ML20196B385

Summary Rept of Facility Changes,Test & Experiments,Per 10CFR50.59 for Period 970901-990331.Summary of Commitment Changes Made During Same Time Period Also Encl

ML20196B385
Person / Time
Site:	Cooper
Issue date:	06/17/1999
From:	Swailes J NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NLS990061, NUDOCS 9906230185
Download: ML20196B385 (178)

Text

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Nebraska Public Power District -

NLS990061 Nebraskai Enerry Leader June 17,1999 U. S. Nuclear Regulatory Commission I i

Attention: Document Control Desk Washington, D.C. 20555-Ovol {

Gentlemen:

Subject:

10 CFR 50.59(b) Summary Report Cooper Nuclear Station NRC Docket No. 50-298, DPR-46 In accordance with the provisions of 10 CFR 50.59(b)(2), the Nebraska Public Power District submits a summary report of facility changes, tests, and experiments completed in accordance with the l

requirements of 10 CFR 50.59. This report covers the time period from September 1,1997, to l March 31,1999, and also includes those changes incorporated in the seventeenth revision of the Updated Safety Analysis Report. Also enclosed is a summary of commitment changes made during i the same time period in accordance with the Nuclear Energy Institute Guideline for Managing NRC Commitments, Revision 2, dated December 19,1995.

In accordance with 10 CFR 50.4, the original report is enclosed for your use, and copies are being transmitted to the NRC Regional Office and the NRC Resident Inspector for Cooper Nuclear Station.

Should you have any questions or comments regarding this report, please cot. tact me.

Sincerely,

. On'/ 1

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Vice esid tof Nuclear Energy

/lb Enclosure cc: Regional Administrator, w/ enclosure USNRC - Region IV h

Senior Resident Inspector, w/ enclosure USNRC 9906230185 990617 PDR ADOCK 05000298 R pon NPG Distribution, w/o enclosure Cooper Nudear Station P.o. Box 98 / Brownville. NE 683214098 Telephone: (402) 825-38n / Fax: (402) 825-52n http//www nppd com k .

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l ATTACHMENT 3 LIST OF MRC COMMYTMINTS l Correspondence No: NLS990061 l The following table identifies those actions committed to by the District in this document. Any other actions discussed in the submittal represent intended or planned actions by the District. They are described to the NRC for the NRC's f information and are not regulatory commitments. Please notify the NL&S Manager at Cooper Nuclear Station of any questions regarding this document or any asscciated ,

regulatory commitments. '

l COMMITTED DATE COMMITMENT OR OUTAGE None i

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1 l PROCEDURE NUMBER 0.42 l REVISION NUMBER 6 l PAGE 9 OF 13 l

l DESIGN CilANGES (DCs), SolTWARE DESIGN CllANGES (SDCs),

MINOR MODIFICATION PACKAGES (MMPs),

MODIFICATION PACKAGES (MPs), ENGINEERING EVALUATIONS (EEs),

CilANGE EVALUATION DOCUMENTS (CEDs)

DC 91-094. 91-094A. and 91-094A Amendment 1 TITLE: Video Capture System (VCS) and Panffiltfl.com (PTZ) Installation.

DESCRIPTION: This DC installed a Mark X VCS to enhance the security system at CNS. The VCS is a video monitoring ,

system which provides still video images of the alarmed zone prior to, during, r.nd aller an alann is l received. The DC was installed in two phases. In the first phase (DC 91-094) a serial connection was '

made to couple the VCS to the existing security computer to provide for system testing and personnel training. The first phase provided operation from the Central Alarm Station (CAS) only. In the second phase (DC 91-094 A) the serial connection was removed and additional system components were installed to enable independent VCS system operation. In phase two, duplicate VCS operational capabilities were also provided in the becondary Alarm Station. DC 91-094A Amendment I provided for the installation of new video amplifiers inside CAS and modification of the security multiplexer power supplies. This amendment also provided for a portion of a security tower to be encased with concrete to proside additional structural protection. In addition to the VCS work, DC 91-094A upgraded security camera 35 from a fixed camera to a PTZ unit to provide viewing of areas previously unable to be obsen ed from I the security stations. Compensatory security measures were required to be taken periodically during the installation period.

SAFTITY EVALUATION: This DC has no effect on any nuclear safety related systems. It does not affect any accidents preiiously analyzed in the USAR. Installation of the VCS is limited to interface with the existing security system j only. This DC does not involve any work associated with equipment which is safety related or supports safety related equipment. Failure of any equipment installed per this DC will affect the security system only; therefore, no new failure modes will be introduced to a safety related system. The security system I and electrical equipment involved in this DC are non-essential and will not cause an accident or malfunction of a different type than previously evaluated. No Technical Specification parameters are affected, thus there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

MP 92-023 TITLE: Drip Leg Drain Modification, Low Pressure Steam Supply to 1-A Reactor Feed Pump Turbine DESCRIPTION: This MP replaced a short, horizontal section of the 10" low pressure turbine piping to the 1-A Reactor FealPump Turbine (RFPT), and installed a drip leg drain and associated trap station. This modification was necessary to enhance moisture removal from this steam line and reduce the rate of crosion obsen ed in the 1-A RFPT.

SAFETY EVALUATION: This installation will improve the moisture removal from the low pressure turbine steam line and reduce further turbine casing erosion on the 1-A RFPT in the future, thus enhancing the long term reliability of the equipment. Minimizing moisture entering the RFPTs will lower the risk of any operational failure in this equipment. This activity will not result in any radiclogical effects. Decreasing the chances of a IUTT failure will decrease the probability that : y other equipment important to safety will be adversely atrected The RFPT steam supply does not direcdy or indirectly affect the Reactor Feed Pump pressure boundary. This MP will not change the design, function, operation, or reliability of equipment important to safety, nor will it induce ary equipment malfunctions or failures. It will not alter the accident mitigating capability of any plant equipment or systems important to safety and will not change their failure modes.

Installation of the drip leg does not create any accident scenarios different than currently analyzed in the SAR. This activity will not introduce any new failure modes in this equipment since it is not changing l

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the form, fit, function, or operating parameters of the low pressure steam system. It does not affect any assumptions, calculations, procedures, or design specifications used to establish the basis for defining the plant's margin of safety. The existing margins of safety as dermed in the basis for any Technical Specification are not reduced.

MP 93-155 TITLE: Optimum Water Chemistry, Phase I DESCRIPTION: This MP authorized the installation of taps (tie-ins) in the piping of various non-essential plant systems which will allow the future implementation of the Optimum Water Chemistry pmgram at CNS. The tie-ins occurred in the Class IVP piping of the Condensate System, Reactor Feedwater System, and l Instrument Air System, and in the Class IIIN piping of the Air Removal System.

SAFETY EVALUATION: This activity does not change the operating parameters of any plant equipment, structure, or appurtenances It does not alter the operation, maintenance, or response of any plant equipment, systems, or structures. Therefore, it cannot increase the probability of a plant event previously evaluated in the SAR. Neither the taps nor the systems in which the taps are installed are needed or credited fbr mitigating any plant events or malfunctions ofequipment important to safety. The new taps are passive components whose design, fabrication, installation, inspection, and testing meet or exceed those of the system they will become a part of. Therefore, this negates any credible increase in the probability of an equipment malfunction. The new taps are isolated and will not create any new interfaces with any existing plant equipment or systems, nor are they capable of creating any new and unanalyzed failure modes for any plant equipment, systems, or structures. This activity does not subject any equipment to higher stress l

levels, nor does it alter the operation or maintenance of any plant equipment. Therefore, no new or l difTerent types of equipment malfunction are introduced. The installation of the injection taps will not alter system performance, reliability, or operational margins; therefore, there is no reduction in the margin of safety.

MP 94-194 A TITLE: Sersice Water (SW) Pump Gland Water Flow Switch Upgrade J

DESCRIPTION: This mothfication replaced the SW Gland Water Flow Indicating Switches (SW-FIS-361 A, B, C, D) with a nonintrusive type Cow indicating system. The new system increases reliability of operation while -

reducing maintenance cost over the previous switches. l SAFETY EVALUATION: Neither the SW System nor the SW Ghmd Water System initiate any previously evaluated accidents or transients. The replacement of the gland water flow indicating switches will not change the function of the gland water system. The flow indicating switches provide kical indication and remote alarms, while the accident miti:;ation feature of the gland water system is not changed. The replacement flow element meets the seismic and pressure retaining requirements for the gland water system. The probability of a loss of gland water flow to a given SW pump is independent of the monitoring system. The gland water system piping has a design pressure which is less than the rated pressure of the flow elements. A failure of any given flow element would result in the loss of the associated SW pump. Additionally, if the failure

, was due to a gland water line rupture, the potential exists for the loss of the companion SW pump in the l given division. Ilowever, these postulated failures do not increase the consequences of a malfunction j because the redundant division is not affected and only one SW pump is required to meet post-Loss of j Coolant Accident heat load. The modification was installed within the gland water system while the l subject SW pump was out of service. A failure of the pressure retaining aspect of a flow element would be the same for either the previous or replacement units. The SW system is designed such that no single failure will prevent the system from performing its safety function. Any failure of the Ow element or indicating switch would be limited to the afTected division, and the other division wouts L: available to perform the intended safety function. Therefore, the margin of safety is not decreased.

I MMP 95-093 TITLE: Refuel Bridge Air Compressor Replacement I

DESCRIPTION: The purpose of this MMP was to replace the air compressor on the air supply unit for the refuel bridge. f The original compressor had cracked piston rings. It also removed the angle iron stub pieces and i associated grating from the washdown area on the 1001' level of the Reactor Building.

SAFETY EVALUATION: The air compressor and grating in the washdown area are tools and support equipment that are not considered in any accident scenarios. Although the refuel bridge grapple is operated by the air {

compressor, the air compressor was replaced with a comparable component. The angle irons and grating in the washdown area are stand alone pieces ofequipment that do not interact with any other equipment or atreet the function of any equipment important to safety. The air compressor and the washdown area grating are not defined in the basis for any Technical Specification; therefore, the margin of safety is not alTected.

MP 95-103 TITLE: Replacement of1IPCI-MOV-MO!4

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DESCRIPTION: IIPCI-MOV-MO!4, the Iligh Pirsstar Coolant Injection (IIPCI) Turbine Steam Supply Isolation Valve, had a history of steam leakage due to wear on the valve seat and disk. Therefore, this MP replaced IIPCI-MOV-M014 with an upgraded valve having tightcr clearances and hard-faced, wear resistant body / disk guides.

SAFETY EVALUATIOi?: The new valve will meet the same system design requirements and safety classifications as the existing valve. The operating conditions for the IIPCI system will not be affected. The modification will be implemented during cold shutdown and the valve stroke time will be tested to verify it remains within the insenice testing limit of 20 seconds. 'Ihe subject valve is not an accident initiator. The operating characteristics and safety functions of the !IPCI system will remain unchanged so calculated radiological (kases are unaffected. The replacement valve will perform the same design basis functions as the existing valve and meet or exceed material requirements. Therefore, the probability of equipment malfunction is not increased and no new types of plant events or equipment malfunctions are introduced. Post-modification testing will verify that the safety margin associated with IIPCI-MOV-MOl 4 and the iIPCI system remain within allowable limits. The modification will not reduce the margin of safety as defined in the basis for any Technical Specification.

MP 95-159 TITLE: Residual Heat Removal (RIIR) Ileat Exchanger "A" and "B" Senice Water Vent Line Upgrade DESCRIPTION: A Problem Identification Report determined that the 1" diameter vent lines kicated in the RIIR IIcat Exchanger Rooms 1-A and 1-B were not adequately supported. This MP provided a Code qualification including pipe stress analysis and new pipe support design and modification. A new support was installed on each of the vent lines.

SAFETY EVALUATION: Installation activities do not introduce any credible initiators of previously evaluated plant events. The modification improves system perfonnance by resulting in lower pipe stresses and support loading. This MP ensures that the modtfied supporting scheme will fulfill the USAR requirements for Class 1 S piping.

This modification, dunng and after implementation, will not impact any plant events previously evaluated in the SAR, and therefore, will not affect any associated radiological releases. This activity will only increase the load capacity of the supports and consequently will not affect the possible conscquences of a malfunction of equipment important to safety. No new failure modes are being introduced, therefbre, this activity will not introduce any other credible plant events or equipment malfunctions of a ditrerent type than previously evaluated in the SAR. This activity does not alTect any margin of safety specified I

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in the Ta:hnical Specifications because stress levels developed in the subject system will be lower than previous levels. Duringinstallation, the piping stresses remain at acceptable levels.

MP 96-Ol l TITLE: Main Steam Isolation Valves (MSIVs) Packing Replacement DESCRIPTION: This MP was generated to replace the packing in the MSIVs with ARGO packing and to install a new stem material which is more resistant to galling. MS-AOV-A086C was previously modified with the new stem material and packing configuration under Minor Modification Package 95-163 in Refueling Outage 16. The new packing configuration provides better sealing and helps mitigate the stem damage experienced with the old packing. MP 96-011 subsequently authorized the installation of the new packing / stem material in the remaining seven MSIVs. MS-AOV-A086A, B, and D were modified per MP 96-011 in Refueling Outage 17. Modifications to MS-AOV-A080A, B, C, and D were completed during Refueling Outage 18. (Note: the modifications completed prior to Refueling Outage 18 were previously reported).

