drain header. During an / l outage, the Component Cooling drain header inside contain ment is normally left open i such that any planned drainsge activities or unplanned /unks own leakage can go to the . l Component Cooling drain sump outside containment. This Wai:a header is allowed to

i. - remain oper an exception to containment closure except while fuel al erations t are in progress. During this time, inputs to the Component Cooling drain header can fill this piping until it is water solid. When the drain piping fills to the point where water enters the discharge of the Component Cooling relief valves inside containment, these valves will leak out of the cap screws and other threaded connections. He new loop seal will be l used with the existing Component Cooling drain header to provide a path from inside

containment to the Component Cooling Drain Sump. This will allow the drain header to remain open during all outage modes including core alterations.

l ! Evaluation: There are no unreviewed safety questions associated with this modification. This l ' modification will not prevent any existing system or component from completing its l design function. No Technical Specification changes are required. Changes are required for UFSAR Sections 9.2.2.6 and 9.2.2.6.1. l 47 Type: MinorModification ' Unit: 1

Title:

Minor Modification CE-61311, Delete valve position transmitters from valves ICF6 and ICF13.

Description:

Minor Modification CE-61311 deletes valve position transmitters ICFVP0060 and ICFVP0130 from valves ICF6 and ICF13. Valves ICF6 and ICF13 are Feedwater Pump Recirculation Valves. The existing transmitters are obsolete and the position signals they provide to the Operator Aid Computer are no longer required. Evaluation: Dese transmitters are not nuclear safety related. They have no effect on any accident evaluated in the UFSAR. The transmitters have no control function. Ahernate position , indication is available. Dere are no Unreviewed Safety Questions associated with this l modification. No Technical Specifications changes are required. A revision is required for

UPSAR Figure 10-27 (piping flow drawing).

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!- 7 - U.S. Nuclear Regulatory Coninaission Apdl 1,1999

       < , Pase 71 of 247.-

t 245 Type: MinorModification Unit: 2 i

Title:

Minor Modification CE-61312 Delete valve position transmitters 2CFVP0060 and 2CFVP0130 from valves 2CF6 and 2CF13 l

Description:

Minor Modification CE 61312 deletes valve position transmitters 2CFVP0060 and j 2CPVP0130 from valves 2CF6 and 2CF13. The existing transmitters are obsolete and the position signals they provide to the Operator Aid Computer are no longer required. Each transmitter is connected to its valve by a mechanical linkage. There are electrical - connections on the transmitter for 120 Volt power and for analog signal input to the Operator Aid Computer. His ' modification mechanically and electrically disconnects each i _. transmitter and removes it from its local panel, Other position indication for valves 2CF6 and 2CF13 is provided on the Main Control Boards.. I~ Evaluation: There are no analyzed accidents that are dependent on Operator Aid Co*rliputer position

indication for valves 2CF6 and 2CF13. These two valve positioners do not provide any control function. There are no unreviewed safety questions associated with this modification. No UFSAR changes are required. No Technical Specification changes are required.

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I 249 Type: Minor Modification Unit: I

                     *Iltle: Minor Modification CE-61313, Addition of Recirculated Cooling Water System cooling line and Conventional Sampling System Heat Exchanger -

Description:

This minor modification will route a supply and return line in the Recirculated Cooling I' Water System to provide cooling for the new Feedwater Corrosion Monitor Sample Cooling Heat Exchanger in the Conventional Sampling System. The piping will originate at the supply header for the Moisture Separator Reheater, First Stage Reheater, and the Second Stage Reheater Drain Tanks and Low Pressure Turbines Crossover Rough Cooling Heat Exchangers. It will be routed to the new heat exchanger in the vicinity of the local feedwater sample point at the Electrochemical Potential Monitor. The return

                             , piping will extend from the new heat exchanger back to the return of the same header.

Isolation valves will be provided in the supply and return lines of the new heat exchanger, and a relief valve will be provided between the two isolation valves. No connections to the Conventional Sampling System side of the Feedwater Corrosion Monitor Sample Cooling Heat Exchanger will be made per this modification. Evaluation: Dere are no unreviewed safety questions associated with this modification. He modification involves components in the Rm:irculated Cooling Water System, with a heat exchanger and connections provided for connection to the Conventional Sampling . System. None of the affected components are initiators of any accidents evaluated in the I

                             , UFSAR. No Technical Specification Changes are required. Changes are required for UFSAR Sections 9.3.2.2.2 and 9.3.2.2.2.4.

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H l U.S. Nuclear Regulatary r===d-ta=

     ' Apre 1,1999 Pase 72 of 247i f

250 Type: MinorModification Unit: 2

                 'I1tle: Minor Modification CE-61314, Addition of Recirculated Cooling Water System cooling line and Conventional Sampling System Heat Exchanger ~
         -E-- C"r: Dis minor modification will route a supply and return line in the Recirculated Cooling Water System to provide cooling for the new Feedwater Corrosion Monitor Sample Cooling Heat Exchanger in the Conventional Sampling System. %e piping will originate i                         at the supply header for the Moisture Separator Reheater, First Stage Reheater, and the Second Stage Reheater Drain Tanks and Lew Pressure Turbines Crossover Rough Cooling Heat Exchangers. It will be routed to the new heat exchanger in the vicinity of the local feedwater sample point at the Electrochemical Potential Monitor. De return piping will extend from the new heat exchanger back to the return of the same header.

- Isolation valves will be provided in the supply and return lines of the new heat exchanger, l and a relief valve will be provided between the two isolEion valves. No connections to l

the Conventional Sampling System side of the Feedwater Corrosion Monitor Sample ^ j Cooling Heat Exchanger will be made per this modification. Evaluation: There are no unreviewed safety questions associated with this modification. The modification involves components in the Recirculated Cooling Water System, with a heat exchanger and connections provided for connection to the Conventional Sampling l System. None of the affected components are initiators of any accidents evaluated in the l UFSAR. No Technical Specification Changes are required. Changes are required for UPSAR Sections 9.3.2.2.2 and 9.3.2.2.2.4. 84 Type: Minor Modification Unit: 1

Title:

Minor Modification CE-61324, Replace flow switch IWLFS7350 for Radiation Monitor IEMF52 with a new switch that is external to the sample line

Description:

Minor Modification CE-61324 replaces flow switch IWLFS7350 for Radiation Monitor  ; IEMFS2 with a new switch that is external to the sample line, Prior to this modification, l the clean area floor drain discharge radition monitor (IEMFS2) had long standing problems with flow switch IWLFS7350 which monitors the Liquid Radwaste system flow to the radiation monitor. De problem was that solid material frequently fouled the flow switch.His resulted in control room alarms and subsequent maintenance. The flow switch provided the input to an annunciator window that alarms on loss of flow. Evaluation: The modification does not involve an unreviewed safety question because the overall operating scheme of the Liquid Radwaste System has not been changed. Only the method by which the amount of sample flow to radiation monitor IEMF52 is monitored is changed. The added component is not nuclear safety related. No technical specification changes are required. UFSAR Figure 11-13 will be revised. 1

N r U.S. Nuclear Regulatory Conunission April 1,1999 Page 73 of 247 85 Type: Minor Modification Unit: 2

Title:

Minor Modification CE-61325, Replace flow switch 2WLFS7350 for Radiation Monitor 2 EMF 52 with a new switch that is external to the sample line

Description:

Minor Modification CE-61325 replaces flow switch 2WLPS7350 for Radiation Monitor 2EMP52 with a new switch that is external to the sample line. Prior to this modification, the clean area floor drain discharge radition monitor (2 EMF 52) had long standing problems with flow switch 2WLFS7350 which monitors the Liquid Radwaste system flow to the radiation monitor. The problem was that solid material frequently fouled the flow switch. This resulted in control room alarms and subsequent maintenance. The flow j switch provided the input to an annunciator window that alarms on loss of flow, i Evaluation: The modification does not involve an unreviewed safety question because the overall operating scheme of the Liquid Radwaste System has not been changed. Only the method by which the amount of sample flow to radiation monitor 2 EMF 52 is monitored is k changed. The added component is not nuclear safety related. No technical specification I changes are required. UFSAR Figure 1120 will be revised. 48 Type: MinorModification Unit: 1 1

               'Dtle: Minor Modification CE-61326, Add an emergency eyewash and shower at the Containment Chilled Water chemical addition area

Description:

Minor Modification CE-61326 adds an emergency eyewash and shower at the Containment Chilled Water chemical addition area. This modification replaces a temporary portable emergency eyewash and shower with a permanent facility. The modification affects piping and equipment located in the Unit 1 Containment Mechanical Equipment Building .The Makeup Demineralized Water System will provide the source of water for the the eyewash station and shower. Evaluation: The systems and components affected by this modification are not nuclear safety related and have no effect on any of the accidents analyzed in the UFSAR. No unreviewed safety questions are created as a result of this modification. No Technical Specification changes are required. A revision is required for UFSAR Figure 9-45 (piping flow drawing). i i I I I l l

U.S. Nuclear Regulatory Conunission Apdf 1,1999 j Page 74 of 247 l 162 Type: Minor Modification Unit: 2

Title:

Minor Modification CE-61327, Add an emergency eyewash and shower at the Containment Chilled Water chemical addition area

Description:

Minor Modification CE-61327 adds an emergency eyewash and shower at the Containment Chilled Water chemical addition area. This modification replaces a temporary portable emergency eyewash and shower with a permanent facility. He modification affects piping and equipment located in the Unit 1 Containment Mechanical Equipment Building . The Makeup Demineralized Water System will provide the source of water for the eyewash station and shower. Evaluation: he systems and components affected by this modification are not nuclear safety related and have no effect on any of the accidents analyzed in the UFSAR. No unreviewed safety questions are created as a result of this modification. No Technical Specification changes are required. A revision is required for UFSAR Figure 9-45 (piping flow drawing). 49 Type: Minor Modification Unit: 1

Title:

Minor Modification CE-61331, Provide two Emergency Eyewash / Shower Stations in the Unit 1 Condensate Polisher Demineralizer Room

Description:

Minor Modification CE-61331 provides two Emergency Eyewash / Shower Stations in the Unit 1 Condensate Polisher Demineralizer Room.nis modification replaces a temporary portable emergency eyewash and shower with two permanent facilities at the Unit 1 Condensate Polisher Demineralizer Room chemical addition area. One facility will be located at the chemical addition area platform and the other on the floor. This modification affects piping and equipment located in the Unit i Turbine Building basement. The Makeup Demineralized Water System will provide the source of water. His modification will satisfy an OSHA requirement for locations utilizing chemicals and chemical products. The Makeup Demineralized Water System is not required for maintenance of plant safety in the event of an accident. Evaluation: The Makeup Demineralized Water System is not required for maintenance of plant safety in the event of an accident. Here are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. A change is required for UFSAR Figure 9-49 (piping flow drawing). l l i l I l i l

U.S. Nuclear Regulatory Commission I April 1,1999 l Page 75 of 247 50 Type: Minor Modification Unit: 2 i

Title:

Minor Modification CE-61332, provides two Emergency Eyewash / Shower Stations in the  ! Unit 2 Condensate Polisher Demineralizer Room. l

Description:

Minor Modification CE-61331 provides two Emergency Eyewash / Shower Stations in the j Unit 2 Condensate Polisher Demineralizer Room. This modification replaces a temporary i portable emergency eyewash and shower with two permanent facilities at the Unit 2 $ Cor.densate Polisher Demineralizer Room chemical addition area. One facility will be located at the chemical addition area platform and the other on the floor. This modification affects piping and equipment located in the Unit 2 Turbine Building 3 basement. He Drinking Water System will provide the source of water. This  ! modification will satisfy an OSliA requirement for locations utilizing chemicals and chemical products. The Makeup Demineralized Water System is not required for maintenance of plant safety in the event of an accident. > Evaluation: The Drinking Water System is not required for maintenance of plant safety in the event of an accident. There are no unreviewed safety questions associated with this modification. No Technical Specifications changes are required. A change is required for UFSAR Table 1-4 (piping and instrumentation diagrams) and UFSAR Figure 9-52. 87 Type: Minor Modification Unit: 1

Title:

Minor Modification CE-61333, Provide two emergency eyewash / shower stations at the Component Cooling System Chemical liandling areas.  ; i

Description:

Minor Modification CE-61333 provides two emergency eyewash stations at the Unit 1 j Component Cooling System Chemical IIandling areas. This modification affects piping { and equipment in the Auxiliary Building. He water will be supplied by the Makeup , Demineralized Water System. ] Evaluation: There is no unreviewed safety question associated with this modification. This l modification will not prevent any existing system or component from completing its design function in support of plant operation. There are no analyzed accidents applicable to the eyewash stations which are added per this modification. No Technical Specification changes are required. UFSAR Figure 9-49 (piping flow drawing) will be revised. , i l

p l l U.S. Nuclear Regulatory Commission Apdf 1,1999 Page 76 of 247 -

  '88     Type: Minor Modification                                              Unit: 2

Title:

Minor Modification CE-61334, Provide two emergency eyewash / shower stations at the Component Cooling System Chemical Handling areas.

Description:

Minor Modification CE-61334 provides two emergency eyewash stations at the Unit 2 Component Cooiing System Chemical Handling areas. This modification affects piping and equipment in the Auxiliary Building. The water will be supplied by the Makeup Demineralized Water System. Evaluation: There is no unreviewed safety question associated with this modification. This modification will not prevent any existing system or component from completing its

                  ' design function in support of plant operation. There are no analyzed accidents applicable to the eyewash stations which are added per this modification. No Technical Specification changes are required. UFSAR Figure 9-45 (piping flow drawing) will be revised.

1

I l i. U.S. Nuclear Regulatory Commission April 1,1999 j Page 77 of 247 309 Type: MinorModification Unit: 1 l

Title:

Minor Modification CE-61347 Replace flow gauge INSPG5120 with a higher range l flow gauge

Description:

Minor Modification CE-61347 replaces flow gauge INSPG5120 with a higher range flow gauge. Flow gauge INSPG5120 indicates flow in the common recirculation line from the Containment Spray pumps discharge back to the Refueling Water Storage Tank. It is currently sized for a 0-700 gallons per minute (gpm) range and is used primarily for in. Service Testing for the Containment Spray pumps. The recirculation line is capable of higher flow rates than 700 gpm, which can over-range the flow gauge. Constant over-ranging can adversely affect the calibration of the gauge. Since it is used during pump testing to set the Containment Spray pump flow rate, any out-of-tolerance found during calibration has the possibility of rendering both trains of Containment Spray inoperable. Additionally, the current limitation of 700 gpm requires the testing to be conducted very near the shut-off flow of the pumps. This low flow limits the pumps to less than two , hours operation per day. A higher range will allow the pump testing to be conducted at a i higher flow rate, thus removing the two hour restriction and reducing wear on the pumps. His modification will replace flow gauge INSPG5120 with one having a larger range, specifically,0-1000 gpm. The gauge will be the same manufacturer and model number as the existing gauge and will be mounted in the same bracket. Neither flow gauge INSPG5120 nor the instrument tubing attached to the flow gauge are nuclear safety-related. Since the flow gauge is located in the Auxiliary Building, it is seismically mounted.he new flow gauge is the same weight and dimensions as the existing gauge and will be mounted in the existing seismic mounting bracket. Consequently, there are no seismic concerns. Since the recirculation line is isolated during normal operation, failure of the instrument will not adversely affect any safety-related system, structure, or component. It should be noted, however, that since the instrument is used for pump testing on the Containment Spray pumps, undetected or unaccounted for inaccuracies in the instrument could affect the validity of the tests and hence the operability of the pumps. Ilowever, since the new gauge is the same manufacturer and model number as the existing gauge, and has the same performance specifications, all instrument inaccuracies associated with the new gauge will be equal to or better than the existing gauge. Evaluation: There are no unreviewed safety questions associated with this modification. The new gauge is considered to be an equivalent replacement for the old gauge. Neither the flow gauge nor the recirculation line are initiators of any accident evaluated in the UFSAR. In addition, the recirculation line is isolated during normal plant operation. No Technical Specification changes are required. No UFSAR changes are required. i

l-U.S. Nuclear Regulatory Co== a s

    ~ Aptil1,1999 Page 78 of 247 300     Type: Minor Modification                                              Unk: 2 '
             , I1tle: Minor Modification CF 61348, Replace flow gauge 2NSPG5120 with a higher range gauge      '

a . . .

Description:

Minor Modification CE-61348 replaces floWgauge 2NSPG5120 with a higher range gauge. Flow gauge 2NSPG5120 indicates flow in the common recirculation line from the Containment Spray pumps discharge back to the Refueling Water Storage Tank. It is currently sized for a 0-700 gadons per minute (gpm) range and is used primarily for In-Service Testing for the Containment Spray pumps. The recirculation line is capable of higher flow rates than 700 gpm, which can over-range the flow gauge. Constant over-ranging can adversely affect the calibration of the gange. Since it is used during pump testing to set the Containment Spray pump flow rate, any out-of-tolerance found during alibration has the possibility of rendering both trains of Containment Spray inoperable. Additionally, the current limitation of 700 gpm requires the testing to be conducted very near the shut-off flow of the pumps. This low flow limits the pumps to less than two hours operation per day. A higher range will allow the pump testing to be conducted at a higher flow rate, thus removing the two hour restriction and reducing wear on the pumps. This modification will replace flow gauge 2NSPG5120 with one having a larger range, specifically,01000 gpm. The gauge will be the same manufacturer and model number as the existing gauge and will be mounted in the same bracket. Neither flow gauge 2NSPG5120 nor the instrument tubing attached to the flow gauge are nuclear safety-related. Since the flow gauge is located in the Auxiliary Building, it is seismically mounted. The new flow gauge is the same weight and dimensions as the existing gauge and will be mounted in the existing seismic mounting bracket. Consequently, there are no seismic concerns. Since the recirculation line is isolated during normal operation, failure of the instrument will not adversely affect any safety-related system, structure, or component. It should be noted, however, that since the instrument is used for pump testing on the Containment Spray pumps, undetected or unaccounted for inaccuracies in the instrument could affect the validity of the tests and hence the operability of the

                    . pumps. However, since the new gauge is the same manufacturer and model number as the existing gauge, and has the same performance specifications, all instrument inaccuracies associated with the new gauge will be equal to or better than the existing gauge.

Evaluation: There are no unreviewed safety questions associated with this modification. The new gauge is considered to be an equivalent replacement for the old gauge. Neither the flow gauge nor the recirculation line are initiators of any accident evaluated in the UFSAR. In addition, the recirculation line is isolated during normal plant operation. No Technical Specification changes are required. No UFSAR changes are required. L

(. a U.S. Nuclear Regulatory Cos==I== tan April 1,1999 Page 79 of 247 f

                  ~ 251       Type: Minor Modification -                                                   Unit: 2
                              'I1 tie: Minor Modification CE-61359

Description:

Minor Modification CE-61359 deletes unused oil fill and drain piping on Unit 2 Reactor

                                          ' Coolant Pump Motor 2C. ne ' piping between valve 2NC203 (upper bearing fill and drain line) and the flexible metal hose connection will be deleted. He piping between vendor valve 2NC205 (lower bearing fill and drain line) and the flexible metal hose connection will be deleted. He flexible hose for these two applications will also be deleted and the remaining downstream piping will be blind flanged. A solid inspection cover will be installed over the access hole that results from removal of the oil piping from the upper bearing oil cooler enclosure. All hangers or appropriate sections of gang hangers that are used to support the piping will also be removed.

Evaluation: The oil fill and drain piping to be removed are not t : quired to mitigate the consequences of an accident during unit operation, and no new acc.Jents are postulated. The oil fill and drain piping is not required for lubrication of the reactor coolant pump motor during operation. Therefore, this modification does not degrade reactor coolant pump reliability or operation.' Although the reactor coolant pump motor oil piping and components are designated as Reactor Coolant System components, the oil piping is not connected to the p6aary reactor coolant fluid pressure boundary. The design function of the reactor coolant pump is to provide an adequate core cooling flow rate for sufficient heat transfer, to maintain a Departure from Nucleate Boiling Ratio (DNBR) greater than 13 within the parameters of operation. Sufficient pump rotation inertia is provided by a flywheel, in conjunction with the impeller and motor assembly, to provide adequate flow during coast-down. This forced flow following an assumed loss of pump power, and the subsequent natural circulation effect, provides the core with adequate cooling. The removal of the oil fill and drain piping will have no effect on the quantity of oil contained in the Reactor Coolant Pump Motor-2C, and therefore pump

                                           " coast-down" is not comprmdsed.

Implementation of this modification will not prevent the Reactor Coolant System fron$ mitigating the consequences of a Design Basis Event. The lubrication of the reactor coolant pump motor during operation is not degraded, thus operation of the motor is not degraded. Herefore, the probability of an accident resulting from a pump failure, such as a decrease in reactor coolant system flow rate (UPSAR 15.3), is not increased. The effect on the Reactor Coolant System or any system structure or component in the vicinity of the Reactor Coolant Pump Motor is not significant, and thus no safety margins are reduced. No unreviewed safety questions are created by or involved with this modification. No changes to the technical specifications or UPSAR are required. i -up .T , . m_____

r- i

        ?
              . U.S. Nuclear Regulatory Co-=8-Ion Apsti 1,1999 Page 80 of 247 207 - Type: Minor Modification .                                               Unit: 2

Title:

Minor Modification CE-61360

Description:

Minor Modification CE-61360 deletes unused oil fill and drain piping on Unit 2 Reactor Coolant Pump Motor 2D. The piping, hangers and appropriate sections of gang hangers between valve 2NC206 (upper bearing fill and drain line) and the flexible metal hose connection will be deleted. The flex hose downstream of valve 2NC206 and the downstream piping and its (6) supports will be deleted. A solid inspection cover will be installed over the access hole that results from removal of the oil piping. Removal of the piping will require it to be cut and fitted with a screwed pipe cap. He piping between vendor valve 2NC208 (lower bearing fill and drain line) and the flexible metal hose connection will be deleted. The flex hose downstream of valve 2NC208 will also be deleted and the remaining (downstream) piping will be cut and fitted

                                  'with a screwed (pipe) cap. Also 4 supports will be deleted and one support designated as excess structural steel.

The removal of fill and drain piping and associated support (steel) is required to develop a configuration that facilitates the removal and replacement of Reactor Coolant Pump - Motor 2D. The piping which will be removed is not currently used for filling the oil

reservoirs.- Draining the reservoirs will be performed, during outages, by alternate means via temporary tubing and remaining " permanent" piping with quick disconnects available near the lower motor bearings. De temporary tubing used to drain the oil reservoirs is removed during the' outage, prior to unit operation, ne removal of piping and hanger (material) will significantly improve the ccafiguration of the equipment since it will ~ .

reduce clutter and provide a simpler access for maintenance of Reactor Coolant Pump

                               - Motor 2D.

The piping stress analysis calculations have been revised documenting the resuking configuration as acceptable for seismic service. The weight of steel removed from the motors is negligible compared to the motor weight. He effect of the removal of this mass does not invalidate basis dynamic analysis or any existing safety analysis. Due to the negligible amount of steel removed as compared to the motor weight, the change in motor vibration was determined to be negligible. Evaluation: The oil fill and drain piping to be removed are not required to mitigate the consequences of an accident during unit operation, and no new accidents are postulated. The oil fill and

                                . drain piping is not required for lubrication of the reactor coolant pump motor during operation. 't iierefore, this modification does not degrade reactor coolant pump reliability or operation. Although the reactor coolant pump motor oil piping and components are designated as Reactor Coolant System, the oil piping is not connected to the primary reactor coolant fluid pressure boundary.

De design function of the reactor coolant pump is to provide an adequate core cooling flow rate for sufficient heat transfer, to maintain a Departure from Nucleate Boiling Ratio (DNBR) greater than 1.3 within the parameters of operation. Sufficient pump rotation

                               . inertia is provided by a flywheel, in conjunction with the impeller and motor assembly, to provide adequate flow during coast-down. This forced flow following an assumed loss of

U.S. Nuclear Regulatory Conunission April 1,1999 ' Page 81 of 247 - J pump power, and the subsequent natural circulation effect, provides tim core with adequate cooling. The removal of the oil fill and drain piping will have no effect on the quantity of oil contained in the Reactor Coolant Pump Motor-2D, and therefore pump

                  -* coast-down" is not compromised.

l Implementation of this modification will not prevent the Reactor Coolant System from mitigating the consequences of a Design Basis Event. The lubrication of the reactor coolant pump motor during operation is not 6egraded, thus operation of the motor is not

                 - degraded. 'Iherefore, the probability of an accident resulting from a pump failure, such as a decrease in reactor coolant system flow rate (UFSAR 15.3), is not increased. The effect on the Reactor Coolant System or any system structure or component in the vicinity of the Reactor Coolant Pump Motor is not significant, and thus no safety margins are reduced.

No unreviewed safety questions r.re created by or involved with this modification. No changes to the technical specifications or UFSAR are required. l 1 l i i I i 4 L.

I U.S.Nsclear Regulatory Commission Apdf 1,1999 Page 82 of 247

 '173,   Type: MinorModification                                                  Unk: 2

Title:

MinorModificationCE-61362

Description:

Minor Modification CE-61362 removes unused oil fill and drain piping on Unit 2 Reactor Coolant Pump Motor 2A. De piping, hangers and appropriate sections of gang hangers between valve 2NC197 (upper bearing fill and drain line) and the flexible metal hose connection will be deleted. He flex hose downstream of valve 2NC197 and the downstream piping and its (4) supports will be deleted. A solid inspection cover will be installed over the access hole that results from removal of the oil piping. Removal of the piping will require it to be cut and fitted with a socket weld pipe cap. The piping between vendor valve 2NC199 (lower bearing fill and drain line) and the flexible metal hose connection will be deleted. De flex hose downstream of valve 2NC199 will also be deleted and the remaining (downstream) piping will be blind flanged. Dreaded pipe plugs will be used to isolate the Duke and vendor supplied valves. All hangers, or the appropriate sections of gang hangers, that were used to . support the piping removed, will also be removed. The removal of fill and drain piping and associated support (steel) is required to develop a configuration that facilitates the removal and replacement of Reactor Coolant Pump Motor 2A. The piping which will be removed is not currently used for filling the oil reservcirs. Draining the reservoirs will be performed, during outages, by alternate means via temporary tubing and remaining " permanent" piping with quick disconnects available near the lower motor bearings. The temporary tubing used to drain the oil reservoirs is removed during the outage, prior to unit operation. The removal of piping and hanger ) (material) will significantly improve the configuration of the equipment since it will reduce clutter and provide a simpler access for maintenance of Reactor Coolant Pump Motor 2A, The piping stress analysis calculations have been revised documenting the resulting configuration as acceptable for seismic service. The weight of steel removed from the motors is negligible compared to the motor weight. He effect of the removal of this mass does not invalidate basis dynamic analysis or any existing safety analysis. Due to the negligible amount of steel removed as compared to the motor weight, the change in motor vibration was determined to be negligible. Evaluation: He oil fill and drain piping to be removed are not required to mitigate the consequences of an accident during unit operation, and no new accidents are postulated. The oil fill and 3 drain piping is not required for lubrication of the reactor coolant pump motor during l cperation. Therefore, this modification does not degrade reactor coolant pump reliability or operation. Although the reactor coolant pump motor oil piping and components are

designated as Reactor Coolant System, the oil piping is not connected to the primary reactor coolant fluid pressure boundary.

The design function of the reactor coolant pump is to provide an adequate core cooling flow rate for sufficient heat transfer, to maintain a Departure from Nucleate Boiling Ratio (DNER) greater than 1.3 within the parameters of operation. Sufficient pump rotation inertia is provided by a flywheel, in conjunction with the impeller and motor assembly, to l I l l

                                                                                                               )

V l 1-t

                                                                                                                 \

U.S. Nuclear Regulatory Commission April 1,1999 Page 83 of 247 ' l f provide adequate flow during coast-down. 'Ihis forced flow following an assumed loss of pump power, and the subsequent natural circulation effect, provides the core with i adequate cooling. The removal of the oil fill and drain piping will have no effect on the l quantity of oil contained in the Reactor Coolant Pump Motor-2A , and therefore pump i I

                    " coast-down" is not compromised.

Implementation of this modification will not prevent the Reactor Coolant System from mitigating the consequences of a Design Basis Event. The lubrication of the reactor l coolant pump motor during operation is not degraded, thus operation of the motor is not I degraded. Therefore, the probability of an accident resulting from a pump failure, such as a decrease in reactor coolant system flow rate (UFSAR 15.3), is not increased. The effect on the Reactor Coolant System or any system structure or component in the vicinity of the Reactor Coolant Pump Motor is not significant, and thus no safety margins are reduced. i No unreviewed safety questions are created by or involved with this modification. No i changes to the technical specifications or UFSAR are required. l l i 1 l l l l l l l l' l 1 l l i l i

       . .U.S. Nuclear Regulatory Co==I==Ian April 1,1999 Page 84 of 247 T

i' 174. Type: MinorModification Unit: 2 I 'I1 tie: Minor Modification CE-61363 -

Description:

Minor Modification CE-61363 removes unused oil fill and drain piping on Unit 2 Reactor Coolant Pump Motor 2B. The piping between valve 2NC200 (upper bearing fill and

                         ' drain line) and the flexible metal hose connection wCl be deleted. He piping between vendor valve 2NC202 (lower bearing fill and drain line) and the flexible metal hose '
                         . connection will be deleted. 'Ihe flex hose for these two applications will be deleted and 3

the remaining downstream piping will be blind flanged. Threaded pipe plugs will be I l' used to isolate the Duke and vendor supplied valves. A solid inspection cover will be installed over the access hole that results from removal of oil piping from the upper bearing oil cooler enclosure. All hangers, or the appropriate sections of gang hangers, l that were used to support the piping removed, will also be removed. I The removal of fill and drain piping and associated support (steel) is required to develop a configuration that facilitates the removal and replacement of Reactor Coolant Pump Motor 2B. The piping which will be removed is not currently used for filling the oil reservoirs. Draining the reservoirs will be performed, during outages, by alternate means via temporary tubing and remaining

• permanent" piping with quick disconnects available near the lower motor bearings. De temporary tubing used to drain the oil reservoirs is removed during the outage, prior to unit operation. The removal of piping and hanger (material) will significantly improve the configuration of the equipment since it will reduce clutter and provide a simpler access for maintenance of Reactor Coolant Pump Motor 2B.

The piping stress analysis calculations have been revised documenting the resulting configuration as acceptable for seismic service. The weight of steel removed from the motors is negligible compared to the motor weight. The effect of the removal of this mass does not invalidate basis dynamic analysis or any existing safety analysis. Due to the negligible amount of steel removed as compared to the motor weight, the change in motor vibration was determined to be negligible. Evaluation: De oil fill and drain piping to be removed are not required to mitigate the consequences of an accident during unit operation, and no.new accidents are postulated. The oil fill and drain piping is not required for lubrication of the reactor coolant pump motor during operation. Therefore, this modification does not degrade reactor coolant pump reliability or operation. Although the reactor coolant pump motor oil piping and co.cponents are designated as Reactor Coolant System components, the oil piping is not connected to the primary reactor coolant fluid pressure boundary. The design function of the reactor coolant pump is to provide an adequate core cooling flow rate for sufficient heat transfer, to maintain a Departure from Nucleate Boiling Ratio (DNBR) greater than 1.3 within the parameters of operation. Sufficient pump rotation inertia is provided by a flywheel, in conjunction with the impeller and motor assembly, to provide adequate flow during coast-down. This forced flow following an assumed loss of pump power, and the subsequent natural circulation effect, provides the core with adequate cooling. De removal of the oil fill and drain piping will have no effect on the quantity of oil contained in the Reactor Coolant Pump Motor-2B , and therefore pump

e U.S. Nuclear Regulatory Commission April 1,1999 Page 85 of 247 t

                  " coast-down" is not compromised.

Implementation of this modification will not prevent the Reactor Coolant System from mitigating the consequences of a Design Basis Event. The lubrication of the reactor coolant pump motor during operation is not degraded, thus operation of the motor is not degraded. Herefore, the probability of an accident resulting from a pump failure, such as a decrease in reactor coolant system flow rate (UFSAR 15.3), is not increased. The effect on the Reactor Coolant System or any system structure or component in the vicinity of the Peactor Coolant Pump Motor is not significant, and thus no safety margins are reduced.

                 . Based on this USQ Evaluation, no unreviewed safety questions are created by or involved with this modification. No changes to the Technical Specifications or UFSAR are required.

51 Type: Minor Modification Unit: 0

Title:

Minor Modification CE-61367, Provide Vent Line for the Containmcot Spray Pmnp Suction Piping.

Description:

Minor Modification CE-61367, Provides a vent line for the Containment Spray Pump Suction piping. This modification provides vent lines on the Containment Spray Pump flush lines from the Reactor Makeup Water Storage Tank. The vent lines will be installed on the flanged connections in the flush lines. These vent lines are needed because there is a lack of appropriate vents for venting the Residual Heat Removal and Containment Spray System Pumps suction piping. Evaluation: he safety related function of the Residual Heat Removal and Containment Spray Systems are not affected by the addition of these vents. The modification will provide better system venting for the Residual Heat Removal and Containment Spray Systems but does not change the function or the design basis of the systems involved. There are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. A change is required for UFSAR Figure 6-109 (piping flow drawing) I l l t

I .'

U.S. Nuclear Regulatory Commission April 1,1999 Page 86 of 247 65 Type

MinorModification Unit: 1

Title:

Minor Modification CE-61383, Allow flexible tubing to remain installed on li,esidual Heat Removal Heat Exchanger Vent Lines

Description:

Minor Modification CE-61383, Allow flexible tubing to remain installed on Residual Heat Removal Heat Exchanger Vent Lines, adds notes to flow diagrams to allow flexible tubing to remain attached to selected vent lines. The lines involved are those that contain valves IND-070, IND-076, and IND-077 associated with Residual Heat Removal Heat Exchanger I A: and valves IND-072, IND-079, and IND-080 associated with Residual Heat Removal Heat Exchanger IB. Evaluation: 'Ihis modification does not change the design or function of the Residual Heat Removal System. The additional weight of the tubing or hose is negligible to the applicable piping qualification calculation. There are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. No UFSAR changes are required. i l _1

g

       . U.S. Nuclear Regulatory Conumission

(- Apdf 1,1999 l Page 87 of 247 283~ Type: Minor Modification Unit: 1

                  'Iltle: Minor Modification CE-61396, Modify the position of valve ICA6 and revise the Unit 1 Upper Surge Tank lo-Level setpoint

Description:

Minor Modification CE-613% changes the position of valve ICA6 so that the valve is normally closed with the associated motor operator breaker in the "off" position, in addition, the Condensate Storage System Upper Surge Tank (UST) Io-Level setpoint will be revised to provide sufficient time for Operations to respond to the alarm and take y appropriate action to avoid an automatic Nuclear Service Water System / Auxiliary Feedwater System autoswap or Auxiliary Feedwater System Pump trip. The above changes to the Auxiliary Feedwater System and Condensate Storage Systems are needed due to the problems with the Auxiliary Feedwater System Condensate Storage Tank (CACST) supply. Currently, valve ICA6 is closed with power removed to isolate the CACST from the Auxiliary Feedwater pump supply piping. This valve is closed to l prevent the introduction of air into the Auxiliary Feedwater System pump suction piping. l Due to the piping configuration of the Auxiliary Feedwater and Condensate Storage Systems, during certain plant conditions and upon depletion of the CACST air could be drawn into the Auxiliary Feedwater suction piping which could result in Auxiliary Feedwater pump damage On 5/8/97 concerns were identified with vortex formation in the CACST and USTs that could lead to the introduction of air into the suction piping of the Auxiliary Feedwater pumps, potentially disabling the pumps. An Operability Evaluation was performed, with support from the Auxiliary Feedwater pump vendor, which concluded that vortex formation is not an operability concern. . During the process of evaluating the vortex concern, a separate mechanism was identified by which air could potentially enter the Auxiliary Feedwater suction piping. This ] mechanism involved the depletion of the CACST and the failure of valve ICA6 (CACST to Auxiliary Feedwater Pump Isolation Valve) to automatically close on a low CACST level. In this situation, the USTs would supply the Auxiliary Feedwater pumps. However, if condenser vacuum is not broken, the relative elevation head of the UST's with respect to thejunction of the Auxiliary Feedwater supply piping from the CACST and USTs is not sufficient to maintain the pressure at this junction above atmospheric pressure over the full range of possible Auxiliary Feedwater flow rates. Also, due to the

                          - elevation difference between thisjunction and the pressure switches that activate the automatic swapover to the assured Auxiliary Feedwater suction sources of the Nuclear
                          ' Service Water System and the setpoint of these pressure switches, the swapover is not assured if the CACST/USTs junction pressure is less then atmospheric over the full range of possible Auxiliary Feedwater flows. This could lead to the introduction of air into the Auxiliary Feedwater suction piping and possibly into the Auxiliary Feedwater pumps from the depleted and unisolated CACST, potentially disabling the pumps.

The analysis of the quantity of air that could be introduced into the Auxiliary Feedwater piping and the resulting effects on Auxiliary Feedwater pump operation required tie use of resources outside of Duke Power Company. Framatome Technologies, Inc. (Fil) was i l 1-1

C 1 .. .. . l . U.S. Nuclear R - datory Commission i l , Apdl 1,1999 l- Page 88 of 247. 1 l contracted to verify the Catawba analysis of possible air introduction upon CACST depletion as described above, and also to analyze the entire non-safety Auxiliary

                     . Feedwater suction source design to determine if any other potential problems exist. De     i evaluation performed by FTI confirmed the problem associated with air introduction into
                    -1 the Auxiliary Feedwater System.-

The isolation of the CACST as a Auxiliary Feedwater System suction source eliminates the possibility of air introduction into the Auxiliary Feedwater suction piping upon CACST depletion previously described. %e CACST is a non-safety related source of water for the Auxiliary Feedwater System. De volume of water in the CACST is not credited in the evaluation of any Design Basis Event. He Nuclear Service Water System is the safety related source of water for the Auxiliary Feedwater System, and though normally isolated from the Auxiliary Feedwater System aligns automatically upon detection of low Auxiliaq Feedwater suction pressure. For Design Events in which the Nuclear Service Water System is not available to supply water to the Auxiliary Feedwater System (Loss of all AC Power Security, Fire), the Recirculated Cooling Water System contains a sufficient volume of water in the embedded Recirculated Cooling Water System header in the Turbine Building to maintain the unit at Hot Standby for the required maximum 72 hour duration. The Recirculated Cooling Water System supply is normally isolated, but aligns automatically on detection oflow Auxiliary Feedwater suction pressure.

                                                                                                    ~

It should be noted that as the CACST is shared between the two units, only one half ofits 42,500 gallon capacity (or 21,250 gallons) can be assumed to be available to each unit during a dual unit event (eg: Loss of Offsite Power). The unit specific USTs have a combined 85,000 gallon capacity and the unit specific Condenser Hotwell contains 170,000 gallons at normal operating level. Therefore, the great majority of the non- , safety, condensate grade water supply to the Auxiliary Feedwater System will remain 1 aligned to the system. Also, the current Technical Specification associated with the required 225,000 gallon i volume of condensate grade water in the CACST, USTs, and Condenser Hotwell (T S. I 3.7.1.5, Condensate Storage System) exists only on Unit 2. No such Technical l Specification exists on Unit 1. ne Unit 2 Technical Specification does not specify any required volume in the specific tanks, only a combined volume. It is possible to meet the  ; Technical Specification requirement with the USTs and Condenser Hotwell only (i.e. 1 CACST empty). Also, Improved Technical Specifications (ITS) are scheduled to be implemented in early 1999 so they will be addressed in this evaluation. The applicable Technical Specification for both units will be 3.7.6 (Condensate Storage System). The Limiting Condition for Operation (LCO) for this specification states that the Condensate Storage System (CSS) ' Inventory shall be greater than or equal to 225,000 gallons. This specification does not list any specific tanks so the same logic may be applied, that is the USTs and the Hotwell's combined inventory exceed the ITS requirement. However, the specific tanks that form the CSS inventory are listed in the bases for ITS 3.7.6 (CSS). The tanks listed are the USTs, the CACST and the Condenser Hotwell. The tanks are listed only once in the bases Background section and in the remaining sections of the bases (Applicable J

m U.S. Nuclear Regulatory Conumission - April 1,1999 ~ Page 89 of 247 - Safety Analyses, LCO, Applicablility, Actions, Surveillance Re uirements and References) are referred to as the CSS. The process for revising the ITS allows the bases for a specification to be revised by the 10CFR50.59 process. De 10CFR50.59 for this modification will document the evaluation for revising the Background Section of the bases for ITS 3.7.6. nis revision will delete the CACST from the list of sources that form the CSS. The USTs and the Condenser Hotwell will be the normal sources that form the CSS. Evaluation: This modification does not add or delete any automatic or manual safety related feature of the Auxiliary Feedwater System, nor does it convert an automatic safety related feature to manual or vice versa. %e modification does not introduce an unwanted or previously unreviewed system interaction, but instead eliminates such an interaction. His modification does not alter the QA condition, seismic or environmental qualification of any component in the Auxiliary Feedwater System as the CACST is a non-safety, non-seismic tank. No adverse effects on the safety related function of the Auxiliary Feedwater System or any interfacing systems are created by this activity. The Condensate Storage System will continue to provide the Technical Specification required condensate

                      ~ inventory of greater than or equal to 225,000 gallons. Revising the UST Lo-Ixvel Setpoint will not affect any safety-related system or function. Increasing the setpoint will allow the Operations Group more time to react to a reduced inventory in the CSS. T'ae Nuclear Service Water System will continue to be the safety related supply for the Auxiliary Feedwater System a:M this fsoction is rni affected. No new failure modes are created by this nr.sdification. No unreviewed safety questions se created as a result of this modification which isolates the CACST from the Auxiliary Feedwater System supply piping and revises the UST Lo-Level Setpoint. Changes to the current Technical Specifications are not required. Changes to the ITS Specification 3.7.6 are not required; however, the bases for this specification will be revised to reflect that the CACST will no longer be a normal CSS source. This modification does result in the plant configuration being different from that described in the UFSAR (specifically Section 10.4.9.2 and
                      . Appendix 10 Chapter 10 Tables and Figures). He affected sections and figures will be revised to show the new configuration of the Auxiliary Feedwater and Condensate Storage Systems.

i L i 1 l l u t - I l

o U.S. Nuclear Regulatory Comunission April 1,1999 Pase 90 of 247

   ~284       Type: MinorModification                                             Unit: 2 T1tle: Minor Modification CE-61397, Modify the position of valve 2CA6 and revise the Unit 2 Upper Surge Tank Lo-Level setpoint

Description:

Minor Modification CE-61397 changes the position of valve 2CA6 so that the valve is normally closed with the associated motor operator breaker in the "off" position. In addition, the Condensate Storage System Upper Surge Tank (UST) 1.4-level setpoint will be revised to provide sufficient time for Operations to respond to the alarm and take appropriate action to avoid an automatic Nuclear Service Water System / Auxiliary Feedwater System autoswap or Auxiliary Feedwater System Pump trip. The above changes to the Auxiliary Feedwater System and Condensate Storage Systems are needed due to the problems with the Auxiliary Feedwater System Condensate Storage Tank (CACST) supply. Currently, valve 2CA6 is closed with power removed to isolate the CACST from the Auxiliary Feedwater pump supply piping. This valve is closed to prevent the introduction of air into the Auxiliary Feedwater System pump suction piping. Due to the piping configuration of the Auxiliary Feedwater and Condensate Storage Systems, during certain plant conditions and upon depletion of the CACST air could be drawn into the Auxiliary Feedwater suction piping which could result in Auxiliary Feedwater pump damage On 5/8/97 concerns were identified with vortex formation in the CACST and USTs that could lead to the intr (du':tica &J:, into # muon ikin @le kniliary Feedwater

                                                     ~

pumps, potentially disabling t.x pJaps. An Operability isvaloatma ejerformed, with stspport from the Auxiliary Feedwater pump vendor, which concluded eat vortex formation is not an operability ecmsn. During the process of evaluating the vortex concern, a separate mech.adsm was identified by which air could potentially enter the Auxiliary Feedwater suction piping. This mechanism involved the depletion of the CACST and the failure of valve 2CA6 (CACST to Auxiliary Feedwater Pump Isolation Valve) to automatically close on a low CACST level. In this situation, the USTs would supply the Auxiliary Feedwater pumps. However, if condenser vacuum is not broken, the relative elevation head of the UST's with respect to the jtmetion of the Auxiliary Feedwater supply piping from the CACST and USTs is not sufficient to maintain the pressure at thisjunction above atmospheric pressure over the full range of possible Auxiliary Feedwater flow rates. Also, due to the elevation difference between this junction and the pressure switches that activate the automatic swapover to the assured Auxiliary Feedwater suction sources of the Nuclear Service Water System and the setpoint of these pressure switches, the swapover is not assured if the CACST/USTs junction pressure is less then atmospheric over the full range of possible Auxiliary Feedwater flows. This could lead to the introduction of air into the Auxiliary Feedwater suction piping and possibly into the Auxiliary Feedwater pumps from the depleted and unisolated CACST, potentially disabling the pumps. The analysis of the quantity of air that could be introduced into the Auxiliary Feedwater piping and the resulting effects on Auxiliary Feedwater pump operation required the use of resources outside of Duke Power Company. Framatome Technologies, Inc. (14TI) was

k, U.S. Nuclear Regulatory Con ==la= Ion

        ' Apdl1,1999 Pase 91 of 247 contracted to verify the Catawba analysis of possible air intro ction upon CACST depletion as described above, and also to analyze the entire non-safety Auxiliary Feedwater suction source design to determine if any other potential problems exist. He evaluation performed by FTl confirmed the problem associated with air introduction into the Auxiliary Feedwater System, l

The isolation of the CACST as a Auxiliary Feedwater System suction source eliminates the possibility of air introduction into the Auxiliary Feedwater suction piping upon i CACST depletion previously described. The CACST is a non-safety related source of water for the Auxiliary Feedwater System. De volume of water in the CACST is not credited in the evaluation of any Design Basis Event. De Nuclear Service Water System is the safety related source of water for the Auxiliary Feedwater System, and though normally isolated from the Auxiliary Feedwater System aligns automatically upon detection of low Auxiliary Feedwater suction pressure. For Design Events in which the 1

                        - Nuclear Service Water System is not availab?e to supply water to the Auxiliary Feedwater System (Loss of all AC Power, Security, Fire), the Recirculated Cooling Water System contains a sufficient volume of water in the embedded Recirculated Cooling Water System header in the Turbine Building to maintain the unit at Hot Standby for the required maximum 72 hour duration. The Recirculated Cooling Water System supply is normally isolated, but aligns automatically on detection of low Auriliary Feedwater suction pressure.                                                                           ;

it should be noted that as the CACST is shared between the two units, only one half of its 42,500 gallon capacity (or 21,250 gallons) can be assumed to be available to each unit ]; hiing a dual unit event (eg: less of Offsite Power). The unit specific USTs have a j combined 85,000 gallon capacity and the unit specific Condenser Hotwell contains  ! 170,000 gallons at normal operating level. Ecrefore, the great majority of the non- l safety, condensate grade water supply to the Auxiliary Feedwater System will remain aligned to the system. Also, the current Technical Specification associated with the required 225,000 gallon volume of condensate grade water in the CACST, USTs, and Condenser Hotwell (T.S. L 3.7.1.5, Condensate Storage System) exists only on Unit 2. No such Technical Specification exists on Unit 1. De Unit 2 Technical Specification does not specify any ) required volume in the specific tanks, only a combined volume. It is possible to meet the l Technical Specification requirement with the USTs and Condenser Hotwell only (i.e. ) CACST empty). Also, Improved Technical Specifications (ITS) are scheduled to be implemented in early 1999 so they will be addressed in this evaluation. The applicable Technical Specification for both units will be 3.7.6 (Condensate Storage System). De Limiting Condition for Operation (LCO) for this specification states that the Condensate Storage System (CSS) Inventory shall be greater than or equal to 225,000 gallons. This specification does not

                        . list any specific tanks so the same logic may be applied, that is the USTs and the Hotwell's combined inventory exceed the ITS requirement. However, the specific tanks that form the CSS inventory are listed in the bases for ITS 3.7.6 (CSS). The tanks listed are the USTs, the CACST and the Condenser Hotwell. The tanks are listed only once in the bases Background section and in the remaining sections of the bases (Applicable b.