SAFETY EVALUATION: This modification will not cause the MSIVs to close, will not cause a Main Steam line to break, or cause a Loss of Coolant Accident. It will not ham any efTect on any other analyzed accident. It simply improves the sealing of the MSIV's stem packing. A minor amount of metal is machined from the valve bonnet but has a negligible effect on the strength of the bonnet. Secondary Containment will be maintained when required during the installation. The stem friction will be reduced with the new seal design, so valve )

stroking will not be adversely afTected. This activity does not afTect the accident mitigating capabilities I of any system. The only equipment affected by this modification is the MSIVs. Improving the stem sealing and reducing the stem friction will not increase the probability of inadvertent closure of a MSIV nor willit increase the probability of the MSIVs not closing during a containment isolation. During the modification installation, the valve work will proceed until Secondary Containment is affected. When Secondary Containment is affected, fuel movement will be restricted until Secondary Containment is reestablished by installing a blank over the valve and leak testing it. This modification does not alter the basic design or function of the MSIVs, nor does it form any new connections or interfaces with any other plant systems or components. If the MSIVs failed, this change would not increase the consequences of that failure. The packing change has no impact on the seismic qualification of the valves. Proper operation of the valve will be verified by satisfactory completion of post-modification testing. No new l and unanalyzed failure modes are created. Changing the packing of the MSIVs does not have any effect l on the margin of safety defined in the Technical Specifications. Secondary Containment will be i maintained when fuel is being moved as required by Technical Specifications.

l MP 96-030 l

! TITLE: RF-AOV-FCVI 1 A/B (Reactor Feed Pump Minimum Flow) Replacement DESCRIPTION: The Reactor Feed Pump (RFP) minimum flow valves had a history ofleaking by. Therefore, this MP installed replacement air-operated valves for minimum flow valves RF-AOV-FCV11 A and RF-AOV-FCVI1B and also replaced a portion of the instmment air tubing from the existing air supply. In addition, differential pressure units (DPUs) associated with the instruments RF-DPIS-11 A and 11B were replaced with new DPUs that utilize a smaller measurement range to increase instrument accuracy and the reliability of actuation for the minimum flow valves.

SAFETY EVALUATION: The RFP control system is not impacted by this modification. Improper operation of the new RFP minimum flow valves or the modified DPUs will not introduce any new failure modes impacting operation of the RFPs. Therefore, design, installation, testing, and operation with the changes made by j this MP will not increase the probability of a plant event previously evaluated in the USAR. The work performed by this MP does not impact the ability of the Main Condenser to act as a " hold-up" tank for  ;

non-condensible gases. It does not impact any radiation monitoring equipment and the components I modified by this MP are not used in any way to minimize personnel or equipment exposure. Therefore, 4

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the changes made by this MP do not increase the consequences of a previously evaluated plant event or malfunction of equipment important to safety. The new smaller range DPUs will proside a more accurate method ofobtaining the differential pressure that indicates a low flow condition and will be designed / rated for the normal differential pressure that could exist across the flow element. Only the Reactor Feed system and the interface systems located in the Turbine Building basement are impacted by this mcxlification. The control circuit for the RFP minimum flow valves is not altered by this MP. The failure  !

modes for the new installation are sin ilar to the failure modes of the existing installation. No new components are being installed that did not previously exist. The operation of the RFP minimum flow control system is not addressed in the Technical Specifications or the basis for any Technical Specification. Therefore, this MP does not impact the margin of safety as defined in the basis of any Technical Specifications.

MP 96-075 TITLE: Stnictural Repair of Reactor Building Crane DESCRIPTION: A Deficiency Report identified the existence of a crack on the rail splice of the Reactor Building Crane where two sections of the 175# crane rail are welded together. In addition to the cracked splice weld, the j trolley rail had a broken clamp stud. This MP placed clamps on both sides of the crack and replaced the {

broken stud. 4 SAFETY EVALUATION: None of the accidents evaluated in the USAR are affected by this modification since the repair actisity will ensure that the crane functions properly and within the original design parameters. The probability i ofoccurrence or consequences of an accident or malfunction of equipment important to safety previously )

evah sted in the SAR is not increased and no new or different types of accidents or malfunctions are l createm No equipment important to safety is affected during or as a result of this repair work.

Implementation of the repair work will be performed in accordance with site approved procedures. No new feature or equipment is added which would require a new accident analysis. This is only a repair of the crane to ensure proper operability. This modification will not reduce the margin of safety as defined in the basis for any Technical Specification since there is no Technical Specification margin of safety involved.

MP 96-103 TITLE: Diesel Generator Exhaust Gas Bypass Modification DESCRIPTION: This modificatun replaced the Diesel Generator (DG) MufIler Bypass Valves (MBV) with a spool piece to allow DG exhaust to exit the exhaust gas bypass line during all modes of DG operation This MP changed the existing gas bypass system from an active to a passive design. This moditication also l performed several piping and support modifications to maintain the pipe / pipe support stresses within the l l Code allowables. I SAFETY EVALUATION: The materials and components installed by this MP will be " essential" and are rated for the temperature and pressure / design requirements of the DG exhaust system. The components of this modification are not associated with any of the accident / plant event initiators discussed in the USAR. The removal of components will have no adverse effect on the balance of the DG control circuitry. The changes to the DG cxhaust system piping and pipe supports will enhance the structural integrity of the system and ensure that the piping system meets Class IS seismic requirements. This MP will not result in any additional i Operator actions under a plant event accident scenario and will not change local DG operation. It will not adversely affect the ability of the DGs to perform their safety function. This MP will not impact the operation of the DG or any plant equipment that is required to operate in order to mitigate the consequences of an accident. The changes to the Exhaust Gas Bypass System will incorporate a passive design which will maintain the " essential" exhaust gas path to the atmosphere under all normal and accident conditions. The DG cxhaust system has no interaction with any other system. The installation of this MP will not impact the ability of the DGs to provide a singic failure proof source of on-site AC

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power adequate for maintaining the safe shutdown of the reactor following abnormal operational wi.as and postulated accidents. This MP removes an automatic function from the DG system, but does not add any new automatic function that could impact the operation of the DG system or any other system. None of the changes a0'ect the ability of the DG system to respond to an emergency start signal.

The exhaust flow path through the exhaust gas bypass line will be functional during all modes of operation and any accident scenario. The function / significance of the DG Exhaust Bypass System is not describert in the Technical Specifications or the bases of any Technical Specification. The requirements for maintenance of the DG system as defined in the Technical Specifications are not impacted by this modification.

MP 96-127 (Unreviewed Safety Question Evaluation (USQE) 1998-0004)

TITLE: Appendix "R" Emergency Battery Lights Upgrade (DG1, DG2, intake Structure, and Service Water Pump Room)

DESCRIPTION: The existing Exide Emergency Battery Light (EBL) units have a long history of failures and unsatisfactory support from the vendor. Therefore, this modification replaced a portion of the existing Exide Appendix R EBL units with the more reliable Big Beam Emergency System EBL units. EBLs were replaced in the following areas: Diesel Generators, Inteke Structure, and Service Water Pump Room.

SAFETY EVALUATION: This modification and the activities required to implement this modification are not accident initiators.

The only initiator which could be affected is the loss of ofTsite power and the minimal loading placed on the auxiliary electrical system has been shown to be acceptable. Work control procedures invoked ensure that the field work does not increase the probability of analyzed accidents or malfunctions of equipment important to safety. The accident for which the EBLs are required to function is an Appendix R fire that causes a loss of power to the normal plant lighting systems. This modification enhances the reliability of the Appendix R EBL units. This change meets the Appendix R requirement which requires the EBL urits to opante in order to perform the CNS safe shutdown task (the safe shutdown scenario is associated with the loss of offsite AC power). The hications of safe shutdown equipment and egress routes which require emergency lighting have not been changed. Therefbre, the availability of the new EBL units during Appendix R events is the same as the existing units. The new lighting units proside increased margin for battery life and, therefore, from a design margin standpoint result in a net decrease in the probability of a malfunction ofequipment important to safety. The EBL functionality and structural supports will be in accordance with Class I requiremenu in the vicinity of Class 1 equipment and Class Il requirements in the vicinity of Class II equipment. The electrical loads connected to the lighting panels have been evaluated for the additional 15W per EBL unit. The EBL units are not a source of a postulated plant event and the failure of a lighting unit will not result in a postulated plant event. This modification only impacts non-essential lighting and electrical sources. Design provisions for scismic considerations i and credit for existing electrical load shed ensures that no new accident types are created. The loss of non-essentiallighting has already been evaluated for CNS and, therefbre, this change does not create any new possibility of safety equipment malfunction. Controls implemented for the field work ensure that the plant equipment is operated within the bounds of failures as analyzed in the USAR and the possibility of a different type of malfunction is avoided. EBL units and normal plant lighting are not addressed in the Technical Specifications. The margin ofsafety remains unchanged by this modification or associated field work.

MP 96-150 (USQE 1998-0005)

TITLE: DGl(2)/EGl(2) Output Breaker Mode Selector Switch Modification DESCRIPTION: This modification replaced the existing Diesel Generator (DG) Output Breaker Mode Selector Switch SS/EGl(2) with a selector switch which provides an additional AUTO contact. This enables DG Breaker 6-A

EGl(2) to automatically close during DG monthly surveillance testing in response to a Loss of Coolant Accident / Loss of Offsite Power (LOCA/ LOOP) condition.

SAFETY EVALUATION: The DGs are required to suppon mitigation of plant events and do not affect the probability of event initiation, except that noted operator errors during the surveillance of the DGs could cause a loss of power event. The design and procedural controls intended to preclude a loss of auxiliary power during surveillance testing are unchanged by this modification. Therefore, this modification does not impact the probability of an accident or operational transient. The selector switch replacement and added wire jumpers enable the DG Output Breaker EGl(2) to automatically perform its intended function, at any transient point, during the DG monthly surveillance testing. The existing DG design is not adversely impacted by the nxxiification to the output breaker; therefore, no functional failures are introduced which would increase the consequences of any previously evaluated USAR accidents. No new failure mcxies are introduced that would impact the ability of the DG Output Breaker to automatically or manually respond to an emergency event (LOCA/ LOOP or LOOP alone). DG starting time to rated speed and voltage has not been altered. No undue equipment challenge has been created by this modification. All components and their nxxmting continue to meet Seismic Class I criteria. The implementation of this MP (kies not add any active components that could degrade the DG or breaker control reliability, it does nu alter or reduce protection to withstand the effects of natural phenomena, missiles, or any extemal events.

Therefore, there is no increase in the probability of a presiously evaluated malfunction of equipment important to safety. This modification does not change the consequences of a failed DG Output Breaker.

The two DGs are redundant; this modification does not crosa tic one DG to the other nor introduce any common cause failure potential. As with the previous design, the DG Output Breaker will support the DG as a sing!c failure proofsource of onsite power during DG monthly surveillance testing. The design criteria of MP 96-150 includes provisions to ensure that separation is maintained. No new types of accidents or equipment malfunctions are introduced. The basis for Technical Specification 3.8.1.3 includes a note intended to ensure that DG tests are conducted on only one diesel at a time to avoid common cause failures that might result from offsite circuit or grid perturbations. This requirement will still be followed during post-modification testing and during routine surveillance tests after completion of the modification. Other DG related Bases do not describe features that are affected by this modification. No adverse consequences result that could affect the ability of the DGs to support emergency response to a LOCA/ LOOP or a LOOP event alone. Ilowever, while the DG is paralleled to Critical Bus IF(IG), the response to a LOOP-only event is not as in Standby status. This is an existing condition that is not impacted by th;3 modification. During the actual modification implementation activities, the Technical Specification Limiting Conditions for Operation will be observed and any actions +

taken as appropriate. Therefore, it is judged that implementation of this MP does not reduce the margin ofsafety as defined in the basis for any Technical Specification in that DGl(2) retain the designed safe shutdown features and the emergency operation capability remains as specified in the USAR.

EE 97-015 TITLE: Gai-tronics Installation DESCRIPTION: A Problem Identification Report identified 117 Maintenance Work Requests (MWRs) that were written to perform work on the plant Gai-tronics and sound power systems. These requests included repair, replacement, removal, and installation of units, cables, and related equipment for the Gai-tronics and sound power systems. Many of these work requests represented changes or unauthorized modifications that were performed witixxit a supporting safety evaluation or design control documentation. Each of the MWRs was evaluated by EE 97-015. 'Ihc EE exammed c!cctrical implications such as fuse loading, part replacement and repairs, cabic and conduit routing, structural implications, and seismic interactions. The EE concluded that the modifications were acceptable and did not create any condition that would afTect the operability of the system. They do not represent any safety concern, have no impact on the safe shutdown of the plant, and in no way degrade the safety of CNS with respect to personnel, equipment, or nuclear safety.