1

              ,              4 U.S. Nuclear Regulatory Comunission
         . Apdf 1,1999 Pane 92 of 247 Safety Analyses, LCO, Applicablility, Actions, Surveillance Requirements and References) are referred to as the CSS. The process for revising the ITS allows the bases
                          ' for a specification to be revised by the 10CFR50.59 process. The 10CFR50.59 for this modification will document the evaluation for revising the Background Section of the bases for ITS 3.7.6. This revision will delete the CACST from the list of sources that form the CSS. The USTs and the Condenser Hotwell will be the normal sources that form the CSS.
            . Evaluation: This modification does not add or delete any automatic or manual safety related feature of the Auxiliary Feedwater System, nor does it convert an automatic safety related feature to   _,
                                                                                                                           ~

manual or vice versa. The modification does not introduce an unwanted or previously unreviewed system interaction, but instead eliminates such an interaction. This modification does not alter the QA condidon, seismic or environmental qualification of any component in the Auxiliary Feedwater System as the CACST is a non-safety, non-seismic tank. No adverse effects on the safety related function of the Auxiliary Feedwater

     -(                     System or any interfacing systems are created by this activity. The Condensate Storage System will continue to provide the Technical Specification required condensate inventory of greater than or equal to 225,000 gallons. Revising the UST Lo-Level Setpoint will not affect any safety-related system or function. Increasing the setpoint will allow the Operations Group more time to react to a reduced inventory in the CSS. The Nuclear Service Water System will continue to be the safety related supply for the Auxiliary Feedwater System and this function is not affected. No new failure modes are created by this modification. No unreviewed safety questions are created as a result of this modification which isolates the CACST from the Auxiliary Feedwater System supply piping and revises the UST Lo-l.evel Setpoint. Changes to the current Technical Specihcations are not required. Changes to the ITS Specificition 3,7.6 are not required; j

, however, the bases for this specification will be revised to reacct that tht CACST will no , longer be a normal CSS source. This modification does result in the plant confiEwatson i being different from that described in the UFSAR (specifically Section 10A.9.2 and I Appendix 10 Chapter 10 Tables and Figures). The affected sections and figures will be i revised to show the new configuration of the Auxiliary Feedwater and Condensate I Storage Systems. l l l i L l (

i U.S. Nuclear Regulatory Com.nission April I,1999 Par,e 93 of 247 l 304- Type: Minor Modification Unit: 1

             *Iltle: Minor Modification CE-61405, Provide Emergency Eyewash / Shower Station at the Plant Heating System Chemical Addition Area

Description:

Minor Modification CE 61405 adds an Emergency Eyewash / Shower Station at the Plant ! Heating System Chemical Addition Area. This replaces a temporary portable eyewash l station with a permanent eyewash and shower facility. This modification affects piping in l the Unit 1 Turbine Building. The Makeup Demineralized Water System will provide the souce of water for this modification. This modification will satisfy an OSHA requirement' for locations utilizing chemicals and chemical products. l Evaluation: The Makeup Demineralized Water System is not required for maintenance of plant safety

in the event of an accident. There are no unreviewed safety questions associated with this l modification. No Technical Specification changes are required. A change is required for l UFSAR Figure 9 49 (piping flow drawing).

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                                                                                                                 )

I i

Il U.S. Nuclear Regulaton Comunission ( Apdf1,1999 Pame 94 of 247 ' 218 Type: MinorModification Unit: 2' Thie: Minor Modification CB-61425 including variation notice 61425A: Modify D/G 2B lube oil supply tubing to the turbocharger and install locking devices on valves 2LD147 and 2LD 148

Description:

Minor ModMication CE-61425 and varaition notice 61425A will modify the Diesel Generator 2B lube oil supply tubing to the turbocharger and install locking devices on valves 2LD147 and 2LD148. The three quarter inch lube oil supply tubing to the left and right banks of the Diesel Generator 2B Turbocharger will be replaced with one inch tubing. Valves 2LD147 (Left Bank) and 2LD148 (Right Bank) will be added as throttle valves so the supply pressure can be adjusted as required to maintain the appropriate I setting. - A locking device will be added to valves 2LD147 and 2LD148 to prevent tampering and mispositioning of the set throttled position. Diesel Generator 2B is out of service for refueling outage inspections. During the maintenance break in run, the Diesel Engine Turbocharger Low Lube Oil Pressure alarm came in at the 20 psig decreasing setpoint. The lube oil supply pressure is not adequate for proper Diesel operation. Evaluation: The new valves (2LD 147 and 2LD 148) are one inch full port ball valves. These valves

                  . are stainless steel nuclear safety related valves with design conditions of 750 psig at 100 degrees F and will meet the design temperature and pressure requirements of the interfacing system (100 psig and 200 degrees F). These valves are suitable for a low pressure throttling application. The new one inch lube oil supply tubing will also meet the design requirements. The new ball valves and tubing are suitable for use in the Diesel Lube Oil System. The addition of the weight of the new components has been evaluated for impact to the stress analysis and new support / restraints have been designed. Adding the new ball valves and one inch tubing will not affect design function of the Diesel Lube   .,

Oil System during any phase of operation. Increasing the size of the turbocharger lube oit j supply tubing from three quarter inch to one inch combined with the addition of the ball j valves for throttling will continue to allow the oil pressure to be set properly. Throttling j of the ball valves to maintain the correct oil pressure will ensure the tube oil supply pressure to the turbocharger is adequate. The handle assembly for the 2LD147 and . I 2LD148 application ofitem Number DMV-904 will be modified to allow installation of a locking device. De removal of the mr.jority of the valve handle and addition of a locking plate and hardware will resuk in a small weight and center of gravity change for the j valve. These valve configuration changes are considered minor, and their effect on the structural integrity and vibration frequency of the valve is considered negligible Dus the seismic qualification for this valve will not be affected. The locking device will be used to lock the valve at a specific throttle position and will not affect th pressure boundary integrity of the valve or the ability of the valve to function properly. Nuclear safety related materials will be used to fabricate the locking device. The valves will be throttled and set at the proper position by an approved maintenance procedure. During Diesel , Generator operation, turbocharger lube oil supply pressure is monitored. Also, j turbocharger lube oil supply pressure is trended by Engineering for any fluctuations in , pressure. De affected portions of the Diesel Lube Oil Systems will continue to function as described in the UFSAR and the Diesel Lube Oil System design basis specification. There are no unreviewed safety questions associated with this modification. No Technical

U.S. Nuclear Regulatory Comunission Apdl I,1999

Pane 95 of 247 -

Specification changes are required. No UFSAR changes are required 227 - Type: Minor Modification Unit: 2

Title:

Minor Modification CE-61428, Install tubing support for Diesel Generator 2B

                    ' Turbocharger lube oil tubing

Description:

Minor Modification CE-61428 installs a tubing support on the lube oil supply and lastrumentatiion tubing for the left bank Diesel Generator 2B Turbocharger. His new I

                    . support replaces a vendor support that was inadvertently omitted during implementation

of minor modification CE-61425.

Evaluation: This support will perform the same function as the vendor support previously performed. Nuclear safety related materials will be used to fabricate the tubing suppost.The support will be installed in accordance with approved station procedures, ne affected portions of the Diesel Generator Lube Oil System will continue to function as described in the UFSAR. No Unreviewed Safety Questions are introduced by the installation of this modificatrion. No Technical Specification changes are required. No UFSAll changes are required. 274 Type: Minor Modification Unit: 1

Title:

Minor Modification CB-61429, Replace valve INW69B

Description:

Minor Modification CE-61429 replaces valve INW69B. The existing valve is a two inch valve, item number 04D-244. It will be replaced with a one inch valve, item number 04D-245. The current two inch valve has frequent operational problems. Seat leakage at valve INW69B results in frequent makeup to the Containment Penetration Valve Water Injection System Tank. Also the leakage threatens Test Acceptance Criteria leakage limits which could affect Technical Specification operability. A pipe hanger associated with this valve will be revised to =ccamadate the new valve. Evaluation: De valve function will not be affected by this modification. It will continue to function as described in the UFSAR and the design basis specification. The function and . 1 operation of the Containment Penetration Valve Water Injection System will not be j affected. Here are no unreviewed safety questions associated with this modification. No  ; Technical Specification changes are required. UFSAR Table 3-104 will be revised. I

                                                                                                                  )

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l l U.S. Nuclear Regulatory Commission April 1,1999 Page 96 of 247 l

    -247     Type: Minor Modification                                              Unit: 2

, Tide: Minor Modification C&61430, Delete damaged and unreliable Resistance Temperature l Detectors from the Ice Condenser Temperature Monitoring Subsystem i

Description:

Minor Modification CE-61430 will delete equipment from the Ice Condenser Temperature Monitoring Subsystem along with the associated wiring and conduits. The equipment to be deleted will be comprised of three groups of resistance temperature detectors identified as floor cooling, wall panel mounted, and wear slab mounted resistance temperature detectors. These items are damaged and unreliable. Repair or l upgrade of these items is not cost justifiable. The resistance temgierature detectors give local indication ofice condenser temperature. These are non Technical Specification l related resistance temperature detectors that do not serve e useful function. l Evaluation: There are no unreviewed safety questions associated with this modification. These resistance temperature detectors do not serve any accident function. No Technical Specification changes are required. No UFSAR changes are required. l 307 Type: Minor Modification Unit: 2

Title:

Minor Modification C&61454 Delete vent valve 2CA290 l Descripdon: Minor Modification C&61454 will delete vent valve 2CA290 and the associated piping and tee. This valve is a construction vent that is no longer used. Currently, there is a leak in one of the socket welds at the location of the tee for the vent line associated with valve 2CA290. Since this vent line is no longer used, the decision was made to delete the vent line and replace the tee with straight piping. Vent valve 2CA290 was originally installed during the construction of the plant for use during system hydrostatic testing. Neither the Operations Group nor the Systems Group use this valve so deleting the valve will not impact operation of the plant. There are no changes to any system design conditions or parameters associated with this change. The replacement piping will be of the same design specification in order to maintain the pressure boundary. The reduction in weight was evaluated for impact to the stress analysis and no support / restraint modifications were required. Removal of vent valve 2CA290 will not affect the operation of the Auxiliary Feedwater System and will not affect the function or design of the Main Feedwater System reverse purge. Removal of the associated piping tee has been evaluated and determined to not significantly reduce the piping pressure losses or increase Auxiliary Feedwater System flow rates. Therefore, there are no changes to inputs for any accident analyses. Evaluation: This modification does not involve an unreviewed safety question. 'Ihe Auxiliary Feedwater System is not an accident initiator. Deletion of this vent valve will have no affect on system operation. The replacement piping is equivalent to the piping that was removed. No Technical Specification changes are required. No UFSAR changes are required. l l

F-U.S. Nuclear Regulatory Commission April 1,1999 Page 97 of 247 l J

                                                                                         .                        i i

42 Type: Minor Modification CB-9451 Unit: 0 l I

Title:

Minor Modification CE-09451, Replace selected Nuclear Service Water System vents and drains with a two inch stainless steel threaded configuration

Description:

Minor Modification CE-9451 replaces selected Nuclear Service Water vents and drains with a two inch stainless steel threaded configuration. The Nuclear Service Water System small bore piping is extremely susceptible to fouling, clogging, and through-wall pitting. This modification will change out specific valves and piping to discourage these degradation mechanisms. This inodification will change out a number of drain or vent valves and their associated piping in the Nuclear Service Water System. These sections will be replaced with stainless steel piping and new ball valves. He old connections will be completely removed at the header. A new carbon steel threaded two inch half coupling will be welded into the header and a piece of two inch stainless steel piping will be threaded into the coupling. The new ball valve is stainless steel and threaded on both . ends and will be threaded into the piping and fitted with a threaded stainless steel nipple l and pipe cap. nreaded fittings are permitted by the Catawba Nuclear Station piping  ! specification. By employing threaded fittings at these drain / vent connections, the threat l of Microbiologically Induced Corrosion in the heat affected zones of stainless steel welds l is eliminated. Stainless steel is less susceptible to general corrosion and pitting. Also by using a ball valve in place of a globe valve, the piping can be cleaned out if it becomes clogged with sitt or corrosion debris. This modification will enhance Nuclear Service l Water System maintenance operations. l t The configuration of the Nuclear Service Water System will remain unchanged.  ! Replacement materials will meet the same ASME or ANSI requirements as the original i parts. The only changes to the system flow diagrams are to show a ball valve instead of a  ; globe valve, increased pipe diameter, if applicable, and to reference the stainless steel material under the appropriate design parameter. Evaluation: This modification does not change any safety aspects of the Nuclear Service Water system. The drains / vents involved serve no safety related functions except as system pressure boundaries. All design parameters will remain the same, except material designation:. He applicable piping classification will remain unchanged. All replacement materials will be procured to the applicable ASMS or ANSI codes. No operating characteristics or failure modes of any systems, structures, or components will change as a result of this modification. Here are no unreviewed safety questions associated with this modification. The changes do not affect the ability of the system to meet its design function nor does it increase the probability of any accidents or equipment malfunction. No technical specification changes are required. UFSAR Figures 9-22 and 9-24 (piping flow drawings) will need to be revised. I t

  ' U.S. Nuclear Regulatory C: * 'on April 1,1999 Page 98 of 247 l

I 191 Type: MiscellaneousItems Unit: 0 i

Title:

" Operable but Degraded" operability evaluation and Compensatory Actions required by Problem Invest 3gation Process Report 0018-0967

Description:

To prevent detrimental interaction with the Emergency Core Cooling System (ECCS) ,  ! the Positive Displacement Pumps will be removed from service and isolated from the j ECCS flowpath. Isolation boundaries include overpressure protection. The Positive = Displacement Pumps and associated piping will be maintained in a water solid condition. l Evaluation: 'Ihe positive displacement pumps were designed to be used during normal plant operation . to provide the charging and seal water flow to the reactor coolant system. The positive displacernent pumps receive a non safety signal to stop running 10 seconds after a safety injection signal actuates. The positive displacement pumps do not receive emergency power. These pumps are not required for accident mitigation. There is no unreviewed safety question associated with this " Operable but Degraded" operability evaluation. No Technical Specification changes are required. No UFSAR changes are required. 192 Type: MiscellaneousItems Unit: 0

Title:

" Operable but Degraded" operability evaluation and Compensatory Actions required by Problem Investigation Process Report 0-C98-0967 Revision 1

Description:

To prevent detrimental interaction with the Emergency Core Cooling System (ECCS) , j the Positive Displacement Pumps will be removed from service and isolated from the j ECCS flowpath. Isolation boundaries include overpressure protection. The Positive j Displacement Pumps and associated piping will be maintained either filled or drained. 1 Evaluation: The positive displacement pumps were designed to be used during normal plant operation to provide the charging and seal water flow to the reactor coolant system. The positive displacement pumps receive a non safety signal to stop rur.ning 10 seconds hfter a safety injection signe.1 actuates. The positive displacement pumps do not receive emergency power.These pumps are not required for accident mitigation. There is no unreviewed  ; safety question associated with this " Operable but Degraded" operability evaluation. No j Technical Specification changes are required. No UFSAR changes are required. l i 1 1 I'

r; N.S. Nuclear P ,, 'to y Comunission April 1,1999 - l -- Pase 99 of 247. 196 : Type: Miscellaneous items Unit: 0 l [

Title:

Administrative Changes to the Selected Licensee Commitments resulting from the l conversion to the Improved Technical Specifications Descriptioni Dis evaluation was ' erformed p to address the changes required to the Selected Licensee

Commitment manual associated with the conversion to the Improved Technical Specifications. He changes primarily involve rdocation of existing Technical Specification requirements to the Selected Licensee Commitments Manual (these must be

; approved by the NRC as a part of the Improved Technical Specification conversion l

                      ~ heense amendment submittal). The changes also involve other administrative changes
                     . made to the Selected IJcensee Commitment Manual as a part of the Improved Technical l                       Specification implementation effort, such as reference changes, editorial changes to promote clarity, and the revision of certain Selected Licensee Commitment requirements to make them consistent with improved Technical Specification requirements. The following is a discussion of the changes:

A-1) In accordance with the conversion to Improved Technical Specifications (ITS), certain current Technical Specifications (CTS) requirements are relocated to the Selected Licensee Commitments (Slf) Manual. These existing requirements are those which did not meet the criteria for inclusion within the ITS as defined by 10 CFR 50.36 (c)(2)(ii) and items within individual CTS specifications which were of a detail nature or not

                     . necessary to deruonstrate operability consistent with the safety analysis. All reformatting and renumbering of these relocated requirements are in accordance with the existing SLC Manual. The reformatting, renumbering, and rewording process involves no technical changes to CTS which are being relocated to the SLC Manual. De Bases of the CTS are relocated with the relocated requirement. Minor changes may be made to the Bases to incorporate terminology used within the SLC or to provide additional detail consistent
                      -with the UFSAR for those cases where the CTS Bases did not address the relocated item.

A-2) De apphcability section 16.2 of the SLC Manual is revised consistent with the ITS. The existing requirements were consistent with the CTS to avoid confusion when implementing SIE Manual requirements. Derefore, these changes were necessary to make the SLC Manual consistent with the changes made during the conversion to the ITS. Any changes in the applicability from the CTS to the ITS is justified in the license amendment request associated with the ITS conversion. SLC 16.2.6 was not revised since the format of the ITS and SLC are different and the additional ITS language related to completion times and frequencies is not compatible with the SLC. De 25% freqbency extension is already included in SLC 16.2.6 consistent with the CTS and the ITS.' A 3) He SIE definition for OPERABLE is revised explicitly to state " normal or ernergency electrical power

• The non-specific necessary . electrical power" requirement is intended to be a requirement for only one source of power to be able to declare l OPERABILITY. Similarly,"specified function" could be misinterpreted. The SLC is revised to address 'specified safety function (s)" and not unintentionally encompass any non - safety functions a system may also perform. De revised definition provides

                    . clarification of the current requirement without any modification ofintent. This change is j-                      consistent with the definition used in the ITS.

i l l.

r , u a e i i , d.S. Nuclear Regulatory Comissission April 1,1999 Page 100 of 247 e L A-4) ne SLC definition of OPERATIONAL MODE has been clarified to include "with fuel in the reactor vessel." This is editorial in nature since the statement is already included in SIL Manual Table 16.3-1. Herefore, the change to the definition does not ' l l constitute a change from existing requirements or practices. This change is consistent l with the definition used in the ITS. A-5) New definitions are added to the SLC Manual consistent with existing delimitions

                         'used in the CTS. Dese definitions are associated with requirements which are being

! relocated to the SIE Manuals No changes are being made to the definitions which are i being relocated. Some definitions are also being added to the Sir Manual which are also I retained in the ITS. These definitions are added consistent with the language used in the ITS. Any changes to these definitions from the CTS to the ITS isjustified in the license amendment request associated with the ITS conversion. Existing Sif definitions have minor editorial revisions for consistency with the ITS. Separate Discussion of Changes are provided for changes which are more than editorial. A-6) The SLC definition of STAGGERED TEST BASIS has_bcen modified to be consistent with its usage throughout the ITS, ne intent of the frequency of testing components on a STAGGERED TEST BASIS is not changed. The ITS definition allows the Surveillance intetval to be specified independent of the number of subsystems. For example, for a three channel system, the CI'S would specify quarterly testing on a STAGGERED TEST BASIS which results in one channel being tested each month (three equal subintervals). Under the ITS definition, the frequency would be monthly on a STAGGERED TEST BASIS and one channel would be tested each month so that at the end of three months, all channels are tested. Thus, there are no net changes in the staggered interval. His represents an editorial preference to the current SLC Manual presentation and is an administrative change. A-7) The average reactor coolant temperature threshold for MODES I and 2 provided in SLC Table 16.3-1 is changed to NA (not applicable). Individual ITS and SLC specify any applicable average reactor coolant temperature limits in the Applicability when necessary, therefore, the 350 degree F temperature threshold does not provide any useful

                                                                     ~

information. Removal of this limitation does not substantively change the requirements for operation in MODE 1 or MODE 2 since the reactivity threshold is unchanged. His change is considered administrative in nature. This change is consistent with the ITS. A-8) The definitions of Hot Shutdown and Cold Shutdown in SLC Table 16.3-1 have been revised to provide clarity, completeness and avoid any potential misinterpretation. Specifically, the new footnote added stating *all reactor vessel head closure bolts fully tensioned" eliminates a potential overlap in defined MODES. For example, when the vessel head is detensioned, both the definition of Refueling and Cold Shutdown could apply, dependent on temperature, it is not the intent of the Sir Manual to allow an option of whether to apply Refueling applicable COMMITMENTS or to apply Cold Shutdown applicable COMMITMENTS. His change is editorial in nature since the intent of the existing requirement is clarified to reflect actual industry practice. This change is consistent with the ITS.

b U.S. Nuclear Regulatory Cosamission

      . April 1,1999 Page 101 of 247 p                        A-9) ne note 4, in SIL Table 16.3 1. which includes with the head removed" has been revised. Since the vessel head can only be removed if the head closure boks are less than
                                                                                                                      )

fully tensioned, there is no purpose in including "or with the head removed" as part of the clarification to the MODE 6 definition. This change is consistent with the ITS.

                      ' A 10) CTS 4.4.2.1 requires surveillance testing of code safety valves in accordance with Specification 4.0.5. Sir 16.5-2 requires the testing be done in accordance with the l

Inservice Testing Program. Specification 4.0.5 of the CTS is moved to the -! Administrative Controls (ITS 5.5.8) and requires that a testing program in accoalance with 10 CPR 50.55a be established. De details of the testing are contained within the program. His change is administrative and is consistent with the ITS.

A-ll) CTS 4.4.10 requires surveillance testing of ASME code class 1,2, and 3

! components in accordance with Specification 4.0.5 and requires testing the RCP flywheel per Regulatory Guide 1.14. SIf 16.5-5 requires the testing of code class 1,2 and 3 components be done in accordance with the Inservice Testing Program. RCP flywheel inspection requirements are retained in the ITS. Specification 4.0.5 of the CTS is moved I to the Administrative Controls and requires that a testing program in accordance with 10 l CFR 50.55a be established. The details of the testing are contained within the program. l_ This change is administrative and is consistent with the ITS. A-12) CTS 4.5.2.c requires a visual inspection of containment to verify loose debris which could be transported to the containment sump has been removed. This requirement is relocated to SLC 16.6-1. A statement of Commitment and Remedial Actions are also added consistent with the relocated Testing Requirement since the LCO statement and Actions of CTS 3.5.2 are not directly applicable to this relocated Surveillance. The remainder of CTS 3.5.2 is retained in ITS 3.5.2 and any changes are discussed in the ITS submittal. %e addition of the Commitment statement and Remedial Action is considered administrative since the Surveillance Requirement to verify absence of debris is performed during shutdown and successful performance would inherently require removal l of anydebris. l A-13) CTS 3.1.3.3 specifies requirements for rod position indication in Modes 3,4, and

5. These requirements are retained in Sif 16.7-6. The associated surveillance requirement 4.1.3.3 is retained in the ITS, therefore, testing requirements performed to satisfy the ITS requirements are applicable to this SLC and are not duplicated, l-A 14) References to CTS 6.9.2 for Special Reports is deleted from specifications within the Sir Manual As part of the conversion to ITS, this CTS section was deleted from the Technical Specifications. %e relocated requirements and existing SIf already contain the requirements for submitting the report and the only additional information provided by CTS 6.9.2 was that the report be filed pursuant to 10 CFR 50.4. All correspondence with NRC is prepared pursuant to applicable CFRs, therefore, this statement is redundant and unnecessary.

                      ' A-15) References to CTS 3.0.3 exceptions in specifications relocated to the SIL Manual are deleted. Dere are no corresponding SLC requirements which require that an i                       exception be taken, therefore, the removal of this statement is administrative.

(; u. 1 l L  ! l l: i L

p, j l I,o I U.S. Nuclear Regulatory Com-lalon Apdf 1,1999  ! l . Pase 102 of 247-P [  : A-16) CTS 3.8.1.1 provides actions for the DG operating at power greater than 5750 kW l- and for the cathodic protection system. CTS 4.8.1.1.2.b,4.8.1.1.2.d,4.8.1.1.2.g.1,  ! '~ 4.8.1.1.2.g.11,4.8.1.1.2.g.14,4.8.1.1.2.1.2, and 4.8.1.1.2.i.3 specifies testing requirements )

                     . for the DG which are not retained in the ITS. Rese requirements are relocated to SIE          j l

16.8-2. A statement of Commitment is also added consistent with the specific relocated ] l requirements since the LCO statement of CTS 3.8.1.1 is not directly specific to the J relocated requirements. De remainder of CTS 3.8.1.1 is retained in ITS 3.8 and any changes are discussed in the ITS submittal. The addition of the specific Commitment

   '                   statement and Remedial Action is considered administrative since these are consistent I

with the requirements and actions taken in the CTS. A-17) CTS 3.1.2.2 and 3.1.2.4 specify requirements for two boron injection flow paths and two charging pumps in Modes I,2,3, and 4. An exception to these requirements in Mode 4 only requires one flow path and one charging pump operable when the RCS temperature is <=285 F due to low temperature overpressure considerations. CTS 3.1.2.1 and CTS 3.1.2.3 specify requirements for one boron injection flow path and one charging pump in Modes 5 and 6. He Applicability of all of these CTS requirements are modified consistent with the exception to require only one pump and flow path operable in Mode 4 when RCS temperature is <= 285 F. These requirements are relocated to the SLC Manual as 16.9-7 through 16.9-10. These changes are administrative since no technical changes are made to the existing requirements. These changes are consistent with similar changes made to the ITS. A 18) CTS 4.7.5.d requires recording Lake Wylie temperature on a daily basis during the summer months. His requirement is relocated to SLC 16.9-14. A statement of Commitment and Remedial Action are also added consistent with the relocated Testing Requirement since the LCO statement and Actions of CTS 3.7.5 are not directly applicable to this relocated Surveillance. He remainder of CTS 3.7.5 is retained in ITS 3.7.8 and any changes are discussed in the ITS submittal. The addition of the Commitment Statement and Remedial Action is considered administrative since the Surveillance Requirement to record lake temperature is not included in the ILO requirement. The Remedial Action is consistent with the operation of the system as described in the UPSAR. A 19) CTS 3.3.3.7 requires the control room area ventilation system be placed in filtered l operation when the chlorine detectors are inoperable. The system operates in this mode continuously and is not designed to operate without filtration. SLC 16.6-4 clarifies this ambiguous action to indicate that the system operation is in the high chlorine protection mode which is filtered but does not use outside pressurir.ing air. His is consistent with  ! the current system design and operating practice. , A-20) CTS 4.7-8.e requires the NRC Regional Administrator be notified prior to using a i different snubber sampling plan. This requirernent is revised to delete the reference to the Regional Administrator. The requirements in 10 CFR 50.4 require correspondence be directed to the NRC Document Control Desk. As a matter of practice, all correspondence is copied to the Regional Administrator. Derefore, this change is administrative and in compliance with regulations.

U.S. Nuclear Regulatory Commission April 1,1999 Page 103 of 247 Evaluation: The changes described above were evaluated and found to contain no unreviewed safety ) questions. No technical requirement of any Selected Licensee Commitment was affected by these changes. He manner in which the plant and accident mitigating equipment are designed, operated, and maintained was not affected by these changes. No accident I' probabilities or consequences were affected by these administrative changes. Similarily, no probabilities or consequences of equipment malfunctions were impacted. No possibility of e new type accident or equipment malfunction was created. No safety margins were reduced by these changes. No Technical Specification changes are required (other than the change from the Current Technical Specifications to the Improved Technical Specifications). Changes to the Selected Licensee Commitment Manual j (UFSAR Chapter 16 are required).  ; l 212 Type: MiscellaneousItems Unit: 0 l

Title:

Analysis of Tygon Tubing used for the Unit Vent Continuous Sampler .- I

Description:

Tygon tubing is being used is being used for the Unit 1 and Unit 2 Vent continuous samplers and associated equipment. The samplers are required per Chapter 16 of the  ; UFSAR (Selected Licensee Commitments). The continuous samplers obtain their l samples via tubing connections to sample taps at the inlet and out.let of the Radiation Monitor EMF 35,36 and 37 skid package (Unit Vent Airborne Radiation Monitors). The Unit Vent Airborne Radiation Monitors and the Unit Vent Continuous Samplers receive a sample stream from the Unit Vent. The Continuous Samplers are provided to facilitate a daily surveillance of the gaseous effluents released through the Unit Vent. The Radiation Monitors and the Continuous Samplers are not safety related. The sample stream for the EMF 35,36 and 37 skid package is provided through non-safety related sample lines. The semple stream for the Continuous Samplers is provided via tygon tubing connected to sample taps on the skid sample lines. Evaluation: He use of tygon tubing for the Unit I and Unit 2 Unit Vent Continuous Samplers and associated equipment has no impact on nuclear safety. The use of tygon tubing neither increases or decreases the margin of safety as described in the Technical Specifications. De use of tygon tubing in the Unit Vent Continuous Sampler application affects only non-safety related plant equipment. There are no unreviwed safety questions associated with use of tygon tubing for this application. No Technical Specification changes are required. No UFSAR changes are required. l t-

E 1 U.S. Nuclear Regulatory Comunission Apdl 1,1999

  ; Page 104 of 247 6

203 . Type: MiscellaneousItems Unit: 0

Title:

Calculation CNC-1553.26-00-0193 Revision 0, Increased Burnable Poison Rod Assembly B4C Concentrations

Description:

Calculation CNC-1553.26-00-0193 documents the analysis for the increased Burnable Poison Rod Assembly B4C concentrations used in Catawba Core Designs. , i Evaluation: Here are no unreviewed safety questions associated with the increase in B4C poison i concentrations in the BPRA rodlets up to 4.0 W/O. FCF Analysis has shown that the change does not affect component lifetimes. He increase in B4C concentration does not increase the probability of an accident analyzed in the UFSAR. Fuel assemblies and BPRAs are not accident initiators. His change does not affect the ability to detect a fuel misionding scenario.' No new credible failure modes are created. UFSAR Sections 4.2 and 4.4 required changes per this calculation. These changes have already been made and were included in the 1997 UFSAR Update. No Technical Specification Changes were

                                                                                                                    ~

required. 204- Type: Miscellaneous Items Unit: 0

                 'Iltle: Calculation CNC-1553.26-00-0194 Revision 1. Replacement of Secondary Sources

Description:

Calculation CNC-1553.26-00-0194 Revision 1 addresses the replacement secondary sources to be used at Catawba. He replacement secondary source will be a double - encapsulated design. Evaluation: There are no unreviewed safety questions associated with ti.e change in design for the 1 replacement secondary sources. The probability or consequences of accidents analyzed in j the UFSAR are not affected. Westinghouse analyses have shown that the change does not i affect component performance and should maintain designed lifetimes. Fuel assemblies  ; and Secondary Sources are not accident initiators. Changes are required for the UFSAR  !

                         - Section 4.2.1.3.3. No changes to the Technical Specifications are required.                ;

i I l' ? u-

I f. l. f- U.S. Nuclear Regulatory Conunission '. l April 1,1999 l Page 105 of 247 < 167 Type: MiscellaneousItems Unit: 2.

Title:

Catawba Nuclear Station Pump and Valve Inservice Testing Program (Rev 24)

Description:

He Catawba Nuclear Station Pump and Valve Inservice Testing Program (Rev 24) was

                      . changed to allow adding four Nuclear Service Water System valves (2RN846A, 2RN847A,2RN848B,2RN849B) to the program. Also a revision will be made to l                      : Section 4.5 on thermal expansion check valves. Valves 2RN846A and 2RN848B are the

[ Diesel Generator 2A and 2B Heat Exchanger returns to the Standby Nuclear Service Water Pond, respectively. Valves 2RN847A and 2RN849B are the Diesel Generator 2A and 2B Heat Exchanger retums to Lake Wylie, respectively. These valves were previously part of the IST program but were removed temporanly from the program to allow repair of a faulted section of piping. Minor Modification CE-8592 removed from service the 2A/2B Diesel Generator Cooling Water Heat Exchanger return to Lake - Wylie. Valves 2RN846A and 2RN848B were opened with power removed,and valves 2RN847A and 2RN849B were closed with power removed, in these positions, testing was not required. Nuclear Station Modification CN-50475/00 repaired the faulted section l of piping and returned the valves and piping back to the original operating scheme. Therefore, testing of valves 2RN846A,2RN847A,2RN848B and 2RN849B had to resume. These valves will be added back to the program. IWV testing has already resumed. Section 4.5, Thermal Expansion Check Valves, which was recently added to the Program Document, is being revised to add more detail. The previous revision did not provide a l complete list of thermal expansion check valves and did not provide sufficient ! categorization of the various applications of these check valves. Though the scope is ! broader, additional testing is not required. All thermal expansion check valves which have testing requirements beyond thermal expansion protection are being tested for that function. Those thermal expansion check valves which do not have testable requirements will be added to sections 3.2 and 4.2 of the Program Manual with a note referring the , reader to section 4.5 of the Program Document. This will insure a comprehensive list of l valves is reflected in the valve sections and provide a tie back to the Program Document [ which provides the logic for not testing certain functions. t Evaluation: No unreviewed safety questions are created as a result of this activity. Both of these

                      - changes are editorial. The required testing is being performed. %ese changes do not change the facility as described in the UFSAR. There is no change to the test procedures, methods, or acceptance criteria of testing already being performed for these valves. No Technical Specification changes are required. No UFSAR revisions are required.

l

- U.S. Nuclear Regulatory wh Apdf 1,1999 Pase IN of 247 - f 168. Type: MiscellaneousItems Unit: 0 Dele: Catawba Nuclear Station Pump and Valve Inservice Testing Program (Rev 24)

Description:

The CMawba Nuclear Station Pump and Valve Inservice Testing Program (Rev 24) was changed to allow the following changes: Change actuator type on valves 1(2)SV009, 1(2)SV010,1(2)SV11,1(2)SV12; change valves 1(2)WU)21 and 1(2)WLO24 to Category B, change valves 1(2)NV206 and 1(2)NV218 to passive. Also, three

                     . justification for deferrals will be revised to discuss the closing function.

Evaluation: No unreviewed safety questions are created as a result of this activity. These changes are editorial. The required testing is being performed. No Technical Specification changes are required. No UFSAR revisions are required. 211 Type: Miscellaneous Items Unit: 0

           . Utle: Catawba Nuclear Station Pump and Valve Inservice Testing Program Revision 24

Description:

The Catawba Nuclear Station Inservice Testing Program will be changed to allow the following: For valves 1(2)VB85 - Change to Active, add full stroke closed, JFD needed to defer testing For valves 1(2)VS56 - Change to Active, add full stroke closed, JFD needed to defer testing For Valves 1(2)VY16 - Change to Active, add full stroke closed, JFD needed to defer testing Valves 1(2)VB85,1(2)VS56 and 1(2)VY16 are check valves which are containment isolation valves. These valves had previously been categorized as A/C Active with a full stroke open and closed (FSO/C) with the closed test deferred to refueling and verified closed with Type C testing. A 10CFR50.59 evaluation was conducted to remove the FSO/C, make the valve Passive and only require Type C tests. This was based on the expectation that the Type C test would be conducted at the end of the refueling outage to ensure tha valves were closed, thus making the valves Passive with no full stroke closed

                   . requirement. It has been determined that Typ ' C testing can also be conducted at the beginning of the outage. These penetratios wuld be used during the outage (valves could be opened) requiring a movement to close to return to its safety position. "Ihe new requirements will be as described before, without the requirement to open, which are:

FSC-Q, Active with a JFD to defer the stroke closed testing to refueling and conducted

                   ' with the Type C test.

Evaluation: 'Ihere are no unreviewed safety questions associated with this activity. No Technical Specification changes are required. No UFSAR changes are required.

,u U.S. Nuclear Regulatory Comunission April 1,1999 Pane 107 of 247 303      Type: Miscellaneous Items                                           Unit: 0 T1tle: Catawba Nuclear Station Pump and Valve Inservice Testing Program Revision 25

Description:

'Ihe following changes were made in the Catawba Nuclear Station Pump and Valve Inservice Testing Program in Revision 25. For valves INDl!6,2ND116, INDil7 and
                      . 2ND117 Add "LT" lesting requirement. For valve 2 Nil 01 - delete the LT testing requirement. For valves INW17, INW21, INW24, INW27, INW40, INW43, INW47, INW50, INW70, INW74, INW77, INW80, INW86, INW89, INW92, INW95, INW98, INW101, INW109, INW111, INW114, INW120, INW121, INW123, INW124, INW125, INW126, INW127, INW128, INW129, INW130, INW131, INW132, INW133, INW135, INW136, INW138, INW139, INW140, INW141,
                      ' INW147, INW148, INW159, INW160, INW163, INW164, INWI68, INW169, INWl71, INW172, INW178,1 NWI79, INW183, INW184, INW188, INW189, INW193, INW194, INWi%, INW197, INW201, INW202, INW205, INW206, INW209, INW210, INW213, INW214, INW218, INW219, INW223, INW224, INW230, INW231, INW235, INW236, INW240, INW241, INW245, INW246 and 2NW 17, 2NW21, 2NW24, 2NW27, 2NW40, 2NW43, 2NW47, 2NW50, 2NW70, 2NW74, 2NW77, 2NW80, 2NW86, 2NW89, 2NW92, 2NW95, 2NW98, 2NW 101, 2NW 109, 2NW i l l, 2NW 1 14. 2NW 120, 2NW 121, 2 NW 123, 2NW 124, 2NW 125, 2NW 126, 2NW 127, 2NW 128, 2NW 129, 2NW 130, 2NW 131, 2NW 132, 2NW 133, 2NW 135, 2NW 136, 2NW 138, 2NW 139, 2NW 140, 2NW 141, 2NW 147, 2NW 148, 2NW 159, 2NW 160, 2NW 163, 2NW 164, 2NW 168, 2NW 169, 2NW 171, 2NW 172, 2NW 178, 2NW 179, 2NW 183, 2NW 184, 2NW 188, 2NW 189, 2NW 193, 2NW 194, 2NW 196, 2NW 197, 2NW201, 2NW202, 2NW205, 2NW206, 2NW209, 2NW210, 2NW213, 2NW214, 2NW218, 2NW219, 2NW223, 2NW224, 2NW230, 2NW231, 2NW235,2NW236,2NW240,2NW241,2NW245,2NW246 the closing function was removed as an IST requirement. For valves ICA23, ICA28, ICA33,2CA23,2CA28 and 2CA33 the closing function was removed as an IST requirement. For valves ICA37, ICA41, ICA45, ICA49, ICA53, ICA57, ICA61, ICA65 and 2CA37,2CA41,2CA45, 2CA49,2CA53,2CA57,2CA61,2CA65 Justification for Deferal (JFD) CN-CA-01 was revised to add these noting sample disassembly for reverse flow testing. 'the following
                     . new generic relief requests were added to the program: CN-GRV-01, CN-GRV-02, CN-GRV-03.

Evaluation: These changes do not change the facility as described in the UFSAR. There are no changes to test procedures, methods, or acceptance criteria of testing already being performed for these valves. Therefore the activities being evaluated are not tests or experiments, nor do they appear significant enough for inclusion in the UFSAR. These changes are considered editorial changes that do not create any new failure modes or operating characteristics. There are no umeviewed safety questions associated with these changes. No Technical Specification changes are required. No UFSAR changes are required. L._.. .

p  ; I U.S. Nuclear Regulatory Commission April 1,1999 Page 108 of 247 l 200 Type: MiscellaneousItems Unit: 0  !

Title:

Catawba Unit 1 Fuel Cycle 11 Reload Safety Evaluation

Description:

A safety evaluation was performed for the Catawba Nuclear Station Unit I Cyclell (CICI1) core reload in calculation file CNC-1552.08-00-0274. This calculation file addresses the issues as they are affected by the new core. The impact of any other plant changes which might be made concurrent with the refueling outage are not addressed in the calculation.

The CICII Reload Design Safety Analysis Review (REDSAR), performed in accordance with the Nuclear Engineering Division workplace procedure NE-102, *Workplace Procedure for Nuclear Fuel Management," serves as the safety review for the unreviewed safety question evaluation. The Nuclear Design and Reactor Support (NDRS) section of the REDSAR checklist indicates the need to further evaluate the power shapes at the limiting state-points for the steam line break and dropped rod departure from nucleate boiling (DNB) evaluations. These evaluations, documented in CNC-1552.0840274, f combined with the REDSAR prove that the UFSAR Chapter 15 accident analyses remain bounding with respect to CICI1 safety analysis physics parameters.

l l Also in CNC-1552.08-00-0274, the results of the boration requirements evaluation show , j- that the current Core Operating Limits Report limits ensure that sufficient borated water i t exists in either the Boric Acid Tank or the Refueling Water Storage Tank to borate to the required shutdown boron concentrations except for the Doric Acid Tank limit to remain at 1.3% shutdown margin in going from IIFP to 200 degrees F. The Boric Acid Tank l volume will be increased to 11,851 gallons to achieve the 1.3% shutdown margin for ! Modes 1 to 4. l l j Evaluation: The CICI1 core reload is similar to past cycle core designs, with a design generated using NRC approved methods. Additionally, no Technical Specification changes specifically related to the operation of the new CICll core are required The unreviewed safety question evaluation resulted in the conclusion that there are no unreviewed safety questions concerning the CIC11 core reload. No UFSAR changes are required. I ll

i l U.S. Nuclear Regulatory Conunission April 1,1999 Page 109 of 247 i 256 Type: Miscellaneousitems - Unit: 2 T1tle: Catawba Unit 2 Cycle 10 Core Reload i Descripdon: ne Catawba Unit 2 Cycle 10 Core Reload is similar to previous cycle core designs. It was generated using NRC approved methods. The core components for this reload are l similar in design to those used in previously approved cycles and are arranged in a , typical configuration. The operational limits for this reload have been developed using methods and codes previously approved by the NRC and continue to reflect the limitations imposed by the safety and design / performance analyses. Rod position limits ensure that adequate shutdown margin is available at all times in core life. Evaluation: There are no unreviewed safety questions associated with this core reload. No Technical  ! Specification changes are required which are specifically related to the operation of the j cycle 10 core. No UFSAR changes are required. [ 186 Type: Miscellaneousitems Unit: 0

Title:

Change to Selected Licensee Commitment 16.11-2

Description:

Selected Licensee Commiunent 16.11-2, Table 16.11-3 item 3.d will be revised to delete the Analog Channel Operatical Test (ACOT) for the Monitor Tank Building Waste Liquid Effluent Line. Currently this item requires a quarterly ACOT to be performed. Since this instrument loop only provides indications and does not provide any control or I alarm function an ACOT is not required. l Evaluation: There are no unreviewed safety questions associated with this activity. Deleting this ACOT will not change the function of the instrument loop or the associated system (liquid radwaste system.) No Technical Specification changes are required. A UFSAR change to l Selected Licensee Commitment 16.11 2 Table 16.11-3, item 3.d is required. I t I l 1 I l l ! I L i m 1

V i I U.S. Nuclear Regulatory Comndssion Apdl 1,1999 Paste 110 of 247 f 205- Type: MiscellaneousIte;ns Unit: 0

Title:

Change to Selected Licensee Commitment 16.7 3 concerning Meterological Equipment

Description:

His change to Selected Licensee Commitment 16.7 3 (Meterological Equipment) changes the testing requirements for Wind Speed, Wind Direction and Air Temperature Instrumentation. He testing requirements for a CHANNEL CALIBRATION, as it pertains to the Meteorological Monitoring System Wind Speed, Wind Direction and Temperatures channels will be revised to exclude tower signal cables. He definition of a CHANNEL CALIBRAT'ON in Technical Specification states in part; ". the CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated." The station's Meteorological Tower is designed to lower the sensors to ground level, via an instrument elevator system. When the elevator is lowered, the sensors are disconnected from the tower signal cables. Therefore, the tower signal cables can not be " encompassed" as part of a CHANNEL CALIBRATION. The types of signals produced by the Meteorological Wind Speed and Temperature Sensors negate any adverse effect that the tower signal cables could induce on channel accuracy. The intent of the CHANNEL CALIBRATION testing requirements, as defined in Catawba's Technical Specification, will be satisfied by employing a Meteorological System specific Instrument Calibration method.1) The Wind Speed and Wind Direction sensors will be either bench calibrated, or calibrated on the tower in a test configuration, which excludes the tower cables. The Temperature Sensors and the Temperature Signal processing module will be bench calibrated as a unit. 2) Test signals will be injected into the Signal processors using test equipment, or a signal will be generated from the Sensor at the tower, via a tower test cable, and all downstream components will be checked / calif n ed. 3) The tower sensor will be returned to their normal operating position, and a CHANNEL CHECK will be performed by the maintenance technicians. This check will verify continuity of the tower signal cables and that they are functioning correctly. Neither the Wind Speed Cups and Sensors nor the Wind Direction Vancs and Sensors require testing as a unit in a wind tunnel. There is no requirement to calibrate the Wind Speed channels if the Wind Speed Cups are replaced, nor to calibrate the Wind Direction channels if the Wind Vanes are replaced. Evaluation: There are no unreviewed safety questions associated with this Selected Licensee Conunitment change. The Meterological Monitoring System does not perform any contr'ol functions and is not a part of any process system boundary. The system is a passive monitoring system and cannot act as an accident initiator. No Technical Specification changes are required. Changes are required for UFSAR Section 2.3.3.2 and Chapter 16 (Selected Licensee Commitments) Section 16.7 3.

l .-

      , i U.S. Nuclear Regulatory Comunission April 1,1999 -

L Page 111 of 247 {  ;,; , s t' 193 Type: MiscellaneousItems' Unit: 0 l

Title:

Change to Selected Licensee Commitment 16.7-4 -Imose Parts Detection

Description:

Selected Licensee Commitment 16.7-4 Remedial Action requirements for inoperable Imose Parts Monitoring System channels will be revised. The current requirements of "One or more channel inoperable" will be changed to "All channels within one or more collection regions inoperable". In addition, the collection regions will be listed. Also the testing requirements for an Analog Channel Operational Test (ACOT) will be changed. The current definition for ACOT, in the Technical Specifications, assumes that each channel is a separate and discrete entity. The Loose Parts Monitoring System is a dual digital computer based system. Each channel begins as a discrete analog circuit

                          ' where individual signals are digitized and processed in the computers. The signals are kept separate for most of the processing but combined in the Alarm Detection section.