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l SAFETY EVALUATION: Although credit is taken for the Gai-tronics system in the Fire Protection Safety Evaluation Report and l it is assumed available in Emergency Plan Implementing Procedures, the system is considered non-essential. The changes will not afTect operability of the Gai-tronics system and will not impact or afTect the operability of any system as previously discussed in any plant event evaluated in the SAR. The changes were evaluated for impact on any equipment important to safety as previously evaluated in the SAR. The evaluation concluded that the changes would not impact or interact with any equipment imponant to safety; therefore, the probability of neunence or consequences of a malfunction of equipment important safety are not increased. No r,ew types of plant events or equipment malfunctions are created. There is no margin of safety afTected by these changes.

EE 97-016 TITLE: Repair of Deficient Dead Weight Support on Control Rod Drive (CRD) Piping I

DESCRIPTION: Deficient anchor bolt installations were discovered on a dead weight suppon on a small bore CRD pipe.

The existing four bolt ceiling base plate of this support was replaced to correct a deficiency in the anchor bolt installation. The existing rod hanger was attached to the ceiling through the use of a standard, single anchor attachment point that utilized one of the existing non-deficient shell type anchors. The piping system is non-essential, scismic class IIS.

SAFETY EVALUATION: The repair of this suppon will not change the design, function, or reliability of the CRD system. The dead load support on this non-essential system cannot be an initiator of any accident evaluated in the SAR.

This activity does not afTect or interact with any systems or components required to mitigate the consequences of any plant event. The dead load capability of the support is maintained above acceptable loading criteria. Repair of this non-essential suppon does not involve any interaction with equipment impodant to safety or alter the accident mitigation capability of any system important to safety. The repair does not induce any new failure modes on any plant system or components. Therefore, no new types of accidents or equipment malfunctions are created. This activity does not afTect any assumptions, calculations, procedures, or design specifications used to establish the margin of safety as defined in the basis of the Technical Specifications.

l MP 97-018 TITLE: Digital Electro flydraulic (DEIf) Self-Calibrating Servo Cards, and Plant Management Information System (PMIS)/DEII Data Point and Annunciator DESCRIPTION: This modification provided an alarm indication of the DEH power supply status. A deficiency previously existed in that there was a lack of Control Room annunciation or Operator indication as to the iailure of the turbine DEH system. A DEII power supply failure can directly affect the control of reactor pressure, as the 7300 series analog cards supplied from this power source drive the turbine governor valves. Three new test cards were irstalled that connect to PMIS and will monitor critical DEH control functions. This i .MP also upgraded the turbine governor and bypans valve servo drive cards in order to improve turbine I

performance and reduce turbine maintenance. The new seno cards provide a self-calibrating function that will reduce outage maintenance time associated with preparing the turbine governor valves for operation following an outage.

SAFETY EVALUATION: The analyzed transients of interest are: (1) Dell pressme controller fail low, (2) turbine trip, and (3) generntor trip / load reject. Individual governor or bypass valve failure is not identified as an event initiator. Failure of the DEH control system could lead to a turbine trip. Ilowever, external signal isolation is provided with the replacement test cards. The transients analysis ofinterest were performed assuming normal bypass valve operation and without bypass valve operation. Given a turbine trip, the governor valves will close and remain closed until the trip is reset. This action is independent of servo control components, including the replacement servo cards. Following a turbine trip, the bypass valves will open (if >25% power) and remain open until power decreases below 25%, plus a 5 second time j l

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delay. After the time delay, the bypass valves revert to a pressure control mode via the servo valve and -

card. The failure of a single servo card would reduce the bypass valve capability. IIowever, this is bounded by the analyzed events, i.e., no bypass valve operation. If the DEII system should fail and prevent normal bypass valve operation / pressure control, the event would represent a turbine tr;p without bypass, which is an analyzed event. Therefore, the probability of occurrence or consequences of a presiously evaluated plant event are not increased. Failure or incorrect operation of an indisidual servo card is not expected to occur more frequently than the existing servo cards, and should be more reliable 1 as the result of the new programmable controls. The enhanced servo cards have performed satisfactorily l

at other nuclear facilities and reliable operation is expected at CNS. The test cards replace existing banana jack terminals. The new configuration will make data gathering easier (via computer terminal) and less susceptible to signal shorting via test leads. Therefore, the probability of a malfunction of equipment important to safety is not increased. Consequences of malfunctions of tim replacement equipment are the same as the existing equipment. Failures of the monitoring circuit will not adversely alTect Deli because isolation resistors are installed, and faults or failures of the DEH will not be transmitted to the monitoring system. The new equipment will perform the same function as the existing components and no new modes of operation are being added. Therefore, no new types of plant events or equipment malfunctions are created. This mothfication makes no changes to the valve operating curves for the turbine govemor or bypass valves. The new cards will be calibrated to the same input and output signals as the previous cards. Operation of the governor and bypass valves will remain within the limits I of the fuel reload .r.alysis Thus, safety margin remains the same.

FE 97-027 TITLE: Evaluation of Power Supply to the Safety System Status Panel (SSSP)

DESCRIPTION: his(kicument evaluated a potential unauthorized modification of the SSSP. The USAR stated that the SSSP was powered from the No Break Power Panel (NDPP), while it is actually powered from 24 VDC Panel DC-A. The SSSP consists of a 5x5 array ofindicating lights (25 windows). The USAR indicated that there are 16 systems displayed on the SSSP and that the panel had eleven spare indicating lights.

There are cunently no spare winck>ws on this panel. Nine windows that were " spare" are used to indicate the" test" status ofnine systems. This evaluation concluded that the instr.' led configuration is acceptable for "use-as-is," and the USAR was updated accordingly per USAR Change Request 97-170.

SAFETY EVALUATION: The existing infonnation in the USAR appears to be based on preliminary design information and it was not updated when the final design was completed. The SSSP is not required to be functional in order to shutdown the reactor or to maintain the reactor in a safe shutdown condition, and it is not associated with any of the accident initiators discussed in the SAR. Therefore, there is no increase in the probability of a pieviously evaluated plant event. The SSSP is not connected to any radiation monitoring equipment and it is not used to document any parameters associated with radiation monitoring. The SSSP is not twed in any manner to minimize radiation exposure of personnel or equipment. Therefore, there is no increase in the consequences of any plant event previously evaluated in the SAR. This EE does not require any physical changes to the plant and does not require any changes to plant drawings or procedures. This evaluation does not change the manner by which the SSSP is utilized or maintained.

The 24 VDC system is used to supply power to process radiation monitoring equipment as well as the SSSP. The installation of the SSSP as a 24 VDC load has not and does not impact the operation of the process radiation power supplies. Therefore, there is no increase in the probability of occurrence or consequences of a malfunction of equipment important to safety. The SSSP is a non-essential panel that is not required to perform a safety-related function. It represents an Operator aide that is used to monitor the status of various safety systems. Loss of the 24 VDC power distribution system is addressed by existing plant procedure: and it will not adversely impact the ability to shutdown the plant and maintain it in a safe shutdown condition. The SSSP is connected to "24 VDC System A" and does not interface with"24 VDC System B" Therefore, there are no equipment separation concerns. The panel does not have any automatic functions and is not connected to any safety-related system. Therefore, no new types of plant events or malfunctions of equipment important to safety are created. The 24 VDC system used to supply the SSSP is not a Class 1E power source and it is not discussed in the basis for any Technical 9

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i Specification. The evaluated changes will not impact the power supplies to any neutron monitoring equipment. Therefore, these changes will not impact the margin of safety as discussed in the basis for any Technical Specification.

MP 97-032 l

l TITLE: Extension of the Third Floor Conference Room in the Administration Building l DESCRIPTION: This MP facilitated expansion of the conference room on the third floor of the Administration Building.

l The work involved demolition and rebuilding of two non-bearing walls, addition / relocation of fire protection spnnkler heads, relocation'mahfication of an iIVAC air register, and relocation of an electrical circuit and sev:ral electrical outlets.

SAFETY i EVALUATION: Structural and ilVAC Changes: The scope of work is confined in a Seismic Class II building with no l safety related SSC being a!Tected. It does not afTect any SSC that is credited as being an event initiator I l or with mitigating the consequences of a plant event. The plant's ability to contain radioactive material either during normal operation or post event is not affected. The probability of occurrence or consequences of a malfunction of equipment important to safety is not increased because this activity is not related to any equipment important to safety. This activity does not afTect any parameter whose margin of safety is addressed in the basis for any Technical Specification.

Fire Protection: The probability of fire is not increased; no ignition sources are being added. The consequences of a fire are not increased because the hydraulic design and, therefore, suppression capabilities of the system, are maintained. The probability or consequences of equipment malfunction are not increased because the fire suppression efTects analysis has analyzed equipment failures as a result of fire suppression system actuation and the applicable codes and standards for system installation and alteration will be followed to preclude failure of the system itself. The possibility of a different type of malfunction or plant event than previously evaluated is not created because this is a modification to an existing system for which the failure modes and efTects have alreaay been evaluated. Technical Specification margins are unalTected. The aheration is so minor that it does not change the conclusions or assumptions of the system demand calculation.

Electrical: This activity is restricted to a Seismic Class 11 facility and does not interface with any SSE involmiin the initiation or mitigation of a plant event, or affect the plant's ability to contain a radioactive release. Relocation of receptacles and lighting fixtures will not decrease the reliability of any system l important to safety. The electrical loading on the lighting panel is not affected, as no loads are added or l remomi. This activity does not interface with any equipment important to safety. This portion of the MP alters an existing system; plant events and effects have been previously evaluated. No margins of safety are affected.

MP 97-035 l TITLE: REC-MOV-714MV Actuator Upgrade DESCRIPTION: This MP replaced the motor pinion and worm shaft gear set on REC-MOV-714MV to pre ,ide additional motor gearing capability and increase actuator torque output to ensure this motor operated valve will meet Generic Letter 89-10 closure requirements following issuance ofimpending Limitorque revised motor sizing criteria. It also resulted in a revision to the USAR to change the total opening time of REC-MOV-714MV and REC-MOV-711MV from 60 seconds to 120 seconds following a Loss of Coolant Accident and restoration of power to the emergency buses.

SAFETY EVALUATION: The replacement gear set will meet the same system design requirements and safety classifications as the previous Fear set. The operating conditions for the systems interfacing with REC-714MV will not be moaified and the control logic for REC-714MV will not be altered. The modification will not alter operating conditions or system parameters or introduce any new failure modes. No additional demands

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are placed on Operations personnel. The new operator configuration will not increase the probability of a previously evaluated accident. 'lhe valve's accident mitigating functions or ability to perfonn its design basis safety related functions are not reduced in any way. The modification does not alter the function or configuration of any components, equipment, structure, or systems required to mitigate postulated accidents or achieve safe shutdown. The modification increases torque and thrust limitation of the REC-714MV valve actuator. The increase in torque will ensure REC-MOV-714MV meets Generic Letter 89-10 closure requirements. Automatic and manual responses for REC-714MV remain unchanged. The resulting increase in stroke time does not increase the consequences of a plant event previously evaluated ,

in the SAR. Nuclear Engineering Department Calculation 92-093 has determined that the REC system is not required up to 30 minutes after an accident. The additional stroke time of approximately 20 seconds will not adversely impact the operation of Reactor Equipment Cooling (REC) or any other safety I system that relies on REC. The modification is to be performed with the plant at power and one train of I I

REC temporarily out of senice while the actuator is modified. Consequently, the plant is required to enter a Technical Specification Limiting Condition for Operation (LCO). The valve actuator handwheel will still be active so the valve could be manually stroked if needed. The valve components to be installed  !

will meet the same performance requirements, design standards, safety classifications, and material requirements as the existing components. This activity does not reduce the margin of safety as defined in the basis for any Technical Specification. It involves only a gear set replacement and the system i

parameters and protective component setpoints are not afTected. Changing the estimated opening time for REC-MOV-711MVn 14MV in the USAR from 60 to 120 seconds does not represent a reduction in the margin of safety. The 120 second time is justified based on not needing REC for five minutes after a LOCA and also based on a review of the diesel loading analysis. Although one redundant safety train ofREC supply will be temporarily removed from service, the margin of safety will not be reduced as the plant will enter a LCO as required by Technical Specifications in response to our licensing conunitment to meet single failure criteria. Senice Water (SW) backup is available through the REC /SW crosstie for REC cooling if failure of REC-MOV-7 I I MV would occur while REC-MOV-714MV is out of senice.

While this requires manual operator action frorn the control room, there is suflicient time available to take this action before REC cooling to the critical loops is required.

l MP 97-036 TITLE: MCC-DG1 Transformer Installation DESCRIPTION: The Plant AC Voltage Study indicated that Motor Control Centers (MCCs) LX and DG1 have the lowest margin between the minimum voltages required to operate safety-related components and the voltages that would be required during minimum ofTsite grid voltage conditions. The margin indicated extremely limited load growth potential not only for MCCs LX and dol, but throughout the entire Division I portion of the Electrical Distribution System. This MP changed the power supply to Motor Control Center (MCC) DG 1 from MCC-LX to a new 4160/480V stepdown transformer powered from the 4160V j bus between 4160V breakers EGI and FE. A new 250 KVA transformer and asscciated cables and conduit were installed in the DGl room to serve as the new power supply to MCC-DG 1. Implementation .

of this MP required DG1 to be temporarily removed from senice. The applicable Technical l Specification Limiting Condition for Operation was observed during that time period.

SAFETY EVALUATION: This activity does not involve an initiator of any plant event previously evaluated in the SAR, nor does j it reduce the ability of any system, structure, or component credited with mitigating the consequences of j a plant event or malfunction of equipment important to safety from performing that function. This MP enhances the reliability ofcomponents supporting event mitigating systems by ensuring adequate voltage is evailable to support operation ofsafety related electrical loads under degraded voltage conditions. This l a tivity does not interface with the reactor coolant, primary containment, or secondary containment 1 essure boundaries. Existing accident analysis and coping strategies are unalrected by this modification.

A loss ofoffsite power with a failed DG 1 will envelope the worst case consequences caused by a failed MCC-DGl transformer. The new configuration provides a more reliable power supply to the DGl auxiliary loads than the previous configuration. Although this MP adds more components to achieve this configuration, the DGl equipment layout and kication provide a more reliable system. Based on the

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design features and equipment qualifications, the probability of a malfunction of equipment important to l safety is not increased. This activity involves safety system auxiliary systems (DGl and Standby Gas I

Treatment) whose malfunction involves either a loss of AC power to safety related loads or fire.

Ilowever, a design in accordance with appropriate standards and the use of appropriately qualified materials minimizes the potential for any such malfunction. The design incorporated by this activity is based on a similar but separate installation in Division 11 which has demonstrated reliable operation for over ten yeris. In addition, with the exception of a ground relay which is physically isolated from the remainder of this installation, this design utilizes only passive electrical devices whose spontanecus failum is unlikely. A failure of the DGl 250 KVA transformer would cause DG1 to become inoperable; however, DG2 would be available. The design and installation of this MP meets single failure criteria.

His design and its potential failures are bounded by existing analysis and the possibility of a plant event of a different type is not created. The installation of the 250 KVA transformer meets IS seismic, electrical separation, and single failure requirements, and the ability to withstand natural phenomena.

Therefore, the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated is not created. This activity does not involve any parameter included in the basis for any Technical Specification. This modification is implemented to increase the margin of safety and assure the operation of safety related equipment at the Technical Specification second level undenoltage value.

MP 97-060 TITLE: Seismic Upgrades for EE-SWGR-41600 DESCRIPTION: His mcxlification addressed scismic issues relating to EE-SWGR-41600 that were identified during the l USI A-46 seismic walkdowns. EE-SWGR-4160G is one of two 4KV switchgear which controls I emergency power distributian and is required to maintain its function aller a seismic event. The south end of EE-SWGR-41600 is in direct with a concrete beam. During a seismic event, the switchgear could 3

pound against the concrete beam and cause relays in the switchgear to chatter. This MP installed a steel I brace connecting the switchgear to the concrete beam, thereby preventing any relative displacement and relay chatter. In addition, two fluorescent light fixtures hung on chains behind EE-SWGR-41600 were relocated to prevent them from swinging and striking the switchgear during a seismic event and causing relays in the switchgear to chatter.

SAFETY EVALUATION: Since this installation serves only to reduce the potential for seismic induced relay chatter, it will not increase the probability ofoccurrence or consequences of a previously evaluated plant event. Installation of the brace will be performed while the breaker 1 GS and emergency transformer on top are removed from service; therefore, the work will not a!Tect any equipment important to safety. Rekicating the light fixtures is a minor maintenance activity that can be performed while the switchgear is in senice without increasing the probability ofoccurrence of a plant event. The modification will increase the ability of the relays to perfonn their intended function and remain operational during a seismic event; therefore, there is no increase in the probability ofoccurrence or consequences of a malfunction of equipment important to safety. In addition, the work will not affect any other equipment important to safety previously evaluated in the SAR. The work will not change or alter the operational characteristics of the switchgear and relays. Therefore, no new or ditTerent types of accidents or equipment malfunctions are created. The safety function of the relays will only be enhanced and will not reduce the margin of safety.

l MP 97-065 TITLE: Pressure Taps for RIIR-MO39A and RIIR-920MV DESCRIPTION: Pressure taps containing a single isolation valve and threaded cap were installed upstream of RIIR-MO39A and downstream of R1IR-921MV. These taps are required to facilitate VOTES testing as required by Generic Letter 96-05 and associated Generic Letter 89-10.

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SAFETY EVALUATION: All work will be performed in accordance with approved plant procedures; therefore, the probability of an unwarranted SCRAM during installation under power has not been increased. Conservative design has been applied to the extension of an existing high energy line, ensuring that any increase in rupture probability is essentially nil. This activity does not alter the reliability, functionality, or design margins of either the Residal lleat Removal or Main Steam Systems, nor does it alter any existing plant procedures used to mitigate the consequences of an accident. Therefore, this activity is incapable of increasing the consequences of a previously evaluated plant event. It does not alter any operational or accident parameters, including environmental parameters, nor does it alter any plant operational or emergency procedures. Therefore, it is incapable of increasing the probability of a previously evaluated malfunction of equipment important to safety. Careful design considerations have been made to ensure that a malfunction of the equipment added by this modification will not result in conditions beyond those assumed in any accident analyses or evaluations. This modification does not create any new interfaces, nor does it add any new systems or functions to existing systems or structures. Therefore, it cannot create any new or different types ofplant events. No new failure modes are created. This activity does not affect any assumptions or margins used in the design of the plant or to evaluate the consequences of a postulated accident. Rather, by enabling more accurate testing of critical, remote operated valves, existing margins will be enhanced.

MP 97-070 TITLE: Safety Relief Valve (SRV) Stellite 21 Pilot Disc Replacement DESCRIPTION: This modification replaced the stellite / platinum pilot discs in all cis,ht Main Steam SRVs with stellite 21 pilot discs. Stellite 21 was used due to its resistance to corrosion in order to reduce or climinate the corrosion bonding of the pilot discs to the pilot seat and, therefore, climinate the setpoint drift that has  ;

l been experienced. I SAFETY l EVALUATION: This nxxhfication has no bearing on the probability of any accident or anticipated operational occurrence

! previously analyzed in the USAR. It does not adversely impact the design, material, or construction

standards applicable to the system or equipment being modified. The modification does not change the l SRV's accuracy or response time, and does not place the SRV outside its design envelope. The change in pilot disc material does not impact the consequences of previously evaluated accidents. The l nxxiification does not degrade a fission product barrier, does not change the response time of the valve, and does not increase any accident radiation dose calculation result or increase onsite or ofTsite r adiation dose, or increase personnel hazards. The purpose of the modification is to increase the probability of the proper functioning of the SRVs and will not cause the systems to be operated outside their design or  !

testing limits. Therefore, the use of a stellite 21 pilot disc will not increase the probability of a malfunction of a device important to safety. The modification does not change SRV automatic or manual operation, change seismic or environmental qualification limits, impact performance of a support system, change testing time intervals or requirements, or change equipment protection features. The only events l which are related to the SRV would be either a stuck open SRV or the inadvertent opening of an SRV.

In neither case would such an event be directly traceable to the material of the pilot disc. The modification will not introduce any new failure modes or create a new component or system interaction.

The material change is intended to nxiuce the possibility of malfunction and increase the potential for the SRV to actuate at the desired setpoint. The margin of safety related to SRV actuation contained in the Technical Specifications is based upon the analysis of the accidents and transients discussed in the USAR.

Based upon operational industry data and manufacturer information, it was determined that the likelih(xxl of the modification to contribute to increased corrosion bonding was low and hence will not reduce the margin of safety.

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1 MP 97-084 TITLE: Crossunder Piping Manways to Moisture Separators DESCRIPTION: This MP installed manways near the four mitered elbows at the inlets to all four moisture separators on the main steam crossunder lives between the high pressure turbine and moisture seperators. Installation {

of manways below the elbows was determined to be the most economical method of accessing the areas requiring Inside Diameter (l.D.) weld build-up repair. Failure to perform the manway installation and the required I.D. weld build-up would have resulted in an expected failure of the steam piping pressure boundary. q l

SAFETY EVALUATION: Installation of the manways will not interface with any other plant equipment and meets the original

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design requirements. The USAR evaluates the worst case double end rupture of main steam and '

feedwater piping which would bound any accident due to this installation. The reload analysis bounds any condition which would be seen by the loss of feedwater heating. The failure of the new manways will not increase any off-site dose and remains bounded by the existing analysis. The manways will have no interaction with any equipment required for the safe shutdown of the plant. The manways are used only for access to the interior of the piping. The additional metal of the manways will strengthen these sections of the pipe. He manway material is equivalent or better than the material to be removed. The manways  !

are independent of any safety related system and have no mechanical function. Failure of the manways will not create any new accident scenarios. The margin of safety will not change because the existing analysis bounds the installation of the manways.

MP 97-090 TITLE: Motor Operator Upgrades for CS-MO-MOSA and MO5B DESCRIPTION: In response to NRC concerns that the guidelines established by Limiterque Corporation for sizing AC-powered motor operators may be nonconservative in determining the torque output of the operator, Limitorque is in the process of developing new motor sizing criteria. The Core Spray (CS) system minimum flow valve operators, CS-MO-MOSA and MO58, were reevaluated based on the new sizing criteria and found to have negative torque margins under degraded voltage conditions. Therefore, this MP replaced the motors and gear sets for CS-MO-MOSA and CS-MO-MOSB to ensure the CS minimum flow valves are capable of performing their design basis functions during design basis events.

SAFETY i EVALUATION: This activity restores the design margin for the components and does not alter the method by which they l perfonn their design basis functions. SORC approved plant procedures will be followed to ensure that l systems required to mitigate a plant event are not impacted and to prevent any malfunction of equipment ]

important to safety. The new motors and gear sets will meet the same material, mechanical, and electrical requirements as the previously existing components. No new failure modes are introduced and redundancy / separation criteria are unaffected. Installation of the new motors will have a minor elTect on the 480 VAC electrical distribution system and the Motor Control Center (MCC) loads. MCC loads have  !

been evaluated and remain within the capacity of the feeder breaker, feeder cable, and the MCC bus. l Work will take place on one CS subsystem at a time with the plant at power. The safety system afTected by this activity will not be required to mitigate a plant event during performance of this activity. With one CS subsystem inoperable, the plant will be in a Limiting Condition for Operation per Technical Specifications. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced during the performance of this MP.

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j MP 97-100 (USQE 1998-0072)

TITLE: Z-Sump Power Supply Modifications l DESCRIPTION: This MP provided a redundant, essential power supply to the Z- sump system. This power supply will enable the Z-sump system to meets its safety performance requirements as a support system to the Standby Gas Treatment (SGT) system. It will enable SGT to function properly during a Loss of Coolant l Accident (LOCA)/High Energy Line Break (IIELB) concurrent with a Loss of OtTsite Power (LOOP),

l or Refueling Accident as required in the USAR. Applicable equipment was upgraded to Essential -

Class I Seismic, designed for the Operational Basis Earthquake (OBE) event.

EVALUATION: This activity does not involve any systems or components previously identified in SAR evaluations as event initiators, except for the AC power distribution system. This activity adds electrical lcads to the essential AC power distribution system and a fault could cause a loss of power, but the design prosisions preclude a common mode failure of both electrical divisions. This modification is intended to upgrade the ability of an implicitly credited safety system, the Z-sump, to perform its intended consequence mitigation function. It provides dedicated, essential power to the Z-sump pumps and provides fbr delayed access to essential power to the heat tracing function on the Z-sump discharge line. A new flow restricting feature on the OIT-Gas drain line to the Z sump will ensure that the Z-sump pumps are not challenged in their safety function to mitigste event conseguences. Seismic upgrade of the OIT-Gas dP l equalization line and Z-sump pressure instrument line outside of the Z-sump will ensure that these components do not create a breach of the Z-sump during seismic events. Therefore, overall this modification enhances the reliabihty ofcomponents supporting event mitigation systems by ensuring SGT is available during a LOCA or llELB concurrent with a LOOP or a Refueling Floor Accident. Existing accident analysis and coping strategies are not adversely afTected by this modification. All possible failures caused by installation activities at risk are covered by the analysis of a single equipment failure or single operator error identified in the USAR. The Off-Gas dP equalization line will be upgraded to seismic IS-OBE to ensure that the Z-sump boundary is maintained during a design basis seismic event.

A flow restriction modification in the OtT-Gas drain line will ensure that the capacity of the Z-sump pumps is not challenged. The component malfunction consequences of added components are addressed consistent with the single failure and isolation design provisions applicable to the system design basis.