Because of the detection logic, used during normal operation, each individual channel cannot be tested separately. However, the intent of the ACOT testing requirements, as defined in Catawba's Technical Specification, can be met by an alternative test method as follows: 1) Each Analog Channel will be checked by monitoring the actual field signals.

2) Each Digital Channel will be checked by verifying that computer generated data is within acceptable limits for unit operating conditions. 3) De Alarm Detection Logic will be tested by injecting a signal into a system test channel and verifying the control room annunciator actuates.

Evaluation: The Imose Parts Monitoring System (LPMS) is an electronic system that monitors the Reactor Coolant System for metal-to-metal impacts in the primary coolant loop and alerts the Contro1 Room Operators. He Control Room Operator may then take the appropriate measures to further identify the severity of the event. Once a loose part is detected and is identified as an authentic loose part, analysis can be performed to estimate the mass and location of the part. With the mass and location information, damage estimates can be evaluated to determine the correct course of action that will minimize damage while maximir.ing plant efficiency. However, the LPMS was not installed to establish initial conditions for design basis accidents or to mitigate the consequences of an accident or transient. De LPMS can provide early detection for the presence ofloose parts that l could lead to Steam Generator Tube Rupture or flow blockage that could result in DNB. His is a benefit of the system and not a requirement for the system. He LPMS is non-safety, and does not perform any control or mitigation functions. His change includes both administrative reporting requirements and work to be performed on control room equipment. No field (sensor) work on the Reactor Coolant System will be affected by this change. The LPMS is a passive monitoring system, not an accident initiation, or l- mitigation system. While LPMS interfaces with equipment important to safety (Reactor

                         - Coolant System piping, Steam Generators), these changes do not alter these interfaces.

There are no unreviewed safety questions associated with this change. No Technical Specification changes are required. A change is required for Selected Licensee Commitment 16.7-4. 1

r U.S. Nuclear Regulatory Commission April I,1999 Page 112 of 247 190 Type: Miscellaneousitems' Unit: 0

Title:

Chemistry Management Procedure 3.4.17.1, Revision 31

Description:

Chemistry Management Procedure 3.4.17.1 Revision 31 ir.corporates two sets of changes. The upper limit of boron concentration in the Spent Fuel Pool will be set to 3500 parts per million (ppm) with an operating range of 3200-3300 ppm. This is based on operational considerations. Also, administrative limits are being placed on the dose equivalent I-131 (DEI) specific activity in the Reactor Coolant System. Specifically, the upper limit on equilibrium DEI specific activity is .11 microcuries/ gram while the upper limit on transient DEI specific activity is 15 microcuries/ gram. These limits are conditions of operability as given in an operability evaluation associated with Problem investigation Process report 0-C97 3621 Evaluation: No physical changes were made to the plant. With reference to the Spent Fuel Pool boron concentration, the operating range and upper limit are above the limit specified in the Core Operating Limits Report for the current fuel cycles. With reference to the DEI limit. these limits are conditions of operability stated in an

• Operable but Degraded" operability evaluation. These limits were placed after single failures were identified which may degrade the ability of the operators to throttle auxiliary feedwater flow and prevent steam generator overfill. (Reference LER 413/97-009-01). 'Ihe degraded condition will be fixed by a plant modification. The limits on DEI are less than (1) the corresponding values assumed in the safety analyses of all associated design basis events (2) the limits of Technical Specification 3/4.4.8, and (3) the license conditions associated with safety evaluation report for the facility evaluation amendment 159/151. Neither any plant equipment nor operation of any plant equipment is affected by these changes to Chemistry Management Procedure 3.4.17.1. There are no unreviewed safety questions associated with this change. No Technical Specification changes are required. No UFSAR changes are required.

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4 U.S. Nuclear RegulatorfComunission

                    ' April 1,1999 Page 113 of 247 209         Type: Miscellaneousitems                                                                      Unit: 0
                                   - litle: Closing of Valve CM-33 as an interim conservative measure

Description:

Valve CM-33 (Hotwil High Level Control Valve) provides mini-flow protection for the

non-safety related CW= Hotwell Pumps and a means of hotwell inventory control on high hotwell level. Valve CM-33 has the potential to supply hot water to the Auxiliary Feedwater System suction in excess of the Auxiliary Feedwater System piping design temperature in the event of a loss of condenser vacuum accident. 'the Auxiliary Feedwater System is not required for mitigation of the loss of vacuum accident; however,

                                             ' the Auxiliary Feedwater System is required to provide core cooling in the long term following the accident. This potential adverse system interaction could render the Auxiliary Feedwater System inoperable. To prevent this from happening, valve CM-33 l                                              .will be isolated by closing other valves in the hotwell high level control flowpath..

Evaluation: During normal operation miminum flow protection can be provided by the Main Feedwater Pump Re-circulation valves. Other means can be used to reduce hotwell inventory if required during normal operation. Due to heater drain pump operation post trip, the hotwell pumps will be " dead-headed" for 2.5 minutes and the hotwell pump discharge design pressure will be exceeded by 1.5%. The " dead-head" condition is acceptable per vendor evaluation.' The piping over-pressurization is allowed by ANSI B31.1 for short term transient operation. The affected portions of the Condensate Makeup (CM) System are not required to mitigate any accidents evaluated in the UFSAR, There are no unreviewed safety questions associated with the isolation of valve . CM-33. No Technical Specification changes are required. No UFSAR changes are required until a final resolution of the issue is reached.

p-L I U.S. Nuclear Regulatory Conunission April I,1999 Page 114 of 247 l { 275 Type: Miscellaneous Items Unit: 0 l

           'lltle: Compensatory Action for Auxiliary Building Ventilation System High Flow and Iodine l                     Penetration Concern

Description:

This compensatory action was developed to support an " Operable but Degraded"

       ,             condition on the Auxiliary Building Ventilation System. The " Operable but Degraded" condition exists because more restrictive limits have been imposed on the Auxiliary Building Ventilation System than those allowed by Technical Specifications. Currently, the Technical Specifications allow the Auxiliary Building Fihered Exhaust Units to operate at a flow rate between 27,000 and 33,000 cubic feet per minute (cfm). Because the Auxiliary Building Filtered Exhaust Trains share common ductwork, the procedure to verify these flow limits (PT/0/A/4450/001C) tests the filters with both trains in operation.

This presents a problem if the dual train alignment flow rates are in the upper end of the allowable flow range and one of the trains trips. This will cause the flow rate in the operating train to increase due to the overall lower flow and resistance in the ductwork. In such a situation, the flow in the single train alignment could exceed the maximum design basis flow rate associated with a 95% efficient filter and dose analysis assumptions could be impacted. Hence there is a potential to operate the Auxiliary Building Ventilation System within the Technical Specification limits but violate the design bases of the system if a train were to trip. Therefore,in conjunction with the operable but degraded evaluation, a compensatory action is needed to document the proposed

temporary restrictions to the Technical Specification limits to ensure the 95% filter efficiency is maintained, ne additional restrictions include (1) limiting the dual train alignment maximum flow rate through each of the filters to be less than or equal to 32,000 cfm (The current limit is 33,000 'efm) and (2) limiting the allowable carbon penetration to less than 3% (current limit is 4%) Performance testing completed on the Auxiliary Building Ventilation System filter units shows that all of the units are currently operating with air flow less than 32,000 cfm. Penetration tests are completed monthly on the

                   ' Auxiliary Building Ventilation System and the most recent results of these tests show that the iodine penetration levels for all filter units is less than 3%.

Evaluation: This compensatory action can be implemented without creating unreviewed safety questions. He Auxiliary Building Ventilation System is not an accident initiator and the changes described in this compensatory action will not reduce the ability of the system to effectively filter radioactivity exhausted ham the Auxiliary Building. All other aspects of the Auxiliary Building Ventilation System equipment operation will be unchanged. No changes to the UFSAR will be required. Temporary restrictions will be required to the Technical Specification. Permanent Technical Specification changes will be pursued if necessary. I.

U.S. Nuclear Regulatory Commission - Apdf 1,1999 Paze 115 of 247 f 240 Type: Miscellaneous Items Unit: 0

Title:

Compensatory Action for the Lower Personnel Airlock, Revision 1

Description:

An evaluation was performed tojustify a permanent Compensatory Action to satisfy internal station requirements for Lower Personnel Airlock closure during the "High Decay Heat, L. oops Not Filled" portion of Mode 5. Revision 1 of this Compensatory Action has been modified to have the Qualified Personnel Airlock Door Operator use Enclosure 4.13 of the Personnel Airlock Procedure rather than Enclosure 4.11. This change will allow for better separation of the Compensatory Action Procedure from the Normal Procedure used by operators to return the Personnel Airlock to service. This Compensatory Action is justified based on the large number of workers involved in Outage activities. Maintaining the Personnel Airlock oper, provides improved access and egress for personnel, reduces wear and tear on Personnel Airlock seals, and avoids the accumulation of materials both inside and outside containment. A qualified personnel airlock door operator, using an existing procedure, is provided as a Compensatory Action (vs. a - normally closed or operable Personnel Airlock) to ensure an orderly Containment evacuation and closure of the inner door, considering thermal margin time each day of the outage. This evaluation also documented the safety system operability requirements and recovery actions associated with this activity. Evaluation: This compensatory action does not change the fr.cility as described in the UFSAR, nor does it change the procedures or methods of operation as described in the UFSAR. It is not a test or experiment. The probability of an accident previously evaluated in the UFSAR will not be increased because the personnel airlock is not an accident initiator.  ! The Fuel llandling Accident described in the UFSAR is not affected since the Personnel l Airlock is closed during core alterations. The probability ofloss of residual heat removal j is not affected because the Compensatory Action does not impact that system. The l consequences of an accident previously evaluated in the UFSAR will not be increased because adequate " Defense in Depth" exists. There is no unreviewed safety question associated with this Compensatory Action. No Technical Specification changes are required. No UFSAR Changes are required. i i i

i o f U.S. Nuclear Regulatory Commission

April 1,1999 l Page 116 of 247 L

( l l , 241 Type: Miscellaneousitems Unit: 0 - l l Titic: Compensatory Action for the Upper Personnel Airlock, Revision 1 t

Description:

An evaluation was performed to justify a permanent Compensatory Action to satisfy j internal station requirements for Upper Personnel Airlock closure during the "High Decay Heat, Loops Not Filled" portion of Mode 5. Revision 1 of this Compensatory Action has been modified to have the Qualified Personnel Airlock Door Operator use Enclosure 4.13 of the Personnel Airlock Procedure rather than Enclosure 4.11. This change will allow for better separation of the Compensatory Action Procedure from the Normal Procedure used by operators to return the Personnel Airlock to service. This Compensatory Action is justified based on the large number of workers involved in Outage activities. Maintaining the Personnel Airlock open provides improved access and egress for personnel, reduces wear and tear on Personnel Airlock seals, and avoids the accumulation of materials both inside and outside containment. A qualified personnel airlock door operator, using an existinp procedure, is provided as a Compensatory Action (vs. a normally closed or operable Personnel Airlock) to ensure an orderly Containment evacuation and closure of the inner door, considering thermal margin time cach day of the outage. This evaluation also documented the safety system operability requirements and recovery actions associated with this activity. Evaluation: This compensatory action does not change the facility as described in the UFSAR, nor does it change the procedures or methods of operation as described in the UFSAR, it is not a test or experiment. The probability of an accident previously evaluated in the UFSAR will not be increased because the personnel airlock is not an accident initiator. The Fuel Handling Accident described in the UFSAR is not affected since the Personnel Airlock is closed during core alterations. The probability of loss of residual heat removal is not affected because the Compensatory Action does not impact that system. The consequences of an accident previously evaluated in the UFSAR will no' be increased because adequate " Defense in Depth" exists. There is no unreviewed sar ety question associated with this Compensatory Action. No Technical Specification c ianges are required. No UFSAR Changes are required, l l l

7-. I U.S. Nuclear Regulatory Commission l Apdl 1,1999 Page 117 of 247 273 Type: MiscellaneousItems Unit: 0 l

Title:

Compensatory Actions associated with PIP 0-C98-4098 i l

Description:

Problem Investigation Process (PIP) 0-C98-4098 developed a compensatory action to

                    ' secure the Residual Heat Removal / Containment Spray Sump Pumps in the " Standby" position to simulate the effects of an interlock described in the UFSAR which was never installed. The Safety Evaluation concluded that operation of this subsystem in this manner would create a minor nonconformance with the UFSAR descriptiion of how this subsystem is operated but that this was not considered to be a significant input into the

! licensing of this subsystem by the NRC. l Evaluation: No changes to the UFSAR are required because this is a temporary condition that is being tracked as an " Operable but Degraded" Item. A long term solution to the problem will be

developed and the appropriate UFSAR changes willk made at that time. No Technical

! Specification changes at: required. No Unreviewed Safety Questions were identified. l 4 1 i l l 1 l l s .

F 1 U.S. Nuclear Regulatory Commission April I,1999 Page 118 of 247 308 Type: Miscellaneous items Unit: 0

Title:

Compensatory Actions associated with the Annulus Ventilation System Surveillance I 4.6.1.8.d.4

Description:

Catawba Nuclear Station Technical Specification Surveillance Requirement 4.6.1.8.d.(4) I was identified as being potentially non conservative with respect to the offsite and operator dose analyses. Surveillance Requirement (SR) 4.6.1.8.d.4 requires that each train of the Annulus Ventilation System be tested to ensure that it can draw the pressure in the annulus to -0.5 in w.g. "within one minute after a start signal." The acceptance criterion for the associated surveillance procedures is that each train of the Annulus Ventilation System be tested to show that it can draw the annulus pressure to a pressure of-1.25 in. w.g. or lower within 16 seconds after a fan start signal. This is based on a pressure setpoint of - 1.19 in. w.g. adjusted only for hydrostatic effects, rounded to -1.2 in.w.g and adjusted by -0.05 in.w.g. to account for uncenainties of the test instrumentation. The Annulus Ventilation Systems of both units were considered to be operable but degraded. The associated compensatory measure sets temporary restrictions in place of the criteria of SR 4.6.1.8.d(4). Specifically a value of -1.19 in. w.g. is set in place of-0.5 in. w.g. for the annulus drawdown pressure. In addition, a value of 16 seconds after a fan start signal for the annulus drawdown time is used in place of the values cited in SR 4.6.1.8.d(4). In Revision I to Generic Letter (GL) 91-18, the NRC has set forth guidelines for the , evaluations of compensatory measures for degraded and nonconforming conditions l pursuant to 10 CFR 50.59. In these guidelines, the NRC has agreed that "The intent of the j 10 CFR 50.59 cvaluation is to determine whether the compensatory action itself(not the l degraded condition) impacts other aspects of the facility described in the SAR." The NRC further stated that "In considering whether a compensatory measure may affect other aspects of the facility, a licensee should pay particular attention to ancillary aspects of the compensatory measure that may result from actions taken to directly compensate for the degraded condition." . Pursuant to these guidelines, the " degraded condition" (here, the adequacy of SR 4.8.1.6.d(4)) will not be considered in the safety review and USQ evaluation of the compensatory measure (lowering of the annulus drawdown pressure requirement from -0.5 in, w.g. to -1.19 in. w.g. and a reduction of the one minute drawdown requirement of SR 4.6.1.8.d(4) to 16 seconds after the fan stan signal). The compensatory measure itself will be evaluated for ancillary aspects on the plant.

                     'Ihe compensatory measure substitutes a value of 16 seconds in place of I minute (60 seconds) for the time to dra.w the annulus drawdown time and a value of- 1.19 in.w.g. in place of-0.5 in.w.g. for the annulus drawdown pressure by one train of the Annulus Ventilation System (SR 4.6.l.8.d(4)). It is apparent that the compensatory measure sets a requirement more stringent than the associated surveillance requirement. The compensatory measure does nothing more than this, it does not chrige the actual conduct of the " drawdown" test. The configuration of the Annulus Ventilation System is not affected by the compensatory action. The systems which are interfaced with the Annulus I ..

r U.S. Nuclear Regulatory Conunission Apdf 1,1999

    ~ Pase 119 of 247 Ventilation System (e.g., electric power supplies, the Solid State Protection System, etc.) .

are not affected by the compensatory measure. Therefore, the post accident operation of the Annulus Ventilation System is not affected by the compensatory measure. Neither the configuration nor the operation of any system, structure, or component in the plant is affected by the compensatory measure. In summary, there are no ancillary aspects

                    - associated with the compensatory measure.

Evaluation: There are no unreviewed safety questions associated with this Compensatory Action. The Annulus Ventilation System is not identified as an accident initiator in the SAR. Neither the configuration nor the operation of any system in the plant is affected by the compensatory measure. The compensatory measure has no effect on any equipment associated with poveer generation, conversion, or transmission, shutdown cooling, the processing of radioactive material, fuel handling, or weir gate movement. Therefore, the temporary restrictions placed upon SR 4.6.1.8.d 4 will not increase the probability of occurrence of an accident previously evaluated in the SAR. Since this is an interim measure, no UFSAR changes are necessary and no Technical Specification changes are required. I i i

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April 1,1999 -

Page 120'of 247 ' f 194. - Type: MiscellaneousItems Unit: 0-

                     'ntle: Compensatory Actions required by an

• Operable but Degraded" operability evaluation associated whh Problem Investigation Process Report 0-C97-1579

Description:

De Operability Evaluation associated with Problem Investigation Process Report (PIP) number O-C971579 concluded that the Auxiliary Feedwater System is " Operable But Degraded", with an associated Compensatory Action that the Auxiliary Feedwater Condensate Storage Tank (CACST) be isolated from the Auxiliary Feedwater System supply piping by closing and removing power from valves 1(2)CA6. De Auxiliary Feedwater System pumps are normally aligned to a non-safety, condensate grade' water supply. The safety-related supply of water to the Auxiliary Fredwater System pumps is the Nuclear Service Water System which is normally isolated from the Auxiliary Feedwater System, but automatically aligns upon a loss-of-normal-suction signal; An additional, non-safety water supply is provided to the Turbine Driven Auxiliary Feedwater Pump from the Condenser Circulating Water System for use in certain Design Events (Standby Shutdown Facility Events). The non-safety, condensate Auxiliary Feedwater System supply consists of one CACST (42,500 gallon capacity shared between the two units vented to atmosphere), two Upper Surge Tanks (USTs) (85,000 gallon combined capacity - unit specific - normally under vacuum), and one Condenser Hotwell (170,000 gallons available at normal operating level - unit specific - normally under vacuum). Of these three supplies, the USTs are at the highest elevation, followed by the CACST, and the Hotwell. However, if condenser vacuum is not broken before or during CA System operation, the relative elevation of the USTs is lower than the CACST.

   ,                        - PIP O-C97 1579 was written on 5/8/97 due to concerns with vortex formation in the CACST and USTs that could lead to the introduction of air into the suction piping of the Auxiliary Feedwater System pumps, potentially disabling the pumps. An Operability Evaluation was performed, with support from the Auxiliary Feedwater System pump vendor, which concluded that vortex formation is not an operability concern.

During the process of evaluating the vortex concern, a separate mechanism was identified

                            - by which air could potentially enter the Auxiliary Feedwater System suction piping. This mechanism involved the depletion of the CACST and the failure of 1(2)CA6 (CACST to Auxiliary Feedwater System Pump Isolation Valve) to automatically close on a low CACST level. In this situation, the USTs would supply the Auxiliary Feedwater System pumps. However, if condenser vacuum is not broken, the relative elevation head of the UST's with respect to thejunction of the Auxiliary Feedwater System supply piping from the CACST and USTs is not sufficient to maintain the pressure at this junction above atmospheric pressure over the full range of possible Auxiliary Feedwater System flow rates. Also, due to the elevation difference between this junction and the pressure switches that activate the automatic swapover to the assured Auxiliary Feedwater System suction sources of the Nuclear Service Water System and the setpoint of these pressure switches, the swapover is not assured if the CACST/USTjunction pressure is less than atmospheric over the full range of possible Auxiliary Feedwater System flows. This
                 \

p t-L 4 U.S. Nuclear Regulatory Commission - l ' Apdf 1,1999 ' Paae 121 of 247 f could lead to the introduction of air into the Auxiliary Feedwater System suction piping and possibly into the Auxiliary Feedwater System pumps from the depleted and unisolated CACST, potentially disabling the pumps. l. De analysis of the quantity of air that could be introduced into the Auxiliary Feedwater System piping and the resulting effects on Auxiliary Feedwater System pump operation required the use of resources outside of Duke Power Company and could not be completed within the normal time frame for completing the Operability Evaluations. Therfore, the Auxiliary Feedwater System on each unit was declared operable, with the recommendation that the CACST be isolated from the Auxiliary Feedwater System by closing and removing power from valves 1(2)CA-6 as a conservative treasure to eliminate this potential failure mechanism. l Various options of configuration and operational control were evaluated. De criteria for ! a Temporary Station Modification was not met and this option was not chosen. A Minor l Modification was not pursued due to the anticipated short time frame that these changes I would have to be in place. It was decided that the configuration control of these valves would be maintained by the Removal and Restoration (R&R) process (R&R 17-643 on ~~ , Unit I and R&R 27-1139 on Unit 2). l i' A vendor was then contracted to verify the CNS analysis of possible air introduction upon CACST depletion as described above, and also to analyze the entire non-safety Auxiliary Feedwater System suction source design to determine if any other potential problems exist. The expected duration of the vendor analysis work was originally anticipated to be approximately four months; however it was subsequently decided by Engineering to have

                   . the vendor perform a unit specific analysis and provide nuclear safety related calculations l                   - to support the analyses. This extended the duration of the work such that the completion date was extended until March 1998. A Corrective Action was assigned to Engineering to l                      track the vendor study and initiate any permanent changes to the station that resulted from l                      this analysis.

After CNS Engineering received preliminary results from the vendor analysis verifying

. that the possibility of air introduction upon CACST depletion does exist, a review of Operability Evaluation associated with PIP 0-C97-1579 was performed. This review has l

concluded that the Auxiliary Feedwater System is

• Operable But Degraded", and that the action ofisolating the CACST from the Auxiliary Feedwater System supply piping by closing and removing power from valves 1(2)CA6 is a Compensatory Action.

i . Evaluation: There are no unreviewed safety questions associated with this compensatory action. The safety related function of the Auxiliary Feedwater System is to provide a nuclear safety related source of emergency feedwater to the steam generators to maintain secondary side 1: level at times when the normal feedwater system is not available. In this function, the

                    - Auxiliary Feedwater System is relied upon to remove primary coolant stored energy and residual core energy, and to prevent overpressurization of the Reactor Coolant System and the resultant reactor coolant expansion that could result in fuel damage.

The Auxiliary Feedwater System is designed to start automatically in the event of loss of ]' offsite electrical power, trip of both main feedwater pumps, safety injection signal, low-low steam generator water level or AMSAC signal; any of which may result in, coincide b l

a

                     /

U.S. Nuclear Regulatory Cosamission

           -- April 1,1999 Page 122of 247

• t

                           ! with, or be caused by a reactor trip. He system is designed to function to mitigate the consequences of all Design Basis Even:s and Design Events listed in Section " System Design Basis" of the Auxiliary Feedwater System Design Basis Specification.

The Auxiliary Feedwater System will provide the required minimum flow to the steam i generators regardless of any single failure. Sufficient feedwater can be provided at required temperature and pressure for the required Operator delay period following an

                                                                                                   ~

j auto-start. l J Since the Auxiliary Feedwater System is the only source of makeup water to the steam generators for decay heat removal when the main feedwater system becomes unavailable, it has been designed with special considerations. The use of redundancy, diversity, and separation has been incorporated into the design of the Auxiliary Feedwater System to ensure its capability to function. Redundancy is provided by using three Auxiliary Feedwater System pumps powered from separate and diverse power sources. Two motor driven pumps are powered from two separate trains of emergency onsi:e electrical power, each supplying feedwater normally to two steam generator's. One turbine driven pump, j supplying feedwater to all four steam generators, is driven from steam contained in either j of the two steam generators. Diversity is provided by using several water sources, two j types of pump drivers, and adequate valving for source selection, isolation, and cross- i connection. Separation is provided with separated power, instrumentation, and control subsystems with appropriate measures precluding interaction between subsystems.  ! The isolation of the CACST as a Auxiliary Feedwater System suction source eliminates . the possibility of air introduction into the Auxiliary Feedwater System suction piping upon CACST depletion previously described. As described previously, the CACST is a non-safety related source of water for the Auxiliary Feedwater System. The volume of water in the CACST is not credited in the evaluation of any Design Basis Event. De Nuclear Service Water System is the safety related source of water for the Auxiliary Feedwater System, and though normally isolated from the Auxiliary Feedwater System aligns automatically upon detection oflow Auxiliary Feedwater System suction pressure. For Design Events in which the Nuclear Service Water System is not available to supply water to the Auxiliary Feedwater System (less of All AC Power, Security, Fire), the Condenser Circulating Water System contains a sufficient volume of water an the enbedded header in the Turbine Building to maintain the unit at Hot Standby for the required maximum 72 hour duration. The Condenser Circulating Water System supply is normally isolated, but aligns automatically on detection of low Auxiliary Feedwater System suction pressure. It should be noted that as the CACST is shared between the two units, only one half ofits l 42,500 gallon capacity (or 21,250 gallons) can be assumed to be available to each unit l' . during a dual unit event (e.g. Loss of Offsite Power). By comparison, the unit specific l USTs have a combined 85,000 gallon capacity and the unit specific Condenser Hotwell contains 170,000 gallons at normal operating level. Herefore, the great majority of the non-safety, condensate grade water supply to the Auxiliary Feedwater System will remain aligned to the system. It should also be noted that the Technical Specification associated with the required L

7

      . U.S. Nuclear Regulatory Coenmission
      ' April 1,1999 Page 123 of 247 225,000 gallon volume of condensate grade water in the CAC5h', USTS, and Condenser Hotwell (T.S. 3.7.1.5, Condensate Storage System) exists oray on Unit 2. No such l                       Technical Specification exists on Unit 1. This Technical Srecification does not specify L                       any required volume in the specific tanks, only a combined volume. It is possible to meet the Technical Specification requirement with the USTs and Condenser Hotwell only (i.e.

_ CACSTempty). This activity does not add or delete any automatic or manual safety related feature of the Auxiliary Feedwater System, nor does it convert an automatic safety related feature to manual or vice versa. He activity does not introduce an unwanted or previously unreviewed system interaction, but instead eliminates such an interaction. This activity does not alter the QA condition, or seismic or environmemal qualification of any component in the Auxiliary Feedwater System as the CACST is a non-safety, non-seismic , tank. No adverse effects on the safety related function of the Auxiliary Feedwater System or any interfacing systems are created by this activity. No new failure modes are created

by this activity, L-l The Auxiliary Feedwater System is designed to provide a nuclear safety related source of I feedwater to the S/Gs to maintain secondary side level in the event of a loss of normal feedwater. Dere are no UFSAR Chapter 15 events which assume Auxiliary Feedwater

      ,                System contributes as an accident initiator. No automatic function of any safety related SSC is affected. As there is no adverse effect on the performance of the Auxiliary Feedwater System or any other plant system, this activity will not increase the probability of accidents evaluated in the SAR. No equipment important to safety evaluated in the SAR is adversely affected. While the volume of water contained in the CACST will no longer be normally aligned to the Auxiliary Feedwater System, this supply is not safety related, in a seismic event, the CACST, along with the other sources of non-safety, condensate supply to the Auxiliary Feedwater System, are assumed to be unavailable.

De Auxiliary Feedwater System pumps remain protected from the consequences of a loss of normal, non-safety suction sources by the automatic swap to the assured source of the ' Nuclear Service Water System (if the Auxiliary Feedwater System is in auto-start mode of operation) and by the automatic trip (if the system is in the manual mode of operation). Derefore, the probhbility of a malfunction of equipnwnt important to safety evaluated in the SAR is not increased No Technical Specification changes are required. No UFSAR changes are required. ll 1

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U.S. Nuclear Regulatory Commission April 1,1999 ] Pane 124 of 247 : 195 . Type: Miscellaneous items Unit: 0 g

                          'litle: th,-tory Actions required by an " Operable but Degraded" operability evaluation associated with Problem Investigation Process Report 0-C98-1726

Description:

Problem Investigation Process Report 0-C98-1726 identified a need to evaluate the potential single failure of the non-nuclear safety related valve CM127 for both Units. Valve CM127 is used during normal full power operation to provide mini-flow protection for the non-nuclear safety related CA-" Booster Pumps. During a power reduction this valve omed as a result of an incorrect setpoint at approximately 50% power. As a

                                  . result of the valve opening, hot water was admitted to the Upper Surge Tank . The Upper         j Surge Tank is currently the preferred non-nuclear safety related source for the Auxiliary   ,

Feedwater System. The elevated temperatures resulted in declaring all three Auxiliary Feedwater Pumps inoperable. This compensatory action is a temporary measure to prevent recurrence of the event. The evaluation addresses the effect of this flow path on j '. the Condensate Booster Pumps and the Secondary Systems. The evaluation states that during normal operation minimum flow protection can be provided by the Main Feedwater Pump Recirculation valves and that this change along with a change in power level for placing the "C" Heater Drain Pumps in service does not materially change

                                                                                                                                  ]

operation of the system from that described in the UFSAR. The Compensatory Actiori to maintain the Auxiliary Feedwater System in an " Operable but Degraded" condition will

                                                                                                                                  )
                                   . be imposed any time the Unit is in Modes 1,2, or 3 and has the following requirements:

1) Isolate the Condensate Makeup recirculation flowpath to the Upper Surge Tank by closing valves CM128 and CM125 in Modes 1,2,and 3. 2) The C Heater Drain Pumps are not to be in service below 70% power. The purpose of the compensatory action is to ensure that a single failure of the non-nuclear safety related component CM127 does not

         ~

input high flow and high energy water to the Upper Surge Tank. The last item of the Compensatory Action minimizes the potential for a low-flow Condensate Booster Pump j trip if the reactor trips below the P-9 setpoint. -) Evaluation: his compensatory action may affect two accidents described in the UFSAR. %ese are (1) a feedwater system pipe break and (2) a loss of feedwater. An evaluation concluded that the plant would operate within the limitations of the piping design and the ANSIB31.1 Code. Herefore the Feedwater System Pipe Break accident is not affected. The evaluation stated that the operation with the CM127 flowpath isolated may result in a loss of Condensate Booster Pumps during a loss of stator cooling runback. A Condensate Booster Pump trip under these conditions may result in a trip of the Main Fredwater

                                   - Pumps. Therefore, a stator cooling runback could result in a loss of feedwater. The evaluation stated that the only loss of stator cooling runback which has been experienced at Catawba resulted in a reactor trip due to loss of both Main reedwater Pumps. This parallels the SER acknowledgenent that Stator Cooling malfunctions may result in a Turbine Trip. It is acknowledged that Overpower Delta Temperature (OPDT) and Overtemperature Delta Temperature (OTDT) runbacks are also capable of continuously running the turbine back to power levels below 40% However, there is no credible              j l-                                     scenario which would cause a continuous OPDT or OTDT runback for any significant              1

j. power range. All of these runbacks that have occurred at Catawba have been of very l l- short duration affecting power levels by only a few percent. This Compensatory Action l will not affect the initiation of any runbacks for any reason therefore the frequency of runbacks is not affected. Therefore, the probability of the loss of feedwater accident or l

l l

[- i. I U.S. Nuclear Regulatory Comunission Apdl1,1999 Page 125 of 247 l the feedwater system pipe break accident evaluated in the UFSAR is not increased. No Technical Specification changes are required. No UFSAR changes are required. s 1 l l l 1 l l [ t

      . .e T

U.S. Nuclear Regulatory Conunission April 1,1999 Page 126 of 247 229 Type: MiscellaneousItems Unit: 0 Titiet Compensatry Action Guidlines - Plant Access Doors, Revision 15

Description:

Revision 15 of this Compensator) Action was initiated to incorporate several changes associated with Auxiliary Buildin,t IMors. These changes included: Deleting the requirement for a specia'. compensatory action if doors AX214 A or AX214B were opened and corresponding Unit was in Modes 1-4.The requirement to have a special compensatory action was originally imposed assuming the ECCS pressure boundary would be moved back to these doors. This has not been accessary so these doors did not become part of the ECCS pressure boundary and no restrictions are necessary. Therefore, note 7 was deleted from these doors. Doors AX260B, AX 260G and AX260ll were added. These doors were not part of die compercatory action in the past but are needed to support painting work in the area. Security requirements on doors AX373B and AX393C were deleted. These doors are no longer part of the security boundary a security watch is no longer required when they are open. A Compensatory Action was revised to reflect changes associated with the Security Computer Upgrade modification. This modification made several changes to security doors such as deleting the security requirements on doors AX353 and AX394, and revising the Controlled Access Door numbers on AX600 and AX629. Miscellaneous editorial / typographical corrections such as changing FSAR to UFSAR and fixing typing and grammatical errors. This revision only affects some of the information associated with door numbers AX214A, AX214B, AX260B, AX2600, AX260ll, AX353, AX393B, AX393C, AX394, AX600 and AX629. Evaluation: Compensatory Action Guidelines " Plant Access Doors Revision 15" serves as the overall review of doors within the Auxiliary Building. The guidelines identify the design functions each of the doors. The guidelines allow activities which could prevent the door from closing normally to occur if they do not challenge a safety-related function or if the feature they are designed to protect can be removed from service and declared inoperable. The guidelines do allow doors to be restricted from closing with certain ) Compensatory Actions. These Compensatory Actions are limited to Fire Boundary Doors which require a Fire Watch in accordance with the Selected Licensee Commitments, Tornado Doors which must be capable of being closed within one hour, and Security Doors which require Security Access Control to be established. With the requirements of the Compensatory Action Guidelines Plant Access Doors satisfied, no Unreviewed Safety Questions are created and the margin to any Technical Specification is not challenged or reduced. No Technical Specification changes are required. No UFSAR changes are required. E

m l

    ' U.S. Nuclear Rip ^g Comunission April 1,1999 Pase 127 of 247 320     Type: MiscellaneousItems                                              Unit: 0 l               'Hele: Core Operating Limits Report (COLR) changes associated with conversion to the

! 1 Improved Technical Specifications Descdption: The Catawba Unit i Fuel Cycle (CICII) and Catawba Unit 2 Fuel Cyclel0 (C2C10) l Core Operating Limits Report (COLR) changes that are addressed by this evaluation are the resuk of the implementation ofImproved Technical Specifications at Catawba. The majority of the changes involve wording and nomenclature changes that were made to improve clarity and to maintain consistency between Improved Technical Specification (ITS) and the COLR. Because of technical specification numbering and ordering changes in the ITS, similar changes were made in the COLR. The changes made to the CICI1 and C2C10 COLRs do not resuh in a change to the intent, interpretation or understanding

. of the technical content contained in these reports. All limits present in current COLRs L and technical specifications were preserved in the creation of the CIC11 and C2C10 COLRs. The changes made also do not involve any changes to the operation, design basis or function of any structure system or component.

The implementation ofITS for CIC11 and C2C10 also required that several specifications, or specification requirements be relocated from current technical specifications (CTS) to the COLR. The items relocated were: 1, Shutdown margin requirement of Modes 1 5.

2. Mode 6 reactivity requirement (per ITS SER) -

3. The requirement for the measured value of FQ(X,Y,Z) to be increased by 3% to account for manufacturing tolerances and 5% for measurement uncertainties. (CTS 3/4.2.2 -ITS 3.2.1)

The relocation of Current Technical Specification (CTS) requirements to the ITS COLR do not impact normal operation of the plant or affect any probability or consequences of accidents evaluated in the SAR since the requirements relocated to the ITS COLR are unchanged from those specified in CTS. Therefore, the consequences of accidents in which these parameters are important are not impacted and the margin to safety remains unchanged. ' The deletion of the minimum cold leg accumulator boron concentration requirement to ensure post. LOCA suberiticality was removed from the ITS COLR for this specification (ITS 3.5.1 - CTS 3/4.5.1.1) because ITS does not require any comparisons to be ! performed against this limit. This change is considered administrative since it was NRC 1- approved in the ITS SER. ! The implementation ofITS does require that an additional moderator temperature coefficient (MTC) surveillance be performed when the reactor core reaches 60 ppmb if the 300 ppmb surveillance limit is violated.- This new surveillance is required by ITS and its limit is included in the C)Cll and C2C10 COLRs. The addition of this limit provides confirmation that input assumptions to SAR accidents which are sensitive to this l I

U.S. Nuclear Regulatory Conunission Apdf 1,1999 - Page 128 of 247 parameter remain bounding. By confirming this input assumption, the consequences of accidents evaluated in the SAR will continue to be bounding. The addition of this new surveillance requirement also does not impact the operation of the plant or affect any setpoints used to mhigate the consequences of any accident. Evaluation: The COLR changes performed to implenwnt Improved Technical Specifications do not change the intent, interpretation or understanding of the technical content of die information being provided in the COLR. 'Ihe changes performed do not result in the modification of equipment important to safety, or impact the function, design bases performance or operation of safety related equipment or components. All operational and safety limits specified in Current Technical Specifications and the CICI1 and C2C10 Core Operating Limits Reports are preserved in the conversion of tiene reports to ITS COLRs. Therefore, the margin to safety as defined by the bases to Technical Specifications is not reduced. The COLR changes do not affect the performance of equipment important to safety, or setpoints or limits that would impact the probability of an initiating event. Therefore, the probability or consequences of a malfunction of equipment important to safety is also not increased. 'Ihere are no unreviewed safety questions that exist resulting from the conversion of CICl I and C2C10 COLRs to ITS COLRS. No Technical Specification changes are required other than those already approved as a part of the ITS related License Amendment. No UFSAR changes are required. i I j

p !~' ' U.S. N'aclear Regulatory Conunission Apdf 1,1999 l Pase 129 of 247 P 210 Type: MiscellaneousItems - Unit: I

               'Iltle: Erection of one set of scaffolding in Unit 1 Upper Containment during operation

Description:

One set of scaffolding (10 feet high) will be installed in Unit 1 Upper Containment. The scaffolding will be used in support of Limitorque valve operator maintenance. The scaffolding will be 10 feet high with a base dimensions of 5 feet by 5 feet. Three aluminum walkboards (19" wide by 7'long) will be used as a work platform with the scaffolding assembly. The scaffold will be erected on the Hydrogen Recombiners and Skimmer Fans platform at elevation 645'+9". This evaluation addresses the following - concerns:

1) Seismic Interaction: The scaffolding will be installed and secured in accordance with the Duke Power Scaffold Manual, which will ensure seismic concerns are controlled to prevent interaction with other components / equipment in containment.

2) Post Accident Recirculation: There are six eight inch diameter penetrations in the refueling canal wall which serve as a recirculation path to the containment sump. The scaffolding components will be secured such that the scaffolding can not become loose -

and block these recirculation paths in case of a seismic event.

3) Aluminum in Containment: Use of aluminum walkboards in containment has been evaluated as being acceptable (reference calculation CNC-1223.02-00@l2).

4) Combustible Material in Containment: No combustible material will be used to erect the scaffold.

i Evaluation: There are no unreviewed safety questions associated with this activity. The scaffold will be constructed within guidelines that prevent the item from causing or affecting mitigation of accidents evaluated in the UFSAR. No Technical Specification changes are required. No UFSAR changes are required. I1 l l , i t  ; l i l l l I w.

U.S. Nuclear Regulatory Con ==la la= AprE 1,1999 Pase 130 of 247 201' . Type: Miscellaneous items Unit: 0

               . '11tle: Nuclear Design Methodology using CASMO-3/SIMULA*lE-3P, DPC-NE-1004A .

Description:

This evaluation was performed to determine if an unreviewed safety question exists when the current methodology is applied to a fuel design that differs from those previously benchmarked and documented in Topical Report DPC-NE-1004A. For Duke Power

                       . Westinghouse design nuclear power plants " Duke Power Company, Nuclear Design Methodology Using CASMO-3/SIMULA'IE-3P,DPC-NE-1004A, November 1992," is considered applicable for Westinghouse OFA, Standard, and FCF Mark B&W (similar to Westinghouse STD) fuels. De November 1992 SER to this Topical stipulated that "the '

application of CASMO-3 and SIMULA'IE-3P to fuel designs that differ significantly from those included in the topical data base should be supported by additional code validation to ensure that the DPC NE-1004 methodology and uncertainties apply." ne new fuel type is Westinghouse Performance Plus fuel with integral burnable absorber (IFBA). His is a thin coating of ZrB2 applied directly to fuel pellets of selected fuel rods. This had not been previously modeled at Duke Power. Supporting calculations documented the performance of a benchmark analysis and a calculation of the uncertainty factors for Performanxec Plus Fuel with IFBA. From these calculations, it was determined that IFB A fuel can be acceptably modeled using the methodology of DPC-NE-1004A. Evaluation: There are no unreviewed safety questions associated with this activity. The change is analytical in nature and does not affect system performance, operation or equipment. 8 This change cannot initiate an accident. De current UFSAR accident evaluations which may rely on any part of this methodology will remain unaffected when modeling the new type of fuel because the uncertainty factors previously used were shown to remain valid . and conservative. No Technical Specification changes are required. No UFSAR changes ) are required. 184 Type: Miscellaneousitems Unit: 1

Title:

On-line leak repair corrective maintenance MP/0/A/7650/063 for a pipe cap after Valve ICF-122

Description:

This on-line leak repair will fix a leaking pipe cap near drain / vent valve ICF-122. (The CF System is the Condensate Feedwater System). The pipe cap has minor external

                                                                  ~

leakage. His repair will not adversely affect the operation of the system. The sealant will not enter the system. The repair will be accomplished using an approved maintenance l procedure which was developed using an industry guideline (EPRI NP-6523d). Evaluation: There are no unreviewed safety questions associated with this leak repair. This is

l. ,

considered to be a normal maintenance procedure, ne worst case failure scenario of the leak repair process was evaluated and it was determined that the probability or consequence of an accident was not increased by this operation. No Technical l Specification changes are required. No UFSAR changes are required. l i i i s

m-f i L U.S. Nuclear Regulatory Ceaunission April I,1999 l Page 131 of 247 : 185 Type: Miscellaneousitems Unit: 1 11tle: On-line leak repair corrective maintenance MP/0/An650A)63 for a pipe cap after Valve ICF-132

Description:

His on-line leak repair will fix a leaking pipe cap near drain / vent valve ICF-132. (The CF System is the Condensne Feedwater System). The pipe cap has minor external leakage. His repair will not adversely affect the operation of the system. The sealant will not enter the system. The repair will be accomplished using an approved maintenance procedure which was developed using an industry guideline (EPRI NP-6523d). Evaluation: There are no unreviewed safety questions associated with this leak repair. This is considered to be a normal maintenance procedure. The worst case failure scenario of the leak repair process was evaluated and it was determined that the probability or consequences of an accident was not increased by this operation. No Technical Specification changes are required. No UFSAR changes are required. 1 197 Typer-Miscellaneous Items Unit: 2

Title:

Procedure TN/2/A/1373/00/01E, Implementation Procedure for Nuclear Station Modification CN 21373/00, Revision 0, Work Unit i

Description:

Procedure TN/2/A/1373/004)1E installs the modifications to the Main Steam Isolation Valve control circuitry designed by Nuclear Station Modification CN-21373. This work will be performed during Modes 5,6 or No Mode when all equipment affected by this i procedure is not required to be operable. All equipment affected is isolated and de- j energized while the work is being performed. All changes are verified to function j properly through post modification testing prior to return to service. Evaluation: There are no unreviewed safety questions associated with this implementing procedure for l Nuclear Station Modification CN-21373. The only applicable accident is the inadvertent , closure of a Main Steam Isolation valve. The modification will be installed when plant conditions do not require Main Steam Isolation valves to be operabic. The modification will be fully tested before the Main Steam Isolation valves are returned to service. No Technical Specification changes are required . No UFSAR changes are required. ] i

y l' I I I . U.S. Nucienr Regulatory Comunission April 1,1999 . L Pase 132 of 247  : l j 189 ~ Type: Miscellaneousitems - Unit: 0 l

Title:

Revision 0-A of DPC-NF-2010A " Nuclear Physics Mediodology for Reload Design" (June 1985)

Description:

As part of Duke Power's effort to maintain updated and accurate Topical Reports, a review of DPC-NF-2010A was conducted to identify any changes that have been made  ! since its original approval in June 1985. Two changes were identified that had not previously been addressed per 10CFR50.59 or approved by the NRC. The changes evaluated in this 10CFR50.59 relate to the calculation of core shutdown margin and to the calculation of maximum ejected control rod worth methodologies described in DPC-NE- - j 1003A. The changes being evaluated are simply refinements of the original  ; methodologies. 1 ,  ; In the case of the shutdown margin calculation, generic penahies were added to die l calculation of core power defect and to the rod insertion allowance to account for die j l possibility of operation at off-nominal conditions within permitted operating conditions. ' The Topical reserves the possibility of specifically calculating shutdown margin under , unusual conditions, such as when extended operation at off-nominal conditions occurs l (e.g. Axial Offset Anomaly). f , For the ejected rod worth analysis, the Topical originally stated that " single rods in  ! l Control Banks D, C, and B are removed in subsequent cases and the worth of the ejected l rod is calculated . "; however the Bank B rods and the Group i rods of Bank C (eighth core locations B-10 & B-08 respectively) have proven to be of such low worth relative to l the other locations (H-08, F-10, and D-12) that it is not necessary to analyze these '  ! locations. Also, Bank B is at most very lightly inserted, even at zero power conditions,  ; j which funher reduces the available worth at B-10. l The SER issued for DPC-NF-2010A does not address, nor apply restrictions on the iteration technique used to calculate the predicted critical heights. i Evaluation: These changes were made to update the discussion to reflect the current design methodology for Catawba Nuclear Station. ' Die changes did not produce any unresolved safety questions or changes to Technical Specifications. Since these changes do not i involve the modification of equipment imponant to safety, or impact the function, design bases performance or operation of safety related equipment and components, the margin to safety as defined by the bases to Technical Specifications is not reduced. For these reasons, the probability or consequences of a malfunction of equipment important to safety is also not increased. The SER issued for DPC-NF-2010A mentioned, but did not i apply restrictions on the methods of calculation for either of these changes. i

U \ l l l U.S. Nuclear Regulatory Cosamle= Ion  ! !. April 1,1999 l Pane 133 of 247 l I f

255 Type

Miscellaneous items Unit: 0 l

l litle: Revision to Selected Licensee Commitments Section 16.9.3 and 16.9.6

Description:

This Selected Licensee Commitment revision concerns changes to Section 16.9.5 that are essentially editorial with the exception of the removal of heat detectors in zones 72 and 80 (Control Room) sind zones 177-180 (Lower Containment Filter Beds) from Table 16.9-3. The following is a summary of the changes: 1. The Remedial Actions section will add a j reference to the Lower Containment Technical Specfication for temperature monitoring in the lower containment. 2. *Ihe frequency for the functional test (formerly refered to as a Trip Actuation Device Operational Test (TADOT) will be changed for the non-accessible heat detectors from "during each refueling outage" to "at least once per eighteen months".'