Design provisions for separation and redundancy ensure that the MP is in accordance with the set of plant events reviewed in the SAR. The possibility of a different type of plant event or equipment malfunction is not created. This activity does not degrade the performance of SGT as defined in the basis for the Technical Specifications. It introduces additional design features to ensure that the flow of SGT is consistent with the assumed values in the accident analyses and in the Technical Specifications. The electricalloads being added to the Motor Control Centers are already considered in or evaluated against

the loading calculations for the diesel generators and were found to be within the design capability fbr i these components. As such, it is concluded that this modification assures the operation of safety related l equipment within the margin of safety as defined in the basis of the Technical Specifications.

EE 97-164 TITLE: Evaluation of Missing Bolts on Nozzle Shield Doors at Sacrificial Shield Wall l DESCRIPTION: Missing bolts were discovenxiin the naule shield (kiors in the upper sacrificial shield wall. This EE was generated to evaluate missing bolts in the reinforcing straps fbr the Reactor Feed Nozzle and Core Spray Nozzle shield doors. It included instructions fbr replacement of any missing bolts by allowing enlargement of the bolt holes to accommodate new bolts. The EE also determined the minimum number of bolts required to be installed in the shield ckx)r reinforcing straps to ensure the door straps will perform their intended function.

SAFETY EVALUATION: The upper sacrificial shield wall doors are non-essential structural components that provide radiation ,

shielding in the large a:ctangular openings for the Reactor Feed and Core Spray nozzles. The doors were E J

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analyzed for Seismic !!M concerns and evaluated to ensure they will resist the pressure fmm a postulated pipe break in the armulus and not become missiles. The evaluation concluded that the doors would  !

perform as intended with the minimum number of bolts as identified by this EE. The reinstallation of missing bolts will not affect any design margin of safety or impact any equipment important to safety. In addition, the doors are not credited as event initiators and do not affect any equipment or systems credited with terminating transients whose failure could result in a plant event. For these same reasons, the radiological consequences of a plant event as previously evaluated in the SAR will not be increased. The functional and design requirements of the doors have not been changed. Since the doors do not impact any equipment important to safety or any equipment or systems credited with mitigating the consequences of a malfunction of equipment important to safety, the probability of occurrence or consequences of a  ;

previously evaluated equipment malfunction are not increased. Since the doors will not create a Seismic '

IIS concern and will maintain their design function, no new types of plant events or equipment malfunctions are introduced. The margin of safety for the design of the shield doors is not described in the basis for any Technical Specification; therefore, the margin of safety as defined in the basis for any J l Technical Specification is not reduced.

EE 97-170 l

TITLE: Installation of Emergency Operating Procedure (EOP) Wrench for Control Rod Drive (CRD) Scram Discharge Volume (SDV) Drain Valve DESCRIPTION: A Problem Identification Report documented an unauthorized modification due to the installation of a wrench on the north wall of the Reactor Building for EOP purposes. The wrench is required per EOP 1 5.8.3 (Altemate Rod Insertion Methods) to manually open the CRD SDV drain valves when Instrument l Air is not available. The wrench has a lanyard and is secured by inserting its 6" long handle inside a 13"  ;

long vertically wall mounted strut. It was determined that the wrench in its current mounting l configuration will not impact any essential or non-essential equipment. Therefore, it was determined to '

be acceptable for "use-as-is."

SAFETY EVALUATION: The subject wrench is not required to be used to support normal plant activities or activities needed to j bring the reactor to a safe shutdown condition. Use of this wTench is controlled by specific steps in EOP 5.8.3 which is used when normal scram methods and the Altemate Rod Insertion system have failed to function. The wrench is secured such that it will not be dislodged by a Design Basis seismic event and, j therefore, will not impact its adjacent CRD equipment. Manual operation of the SDV drain valves is not required to mitigate the consequences of any plant event discussed in the SAR. This installation of this wrench does not impact the normal or safety-related functions of any safety-related equipment in the area.

The wrench remains in its storage location unless needed to support operator action in EOP 5.3 3.

Therefore, there is no increase in the probability of occurrence or consequences of a malfunction of equipment important to safety. The permanent storage of the wrench in the Reactor Building does not create the pmsibility of a plant event of a ddTerent type than previously evaluated in the SAR. It has been determined that the support anchoring method is adequate to meet Seismic 11/1 requirements. The presence of the nrench does not impact the operation of any other equipment in the area Therefore, no new types of equipment malfunctions are introduced. Manual operation of the SDV drain valves is not addressed in the Technical Specifications. Therefore, there is no reduction in the margin of safety as '

defined in the basis for any Technical Specification. I 1

l EE 97-171 l l

l TITLE: Core Spray Motor Ilas Angle Welded to Bottom Edge for Pipe Support

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l DESCRIPTION: A Problem Identification Report identified an unauthorized mochfication where the Core Spray Pump B Motor has angle iron welded to the bottom edge of the motor bent in an "1." shape to support the pump vent and drain pipes. A similar condition was found on Core Spray Pump A. This EE determined that the installation is acceptable for use-as-is.

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I SAFETY I j EVALUATION: The installed pipe suppons do not affect the known accident initiators. The pipe span, member stress on the installed pipe supports, and the additional weight of the pipes and pipe supports were determined to i

be acceptable. The pipe supports have been seismically qualified to function during a design basis l

seismic event. They will not fail during a seismic event and impact any other safety-related equipment in the area. 'the installed pipe supports will not afTect the function of the Core Spray Pump B or A Motor and Pump. They are not required to mitigate the consequences of any plant event discussed in the SAR.

The ofTsite dose (normal or accident) will not be afrected by the installed pipe supports. The supports do not create the possibility of a plant event or equipment malfunction of a ditTerent type than previously evaluated. While the Limiting Conditions for Operation and the Surveillance Requirements for the Core Spray System are discussed in the Technical Specifications, the subject pipe supports are not addressed and have no impact on the margin of safety as defined in the basis for any Technical Specification.

1 EE 97-175 j TITLE: RW-MOT-WIIUPC Removal / Spool Piece Replacement l

l DESCRIPTION: This EE documented the removal of the nonessential waste hold up pump cooler, RW-MOT-WIIUPC, from the radwaste system and the installation of a replacement section of pipe. This included removal of the coolcr's motor, motor support, and cooler's support. The cooler was removed because it routinely leaks and has not been required for use since initial installation. It has been used only as a fiow path and not for cooling purposes. It was originally installed to protect the demineralizer filter resins from degradation that would result if the water temperature exceeded 150*F, which does not occur based upon operating experience.

SAFETY EVALUATION: Removal of the cooler and cooler motor determination will not change the overall design, function, or reliability of the radwaste system or the station lighting system. The nonessential cooler and cooler motor l are not analyzed as part of any plat.t events; thus, their removal is not an accident initiator. The affected

! systems do not perform any accident mitigation functions; therefore, this activity will not result in any l increased radiological effects. This activity will not induce any equipment malfunctions or failures. It i

does not create any new failure modes since it does not change the operation of the radwaste system or station lighting system, nor does it change any existing interfaces between these systems and any other equipment and/or systems. The existing margins of safety as defined in the basis for any Technical Specification are unaffected by :emoval of the waste hold up pump cooler as this activity does not affect any assumptions, calculations, pnx:edures, or design specifications used to establish the basis for defining the plant's margin of safety.

EE 97-202 TITLE: Removal of Service Water (SW) Pumps' Suction Screens DESCRIPTION: A Problem Identification Repon documented discrepancies on Byron Jackson drawings which show a pump suction screen installed for each of the station SW pumps. Contrary to the drawings, these suction l screens have been removed. The removal of the screens was considered an unauthorized modification and this EE was generated to evaluate the acceptability of the modification. The suction screens were installed during original plant construction and were intended to be installed only during the pre-operational testing of the SW System. The EE concluded that the modification is acceptable from a design standpoint and that the SW pumps have retained their original design function.

l EVALUATION: This activity will not afTect the perfonnance or reliability of the SW System or Residual !Icat Removal (RIIR) SW Booster System, or afTect any system interface in a way that could lead to an accident.

Systems and equipment will not be degraded or operated outside their design or test limits. This activity does not affect or increase the radiation dose associated with the plant's response to any accident or l equipment malfunction. The SW and RIIRSW Booster Systems will retain their design functions.

I Filtering for the nyer water which is supplied to the SW pumps is provided by the trash racks, traveling J

screens, and the sparger/ screen wash systems. Filtering is also prmided at the discharge of the SW pumps by two essential strainers. Removal of the SW pumps' suction screens has not affected the ability of the SW and RHRSW Booster Systems to perform the essential auxiliary functions required for each of the operating modes and plant events they are required to function under. The function of equipment designed to control the release ofradiation is not affected by this activity. This actisity has no affect on safety related equipment and will not increase the probability of occurrence or consequences of a malfunction ofequipment important to safety. The suction screens served a non-essential function prior to plant stanup during preoperational testing of the SW System. The SW and RIIRSW Booster Systems will retain their safety objective and the SW pumps will function as designed during planned operations, transients, accidents, and special events. Removal of the screens does not affect any design requirements.

Therefore, no new types of accidents or equipment malfunctions are created. The removal of the SW pumps' suction screens does not affect the margin of safety established for the SW System and the REIRSW System. Removal of the screens does not affect the redundancy prosided by each system or the ability ofeach SW pump to perfonn as designed in an accident condition. Therefore, this actisity will not reduce the margin of safety as dermed in the basis for any Technical Specification.

EE 97-307 TITLE: B Sump Pump Instrument Air Isolation DESCRIPTION: This EE documented the pernument installation of a manual isolation valve for the B Sump Instrument Air (IA) Supply which was previously installed under Plant Temporary Modification 97-41. The valve was installed to suppait maintenance activities to the B Sump Pumps. It allows isolation ofIA during B Sump maintenance without valving out IA to Reactor Core Isolation Cooling (RCIC) and Senice Water (SW) air operated components.

SAFETY EVALUATION: The installed valve meets or exceeds the system design requirements of the IA system. Failure of the valve would have no different etTect on the IA or Radwaste systems than would failure of the tubing that the valve is replacing. This portion of the IA system is non-safety related and non-seismic. The Radwaste air operated valves that are supplied through this valve are not required to perfonn any safety function.

This activity will not result in any increased radiological effects. Probability of failure of the valve pressure boundary is no more likely than the failure of the tubing it replaces. Failure of this valve could potentially affect the operation of the Reactor Building and Torus drain lines into B sump; however, the operation of these drain valves does not perform a safety function. This activity does not change the design, function, operation, or reliability of any equipment important to safety, nor will it induce any equipment malfunctions or failures. The affected portion of the IA system does not perform any accident mitigation function; furthermore, this activity will not alter the accident mitigating capability of any equipment or systems. This activity does not create any new accident scenarios or introduce any new failure modes as it is not changing the form, fit, function, or operating parameters of the IA or Radwaste systems. The existing margins of safety as defined in the basis for any Technical Specification are unafTected as this modification does affect any assumptions, calculations, procedures, or design specifications used to establish the basis for defining the plant's margin of safety.

EE 97-315 TITLE: Repair of Pipe Support FDR-S483 DESCRIPTION: A nonconformity was found on pipe support FDR-S483. This suppon restrains 3" Radioactive Fhor Drain piping which is routed in the area of the torus. The support is a seismic restraint and is intended to restrain the pipe in the lateral direction. It employs a strut and clamp design; however, there was no load pin to connect the strut and the clamp, thus rendering the support ineffective. Interference with an adjacent large bore Senice Water pipe caused a misalignment between the clamp and the strut which made simply replacing the pin impossible. Additionally, a portion of the strut body was missing. The existing clamp was rotated on the pipe and a lug welded to the side of the clamp to create an attachment for the strut paddle. A strut paddle and pipe were fabricated to replace the missing part.

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l SAFETY EVALUATION: Repair of the pipe support will not change the design, function, or reliability of the Radioactive Floor Drain system. The seismic loading on this non-essential system cannot be the initiator of any accident evaluated in the SAR. This activity does not afTect or interact with any system or components required to mitigate the consequences of any plant event. The dynamic load capacity of the suppon is maintained above acceptable loading criteria and the support meets seismic II over I requirements. Repair of the non-essential support does not involve any interaction with equipment important to safety or alter the accident mitigation capability of any system important to safety. This repair does not induce any new failure modes on any plant system or components. Therefore, no new types of accidents or equipment malfunctions are created. This activity does not alTect any assumptions, calculations, procedures, or design specifications used to establish the margin of safety as dermed in the basis of the Technical Specifications.

EE 97-351 TITLE: RWCU Pumps A and B Oiler Piping Support DESCRIPTION: A Problem Identification Report identified the need to support the Reactor Water Cleanup (RWCU)

Punip A and B Oiler piping to maintain the proper orientation. A problem was created when the fill bottle became misaligned through various maintenance activities or pump vibration. The alignment of the oiler is critical since a misalignment can lead to an inadvertent draining of the oil. This EE installed a suppon on the oiler bottle piping to maintain the oiler in a horizontal orientation.