3. Six heat detectors will be deleted for zones 72 and 80 (Table 16.9.3) 4. Two heat

detectors will be deleted for zones 177,178,179, and 180 (Table 16.9-3). 5. Testing l requirements will be relocated for the CO2 System detectors (from Section 16.9.6 to 16.9.3.). Testing requirements for spot type heat detectors will be deleted from Section 16.9.6. I Evaluation

All of the heat detectors being removed from the scope of the Selected Licensee Commitments are coupled with an existing smoke detector. The presence of the adjacent smoke detectors, along with other means of early warning (continuous occupancy of the I Control Room and ventilation system thermal detectors) result in adequate fire detection j capabilities for these zones. The relocation of CO2 system detectors to SLC 16.9.3 (CO2 Systems) couples diese detectors with their associated system requirements and establishes an effective testing methodology. There are no unreviewed safety questions  !

associated with this change to the Selected Licensee Commitments. No Technical

Specification changes are required. No UFSAR changes are required l

l l i l L

U.S. Nuclear Regulatory Ca==8==Aa= April 1,1999 j Page 134 of 247 - '198- Type: Miscellaneousitems. Unit: 0

Title:

Revisions to Selected Licensee Commitments 16.9.2 and 16.9.5

Description:

This revision makes editorial changes to Selected Licensee Commitments 16.9.2 and

               . 16.9.5. De changes to Selected Licensee Commitments 16.9.2 are: (1) the title of the section will be changed from " Spray and/or Sprinkler Systems" to " Sprinkler Systems".

Catawba Nuclear Station does not have any " spray" systems within the scope of the Selected Licensee Commitments committed areas. (2) the reference to " automatic" valves in the Testing Requirernents will be removed. Catawba Nuclear Station does not have any

                 " automatic" valves in the Selected Licensee Commitment committed flow paths. (3) De system functional test will be accurately described as an " inspectors test connection flow test". His is more accurate than the current wording of " simulated automatic actuation".

The changes to Selected Licensee Commitments 16.9.5. is to remove the test requirement for fire door supervision systems. Catawba does not have any electrically supervised Selected Licensee Commitment committed doors. This change makes the Selected Licensee Commitment accurately reflect Catawba's configuration. Evaluation: These changes have no impact on any accident analyzed in the UFSAR. There is no unreviewed safety question associated with this change to the Selected Licensee Commitments. No Technical Specification changes are required. Changes to the Selected Licensee Commitment Manual (UFSAR Chapter 16) are required. A.

U.S. Nuclear Regulatory Commission Apdl 1,1999 Page 135 of 247 F 321 Type: Miscellaneous Items Unit: 0

Title:

Software Upgrade to COMET 02 Software Descripuen: This change involves an upgrade to software used for offline core power distribution monitoring. He current software, COMETVI,is being replaced by COMET 02. Evaluation: COMETV2 is an improved version of COMEIDI. The fundamental methodologies employed are unchanged. The new software incorporates vadous enhancements piedominantly editorial in nature, with one exception. The capability was added to apply a burnup-dependent penalty factor to certain Improved Technical Specification (ITS) power distribution surveillance calculations (see ITS 3.2.1 and 3.2.2, Reference 5). Currently the code uses a single penalty for an entire cycle. Recently submitted licensing amendments address the use and methodology of the burnup<lependent penalty. COMET 02 was certified per Duke Power's directive for software certification and verified to yield the same results as COMETUI, excepting the new modifications. The modifications were verified and, as applicable, are in compliance with ITS or submitted revisions. COMET 02 is considered equivalent to COMET 01. This change involves no taaterial changes to the plant. The COMET software and resident workstation are not part of any system, structure, or component (SSC) important to safety and do not directly affect any SSCs. The three systems indirectly associated, the Movable Incore Detector System, the Excore Power Range Detectors, and the Reactor Protection System, are all unaffected by this change. The only safety significant function performed by, or involved with, this system is generation of data to evaluate ITS 3.2.1, 3.2.2, and 3.2.4, and to periodically calibrate the power range AFD indications, as required per SR 3.3.1.6 (Table 3.3.1-1). As indicated, the new software is functionally equivalent to the replaced software and yields the same analytical results. Therefore assurance of the fuel integrity limits associated with the referenced ITSs and with the OPDT and OTDT reactor trips (AFD parameter input) are not compromised. This change does not impact any plant parameters, safety limits or setpoints that potentially affect the fission product barriers. His change does not involve an unreviewed safety question. No changes to the Technical Specifications are required. UFSAR Sections 4.3.2.2.7 and 4.3.6 will require changes to update COMET references. I 1 l l l l 1 l

Fr l I U.S. Nuclear Regulatory Commission April 1,1999 Page 136 of 247 183 Type: Miscellaneous Items Unit: 2

Title:

Temporary Station Modification Work Order 98025036-01 which disables instrument 2WLPS I 10 as an input to annunciator 2AD18.02-12

Description:

Temporary Station Modification (TSM) Work Order 98025036-01 disables instrument 2WLP5110 as an input to annunciator 2AD18.02-12. his TSM disconnects wiring and places ajumper in the current alarm circuitry for the lower containment resistance temperature detector (RTD) 2WRD5110. This will disable the 2WLP5110 instrument loop input to annunciator 2AD18.02-12 and clear the annunciator light. He temporary modification will remain in place until the Unit 2 End of Cycle 9 refueling outage. During that outage Minor Modification CE-61344 will remove the circuitry associated with the containment local temperature annunciator alarms. This instrument is causing the annunciator to be in alarm because of a local hot spot believed to be caused by degradation of insulation on Steam Generator 2B No safety related electrical equipment is located in the immediate area of this instrument. Evaluation: There are no unreviewed safety questions associated with this modification. The probability or consequences of an accident is not increased by this temporary modification. The modification will not adversely affect the Operator Aid Computer lower contain nent local temperature indications or the average temperature alarms. The , local control room temperature gauges will not be adversely affected by the modification. Lower Coritainment average temperature surveillance 4.6.1.5.2 will not be affected. Instrument loop 2WRD5110 will remain available as an input for this surveillance. No Technical Specification changes are required. No UFSAR changes are required. 187 Type: Miscellaneous Items Unit: 1

Title:

Temporary Station Modification Work Order 98031580, install a gag on valve IRC31 l

Description:

Temporary Station Modification Work Order 98031580, installs a gag on valve IRC31. 1 This Temporary Station Modificatic, (TSM) places a gag on valve 1 RC31, Cooling Tower I A makeup water flow control valve while repairs are made to fix a problem with the valve operator. During the repair the valve will maintained in the full open position. Evaluation: Dere is no unreviewed safety question associated with this activity. Valve IRC31 is not an accident initiator and is no involved involved in mitigating any accident. He function of the valve will be done by mainipulating a manual valve if necessary. No Technical  ; Specification changes are required. No UFSAR changes are required. I i I l I

U.S. Nuclear Regulatory Commission April 1,1999 Page 137 of 247 f 206 Type: Miscellaneous Items Unit: 0

Title:

Temporary Station Modification Work Order 98057995-02 and 98057995-03, Oag valve 2RC31 in the oper position

Description:

Temporary Station Modification Work Orders 98057995-02 and 98057995-03, install a gag on valve 2RC31 in the open position. Work Order 98057995 will repair a problem on the operator for valve 2RC31 Cooling Tower 2A makeup water flow control valve. While the operator is being repaired, the make-up flow path through 2RC31 will need to be in service. This Temporary Station Modification will instaill a gag on valve 2RC31 in its fully open position to allow the 2A Cooling Tower to remain in service. Chemistry personnel will control flow using the manual butterfly valve 2RC95. Evaluation: There is no unreviewed safety question associated with this activilty. Valve 2RC31 is not an accident initiator and is not involved in mitigating any accident. The function of the valve will be done by mainipulating a manual valve if necessary. No Technical Specification changes are required. No UFSAR changes are requked. 202 Type: Miscellaneous items Unit: 0

Title:

Temporary Station Modification Work Order 98059997, Repair of A Train YC Chiller Oil Pressure Sensing Line

Description:

Temporary Station Modification Work Order 98059997, involved repair of the A Train YC Chiller Oil Pressure Sensing Line. The modification installed a gasket secured by a hose clamp on the discharge oil pressure sensing line on Control Room Ventilation Chiller A. This line has a pinhole leak in it which is caused by a metal band wearing a hole through the copper tubing. This line serves the oil regulator, an oil pressure switch, and the oil pressure gauge on the control panel. The pressure switch provides a permissive for the chiller to operate , therefore significant leakage on this line could potentially render the chiller incapable of performing its design function. This temporary modification will repair the line and ensure the operability of the Chiller. The repair will be made by installing a gasket around the hole in the tubing and securing it with a hose clamp. Evaluation: This modification will temporarily repair an oil line. The repair method has been evaluated and found to be acceptable from an Engineering perspective. The cause of the leak has been identified and removed. The repair method has been determined adequate for the stresses involved. The repair is not significant from a seismic interaction standpoint. There is no unreviewed safety question associated with this temporary repair. No Technical Specification changes are required . No UFSAR changes are required.

p U.S. Nuclear Regulatory Conunission Apdl 1,1999 Pane 138 of 247' l' l 236 Type: Miscellaneousitems Unit: 2 i

             'Iltle: Temporary Station Modification Work Order 98064276 Addition of a spool piece on the              ,

t Makeup Demineralized Water System for Outage support I

Description:

Temporary Station Modification Work Order 98064276 adds a temporary spool piece, consisting of polyethylene piping and an isolation valve, between the inlet and outlet headers of the Demineralized Water System Makeup Demineralizer Vacuum Demerators (MUDVAC) to allow the Demineralized Water System to supply a greater flow rate of

                     ~

makeup water to the Steam Castors during refueling outage 2EOC9. De only piping l affected by this modification is in the Demineralized Water System. The normal

                       ' deaerated makeup to the Steam Generators is limited to the 35 gallons per minute (gpm) output of the MUDVAC. His modification will allow the makeup rate to the Steam Generators to be increased to 100 gpm by use of a vendor supplied chemical deacration
                       ' unit which will be installed in the system upstream of the demineralizers. The design temperature of the MUDVAC discharge is 140 degrees F. The polyethylene piping is rated at 140 degrees F. The actual discharge temperature of the MUDVAC, when supplying makeup to the RMWST as designed,is 135 degrees F. The pressure rating on the inlet side of the MUDVAC is 115 psia while the outlet pressure is 165 psia. The actual pressure of the installed Temporary Station Modification will be approximately 80 psig which is well below the design pressure of either piping. The pressure rating of the piping material is 155 psia. This will not result in any pressure problem associated with normal system operation. The valve and system operation will be governed by the Chemistry operating procedure, Evaluation: He Demineralized Water System supplies filtered demineralized water to the Upper Surge Tanks for normal makeup to the secondary system. System makeup and flush water is supplied to other station systems by the Demineralized Water System through the          j Demineralized Water System Storage Tank and a Turbine and Service Building                -i i

l distribution header. The subject Temporary Station Modification will not affect the  ; normal distribution of water to these systems. The installation of this Temporary Station j Modification will result in a temporary inability of the Demineralized Water System to  ; I supply deaerated makeup water to the Steam Generators and Reactor Makeup Water Storage Tanks via the MUDVACs. After installation, this function will be available. The piping associated with this change is classified as non-nuclear safety related. There are no Design Basis Events which are applicable to this application or operation. No unreviewed safety questions are introduced as a result of this temporary station modification. No Technical Specification changes are required. No UFSAR changes are required.

U.S. Nuclear R,-N- i Co===s s-April 1,1999 ' ' Pane 139 of 247 208L Type: Miscellaneousitems Unit: 1 T1tle: Temporary Station Modification Work Order 98066402-01

Description:

Temporary Station Modification 98066402-01 is being issued to allow Unit I to be operated with the flow path used for Steam Generator Auxiliary Feedwater Nozzle i Tempeting Flow isolated. Indications are that past Unit operation with high tempering flow rates (200-250 gpm) have resulted in some piping erosion. Tempering flow rates were reduced after fuel cycle 6 (cuirent fuel cycle is 11), and are now in the 100-125 gpm range. However, some portions of this piping may still be adversely affected at these lower flow rates. Although there is no piping leak at this time, the isolation of tempering flow will remove the source of potential piping degradation. If a piping leak should occur in the area of concern, the unit may have to be shutdown to repair the leak. Therefore,

                 . this activity inssociated with unit reliability, and is not in response to any identified operability issue.                                              -

The purpose of the Auxiliary Feedwater Nonle Tempering Flow is to maintain the Auxiliary Feedwater nonles at final feedwater temperature to reduce the thermal shock to these nozzles associated with Auxiliary Feedwater System autostarts and testing, and the swapping of Main Feedwater from the Main Feedwater nozzles to the Auxiliary Feedwater noules during a unit shutdown. Per the Steam Generator manufacturer's documentation,100 cycles of Main Feedwater to Auxiliary Feedwater nozzle swaps ' without tempering flow are allowed. Any Auxiliary Feedwater autostart or Main 3 Feedwater to Auxiliary Feedwater nozzle swap that occurs while this temporary modification in place will comprise such a cycle. Since this temporary station  ; modification will be in place only for the remainder of the current fuel cycle, exceeding the allowable number of cycles is unlikely.

                                                                                                                 -1 UFSAR Section 10.4.9.2 states that Auxiliary Feedwater Nozzle Tempering Flow is                  ,

provided at all times when Main Feedwater bypass flow to the Auxiliary Feedwater  ! nozzle is isolated, except when a feedwater isolation signal is activated. Due to the short j duration of the Tempering Flow isolation, less than one fuel cycle, a UFSAR change to reflect this isolation is not required. , Main Feedwater System Reverse Purge will not be affected by this activity. The feedwater isolation function associated with the Tempering Flow Valves is not affected as all valves affected will be maintained in their safety position of " closed". l

   - Evaluation: Here are no unreviewed safety questions associated with this temporary modification.

His modification will reduce the probability of an erosion induced piping failure. He Steam Generator manufacturer states that 100 cycles of Main Feedwater to Auxiliary Feedwater nozzle swaps without tempering flow are allowed. Due to the temporary nature of this modification no UFSAR changes are required. No Technical Specification changes are required. l l o

i U.S. Nuclear Regulatory Commission April 1,1999 l Page 140 of 247 i 199 Type: Miscellaneous Items Unit: 0

Title:

Temporary Station Modification Work Order 9807929441

Description:

Temporary Station Modification Work Order 98079294-01 installs an electricaljumper to defeat the Nuclear Service Water Pit 2B two out of three Standby Nuclear Service Water Pond swap logic. The purpose of this modification is to disable the Emergency Lo 12 vel signal for the Nuclear Service Water Pit 2B during the 2EOC9 refueling outage to allow the pit to be drained for maintenance on Nuclear Service Water Pump 28. His temporary modification will allow draining of Pit 2B without initiating the Standby Nuclear Service Water Pond swap logic which positions the associted valves and starts the Nuclear Service Water Pumps. This evaluation addresses only the placement of the jumper (not the plant evolution). Evaluation: He Nuclear Service Water System will be aligned to the Standby Nuclear Service Water Pond prior to the installation of the temporary station modification to satisfy Technical Specification 3/4.3.2 Table 3.3-3,14g. leop B of the Nuclear Service Water System will  ! be declared inoperable prior to installing the modification to satisfy Technical Specification 3/4.7.4. Therefore, there are no unreviewed safety questions associated with this temporary modification. No Technical Specification changes are necessary. No UFSAR changes are necessary. j i 239 Type: Misce!!aneousitems Unit: 2

Title:

Temporary Station Modification Work Order 98083195-01, Installation of blanks in CPFU+2A

Description:

Temporary Station Modification Work Order 9808319541 addresses installation of blank flanges in the suction and discharge ductwork of the Containment Purge Filter Unit 2A (CPFU-2A). This modification will provide ductwork separation between Train A and Train B of the Containment Purge Ventilation System and enable operation of CPFU-2B during CPFU-2A maintenance activities. The Containment Purge Ventilation System is designed to maintain the environment of the containment within acceptable limits for personnel access during inspection, testing, maintenance and refueling operations and to limit the release of any contamination to the environment. His only safety related design basis function of this system is to assure that its containment penetrations can isolated. Evaluation: The Containment Purge Ventilation System is not an accident initiator or an accident mitigator. The blank flanges will be installed per seismic design criteria. This modification is intended to provide train isolation to enable maintenance work on the system. There are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. No UFSAR changes are required.

b i I I U.S. Nuclear Regulatory Commission l April 1,1999 Pase 141 of 247 l :169 Type: Miscellaneousitems Unit: 0 , l .. , . 1

Title:

Topical Report DPC-NE-201 IPA *DPC Nuclear Design Methodology for Core

Operating Limits for Westinghouse Reactors" l-

Description:

As part of an effort to maintain updated and accurate Topical Reports, a review of DPC. NE-201 IPA,

• Nuclear Design Methodology for Core Operating Limits for Westinghouse .

i

                      . Reactors". March 1990, was cW~i to identify any changes that have been made since the original approval. Three changes, summarized below, were identified that had not previously been addressed by a 10CFR50.59 evaluation, NRC SER, or approved Technical Specification change.

1. Update the methods for generating xenon transients in order to analyze more severe axial offsets in the maneuvering analysis. The third sentence of paragraph three in  ;

Section 2.3 was changed from : "The transient is initiated with an instantaneous change in l power level, control rod position and soluble boron concentration." to "The transient is initiated with some combination of instantaneous changes in power level, control rod position, and soluble boron concentration." l Due to fuel management changes, it became necessary to create more severe transients to ensure greater coverage of the AFD/ rod-insertion power level operating space.

                       'Ihe above change does not involve an Unreviewed Safety Question because accident initiating events, monitoring of parameters that ensure operation within safety analysis assumptions, SSCs, and UFSAR Chapter 15 safety calculations are not affected by the use of a conservatively large range of analyzed axial offset conditions in maneuvering analysis calculations.

2. Remove the model bias term in the square-root sum of the squares term of the SC factor in the CFM calculations. Change the equation in the definition of SC in Section 4.5 to delete a term that it is not statistically correct to include. It is not statistically correct to account for a model bias in the statistical combination of other uncertainty factors. Model specific biases are documented in the appropriate Topical Reports and are accounted for in safety related calculations.

The above change does not involve an Unreviewed Safety Question because accident I initiating events, monitoring of parameters that ensure operation within safety analysis I assumptions, SSCs, and Chapter 15 safety calculations are not affected by the correct I

                     . statistical combination of uncertainty factors for use in CFM maneuvering analysis              ,

calculations. J

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

positions. The first sentence of the paragraph describing FQD(x, y, z) in Section 6.1 was j changed from "FQD(x,y,z) at full power is calculated from the nominal, all rods out,  ! equilibrium power distribution that is used to start one of the xenon transients" to

                       *FQD(x,y,z) at full power is calculated from the nominal, all rods out or expected operating rod position, equilibrium power distribution that is used to start one of the i                                                                                                                       !

l. 1

                                                                                                                       )

p

 .L U.S. Nuclear Regulatory Commission April 1,1999 Page 142 of 247 xenon transients".

Surveillance of FQ specified in the topical requires calculated power distributions and margins to limits to determine a maximum allowed FQ. Generating calculated power distributions at expected operating rod positions provides for accurate monitoring of the core against operating limits, because rod position and nuclide distribution is more l accurately modeled. This change reduces uncertainty between incore measured and calculated power distributions. q ne above change does not involve an Unreviewed Safety Question because: accident initiating events, the monitoring of other parameters that ensure operation with safety analysis assumptions, SSCs, and Chapter 15 safety calculations are not affected by the use of expected operating rod positions in the calculation of design FQ values. He monitoring of FQ is improved since the uncertainty between mcore measured and calculated power distributions is reduced. Evaluation: Revision Oa of Topical Report DPC-NE-201 IPA was performed to reflect current methods provided in other NRC approved Topical Reports, to reflect curren: Technical Specifications, and to correct administrative errors. No unreviewed safety questions were created as a result of this Topical Report revision. No Technical Specification changes are required. No UFSAR revisions are required. l 188 Type: MiscellaneousItems Unit: 0 l

Title:

UFSAR changes to resolve editorial and technical discrepancies as identified in the Problem Identification Process Report 0-C97 2515

Description:

In Section 3.9.1.1

• Mechanical Systems and Components / Design Transients" the expected normal heatup and cooldown rates will be changed from 50 degrees F per hour j to 30 degreces F per hour for heatup and 75 degrees F per hour for cooldown. Also, in Section 5.3.3.1 " Reactor Vessel Integrity / Design", the normal and emergency heatup and cooldown rates were changed from 75 degrees F per hour normal and 100 degrees F per hour emergency to 30 degrees F per hour for normal heatup and 75 degrees F per hour for normal cooldown; and 60 degrees F per hour for emergency heatup and 100 degrees F per hou.r for emergency cooldown. In addition, in Section 10.4.4.2
• Turbine Bypass System / System Description", the information on normal reactor coolant system cooldown rate will be deleted.

Evaluation: here is no unreviewed safety question associated with these changes. These changes correct errors in the UFSAR. The performance of systems or components will not be i affected by these changes. These changes will not have any effect on any accident described in the UFSAR. No Technical Specification Changes are required. Changes are required for UFSAR Sections 3.9.1.1, 5.3.3.1, and 10.4.4.2. l l b

w 1 U.S. Nuclear Regulatory Comunission EAprilli1999 Page 143 of 247 260 Type: Nuclear Station Modification Unit: 2

Title:

Nuclear Station Modification CN-21373, Modify Main Steam Isolation Valve Control Circuitry ,

Description:

Nuclear Station Modification CN-21373, Modify Main Steam Isolation Valve Control Circuitry, makes the Main Steam Isolation Valves less likely to close following failure of certain circuit components (e.g., optical isolators, relays) to improve unit operating reliability and reduce the associated plant transients. His will be accomplished while not degrading the safety related circuit logic for the Solid State Protection System / Engineered Safety Features Actuation System automatic or manual Main Steam Isolation functions. The control circuits for the Main Steam Isolation Valves contain both non-safety-related and safety related portions. Currently, battery backed AC power is provided to the push-button portion of the control circuit. This modification will change this feature. Digital Optical Isolator failures will be minimized in the new design as DC input optical isolators will be used which have been proven to be more reliable. There will be no change to the solenoid valves that supply and vent air to open/close the Main Steam Isolation Valves. These will remain as an " energized to open the Main Steam Isolation Valve" design. This means the supply solenoid valves are open when energized and the vent solenoid valves are closed when energized. The new design uses normally de-energized optical isolators for safety Train A/B isolation and separation. The non-safety pushbutton to close the MSIVs will be moved to I the A Train control circuit, safety related side of the circuit which allows for elimination j of the normally energized optical isolators. The pushbutton actuates the solenoid valves l in both trains to open the Main Steam Isolation Valves. However, to close the valves, ) only the train A solenoids are actuated. Evaluation: The Main Steam Isolation Valves are accident initiators. Turbine Trips and Inadvertent I Closure of Main Steam Isolation Valves result from direct closure of a Main Steam Isolation Valve or indirectly through subsequent protective action. De changes in the control circuitry have reduced the likelihood of closure of the Main Steam Isolation  ; Valves due to a single failure of a component in the control circuitry. Herefore, the i probability of an accident is not increased. De accident mitigation function of the Mam ) Steam Isolation Valves is closure to ensure no more than one Steam Generator blows down following a Steam System Piping Failure. Neither the actuation signals, stroke time, or sealing ability of the valves is affected by this modification. No new failure  ; i modes or effects are being introduced. The control circuits are still single failure proof with respect to the safety function of Main Steam Isolation Valve closure on demand. The characteristics of the Main Steam Isolation Valves are maintained as described in the bases of the Technical Specifications. No unreviewed safety question is associated with this modification. No change to the Technical Specifications is required. A change to Section 10.3 of the UFSAR is required to describe the revised control circuit operation for the Main Steam Isolation Valves. E

p i UJ!i. Nuclear R, " "- i Comunission April 1,1999 - Pase 144 of 247' t 258 Type: Nuclear Station Modification - Unit: 'l I litle: Nucleat Station Modification CN 11271, Modify Chemical and Volume Control System Pump Suction Pulsation Dampener .

Description:

Nuclear Station Modification CN ll271, Modify Chemical and Volume Control System Pump Suction Pulsation Dampener. The Chemical and Volume Control System has three

  !                    charging pumps one of which is a positive displacement (reciprocating) pump. This pump has a history of unavailability which is primarily due to failures of the suction pulsation dampener his modification will e&ctively resize the current dampener by -

providing a water charging connection to the side of the dampene. Changes per this ! modification will consist of a charging turning connection and water charging cylinders for adding water to the dampener. The charging loop will include a quick disconnect to allow the connection of a vacuum pump. isolation valves will be provided for the water charging cylinder, quick disconnect, and a gas charge (Schradrer) valve. The existing pressure gauge on the charging loop will be replaced with a compound gauge to read vacuum or positive pressure. In order to verify that the new gas and water charging of the pulsation dampener is effective, suction and discharge piping on the positive displacement pump will be instrumented. Pressure transducers will be added to the piping, off existing suction and discharge pressure gauges, that will monitor the " dynamic pressure pulsations" in the piping. Two peak level meters per unit will be added to receive signals from the new pressure transducers. The Unit I peak level meters will be provided with permanent power. Evaluation: These changes will not adversely affect the design conditions of the Chemical and l l Volume Control System. The ability and reliability of the system will not be adversely affected. No accident initiators are adversely affected. No fission product barrier or l source term evaluation is adversely affected. Imading of affected electrical systems will 1 remain within design specifications. Therefore the probability or consequences of an l accident previously evaluated in the UFSAR will not be increased. He modification will not degrade compliance with seismic design criteria. During a loss of coolant accident , l the Chemical and Volume Control System is isolated except for the centrifugal charging i pumps and the piping in the safety injection path. Derefore the probability or consequences of a malfunction of equipment important to safety addressed in the UFSAR is not affected. Certain non nuclear safety related pressure transducers will be qualified l for nuclear safety related service. The modification is in compliance with Appendix R l Criteria. No new failure modes are created. Derefore the probability of an accident or malfunction different from those already evaluated in the UFSAR will not be increased. L The modification will not affect any safety limits, set points, or operating parameters. No fission product barriers are adversely affected. Derefore the margin of safety as defined in the bases of any Technical Specifications will not be reduced. There are no Unreviewed Safety Questions associated with this modification. No Technical Specification Changes are required. No UFSAR changes are required. l' l l I I I j l 1  ! i _ J

1- 1 l t

  . U.S. Nuclear Regulatory Commission Apeil 1,1999 Page 145 of 247 f

1 Type: Nuclear Sution Modification . Unit: 1 71tle: Nuclear Station Modification CN-11335, Addition of Automated Sampling to the Nuclear j Sampling System l

Description:

Nuclear Station Modification CN-11335, Addition of Automated Sampl.ing to the Nuclear Sampling System, provides automated sampling equipment. A panel containing equipment for sample analysis will be installed in the Unit i Sampling Rooms. Connections will be made to existing Nuclear Sampling System sample lines in the vicinity of the Sample Room. The connecting tubing will be equipped with solenoid valves to select and route samples to the panel for analysis. The process of sample j collection and analysis will be controlled from a dedicated workstation located in the

                                                                                                                   )

Units Sample Room. 'Ihe equipment affected by the modifications is not nuclear safety I related.

       .isvaluation: This Nuclear Station Modification does not create an unreviewed safety question. None of the equipment affect by this modification is nuclear safety related. Postulated breaks in the sample lines are bounded by accident analyses already contained in the UFSAR. No Technical Specification changes are required. The ability to perform sampling and analysis using existing equipment will not be affected by the modification. FSAR changes are required to sections 9.3.2 and 9.3.4. Changes for FSAR figures 9-49,9-62,9-78,9-79,9-80,9-81, and 9-122 will be required l

l l l t_

U.S. Nuclear Regulatory Conunission Apdf 1,1999 Page 146 of 247 2 Type: Nuclear Station Modification Unit: 1

Title:

Nuclear Station Modification CN-11341, Replace Nuclear Service Water System piping and valves with stainless steel components

Description:

Nuclear Station Modification CN-11341. Replace Nuclear Service Water System piping and valves with stainless steel components, will improve corrosion resistance in limited part's of the Nuclear Service Water System. This modification replaces some piping and valves that supply the Spent Fuel Pool Cooling System and Containment Seal Water Injection System assured makeup from the Nuclear Service Water System, with stainless , steel components. This is an attempt to remedy problems with these lines becoming corroded as evidenced by the ilushing frequency increasing from quarterly to monthly. Also, vent valve IRN-155 will be deleted. This valve was used in an earlier design to provide cooling flow to the Spent Fuel Pool Cooling System Pump Motor Coolers. The Component Cooling System now provides cooling water to the Spent Fuel Pool Cooling System Pump Motor Coolers and valve IRN-155 is no longer needed. A pipe cap will provide an equivalent nuclear safety related pressure boundary. Evaluation: The components being added are functionally equivalent from a performance perspective and should be an improvement with respect to corrosion resistance. The function of the valve being removed has been incorporated into another system. No Unreviewed Safety Questiens are created as a result of this modification. No Technical Specification changes are required. UFSAR Figures 9-26 and 9-30 (flow diagrams) will be revised to show the new pipe specification for the stainless material, new valve types (globe valve changing to ball valve) and the elimination of valve IRN-155. f l h'

[; U.S. Nuclear Regulatory Conunission - I

    - Apdf 1,1999 ~

Page 147 of 247 p 259 Type: Nuclear Station Modification Unit: 1

Title:

Nuclear Station Modification CN-11383, Auxiliary Feedwater System flow reduction l modification

Description:

Nuclear Station Modification CN-11383 will reset the Auxiliary Feedwater System flow control valve travel stops to %rottle down" on Auxiliary Feedwater System flow. Recent

                       . concerns have been identified about potential air entrainment/vortexing related to the    ,

Auxiliary Feedwater Condensate Storage Tank and Upper Surge Tank. The flow

                       ' reduction will provide lower flow losses (frictional pressure drop) to assure that if the ]

Nuclear Service Water Systen swapover does occur, the Condensate supply check valves

                        - will stay closed due to adequate backpressure, thereby assuring no air entrainment from the non-safety related condensate sources. Also, reducing auxiliary feedwater flows will provide additional margin to premature swapover to the Nuclear Service Water System (assurred source), with respect to the existing higher flows following a normal Reactor /I'urbine Trip. This modification involves no field work. Only a flow balance per the Auxiliary Feedwater System flow balance procedure and adjustment of travel stops (to obtain the required Cv values for various Auxiliary Feedwater System flow control valves) is required. Net Positive Suction Head considerations are improved by the modification because flow is reduced, thereby increasing available net positive suction head and reducing required net positive suction head.

Evaluation: This modification preserves the same requirement of "no operatoraction required for thirty minutes" via a passive system design. Control valve positions were selected to provide inherent runout protection for all accidents analyzed where the Auxiliary Feedwater System is credited as an accident mitigation system under modification CN-11371. Since this modification reduces Auxiliary Fcedwater System flow, more margin to runout is provided. There are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. No UFSAR changes are required. I I i i i

n U.S. Nuclear Regulatory Ceaunission

  + Apail1,1999 Paste 148 of 247 234    Type: Nuclear Station Modification                                       Unit: 2                       !

L . L '11tle: Nuclear Station Modification CN-21379, replace 6.9 Kilovolt switchgear tie breakers

, with faster vacuum breakers 1

L

Description:

Nuclear Station Modification CN-21379 Revision 0 will replace the current 6900 Volt tie l breakers with faster vacuum breakers, because the current fast transfer schemes have the l l, _ potential to subject the switchgear loads (motors) to excessive restart torque. The l l modification will also install switches to provide the function of defeating the automatic L fast transfer when the units are offline. The Unit Main Power System is part of the Onsite Power System which directly interfaces with the Offsite Power System. The balance of the Onsite Power System consists of the Diesel Generators, batteries, controls and auxiliary power system. The Unit Main Power System includes the main generator, isolated phase busses, Generator Power Circuit Breakers and associated motor operated disconnects, main step-up transformers, four unit auxiliary (20.9/6.9 KV) transformers and one auxiliary (20.9/13.8 KV) transformer. The Unit Main Power System starts with the main generator which feeds two trains (A and B) of transformers, breakers and conductors. 'Ihe main generator feeds: two "20.9 KV to 230 KV half sized" unit step up transformers, and four "20.9 KV to 6.9 KV half sized" unit step down auxiliary transformers, and one auxiliary "20.9 KV to 13.8 KV step down" transformer through the isolated phase busses and two Generator Power Circuit Breakers.

                      "Ihe Offsite Power System consists of the entire station switchyard including Power

Circuit Breakers, associated Motor Operated Disconnects, conductors, and protective

! relaying. I Power is fed from the Onsite Power System to the 6900 Volt Normal Auxiliary Power ) System. The 6900 Volt Normal Auxiliary Power System includes four switchgear assemblica of a split bus design including a short leg split from a long leg on each bus. The long and short legs of a respective switchgear are normally powered from opposite l trains of the Onsite Power System. These switchgear assemblies are named 2TA,2TB, 2'IE, and 2TD. Normal power for these switchgear is supplied from 20.9/6.9 KV transformers 2TI A,2T2A,2T2B, and 2T1B. In order of short leg then long leg, the l power sources on 2TA,2TB,210,2TD are trains AB, BA. AB, B A.

                                                                                                                   ]

l The 6900 Volt tie breakers provide the capability to tie together the short and long leg of i l a particular switchgear assembly such that both legs are powered by the same train of the  ! Onsite Power System. The 6900 Volt Normal Auxiliary Powe System feeds the 4160 Volt Essential Auxiliary Power System. Normal aligr. ment has esin A 20.9/6.9 KV i l transformer 2T2A feeding switchgear 2TA (short leg) which feeds iNsential bus 2 ETA { l through 6.9/4.16 KV transformer 2ATC; and train B 20.9/6.9 KV transformer 2T1B feeding switchgear 2TD (short leg) which feeds Essential Bus 2ETB through 6.9/4.16 KV l transformer 2ATD. l

                     . It is through these 6900 Volt tie breakers that the 4160 Volt Essential busses can be supplied from the opposite 6900 Volt train. A " Zone lockout" is protective action provided in the switchyard that isolates one side of the Offsite Power System including the genesator from a fault by opening the switchyard and associated Generator Power 1

1 1 l l i U.S. Nuclear Regulatory Comunission April 1,1999 j _ Page 149 of 247 Circuit Breakers. Following this isolation, the tie breakers can power the de-energized 6900 Volt switchgear from the opposite train of Unit Main Power System with the resulting alignment having one 6900 Volt switchgear completely supplied from one auxiliary transformer. This automatic transfer will also occur for any reason an under

                                                                                                                   )

j voltage condition exists on a particular short or long leg provided the under voltage is not i due to a fault on the bus.

                    ' The subject tie-breakers are non-safety related devices. Essential Power is provided to the safety related 4160 volt Essential Auxiliary Power System via the Class IE Diesel
                    . Generators. Therefore, the normal power source to the station auxiliaries is mainly important to the extent that plant transients are not created through intermption of power.

Interruption of power through the normal feed paths discussed above (2ATC and 2ATD), would create a blackout condition such that the load sequencer would start and load the Diesel Generator onto the affected Essential Bus. Additionally, interruption of power j would terrninate powered operation of the reactor coolant pumps and they would begin a coastdown due to the rotating inertia of the flywheel. Loss of forced flow in one loop will . cause a Reactor Trip from an initial power level exceeding 48 % (P8). Below 48 % but j l

                    - greater than 10 % (P10), two loops must lose forced flow to result in a Reactor Trip. Per     l Tech Spec Table 3.3.1-1 and UFSAR Table 7-1, a 2 out of 4 undervoltage condition on          j reactor coolant pump motor voltage, will cause a Reactor Trip at less than or equal to 77    4
                      % normal voltage per UFSAR Section 7.2.1.1.2. This protection is provided by the Reactor Coolant Pump Monitor System. Unsuccessful transfer of power will result in a loss of flow. A slow transfer of power will result in a Reactor Trip. A Reactor Trip causes a Turbine Trip from any power level if the Turbine is not tripped. Thus, the 6900 Volt Normal Auxiliary Power System is identified as an accident initiator. The Unit Main Power System is also an accident initiator since its transformers feed the 6900 Volt switchgear.

Additionally, through auto-starting the Diesel Generators, challenges to accident mitigation equipment are increased if normal power equipment is made less reliable. Thus, the tie breakers are evaluated to be equipment important to safety as they can initiate a transient (Reactor Trip / Turbine Trip) and cause challenges to accident mitigation equipment (Diesel Generators) if they fail to operate as designed when called upon. In addition to the above equipment other equipment such as Hotwell Pumps. Condensate Booster Pumps, and Condenser Circulating Water Pumps are powered from the 6900 Volt switchgear. Evaluation: There are no unreviewed safety questions associated with this modification. All , scenarios involving possible motor restart torque potential are improved by this l l modification due to the faster transfer time provided by the replacement breakers. The I effects of this modification on breaker coordination has been evaluated. It was determined l [ that the effects, if any, would be positive. No Technical Specification changes are l required. A change is required for UFSAR Section 8.3.1.1.1.3. l

y l

  . U.S. Nuclear RC ' y Commission April 1,1999 Page 150 of 247 '

l' 219! : Type: Nuclear Station Modification Unit: 2

             'Iltle: Nuclear Station Modification CN-21380/0, Control Circuit Modification for Four Chemical and Volume Control System Valves

Description:

Nuclear Station Modification CN-21380/0, Control Circuit Modification for Four Chemical and Volume Control System Valves, will modify the control circuits for valves 2NV-294,2NV-309,2NV-148, and 2NV-849. Dese valves are part of the Chemical and Volume Control System that are used to establish charging flow, letdown flow, and reactor coolant pump seal water injection flow. These changes are deemed beneficial in order to provide the ability to control these valves following loss of non-safety power and/or loss of essential power. All of the valves are air-operated and have non-safety controls as they all fail safe to the

                     . open position on loss of air or power. In addition, all except 2NV-849 have "AND logic" in their control circuits vehich requires both trains of essential power to be available in order to have control from the control room. De design for the "AND logic" is to ensure that the Control Room controls are disabled assuming a single failure of one of the Auxiliary Shutdown Panels to swap to LOCAL control. This failure could be a power failure from essentialpower associated with the A or B Auxiliary Shutdown Panel controls. The end result is that control from the Control Room should not interfere with proper operation of the Auxiliary Shutdown Panel controls.

A terminal box and plug arrangement is provided in the Auxiliary Feodwater Pump Turbine Control Panel. This plug will have two matching receptacles. %e normal position will allow for Control Room control of the valves, while the othe r will be utilized . upon transfer to the Auxiliary Shutdown Complex. In addition non-safet/, non - interruptible (battery backed) power will be provided to the controis for the valves.- Evaluation: De controls of these Chemical and Volume Control System valves are only being I modified to allow for Control Room control during unavailability of one train of essential

                                                                                                                       ]

power. %e actual control of the valves is unchanged during normal and abnormal (less j Of Control Room (LOCR)] situations. Thus, these accidents are not made more likely to occur. The plug will be located in a Vital Area (Auxiliary Feedwater Pump Room) inside a terminal box. Plug withdrawal during normal operation leading to some charging or j letdown transient is deemed not credible. Therefore, this modification will not increase the probability of an accident previously evaluated in the UFSAR. Since the procedure step to operate the new plug is assumed to be carried out without failure, and the plug is not deemed to have any failure mo:les which could prevent disablement of Control Room controls, the probability of a malfunction of equipnent important to safety evaluated in the UFSAR is not increased. No Appendix R concerns are involved with this modification. No seismic interaction concerns are created. leads on affected power supplies remain within design specifications. Derefore, there is no increase in the probability of malfunctions of equipment important to safety. Also, no malfunctions or accidents of a different type than l cvaluated in the UFSAR are created. Response to LOCR events will not be degraded since the new plug will be installed by l

g F. i U.S. Nuclear Regulatory Commission Apdf .1,1999 Page 151 of 247 procedure to disable Control Room controls. Response to accident mitigation will not be degraded as the controls for the subject valves have not been modified to alter the post accident response. Response of valves 2NV-294,2NV 309, and 2NV-148 to loss of essential power or less of Offsite Power (LOOP) will keep these valves controlling to the control signal via the Process Control System due to the presence of the non-interruptible non-safety power source (absent a seismic event) and absence of essentially powered optical isolators in the new circuits. De non-safety non-interruptible power supply will result in more reliable operation. The valves will still fail open on loss of air or power. Derefore, the consequences of accidents and malfunctions of equipment are not increased. Margin of safety is not affected as no fission product barriers are modified. Reactor coolant pump seal injection via valves 2NV-294 and 2NV 309 will function as before with the exceptions noted above for various power failures. The valves still fail open on loss of air or power. The power is just more reliabic. There is no unreviewed safety question associated with this modification. No change to the Technical Specifications is required. A change to UFSAR Section 7.4.7.1 is required to describe the method of defeating Control Room Controls for the four valves..

                                                                                                             )

i l l i l-l

l U.S. Nuclear Regulatory Commission April 1,1999 Page 152 of 247 244 Type: Nuclear Station Modification Unit: 2

Title:

Nuclear Station Modification CN-21381 Revision 0, Add bonnet vents to several Residual Heat Removal, Containment Spray, and Safety Injection System valves to eliminate pressure locking concerns. l

Description:

Due to pressure locidng concerns identified during the response to NRC Generic letter 95-07, bonnet vents have been determined to be necessary to maintain long term operability on certain specified valves. Valve pressure locking occurs when high pressure fluid is trapped in the bonnet of a closed gate valve. The bonnet on such a valve could i become pressurized to the extent that the valve may be unable to open if required to I

                   , perform its safety function, A bonnet vent relief vent path on each valve will be installed with a small globe valve in the line to isolate the path if necessary. The valves affected by this modification are: 2NS-IB and 2NS-18A (Containment Recirculation Sump to Building Spray Pump Suction Isolation Valves),2NI-136B (Residual Heat Removal Pump to Centrifugal Charging Pump isolation valve), and 2NI-9A and 2NI-10 (Centrifugal Charging Pump to Cold leg injection isolation valve),

in addition, this modification is adding vent lines to the piping on the Residual Heat Removal Pump suction. nese process line vents will aid in draining the adjoining Residual Heat Removal System piping. The subject valves are flexible wedge valves with two seating surfaces. The bonnet vents will bypass one seat and relieve / equalize pressure between the process pipe and the bonnet. The vent relief path will be a 1/2 stainless steel nuclear safety related line. It will be compatible with temperature and pressure requirements of the interfacing system. The bonnet of each valve will have a 1/2 inch half l coupling installed to which the 1/2 inch piping will be attached. The bonnet vent isolation  ! valves provided will be 1/2 inch globe valves and will be locked open. If a Motor Operated Valve develops a seat leak, the globe valves can be closed. De applications associated with these valves will vent to the higher pressure (upstream) side of the valve. 1 There will be no change to the operation of the valve or associated system. The addition of the weight of the new components has been evaluated for impact to the stress analysis and support / restraints. No electrical considerations are involved with this modification. I Material compatibility requirements of UFSAR 6.3.2.4 are satisfied with the specified l piping and valve materials. All of the modified valves are active valves described in l UFSAR Table 3-104. ECCS leakage will not be adversely affected with the addition of the isolation valves and process line vent valves. Hey are of a packless design . l Evaluation: There is no unreviewed safety question associated with this modification. None of the valves involved are accident initiators but all are accident mitigators. Piping stress and support / restraint design have been evaluated for the involved items. No UFSAR changes are required because only Unit I flow diagrams are represented in the UFSAR. No Technical Specification changes are required. ! 1 4

r.

l j I

U.S.Noclear Regulatory Commisalon ) April 1,1999 Page 153 of 247 . i

                                                                                                 .                              i I

178 Type: Nuclear Station Modification Unit: 2

Title:

Nuclear Station Modification CN-21388 l

Description:

NSM CN-21388/0 will replace the existing 125 VDC Diesel Generator Auxiliary Power

                            . (EPQ System) Batteries The replacement batteries will have 40 empere-hours of additional capacity compared to the existing batteries. The replacement batteries will
                            ' have two additional cells per battery for improved capacity. The new batteries will also have nickel plated negative plates to retard migration ofimn at elevated temperatures for
                             . better reliability. '

Additionally, cooling units 2VDAH2A and 2VDAH2B will be removed along with the associated ducting and power. Air temperature thermometers 2VDTH5240 and 2VDTH5250 will also be removed . These thermometers currently exist mounted to the wall above the battery racks. An interference exists since the new battery cells are taller l- than the current cells. M

. The 125VDC Diesel Essential Auxiliary Power (EPQ) System provides a separate and l independent train of 125 voitde power to each diesel generator. Each train consists of a 125VDC battery and a battery charger powered from its associated train of 600 volt essential auxiliary power. The 125VDC Diesel Essential Auxiliary Power System is shown in UFSAR Figure 8-26.

i h diesel generator batteries are Nickel Cadmium Class IE batteries. hse batteries are sized to carry their assigned loads for two hours. W load duty cycle used to size the diesel generator batteries is shown in UFSAR Figure 8-27. l The EPQ System provides direct support to ensure the proper operation of the emergency D/G. Each train of the 4160VAC Essential Auxiliary Power System is supplied with emergency standby power from an independent diesel generator. Each diesel generator is designed to attain rated voltage and frequency and to accept load within 11 seconds after receipt of a start signal. The system has increased capacity, extended service life, and i improved reliability. The additional weight of the batteries has been qualified for the rack I and floor loadings including seismic considerations. Charger sizing and Diesel Building j

                            - Ventilation has been evaluated to be acceptable. Cable sizing and circuit breaker                 l interrupting capability has been evaluated to be acceptable.                                    -

l Proper design considerations have been made so as to assure no new failure modes are  ! created. The resulting design is more reliable and robust without placing any l unmanageable demands on supporting systems (ventilation, structural supports, chargers, i cables and breaker operation).  ! l ~ Evaluation: Modification CN-21388/0 does not introduce any Unreviewed Safety Questions. No Technical Specification changes are required. UFSAR text changes are required to ] section 8.3.2.1.2.2 to properly describe the rating of the batteries; and to section 9.4.4.2 to  ; describe the changes to the Diesel Building Ventilation System. I

  \
    . U.S. Nuclear Regulatory Commission April 1,1999 - .                                                         ,

Paste 154 of 247.

      -235 ' Type: Nuclear Station Modification                                       Unit: PION 10/5/98
               'lltle: Nuclear Station Modification CN-21388 Revision 0, Replace the Diesel Generator 125 Voh DC Batteries

Description:

Nuclear Station Modification CN-21388 Revision 0, will replace the 125 Volt DC Batteries of the Diesel Generator Auxiliary Power System. The replacement batteries will have 40 ampere hours of additional capacity compared to the existing batteries. Ihe - replacement batteries will have two additional cells per battery for improved capacity. The replacement batteries will have nickel plated negative plates to retard migration of iron at elevated temperatures for improved reliability. Also cooling units 2VDAH2A and 2VDAH2B will be removed along with the associated ducting and power. Air temperature thermometers 2VDTH5240 and 2VDTH5250 will also be removed. These thermometers currently are mounted on the wall above the battery racks and are causing an interference since the replacement batteries are taller than l the current cells. The 125 Volt DC Diesel Essential Power System provides a separate l and independent train of 125 volt DC power to each diesel generator. Each train consists l of a 125 Volt DC battery and a battery charger powered from its associated train of 600 volt essential auxiliary power. The Diesel Generator Batteries are Nickel Cadmium Class IE batteries which are sized to carry their assigned loads for two hours. The Diesel Generator Auxiliary Power System provides direct support to ensure the proper operation of the emergency diesel generator. Each train of the 4160 Volt essential power system is j supplied with emergency standby power from an independent diesel generator. Each i diesel generator is designed to attain rrted voltage and accept load with 11 seconds after !- receipt of a start signal. The system has increased capacity, extended service life, and i improved reliability. The additional weight of the batteries has been quailified for the l rack and floor loadings included in the seismic considerations. Charger sizing and the Diesel Building Ventilation System has been evaluated to be acceptable. cable sizing and circuit braker interrupting capability has been evaluated to be acceptable. Evaluation: The replacement battery is more reliable than the previously installed battery. The replacement battery does not place any unmanageable demand on supporting systems (ventilation, structural supports, chargers, cables and breaker operation). This modification does not introduce any unreviewed safety questions. No Technical i Specification changes are required. Changes are required for UFSAR Sections 8.3.2.1.2.2 and 9.4.4.2. I i l l l- , l 1 l a

U.S. Nuclear Regulatory Conunission April I,1999 Page 155 of 247 3 Type: Nuclear Station Modification Unit: 0

Title:

Nuclear Station Modification CN-50431/01, Provide Instrument Air System Dryer Automatic Operation, Flow Instrumentation, Emergency Lighting and Relief Valve Replacement

Description:

Nuclear StationModification CN-50431/01, Provide Instrument Air System Dryer Automatic Operation, Flow Instrumentation, Emergency Lighting and Relief Valve Replacement, provides a full flow bypass line around the Instrument Air Dessicant Dryers that were installed per modification CN-50431/00. This line will allow for air to bypass the dryers in case a failure of the dryer package were to occur that could completely block the flow path. In addition to the full flow bypass line, this modification will add a computer alarm and annunciator window controlled off of the bypass valve limit switches. Other annunciators will be added for more complete Control Room status of instrument air system equipment / components. Modification CN-50431/01 will also install flow monitoring instrumentation on the discharge piping ofInstrument Air System Compressors D, E, and F between the compressors and the receiver tanks. Additionally, emergency lighting which currently exists in the area of the Instrument Air Compressors will be relocated to optimize the available lighting at the Air Compressor control panel. Lighting will also be added to the area where the dryers are located. Additional relief valve capacity was deemed necessary to resolve overpressure protection concerns. Valves IVI-007, IVI-019, and IVI-031 (Instrurnent Air System Receiver , Tanks Relief Valves) will be replaced as part of this modification with relief valves of , higher capacity. Addition of these valves resolves the concerns associated with l overpressure protection. Also, these valves do not introduce any undesirable effects to the Instrument Air System. Evaluation: No Unreviewed Safety Questions are created this modification. This modification will have no effect on the reliability of the Instrument Air System. The modification will eliminate the possibility of loss ofinstrument air due to flow blockages associated with the dryer package. No Technical Specification changes are required. This modification results in several changes to sections and figures in the UFSAR. Changes are being nude to Table 8-6 to correct a typographical error made on an earlier revision. Text changes are being made to section 9.3.1

• Compressed Air System" to properly reflect as-built plant conditions. Section 9.3.1.3 " Safety Evaluation" is revised to clarify power supply alignment following modification CN-50431/00. The Instrument Air System flow diagrams, UFSAR Figures 9-70 and 9-71, are being revised as part of the annual update ,

process which will include a revision to Table 1-4. I i l i

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! Apdf 1,1999 Page 156 of 247 l: . l .4 Type: Nuclear Station Modification Unit: 0 -

               'I1tle: Nuclear Station Modification CN-50431/02 Part 1, Instrument Air System Electrical Abandonment and Cable Voiding ll

Description:

Nuclear Station Modification CN-50431/02 Part 1, instrument Air System Electrical Abandonment and Cable Voiding, abandons electrical equipment which is no longer used. "Ihis modification will electrically abandon (spare the load center breakers) the breaker which previously supplied the power for Instrument Air System Compressor A, and the breaker which previously supplied the power for Instrument Air System l Compressor C. These compressors are no longer needed since Modification CN-50431/00 removed them from service and installed new centrifugal compressors. Also, all associated power and control cables will be removed. Modification CN-50431/00 disconnected the power cables from these breakers rendering them incapable of adding load to the associated blackout load centers.- Additionally, other associated support loads such as refrigerant dryers, air compressor control panels, solenoid valves for recirculated cooling water system cooling (to the previous reciprocating compressors) will be abandoned. This modification will update documents (one line and three line drawings, AC and DC electrical elementary drawings, databases, etc.) which describe the associated cables. Additionally, the affected power and control cables will either be completely voided, only disconnected at tiw abandoned - equipment end, or abandoned in place. No physical changes are being made to any equipment that is currently capable of being operated. This anodification is mainly a document update, however some cables are being completely removed. Evaluation: No Unreviewed Safety Questions are created by this modification.There is no effect on any component that is currently capable of being operated. Unused items are being abandoned or removed. No Technical Specification changes are required. UFSAR . changes are required to Figures 8-01 and 8-21. i 1 1 i

r U.S. Nuclear Regulatory Commission April 1,1999 Page 157 of 247 261 Type: Nuclear Station Modification Unit: 0

Title:

Nuclear Station Modification CN-50449, Provide a catchment for the Steam Generator Drain Tank

Description:

Nuclear Station Modification CN-50449 will provide a catchment for the Steam Generator Drain Tank. The Liquid Radwaste System collects, segregates and processes all liquid waste which is potentially radioactive. One of the subsystems' to the Liquid Radwaste System is the Steam Generator Drain Tank subsystem. This subsystem collects, holds, recirculates and samples waste water generated by the draining and flushing of a Steam Generator for maintenance, testing, or inspection. This subsystem includes a seismic category I building which consists of two 50000 gallons tanks that are actually compartments of the building. His modification will constmet a catchment for the Steam Generator Drain Tank discharge piping. This will contain any leakage from Steam Generator Drain Tank discharge piping and ensure compliance with regulatory Guide 1.143. Evaluation: The Steam Generator Drain Tank pressure boundary is not degraded by this modification. No other equipment used to process radioactive fluid is adversly affected. No equipment used to mitigate any accident is involved in this modification. There are no unreviewed safety questions associated with this modification. No Technical Specification changes are required. No UFSAR changes are required. I' L L

r J U.S. Nuclear Regulatory Commission April 1,1999 Page 158 of 247 5 Type: Nuclear Station Modification Unit: 0

Title:

Nuclear Station Modification CN-50455, Control Room Renovation

Description:

Nuclear Station Modification CN-50455, Control Room Renovation, modifies the Control Room " Horseshoe Area" layout to better meet the needs of Control Room personnel. Before the modification workstations for the Reactor Operators (ROs) did not face the Control Boards. Senior Reactor Operators did not have a suitable workstation. Control Board MC-15 was deleted to accommodate these changes. A specified area of the Control Room floor was raised seven inches. New furniture was installed with work stations. These changes provide a better man-machine interface (workstations) and human factors considerations. New computer workstations were installed for the Reactor Operators. Communications is improved between the Control Room SRO and ROs. All items previously located on MC-15 will be relocated in the Control Room. These include the Containment Video Monitor, Camera Controls, Fire Brigade Radio microphones and speakers, Reg Guide 1.47 Annunciator Panel, Event Recorder Work Stations, phones, flashlights, and other video monitors. Also, a new emergency communications console was installed to replace the existing one allowing all associated equipment to be located together. The Control Room changes provide more and better man / machine interfaces for the Control Room operators. The Main Control Board is not deg aded with the new attachments. No degradation is imposed on the Control Room Chilled Water and Control Room Ventilation Systems, control room pressure boundary, or equipment contained within the Control Room envelope. The ability of the Operators to monitor ESF Systems status with respect to

                 " Bypassed Conditions" via the Regulatory Guide 1.47 panel integrated within the OAC is not degraded with respect to availability ofinformation or reliability of information.

Also, the heat loads added to the Control Room have been evaluated and a determined not to impact the Control Room Chilled Water and Control Room Ventilation Systems ability to maintain the Control Room within design parameters Evaluation: No Unreviewed Safety Questions are identified with this modification. The modification does not affect any accident initiators. The only accident mitigation equipment involved is the Control Boards and the Control Room Chilled Water and Control Room Ventilation Systems. The Control Boards will have devices rearranged on them due to the deletion of MC-15. The Control Boards are not adversely affected by rearranging the devices mounted on them. Heat Loads were evaluated not to have an adverse effect on the Control Room Ventilation System. No Technical Specification changes are necessary. A change to UFSAR Section 7.8.3 will be required.

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         - U.S. Nuclear Regulatory Coninaission April 1,'1999 ~
         - Pete 159 of 247 296- . Type: Nuclear Station Modification                                         Unit: 0

Title:

Nuclear Station Modification CN.50463, Replace Security Computer and Access Control System .

Description:

Nuclear Station Modification CN-50463 replaces the Security Computer and Access

                             ' Control System with a fault tolerant / redundant distributed network system. His includes replacing card readers, local door controllers, multiplexer, and the security computer.

This modification adds new equipment including biometic palm readers, magnetic stripe card readers at various plant locations, remote multiplexer units, fault tolerant / redundant , i computers, a networked badging / database system, new central alarm station and secondary alarm station console equipment and modifications to the Personnel Access Portal layout. Additionally, the site assembly function is being automated. Evaluation: There are no unreviewed safety questions associated with this modification. He modification replaces the plant security system with more modern equipment. The new system will perform all the functions of the system that is being replaced. The Security Computer and associated equipment is not an accident initiator as evaluated in the SAR. However, a radiological sabotage event is evaluated in the SAR. Access control, detection of abnormal activities within the protected and/or vital areas, and response capabilitites are not degraded by this modification. Therefore the modification has no degrading effect on the probability of a radiological sabotage event and the probability of an accident evaluated in the SAR is not increased. Failures associated with

                             . the Security Computer System can not cause plant transients or affect the performance of any systems, structures, or components important to safety evaluated in the SAR.

Accordingly, no accidents of a different type than those evaluated in the SAR are created. The systems, structures, and components being modified by this modification includes modification of Category I structures through the attachment of magnetic card readers at various places within the protected area (mostly vital doors). These attachments do not degrade these structures. These functions are passive in nature, providing protection and support for other safety related systems, structures, and components. Additionally, there j are many cables being routed through various fire areas throughout the plant. An ] evaluation has been performed pursuant to Appendix R requirements to assure that those assumptions are not degraded. Since no Appendix R concerns are identified, no new malfunctions of equipment are created. Additionally, Vital area determination and features that define Vital area boundaries continue to meet applicable requirements. The response of the plant to sabotage, fire or other events requiring use of the Safe Shutdown System is not degraded. The response of the plant to a Blackout is not degraded.  ; Accordingly, the possibility of a malfunction different than evaluated in the SAR is not

created. It follows that the probability of a malfunction of equipment important to safety evaluated in the SAR is not increased.

l No accident mitigation equipment is directly affected (other than attachment to nuclear l safety related walls) by the proposed modification. As discussed above, no Appendix R  ! j concerns are created by this modification. Thus, the ability to achieve and maintain hot ! - standby and subsequently achieve cold shutdown is not degraded by this modification. A l security event (sabotage) is assumed to be terminated by use of security force response.  ; Access control, detection of abnormal activities within the protected and/or vital areas, I i f t .1

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    ' April 1,1999                                                                                                l

Pate 160 of 247 - j i

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and response capabilitites are not degraded by this inodification. Assuming the Control l Room is ineffective, plant control and shutdown is effected from the Safe Shutdown I System. No accident mitigation equipment is degraded by this modification. Acrefore, the consequences of accidents previously evaluated in the SAR are not increased. Since the ability to achieve and maintain hot standby and subsequently achieve cold shutdown is not degraded by this modification, the consequences of a malfunction of equipment important to safety evaluated in the SAR are not increased.' Margin of safety is related to the confidence in the fission product barriers. This modification does not affect the fuel, cladding, reactor coolant pressure boundary, or containment including the annulus / annulus ventilation system. Fission product barriers are no more susceptible to sabotage than before this modification. No assumptions made in accident analyses are affected by this modification. No setpoints or limiting safety system settings are affected by this modification. Herefore, the margin of safety defined in the basis to the Tech Specs is not reduced. No Technical Specification changes are required. No UFSAR changes are required. A change to the Security Plan is required pursuant to 10CFR50.54(p). 289 Type: Procedure Unit: 0

Title:

Procedure PT/1/A/4150/13B, Calorimetric Reactor Coolant Flow Measurement

Description:

Procedure PT/1/A/4150/13B, Calorimetric Reactor Coolant Flow Measurement, was revised to incorporate the following changes:

1. Allow the use of the Operator Aid Computer to acquire feedwater venturi differential pressures and feedwater temperatures instead of using the "dLog" data acquisition system. Previous testing has shown that the use of the Operator Aid Computer for data acquisition produces results that differ from the dLog data by less than 0.05%.

2. Allow the use of an alternative computer program to calculate feedwater flow and reactor coolant flow. This program performs the same functions as the "CF Flow" and "NC Flow" programs but combines them into a visual basic program that runs under a Microsoft EXCEL spreadsheet. This program has been benchmarked and determined to correctly calculate reactor coolant flow l Evaluation: here are no unreviewed safety questions as a result of this procedure revision. De test  ;

measures total reactor coolant flow by performing a calorimetric heat balance. De actual surveillance requirement is satisfied by measuring flow as indicated on the Operator Aid Computer. This revision allows all data to be gathered from the Operator Aid Computer rather than other acquisition equipment. Performance of this procedure does not affect any systems or equipment that are involved in the initiation or mitigation of any Chapter i 15 accident. No Technical Specification changes are required. No UFSAR changes are required. i l i I

t i

  ~
    . U.S. Nuclear Regulatory Commission l    . April 1,1999 .

Page 161 of 247 f 52 Type: Procedure Unit: 0 I .

Title:

Procedure CPMB/8800/014 Revision 14, Chemistry Procedure for the detennination of l Steam Generator Tube leak Rate

Description:

Chemistry Procedure CPMB/8800/014 which measures primary to secondary leakage was revised to include a density correction factor which brings consistency between the Chemistry Procedure and the licensing basis dose calculations for which primary to secondary leakage is used as an input. This procedure change addresses issues from NRC Information Notice 97-79. Upon review of the Information Notice and the Chemistry ' Procedure,it was determined that the Chemistry Procedure uses a Reactor Coolant _

                      ! specific activity based on room temperature, biasing the calculated primary to secondary leakage low (i.e. below the actual value for primary to secondary leakage). Without this procedure change, the primary to secondary leak rate could be higher than included in the dose calculations. With this procedure change, there is consistency between the Chemistry Procedure and the dose calculations.

Evaluation: There are no unreviewed safety questions associated with this procedure change. The plant licensing basis will be maintained as a result of this change. No Technical Specification changes are required. No UFSAR Changes are required. 107' Type: Procedure Unit: 0

Title:

Procedure IPMA/3890/01 ' applying to work performed under work order 98054627 concerning removal of two thermometers from thermowells in the Control Room Ventilation System Descriptioni Procedure IPMA/3890/01 applying to work performed under work order 98054627 concerns removal of two thermometers from thermowells in the Control Room

                      . Ventilation System. De two thermometers will removed and test instruments will be installed in their place. The thermometers will be reinstalled afte !be ew,iution. This activity is being performed to support maintenance personnel in trimming the refrigerant charge for Control Room Ventilation System Chiller B.

Evaluation: There are no unreviewed safety questions associated with this procedure. He thermometers provide local indication of chilled water inlet and outlet temperatures for Control Room Ventilation System Chiller B. The thermometers are not a part of the system pressure boundary. No Technical Specification changes are necessary. No UFSAR revisions are required. l L 1 l L':

i-I, U.S. Nuclear Regulatory Ce====8==ta= AprilI,1999 l , ' I' age 162 of 247 p 1 165- Type: Procedure Unit: 0 >

Title:

Procedure IPMA/4974/021, Procedure for Motor On-line Testing

Description:

Procedure IPMA/4974/021, Procedure for Motor On-line Testing, is being created to i l addrm . Fon-line" testing of motors. ."On-line" means that the motor is running during the tesi e When accessible, motor shaft speed will be measured. Shaft speed, along with current data can determine motor loading. Phase currents will be compared to assure all three phases are balanced. Current imbalance may indicate a problem with the motor. Current data will be examined for " sidebands" which may indicate broken rotor bars. Temperature and thermography data will also provide indications of problems with the motor stator or bearings. I Evaluation: The testing which will be donc per this procedure is non-intrusive and will not affect the reliability of the motor being tested. No unreviewed safety questions are created as a result of this procedure. No Technical Specification changes are required. No UFSAR revisions are required.

          -175-     Type: Procedure                                                         Unit: 1

Title:

Procedure IP/1/A/3222/100 Revision 0, Installation, Testing, and Removal of the Low Temperature Over Pressure (LTOP) Computer. l

Description:

Procedure IP/1/A/3222/100 Revision 0 addresser Installation, Testing, and Removt.' of the Low Temperature Over Pressure (LTOP) Computer. The purpose of the procedure is to establish a safe and correct method for installation of additional monitoring instrumentation in the Control Room to assist in monitoring plant conditions and the l potential for challenging the LTOP Setpoints. The system was designed to augment the

                            - information that is currently provided via an Operator Aid Computer Graphic. This l                              system was originally designed to be used during times when the Operator Aid Computer was out of service.

Evaluation: The LTOP instrumentation is designed to ensure that the primary system will not over-pressurize during low temperature operation. The instruments monitor reactor coolant i system temperature and pressure and employ a switch interlock on the Control Board as a l . part ofits design. The procedure and additional monitoring do not change the function or setpoints of the LTOP circuitry. The monitor brings out isolated signals to a laptop

                            ' computer mounted on the Control Board and displays the values graphically. Alarms are l                              provided on the computer to alert the operators when a setpoint has been exceeded The l_                             local setpoints are designed to anticipate the potential for a condition that might challenge j.

the PORVs. There are no postulated accidents that involve failure of the LTOP System. This procedure does not affect the the operation of the current LTOP System. All physical l connections are via isolated circuit cards. *Ihere is no unreviewed safety question

j. associated with this procedure revision. No Technical Specification changes are required.

l No UFSAR changes are required. l

E t b y U.S. Nuclear R,_ " ^- y Ch

      . April 1,1999 .

Pese 163 of 247 . l 176 Type: Procedure Unit: 2 .

                  'Iltle: Procedure IP/2/A/3222/100 Revision 0, Installation Testing. and Removal of the Low Temperature Over Pressure (LTOP) Computer.

Description:

Procedure IP/2/A/3222/100 Revision 0 addresses installation Testing, and Removal of i the law Temperature Over Pressure (L'IDP) Computer. De purpose of the procedure is to establish a safe and correct method for installation of additional monitoring instrumentation in the Control Room to assist in monitoring plant conditions and the potential for challenging the LTOP Setpoints. The system was designed to augment the information that is currently provided via an Operator Aid C ;=*r Graphic. This system was originally designed to be used during times when the Operator Aid Computer was out'of service. Evaluation: The LTOP instrumentation is designed to ensure that the primary system will not over-pressurize during low temperature operation. The instruments monitor reactor coolant system temperature and pressure and employ a switch interlock on the Control Board as a part ofits design. He procedure and additional monitoring do not change the function or setpoints of the LTOP circuitry. De monitor brings out isolated signals to a laptop computer mounted on the Control Board and displays the values graphically. Alarms are

provided on the computer to alert the operarors when a setpoint has been exceeded. The local setpoints are designed to anticipate the potential for a condition that might challenge the PORVs. There are no postulated accidents that involve failure of the LTOP System.

This procedure does not affect the the operation of the current LTOP System. All physical connections are via isolated circuit cards. nere is no unreviewed safety question associated with this procedure revision. No Technical Specification changes are required. No UFSAR changes are required. l 166 Type: Procedure Unit: 0

                 'Iltle: Procedure MP/0/A/4450080 Enclosure 13.2 (Adjustment of Filter Flow)

Description:

his procedure addressed a special instructions data sheet which is associated with procedure MP/0/A/7450480 and Work Order 98047433. These are necessary to ensure that the operable trains of Auxiliary Building Ventilation remain operable while the air flow rate is being adjusted on the train of Auxiliary Building Ventilation in maintenance. He work activity will adjust the air flow on one Train of Auxiliary Building Ventilation to within the allowable limits of the applicable Technical Specification and Performance !- Test PT/0/A/445001C. While this work is being completed on the train in maintenance, the opposite train will continue to operate as designed to provide filtered exhaust for

                          . potentially contaminated areas of the Auxiliary Building during a design basis accident, i Ev'aluation: The Auxiliary Buildies Ventilation System is not an accident initiator. The items l'                           described in the proceduse will not affect the systems ability to perform its design basis

! function. No unreviewed safety questions are created as a result of this procedure. No Technical Specification changes are required. No UFSAR revisions are required. I J

p g 1 U.S. Nuclear Regulatory Comanssion Apdl 1,1999 L Page 164 of 247

  '53      Type: Procedure                                                         Unit: 0

Title:

Procedure MPMAn300/014 Revision 4, Ventilation Equipment Bearing Inspection and Lubrication

Description:

His change to Procedure MPMAU300/014 will allow Heating Ventilation and Air Conditioning Fan and Motor bearings to be lubricated in accordance with the Catawba )

                  . Nuclear Station Lubrication Manual. Some fan pillow block torque values have been changed to " tighten securely" since associated fan bolts are installed using lock washers.

Other fans have had torque values changed from "none required" to ' tighten securely". This equipment will continue to meet Technical Specification performance standards. Evaluation: There is no unreviewed safety question associated with this procedure revision. The revision clarifies how mounting bolts are secured. No new failure modes were identified. No Technical Specification changes are required. No revision to the UFSAR is required. 181 - Type: Procedure Unit: 0

Title:

Procedure MPMAn450/048, Revision 2

Description:

Procedure MPMAn450/048, Revision 2, included adding enclosures 13.2.1,13.2.2, 13.2.3, 13.2.4 and making editorial changes. The enclosures will be utilized to isolate the Annulus Ventilation System pressure boundary from the Unit Vent whenever certain specific maintenance is performed. In order to isolate a specific Annulus Ventilation System Train pressure boundary anytime the ductwork or filter unit is open, the train specific unit vent isolation backdraft damper must be secured in the closed position. This is necessary to ensure Auxiliary Building tornado protection is maintained while the train specific ductwork or filter unit is open for maintenance during any mode of plant I operation.  ! 1 Evaluation: There are no unreviewed safety questions associated with this procedure change. Each l Train of the Annulus Ventilation System will be removed from service when these l procedure enclosures are used.Re Annulus Ventilation System is not an accident initiator and the actions identified by this procedure will not increase the probability of an accident. The redundant train will be available when the enclosures are used. No l Technical Specification changes are required. No UFSAR changes are required.

l- .i l U.S. Nuclear Regulatory Comunission - Apdl 1,1999 Page 165 of 247 ? i 118 Type: Procedure Unit: 0

Title:

Procedure MPMAn450/049 Revision 4

Description:

Procedure MPMAn450019 Revision 4 addresses corrective maintenance on Switchgear Air Handling Units. The procedure was created to provide a station document for guidance on repairing or refurbishing the Control Room Ventilation System Switchgear Air Handling Units. This procedure does not authorize any work that was not previously allowed. The procedure does change who may perform the work and under which program the work may be performed. Also several procedure sections were enhanced and

   .                       clarified. There were no significant technical changes.

1-Evaluation: There are no unreviewed safety questions associated with this procedure revision. No Technical Specification changes are required. No UFSAR changes are required. i 224 Type: Procedure Unit: 0

Title:

Procedure MPMAn450/079 Revision 4 Air Flow Monitoring Manifold and Radiation Monitor Sampling -

Description:

Procedure MPMAn450/079 Revision 4. " Air Flow Monitoring Manifold and Radiation Monitor Sampling", makes several editorial and process related changes as well as one substantive change. The substantive change is associated with tornado watch or warning conditions and opened ductwork access doors in the Auxiliary Building. A procedure enclosure was added for any special engineering instructions needed to perform maintenance work. Use of this enclosure will require a separate evaluation for the 2 specific work being performed. Therefore this change is process related and this .j evaluation does not cover specific uses of the enclosure. The remainder of the procedure I' changes do not impact any nuclear safety related systems, structures, or components. Evaluation: Procedure MPMAS450/079 provides guidance for inspecting and cleaning Air Flow Monitoring Manifolds and Radiation Monitor Sampling Manifolds. In order to properly j clean these components, the ductwork in which they are installed must be opened. This is done through access doors typically located upstream and downstream of the Air Flow I Monitoring Manifold. In several cases opening these access doors creates additional i openings in the Auxiliary Build;ng boundary. Therefore a caution note was added i requiring ductwork access doors opened within the Auxiliary Building during these i maintenance activities to be closed within one hour of a tornado watch or warning. Tius  ; action is consistent with other tornado compensatory measures in place at Catawba. There is no unreviewed safety question associated with this procedure change. No l Technical Specification changes are required. No UFSAR changes are required. i 1

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U.S. Nuclear Regulatory Coh April 1,1999 Pase 166 of 247 264 Type: Procedure Unit: 0

             'I1tle: Procedure MPMAn450/080 Enclosure 13.2 (SpecialInstructions Datasheet)

Description:

Procedure MPMAn450/080 Enclosure 13.2 associated with Work Order 98098745 was necessary to ensure the operable trains of Auxiliary Building Ventilation remain operable while air flow rate is being adjusted on the Auxiliary Building Ventilation train in maintenance. This work activity will adjust the air flow on one train of Auxiliary

Building Ventilation to within the allowable limits of Technical Specification Surveillance 4.7.7.e and performance test PTMA/4450/01C. While this work is being

                     ; completed on the train in maintenance, the opposite train of Auxiliary Building Ventilation will continuue to operate as designed to provide filtered exhaust for E                       potentially contaminated areas of the Auxiliary Building during a design basis accident.

Evaluation: Implemenatation of this special instruction data sheet will adjust air flow of the Auxiliary Building Ventilation train in maintenance for filter replacement. The nunual volume damper at the discharge of this filter unit will be adjusted in small increments so that the allowable flow limits of Tecimical Specification Surveillance requirement 4.7.7.e and the Auxiliary Building Ventilation filter performance test continue to be met. The filter unit in maintenance will be declared irioperable while the manual volume damper at the discharge of the filter unit is being adjusted per these special instructions. The Auxiliary Building Ventilation filter units on the opposite unit will remain operable while adjustments are being rnade to the filter unit in maintenance. Small amounts of adjustment of the manual volume damper at the discharge of the filter unit in maintenance will cause only small increases in air flow. The opposite train will be tested prior to any manual volume damper adjustment to ensure adequate margin exists to ensure flow remains within the allowable limits. No increase in the probability of an accident will be created by this testing procedure. The Auxiliary Building Ventilation system is not an accident initiator. Here are no unreviewed safety questions associated with this procedure change. No Technical Specification changes are necessary. No UFSAR changes are required. , 1 l

p .- U.S. Nuclear Regulatory Connaission April 1,1999 Pase 167 of 247 l.

                                          .^

123 Type: Procedure , Unit: 0 2

    ,           11tle: Procedure MPMAn450/080 Enclosure 13.2 (WO's 98070388 and 98070390)

Description:

Procedure MP/0/An450/080 Enclosure 13.2 (WO's 98070388 and 98070390), issues a specialinstruction to isolate the Unit 2 Annulus Ventilation System Train A pressure

                         ' boundary from the Auxiliary Building while a carbon sample is pulled with Unit 2 in operation. In order to isolate the Annulus Ventilation System boundary with the filter unit open, a backdraft damper will be tied in the closed position. This will enhance the Auxiliary Building tornado protection while the fiker unit is open for removal of the carbon sample.

{ Evaluation: His procedure will remove one train of the Annulus Ventilation System from service. The affected tr sin will be removed from service per station procedures. De Annulus Ventilation System is not an accident initiator and the actions described in this procedure will not increase the probability of an accident. There is no unreviewed safety question associated with this procedure. No changes to the UFSAR are required. No Technical Specification changes are required.- 124 Type: Procedure Unit: 0

Title:

Procedure MP/0/An450/080 Enclosure 13.2 (WO's 98070896-01 and 98070896-02)

Description:

Procedure MP/0/An450/080 Enclosure 13.2 (WO's 98070896-01 and 980708%-02) describes activities that are necessary to ensure that Auxiliary Building Ventilation System Filter Units ABFU-1 A, ABFU-2A, and ABFU-2B will remain operable when airflow to ABFU 1B is reduced. Performance testing performed on 7/29/98 showed that the air flow of ABFU-1B exceeded its upper limit. The Special Instructions data sheet developed for this procedure provides guidance to allow ABFU lB air flow to be throttled in small increments to reduce air flow. The air flow will be throttled by a manual volume damper located at the discharge of the filter unit ABFU-1B. Each time the manual volume damper is throttled the air flow for the IB unit will decrease. This will cause a corresponding increase in air flow for the I A filter unit. A series of incremental adjustments will be continued until the air flow meets acceptance criteria. Evaluation: Unit i Train B of the Auxiliary Building Ventilation System was declared inoperable during performance of this procedure. De Auxiliary Building Ventilation System is not an accident initiator and the actions per this procedure will not increase the probability of an accident. Dere is no unreviewed safety question associated with this procedure. No changes to the UFSAR are required. No Technical Specification changes are required. I l_

I l U.S. Nuclear Regulatory Conunission Apdf 1,1999 Page 168 of 247 242 Type: Procedure Unit: 0

Title:

Procedure MP/0/An4504)82 Enclosure 13.2 (W/O 98064424-31) De cdption: Procedure MP/0/An450/082 Enclosure 13.2 (W/O 980644244!) describes changes associated with corrective maintenance of the Unit 2 Annulus Ventilation System Unit Vent Backdraft Damper 2AVS-D-5. The damper is normally used to isolate the Auxiliary Building from the Unit Vent whenever certain specifiic maintenance is performed which involves breaching the Annulus Ventilation System ductwork pressure boundary. Evaluation: His procedure will remove Train 2A of the Annulus Ventilation System from service i while the system boundary is opened for maintenance activities. This work will be done in Modes 5,6, or No Mode when the system is not required to be operable. A compensatory action is available in the event of tornadoes. The affected train will bc , removed from service per station procedures. The Annulus Ventilation System is not an ) accident initiator and the actions described in this procedure will not increase the probability of an accident. There is no unreviewed safety question associated with this procedure. No changes to the UFSAR are required. No Technical Specification changes are required. 54 Type: Procedure Unit: 0

Title:

Procedure MP/0/An650/063

Description:

This revision to Procedure MP/0/An650/063 will allow a leak repair option for CFLT5551(a Main Feedwater Level Transmitter) first root valve on the high pressure line. He root valve has body to bonnet leakage which will be repaired by use of a clamp and sealant material being injected into the valve body. Evaluation: There is no unreviewed safey question associated with this procedure. His type ofleak repair is considered a routine maintenance practice. The sealant will not enter the piping system. The procedure will not affect the facility as described in the UFS AR. ne probability or consequences of an accident will not be increased. No Technical Specification change is required. No UFSAR changes are required. 1 l

[. U.S. Nuclear Regulatory Co==Indan April 1,1999

    ~ Pane 169 of 247.

t j 319 Type: Procedure Unit: 0

                'I1tle: Procedure OP/0/AM450005 Change 92A

Description:

Procedure OP/0/A/6450005 Change 92A involves changing the valve checklist position for valves 2VI.Y78 and 2VI-Y79 to clear a removal and restoration sheet associated with

                         ' abandonment of the Upper Head Injection System. De checklist position for these l-                        valves was changed to

• Closed". The valves are instrument air system valves.

1-Evaluation: Control switches for valves 2NV476 and 2NV481 (Upper Head Injection fill and isolation valves) were removed when the Upper Head Injection System was abandoned per modification CN-20299. The permanent position of these valves is closed. 'Ihe position of these valves has been maintained by failing the control air to them by

                                                      ~

maintaining valves 2VI-Y78 and 2VI.Y79 closed per a removal and restoration sheet. This procedure revision changes the valve checklist position to " Closed". This will allow the removal and restoration sheet to be cleared. A modification will be generated to revise the flow diagram to show the positions for valves 2VI-Y78 and 2VI-Y79 as closed. This procedure change does not make any physical change to the plant. No unreviewed safety 3 questions are associated with this procedure change. No Technical Specification changes

                                                                                                   ~

are required. No UFSAR changes are required. 1 b

mm  ; 1 1

      ' U.S. Nuclear Regulatory CM Apsil 1,1999 Pane 176 of 247 j

277 Type: Procedure Unit: 0

Title:

Procedure OP/1,2/A/6150001, Changes to support drawing a steam bubble following : Reactor Coolant System Vacuum Refill

                                                                                                             .          )

Description:

Procedure OP/1,2/A/6150001 describes changes to support drawing a steam bubble following Reactor Coolant System Vacuum Refill. The 10CFR50.59 evaluation originally performed by Westinghouse to support allowing a vacuum to be pulled on the Reactor Coolant System and ce~l systems, including components and instrumentation, was issued as Revision 1 in 1991. Revision 2 of this10CFR50.59 safety

  ,                        evaluation was updated to incorporate new analysis to support elimination of the 300 psig -

pressure acceptance criteria and jogging the reactor coolant pumps previously performed following vacuum refill at Catawba. Revision 3 of this 10CFR50.59 safety evaluation has been updated for drawing a steam bubble following vacuum refill at Catawba and contains suggested procedure guidance. Validation of the final procedure ensures that this guidance has been incorporated in the Reactor Coolant System Fill and Vent Procedure, OP/1,2/A/615001 and associated procedure changes. The SAR documents which were reviewed by Duke and Westinghouse for update to this safety evaluation were the UFSAR, Technical Specifications, and Technical Specification Bases. Evaluation: Following Westinghouse suggested procedure guidance, drawing a steam bubble immediately following vacuum refill does not increase the probability or consequences of any accident analyzed, nor create any riew accident or equipment malfunction. An evaluation was performed to demonstrate that long term natural circulation cooling is possible, given the required vacuum conditions specified in this procedure. It has been concluded that with the resulting air present in the U-tubes, heatup and pressurization following a postulated loss of decay heat removal will result in spillover and continuous contact of primary coolant in half or more of the Steam Generator U-tubes in two steam generators, thus ensuring adequate reactor coolant system cooling per Technical Specification 3.4.1.4.1. The coolant used to fill the reactor coolant system including the Pressurizer meets the Technical Specification for boron concentration during Mode 5. i

i. Nitrogen purge of the Steam Generators to minimize oxygen in the system prior to pulling

vacuum has been shown to result in no new failure modes or otherwise increase the probability of malfunction resulting in a loss of decay heat n moval. Westinghouse analysis to support drawing and operating with a steam bubble in the pressurizer during pressurization has shown that the differential temperature in the pressurizer surge and

                         - spray lines will not exceed thermal hydraulic limits already specifed in the design documents and procedures. Oxygen level in the coolant supplied to the pressurizer is controlled to minimize the potential for corrosion of the pressurizer and surge line during vacuum refill operation. Reactor coolant system pressurization rates are reasonable and       J

l. controllable throughout the procedure. Adequate core cooling is assured by maintaining  ;

l pressurizer level and reactor coolant subcooling margin in the hot legs and at core exit, i No fission product barriers are challenged. No Unreviewed Safety Questions are l l

                        ' introduced. No Technical Specification changes are required. No change to the UFSAR is appropriate since the level of detail present in the current sections does not warrant a detailed discussion of the Reactor Coolant System fill and vent process.

r- 1 l l l I U.S. Nuclear Regulatory Conunission April 1,1999 Page 171 of 247 55 Type: Procedure Unit: 1

Title:

Procedure OP/1/A/6150/001 Revision 80 and OP/1/B/6100/010G Revision 51 )

Description:

ne Unit 1 Reactor Vessel O-Ring has been determined to be leaking as demonstrated by receipt of alarms on the Reactor Vessel Leakoff High Temperature Annunciator. Dese procedure changes are required as a result of changing the Reactor Vessel leakage Detection System alignment from the Inner 0-Ring to the Outer O-Ring. , Evalua'ilon: A Safety Review and Unreviewed Safety Question Analysis were performed for changing the alignment of valves INC23 and INC24 to allow operation on the Outer Reactor Vessel O-Ring. It was concluded that the Inner O-Ring and Outer O-Ring aie equivalent by design and have the same licensing basis and that only one O-Ring is in service at any time. The design basis for leakage across either O Ring is limited by the Technical Specifications for Reactor Coolant System leakage. The design basis for the leakage monitoring system is to assist in identifying and maintaining leakage within the limits  ! established by the Technical Specification. He consequence of leakage past the Outer O-Ring was shown to be bounded by the effects of unidentified leakage past any other component in the Reactor Coolant System. Changes are required for UFSAR Section j 5.2.5.2.1, No Technical Specification changes are required. There is no unreviewed j safety question associated with these procedure changes, j i i I 1 I l l

F , U.S. Nuclear Regulatory Comunisslos

     ' April 1,1999
     ' Pase 172 of 247 269     Type: Procedure                                                        Unit: 1 i
                'Iltle: Procedure OP/1/A/62004)06 Revision 34, Enclosure 4.8, Rescating Cold leg Injection Check Valves

Description:

It is the intent of this procedure to exert safety injection pump discharge pressure between the primary and secondary check valves, via the test valve (s) bypassing each secondary ten inch check valve. Based on operation experience gained in the conduct of this rescating procedure and feedback gained through troubleshooting, the following changes were made to the existing procedure enclosure.1) Step 2.15.8 and all other venting steps have been revised to maintain the vent open for 15 minutes following observation of a i steady stream of water,2) Steps 2.15.3 through 2.15.5 and all other steps to prepare the I vent tubing have been moved to be after step 2.7, to allow the procedure to continue while actions are being taken to ensure all vent connections are ready for venting. The primary ten inch cold leg check valve acts as a reactor coolant system pressure boundary and its associated bypass test valves will not be utilized in this procedure. The procedure utilizes elements of the Cold leg Accumulator makeup procedure as well as high point venting which has been reviewed and accepted by the NRC. 'lhe Unit I A 4 Train of Safety Injection will be declared inoperable for the duration of the procedure. Steps are included to ensure that it is easy for the operator in contact with the Control Room to recognize the need and reclose the high point vent in tir Auxiliary Building prior to the time at which swapover to sump recirculation begint 'ite cautions and procedure steps have been reviewed with respect to credit for m ,tator action to ensure they are simple, concise, and possible under blackout conditiotu It is thus ensured that the high point vents can be closed in several minutes and this vent path will not contribute to offsite or control room dose. Evaluation: There are no unreviewed safety questions associated with this procedure revision due to the existence of automatic isolation features as well as ECCS flow margin over the value required by plant Technical Specifications. The fission product barriers of the pellet, clad, reactor coolant pressure boundary, and containment are not affected by this procedure revision. No Technical Specification changes are required. No UFSAR changes are required, i I< i b  ; i l l 1 l I

p e I U.S. Nuclear Regulatory Commission Apdf 1,1999 Pane 173 of 247 f 105 Type: Procedure ~ Unit: 1

                '11tle: Procedure OP/1/A/6200/006, Enclosure 4.8 rescating Cold Leg Injection Check valves and Enclosure 4.9. Effects of Test Header Isolation Valve Failures on System Operability
        . Desedption: These procedures have been written to rescat the (secondary) ten inch, six inch and two inch Reactor Coolant System Cold leg injection Check Valves by applying a differential
                       ' pressure across these valves utilizing the associated Cold leg Accumulator Makeup Valve, Secondary Check Valve Test Valves, the Safety injection Check Valve Test Panel, and Safety Injection Pump Discharge Pressure. The primary ten inch Cold Leg
                        . Check Valve acts as the Reactor Coolant System pressure boundary, and its associated Bypass Test Valves will not be utilized in this procedure. The procedure utilizes elements of the Cold leg Accumulator Makeup procedure, as well as high point venting similar to ECCS venting which has been previously reviewed and accepted by the NRC.
                         'Ihe 1 A Safety Injection Train will be logged as inoperable for the duration of the procedure.

Evaluation: There are no unreviewed safety questions associated with this procedure due to the existence of automatic isolation features as well as ECCS flow margin over the valve required by the Technical Specifications. The fission product barriers of the Fuel Pellet, Fuel Clad, Reactor Coolant Pressure Boundary, and Containment are not affected by this procedure. No Technical Specification changes are required. No UFSAR changes are required. 97 Type: Procedure Unit: 1

Title:

Procedure OP/1/B/6100/010G

Description:

Procedure OP/1/B/6100/0100, Annunciator Response for Panel I AD-6, makes the following changes: 1) adds instructions for operators to monitor reactor vessel flange leak-off temperature if annunciator I AD-6,Fn is in alarm. 2) adds additional guidance to help define " Excessive" leakage for annunciator I ASD-6, Fn 3) makes other editorial changes. Evaluation: There are no unreviewed safety questions associated with this procedure. This procedure change provides additional annunciator response guidance for operators concerning reactor vessel flange leaks.' No Technical Specification changes are required. No UFSAR changes are required. 1 1

 .i.

r _ 1

   - U.S. Nuclear Regulatory Comunission April 1,1999

_ Page 174 of 247 i l, 254 Type: Procedure Unit: 2 -

              'ntie: Procedure OP/2/N6200X)06 Change 388

Description:

Procedure OP/2/N62004)06 Change 38E supports continued operation with documented seat leakage on valve 2NI 185A by closing the associated bonnet vent valve 2NI-488 which otherwise would have been maintained locked open to ensure pressure between the ' seats of 2NI 185A is vented. Evaluation: There is no unreviewed safety question associated with this procedure change. Closing the bonnet vent has been evaluated by Engineering and will not affect the operation of valve 2NI 185A nor will it affect any accident evaluated in the UFSAR. No Technical Specification changes are required. A UFSAR flow drawing will need to be revised. ' 268 Type: Procedure Unit: 2

Title:

Procedure OP/2/N6200/006 Restricted Change 39A

Description:

Procedure OP/2/N6200/006 Restricted Change 39A is a reissue of existing change 38fi I Change 38E was a restricted change (which expired) to support a major revision to the l procedure. Change 39A supports continued operation with documented leakage on valve 1 2NI-185A by closing the bonnet valve 2NI-488, which otherwise would have been maintained locked open, to ensure pressure between the seats of 2NI 185A is vented. The l UFSAR flow diagram does include this valve as open. Engineering has evalauted the safe operation of valve 2 nil 85A with the bonnet vent closed. Evaluation: There are no unreviewed safety questions associated with this procedure change. This activity only closes the bonnet vent which has been evaluate not to affect the operation of 2NI-185A nor to affect any accident evaluated in the SAR. No Technical Specification changes are required. N i UFSAR changes are required. i l l i t'

I 'r l l [ l U.S. Nuclear Regulatory Commission i April 1,1999 l Page 175 of 247 ' l: [_ 270 Type: Procedure Unit: 2 L

Title:

Procedure OP/2/A/6200/006 Revision 39. Enclosure 4.8, Rescating Cold leg injection Check Valves

Description:

It is the intent of this procedure to exert safety injection pump discharge pressure between j the primary and secondary check valves, via the test valve (s) bypassing each secondary ) ten inch check valve. Based on operation experience gained in the conduct of this  ! L rescating procedure and feedback gained through troubleshooting, the following changes were made to the existing pmcedure enclosure.1) Step 2.15.8 and all other venting steps have been revised to maintain the vent open for 15 minutes following observation of a steady stream of water. 2) Steps 2.15.3 through 2.15.5 and all other steps to prepare the vent tubing have been moved to be after step 2.7, to allow the procedure to continue while

                      . actions are being taken to ensure all vent connections are ready for venting. The primary L

ten inch cold leg check valve acts as a reactor coolant system pressure boundary and its associated bypass test valves will not be utilized in this procedure. The procedure utilizes elements of the Cold leg Accumulator makeup procedure as well as high point venting which has been reviewed and accepted by the NRC. The Unit I A Train of Safety i Injection will be declared inoperable for the duration of the procedure. Steps are included to ensure that it is easy for the operator in contact with the Control Room to recognize the need and reclose the high point vent in the Auxiliary Building prior to the time at which sw pover to sump recirculation begins. The cautions and procedure steps have been reviewed with respect to credit for operator action to ensure they are simple, concise, and i possible under blackout conditi<,ns, it is thus ensured that the high point vents can be i closed in several minutes and this vent path will not contribute to offsite or control room j dose. l Evaluation: There are no unreviewed safety questions associated with this procedure revision due to  ! the existence of automatic isolation features as well as ECCS flow margin over the value  ! required by plant Technical Specifications. No Technical Specification changes are , I required. No UFSAR changes are required. i

m i. h U.S. Nuclear Regulatory th*Ia= L . Apdf 1,1999 l Page 176 of 247 286 Type: Procedure - Unit: 2-

                     'Iltle: Procedure OP/2/N6200006 Revision 40A'

Description:

Procedure OP/2/N6200/006 (Safety injection System) Revision 40A supports continued  ! operation with documented seal leakage on valve 2NI 185A by closing the associated ' bonnet vent valve 2NI-488 which otherwise would Lve been maintained locked open to ensure pressure between the seats of 2NI-185A is vented. The UFSAR flow diagram does include this bonnet vent as open. An Engineering evaluation has documented the safe operation of valve 2NI-185A with the bonnet vent closed. ! ' Evaluation: 'Ihere is no unreviewed safety question associated with this procedure change. This

                             ' activity closes the bonnet vent which has been evaluated not to affect the operation of valve 2NI-185A nor does it have the potential to affect any accident evaluated in the
                             . UFSAR. No Technical Specification changes are required. No UFSAR changes are required.

106 - Type: Procedure Uniti 2

Title:

Procedure OP/2/A/6200/006, Enclosure 4.8 rescating Cold Leg Injection Check valves and Enclosure 4.9. Effects of Test Header Isolation Valve Failures on System Operability

Description:

These procedures have been written to rescat the (secondary) ten inch, six inch and two inch Reactor Coolant System Cold Leg injection Check Valves by applying a differential l pressure across these valves utilizing the associated Cold Leg Accumulator Makeup Valve, Secondary Check Valve, Test Valves, the Safety Injection Check Valve Test Panel, and Safety injection Pump Discharge Pressure. The primary ten inch Cold 1xg Check Valve acts as the Reactor Coolant System pressure boundary, and its associated Bypass Test Valves will not be utilized in this procedure. The procedure utilizes

elements of the Cold leg Accumulator Makeup procedure, as well as high point venting similar to ECCS venting which has been previously reviewed and accepted by the NRC.

                                                              ~

The 2A Safety injection Train will be logged as inoperable for the duration of the procedure. Evaluation: There are no unreviewed safety questions associated with this procedure due to the

                             . existence of automatic isolation features as well as ECCS flow margin over the valve required by the Technical Specifications. The fission product barriers of the Fuel Pellet.