SAFETY EVALUATION: The addition of the support to the oiler piping will result in upgrading the existing configuration and the function and reliability of the RWCU Pump Oiler and its piping system are positively affected. This activity (kies not airect or interact with any system or components required to mitigate the consequences l

of a plant event. The affected system is nonessential and the dynamic load capacity of the support is j maintained above acceptable loading criteria. The added support does not involve any interaction with l equipment important to safety and does not change the design, function, operation, or reliability of any j equipment important to safety. The perfonnance of this activity will be procedurally controlled to ensure that systems required to mitigate a plant event are not impacted and that malfunctions of equipment important to safety are prevented. No new failure modes are introduced. Performance of this EE does not alTect any equipment needed for an accident or transient. This activity does not affect any assumptions, calculations, procedures, or design specifications used to establish the margin of safety as dermed in the basis of the Technical Specifications.

EE 98-001 TITLE: Installation of Larger Filtering Capacitor into Multiplexer (MUX) Cabinet L DESCRIPTION: This EE changed the fihering capacitor in MUX-14 from a 3.0 microfarad capacitor to a 1300 microfarad capacitor. Noise on the input leads coming from sump 1U to MUX-14 was causing spurious alarms in the control room. The larger capacitor climinated t' e noise problem.

SAFETY EVALUATION: The MUX system is not an event initiator. Installation of a larger filter capacitor does not cause an increase in the consequences of a plant event since it has no efTect or impact on the boundaries and barriers to fission product release. This change restores the original intended design function of MUX-

14. Replacing the filter capacitor has no impact on the reliability of equipment important to safety.

MUX 14 is electrically isolated from all safety related equipment. The fiher capacitor does not afTect l automatic initiation of safety systems. Replacement of the capacitor not does not introduce any new l

accident initiators or precursors or create any new failure modes for equipment important to safety. It does not impact the indication or sensing of any parameters that are defined by Technical Specifications or administrative limits. Therefore, there is no reduction in the margin of safety.

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f EE 98-010 TITLE: Qualify Existing 1/4" Copper Tubing Between DW-V 147 and HPCI-V-284 1 DESCRIPTION: A Problem Identification Report identified the existence of I/4" copper tubing installed between valves DW-V 147 and IIPCI-V-284. The tubing was not shown on applicable station drawings. It was determined to serve no function other than a means of getting demineralized water from one place to the l other. The tubing has been evaluated for seismic concerns by Engineering Judgement 94-074 which i

' detemuned that the tubing is acceptable for seismic class IIS/IS ser ice. The copper tubing is only used during surveillance testing and has no efTect on the function of the High Pressure Coolant Injection (IIPCI) Gland Seal Condenser. This Elbletermmed that the installed copper tubing is acceptable for use-as-is and it was added to appropriate drawings.

SAFETY EVALUATION: The purpose and function of the associated surveillance tests are not changed and the procedures are not impacted. Adding the tubing to applicable drawings enhances the drawings and will not afTect the design or function of the HPCI Gland Seal Condenser. There will be no physical or operational changes in the plant. There is no increase in the probability of occurrence or consequences of a plant event or malfunction ofequipment important to safety and no new types of plant events or equipment malfunctions are created. Technical Specifications contain the Limiting Conditions for Operation for instruments that initiate or control the core contamment cooling systems. The !IPCI Gland Seal Condenser ilotwell Level Switches are inchided, however, the setting limit for these instruments is not affected by the addition of the copper tubing. There is no reduction in the margin of safety as defined in the basis for any Technical Specification.

EE 98-015 TITLE: Evaluation of Existing Manual Screen Wash Strainer A & B Drain Valve DESCRIPTION: This document evaluated an unauthorized modification in the Screen Wash System. A drain valve was identified on the manual strainer for Screen Wash Pumps A and B that was not in the valve lineup, did not have a Component Identification Code or identification tag, and was not on any Control Room drawing. The EE determined that the design characteristics of the valve and its associated piping meet or exceed the design requirements of the screen wash strainer / system. The installation of the drain valve (which was subsequently labeled CW-V-318) was found to be acceptable for "use-as-is." The valve was added to USAR Figure XI-6-1, Sheet 1, per USAR Change Request 98-043. It was also added to the valve lineup in Procedure 2.2.3A, Circulating Water System Component Checklist. ,

SAFETY EVALUATION: Documentation of a drain valve installed on the manual strainer for Screen Wash Pumps A and B will not impact the operation or control of the strainer or the Screen Wash system. This strainer is not associated l

with any of the initiators of plant events discussed in the USAR. The Screen Wash system is not required to support any activities required to shutdown or maintain the reactor in a safe shutdown condition.

Therefore, there is no increase in the probability of a previously evaluated plant event. The changes do

, not impact any procedures or equipment used to monitor, control, or minimize radiation / contamination l exposure. This valve is not located near any radiation monitoring equipment. Therefore, there is no increase in the consequences of a previously evaluated plant event or malfunction of equipment important to safety. The existing valve on the manual screen wash strainer is not required for the operation or

, testing of any safety-related equipment. This strainer is not required to support any normal or emergency l reactor shutdowns and is not needed to support any activities required to ensure personnel safety under l normal or emergency conditions. Thus, the addition of this valve does not increase the probability of

' l occurrence of previously evaluated rnalfunctions of equipment important to safety. The manual Screen  !

Wash strainers are operated as passive componerts. They are not required to actively operate for any  !

steady-state / emergency design functions. The subject valve is maintained closed and does not impact the design function or operation of the strainer or the Screen Wash system. The manner in which the Screen Wash system is operated or controlled is not altered by this change. Therefore, no new types of plant events or malf.metions of equipment important to safety are created. Since the Screen Wash system is 20-

not discussed in the Technical Specifications or the basis for any Technical Specification, there is no reduction in the margin of safety.

EE 98-017 TITLE: Weld Repair of Moisture Separator "D" Inlet Steam Pipe DESCRIPTION: During the RFOl7 crosion/ corrosion inspection of Main Steam Pipe Component MS-E-9-724J489,it was discovered that wall thinning had occurred in a localized area which reduced pipe wall thickness below the acceptable minimum. This component is a mitered elbow, which is part of the 36" diameter Main Steam inlet piping for Moisture Separator "D" Based on the minimum measured wall thickness and the calculated wear rate for the subject pipe component, calculations showed that the pipe wall thickness could be reduced below the required thickness prior to RFOl8. Therefore, this EE implemented repairs to ensure the pipe wall meets the crosion/ corrosion minimum thickness and code required thickness to remain operable through RFOl8. The repair was made by adding weld build-up on the outside of the pipe over the thinning area during a plant shutdown.

SAFETY EVALUATION: This repair will have no effect on system performance and will not increase the probability of a previously evaluated plant event. It will have no effect on the function of the piping and the ability to mitigate the consequences of an accident. This activity does not involve equipment important to safety. The pipe component is non-essential and the configuration of the piping will remain unchanged. This repair will restore the pipe to the required thickness. It will not increase the radiological consequences of an equipment malfunction. The repair will not introduce any new accident initiators or failure modes; therefore, no new types of accidents or equipment malfunctions are introduced. This activity will not affect any plant parameters or the level of redundancy of any plant system. No safety margins will be adversely afrected.

EE 98-098 (USQE 1998-0029)

TITLE: Evaluation of Packing Configuration Changes DESCRIPTION: This EE was initiated to evaluate changes made in pump and valve packing materials over the years without formal pre-installation evaluation by Engineering or follow-up revisions to controlled documents.

This EE was generated as part of the Unauthorized Modifications Project. Based on a resiew of pump and valve packing changes performed under Maintenance Work Requests (MWRs),1036 MWRs with madequate supporting documentation were identified as affecting the current packing configurations for

%2 components. This EE determmed that existing packing configurations were acceptable for use-as-is.

Maintenance Procedure 7.2.70, Valve Packing Maintenance, was revised to provide for future packing configuration control.

SAFETY

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EVALUATION: Unauthorized changes to add or replace packing on both active and passive pressure boundary components were evaluated. Evaluation of these unanalyzed conf:gurations to ensure the changes are equivalent to originally installed configurations, or were made in accordance with accepted industry practices, will reduce the probability of unanticipated equipment malfunctions. Following valve packing replacement, post maintenance testing and surveillance testing verify that the component will function as required. Therefore, the probability of a previously evaluated accident is not increased. The graphite

, based material, typically used as a replacement for asbestos, is chemically inert and has superior scaling capabilities. Use of these improved materials at passive pressure boundary interfaces will serve to l pn: vent degradation and leakage that could limit access to vital areas and impede actions to mitigate the l

consequences of accidents. Evaluation of changes for active pressure boundaries provides additional assurance that the performance ofcomponents relied upon to mitigate the consequences of accidents will not be degraded. Two additional failure mechanisms have been noted after asbestos pecking replacements with graphite ribbon materials; packing extrusion into the area between the valve body and i stem hindering valve motion, and adhesion of graphite particles on valve stems resulting in excessive l 1

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packing frict:on llowever, these failure mechanisms were specifically evaluated and it was concluded that packing additions have not been excessive due to packing extrusion and these are no increased risks j ofmalfunction due to graphite particle buildap. Surveillance testing is used to verify that essential and important to safety components are capable of perfomdng their design functions following packing replacement. Therefore, there is no increase in the probability of occurrence or consequences of a malfunction of equipment important to safety. This activity will not increase the radiological consequences of a malfunction of equipment important to safety because it will not result in any changes in modes of operation for any of the affected components. Packing failures could result in packing leakage or binding within active components that could cause them to fail to operate. These types of failures were previously analyzed and evaluated in the USAR. This activity will not create the possibility of a plant event of a difTerent type as the manner ofinteraction with affected components will remain unchanged. No new system components are being added as a result of the packing configuration changes.

All changes are intemal .o the valves. These configuration changes will not introduce any rnw types of malfunctions beyond those previously anticipated and evaluated in the SAR. Evaluation of current packing configurations and improved control of future configurations will provide additional assurance that performance for components relied upon to mitigate the consequences of accidents will not be degraded. Valve stroke time testing is used to ensure that a valve is capable of opening or closing within its design requirements. Therefore, margins of safety associated with operation of these components as used to define the basis for any Technical Specifications will not be reduced.

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EE 98-109 l TITLE: Use of Spiral Wound Gasket Material in the Senice Water (SW) System ,

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DESCRIPTION: This EE was performed to evaluate the use of Flexitallic style spiral wound stainless steel gaskets in the l SW system in place of cloth inserted rubber gaskets as specified in Contract E-69-4. In particular, the l

Residual Ileat Removal Heat Exchanger "A" SW-2 drain line flange has a Flexitallic gasket installed. J Therefore, this EE was initiated tojustify the installation or authorize its removal. The EE concluded that it is satisfactory to use Flexitallic s*yle spiral wound metallic gaskets in all applications of the SW system for all piping sizes. Maintenance Procedure 7.2.71, Bolting and Torque Program, was subsequently revised to reflect approval of the spiral wound metallic gasket for use in all SW piping applications.

SAFETY EVALUATION: This activity will not change the overall design, function, or reliability of the SW system. The function of the stronger gaskets will be to reduce gasket failures and improve the maintenance of flanged fittings; therefore, the gaskets will not be an accident initiator. The SW system will still perform its intended function. The use of the metallic gaskets will not afTect any accident mitigation function nor will it result in any increased radiological effects. This is a maintenance impros ement activity and will not increase the consequences of an accident. It will not change the design, function, or reliability of any equipment important to safety, nor will it induce any equipment malfunctions or failures. The installation of the gaskets rQes on the skill of the craft following established procedures. The metallic gaskets are compatible with the SW piping system and will not adversely impact any equipment important to safety.

This activity is not capable ofincreasing the radiological effects of any equipment failures. No new failure modes are created since this activity is not changing the form, function, or operation of the SW system. Therefore, no new types of accidents or equipment malfunctions are introduced. The use of metallic gaskets (kies not affect any assumptions, calculations, procedures, or design specifications used l to establish the basis for defining the plant's margin of safety.

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[' EH 98-110 l

TITLE: Repair of Reactor Building Railroad Airkick Door l

DESCRIPTION: A Problem Identification Report identified broken and missi ig bolts in the threshold plate of the Reactor Building railroad airlock door, BLDG-DOOR-X100. The bolts required replacement and/or repair in order to maintain secondary containment integrity. This EE was prepared to authorize repair of this condition. The replacement bolts are an upgrade to the original design.

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l SAFETY j EVALUATION: Although the railroad airlock door is an essential door required as part of secondary containment and also acts as a fire door, the replacement of missing and broken bolts with the proposed bolt material and bolt

! size per EE 98-110 will not affect the ability of the door to continue to perform its intended function or

! impact any other equipment important to safety. W functional and design requirements of the door have I not changed and the ability of the door to maintain secondary containment is not afTected. There is no increase in the probability ofoccurrence or consequences of a plant event or malfunction of equipraent 1 l important to safety. The margin of safety is not reduced as the door will continue to maintain the I

requirements for secondary containment. '

EE 98-12Q TITLE: Seal Repair of Reactor Building Doors R101 and R102 DESCRIPTION: Doors R101 and R102 are the primary personnel entrance / egress doors to the Reactor Building and are therefore subjected to heavy usage. They have consistently failed to meet air in-leakage requi:ements of Gurveillance Procedure 6.SC.502. In order to minimize the amount of air in-leakage into the Reactor Building, P-strips were installed on the bottom of both sets of airlock doors. This EE also evaluated the previously unauthorized installation of existing P-strips. The doors were satisfactorily tested aller completion of the seal repair.