Fuel Clad, Reactor Coolant Pressure Boundary, and Containment are not affected by this procedure. No Technical Specification changes are required. No UFSAR changes are required. l L

r"-- I U.S. Nuclear Regulatory Commission

      . Apdf 1,1999 Page 177 of 247 56    Type: Procedure                                                        Unit: 0

Title:

Procedure PT/0/A/4150/12A Revision 10 =

Description:

Procedure PT/0/A/4150/12A Revision 10, Isothennal Temperature Coefficient of Reactivity Measurement, changes the analysis mediod from the endpoint method to the 1 slope method. An Advance Digital Reactivity Computer (ADRC) is used to calculate 1 Isothermal Temperature Coefficient (lTC). Enclosures 13.2 (ITC Measurement Data) and 13.4 (Boron Concentration log) are deleted. Data from ADRC provides equivalent  ; data to that previously recorded on Enclosure 13.2 and Enclosure 13.4 is relocated to procedure PTM/A/4150/001, Controlling Procedure for Startup Physics Testing. This revision does not require operation of plant equipment outside of design specifications. Plant configuration and operation are unchanged due to this revision. The methods used to change Reactor Coolant Temperature remain the same although the required magnitude has changed. The required temperature change decreased from at least 5 degrees F. to at least 1.1 degrees F. with an additional error analysis performed to ensure acceptable measurement results. Also the acceptable difference between heatup and cooldown results has been increased from 0.4 to 1.0 pcm/ degree F. This does not constitute any change to the plant configuration or operation. The ARDC performs data analysis using I the slope method as defined in ANSI /ANS 19.6.1 1985. The change in analysis methods l and reduction of the reactor coolant system temperature change do not increase the j probability of any previously evaluated accidents. The remaining changes are editorial j and do not affect plant operation. Evaluation: There are no unreviewed safety questions as a result of this procedure revision. The revision will not result in plant equipment being operated outside design specifications. This revision in methodology requires no changes to plant configuration and operation. The techniques used to initiate reactor coolant temperature decrease are unchanged. liowever the magnitude of the temperature change decreased from at least 5 degrees F to at least 1.1 degrees F with an error analysis performed to ensure acceptable measurement results. Also, the acceptable difference between heatup and cooldown results does not constitute a plant change. The ARDC uses the slope method to calculate ITC as defined in ANSI /ANS 19.6.1-1985. The change in analysis methods and reduction of reactor coolant temperature change do not increase the probability of any previously evaluated accidents. No Technical Specification changes are required. UFSAR Section 14.3.2.2.2 will be revised. I l l a

U.S. Nuclear Regulatory Comunisalon

 - Apdf 1,1999 Pase 178 of 247 266    Type: Procedure Unit: 0                   ;
            'Iltie: Procedure PTMA/4150/19B, Reactor Coolant System Dilution Following Refueling.             l Revision 1 -

Description:

Procedure FD0/A/4150/19B Revision 1 eliminates the ability to perform dilution in Mode 5. Two concerns have been identified with performing dilution in Mode 5. These are:

1. the ability of the residual heat removal system to remove pump heat from the four reactor coolant pumps in addition to the low decay heat from the core.

2. the consequences of an uncontrolled rod withdrawal accident without containment integrity.

An analysis has shown that the reactor will remain suberitical with an uncontrolled rod withdrawal event if only Shutdown Banks A and B are withdrawn for the dilution, thus resolving the concern over consequences. However, since dilution will not be performed in Mode 5 due to the first concern (pump heat), no effort will be made to explicitly address the second). Revision 1 also changes the control rod configuration for the dilution. Revision 0 had all Shutdown Banks withdrawn. Revision I has only Shutdown Banks A and B withdrawn in Mode 4. This change is to ensure that the reaactor remains suberitical in the event of an uncontrolled rod withdrawal accident in Mode 4. UFSAR analysis of uncontrolled rod withdrawal (Section 15.4.1) states that the limiting case is with the reactor starting at critical conditions at noninal no-load temperature and pressure. This can only be assured if the reactor is prevented from reaching criticality  ; I during an event in Mode 4 or Mode 5. With lower coolant flow and pressure allowed in  ! Modes 4 and 1,, the consequences of an uncontrolled rod withdrawal event that results in criticality could be worse than the limiting care given in the UFSAR. Shutdown Banks A and B provide sufficient trippable rod worth for operators to respond to an unexpected reactivity change. { Evaluation: Since it has been shown that the reactor would remain subcritical if an uncontrolled rod I

                  ' withdrawal occurs in Mode 4 with only Shutdown Banks A and B withdrawn, no j

unreviewed safety question exists for revision I of Procedure l'TMA/415&l9B. No  ; Technical Specification changes are required. No UFSAR changes are required. 1 l  ! l I l j l i l l

U.S.Nucliar Regulatory Commission April 1,1999 Page 179 of 247 r

      %      Type: Procedure                                                         Unit: 0

Title:

Procedure PT/0/A/4200/23, Restricted Change {

Description:

Restricted Change to Procedure PT/0/A/4200/23, Pressurizer Continuous Spray Flow

• Valve Setup; makes the following changes:

1) requires the Unit to be in Mode 2 instead of Mode 3.

2) Starts the data gathering process with throttle valves NC-28 and NC-30 fully open instead of fully closed.

3) Has data gathered in one eighth turn increments until the valve is one quarter turn from closed.

During the previous conduct of this test on 1/4/98, Va've INC-28 was fully closed and the valve did not come offits seat even though the sten rose a full one and one quarter turn. In order to allow the procedure to be performed in I.fode 2, after maintenance is done to free up the valve, the procedure was revised to allow the data to be gathered from a fully open position, closing the valve until it is one quarter twn from closed. From past experience and currently throttled position of valve INC-30 closing the valve, it was determined that the valves should not need to be throttled to less than one quarter turn open. Evaluation: There are no unreviewed safety questions associated with this procedure. These valves (Pressurizer Spray Throttle Valves) are not accident initiators. Performing the throttle

                    - valve setting procedure in Mode 2 instead of Mode 3 does not increase the probability of an accident. The major purpose of this procedure is to set the minimum flow needed to maintain the pressurizer spray line within 125 dergrees of pressurizer steam temperature to decrease the magnitude of the thermal transients on the spray line and nozzle.

Therefore the critical parameter being maintained is temperature. The spray line has its own temperature monitoring device. No Technical Specification changes are required. 3 No UFSAR changes are required. l l

                                                                                                                  )

i m.

I

        , U.S. Nuclear Regulatory Ch April 1,1999 Page 180 of 247 -                                                                                                 )

221 ' Type: Procedure Unit: 0

Title:

Procedure PT/0/N4450/001B, Revision 13, Control Room Area Outside Air Pressure FilterTrains Performance Test

Description:

Procedure PT/0/N4450/001B, Revision 13," Control Room Area Outside Air Pressure Filter Trains Performance Test", describes testing to ensure the operational readiness of

                           . the Control Room Ventilation System Pressurizing Filter Trains. These tests include (1) verfiying that the Pressurizing Filter Trains maintain the proper in place penetration and    i bypass leakage. (2) verfiying that the Pressurizing Filter Trains pass the required system flowrate (3) verflying that the Pressurizing Filter Trains maintain the proper pressure drop. These tests ensure compliance with Technical Specification Surveillance Requirements 4.7.6.c.I,4.7.6.c.3,4.7.6.e.1,4.7.6.f 4.7.6.g.

Evaluation: The Control Room Area Ventilation System is a safety related system whose purpose is to (1) ensure that the control room remains habitable for Operations personnel during and following all credible accident conditions, and (2) ensure that the ambient air temperature does not exceed the allowable temperature for continuous-duty rating for the equipment and instrumentation cooled by the system. In order to ensure that the system can perform these design basis functions the Technical Specifications require periodic monitoring and testing of critical parameters. Procedure FT/0/N4450/001B is a station test procedure that performs the tests which verify several of these Technical Specification surveillances. He purpose of PT/0/N4450/001B is to ensure the operational readiness of the Controf Area Ventilation System, by verifying the surveillance requirements of Technical Specification 4.7.6.c.1,4.7.6.c.3,4.7.6.e.l. 4.7.6.f. 4.7.6.g. These surveillances have an eighteen month frequency but may also be required if certain maintenance is performed on the system filter units or if the filters were exposed to degrading conditions such as painting, chemical release or smoke. The procedure does not alter the operation of the Control Room Ventilation System and all data taken is taken with the system in its normal mode of operation. None of the design bases for the Control Room Ventilation System are affected by the performance of this procedure. Nothing is done that would significantly affect the flow rate through the filter unit, tierefore the residence time, filter efficiencies, and the filter differential pressure will not be affected. The ability of tim system to detect chlorine, smoke or radiation is not affected by this procedure. De procedure does not affect any Control Room Chilled Water System components. The procedure does not require connecting test equipment to an operating filter train. This is

i. acceptable because only one connection point is located such that unfiltered air could be introduced into the system and the procedure contains adequate guidance to seal this connection to prevent air inleakage. All other test connection points are either on the suction side of the filter unit or are in positively pressurized ductwork on the discharge side of the pressurizing fan. This precludes the possibility ofincreasing the amount of unfiltered in leakage into the system. Therefore the design basis and dose analysis are not -

affected by the performance of this test. There are no unreviewed safety questions associated with this procedure revision. No Technical Specification changes are required. No UFSAR revisions are required. 1 i

I - l l i l U.S. Nuclear Regulatory Commission l Apdf 1,1999 j Page 181 of 247 j 109 Type: Procedure Unit: 0

            'Iltle: Procedure PT/0/N4450/001C, Revision 12, Auxiliary Building Filtered Exhaust Filter Unit Perrformance Test                                                                       i i

Description:

Procedure PT/0/N4450/001C, Auxiliary Building Filtered Exhaust Filter Unit l Performance Test Revision 12, ensures that Technical Specification Surveillance Requirements 4.7.7.b.1, 4.7.7.b.3, 4.7.7.d.I,4.7.7.e, and 4.7.7.f are met. These j surveillance requirements have an 18 month frequency. Until recently it was believed that l PT/0/N4450/001C, Auxiliary Building Filtered Exhaust Filter Train Performance Test, ' provided the data to satisfy these surveillance requirements. However, based on strict compliance with the wording of the surveillance requirements it was determined that current testing practice was not in adherence with the Technical Specification , surveillance. This surveillance requirement was established to ensure that the Auxiliary l Building Ventilation system would be able to provide adequate exhaust flow throughout i the Auxiliary Building as the filters became dirty and the filter unit pressure drop increased. While there are processes in place to ensure technical compliance with the Technical Specification, strict compliance with the existing wording was not met. Changes added by this procedure revision will ensure that the pressure drop across the moisture separator is measured and added to the total pressure drop across other j components (prefilter, upstream and downstream HEPA filters, and the carbon adsorber) j within the filter unit. Evaluation: Here are no unreviewed safety questions associated with the proposed changes to performance test FT/0/N4450/001C. He Auxiliary Building Ventilation System is not an accident initiator. This procedure does not affect the ability of the Auxiliary Building Ventilation System to perform its design basis functions. No Technical Specification changes are required. No UFSAR changes are required.

• I i

r: L F' 1 U.S. Nuclear Regulatory Conunission l Apdf 1,1998 i Page 182 of 247_ . l 108 Type: Procedure Unit: 0 Tide: Procedure FD0/A/4450004A, Revision 39, Auxiliary Building Ventilation System l Performance Test l l

Description:

Procedure PT/0/A/4450/004A, Auxiliary Building Ventilation System Performance Test, ensures that Technical Specification Surveillance Requirements 4.7.7.d.3, 4.7.7.d.4, and 4.7.7.d.5 are met. These surveillance requirements have am eighteen month frequency. The following items are verified per this procedure:

1) That the Auxiliary Building Ventilation System starts on a safety injection test signal and directs its exhaust air flow through the HEPA filters and the activated carbon adsorbers.

2) That the Auxiliary Building Ventilation System maintains ECCS Pump

! Rooms at a negative pressure relative to adjacent areas.

3) That the Auxiliary Building Ventilation System filter cooling bypass valves can be manually opened.

4) That the Auxiliary Building Ventilation System Heaters dissipate 40 +/- 4 KW at a nominal voltage of 600 VAC.

Evaluation: There are no unreviewed safety questions associated with this procedure. This test does not prevent the Auxiliary Building Ventilation System from being able to perform its design basis functions. No Technical Specification changes are necessary. No UFSAR revisions are required. I i l I

m U.S. Nuclear Regulatory Commission April 1,1999 Page 183 of 247 220 Type: Procedure Unit: 0

Title:

Procedure PT/0/A/4450/008 Revision 24, Control Room Area Ventilation System Performance Test

Description:

PT/0/A/4450/008 Revision 24,

• Control Room Area Ventilation System Performance Test" ensures the operational readiness of the Control Room Ventilation System by verifying several system parameters. Rese include:

1. Verifying tu t the pressurizing filter train preheaters dissipate the required wattage

2. Verifying that the outside air intake isolation valves close within the allowable time upon receiving a high chlorine test signal

3. Venfying that the system maintains the control room at an acceptable positive pressure relative to adjacent areas during normal system operation

4. Verifying that upon receiving a high smoke concentration test signal that an alarm is received in the control room

5. Verifying that an acceptable differential pressure exists across the smoke detector sensing tubes Evaluation: The Control Room Area Ventilation System is a safety related system whose purpose is to (1) ensure that the control room remains habitable for operations personnel during and following all credible accident conditions, and (2) ensure that the ambient air temperature does not exceed the allowable temperature for continuous-duty rating for the equipment and instrumentation cooled by the system. In order to ensure that the system can perform these design basis functions the Technical Specifications require periodic monitoring and testing of critical parameters. Procedure PT/0/A/4450/008 is the station test procedure ,

that perfonns the tests which verify several of these Technical Specification surveillances. Irr/0/A/4450/008 ensures the operational readiness of the Control Area Ventilation System by verifying the surveillance requirements of Technical Specification 4.7.6.e.2 (smoke portion only),4.7.6.e.3,4.7.6.e.4 and 4.7.6.e.5. These surveillances have an 18 month frequency. In addition to the above surveillance requirements this procedure also verifies that an acceptable differential pressure exists across the smoke detector sensing tubes which is an NFPA-72E concern. This procedure does not alter the operation of the Control Area Ventilation System except for closing each intake separately in enclosures 13.1 and 13.2. Since the control room can be positively pressurized with only one intake open, closing an intake does not affect the design basis of the system. Data is taken with the system in its normal mode of operation. None of the design bases for the Control Area Ventilation System will be affected by the performance of this procedure. Nothing is done that will si Enificantly change the flow rate through the filter unit so the residence time, the filter efficiencies, and the filter differential pressure will not be affected. IIaving test equipment connected to the system ductwork while the system is in service is acceptable because all test connection points are either on the suction side of the filter unit or are in positively pressurized ductwork on the discharge side of the pressurizing fan. His precludes the possibility of increasing the amount of unfiltered inleakage into the system and, therefore, the design basis is not affected by the performance of this test. No Unreviewed Safety Questions are introduced as a result of this procedure revision. l No Technical Specification changes are required. No UFSAR changes are required.  ;

                                                                                                                      'l i

4 l l

r: U.S. Nuclear Regulatory Conunission April 1,1999 ) Page 184 of 247 l l l I l 4 I l i I l l I

l U.S. Nuclear Regulatory Comunission Apdl I,1999

 . Page 185 of 247 1

l 104 Type: Procedure Unit: 0

Title:

Procedure PT/1,2/A/4400/09, Cooling Water Flow Monitoring for Asiatic Clams and Mussels Quarterly Test. l

Description:

Procedure PT/1.2/A/4400/09, Cooling Water Flow Monitoring for Asiatic Clams and l Mussels Quarterly Test. His procedure change is to reduce Component Cooling System Heat Exchanger acceptance criteria to a resistance factor of 425 from the previous value of 456. He Nuclear "ervice Water System is the ultimate heat sink for various nuclear safety related heat loads during normal operation and design basis events. ' The Nuclear

Service Water System rupports Emergency Core Heat Removal operation by providing cooling to the Component Cooling System via the Component Cooling Heat Exchangers and also to the Diesel Generators via the Diesel Generator Engine Jacket Water Cooler System Heat Exchangers. Other nuclear safety related loads include Containment Spray Heat Exchanger and Control Room Chiller Condenser. He Nuclear Service Water System also provides assured makeup to Component Cooling, Spent Fuel Pool, Auxiliary Feedwater Supply and the Containment Seal Water injection System.

Procedures PT/1/A/4400/09 and PT/2/A/4400/09 measure the flow and pressure drop across the Component Cooling System, Diesel Generator Engine Jacket Water Cooling System, and Containment Spray System Heat Exchangers. This test is performed to provide information used for heat exchanger fouling trending and also to alert the station to any severe blockage that may occur in the heat exchangers. The test acceptance Hteria is in the form of a "Cv" flow coefficient. or " resistance factor" This coefficient l arctically links the flowrate and pressure drop over the range of fully developed t , bulent flow, i Due to recent changes in tube materials in the 1 A and 2B Component Cooling Heat l Exchangers,. these components have experienced rises in differential pressure over very , short periods of time (5 to 6 months) that have residted in the heat exchangers approaching the acceptance criteria for the flow monitoring test. As a result of this problem, these heat exchangers have had to be cleaned at an increased frequency. The original acceptance criteria for the Component Cooling Heat Exchangers has been determined to be overly conservative. The acceptance criteria is based on the expected i pressure drop across the heat exchanger at a fouling factor of 0.002. Engineering ! calculation CNC 1223.24-004)0$0, " Component Cooling Heat Exchanger Tubeside Differential Pressure Limitations" evaluated the fouling of the Component Cooling Heat Exchanger tubes and its effect on the heat transfer required for design basis operation. This . calculation documents that the heat exchanger resistance factor can be lowered to a value l of 425 based on a fouling factor of 0.003 and still ensure that the heat transfer required for design basis operation is assured. Therefore the acceptance criteria for the j Component Cooling Heat Exchangers in procedures PT/1/A/4400/09 and PT/2/A/4400/09 can be reduced from 456 to 425. This acceptance criteria is conservative based on a heat exchanger with 0.0030 fouling, j The fouling limit established to support Catawba Nuclear Station accident analysis is l 1

y L ' i l I l U.S. Nuclear Regulatory Commission April 1,1999.' Pase 186 of 247-0.0032. Flow balance test data indicate that changes in Component Cooling Heat Exchanger differential pressure do not significantly impact the flow balance. He

1. increased Component Cooling Heat Exchanger differential pressure associated with this l: change is approximately 0.66 psi at flow balance flow of 5000 gpm.

Evaluation: here are no unreviewed safety question associated with the changes to these procedures.

                     ' He performance of systems supported by the Nuclear Service Water System will not be degraded by this change since the requimd heat transfer rates are maintained. De change will not impact the Nuclear Service Water System flow balance and is not considered to l

impact the performance of the Component Cooling Heat Exchangers. No other equipment

                     . used for any phase of either power generation or conversion or transmission, normal shutdown cooling, fuel handling, or radwaste' treatment is affected.

l Nuclear Service Water System response to an accident is not degraded as the required heat transfer rates assumed in licensing and design calculations are still maintained. No other system used to mitigate any accident is degraded based on the ability of the Nuclear l Service Water System and Component Cooling Systems to maintain their design basis. De frequency of challenges to equipment provided to mitigate accidents is not increased j based a the acceptance criteria assuring that design basis heat transfer is maintained. I ne probability of a malfunction of equipment important to safety evaluated in the UFSAR is not increased since the new acceptance criteria ensures design basis operation of the affected systems. Since the acceptance criteria takes flow and pressure drop into account, there are no new failure modes created. Common cause failure modes are not created for the same reason no new failure modes are created, ne consequences of a malfunction of equipment important to safety evaluated in the UFSAR are not increased based on the acceptance criteria ensuring that design basis heat transfer requirements are still met for this equipment. Likewise the possibility of a malfunction of a different type than any evaluated in the UFSAR is not created. Since there is no effect on a fission product barrier or source tenn, the change will not increase the consequences of an accident evaluated in the UFSAR. Here are no changes to any design limit or setpoint. No fission product barrier is affected. His modification will not change the flow, temperature, or pressure of cooling water supplied to any component. No control, instrument function, or performance of any structure, system,or component is degraded. Herefore the margin of safety as defined in the bases to the Technical Specification has not been reduced. No Technical Specification changes are required. No changes to the UFSAR are required. l l l t l i

7 l 4y s j I L [ U.S. Noelear Regulatory Co==I= ton Apsil 1,1999 -

    - Page 187 of 247                                                                                                   !

! . 98 ' _ Type: Procedure Unit: 0 T1tle: ProcedureIT/1,2/A/4600/02A

Description:

Procedure PT/1,2/A/4600/02A, Units 1 and 2 Mode and Pre-Mode Checklists, changes , the normal pre-accident operating temperature limits utilized in mode and pre-mode l periodic surveillance procedures.

                       . Calculation CNC-1210.04-00-0054, Rev. 2 determined normal pre-accident total loop uncertainties for the upper and lower containment temperature detectors. . These total loop uncertainties are used to calculate a reduced error depending upon the number of               ;

operating air handling units and available temperature (RTD) loops in service. 'Ihe RTDs I and their associated alarms are utilized by the operators to verify and maintain the bulk  ; average containment air temperature within Technical Specification 3.6.1.5 limits. These containment temperature loops are non safety related, however, they are utilized to verify the containment temperature parameter associated with plant Safety Analyses. , The surveillance procedures currently utilize error uncertainties determined per CNC- 1 1210.04-00-0054, Rev. 2 to verify average containment temperature per Technical i Specification 3.6.1.5. These error uncertainties are no longer required for the subject l Mode Surveillance Procedures because several nuclear safety analyses were performed to ;l incorporate these uncertainties. ll Evaluation: The limitations on containment average air temperature ensure that:

                                                                                                                      'l (1) The containment air mass is limited to an initial mass sufficiently low to prevent exceeding the design pressure during a LOCA condition.

(2) The ambient air temperature does not exceed the temperature allowable for the  ! continuous duty rating specified for equipment and instrumentation located in the i containment.- I The containment pressure transient is sensitive to the initially contained air mass during a I LOCA. The containment air mass increases with decreasing temperature. The lower Technical Specification 3.6.1.5 temperature limit of 100 degrees F for the lower compartment and 75 degrees F (60 degrees F when in Mode 2,3, or 4) for the upper compartment limits the peak pressure to 14.7 psig which is less than the containment  : design pressure of 15 psig. The upper temperature limit of 120 degrees F for the lower l compartment and 100 degrees F for the upper compartment influences the peak accident j temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and anticipated operating conditions. l Procedures require error analyses to be performed for indications which are relied upon to l

                      - maintain / verify significant parameters associated with nuclear safety analyses. Both the upper and lower containment temperature limits have been analyzed using limits which incorporate the error uncertainties derived in CNC-1210.04-00-0054, Rev. 2.
                        'Ihe peak containment pressure transient was analyzed utilizing 95 degrees F for the lower containment and 70 degrees F for the upper cortainment. It was demonstrated that the peak containment pressure is not exc-ial for ice masses well below the current Technical Specification 3.6.5.1 minimum values for Catawba Nuclear Station Units I and

2. 'Iherefore application of error uncertainties to the lower temperature limits is not necessary.

                        'Ihe upper temperature limit for lower containment is not utilized in Safety Analyses. The peak lower containment temperature and pressure transient was analyzed utilizing 125 and 135 degrees F. Resdts indicated that there was no significant impact to this analysis.

L

U.S. Nuclear Regulatory Conunission

   - April 1,1999 :
   - Page 188 of 247.

l Therefore application of error uncertainties to the upper temperature limits is not necessary. l Removal of the containment temperature error uncertainties from the subject mode surveillance procedures is acceptable because they have been incorporated into the safety analyses . These changes will allow the plant to operate per the temperature limits of Technical Specification 3.6.1.5. There are no~ unreviewed safety questions associated with this procedure. No Technical Specification changes are required. No UFSAR changes are required. 1 I 4 I 1 I i l l [  ! c )

[S

  ' U.S. Nuclear Regulatory Commission April 1,1999 ;

Pase 189 of 247 99 ' Type: Procedure Unit: 0

Title:

Procedure PT/1,2/A/4600/02B '

Description:

Procedure PT/1,2/A/4600/02B, Units I and 2 Mode and Pre-Mode Checklists, changes the normal pre-accident operating temperature limits utilized in mode and pre-mode periodic surveillance procedures. Calculation CNC-1210.04-00-0054, Rev. 2 determined normal pre-accident total loop

                    . uncertainties for the upper and lower containment temperature detectors. These total loop uncertainties are used to calculate a reduced error depending upon the number of operating air handling units and available temperature (RTD) loops in service. The RTDs and their associated alarms are utilized by the operators to verify and maintain the bulk average containment' air temperature within Technical Specification 3.6.1.5 limits. Rese containment temperature loops are non safety related, however, they are utilized to verify the containment temperature parameter associated with plant Safety Analyses.

The surveillance procedures currently utilize error uncertainties determined per CNC-1210.04-00-0054, Rev. 2 to verify average containment temperature per Technical Specification 3.6.1.5. These error uncertainties are no longer required for the subject Mode Surveillance Procedures because several nuclear safety analyses were performed to incorporate these uncertainties. Evaluation: The limitations on containment average air temperature ensure that:

                    . (1) The containment air mass is limited to an initial mass sufficiently low to prevent exceeding the design pressure during a LOCA condition.

(2) The ambient air temperature does not exceed the temperature allowable for the continuous duty rating specified for equipment and instrumentation located in the containment. The containment pressure transient is sensitive to the initially contained air mass during a LOCA. The containment air mass increases with decreasing temperature. The lower Technical Specification 3.6.1.5 temperature limit of 100 degrees F for the lower compartment and 75 degrees F (60 degrees F when in Mode 2,3, or 4) for the upper compartment limits the peak pressure to 14.7 psig which is less than the containment design pressure of 15 psig. The upper temperature limit of 120 degrees F for the lower compartment and 100 degrees F for the upper compartment influences the peak accident temperature slightly during a LOCA; houver, this limit is based primarily upon equipment protection and anticipated operating conditions. Procedures require error analyses to be performed for indications which are relied upon to maintain / verify significant parameters associated with nuclear safety analyses. Both the upper and lower containment temperature limits have been analyzed using limits which incorporate the error uncertainties derived .in CNC-1210.04 00-0054, Rev. 2. The peak containment pressure transient was analyzed utilizing 95 degrees F for the lower containment and 70 degrees F for the upper containment. It was demonstrated that the _ peak containment pressure is not exceeded for ice masses well below the current Technical Specification 3.6.5.1 minimum values for Catawba Nuclear Station Units I and

2. Herefore application of error uncertainties to the lower temperature I;mits is not necessary.

De upper temperature limit for lower containment is not utilized in Safety Analyses. The peak lower containment temperature and pressure transient was analyred utilizing 125 and 135 degrees F. Results indicated that there was no significant impact to this analysis. i 1

                                                                                                                    =

l l

 . U.S. Nuclear Regulatory Coenmission Apdf 1,1999 Page 190 of 241 1

Therefore application of error uncertainties to the upper temperature limits is not necessary. Removal of the containment temperature error uncertainties from the subject mode surveillance procedures is acceptable because they have been incorporated into the safety analyses . These changes will allow the plant to operate per the temperature limits of J Technical Specification 3.6.1.5. There are no unreviewed safety questions associated I with this procedure. No Technical Specification changes are required. No UFSAR changes are required. i l l l

[ l l U.S. Nuclear Regulatory Comunission April 1,1999 Page 191 of 247 i i 100 Type: Procedure Unit: 0

Title:

Procedure PT/1,2/A/4600/02C

Description:

Procedure PT/1,2/A/4600/02C, Units 1 and 2 Mode and Pre-Mode Checklists. changes the normai pre-accident operating temperature limits utilized in mode and pre-mode I periodic surveillance procedures. I Calculation CNC-1210.04-00-0054, Rev. 2 determined normal pre-accident total loop uncertainties for the upper and lower containment temperature detectors. These total loop uncertainties are used to calculate a reduced error depending upon the number of operating air handling units and available temperature (RTD) loops in service. The RTDs I and their associated alarms are utilized by the operators to verify and maintain the bulk average containment air temperature within Technical Specification 3.6.1.5 limits. These containment temperature loops are non safety related, however, they.are utilized to verify the containment temperature parameter associated with plant Safety Analyses. The surveillance procedures currently utilize error uncertainties determined per CNC-1210.04-00-0054, Rev. 2 to verify average containment temperature per Technical Specification 3.6.1.5. These error uncertainties are no longer required for the subject Mode Surveillance Procedures because several nuclear safety analyses were performed to incorporate these uncertainties. Evaluation: The limitations on containment average air temperature ensure that: (1) The containment air mass is limited to an initial mass sufficiently low to prevent ) exceeding the design pressure during a LOCA condition.  ! (2) The ambient air temperature does not exceed tS temperature allowable for the continuous duty rating specified for equipment and ixtumentation located in the containment. The containment pressure transient is sensitive to the initially contained air mass during a LOCA. The containment air mass increases with decreasing temperature. The lower Technical Specification 3.6.1.5 temperature limit of 100 degrees F for the lower , compartment and 75 degrees F (60 degrees F when in Mode 2,3, or 4) for the upper i compartment limits the peak pressure to 14.7 psig which is less than the containment  ; design pressure of 15 psig. The upper temperature limit of 120 degrees F for the lower i compartment and 100 degrees F for the upper compartment influences the peak accident temperature slightly during a LOCA: however, this limit is based primarily upon cluipment protection and anticipated operating conditions. Procedures require error analyses to be performed for indications which are relied upon to  ! maintain / verify significant parameters associated with nuclear safety analyses. Both the upper and lower containment temperature limits have been analyzed using limits which incorporate the error uncertainties derived in CNC-1210.04-00-0054, Rev. 2. The peak containment pressure transient was analyzed utilizing 95 degrees F for the lower  ! containment and 70 degrees F for the upper containment. It was demonstrated that the peak containment pressure is not exceeded for ice masses well below the current  ; Technical Specification 3.6.5.1 minimum values for Catawba Nuclear Station Units 1 and

2. Therefore application of error uncertainties to the lower temperature limits is not necessary.

The upper temperature limit for lower containment is not utilized in Safety Analyses. The peak lower containment temperature and pressure transient was analyzed utilizing 125 and 135 degrees F. Results indicated that there was no significant impact to this analysis.

U.S. Nuclear Regulatory Comunission Apdf 1,1999 . Page 192 of 247 Therefore application of error uncertainties to the upper temperature limits is not n e e m ry. Removal of the containment temperature error uncertainties from the subject mode surveillance procedures is acceptable because they have been incorporated into the safety analyses . These changes will allow the plant to operate per the temperature limits of Technical Specification 3.6.1.5. There are no unreviewed safety questions associated with this procedure. No Technical Specification changes are required. No UFSAR changes are required.

p

       '_U.S. Nuclear Regulatory Co==8-Aa=

April 1,1999 l Pete 193 of 247 - [

        .101     Type: Procedure                                                         Unit: 0 Titlei Procedure PT/1,2/A/46004)2D

Description:

Procedure Pr/1,2/A/46004)2D, Units I and 2 Mode and Pre-Mode Checklists, changes the normal pre-accident operating temperature limits utilized in mode and pre-mode periodic surveillance procedures. Calculation CNC 1210.04-004054, Rev. 2 determined ' normal pre-accident total loop uncertainties for the upper and lower containment temperature detectors. These total loop uncertainties are used to calculate a reduced error depending upon the number of operating air handling units and available temperature (RTD) loops in service. The RTDs and their associated alarms are utilized by the operators to verify and maintain the bulk average containment air temperature within Technical Specification 3.6.1.5 limits. These containment temperature loops are non safety related, however, they are utilized to verify the containment temperature parameter associated with plant Safety Analyses. The surveillance procedures currently utilize error uncertainties determined per CNC-1210.04404054, Rev. 2 to verify average containment temperature per Technical Specification 3.6.1.5. These error uncertainties are no longer required for the subject Mode Surveillance Procedures because several nuclear safety analyses were perforned to incorporate these uncertainties. Evaluation: The limitations on containment average air temperature ensure that: (1) The containment air mass is limited to an initial mass sufficiently low to prevent exceeding the design pressure during a LOCA condition. (2) The ambient air temperature does not exceed the temperature allowable for the continuous duty rating specified for equipment and instrumentation located in the containment. The containment pressure transient is sensitive to the initially contained air mass during a LOCA The containment air mass increases with decreasing temperature. The lower Technical Specification 3.6.1.5 temperature limit of 100 degrees F for the lower compartment and 75 degrees F (60 degrees F when in Mode 2, 3. or 4) for the upper compartment limits the peak pressure to 14.7 psig which is less than the containment design pressure of 15 psig. The upper temperature limit of 120 degrees F for the lower compartment and 100 degrees P for the upper compartment influences the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon

                        . equipment protection and anticipated operating conditions.

Procedures require error analyses to be performed for indications which are relied upon to maintain / verify significant parameters associated with nuclear safety analyses. Both the upper and lower containment temperature limits have been analyzed using limits which incorporate the error uncertainties derived in CNC-1210.04-00-0054, Rev. 2.

                          'Ihe peak containment pressure transient was analyzed utilizing 95 degrees F for the lower containment and 70 degrees F for the upper containment. It was demonstrated that the peak containment pressure is not exceeded for ice masses well below the current               ,

Technical Specification 3.6.5.1 minimum values for Catawba Nuclear Station Units 1 and

                         . 2. Therefore application of error uncertainties to the lower temperature limits is not
                        -necessary.

The upper temperature limit for lower cc,ntainment is not utilized in Safety Analyses. The peak lower containment temperature anC pressure transient was analyzed utilizing 125 and 135 degrees F. Results indicated *. hat there was no significant impact to this analysis. L o Lb.

 .I U.S. Nuclear Regulatory Conunission Apdl 1,1999 Page 194 of 247

! 'Iherefore application of error uncertainties to the upper temperature limits is not necessary. Removal of the containment temperature error uncertainties from the subject mode surveillance procedures is acceptable because they have been incorporated into the safety analyses . 'Ihese changes will allow the plant to operate per the temperature lindts of Technical Specification 3.6.1.5. There are no unreviewed safety questions associated j1 with this procedure. No Technical Specification changes are required. No UFSAR changes are required. 1 l 1 i i l I~ l

l 1 h U.S. Nuclear Regulatory Comunission Apdf 1,1999 Pane 19.5 of 247 - 102- Type: Procedure Unit: 0

Title:

Procedure PT/1,2/A/4600/19A .

Description:

Procedure PT/1,2/A/4600/19A, Units I and 2 Mode and Pre-Mode Checklists, changes the normal pre-accident operating temperature limits utilized in mode and pre-mode periodic surveillance procedures.

                    ' Calculation CNC-1210.04-00-0054, Rev. 2 determined normal pre-accident total loop uncertainties for the upper and lower containment temperature detectors. These total loop uncertainties are used to calculate a reduced error depending upon the number of operating air handling units and available temperature (RTD) loops in service. The RTDs and their associated alarms are utilized by the operators to verify and maintain the bulk average containment air temperature within Technical Specification 3.6.1.5 limits. These containment temperature loops are non safety related, however, they are utilized to verify
                     - the containment temperature parameter associated with plant. Safety Analyses.

The surveillance procedures currently utilize error uncertainties determined per CNC-1210.04-00-0054, Rev. 2 to verify e verage containment temperature per Technica! Specification 3.6.1.5. These error u ncertainties are no longer required for the subject Mode Surveillance Procedures becmse several nuclear safety analyses were performed to incorporate these uncertainties. Evaluation: The limitations on containment average air temperature ensure that: (1) The containment air mass is limited to an initial mass sufficiently low to prevent exceeding the design pressure during a LOCA condition. (2)'the ambient air temperature does not exceed the temperature allowable for the continuous duty rating specified for equipment and instrumentation located in the containment.- The containment pressure transient is sensitive to the initially contained air mass during a LOCA. The containment air mass increases with decreasing temperature. The lower Technical Specification 3.6.1.5 temperature limit of 100 degrees F for the lower compartment and 75 degrees F (60 degrees F when in Mode 2,3, or 4) for the upper compartment limits the peak pressure to 14.7 psig which is less than the ccatainment design pressure of 15 psig. The upper temperature limit of 120 degrees D for the lower compartment and 100 degrees F for the upper compartment influences the peak accident temperature slightly during a LOCA; however, this limit is based r.smarily upon

                    - equipment protection and anticipated operating conditions.

Procedures require error analyses to be performed for indications which are relied upon to  ; maintain / verify significant parameters associated with nuclear safety analyses. Both the l upper and lower containment temperature limits have been analyzed using limits which j incorporate the error uncertainties derived in CNC-1210.04-00-0054. Rev. 2.  ! The peak containment pressure transient was analyzed utilizing 95 degrees F for the lower containment and 70 degrees F for the upper containment. It was demonstrated that the peak containment pressure is not exceeded for ice masses well below the current Txhnical Specification 3.6.5.1 minimum values for Catawba Nuclear Station Units I and

2. Therefore application of error uncertainties to the lower temperature limits is not necessary.

The upper temperature limit for lower containment is not utilized in Safety Analyses. The peak lower containment temperature and pressure transient was analyzed utilizing 125 and 135 degrees F. Results indicated that there was no significant impact to this analysis. 'V i l l

I? U.S. Nuclear Regulatory Convaission April 1,1999 Page 196 of 247 Therefore application of error uncenainties to the upper temperature limits is not necessary. Removal of the containment temperature error uncertainties from the subject mode surveillance procedures is acceptable because they have been incorporated into tie safety - analyses . These changes will allow the plant to operate per the temperature limits of Technical Specification 3.6.1.5. There are no unreviewed safety questions associated with this procedure. No Technical Specification changes are required. No UFSAR changes are required.

p , i l - U.S. Nuclear Regulatory Comunission

   ' Aptil1,1999        .

Pane 197 of 247 1

     '103     Type: Procedure                                                       ' Unit: 0 litle: Procedure PT/1,2/A/4600/19D

Description:

Procedure PT/1,2/A/4600/19D, Units I and 2 Mode and Pre-Mode Checklists, changes the normal pre-accident operating temperature limits utilized in mode and pre-mode periodic surveillance procedures. . Calculation CNC-1210.044)0-0054, Rev. 2 determined normal pre-accident total loop uncertainties for the upper and lower containment temperature detectors. These total loop uncertainties are used to calculate a reduced error depending upon the number of operating air handling units and available temperature (RTD) loops in service. The RTDs and their associated alarms are utilized by the operators to verify and maintain the bulk average containment air temperature within Technical Specification 3.6.1.5 limits. These containment temperature loops are non safety related, however, they ate utilized to verify the containment temperature parameter associated with plant Saf:ty Analyses. The surveillance procedures currently utilize error uncertainties determined per CNC-1210.04-00-0054, Rev. 2 to verify average containment temperature per Technical Specification 3.6.1.5. These error uncertainties are no longer required for the subject Mode Surveillance Procedures because several nuclear safety analyses were performed to incorporate these unceitainties. Evaluation: The limitations on containment average air temperature ensure that: (1) The containment air mass is limited to an initial mass sufficiently low to preveni exceeding the design presrure during a LOCA condition. (2) The ambient air temperature does not exceed the temperature allowable for the continuous duty rating specified for equipment and instrumentation located in the containment. The containment pressure transient is sensitive to the initially contained air mass during a LOCA. The containment air mass increases with decreasing temperature. The lower i Technical Specification 3.6.1.5 temperature limit of 100 degrees F for the lower j compartment and 75 degrees F (60 degrees F when in Mode 2,3, or 4) for the upper j compartment limits the peak pressure to 14.7 psig which is less than the containment design pressure of 15 psig. The upper temperature limit of 120 degrees F for the lower compartment and 100 degrees F for the upper compartment influences the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and anticipated operating conditions. , Procedures require error analyses to be performed for indications which are relied upon to maintain / verify significant parameters associated with nuclear safety analyses. Both the upper and lower containment temperature limits have been analyzed using limits which

                     - incorporate the error uncertainties derived in CNC-1210.04 004)054. Rev. 2.                     ;

The peak containment pressure transient was analyzed utilizing 95 degrees F for the lower ' containment and 70 degrees F for the upper containment. It was demonstrated that the peak containment pressure is not exceeded for ice masses well below the current Technical Specification 3.6.5.1 minimum values for Catawba Nuclear Station Units I and

                      .2. Therefore application of error uncertainties to the lower temperature limits is not necessary.
                     - The upper temperature limit for lower containment is not utilized in Safety Analyses. 7he
                     - peak lower containment temperature and pressure transient was analyzed utilizing 125            l and 13.5 degrees F. Results indicated that there was no significant impact to this analysis.

r s U.S. Nuclear Regulatory Conunission April 1,1999 Paste 198 of 247 - Therefore application of error uncertainties to the upper temperature limits is not necessary. Removal of the containment temperature error uncertainties from the subject mode surveillance procedures is acceptable because they have been incorporated into the safety analyses . These changes will allow the plant to operate per the temperature limits of Technical Specification 3.6.1.5. Here are no unreviewed safety' questions associated with this procedure. No Technical Specification changes are required. No UFSAR changes are required. 279 Type: Procedure Unit: 1

Title:

Procedure PT/1/A/4200/0lN, Reactor Coolant System Pressure Boundary Leak Rate Test, Revision 42

Description:

Procedure PT/1/A/4200/0lN, Reactor Coolant System Pressure Boundary Leak Rate Test, Revision 42 adds new enclosures to allow for background testing the boundary valves. His revision also converts the procedure to a new word processing software. During this revision additional changes were made to enhance the procedure. Some of these enhancements were administrative in nature, others were to provide clarification to the users. None of these changes affected the test mehod or the acceptance criteria of the test. Evaluation: There is no unreviewed safety question associated with this procedure change. No changes to the UFSAR are required. No Technical Specification changes are required. I.......r- i. . . - - ___2__. _ _ _ _ _ _ _ _ _ - -

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    - U.S. Nuclear Regulatory Comunission April 1,1999 Page 199 of 247
 ..                                                                                        t 57    Type: Procedure                                                          Unit: 1

Title:

Procedure IT/1/A/4200/059 Revision 32  ;

Description:

Procedure PT/1/A/4200/059, Nuclear Service Water System to Auxiliary Feedwater

                        ' System Piping Flush, is a procedure to flush this piping to remove clams and settlement
                       . that that could restrict flow to Auxiliary Feedwater System if allowed to accumulate.

Revision 32 makes the following changes: 1.) Add a time limit of 15 minutes to secure the flush following a safety injection or auxiliary feedwater auto start on either unit. 2.) Add requirement for an operator to be stationed in the r.uxiliary feedwater pnp room or auxiliary shutdown panel room while the flush is in progress. 3.) Adds steps to record the names of the operator and person in the control room who can terminate the flush 4.) Adds a step to record the primary and secondary means of communication between the Control Room and the operator performing the flush and a step to establish communications prior to beginning the flush. 5.) Adds a constraint to not have any testing or evolution in progress which would establish flow through a Containment Spray System licat Exchanger while the flush is in progress. _  ! 6.) Adds a step to declare the associated motor driven auxiliary feedwater pump - inoperable while the flush is in progress 7.) Deletes steps associated with flushing Train B from the Train A enclosure and steps associated with flushing Train A from the Train B Enclosure. l

                      ' 8.) Adds a requirement to have both of the train related nuclear service water pumps       !

operable when performing the flush. ' Evaluation: These changes do not affr:t the availability of the Nuclear Servici Water System, the Auxiliary Feedwater System, or the Emergency Core Cooling System. There are no unreviewed safety questions associated with this procedure change. No UFSAR changes are required. No Technical Specification changes are required. l i.

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                                                                                                                     .l
 . U.S. Nuclear Regulatory Co--l-lon
 . April 1,1999 -

Page 200 of 247

58. Type: Procedure Unit: 1

Title:

~ Procedure PT/1/A/4200/069

Description:

Procedure PT/1/A/4200/069, " Steam Generator PORV nitrogen leakage test", tests leakage of the nitrogen system associated with the Steam Generator PORVs.He purpose of this procedure is to test the leakage of the pneumatic circuitry associated with the Steam Generator PORV'S. De Steam Generator PORV's are provided with safety related nitrogen supplies which ensure that certain Design and Lic<.9 sing Basis Events would be adequately mitigated. %e calculation that was used to size the nitrogen bottles for the PORV's assumed a certain value ofleakage through the positioner. However, the circuit - is never tested to assure that the assumption in the calculation is valid. This test

                      ~ procedure will isolate the instrument air supply to the PORV'S, bleed the nitrogen           _

pressure down to the Technical Specification value (this value is temperature corrected), have Operations stroke the valve five times, then let the system sit for at least 8 hours and then record the final bottle pressures and doghouse temperature. The final bottle pressure must be greater than or equal to 80 psig in at least one bottle with a doghouse temperature of less than 130 degrees F. This will validate the assumptions that were made in the calculation that sized the nitrogen bottles. Since the bottles will leak down below the Technical Specification limit of 2l00 psig, the PORV being tested will be declared inoperable at the beginning of the test (in accordance with T/S 3/4.7.1.6). At the end of the test new nitrogen bottles wift be installed and the PORV will be declared operable. j Evaluation: No unreviewed safety questions are created by this procedure change. The PORV being tested is placed in the manual mode and closed. The manual position is the safety related position of the valve. The valve is declared inoperable during the testing and the applicable Technical Specification (3/4.7.1.6) is followed. No UFSAR changes are required. No Technical Specification changes are required. 59 Type: Procedure Unit: I

             'lltle: Procedure PT/1/A/4400/06E Revision 13

Description:

' Procedure PT/1/A/4400/06E, Diesel Generator Cooling Water System Heat Exchanger

           ~

I A Heat Capacity Test, Revision 13 makes the following changes: 1) allow the test

                      ' coordinator to throttle nuclear service water flow to help system conditions stabilize. 2) increase the maximum acceptable nuclear service water flow for collecting data from -

1425 gpm to 1500 gpm. 3) changes the acceptable fouling factor on the heat exchanger to agree with the " Test Acceptance Criteria" sheet.

    - Evaluation: here is no unreviewed safety question as a result of this procedure change. These changes do not affect the availability of the Diesel Generator. The ability of the Diesel Generator Cooling Water System and the Nuclear Service Water System to perform their safety function is not compromised. No UFSAR changes are required. No Technical Specification changes are required.

y U.S. Nuclear Regulatory Conunission Apdl 1,1999

    - Page 201 of 247
       '60    ' Type: Procedure -                                                    Unit: 1                          q

Title:

Procedure PT/1/A/4400/06F Revision 15

Description:

Procedure PT/1/A/4400/06F, Diesel Generator Cooling Water System Heat Exchanger Li IB Heat Capacity Test, Revisiori 15, makes the following changes: 1) allow the test l coordinator to throttle nuclear service water flow to help system conditions stabilize. 2) i increase the maximum acceptable nuclear service water flow for collecting data from 1425 gpm to 1500 gpm. 3) changes the acceptable fouling factor on the heat exchanger to agree with the " Test Acceptance Criteria" sheet. Evaluation: There is no unreviewed safety question as a result of this procedure change. These changes do not affect the availability of the Diesel Generator. The ability of the Diesel Generator Cooling Water System and the Nuclear Service Water System to perform their safety function is not compromised. No UFSAR changes are required. No Technical Specification changes are required. - l 1 I i i ( l' t k. i LL

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        .U.S. Nuclear Regulatory Comunission
        April 1,1999 Page 202 of 247 l

182 Type: Procedure Unit: 1

Title:

Procedure PT/1/A/4450/001 A Revision 09 and PT/2/A/4450/001 A Revision 12

Description:

Procedure PT/1/A/4450/001 A Revision 09 and PT/2/A/44504)01 A Revision 12 were completely rewritten to ensure conformance with current procedure standards. He procedures address testing of the Annulus Ventilation System. The changes to the procedure included the following: 1) Converting the format to the standard Operations format.

2) Test Sections were arranged in Enclosure Format.

I

3) Various human performance notes were added or deleted.

4) Component nomenclature changes were made to enhance human performance.

5) Technical Specification references were added.

6) Information from another procedure addressing filter performance testing was added.

7) A psychrometric chart was added.

8) Two enclosures addressing airflow adjustment fan data and procedure multiple signoffs were deleted.

9) instruction were added to addresss test failures.

10) Information was added to ensure consistent test alignments.

11) Steps were added to ensure electrical power is removed when filter i

units are open for inspection.  ;

12) Steps were added to secure certain dampers during testing.

13) Limits and precautions note was added as guidance for reducing refrigerant concentration after carbon adsorber in place leak rate test.

14) other minor editorial changes were made.

Evaluation: Here are no unreviewed safety questions associated with this procedure change. The Annulus Ventilation System is not an accident initiator and the actions identified by this procedure will not increase the probability of an accident. These revis. ions do not change the basic testing method. No Technical Specification changes are required. No UFSAR changes are required. L L

U.S. Nuclear Regulatory Conunission April 1,1999 Page 203 of 247 223 Type: Procedure Unit: 1

Title:

Procedure PT/1/A/4450/003C Revision 37 and PT/2/A/4450/003C Revision 28, Annulus Ventilation System Performance Test .

Description:

Procedure PT/1/A/4450/003C Revision 37 and PT/2/A/4450/003C Revision 28, Annulus Ventilation System Performance Test are major revisions to update the procedure format j to current standards. Evaluation: The major procedure changes described in the Activity Description Section of this evaluation will not adversely affect the technical aspects of these system performance i tests. The test methods and intent of the procedures did not change.  ; i Steps in the Preheater Power Dissipation Enclosures which required test personnel to contact Engineering regarding differences in phase or amperage data were deleted. Steps were added to the acceptance criteria to declare the heater inoperable when voltage differences exceed 5% and current differences exceed 10%. Large variations in heater phase voltage or amperage are indication that degradation of the heater elements or other electrical components have occurred. Therefore, the acceptance criteria was revised to ensure the heater is declared inoperable when phase voltage differences exceed 5% or phase amperage differences exceed 10%. *Ihe variations were based on past heater test results and engineering judgment. Acceptance Criteria was changed to reflect the Annulus Ventilation System Test Acceptance Criteria for annulus drawdown time during normal plant operation. Calculation CNC 1211.0040-0086, " Annulus Technical Specification Drawdown Time", generated the data which established the test acceptance criteria during non accident conditions to ensure that Surveillance Requirement 4.6.1.8.d.4 (60 seconds) would be met during accident conditions. The total filter train pressure drop associated with the acceptance criteria was changed from 8 inwg to 7.9 inwg to agree with the acceptance criteria in PT/l(2)/A/4450/001 A (Annulus Ventilation Filter Train Performance Test) and other similar filter trains at Catawba. The variation from Surveillance Requirement 4.6.1.8.d.1 (8 inwg) is due to the uncertainty or error associated with the test manometers utilized during testing. This change does not affect the actual test methods or intent of the procedure. Steps were changed in the vacuum decay test enclosures of I'T/2/A/4450/003C to allow single train testing during any mode of plant operation. Additional steps were added to inspect individual dampers prior to testing to reduce the potential for failure. There is no unreviewed safety question associated with these procedure changes. No Technical Specification changes are required. No UFSAR changes are required.

n, U.S. Nuclear Regulatory Coenraission April I,1999 l Pase 204 of 247 222 Type: Procedure Unit: 2- j

Title:

Procedure PT/2/A/4200/001C Revision 114, Containment Isolation Valve leak Rate Test

Description:

Procedure PT/2/A/4200/001C, Revision i14, Containment Isolation Valve 12ak Rate Test, corrects the containment penetration numbers for several penetrations. Evaluation: This procedure change is only an administrative change to correct four penetration numbers that were shown incorrectly in the procedure. There is no actual change to the plant. There are no unreviewed safety questions associated with this procedure revision. No Technical Specification changes are required. A revison is required for UFSAR Table 6 77,

                                                ~

243. Type: Procedure Unit: 2

Title:

Procedure PT/2/A/4200/0lN Restricted Change 38, Reactor Coolant System Pressure Boundary Valve Leak Rate Test

Description:

Restricted Change 38 to Procedure PT/2/A/4200/ DIN involves addition of new. enclosures to allow for testing boundary valves in groups. This revision also involves converting the procedure to a different word processing software which will allow for enhanced readability. Other administrative enhancements and clarifications were made as  ; well. None of the changes affected the test method or acceptance criteria. During { performance of new procedure enclosures and all other enclosures, the procedure ensures i that (a) a valid test flowpath exists, (b) each valve under test is pressurized, and (c) no direct flowpath from the Reactor Coolant System to the Auxiliary Building exists. Evaluation: There is no unreviewed safety question associated with this procedure change. No  ; changes to the UFSAR are required since this test is not described in detail in the UFSAR. No Technical Specification changes are required. e , 1

t U.S. Nuclear Regulatory Commission I Apdf 1,1999 Page 205 of 247 265 Type: Procedure Unit: 2

Title:

Procedure PT/2/A/4200/069 Steam Generator Power Operated Relief Valve (PORV) Nitrogen leak Test

Description:

Procedure PT/2/A/4200/069 Steam Generator PORV Nitrogen Leak Test will test the leakge of the pneumatic circuitry associated with the Steam Generator PORVs. He Steam Generator PORVs are provided with nuclear safety related nitrogen supplies

which ensure that censin Design and Licensing Basis Events would be adequately mitigated. The calculation used to size the nitrogen bottles for the PORVs assumed a

                    . certain value ofleakage through the positioner. However, the circuit is never tested to assure that the assumption in the calculation is valid. This test procedure will isolate the Instrument Air supply to the PORVs, bleed the nitrogen pressure down to the Technical Specification value (temperature corrected) , have the valve stroked five times, allow the system to sit for at least eight hours and then record the final bottle pressure and doghouse temperature. The final bottle pressure must be greater than or equal to 80 psig in at least one bottle with a doghouse temperature of less than 130 degrees F. This will validate the       ,

assumptions that were made in the calculation that sized the nitrogen bottles. Since the  ! bottles will leak down below the Technical Specification limit of 2l00 psig, the PORV ,l being tested will be declared inoperable at the beginning of the test (in accordance with l Tech Spec 3/4.7.1.6). At the end of the test new nitrogen bottles will be installed and the PORV will be declared operable (assuming that it passes the test acceptance criteria). Evaluallon: There are no unreviewed safety questions associated with this procedure. He procedure contains adequate guidance for ensuring that the PORV being tested is in the manual mode and closed. The manual position is the safe position of the valve. He valve under test will be declared inoperable and Tech Spec 3/4.7.1.6 will be followed. No Technical Specification changes are required. No UFSAR changes are required. 4 l i I I l 1 1 1 I t

y i-p l' U.S. Nuclear Regulatory Co==d==L= ! April 1,1999 Page 206 of 247 9 272 . Type: Procedure Unit: 2 [' !~

                 'Iltle: Procedure PT/2/A/4200/31 A, Revision 12A

Description:

Procedure IT/2/A/4200/31 A', Revision 12A adds information to the proocedure so that it will no longer be necessary to measure the differential pressure across the valve actuator for valve 2SV-19. In the current procedure a differential pressure transmitter is installed across the valve actuator. The valve is then partially stroked (30% open) and the actuator differential pressure is recorded with a visicorder. His pressure is compared with a calculated differential pressure value to determine acceptable valve performance. In the revised procedure, an option will be given to have maintenance adjust the nitrogen . regulator for the valve to the calculated pressure. Operations will then pantially stroke the valve 30% open. The only acceptance criteria will be that the valve stem travels more than 0.75 inches. Adjusting the regulator pressure to the calculated value prior to stroking the valve accomplishes the same goal as measuring the differential pressure and

                          . comparing it against a calculated value. In either case acceptable valve performance can be verified.

Evaluation: There are no unreviewed safety questions associated with this procedure revision' This . l ' revision provides a different option for !.-scing. He valve being tested will operate the same way as it does in the current procedure (stroked 30% open) No Technical ' Specification changes are required. No UFSAR Changes are required. i 233 . Type: Procedure Unit: 2

Title:

Procedure PT/2/A/4200/75 Component Cooling System alignments for Generic Letter 89- l 10 Testing, was revised to add two valves.

Description:

Valves 2KC-003A and 2KC 230A were added to the existing procedure.His procedure change does not affect any section of the existing procedure, it only adds the new sections to test these two valves. De procedure provides a controlled system alignnwnt to support

                           ' static and/or differential pressure testing associated with the requirements of Generic -

letter 89-10. -

          . Evaluation: No unreviewed safety questions are introduced by this procedure change. No Technical Specification changes are required. No UFSAR changes are required.

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                                                                                                                          .j

i I I U.S. Nuclear Regulatory Commission ' Apdf 1,1999 Page 207 of 247 f 263 Type: Procedure Unit: 2

Title:

Procedure PT/2/A/4250/03E Revision 23A, Auxiliary Feedwater System discharge I control valve throttling procedure.

Description:

Procedure PT/2/A/4250/03E Revision 23A makes the following changes: 1. Enclosure 13.10 is revised to provide instructions for installation and removal of test instrumentation. The systems to which this instrumentation is installed are required to be operable during performance of the test. This revised enclosure provides detailed direction to ensure the system operability is not affected. 2. Several procedure sections < were revised to eliminate the requirement to perfonn stroke time testing using visual I observation for valves being throttled open by an insignificant amount. One half turn on the handwheel of these valves is equivalent to approximately 0.1 inch of valve travel. 3 These valves are air-operated (fail open) and most of the stroke time (close to open) is associated with the venting of air from the operator to start valve movement. The normal stroke time of these valves is less than 10 seconds and the required stroke time for system operability is less than or equal to 60 seconds. Given the inacuracy of timing these valves by visual observation, requiring such a test for a small adjustment is unnecessary. The procedure already requires the associated steps to be signed by the Auxiliary Feedwater

                . System Engineer or the IST Engineer if the stroke is not performed.

Evaluation: These changes are considered editorial. They do not have a non conservative effect on ) the performance of the procedure. The Auxiliary Feedwater System is not an accident initiator for an UFSAR analyzed accident. There are no unreviewed safety questions associated with this procedure change. No Technical Specification changes are necessary. No UFSAR changes are required. 225 Type: Procedure Unit: 2

Title:

Procedure PT/2/A/4400/01, Emergency Core Cooling System (ECCS) Flow Balance, Revision 28

Description:

Procedure PT/2/A/4400/01, Emergency Core Cooling System (ECCS) 190w Balance,

                 ~ Revision 28, makes enhancements to the existing procedure that allows testing to bc         )

performed in an caiser manner. The changes made to the procedure do not significantly j affect the test procedure nor the data and results obtained from the test. Evaluation: These procedure changes have no adverse impact on plant safety. ne test is performed l with no fuel in the reactor core and does not affect equipment associated with the Spent  ! Fuel Pool. The original purpose of the test is still applicable and the intent of the test is I not comprimised as a result of these procedure changes. He ECCS equipment is operated within specified design limits during the test. No Unreviewed Safety Questions are created by this procedure revision. No Technical Specification changes are necessary. No UFSAR changes are required.

p U.S. Nuclear Regulatory Conunission April 1,1999 - Page 208 of 247 t 61- Type: Procedure Unit: 2

Title:

Procedure PT/2/A/44005)6E Revision 10 -

Description:

Procedure PT/2/A/44004)6E, Diesel Generator Cooling Water System Heat Exchanger 2A Heat Capacity Test, Revision 10, makes the following changes: 1) allow the test coordinator to throttle nuclear service water flow to help system conditions stabilize. 2) increase the maximum acceptable nuclear service water flow for collecting data from 1425 gpm to 1500 gpm. 1 Evaluation: There is no unreviewed safety question as a result of this procedure change. These changes do not affect the availability of the Diesel Generator. The ability of the Diesel Generator Cooling Water System and the Nuclear Service Water System to perform their safety function is not compromised. No UFSAR changes are required. No Technical Specification changes are required. 62 Type: Procedure Unit: 2

Title:

Procedure PT/2/A/4400/06F Revision 11

Description:

Procedure PT/2/A/4400/06F, Diesel Generator Cooling Water System Heat Exchanger 2A Heat Capacity Test, Revision 11, makes the following changes: 1) allow the test coordinator to throttle nuclear service water flow to help system conditions stabilize. 2) increase the maximum acceptable nuclear service water flow for collecting data from

                   ' 1425 gpm to 1500 gpm.

Evaluation: Tiiere is no unreviewed safety question as a result of this procedure change. These changes do not affect the availability of the Diesel Generator. The ability of the Diesel Generator Cooling Water System and the Nuclear Service Water System to perform their 3 safety function is not compromised. No UFSAR changes are required. No Technical Specification changes are required. 121 Type: Procedure Unit: 0

Title:

Procedure SH/0/B/2005/001, Emergency Response Offsite Dose Projections

Description:

Procedure SH/0/B/2005/001, Emergency Response Offsite Dose Projections, is a new procedure. His new standard procedure provides methods for calculating downwind offsite dose in the event of radioactive material release during a Site Emergency Response. Evaluation: This procedure has no effect on any system, structure, or component described in the UFSAR. Dere is no unreviewed safety question associated with this procedure. No changes to the UFSAR are required. No Technical Specification changes are required. l I L

p- ) i; l. L L U.S. Nuclear Regulatory Conunission April 1,1999 l l . Page 209 of 247 i

        '63    . Type: Procedure                                                          Unit: 2

Title:

Procedure Tl/2/B/3222/001 Revision 0 l

Description:

Procedure Tl/2/B/3222M01. Revision 0, provides a safe and correct method for investigation and repair of the Steam Dump Pressure Controller within instrument loop IT507. His procedure is conducted with the Steam Dump Pressure Controller mode

                         . selector switch in the T Avg Mode. His will isolate the equipment that is being tested.
                            %c planned steps are to verify the as found operation of the driver card, to replace the controller card with the card that was originally installed in it, to verify and adjust if necessary the tightness of the connector on the rear of the M/A selector station and to potentially replace the driver card. After the work is complete a functional test will be      ,

performed to ensure that through overlap testing that all three components work properly. l Both automatic and manual functionality will be verified. l Evaluation: No unreviewed safety questions were identified associated with this procedure. No UFSAR changes are required. No Technical Specification changes are required 119 Type: Procedure Unit: 2

Title:

ProcedureTN/2/B/1281/MM/01E

Description:

Procedure TN/2/B/1281/MM/01E is the implementing procedure for Minor Modification CE-61281 Work Unit 1. Completion of this procedure will change the power source for the A Train Boron Dilution Mitigation System interlock in the Reactor Makeup Water

                        . pumps interlock circuit from non-blackout source 2RPA to blackout source 2KXPA.

This will make the A Train Boron Dilution Mitigation System interlock in each Reactor Makeup Water Pump circuit functional during a blackout. The A Train Shutdown Margin

                        ' Computer Alarm and the A Train Shutdown Margin Control Room Annunciator will also be functional during a blackout.                                                              .

Evaluation: His modification will be installed during a time when the Boron Dilution Mitigation l System is not required to be operable. If plant conditions change during the 1 l-implementation of the Modification, Technical Specification 3.3.3.3.11 provides

                          - alternative measures to account for unavailability of the Boron Dilution Mitigation            ,

System.nere are no unreviewed safety questions as a result of this modification. No j Technical Specification changes are required. No UFSAR changes are required. i I i l i i I

                                                                                                                       ^

I N l U.S. Nuclear Regulatory Commission l Apdf 1,1999 l Page 210 of 247 i 120 Type: Procedure Unit: 2

Title:

Procedure TN/2/B/1281/MM/02E

Description:

Procedure TN/2/B/1281/MM/02E is the implementing procedure for Minor Modification - CE-61281 Work Unit 2. Completion of this procedure will change the power source for l the B Train Boron Dilution Mitigation System interlock in the Reactor Makeup Water pumps interlock circuit from non-blackout source 2RPB to blackout source 2KXPB. This will make the B Train Boron Dilution Mitigation System interlock in each Reactor Makeup Water Pump circuit functional during a blackout. The B Train Shutdown Margin Computer Alarm and the B Train Shutdown Margin Control Room Annunciator will also be functional during a blackout. Evaluation: This modification will be installed during a time when the Boron Dilution Mitigation System is not required to be operable. If plant conditions change during the implementation of the Modification, Technical Specification 3.3.3.3.11 provides alternative measures to account for unavailability of the Boron Dilution Mitigation System. There are no unreviewed safety questions as a result of this modification. No Technical Specification changes are required. No UFSAR changes are required. 95 Type: Procedure Unit: 0

Title:

Procedure 'IT/0/A/9200/026, Revision 0

Description:

Procedure *IT/0/A/9200/026 Revision 0, Temporary Test for Auxiliary Building Ventilation System, is a test procedure which is intended to ensure that the Auxiliary l Building Ventilation System can meet Technical Specification Surveillance Requirement j 4.7.7.d.l. Until recently it was believed that another procedure (PT/0/A/4450/001C) J l satisfied Technical Specification Surveillance Requirement 4.7.7.d.l. However,it was ! determined that procedure did not adequately meet the Surveillance Requirement. The i Surveillance Requirement assures that the Auxiliary Building Ventilation System can provide adequate exhaust flow throughout the plant as the filters become dirty and the filter unit pressure drop increases. There were other processes in place to ensure that the Technical Specification Surveillance requirement was met. 1 Evaluation: There are no unreviewed safety questions associated with this procedure.The Auxiliary Building Ventilation System is not an accident initiator and the system would still be able to perform its design basis function during the performance of the test. No Technical Specification changes are required. No UFSAR changes are required. I 1 l l

F' L L l U.S. Nuclear Regulatory Comunission f .Apdl1,1999 L

   . Page 211 of 247 64    Type: Procedure                                                        Unit: 0
              'Iltlei Procedure'IT/0/AS200/092 Revision 0

Description:

Procedure TT/0/AS200/092 Revision 0 was developed after an incident occurred which appeared to render the Control Room Ventilation System inoperable. An access door on air handling unit 2CR-AHU-l was left open for approximately five minutes after maintenance work was performed on the air handling unit. He opposite train was out of service at the time. This test evaluated the ability of the Control Room Ventilation System to perform its design basis function given these conditions. His event was reported in Licensee Event Report 50-41368-001. Evaluation: A Compensatory Action was used during this test to avoid rendering the system inoperable. There are no unreviewed safety questions associated with this procedure. No UFSAR changes are required. No Technical Specification changes are required. 122 Type: Procedure Unit: 0

Title:

Procedure 'IT/0/BS100/070, Powerboost Assessment Test

Description:

Procedure'IT/0/BS100/070 assesses the effectiveness of the Powerboost chemicals to lower Cooling Tower outlet temperature during power operation to increase generator output during the summer months. Powerboost chemicals will be added to the basin of one Cooling Tower. Data will be collected and calculations will be performed to determine the effect the chemicals have on Cooling Tower capability. The addition of the Powerboost chemicals is expected to lower the cooling tower outlet temperature by up to five degress F. An evaluation was performed and it was determined that this temperature drop will have no effect on reactor power. He effect will be similar to an equivalent drop in condenser cooling water temperature caused by weather conditions. Evaluatnon: The operation of the plant will not be changed by the addition of these chemicals. Here is no unreviewed safety question associated with this procedure. No changes to the UFSAR are required. No Technical Specification changes are required.

w

   . U.S. Nuclear Regulatory Co==d-lon Apdf 1,1999 Pase 212 of 247 -

P 177 Type: UFSARChange Unit: 0

Title:

Calculation CNC-1553.26-00-0228 Evaluation of 1998 UFSAR Updates

Description:

He following changes were made to the Catawba UFSAR.

                   . 1) Reactor Coolant System Temperatures updates to Table 4-1 and 4-17 and related entries ( Table 4-1. Items 14-18 and Table 4-17 Coolant Temperatures) were updated to reflect typical Reactor Coolant System flow rate and program average temperature.

2) nose entries related to the thermal design flow (Tables 4-1 Items 9,10, 13 and Table 4-17 Items Total Thermal Flow Rate, Effective %ermal Flow Rate, Average Velocity, and Along Rods) were updated to be consistent with CNS Units 1 and2 Technical Specifications and typical Reactor Coolant System inlet temperatures.

3) Reactor Coolant System Temperatures in Tables 5 1 and 5-5 were .

updated to be consistent with Tables 4-1 and 417.

4) Fuel Pellet length, item 23 in Table 4-4 was updated to clarify the pellet length of the enriched region and the axial blanket region of the fuel rod.

5) Mass of H20/ Uranium ratio was updated to correct an arithmetic error in the calculation of the ratio.

6) Power Density footnote and entries to Table 4-1 (Item 26) and Table 4 17 were updated to clarify that the power density is a function of the fuel assembly pitch and not the fuel's theoretical density,

7) Mass Of UO2 per foot of fuel in Table 44 was updated to reflect current design data.

8) Section 4.4.6.5 was updated to reflect the NRC approved Reactor Coolant System flow measurement technique currently used to verify the RCS flow measurement criterion in CNS Units I and 2 Technical Specifications, Section 3.2.5 DNB Parameters.

The changes to reactor coolant temperatures CNS UFSAR Tables 4-1,4-17,5- 1, and 5-5 were made to reflect typical values for the specified Reactor Coolant System flow rates and bypass flow rates for design conditions. These values are not used in any reload or accident analyses previously evaluated or discussed in the SAR. He coolant flow rate changes made to the CNS UFSAR Tables 4-1 and 4-17 reflect the minimum allowed Reactor Coolant System flow per Technical Specificaticas that are used in generic DNB analyses. These changes do not increase the likelihood of DNB

                   - occurring per NRC approved analyses methodologies discussed in the SAR.

De changes to Table 4-4 to clarify the fuel pellet length of the central enriched fuel region and the axial blanket region are not used in any mechr.nical analyses (pin pressure, LHRTM, clad stress / strain, creep collapse, etc.). Also, the fuel pellet lengths are not explicitly modeled in any Nuclear Design reload analyes. Therefore, there is no adverse impact on the fuel reliability or performance. The correction in Tables 4-1 and 4-4 of the mass of H2O (Uranium ratio reflects the thermal utilir.ation of the fuel).This paramnter is not used in any analyses evaluated in the

(7 l L L U.S.' Nuclear Regulatory Con-d= san April 1,1999

    - Pane 213 of 247.-

f SAR. De power density changes reflect that there is no dependency of this parameter on the theoretical density of the fuel only the fuel assembly pitch. His value is not directly used

                         . in any reload, thermal or mechanical analyses. SIMULATE-3, an NRC approved computer code (DPC-NE-1004P-A), utilizes the specific power not the power density as defined in the UFSAR. Therefore, reload analyses performed by Nuclear Design are not impacted by this change.

The mass of UO2 per foot of fuel rod is not directly input into any reload, thermal, or nwchanical analyses. CASMO-3, an NRC approved computer code, utilizes the radial fuel pellet dimensions, fuel density, and stack height to calculate the metric tons of fuel. SIMULATE-3, an NRC approved computer code (DPC-NE-1004P-A), evaluates the impact on reactivity and power distributions. Since, this parameter is not used in any analysis, this update does not adversely affect the fuel reliability and performance. Section 4.4.6.5 was updated to reflect the NRC approved method for Reactor Coolant

                        - System flow measurement currently used to verify the criterion for Reactor Coolant System flow measurement in Technical Specifications Bases, Section 3.2.5, DNB Parameters.

He changes identified are not directly used in any accident analyses previously evaluated in the SAR. These changes reflect fuel rod design or assumptions consistent with the licensing bases specified in the CNS UFSAR and Technical Specifications. These changes do not result in any new failure mechanisms. He update to flow rate in the CNS UFSAR reflects the minimum Reactor Coolant System flow rate consistent with the Technical Specifications used to ensure DNB does not occur, Evaluation: There are no unreviewed safety questions associated with these UFSAR changes. No Technical Specification Changes are required. Changes are required for UFSAR Sections 1.5,4.1,4.2,4.4, Tables 4-1,4-4,4-17,5-1,5-5. , l 138 Type: UFSAR Change Unit: 0 l

                'Iltle: Change to UFSAR Chapter 5 Table 5-13

Description:

UFSAR Chapter 5, Table 5-13 (page 1 of 2) was changed to correct indicated values for (1) Tndt, (2) Rtndt and (3) Ni. , l

        . Evaluation: Dis change does not affect the operation, design basis or function of any structure, .

system or component. There are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are i required for UFSAR Chapter 5, Table 5 13.  ! i L2

i U.S. Nuclear Regulatory Comadssion Apdf 1,1999 Page 214 of 247 158. Type: UFSAR Citange Unit: 0

Title:

Change to UFSAR Chapter 10, Figure 10-41

Description:

UPSAR Chapter 10, Figure 1041 was revised to change the title of the figure from

                   " System Overview of Control Room System" to " System Overview of the Turbine Control System".

Evaluation: Dere is no unreviewed safety question associated with this UFSAR change. There was no actual change to the plant. No Technical Specification changes are required. A change is required for UFSAR Chapter 10, Figure 10-41. 136 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Chapter 5, Figure 5-13

Description:

UFSAR Chapter 5, Figure 5-13 was changed to replace the current Fig. 5-13 with a new Fig. 5 13 titled " Reactor Coolant Pump Hot Performance Curve". The new curve comes from the Reactor Coolant Pump Technical Manual CNM 1201.01-0157 001. Evaluation: This change does not affect the operation, design basis or function of any structure, system or component. There are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are required for UFSAR Chapter 5. Figure 5-13. 137 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Chapter 5, Table 5-12

Description:

UFSAR Chapter 5, Table 5-12 " Toughness Properties for the Catawba Unit 1 Reactor Vessel" was revised to correct two erroneous entries in the " Shelf Energy" column. Evaluation: This change does not affect the operation, design basis or function of any structure, system or component. Here are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are  ; required for UFSAR Chapter 5, Table 5 12. I I 1

r U.S. Nuclear Regulatory Conunission )' April 1,1999 Page 215 of 247 - t 139 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Chapter 5, Table 5-14

Description:

UFSAR Chapter 5 Table 5-14 was changed to correct various values for "RT" and "RT pts". Evaluation: This change does not affect the operation, design basis or function of any structure, system or component. There are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are required for UFSAR Chapter 5. Table 5-14. 140 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Chapter 5. Table 5-15

Description:

UFSAR Chapter 5, Table 5-15 " Fracture Toughness Properties of Unit 2 Reactor Vessel" was changed to correct various values for RTpts. Evaluation: This change does not affect the operation, design basis or function of any structure, system or component. The.e are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are required for UFSAR Chapter 5, Table 5-15. 146 Type: UFSAR Change Unit: 0  !

          'Iltle: Change to UFSAR Chapter 5, Table 5-25

Description:

UFSAR Chapter 5, Table 5-25 was changed to revise the Unit 2 Steam Generator total heat transfer area from 48,300 square feet to 48,165 square feet. This information comes from the manufacturer's drawings. Evaluation: This change does not affect the operation, design basis or function of any structure, system or component. 'Ihere are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are required for UFSAR Chapter 5. Table 5-25.

p L I i _U.S. Nuclear Regulatory Convaission April 1,1999

  ' Page 216 of 247 :
    '147     Type: UFSAR Change                                                      Unit: 0 l

Title:

Change to UFSAR Section 6.3.2.8 and Table 6-93

Description:

This UFSAR revision incorporates changes to the validation of the Refueling Water Storage Tank lo and lo-lo level setpoints due to updated Emergency Core Cooling System (ECCS) flow rates and operator response times. UFSAR Table 6-93 is also updated to include all of the steps shown in the procedure for the transfer to cold leg recirculation. The steps that are added prior to the stopping of the Containment Spray Pumps are accounted for in the operator response times assumed in the calculation of the Refueling

                     - Water Storage Tank level setpoints. The steps after the stopping of the Containment Spray Pumps are resequenced to agree with the procedure for the transfer to cold leg recirculation. The results of the validation of the Refueling Water Storage Tank level setpoints show that adequate time is provided to ensure the completion of all operator actions necessary for the transfer to cold leg recirculation prior to the loss of all usable Refueling Water Storage Tank inventory and loss of suction to the ECCS pumps. Since the Refueling Water Storage Tank level setpoints are not changed, the net positive suction head provided for the ECCS pumps is unaffected by these UFSAR revisions. In addition, the primary system boron concentrations following a loss of coolant accident (LOCA) are not affected, and therefore the core will remain suberitical. The LOCA blowdown, refill, and reflood phases occur before the initiation of the transfer to cold leg recirculation, and thus the UFSAR revisions do not impact fuel clad integrity. The operator response times and ECCS flow rates assumed in the analysis of the containment pressure response during a LOCA conservatively bound those contained in the UFSAR revisions, and thus the containment pressure response is not impacted.

Evaluation: There are no unreviewed safety questions associated with the revisions to UFSAR Section 6.3.2.8 and Table 6 93. No changes to the Technical Specifications are requ9d, nor are any station procedures affected since there is no physical change to the plant. Changes are required for UFSAR Section 6.3.2.8 and Table 6-93. { l 154 Type: UFSARChange Unit: 0

Title:

Change to UFSAR Section 10.2.2

Description:

This revision to UFSAR Section 10.2.2 corrects a statement about turbine overspeed protection. When the turbine overspeed protection system activates it will directly trip the disc-dump valves of the main and intermediate stop valves as well as the control and , intercept valves. I I Evaluation: There is no unreviewed safety question associated with this UFSAR change. No actual i change being made to the plant. No Technical Specification changes are required. A , change is required for UFSAR Section 10.2.2. l

                                                                                                                      ^

I l

y' l f U.S. Nuclear Regulatory Conunission Apdf 1,1999 - Page 217 of 247 f 153 Type: UFSAR Change Unit: 0 1

Title:

Change to UFSAR Section 10.2.2

Description:

His revision to UFSAR Section 10.2.2 corrects the description of the sequence of events leading to a turbine trip on overspeed. De correct information is that when an overspeed condition is detected, the turbine logic provides outputs which 1) de-energize the electrical trip solenoid valve and 2) energize the mechanical trip solenoid valve. Either one or both of these will trip the turbine. Evaluation: There is no unreviewed safety question associated with this UFSAR change. No actual j change is being made to the plant. No Technical Specification changes are required. A J chanEe is required for UFSAR Section 10.2.2. 155 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Section 10.4.5.2

Description:

Section 10.4.5.2 was revised to clarify that additives for the Cooling Towers include 1) 4 Sulfuric Acid ,2) Algaecides, and if needed 3) Silt Dispersants. ) Evaluation: There is no unreviewed safety question associated with this UFSAR change. No actual i change being made to the plant. No Technical Specification changes are required. A change is required for UFSAR Section 10.4.5.2

                                                                                                             ]

I l l l

i L I:

                                               .('
    ; U.S. Nuclear Regulatory Ch April 1,1999 Page 215 of 247
      .'156      Type: UFSARChange                                                      Unit: 0

Title:

Change to UPSAR Section 10.4.6.2

        . E- ."#- This revision to UFSAR Section 10.4.6.2 corrects several inaccuracies associated with the process of backwashing the condensate polishing demineralirers. %e period for -

backwashing the condensate polishing domineralizers is based on accepted practices that . will ensure the service life of the filters. He changes from 17 cubic feet to 16 cubic feet l of resin reflects the current amount of resin that might be collected during normal I operation. The amount of resin is subject to change based on the evolutions being conduceed in the plant. De powered resin is no longer pumped to the Conventional Wastewater System. His decreases the amount of redwaste that would have to be

                         . disposed of should the Conventional Wastewater System become' contaminated. For all
                         . situations, the condensate polishing demineralizer backwash tank is allowed to settle and l~

only the supernatant water is decanted to the Decant Monitor Tank. Sampling of the

                         . Decant Monitor Tank will determine whether the water can be released via the
                        . Conventional Wastewater System or the Liquid Waste Recycle System. The resin in the l

condensate polishing demineralizer backwash tank will be sample and pumped to a suitable container. Evaluation: There is no unreviewed safety question associated with this UFSAR change. No Technical Speci0 cation changes are required. A change is required for UFSAR Section

                          ' 10.4.6.2.

i 159 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Section 12.5.2.1.3

Description:

UFSAR Section 12.5.2.1.3 was changed to note that the Radiation Protection Dosimetry Office was moved frcm the Administration Building to the Service Building. Evaluation: There is no unreviewed safety question associated with this UFSAR change. Neither the lL Administration Building nor the Service Building are Nuclear Safety Related areas. Relocation of Offices has no effect on any accident descrituxi in the UFSAR. No Technical Specification changes are required. A change is required for UFSAR Section 12.5.2.1.3. l i 1- < l j

                                                    ^

2

       - U.S. Nuclear Regulatory ca==

s

       - Apdl 1,1999                                                                                                      I Page 219 of 247 160     'Iype: UFSAR Change                                                     Unit: 0

Title:

Change to UFSAR Section 14.3.2.3.2,14.3.2.3.3,14.3.2.3.4

Description:

UFSAR Sections 14.3.2.3.2,14.3.2.3.3, and 14.3.2.3.4 were revised'to include the Dynanuc Rod Worth Measurement Technique. The Rod Swap Measurement method was left as an option; he Dynamic Rod Worth Measurernent technique and the review / acceptance criteria was added to Section 14.3.2.3 which describes the method used to measure the worth of the control rods. Evaluation: There is no unreviewed safety question associated with this UFSAR change. De Dynamic Rod Worth Measurement Technique has been reviewed and approved by the NRC in WCAP 13360-PA. No Technical Specification changes are required. A change is required for UFSAR Sections 143.2.3.2,14.3.2.33, and 14.3.2.3.4. 132 Type: UFSAR Change Unit: 0

                   'lltle: Change to UFSAR Section 5.3.1.4

Description:

UFSAR Section 5.3.1.4 was changed to delete text about BW1 maintaining full compliance with Regulatory Guides 1.43,1.50,1.71. This statement is not relevant to

                            - this subsection. This deletion 1) has no effect on the operation, design bases, or function of any structure, system or component 2) is supported by either high level design or lower level (supporting) design and/or operation-related documents. 3) does not affect any operation-related information 4) does not affect the Technical specifications, Evaluation: Dis change does not affect the operation, design basis or function of any structure, system or component. There are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are -

required for UFSAR Section 5.3.1.4 131 Type: UFSARChange Unit: 0

                   'lltle: Change to UFSAR Section 5.3.1.7

Description:

UFSAR Section 5.3.1.7 was changed to delete text in the last two paragraphs on page 5 39 that is not relevant or necessary. This deletion 1) has no effect on the operation, design bases, or function of any structure, system or component 2) is supported by either high level design or lower level (supporting) design and/or operation-related documents.

3) does not affect any operation-related information 4) does not affect the Technical i specifications.

Evaluation: Dis change does not affect the operation, design basis or function of any structure, i system or component. Dere are no unreviewed safety questions associated with these  !

UFSAR Changes. No Technical Specification changes are required. Changes are required for UFSAR Section 5.3.1.7 L1 l i

p 1 U.S.Neeleer Regulatory s'e

 .Ap:111,1999 Pase 220 of 247 f

133 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Section 5.3.3.1

Description:

UFSAR Section 5.3.3.1 was changed to add the following sentenence "De heatup rates imposed by plant operating limits are 30 degrees F per hour for normal operation, tad 60 degrees F per hour under abnormal or emergency conditions." This addition 1) has no effect on the operation, design bases, or function of any structure, system or component 2) is supported by either high level design or lower level (supposting) design and/or operation-related documents. 3) does not affect any operation-related information 4) does not affect the Technical Specifications. Evaluation: His change does not affect the operation, design basis or function of any structure, system or component. There are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are required for UFSAR Section 5.3.3.1. 129 Type: UFSAR Change Unit: 0

Title:

Change to Ulu AR Section 5.4.3.2

Description:

UFSAR Section 5.4.3.2 was changed to correct inaccurate technical infonnation. Concerning the Reactor Coolant System, the statement is made that "all joints and connections are welded except for the pressurizer code safety valves". However, the reactor head vent piping attached to the reactor head is also flanged in order to faciliatte removal of the head for refueling. This UFSAR Change adds this exception to the Reactor Coolant System piping description. Eva!uation: This design is an original plant configuration. Originally, the head vent was flanged and was a part of the Upper Head injection System. It was converted by a modification to serve as a reactor vessel head vent. Since this is a standard Westinghouse design and was considered in the original design calculation and license submittals, tiere is no effect on plant design, procedures, or ogeration. There are no unreviewed safety questions associated with this UFSAR Change. No Technical Specification changes are required. UFSAR Section 5.4.3.2 will be revised.

r: U.S. Nuclear Regulatory Conomission Apdl 1,1999 Page 221 of 247 130 Type: UFSAR Change Unit: t;

Title:

Change to UFSAR Section 5.4.3.3.1 Descripden: UFSAR Section 5.4.3.3.1 was changed to delete a reference to a nonexistent Table 5.9 -

                   " Reactor Coolant Chemistry Specification? 'Ihis change also deletes the title of the table and notes "a" and "b" that accompany the title to the table. Instead of including a table of reactor coolant specifications, the section will now reference Catawba Technical Specifications and EPRI Primary Water Chemistry Guidelines. These documents are the basis for the chemistry inonitoring program. References in note "a" to lithium limits during pre-startup testing prior to heat-up and cold hydrostatic testing are dated and not relevant to present operating conditions. Operating limits for lithium are included in the chemistry monitoring program derived from Catawba Technical Specifications and the EPRI Primary Water Chemistry Guidelines previously mentioned. Information in note            ,
                   "b" has been incorporated into the revised section 5.4.3.3.1.

Evaluation: This change does not change the intent, interpretation, or technical content of the section. There are no unrevicwed safety questions associated with this UFSAR Change. No Technical Specification change is required. UFSAR Section 5.4.3.3.1 will be revised. 134 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Section 5.4.3.3.3

Description:

UFSAR Section 5.4.3.3.3 was changed to correct information about the external cleanliness required for new construction and modifications to stainless steel piping in the Reactor Coolant System. The statement about halogen limits was extracted from a specification section that was applicable to internal cleanliness. The specification contains the correct information in another section. This change will insert the correct limits. Evaluation: This change does not change current practice, operating procedures, or plant design. There are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are required for UFSAR Section 5.4.3.3.3. l l i I i

U.S. Nuclear Regulatory Commission April 1,1999 Page 222 of 247 135 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Section 5.4.7.2.6

Description:

UFSAR Section 5.4.7.2.6 was changed to revise the description of the most limiting single failure related to the circulation of reactor coolant. This chnage 1) has no effect on the operation, design bases, or function of any structure, system or component 2) is supported by either high level design or lower level (supporting) design and/or operation-related documents. 3) does not affect any operation-related information 4) does not affect the Technical specifications. Evaluation: Dis change does not affect the operation, design basis or function of any structure, system or component. Here are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are required for UFSAR Section 5.4.7.2.6 143 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Section 6.2

Description:

UFSAR Section 6.2 was changed to add information on the Ilydrogen Mitigation System. The Catawba Nuclear Station Technical Specification Bases includes a description of the . I Hydrogen Mitigation System, but such an explanation was not provided in the UFSAR. ne information being added has been available in a document called the " Red Book". His addition will make the infomation more readily accessible. Evaluation: There are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are required for UFSAR Section l 6.2. l 145 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Section 6.3.2.5

Description:

UFSAR Section 6.3.2.5 was revised to more clearly state that direct hot leg injection from the Residual Heat Removal Pumps is nct credited in the safety analyses. One Train of Residual Heat Removal may be required to supply Residual Heat Removal auxiliary containment spray during this time and this Train should not be aligned for direct hot leg injection. Evaluation: There are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are required for UFSAR Section 6.3.2.5. l l l L.

m U.S. Nuclear R.,- " TiCommission

    ' April 1,1999 Pase 223 of 247 -

148- Type: UFSAR Change Unk: 0

Title:

Change to UFSAR Section 7.1.2.4

Description:

This UFSAR Revision dpdates a list of equipment which is not tested per UFSAR Section 7.3.' Section 7.3 describes testing of the Engineered Safety Features Actuation System output devices (Slave relays) as being performed by actuating their final devices. Actually there are several final devices that'cannot be tested with the plant at power. Section 7.1.2.4 lists some exceptions to the requirements of Section 7.3 but does not give a complete list. This change to the UFSAR will update the list of equipment in Section

                      - 7.1.2.4 to include all the equipment that is currently not being tested as described in Section 7.3.

Evaluation: There are no unreviewed safety questions associated with the revisions to UFSAR Section

                                  ~

7.1.2.4. The purpose of the Engineered Safety Features Actuation System is to mitigate damage to the core and Reactor Coolant System components, and ensure containment integrity in the event of an accident. In order to accomplish these design objectives the >

 -i                     Engineered Safety Features System has initiating signals which are supplied by the
                      ' sensors, transmitters, and logic components making up the various instrumentation channels of the Engineered Safety Features Actuation System (ESFAS). This actuation
                      - logic is required to be periodically tested when the reactor is in operation. Regulatory Guide 1.22 recognizes that there are certain pieces of equipment that can not be tested at power without disrupting reactor operation and provides acceptable methods of including the equipment in periodic testing while avoiding the undesirable effects of operating the equipment. He equipment that can not be tested at power is listed in the UFSAR in section 7.1.2.4 however, the current list of equipment was incomplete. This UFSAR change will add additional pieces of equipment to the listing in section 7.1.2.4. No physical changes to the plant will be made. No changes to the Technical Specifications are required. Changes are required for UFSAR Section 7.1.2.4.

I 149 Type: UFSAR Change , Unit: 0

Title:

Change to UFSAR Section 7.1.2.4.2

Description:

His UFSAR change corrects a discrepancy concerning crediting of the non safety related feedwater pump trip. The UFSAR currently correctly states that the containment integrity analyses do not assume tripping of the feedwater pump and that trip solenoids are not , nuclear safety related. A statement will be added to clarify the fact that the Feedwater Pump trip is credited with tripping in certain accident scenarios. ) Evaluation: This change is a clarification and does not affect the licensing basis of the plant. There are no unreviewed safety questions associated with this UFSAR revision. No changes to the l Technical Specifications are required. Changes are required for UFSAR Section 7.1.2.4.2. i

U.S. Nuclear Regulatory Comnaisision 3 Apdf 1,1999 j Page 224 of 247 150 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Section 7.7.1.2.1

Description:

Special tests (such as Dynamic Rod Worth Measurement) require that control rod banks be inserted and withdrawn at the maximum possible rate (limited by the mechanical properties of the drives and control). The maximum possible mechanical rate is 72 steps per minute in both manual and automatic. This number is specified in several sections of the UPSAR. However, in section 7.7.1.2.1 the bank speed is described at being fixed at 48/64 steps per minute for control and shutdown banks in manual operation. This revision allows (for special tests) for the adjustment of the bank speeds up to the ,  ; maximum 64 steps per minute for Control Banks and 72 steps per minute for Shutdown  ! Banks, while in manual operation. Any procedure that requires that rod speeds be increased will have a 10CFR50.59 evaluation specifically written for it. This 10CFR50.59 evaluation is intended solely for the UFSAR Section 7.7.1.2.1 revision to allow the maximum rod speed to be increased to the maximum mechanical limit while in manual operation. This revision to UFSAR section 7.7.1.2.1 will not adversely impact the plant in any manner. Evaluation: There is no unreviewed safety question associated with this UFSAR change. No Technical Specification changes are required. A change is required for UFSAR Section 7.7.5.2.1 l i l I

t i U.S. Nuclear Regulatory Cha-lan Apdf 1,1999 Pase 225 of 247 152 Type: UPSARChange Unit: 0

             'I1tle: Change to UFSAR Section 9.2.5.3.1
     . Desedption: This change to UPSAR Section 9.2.5.3.1 revises the description of calculations performed by Duke Power Company to confirm the ability of the Standby Nuclear Service Water Pond (SNSWP) to provide cooling water to the plant for up to 30 days in the event of simukaneous Loss of Coolant Accident (LOCA) on one unit, Loss of Offsite Power (IDOP) on bodi units, Loss of Lake Wylie, and shutdown of the non-LOCA unit. These calculations have been refined to more accurately reflect the performance of the Nuclear Service Water System and the SNSWP. Specific changes to the calculation include a more detailed listing of the heat inputs to the pond, corrections to the analysis of the Nuclear Service Water System to more accurately reflect the flow capacity of the system, changes to the most severe meteorological conditions calculation, and inclusion of resuhs of the Duke Power Gothic Computer Program for calculation of the decay heat from the reactor core on the LOCA unit. Previously the decay heat input to the calculation was taken from curves supplied by Westinghouse. Duke has taken responsibility for these calculations and performs them in-house, in all cases, the results of these calculations provided more conservative results. The heat input to the SNSWP was increased as a result of these calculations and the cooling water flow rate was decreased. The results showed that the SNSWP is adequately sized to support the plant under these most severe conditions.
      . Evaluation: There is no unreviewed safety question associated with this UFSAR change. No actual

change being made to the plant. No Technical Specification changes are required. A l change'is required for UFSAR Section 9.2.5.3.1 l

i e i

i JI' ' U.S. Neclear = . "- y r m L Apail1,1999 Page 226 of 247 i p 157 Type: UPSAR Change Unit: O l

Title:

Change to UPSAR Sections 10.'4.7.5.1,10.4.7.5.2,10.4.8.2

Description:

UPSAR Sections 10.4.7.5.1,10.4.7.5.2, and 10.4.8.2 were revised to correct the ' following discrepancies:

                            ' 1) Section 10.4.7.5.1. Alarms and Trips (CM), item 1, Hotwell low level, incorrectly states that the hotwell pumps trip on low-low hotwell level.

This statement is incorrect since this trip function was deleted per minor

                         ,        modification CE-1378 on Unit I and CE-1379 on Unit 2. These modifications were completed in 1990. It is inconectly stated that the IAw. Low flow Condensate Booster Pump trip has a five second time delay. The Unit 2 low flow trip was changed to 20 seconds per modification CE-1623 in 1988. Incorrect time delays for Main Feedwater Pump low-low flow 'and low-low suction pressure trips. Both units currently have 20 second time delays per minor mods CE-2196 on Unit I and CE2197 on Unit 2 completed in 1990 and 1989 respectively.

Since these modifications were previously evaluated under the - modification process they require no further evaluation.'

2) Section 10.4.7.5.2, Controls (CM). Steam Generator feedwater control valves, incorrectly implies that the Main control valves only are operated in automatic, it is incorrectly stated that the Main Feedwater bypass control valves are manually operated at low power levels. Since tie implementation of the Digital Feedwater Control System nmdifications CN 11168 on Unit 11 and CN-20544 on Unit 2, the Main Feedwater bypass control valves are operated in automatic from 0 to 100% power.