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EVALUATION: The installation of the P-strips results in an enhanced method of reducing air in-leakage into the Reactor l

' Building and does not affect the ability of the doors to perform their intended functions or impact any i other equipment important to safety. The doors will continue to function and maintain secondary l containment, act as a barrier when primary containment is opened, act as a fire barrier, and proside a j security barrier to a vital area. The ability of the Standby Gas Treatment system to maintain a pressure of 0.25* of water vacuum in the Reactor Building will be enhanced. The installation enhances the ability ofplant equipment to mitigate the consequences of a malfunction of equipment important to safety. The design function of the doors will only be enhanced. No different types of plant events or equipment malfunctions are introduced. The margin of safety is not atTected as the doors will continue to maintain the requirements for secondary containment and continue to function as a fire and security barrier.

1 EE 98-137 l (USQE 1998-0008)

TITLE: Evaluation of Reactor Protection System (RPS) Pilot Solenoid Valve Relay Replacement per Replacement Component Evaluation (RCE)94-029 DESCRIPTION: RCE 94-029 previously evaluated end justified the replacement of RPS scram solenoid contactors fr om a General Electric (GE) Part # CR105 to a GE Part # CR305. GE recommended replacement of contactors that had been in service longer than 18 years. CR105 contactors were no longer available and the replacement component from the Original Equipment Manufacturer was the CR305 contactor. The RCE incorrectly stated that no changes to the USAR were required. CR105 contactors are specifically l mentioned in USAR Section VII. The RPS contactors were changed from GE Mouel CR105 to GE Model CR305 in April 1993. RCE 94-029 was completed in January 1995 to provide the engineering i basis for future replacements. Since the modification constituted a change to the facility as described in the SAR, it was completed without appropriate documentation. The work performed in 1993 is considered an unauthorized modification. This Engineering Evaluation was written to evaluate the acceptability of RCE 94-029 and the required USAR revision. The RCE documented that the two contactors are equivalent and this EE concluded that the present installation is acceptable for "use-as-is."

, The USAR was revised per USAR Change Request 98-064.

l SAFETY l EVALUATION: The response time of the RPS is not impacted by this change. The replacement contactors meet the design requirements for this installation. The design of the replacement contactors is equivalent to the original contactors and no new failure modes are introduced as a result of this change. This actisity does I .

not change any instrument accuracy or response characteristic in a manner that could cause an accident to more likely occur. In addition, it does not increase the possibility of operator error or add to human factor conditions such that the probability of an accident to occur is increased. No new components are added to any RPS circuits. The ability of the RPS to perform its design function is not impacted. This change does not impact any procedures or components used to control or monitor contamination or radiation. Therefore, there is no increase in the consequences of any previously evaluated accidents.

RCE 94-029 verified that the design characteristics of the contactors are equivalent. The manner by which the scram contactors are controlled and maintained is not impacted. The quality classification of the scram contactors is not affected. The method ofinstalling the CR305 contactors is the same as for the CR105 contactors. Therefore, this activity does not increase the probability of a malfunction of equipment important to safety previously evaluated in the SAR. Since this change does not impact any equipment used to control, monitor, or nunmuze the release or spread of contamination or radiation, there is no increase in the consequences of equipment malfunction. The RPS scram contactors are not being relocated and there are no changes in the scismic qualification requirements for the scram contactors. The design life of these contactors is 20 years and preventive maintenance to replace :these contactors on a routine basis will ensure that design life is not exceeded. The control circuit for these contactors is not being altered and the maximum response time is not impacted. Therefore, no new types of plant events or equipment malfunctions are created. The calibration requirements for the RPS components are discussed in the Technical Specifications. Neither the manner in which the response time is measured or monitored, nor the frequency for response time tests, is being altered by this change. The basis for the applicable Technical Specification section is not impacted and there is no reduction in the margin of safety.

CED 1998-0013 (USQE 1998-0025)

TITLE: CS-MOV-M012A/M0128 Motor Replacement DESCRIPTION: This CED provided for the installation of replacement motors for CS-MOV-M012A/B in order to provide positive torque margin in the closing direction. This replacement is in response to a Limitorque Technical Update which reduced the projected torque output of Limitorque actuators. For motor  ;

operated valves that have low torque margins, the revised sizing criteria could result in reduced margin under design basis, degraded voltage conditions and produce negative torque margins in the closing direction. The replacement motors provide suflicient torque to eliminate these concerns for CS-M012A/B. Only the motor replacement for CS-MOV-M012B was completed during the 1998 refueling

. outage. The work on CS-MOV-M012A remains to be completed in a future outage.

SAFETY EVALUATION: This modification does not change the system in a way that could initiate any of the events or accidents evaluated in the SAR. The modification will not change the control functions of the Core Spray (CS) I system. Work will be controlled and performed in accordance with the applicable Technical Specification Limiting Condition for Operation (LCO) requirements for the plant conditions. This modification will increase system performance and reliability. Installation of the replacement motors is to ensure compliance with the design basis of the Core Standby Cooling System. Replacement of the subject motors will not alter the ability of the valves to provide Primary Containment isolation, nor will overthrust protection be reduced. Therefore, the consequences of an accident previously evaluated in the SAR are not increased. This modification is bounded by previously evaluated malfunctions. It doce not involve the alteration of safety system setpoints nor does it introduce any new failure modes. Operability of safety equipment as stated in the SAR will not be hindered by the failure of either CS-MOl2Allt Each loop of the CS system is physically and electrically separated so that no design basis event can render both trains inoperable. Each train is independent of each other and acts in conjunction with Low Pressure Coolant Injection (LPCI) to provide emergency core cooling. This modification will not alter the perfonnance characteristics of the CS or LPCI safety subsystems. No existing accidents are broadened in their scope and no new types of accidents or equipment malfunctions are created. The replacement motors provide suflicient increases in torque to maintain positive margins under degraded voltage conditions. This ensures operability in the environments during and following a design basis event. The 1

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work will be performed in accordance with the applicable Technical Specification LCO requirements for Improved Technical Specification 3.5.1 or 3.5.2, and 3.6.1.3 for the plant conditions. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

CED 1998-0014 (USQE 1998-0026)

TITLE: RIIR-MOl8 Motor Replacement DESCRIPTION: This CED installed a replacement motor to provide positive torque margin for RHR-MOT-MOl 8. The .

change was implemented in response to a Limitorque Technical Update which reduced the projected l torque output of Limitorque actuators. For motor operated valves that have low torque margins, the revised sizing criteria can produce torque decreases under design basis, degraded voltage conditions that result in negative torque margins. The replacement motor provides suflicient torque to eliminate these concerns. A revision to the CED was issued to allow motor replacement with the valve in any i

configuration (open, closed, or disassembled), provided the Shutdown Cooling (SDC) System can be

! isolated, as required for plant conditions and valve configuration.

l SAFETY l EVALUATION: 'Ihis modification does not impact any system in a way that could initiate any of the events or accidents l cvaluated in the SAR. The valve is a like replaecment, but with a higher capacity. The safety margin will be increased, therefore, the function in the system is the same and previous accident analyses are applicable. The replacement motor for RHR-M018 will improve its ability to provide Primary Containment isolation. During this installation, the integrity of the reactor coolant pressure boundary will not be violated. Compliance with Technical Spechications associated with the Residual Heat Removal (RHR) system will ensure operability and Limiting Condition for Operation requirements are met.

Installation of the replacement motors is to ensure compliance with the design basis of the RHR system.

This compliance ensures that the consequences of previously evaluated accidents are maintained consistent with the design basis. Restrictions are in place to prevent inadvertent draindown or a SDC

, isolation event. Tag-out procedures will ensure that safety systems re aain operable and that the l consequences of malfunctions of equipment important to safety are preserved during the installation l l

activities. Changes will not be made to safety system setpoints or control circuity. The improvements i

are positive in nature and thus subject to previous malfunction analyses. The modification will not afTect any analyses pertaming to the Loss of SDC as described in USAR Appendix G. No equipment important I i to safety will be deenergized or its setpoints altered as part of this modification. Installation of the new }

motor will not hinder the ability of safety systems to perform their intended functions. No existing l

l accidents are broadened in their scope and no new accident types are created. No new failure modes will be introduced by this modification, nor will the installation of the replacement motor propagate failure to any safety system. The replacement motors provide suflicient increase in torque to maintain positive margins under degraded voltage conditions. This improvement will ensure functional operability of the

, valve under nonnal operation and following design basis events. Testing of the motor operated valve is l to be performed in accordance with applicable Technical Specifications and testing procedures. l l Procedures will be followed during the modification to ensure operability of all components required to provide and maintain safe shutdown. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

I CED C-1998-016 l l

('USQE 1998-0096) 1 TITLE: Commercial Configuration Change - Removal of Disconnect Switch 33 D from 12.5KV System l

DESCRIPTION: This document evaluated an unauthorized modification to the 12.5KV system. A Problem Identification Report documented that 12.5KV disconnect 33-D did not exiat, although it was still shown in a system l

l operating procedure and on a station drawing (this station duwing is also a USAR drawing). The

)j l function ofdisconnect switch 33-D is to connect a step-down transformer to the 12.5KV North Overhead l Loop. Once positioned, this switch is not manipulated under any normal or emergency operating l l

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l conditions. Work performed on this portion of the 12.5KV system is under the responsibility of the Nebraska Public Power District Transmission and Distribution department. It is postulated that the pad mounted transformer served by this disconnect may have been removed based on flooding concerns near the Maintenance Training Facility located north of the protected area, or changes to the power distribution for the north meteorological tower. He disconnect was removed when it was no longer needed. Drawing 3009 Sheet I has been revised to show the removal of disconnect switch 33-D from the 12.5KV system (Drawing Change Notice 98-2132).

SAFETY l EVALUATION: The subject disconnect switch has no automatic functions and is not associated with any accident l evaluated in the SAR. No new manual or automatic functions are being added to the plant. Therefore, I

the probability of occurrence of a previously evaluated accident is not increased. This activity does not ,

l change the power source for any components. It does not impact any radiation monitoring equipment or l l any radiological / contamination control procedures, and does not result in the relocation of any l l components in any radiological controlled areas. Therefore, there is no increase in the consequences of a previously evaluated accident or equipment malfunction. The function of disconnect 33-D is no longer required because the load that it protected is no longer installed. The manner in which the 12.5KV l disconnects operate is not altered. None of the equipment supplied by this disconnect is considered to be important to safety. The malfunction of this disconnect is not discussed in the SAR. This change is associated with equipment located outside the protected area and it does not interface with any essential power sources. It does not impact the manner in which the 12.5KV distribution system is operated. The function / control / operation of the installed disconnect 50-D will not be impacted by the removal of l disconnect 33-D. No new types of accidents or equipment malfunctions are created by the removal of l this disconnect. The 12.5KV system is not used in the basis for any Technical Specification, and is not l associated with any limitations, surveillances, or discussions for any Technical Specifications. Therefore,

! there is no reduction in the margin of safety.

CED 1998-0029 l (USQE 1998-9901) l TITLE: REC-MOV-700MV Motor Replacement 1

DESCRIPTION: This CED installed a replacement motor to provide positive torque margin for REC-MO-700MV in the closing direction. This change was implemented in response to a Limitorque Technical Update which reduced the projected torque output of Limitorque actuators. The replacement motor provides sufficient torque to address this concern. The modification was completed during plant shutdown.

SAFETY EVALUATION: This modification is a replacement and does not impact any system in a way that could initiate any of the events or accidents evaluated in the SAR. It is intended as an improvement to the existing device and is a like replacement, but with a higher capacity. The safety margin will be increased, therefore, the function in the system is the same and previous accident analyses are applicable. Changes will not be made to safety system setpoints or control circuitry. The replacement motor will improve its ability to isolate non-safety related cooling loads under accident conditions. This modification will not increase the consequences of an accident as previously evaluated in the SAR. Installation of the replacement motor

! is to ensure compliance with the design basis of the Reactor Equipment Cooling (REC) system. During l installation, tag-out procedures will be in place to ensure that safety systems remain operable and that the probabihty and consequences of malfunctions of eqmpment important to safety are not increased during

the installation activities. Operabihty of the REC critical loops as stated in the SAR will not be altered l by the failure of REC-MO-700MV. No change to previously evaluated malfunctions of equipment important to safety are introduced. This modification will not alter any parameters used in presious accident analyses. Following standard installation procedures ensures that the possibility of new accident types are not created during implementation of the modification. No new failure modes will be introduced by this modification, nor will the installation propagate failure to any safety system. The replacement motor provides suflicient increase in torque to maintain positive margins under degraded voltage conditions. It will improve functional operability of the valve under normal operation and following design basis events. Testing of the motor operated valve will be performed in accordance with applicable

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Technical Specifications and surveillance procedures. Tag-out and operational procedures will be in place to ensure the availability of both REC critical loops. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

CED 1998-0033 (USQE 1998-0014)

TITLE: Realignment of Drain Valves to Eliminate Insulation Deterioration DESCRIPTION: The orientation of the strainer drain valves on the Turbine Building iIcating & Ventilation (IIV) units caused water to drain into the adjacent insulation when the drains were used. This CED added fittings to reorient the valves, along with pipe nipples on the valve discharge to allow the liquid to be expelled away fromexistinginsulation. Valves affected were: ACD-V-212, ACD-V-213, ACD-V-222, ACD-V-223, ACD-V-211, ACD-V-214, and ACD-V-221.