These modifications are complete and require no further evaluation.

3) Section 10.4.8.2, System Description (BB), is being revised to reflect that both Steam Generator Blowdown Heat Exchangers are normally in service instead of one as stated in the UFSAR.

Evaluation: There is no unreviewed safety question associated with this UFSAR change. The Steam l Generator Blowdown Heat Exchangers are a part of non safety related portions of tim Steam Generator Blowdown and Condensate Makeup Systems. 'the Steam Generator Blowdown System is not an accident initiator and the number of Heat Exchangers in operation will not increase the probability of the Condensate Makeup System initiating a loss of feedwater accident or turbine trip. Operation of both heat exchangers would not  ; increase the probability of a reduction in feedwater temperature accident since any l' temperature effects would be offset by the feedwater heaters downstream of the Heat Exchangers. No Technical Specification changes are required. A change is required for l UFSAR Sections 10.4.7.5.1,10.4.7.5.2,10.4.8.2

                 ~

l l. 1

                                                      \

L..

p. l-U.S. Nuclear Regulatory Co==66: l . Apdl1,1999 Pane 227 of 247 142 Type: UFSAR Change Unit: 0 1 .

Title:

Change to UFSAR Sections 5.4.7, 6.2.2.3,6.3.3.2, Table 5-30 Table 6-70, Table 6-87

Description:

UFSAR Sections 5.4.7,6.3.3.2 and Table 5-30: With respect to the Residual Heat Removal Pumps, additional discussion has been provided to describe the results of two

                        - different Residual Heat Removal Pump Net Positive Suction Head (NPSH) calculations performed by Westinghouse in response to two different Nuclear Regulatory Commission questions during the licensing process. Confirmatory analysis performed by Duke Power Company for Residual Heat Removal Pump NPSH at current licensing basis conditions is also described as being bounded by original UFSAR analysis. Table 5-30 NPSH data has been corrected to pump centerline elevation and maximum calculated nmout flow rate (first case) has been conected from 4500 gallons per minute (gpm) to 3800 gpm with corresponding NPSH values from station controlled documents. Assumed " Runout Flow" rate (second case) has been added to Table 5-30.

Section 6.3.3.2 and Table 6-87: With respect to the Centrifugal Charging and Safety injection Pumps, additional discussion has been provided to describe that the confirmatory analysis performed by Duke for pump NPSH at current licensing basis conditions remains bounded by original UFSAR analysis. Table 6-87 required NPSH data has been corrected to agree with values from station controlled documents and converted to pump centerline elevation. Maximum Flow Rate values have been changed to agree with values from station controlled documents. Section 6.2.2.3 and Table 6-70: With respect to the Containment Spray Pumps, additional discussion has been provided to describe that the confirmatory analysis performed by Duke for pump NPSH at current licensing basis conditions remains bounded by original UFSAR analysis. Table 6-70 " Containment Spray Pump Maximum Calculated Runout Flow Rate" value has been incorporated to agree with values from station controlled documents. NPSH data (Required and Available) for Design and Maximum Calculated Runout Flow Rates have been incorporated and converted to pump centerline elevation to agree with values from station controlled documents.  ; Evaluation: The UFSAR revision being evaluated provides additional discussion and clarifies existing I information in various sections and tables of the UFSAR. This information relates to the , i.5 design basis evaluation of Emergency Core Cooling System (ECCS) and Containment Spray Pump NPSH issues. No Technical Specification change is associated with this change, since no change is required to the surveillance criteria specified in the Technical Specifications for the ECCS Pumps and Containment Spray Pumps. The original NPSH analyses as conservatively performed by Westinghouse has been demonstrated in Duke  ; Power's confirmatory analyses to remain bounding for all design basis cases (minimum I safeguards, maximum safeguards, and any one pump failed with all others operating). , There is no need identified to throule pump flows or otherwise interfere with automatic l system responses. Therefore, the failure modes and effects described in the l corresponding UFSAR sections are not impacted. The fission product barriers of the pellet, clad, reactor coolant system primary pressure boundary and containment are not affected as a result of this change to the UFS AR. There are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are

                       - required. Changes are required for UFSAR Sections 5.4.7,6.2.2.3,6.3.3.2, Table 5-30, 4

o U.S. Nuclear Regulatory Conunission April 1,1999 Page 228 of 247 Table 6-70, and Table 6-87. 144 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Sections 6.2.2,6.2.2.2,63.2.1,63.2.2,63.2.5, .633 Table 6-89, Table 6-96 Table 9-15

Description:

These changes correct editorial and techniel discrepancies that have been discovered in these sections. Evaluation: There are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are required for UFSAR Section 6.2.2,6.2.2.2,63.2.1,63.2.2,63.2.5,633 Tah!c 6-89, Table 6-96, Table 9-15 151 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Sections 83.1.1.2.2 and 83.1.5.1

Description:

This change to UFSAR Section 83.1.1.2.2, "600 VAC dssential Auxiliary Fower System" consists of adding detail about cable protection practice. The issue originated from a 1988 audit which cited a discrepancy between cable ampacity ratings and the settings of breakers protecting certain cables. A detailed review showed that the existing breaker settings adequately protect the cables. The UFSAR addition says the specific cables can be protected by breakers set slightly higher than the standard 70% derated ampacity value. These settings are documented in the breaker setting calculation. This resolution was preferred because breaker trips had occurred during startup testing when the breakers were set lower. A modification was implemented in 1984 to raite the ! breaker settings after the trips occurred. The need for a UFSAR change was not recognized at the time the modification was implemented. The proposed activity is for a UFSAR change only and no change is being made to plant systems or equipment. Also included with this change is the addition of an editorial note to UFSAR Section 83.1.5.1. Evaluation: There is no unreviewed safety question associated with this UFSAR change. No actual change being made to the plant. No Technical Specification changes are required. A change is required for UFSAR Sections 83.1.1.2.2 and 83.1.5.1. {

n. m -

c

        . U.S. Nuclear Regulatory Conunission April 1,1999 Page 229 of 247 9

180 Type: UFSAR Change Unit: 0 11tle: Change to UFSAR Table 10-6 Desciiption: .UFSAR Table 10-6 specifies normal and maximum turbine exhaust temperatures for conditions of no steam bypass and with steam bypass. The current values listed are not correct. The numbers will be replaced with values based on a thermal performance l

                                                                                                     ~

analysis. Evaluation: 'Ihere is no actual change to the plant associated with this UFSAR change. The affected components are the low pressure turbines and condensers. The turbine exhaust temperatures were recalculated based on the known heat load to the condenser using methods of the Heat Exchanger Institute for Steam Surface Condensers. These values for turbine exhaust temperature are not used as inputs to any UFSAR Chapter 15 accident.'  ! There is no Unreviewed Safety Question associated with this UFSAR Change. No Technical Specification changes are required. UFS AR Table 10-6 will be revised. 179 Type: UFSAR Change Unit: 0

Title:

Change to UFSAR Table 7-14

Description:

This change to UFSAR Table 7-14 removes

• Upper Head injection" from the list of ESF Bypass Indications. Table 7 14 lists the safety related functions for which bypass indication is provided. The UFSAR is being updated to match the plant as-built condition. Nuclear Station Modifications CN-10910 for Unit I and CN-20299 for Unit 2 removed the Upper Head Injection systems and bypass indication logic. Table 7-14 was not revised at that time as it should have been.

Evaluation: There is no unreviewed safety question associated with this UFSAR change. There is no actual change to the plant. No Technical Specification changes are required. UFSAR Table 7-14 will be revised. i i l l

 =:

     ~

y U.S. Nuclear Regulatory Commission

 ' Apdl 1,1999 Page 230 of 247 161     Type: UFSAR Change                                                     Uilt: 0

Title:

Change to UFSAR Table 7 3, Table 15-7 and addition of a New Table

Description:

UFSAR Table 7-3, Table 15-7 and a New Table were revised /added to accomodate removal ofinformation from the Technical Specifications to the UFSAR. "Ihe Reactor Trip System Instmmentation Response Times Table currently resides in the Technical Specifications as Table 3. 3-2. During implementation of the Catawba Improved Technical Specifications this table will be relocated to the UFSAR. Evaluation: There is no unreviewed safety question associated with this UFSAR change. The current safety analyses are bounded because they assume the existence of the additional 1.5 second time delay, this longer response time acts as a penalty but is included in order to allow a single analysis to be used for Catawba Unit I and for McGuire Unit I and Unit

2. The change to a smaller allowable response time is more conservative and provides greater margin. No Technical Specification changes are required (except those involved with issuing the Catawba Improved Technical Specifications). Changes are required for UFSAR Table 7-3 and Table 15-7. Also a new UFSAR Table will be required.

l 141 Type:. UFSAR Change Unit: 0

             'Iltle: Changes to UFSAR Chapters 6 and 15

Description:

Fifty one changes were made to UFSAR Chapter 6 and Chapter 15 associated with the ) 1998 Interim Update. Twenty eight of these were editorial corrections. The changes ) correct omissions from previous UFSAR Updates or revise the results of a UFSAR analysis performed using NRC approved methods. Evaluation: None of these revisions increase the probability of an accident. There are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are required for UFSAR Chapters 6 and Chapter 15. 1 I l I I

p-l l. ! - U.S. Nuclear Regulatory ra==='8 ta= April 1,1999

          - Pane 231 of 247 t

r 126 Type: UFSAR Change y Unit: 0

                    - 'nties Changes to UFSAR Section 4.2.5.2 associated with calculation CNC-1553.26-00-0207

Description:

Dese changes are being made to correct inconsistencies associated with calculation CNC-1553.264)0-0207, the quality control practices for fabncation of fuel. These descriptions were oue<l=W and inconsistent with existing practices. Evaluation: Hese UFSAR changes document fabrication processes presently in place at the vendor facility. De practices mitigate the potential for loading the wrong enriched pellets into a fuel rod, and an operational process which mitigates the insertion of the correct enrichment fuel pellets into fuel rods, but in the wrong sequence. The enrichment controls I at the vendor facility provide the vehicle to ensure that the fuel is built to the specifications for which it has been analyzed for in core operation. De inadvertent loading and operation of one or more fuel assemblies into improper positions, loading a fuel rod during manufacture with one or more pellets of the wrong enrichment, or loading of a full fuel assembly during manufacture with pellets of the wrong enrichment. The changes being made to the UFSAR will not provide any avenue for the increase in the possibility of, nor the formulation of, any new or existing accident. The methods,

                             - segregation controls, and procedures being used at the vendor manufacturing facility will remain the same.

A fuel handling accident is characterized as the loss of the entire gas gap radioactivity

                            - inventory of the damaged fuel assembly. In examining the explanation of the gap inventory, the gap activities are determined utilizing the model suggested in Regulatory Guide 1.25. There are no unreviewed safety questions associated with these UFSAR Changes. No Technical Specification changes are required. Changes are required for UFSAR Sections 4.2.5.2,4.2.5.2.1,4.2.5.2.3.d 4.2.5.2.3.f,4.2.5.2. 3.g, and 4.2.5.2.6.

127 Type: UFSARChange Unit: 0

Title:

Changes to UFSAR Section 5.2.3.4.6

Description:

UFSAR Section 5.2.3.4.6 was changed to correct various inaccurate technical information and delete, add, or change verbiage to clarify information. The corrections and changes

                                                                                                                           )

are limited to those that 1) have no effect on the operation. design bases, or function of any structure, system or component. 2) are supported by either high level design or lower .

                            ' level (supporting) design and/or operation-related documents. 3) do not affect any operation-related information 4) do not affect the Technical Specifications.

Evaluation: Dese changes do not affect the operation, design basis or function of any structure, ] system or component. Dere are no unreviewed safety questions associated with these  ! UFSAR Changes. No Technical Specification changes are required. Changes are l required for UFSAR Section 5.2.3.4.6. j l l l be

U.S. Nuclear Regulatory Conunission April 1,1$99 Paste 232 of 247 128 .- Type: UFSAR Change Unit: 0

Title:

Changes to UFSAR Section 5.4.3.2

Description:

UFSAR Section 5.4.3.2 was changed to correct text to explain that the Nitrogen supply line for the pressurizer relief tank is stainless steel instead of carbon steel. His change 1) has no effect on the operation. design bases, or function of any structure, system or component 2) is supported by either high level design or lower level (supporting) design and/or operation-related documents. 3) does not affect any operation-related information

4) does not affect the Technical specifications.

Evaluation: These changes do not affect the operation, design basis or function of any structure, system or component. Here are no unreviewed safety questions associated with these UFSAR Changes. No Technical Spec,.ification changes are required. Changes are required for UFSAR Section 5.4.3.2. 226 Type: UFSAR Change Unit: 0

Title:

Editorial Changes to the UFSAR identified in 1998

Description:

These changes to the UFSAR involve editorial changes to the UFSAR that were identified ' in the 1998 UFSAR Update process. The changes include corrections of misspellings, corrections of typographical errors, corrections ofincorrect punctuation, correction of capitalizations, and corrections of erroneous numbering. None of these changes have any effect on installed systems, structures, or components of the plant. Evaluation: These changes have no technical impact on the UFSAR. There are no unreviewed safety questions associated with these editorial changes. No Technical Specification changes are required. 295 Type: UFSAR Change Unit: 0

Title:

UFSAR Change to Chapter 10 Table 10-11 l

Description:

UFSAR Chapter 10 Table 10-1 I was changed to show that the Condensate Booster Pump Minimum Continuous flow is 1600 gallons per minute, not 3000 gallons per minute. The manufacturer's drawing shows a value of 1600 gallons per minute. Evaluation: There are no unreviewed safety questions associated with this change to the UFSAR. The change improves the margin of safe operation of the pumps and does not involve radiological consequences. No Technical Specification changes are required. A change is required for UFSAR Chapter 10 Table 10-11. I l l o

     . U.S. Nuclear R4 "- y Comunission April I,1999 Pase 233 of 247 294       Type: UFSAR Change                                                       Unit: 0

Title:

UFSAR Change to Chapter 9 Table 9-1

Description:

UFSAR Chapter 9 Table 9-1.was revised to clarify that there is a Fuel Pool Cooling Domineralizer Strainer on Unit 1 only. De equivalent item on Unit 2 was removed in 1994 per modification CE-3955. His UFSAR change incorporates a change that should have been made with the original modification. Evaluation: There are no unreviewed safety questions associated with this change to the UFSAR. De change to the plant was made and evaluated per 10CFR50.59 under modification CE-3955. This change makes the UFSAR revision that should have been made at the time the modification was done. No Technical Specification changes are required. Changes are required for UFSAR Chapter 9 Table 9-1. 170 ' Type: UFSAR Change Unit: 0

                 'Iltle: UFSAR Change to correct discrepancies associated with core components, fuel assembly design and related NRC approved thermal-hydraulic and mechanical analysis methodologies

Description:

This UFSAR change corrects technical discrepancies found within several se:tions of the J UFSAR. Corrections were made to the following sections to remove reference to a specific value for the DNBR. Section 7.2.2.2.1 " Trip Setpoint Discussion" states in 3 different places the DNBR =

                                                ~

1.30. This section was updated to remove any reference to a specific value. Section 7.7.2 " Analysis" This section states the DNBR limit is equal to 1.30. This section was updated to remove any reference to a specific value.  ; Section 8.2.2.1 " Grid Stability Analysis" This section states the DNBR limit is equal to l 1.30. This section should be updated to remove any reference to a specific value. Section 5.4.1.1 " Design Basis" his section states the DNBR limit is equal to 1.30. This i

section should be updated to remove any reference to a specific value.

Item 5: UFSAR Section 7.2.1.2.4 " Limits, Margins, and Set Points" states " Fuel rod l maximum linear power for determination of protection setpoints = 18.0 kw/ft." He linear ]

                          ' heat rate to melt limit is a function of the burnup and can not be described by a single value. "18.0 kw/ft" was replaced by " limiting value" and a reference back to Section 4.2 was made.
        . Evaluation: There are no unreviewed safety questions associated with this activity. These changes are         i
   ,                        essentially editorial in that a specific value for DNBR is removed and replaced with a      ;

reference to another UFSAR section where the information can be found. No Technical Specification changes are required. A revision is required for UFSAR Section 5.4.1.1, l 7.2.1.2.4, 7.2.2.2.1, 7.7.2, 8.2.2.1, and Table 3-4. , I l

k-- l t U.S. Nuclear Regulatory Cos=-i-laa

   ' Apdf 1,1999 Pane 234 of 247 :

232: Type: UFSAR Change Unit: 0 Mtle: UFSAR Change to Section 10.2.2 concerning bulk hydrogen - p

Description:

This UFSAR update for section 10.2.2, will reflect the current method in which hydrogen . is acquired in support of the Bulk Hydrogen System. Two original design options existed to maintain an adequate supply of hydrogen to the six storage cylinders. Due to poor reliability and high cost, the Hydrogen / Oxygen Gas Generator has been abandoned in place (per Nuclear Station Modification CN-50364. Revision 1). Therefore the option of

                    ~ taking hydrogen delivery from offsite suppliers will be utilized. A description of the control and delivery of hydrogen will be added to UFSAR Section 10.2.2.

! Evaluation: UFSAR Section 1.7 references Regulatory Guide 1.91: " Evaluations of Explosions Postulated to Occur on Transportation Routes near Nuclear Power Plants". In accordance with the methodology of Regulatory Guide 1.91, calculation CNC-1435.00-00-0029 evaluated the risk for the transportation route for the delivery of hydrogen to be sufficiently low and therefore acceptable. No Unreviewed Safety Questions have been created by this change. There has been no increase in the probability of occurrence of an accident previously evaluated in the UFSAR. There has been no increase in the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR. 'Ihe proposed activity.. will not increase the consequences of an accident previously evaluated in the UFSAR, nor

                    ' will the proposed activity increase the consequnces of a malfunction of equipment important to safety previously evaluated in the UFSAR. The proposed activity has not created the possibility for an accident of a dilferent type than any evaluated previously in the UFSAR. The proposed activity has not cicated the possibility for a different type of malfunction of equipment important to safety than any evaluated previously in the

! UFSAR and no reduction in the margin of safety as defined in the basis for any Technical Specification has occurred. No Technical Specification changes are required. A change is l l required for UFSAR Section 10.2.2 i l ! 231 Type: UFSAR Change Unit: 0

            - Mile: UFSAR Change to Section 10.2.2 concerning Main Turbine Controls

Description:

UFSAR Change to Section 10.2.2 revises the description of the operation of a solenoid that was previously a dual coil unit that was replaced with a single coil unit. This involves the operation of the Main Turbine. Evaluation: There are no unreviewed safety questiions associated with this change. The change has [ no significant effect on the operation, design basis or function of the turbine. No Technical Specification changes are required. A change is required for UFSAR Section 10.2.2 L_

y U.S. Nuclear Regulatory Comunission April 1,1999 Page 235 of 247

     '276      Type: UFSAR Change                                                      Unit: 0 111tle: UFSAR Change to Section 10.3.2 and 10.4.9.2 concerning Auxiliary Feedwater Turbine Driven Pump Steam Supply line heat tracing

Description:

UFSAR Sections 10.3.2 and 10.4.9.2 were revised to describe Auxiliary Feedwater Turbine Driven Pump Steam Supply line heat tracing. His heat tracing is relied upon to ensure operational readiness of the system. The heat tracing prevents damage to the

                      ' Auxiliary Feedwater Tubine Driven Pump by preventing the formation of condensate in the pump's steam supply line. De heat tracing is electrical.

Evaluation: Heat Tracing was used on the Auxiliary Feedwater Turbine Driven Pump Steam Supply line to help prevent damage to the pump.The Auxiliary Feedwater Pump Turbine is designed to accept water slugs without sustaining damage; however, water slugging may cause overspeed trips. Industry experience has shown that the turbine is most susceptible to this during turbine startup. This UFSAR change adds information to the UFSAR to i describe the as built configuration of the plant. There are no actual plant changes made in association with this UFSAR Change. No Unreviewed Safety Questions are created by

                                                                                                                     )
                                                                                                                     )

this UFSAR change. No Technical Specification changes are required, Changes are ) required for UFS AR Sections 10.3.2 and 10.4.9.2. 248 ' Type: UFSAR Change Unit: 0 l

Title:

UFSAR Change to Section 10.4.4.4 concerning Atmospheric Dump Valve Testing

Description:

A discrepancy was identified hetween the SER and the UFSAR with reference to the atmospheric dump valves. Currently the atmospheric dump valves are stroked at an eighteen month frequency. Section 10.4.4 of the Catawba SER implies that the atmospheric dump valves as well as the condenser dump valves should be tested quarterly. Section 10.4.4.4 of the UFSAR only mentions testing the condenser dump valves quarterly. This change to UFSAR Section 10.4.4 will delete the requirement for , stroking the condenser dump valves once per quarter. l Evaluation: There are no unreviewed safety questions associated with this UFSAR change. Extending l the condenser dump valve testing frequency will not increase the probability that the l valves will malfunction. The valves will still be tested on an eighteen month frequency as the atmospheric dump valves are. At Catawba there has been no history of valve problems with either the condenser dump valves or the atmospheric dump valves that would preclude extending the test frequency of the condenser dump valves. Here is no credit taken for the condenser dump valves operation in any safety analysis for any i accident described in the UFSAR. No Technical Specification changes are required. ) UFSAR Section 10.4.4.4 will require a change. I l l

1 U.S. Nuclear Regulatory Commission April 1,1999 Page 236 of 247 l l 287 Type: UFSAR Change Unit: O l

Title:

UFSAR Change to Section 5.4.13.3 , 1

Description:

His change to UFSAR Section 5.4.13.3 adds information about Pressurizer Power  ! Operated Relief Valve testing. Pressurizer Power Operated Relief Valves are tested for stroke time and discharge capacity through results from an Electic Power Research Institute (EPRI) PWR Safety and Relief Valve Test Program. This UFSAR change adds that the test program was performed i s response to NUREG 0737 and is documented in EPRI Report NP 2770. Evaluation: Dis change adds information about valve testing. He change has no effect on plant equipment. T.iere a. e no unreviewed safety questions associated with this change. No Technical S;ccificatir,n changes are required. A revision is required for UFSAR Section 5.4.13.3. l 271 Type: UFSAR Change Unit: 0

Title:

UFSAR Change to Section 5.4.2.1.3

Description:

UFSAR Section 5.4.2.1.3 was changed to add the Advanced Amine chemistry control program for the Steam Generators. This process was previously added per chemistry procedure OP/0/B/6250/10. Evaluation: There are no unreviewed safety questions associated with this UFSAR change. This revision has no effect on the operation, design bases, or function of any system, structure or component. No Technical Specification changes are required. No UFSAR changes are required. 317 Type: UFSAR Change Unit: 0

Title:

UFS AR Change to Section 5.4.7.2.5

Description:

UFSAR Section 5.4.7.2.5 was changed to delete a portion of a sentence. He sentence formerly stated "The effects ofleaks will be detected in the control room via the floor drain system alarms and area radiation monitors." He phrase "the floor drain system alarms and" will be removed because these alarms were removed by modification CE-5079. Evaluation: This changes involves correcting inaccurate technical information. There is no change to the operation of the plant or to the design basis or to the function of the affected systems, structures and components. There is no unreviewed ssfety question associated with this change. No Technical Specification changes are required. A change is required for UFSAR Section 5.4.7.2.5. 1 L I

U.S. Nuclear Regulatory Commission April 1,1999 Page 237 of 247-293 Type: UFSAR Change Unit: 0

Title:

UFSAR Change to Section 9.3.4.2.3.1 concerning seal leakoff for the Centrifugal Charging Pumps

Description:

This change to UFSAR Section 9.3.4.2.3.1 corrects the description of the seal leakoff configuration associated with the centrifugal charging pumps. The previous UFSAR description stated that the leakage was contained and not exposed to atmosphere. This statement is not correct since the configuration of the seal leakoff for the centrifugal charging pumps is the same as it is for the safety injection pumps. He seat leakoff is exposed to the pump room atmosphere, however it is contained and collected in drainage system piping. He program for monitoring Emergency Core Cooling System (ECCS) pump seal leakage and the calculational basis for the limits associated with the total ECCS pump seal leakage is not affected by this correction to the UFSAR. he design of the Auxiliary Building Ventilation System is consistent with the existing configuration. Evaluation: There are no unreviewed safety questions associated with this change to the UFSAR. No Technical Specification changes are required. Changes are required for UFSAR Secftion 9.3.4.2.3.1. 230 Type: UFSAR Change Unit: 0

Title:

UFSAR Change to Section 9.5.6.2.2

Description:

This change to UFSAR Section 9.5.6.2.2 deletes the Nuclear Service Water supply temperature and flow rate in the Diesel Generator Starting Air System Aftercoolers. UFSAR Section 9.5.6.2.2 states that the Nuclear Service Water System supplies cooling water to the Diesel Generator Starting Air System Aftercoolers at a flow rate of 12.5 gpm and that the water accepts a temperature increase from 88 to 89.6 degrees F. The flow rate and temperature of the Nuclear Service Water System through the Diesel Generator Starting Air System Aftercoolers is more realistically defined by range of temperatures and flows that bound the operation of the system. It is misleading to show a single operating point that may not be the actual flow or temperature at any given time. Design calculations exist that show the design bases for these systems and associated systems as being maintained over a range of temperatures that include the single operating point currently shown in the UFSAR. Evaluation: There is no unreviewed safety question associated with this UFSAR Change. The emergency diesel generators and their support systems are not accident initiators. No physical change is being made to any system structure or component. The change involves removing a reference to a single operating point that is bounded by the operating limits currently specified in the UFSAR. No Technical Specification changes are i required. A change is required for UFSAR Section 9.5.6.2.2 I 1 L i

U.S. Nuclear Regulatory Conunission Apdl 1,1999 Page 238 of 247 1 297 Type: UFSAR Change Unit: 0

Title:

UFSAR Change to Section 9.5.7.2.1 Revise Lube Oil Sump Tank Volume to reflect as-built conditions

Description:

UFSAR Section 9.5.7.2.1 was changed to revise Diesel Generator Lube Oil Sump Tank Volume to reflect as-buik conditions. During implementation of the Improved Technical Specifications, it was noted that the actual size of the vendor supplied diesel generator tube oil sump tanks is 579 gallons versus 700 gallons. The vendor supplied tank drawings listed the tank as 700 gallons and this value was used in the development of the UFSAR. This change is considered editorial since the volume of oil required to support emergency diesel operability is not being changed nor is the normal operating system volume being changed. No operating system parameters are being affected by this change. Evaluation: There are no unreviewed safety questions associated with this UFSAR change.The i change is editorial. No Technical Specification changes are required. A change is required for UFS AR Section 9.5.7.2.1. l l l 318 Type: UFSAR Change Unit: 0 l

Title:

UFSAR Change to Sections 5.1,5.4.1.3.8, and 5.4.1.5.1

Description:

UFSAR Sections 5.1,5.4.1.3.8, and 5.4.1.5.1 reference a reactor coolant pump overspeed resulting from a turbine overspeed of 20 percent. The turbine overspeed is limited to 111.5 percent by the turbine protection system. Since reactor coolant pump speed is directly proportional to the turbine overspeed, the reactor coolant pump overspeed is also limited to 111.5 percent Evaluation: This changes involves correcting inaccurate technical information. There is no change to the operation of the plant or to the design basis or to the function of the affected systems, r structures and components. There is no unreviewed safety question associated with this l change. No Technical Specification changes are required. A change is required for UFSAR Sections 5.1,5.4.1.3.8, and 5.4.1.5.1. The corrected number (111.5 percent) is lower than the number currently given and the pump is designed to a number greater than l~ 20 percent. 1 I I l l

l l l f U.S. Nuclear Regulatory Commission l April 1,1999 Page 239 of 247 4 290 Type: UFSAR Change Unit: 0

Title:

UFSAR Change to Sections 5.4.10.1.1 and 5.4.15

Description:

This UFSAR change adds information about the station's response to NRC Dulletin 88-11 concerning Pressurizer Surge Line Thermal Stratification. The additional information confirms the adequacy of the existing design. References were added to relevant correspondence and analyses. Evaluation: There is no change to the plant associated with these UFSAR changes. The changes add information about compliance with NRC Bulletin 88-II. nere are no unreviewed safety questions associated with these changes. No Technical Specification changes are required. Changes are required for UFSAR Sections 5.4.10.1.1 and 5.4.15 288 Type: UFSAR Change Unit: 0

Title:

UFSAR Change to Sections 5.4.13.4 and 5.4.15

Description:

This UFS AR change inserts information about how Catawba Nuclear Station complies with NRC Generic letter 9046, Generic Issue 70 about Power Operated Relief Valve and Block Valve Reliability and Generic Issue 94 about Additional Low Temperature Overpressure Protection for Light Water Reactors. A reference was added to a letter which addresses Duke Power's response to Generic letter 90-06. Evaluation: There is no change to the plant associated with this UFSAR change. There is no unreviewed safety question. No Technical Specification changes are required. A change is required for UFSAR Section 5.4.13.4 and 5.4.15. 291 Type: UFSAR Change Unit: 0 J

Title:

UFS AR Change to Sections 5.4.7.2.5 and 5.4.15

Description:

his UFSAR change adds information about the station's response to NRC Generic Letter 87 12 about " Loss of Residual Heat Removal while the Reactor Coolant System is i partially filled l' and NRC Generic Letter 88-17 about

• Loss of Decay IIcat Removal". As I a result of these two generic letters an extensive review was performed on the physical  ;

plant configuration, training programs, administrative procedures and programmatic enhancements. References were added to UFSAR Section 5.4.15 concerning these generic letters. 1 Evaluation: There is no change to the plant associated with these UFSAR changes. The changes add information about compliance with NRC Bulletin 88-11. Here are no unreviewed safety questions associated with these changes. No Technical Specification changes are required. Changes are required for UFSAR Sections 5.4.7.2.5 and 5.4.15. u

1 I i l. f- U.S. Nuclear Regulatory Conunission l April 1,1999 Page 240 of 247 f

   '292     Type: UFSAR Change                                                     Unit: 0

Title:

UFSAR Change to Sections 6.l.1.2,6.1.1.2.1,6.2.6.1,63.3,9.1.4.2.2 Table 6-2, Table 9-15, and Table 9-23

Description:

Dese UFSAR changes resolve editorial discrepancies which have been identified. The discrepancic3 are associated with boa,n concentration of the Refueling Water Storage Tank and Cold Leg Accumulators, as well as the maximum allowable containment leakage. These changes bring the UFSAR discussions into concurrence with existing Station Technical Specifications, owner controlled documents, and supporting safety analyses Which were previously evaluated per 10CFR50.59 or changes to the Technical Specifirstions. Evaluation: There are no unreviewed safety questions associated with these changes. Nr, Technical Specification changes are required. Changes are required for UFSAR Sections 6.1.1.2, 6.1.1.2.1, 6.2.6.1,6.3.3, 9.1.4.2.2. Table 6-2, Table 9-15, and Table 9-23. 312 Type: UFSAR Change Unit: 0

Title:

UFSAR Change to Table 5-1

Description:

This change involves revisions to information in UFSAR Table 5-1 Page 1 of 2 and Table 5 1 page 2 of 2. The specific items that were changed are

1. Total System Volume including Pressurizer and Surge Line

2. System Liquid Volume, including Pressurizer Water level

3. Thermal MW

4. Million BTU /hr

5. Thermal Design Flow, Loop gpm

6. Reactor Million BTU /hr

7. Reactor Coolan' Temperature at Vessel Ou%

8. Reactor Coolant Temperature (Core Average)

9. Best Estimate Flow (gpm)

10. Pump Design Point, Flow (gpm)

11. Pump Design Point,licad

12. Mechanical Design Flow

13. Steam Generator Delta P. psi

14. Pump IIcad, Feet Evaluation: This change involves correcting inaccurate technical information. There is no change to the operation of the plant or to the design basis or to the function of the affected systems, structures and components. There is no unreviewed safety question associated with this change. No Technical Specification changes are required. A change is required for UFSAR Table 5-1 Page 1 of 2 and Table 5-1 Page 2 of 2.

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p g U.S. Nuclear Regulatory Comunission April 1,1999 Page 241 of 247 313 Type: UFSAR Change Unit: 0 Utle: UPSAR Change to Table 5-10

Description:

UFSAR Table 5-10 was revised because it included several statements that implied that the Volume Control Tank is used for detecting Reactor Coolant System Leakage. He actual method used is described in Technical Specification 3.4.6. Evaluation: his changes involves correcting inaccurate technical information. There is no change to the operation of the plant or to the design basis or to the function of the affected systems, structures and components. %ere is no unreviewed safety question associated with this change. No Technical Specification changes are required. A change is required for UFSAR Table 5-10. 314 Type: UFSAR Change Unit: 0

Title:

UFSAR Change to Table 5-29

Description:

Table 5-29 was revised to show that reactor coolant system initial temperature is "< 350" degrees instead of *~ 350 degrees" Evaluation: Dis changes involves correcting inaccurate technical information. There is no change to the operation of the plant or to the design basis or to the function of the affected systems, structures and components. There is no unreviewed safety question associated with this change. No Technical Specification changes are required. A change is required for UFSAR Table 5-29 316 Type: UFSAR Change Unit: 0

Title:

UFSAR Change to Table 5-30

Description:

UFSAR Table 5-30 was changed to correct several values associated with the Residual Heat Removal System Heat Exchangers. Specific changes were:

1. Out Temp, Deg F, Tube Side changed from 119.8 to 119.0.

2. Out Temp, Deg F Shell Side changed from 117.3 to 117.5.

3. Design Flow,1b/hr, Shell Side changed from 2.47510E6 to 2.4810E6.

4. Design Heat Removal Capacity, BTU /hr changed from 30.510E6 to 30.9610E6.

Evaluation: This changes involves correcting inaccurate technical information. There is no change to the operation of the plant or to the design basis or to the function of the affected systems, structures and components. Here is no unreviewed safety question associated with this change. No Technical Specification changes are required. A change is required for UFSAR Table 5-30.

m U.S. Nuclear Regulatory Conunission April 1,1999 Page 242 of 247 311 Type: UFSARChange Unit: 0

Title:

UFSAR Change totable 5 30

Description:

UFSAR Table 5-30 was changed to revise the

• Estimated UA" for the Residual licat Removal licat Exchanger, from 1.91 x 10E6 BTU /hr degree F to 2.07 x 10E6 BTU /hr degree F. This new value was calculated as the product of the heat exchanger area and the overall coefficient of heat transfer.

Evaluation: This change involves correcting inaccurate technical information. There is no change to the operation of the plant or to the design basis or to the function of the affected systems, structures and components. There is no unreviewed safety question associated with this change. No Technical Specification changes are required. A change is required for UFSAR Table 5-30. 315 Type: UFSAR Change Unit: 0

Title:

UFSAR Change to Table 5-31 Page lof 7

Description:

UFSAR Table 5 31 Page lof 7 Failure Modes and Effects, Items 1 and 2, the 1,2NCPT5120 (PB-405 A) and 1,2NCPT5140 (PB-403A) interlock setpoint is stated as 385 psig. The correct value is 385.5 psig. Evaluation: This changes involves correcting inaccurate technical information. There is no change to the operation of the plant or to the design basis or to the function of the affected systems, structures and components. There is no unreviewed safety question associated with this change. No Technical Specification changes are required. A change is required for UFSAR Table 5-31 Page 1 of 7. 1

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l U.S. Nuclear Regulatory Ceaunission l Apdf 1,1999 L Page 243 of 247 1

       '298    Type: UFSAR Change                                                     Unit: 0
                'ntie: UFSAR Changes associated with conversion to the Improved Technical Specifications

Description:

Many changes were made to the UFSAR associated with conversion to the Improved Technical Specifications. The changes primarily involve relocation of existing Technical Specification requirements to the UFSAR (this relocation was approved by the NRC as a part of the Improved Technical Specification conversion process). The changes also 1 involve other administrative changes made to the UFSAR as a part of the improved Technical Specification Implementation effort, such as reference changes. Evaluation: 'Ihe changes described above were evaluated in conjunction with the requirements of 10CFR50.59 and were fouad to involve no unreviewed safety questions. No technical i requirements of any of the relocated technical specification material were affected by the changes. The manner in which the plant and its accident mitigating equipment are designed, operated and maintained was not affected by these changes. No accider.t probabilities or consequences of equipment malfunctions were impacted. The possibility

                       . for any new type of accident or malfunction was not created. No safety margins were reduced by the proposed changes. Therefore the changes were performed under 10CFR50.59. No Technical Specification changes are required other than those approved by the NRC in the applicable license amendment. Changes are required for UFSAR Sections 2.4.13.5, 2.4.14, 3.1, 3.7.4.5, 3.8.1.7, 4.2.1.3.1, 4.3.1.5, 4.4.1.10, 4.6.3, 4.6.4, 5.2.3.2.1, 5.2.5.1, 5.2.5.2.3, 5.3.1.6, 5.4.2.5.4, 5.4.12.1, 5.4.12.3, 6.2.1.1.3.1, 6.2.1.3.2.6, 6.2.6.1, 6.2.6.4, 7.1, 7.1.2.1.1, 7.11.2.4.2, 7.2.1.1.2, 7.2.1.2.4, 7.2.1.2.6, 7.2.2.2.3, 7.3.1.2.6, 7.6.3.3.9, 7.6.6.3.9, 7.6.7.3.9, 7.6.8.3.9, 7.6.9.3.9, 7.6.11.3.9, 7.6.16.3.9, 7.6.21.3.9, 8.1.5.2, 8.3.1.1.3.11, 9.3.4.2.3.8, 9.5.1.1, 9.5.5.4, 9.5.8.4, 10.4.6.2,i1.2.2.2.5.6, i1.3.2.5.2, i1.4.7, 11.5.1.2.1.8, i1.5.2.1, 11.5.2.4,i1.5.2.5, 12.5.3, 13.4.2.1, 13.4.2.2, 13.5.1.2, 14.2.10.1, 14.3.4,15.7.1.2, 15.7.1.3, 15.7.4.2.3, 16.2, and 17.1.1.2. Changes were made to Tables 6-77, 7 3, 7-4, 7-15, and 8-6.

Tables 5-40 and 5-41 were added. l i i i l 4

U.S. Nuclear Regulatory Conunission April I,1999 Page 244 of 247 - 125 Type: UFSAR Change Ualt: 0

Title:

Update of UFSAR Information on Weight of Zircaloy in the MK-BW fuel assembly design and Material for the BWFC burnable poison rod cladding

Description:

Calculation DPC-1553.26-00-0127 describes an Update of UFSAR Information on Weight of Zircaloy in the Mk-BW fuel assembly design and Material for the BWFC burnable poison rod cladding. This analysis evaluated the zircaloy weight and BWFC burnable poison cladding material updates. The zircaloy weight in UFSAR Tables 4-1 and 4-4 is not used in any UFSAR analyses. The mass of zircaloy in a Mk-BW fuel assembly in a core reported in the CNS UFSAR includes the zircaloy in the fuel rods, end caps, grid spacers, guide tubes, and thimbles. The mass of zirconium assumed in hydrogen production analysis (CNC 1552.08-00-0194, Rev.1) is based on the mass of zircaloy in the MK BW assembly minus the spacer grids for the active fuel height. This was conservatively calculated assuming a mass of zircaloy in a MK-BW fuel assembly slightly higher than the proposed update. The mass of zirconium assumed in the hydrogen production analysis (CNC-1552.08-00-0194, Rev.1) envelopes the proposed update. Ilydrogen production in the reactor and inside containment is important for the consideration of possible hydrogen deflagration or burn inside the reactor / containment. A hydrogen deflagation could result in a significant stress to the reactor and containment structures. Hydrogen mitigation and control systems are used to maintain the hydrogen concentration ler.s than the lower flammability limit of four volume percent following a LOCA to minimize the potential for hydrogen deflagration in containment. The amount of hydrogen produced following a LOCA is calculated to ensure that these systems have adequate capacity to maintain the hydrogen concentration less than four volume percent in containment. This data is also used to design the containment spray system to ensure i that containment internal pressure can be maintained less than 15 psig. Evaluation: The mass of zircaloy in a MK-BW fuel assembly used to calculate the mass zirconium in the active fuel height in hydrogen production analysis envelopes the proposed change. Therefore, the severity of the damage to the reactor / containment due to a hydrogen deflagration or burn is bounded by the current UFSAR analyses. This update will not { adversely impact the capability of a Safety System or Component to mitigate the consequences of a accident previously evaluated in the UFSAR. That is, containment integrity will not be compromised. This update will not adversely impact the assumptions used for hydrogen production, control, or mitigation. There are no new pathways for release of radioactive effluents to the environment than previously evaluated in the UFSAR. There are no new failure modes or rr chanisms due to this update. Therefore, there are no Unreviewed Safety Questions associated with the update of the weight of zircaloy in a MK-BW fuel assembly in the CNS UFSAR. Changes are required for UFSAR Table 4-1 and Table 4-4. No Technical Specification changes are required. m

i U.S. Nuclear Regulatory em d-sa. Apdl 1,1999 Pase 245 of 247 9 314 Type: UFSAR Changes Unit: 0

            'Dde: USFAR Change to Table 9      - DescripGom: His UFSAR change involves revising the Diesel Generator Jacket Water System Low Pressure alarm setpoint in Table 9-40 to reflect as-built conditions. De low-pressure alarm is currently set at 10 psig decreasing , although UFSAR Table 9-40 lists the setpoint as 12 psig decreasing.

De Emergency Diesel Engines at Catawba were manufactured by Transamerica Delaval,

                  ' which has been sold to Cooper Energy Services. The engines were supplied, with several system pressure and temperature alarms to annunciate at predetermined setpoints. The purpose of these alarms is to provide warning ofimpending diesel problems and to show system degradation. The setpoint for the jacket water low pressure alarm was originally 12 psig decreasing. The setpoint was changed per approved Duke Power Guidelines in 1984 to 10 psig decreasing. The vendor via a letter dated 5/3/84 also supported the setpoint changes. UFSAR Table 9-40 was not revised in 1984 when the setpoint was lowered.

Evaluation: De Emergency Diesel Generators are provided with a closed loop cooling water system. The system is designed with an engine driven pump that circulates water through a heat exchanger and the engine. Since the pump is engine driven, it is a constant speed positive displacement pump. The purpose of the pump is to circulate cooling water through the j engine to maintain normal operating temperatures. The important Diesel Engine Cooling Water system parameter to monitor to determine system health is engine differential temperature. The vendor states that this value should be less than 15 degrees F . The differential Jacket water temperature across the engines at Catawba is less than 10 degrees F. Cooper Energy Services has again reviewed the 10 psig lowjacket water pressure setpoint

                  . being used on the Catawba Diesel Engines and they do not have any concerns as long as Diesel Engine Cooling Water system engine differential temperature is less than 15            ;

degrees. There have been no modifications to the engines to change the jacket water l system that would alter operating pressure. The engines have always operated with a l Jacket water pressure between 12 and 14 psig . The setpoint for lowJacket water pressure  ! was lowered in 1984 because the annunciator would not clear during engine operation. De instrument has a 2 psig dead band that must be reached to clear the alarm. If the setpoint were 12 psig, the Diesel Engine Cooling Water system would have to reach 14 psig to clear the alarm. With a normal Diesel Engine Cooling Water system operating

                  . pressure of 12 - 14 psig, the annunciator would be in continuour, alarm during engine operation with a 12 psig setpoint.
                  . He purpose of the alarm is to notify Operations that system degradation is occurring          I during normal engine operation. He alarm does not provide a trip'nor is it designed to protect the engine from any type of failure. Jacket Water System pressure and engine differential temperatures are trended per the Diesel Generator System Technical Support Program to ensure system reliability. Operating the Diesel Generator engines with a low       '

Diesel Engine Cooling Water system pressure alarm setpoint of 10 psig does not affect or  ! alter engine reliability or affect or alter the ability to detect when system performance is l I , 2 e ]

F U.S. Nuclear Regulatory Comunission

   - April 1,1999 Page 246 of 247 degrading.

The revision to the UFSAR is only being made to reflect the actual Diesel Engine Cooling Water System low pressure alarm setpoint. The alarm setpoint is approved by the engine vendor Cooper Energy Services. The Diesel Generators will still be able to perform the required safety related function of providing emergency AC power for seven days in response to a Design Basis Event. No unreviewed safety questions are introduced by this change to Table 9-40. No Technical Specification changes are required. A change is required for UFSAR Table 9-40. 301 Type: USFAR Change Unit: 0

Title:

USFAR Change to Section 4.2.1.3.1. Table 4-1. Table 4-4, Table 4-17 and Figure 4-13

Description:

USFAR changes were made to Section 4.2.1.3.1, Table 4-1, Table 4-4 Table 4 17 and Figure 4-13 associated with a replacement hybrid B4C Rod Cluster Control Assembly (RCCA) design change. The functionality of the design was certified in design calculation DPC-1553.26-00-0196, FCF lon-Nitrided Hybrid B4C RCCA Functionality Evaluation. Evaluation: The calculation referenced above concluded that there was no unreviewed safety question. RCCAs are addressed in several accident scenarios. The only one that could potentially be affected by this change is the stuck RCCA accident. This new design will lessen the probability of the occurrence of that accident scenario because of the increased resistance to fretting wear and tip cracking. No Technical Specification changes are required. Changes are required to USFAR Section 4.2.1.3.1. Table 4 1, Table 4-4, Table 4-17 and Figure 4-13. l t

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i U.S. Nuclear R;,i" y Commission April 1,1999 Page 247 of 247 - 9 306 Type: USFARChange Unit: 0

Title:

USFAR Change to Section 9.2.5.3.2

Description:

USFAR Section 9.2.5.3.2 was chsnged to utilize a more conservative approach in determining the most severe historical meteorological conditions. His approach is then

                       - used in the thermal calculations for d, .termining the adequacy of the Standby Nuclear Service Water Pond.

Evaluation: The revision to UFSAR Section 9.2.5.3.2 chnges the description of calculations

                      - performed by Duke Power to confirm the ability of the Standby Nuclear Service Water Pond to provide cooling water to the plant for up to 30 days in the event of a simultane aus loss of Coolant Accident on one Unit, Loss of Offsite power on both Units, Loss of Lake Wylie, and shutdown of the non-LOCA Unit. These calculations have been refined to more accurately reflect the performance of the Nuclear Service Water System and the Standby Nuclear Service Water Pond. Specific changes to the calculations include a more detailed listing of the heat inputs to the pond, corrections to the analysis of the Nuclear Service Water System to more accurately reflect the flow capacity of the system, changes to the most severe meteorological conditions calculation, and inclusion of results from the Duke Power

• Gothic" computer program for cstculation of decay heat from the reactor core on the LOCA unit. Previously the decay heat input to the calculation was taken from curves supplied by Westinghouse. Duke now performs these calculations "in-house". In all cases the results of these calculations provided more conservative results.

The heat input to the Standby Nuclear Service Water Pond was increased as a result of these calculations and the cooling water flow rate was decreased. He results showed that the Standby Nuclear Service Water Pond is adequately sized to support the plant under the most severe conditions. This UFSAR change describes changes to calculations and does not affect the physical plant or the predicted performance of plant systems. There is no unreviewed safety question associated with this UFSAR change. No Technical Specification changes are required. A change is required for UFSAR Section 9.2.5.3.2. l l Y >}}