SAFETY EVALUATION: This activity will not change the overall design, function, or reliability of the IIV system and the configuration will not be an accident initiator. The drain line will still drain non-radioactive steam to the atmosphere. This system is non-essential and does not perform any accident mitigation function; diis activity will not result in any increased radiological efTects. This CED will not change the design, function, operation, or reliability of any equipment important to safety, nor will it induce any equipment malfinictions or failures. This activity will not change or airect any essential equipment. It will not alter the accident mitigation capability of any equipment or systems important to safety, nor change their failure modes. This activity does not create any new failure modes since it isn't changing the operation of the Turbine Building IIV units, nor does it change any existing interfaces between any other equipment and/or systems. Rerouting of the drain lines does not create any new accident scenarios. The design of this modi? .ation has properly considered seismic, thermal, structural, material, and system interaction concerns. Therefore, no new types of malfunctions are created. This activity does not affect any assumptions, calculations, procedures, or design specifications used to establish the basis for defining the plant's margin of safety; therefore, the margins of safety as defined in the basis for any Technical Specification are not affected.

CED 1998-0042 (USQE 1998-9900)

TITLE: RIIR-MO39B and RIIR-M015B/D Interlock Bypass with Alternate Shutdown Control DESCRIPTION: This CED modified the control circuitry for RIIR-MOV-39B to allow the bypass of an interlock with RIIR-MOV-15B/D for post-fire operations from the Alternate Shutdown (ASD) panel. This change was required to provide assurance that the operator can complete activities necessary to operate the Residual <

IIcat Removal (RIIR) system in the suppression pool cooling mode following a fire that requires evacuation of the Cooper Nuclear Station control room.

SAFETY EVALUATION: This modification does not impact any systems in a way that could initiate any of the events or accidents evaluated in the SAR. It will be performed while the reactor is at power; therefore, there is no need to evaluate the impact on a loss of shutdown cooling. Compliance with Technical Specifications for the RIIR system ensures that the operability and Limiting Condition for Operation requirements for RIIR are met. Tag-out provisions ensure that a Loss of Coolant Accident via misoperation of any RIIR valve will not occur. Procedure revisions for post-surveillance restoration of the interlock provide assurance diat the interlock function to minimize the potential for reactor vessel drain down and associated loss of shutdown cooling is maintain 61 in the operating plant configuration for which it is required. This modtfication will allow RIIR-MOV-39B to be opened when either RIIR-MOV-15B or RIIR-MOV-15D are spuriously opened m the event of a hot short. This preserves the design basis for the fire related special event. The interlock bypass is used to mitigate the consequences of an accident in the event post-fire remote control of RIIR is necessary from the ASD panel. In the remote event of a single failare of the bypass that disables RIIR-MO-39B, the opposite division RI-IR system would be available for post

-27 l

l i

event suppression pool cooling. A malfunction of the interlock bypass will not hinder the ability of safety systems to perform their intended functions. The modification does not overnde or introduce any new malfunctions of the safety-related RIIR system equipment. Plant tag-out guidelines will ensure that the consequences of malfunctions of equipment important to safety are preserved during the instellation activities. The modification does not alter the interface between the ASD function and the RHR functions of the plant, does not breach the barriers of any systems, nor make them more susceptible to breach or failure. The impact of the design with regard to reactor vessel drain down is mitigated by restoretion provisions in the ASD test prouxiures to ensure that the interkicks that minimize the potential for reactor drain down are maintained in their safety, standby state. Therefore, no new types of accidents are created.

The RHR-MOV-39B and RHR-MOV-15B interkick bypass for ASD control is a passive device designed for use in the event that post-fire remote shutdown capability is needed. The interlock only controls the operation of RIIR-MOV-39B and does not interface with the operation of any other devices. Therefore, no new types of malfunctions are introduced into the system. Testing for operability will be conducted per surveillance of the ASD panel once per cycle. Testing of the operability of the interlock bypass of RHR-MOV-39B ensures that the basis for Technical Specification 3/4.2.I is preserved. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

CED 1998-0052 TITLE: Altemate Drain Line for Senice Air Moisture Separator SA-MSEP-A DESCRIPTION: Plant Temporary Modification (PTM) 96-24 installed an extra drain line, ball valve, and Chicago fitting between Senice Air Moisture Separator SA-MSEP-A and drain line SA-TP-ACDA. This was installed to provide an interim solution to moisture separator drain line plugging and to allow blowdown of the drain line without losing the drain trap seal. This CED justified leaving this extra drain line as a pennanent modification. Leaving this extra drain line and valve installed will increase the reliability of the moisture separator drain trap. (Note: A summary of PTM 96-24 was previously reported to the NRC in November 1997).

SAFETY EVALUATION: The safety evaluation performed for PTM 96-24 was for a temporary modification; however, the evaluation does not change for making this a permanent modification. All responses to the Unreviewed Safety Question Evaluation were reviewed and found to adequately bound making PTM 96-24 a permanent modification. The only exception is that a USAR change was required to make permanent changes to a USAR drawing to reflect the permanent modification.

CED 1998-0055 TITLE: Permanent Test Connections for HV-AOV-261 AV DESCRIPTION: This CED was developed to make Plant Temporary Modification (PTM)97-012 permanent. This PTM

~

previously added test connections to the air lines of HV-AOV-261 AV in order to be able to quickly attach and remove diagnostic test equipment. Leaving the connections installed will increase the etliciency of troubleshooting efforts and, therefore, increase the availability of HV-AOV-261 AV, (Note: A summary of PIM 97-012 was previously reported to the NRC in November 1997).

SAFETY EVALUATION: The 10CFR50.59 evaluation for PTM 97-012 was reviewed and determined to appropriately bound the permanent modification, with the following exceptions. The previous reference to Custom Technical Specification 3.7.C was changed to improved Technical Specification 3.6 ' (

• wever, it did not change any conclusions of the evaluation. In addition, a USAR change was initi wd to e ic permanent changes to a USAR drawing.

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CED 1998-0062 (USQE 1998-0021)

TITLE: Temporary Modifications to the Bellows Shield and Steam Plugs to Allow the Use of the GERIS 2000 l ID Inspection System DESCRIPTION: The bellows shield was modified to allow the installation of the GERIS 2000 inspection system's track by temporarily n: moving the shield's ramp and cutting oft the ramp's two support arms. This temporary configuration change has not yet been restored. In addition, the main steam plug vent lines were shortened to prevent interference with the GERIS inspection device. The length removed was small and does not affect the function or operation of the steam line plugs. Therefore, the changes to the vent line were left as a permanent modification.

SAFETY EVALUATION: This modification does not increase the probabili'y of a fuel handling accident because the modification of the bellows shield does not alter the fuel bundle's path, visibility, or clearances during transfer. The modification of the steam line plugs does not reduce their structural integrity, reliability, or service conditions, negating any possibility of a vessel drain-down. Any bundle damage due to the removal of the ramp as a result ofmishandling will not exceed those previously assumed for a fuel handling accident.

Even though the ramp's removal could allow a dropped bundle to fall in between the shroud and the vessel wall, the consequences remain bounded by the refueling accident analysis. The cross-sectional flow area and the geometry of the steam line plug vent lines remains unchanged, preventing any increase in vessel drain-down rate in the unlikely event the vent should become ruptured or damaged during refueling activities. The shielding characteristics of the bellows shield and its structural integrity remain unchanged. Shortenmg the steam line plug's vent line does not alter the service conditions or stress levels within the plug or its sealing members, thus the possibility of a plug failure is not increased. The GERIS ring, in conjunction with the bellows shield, will prevent a bundle from striking the reactor pressure vessel flange's sealing surface either during mishandling or in the event of being dropped during transport. This activity does not create any new interfaces with safety related equipment, systems, or struc 'res required to maintain core reactivity control or cooling during refueling activities. No new credible accidents are created as a result of this change. This activity does not alter the function of either the bellows shic!d or the steam line plugs, nor will it result in any plant equipment, structures, or components being subjected to increased stress levels or new and unanalyzed service conditions. The modifications are incapable of creating any new unanalyzed failure modes. The design margins of the bellows shield and steam line plugs will be maintained. These activities will not reduce any margins of safety as previously defined in any design document, plant operathg procedures, or design evaluations.

CED 1998-0088 (USQE 1958-0022)

TITLE: Feed Pump A Drain Line Flange Temporary Repair DESCRIPTION: A flange gasket on the Feedwater Pump A casing drain line failed and wais leaking. The leak was not sufficient enough to require the shutdown of the pump for a pennanent repair; therefore, a leak repair compound was used. This temporary modification injected leak repair compound into a special leak repair flange installed around the existing flange. The leak repair flange was removed during the RE18 outage to allow for a permanent repair.

SAFETY  !

EVALUATION: This activity will not change the overall design, function, or re'iability of the feedwater pump drain line.

Leak scalant is compatible with the CII-2 piping material, so the probability ofline breaks is not affected.

The sealant is compatible with the reactor coolant; however, the volume of sealant injected is controlled so as not to enter the reactor coolant. The function of the drain line to drain the pump casing will not be affected by this temporary repair; thus, the repaired drain line will not be an accident initiator. This system is non-essential and does not perform any accident mitigation function; this activity will not result in any increased radiological effects. The feedwater drain flange has no Safety Design Basis and, therefore, does not perform any safety function. This activity will not change the design, function, l

i 4

l 3

1 operation, or reliability of any equipment important to safety, nor will it induce any equipment malfunctions or failures. This activity is incapable ofincreasing the consequences, including increased radiological efTects, of any equipment failures. Materials of construction and the leak sealant material are j

compatible with the pressure boundary. The temporary repair of the casing drain line does not create any

, accident scenarios different than currently evaluated in the USAR. The design of this modification has i

properly considered seismic, thermal, structural- material, and system interaction concerns. The leak repair flange meets or exceeds the original piping material and strength requirements. Therefore, no new types ofequipment malfunctions are created. This activity does not afTect any assumptions, calculations, procedures, or design specifications used to establish the basis for defining the plant's margin of safety. l The existing margins ofsafety in the basis for any Technical Specification will remain unaffected by this  :

1 temporary repair.

CED 1998-0098 j

(USQE 1998-0047)

TITLE: Reload 18 Analysis for Cycle 19 DESCRIPTION: This CED docurnented the analysis and design review associated with Reload 18 fuel bundles for Cycle 19 operation at CNS. This evaluation was performed to ensure that the Cycle 19 core design meets all 1 of the applicable safety and regulatory requirements. Reload 18 fuel bundles are of the GE9B family of designs. General Electric (G.E.) provided reload design and licensing documents that summarized the reload design and licensing calculations performed by G.E. in accordance with their NRC approved l methodology documented in NEDE-24011-P-A. The subject safety evaluation covers the configuration change associated with the re'oad fuel including the USAR changes required for the Cycle 19 reload and the Cycle 19 Core Operating Limits Report (COLR).

SAFETY EVALUATION: The Cycle 19 reload core thermal limits have been developed with NRC approved methodologies as described in the General Electric Standard Application for Reactor Fuel (GESTAR) 11 and required by Technical Specification 5.6.5. The Cycle 19 reload core will not increase the probability of previously evaluated accidents because the analyses of the core have been carried out with NRC approved methodology and have been correctly implemented as documented in the Cycle 19 COLR. Cycle 19 operation within the thermal limits specified in the COLR does not increase the consequences of any analyzed anticipated operational occurrences or accidents because the mechanical, thermal-hydraulic, loss ofcoolant accident (LOCA), and other accident general design criteria imposed on the fuel are met. The only changes for the Cycle 19 reload are in the burnable absorber and core loading pattern; the fuel mechanical design is the same as previously utilized at Cwper Nuclear Station (CNS) and, therefore, fission product inventory assumptions will remain valid for Reload 18. The analyses and thermal limits address all boundmg previously evaluated core design related equipment malfunctions. The mechanical stability and maintenance of coolable geometry as required by General Design Criteria are assured by adherence to these core thermal limits for the evaluated core design related equipment malfunctions. The ,

consequences of a malfunction of equipment important to safety previously evaluated in the SAR is not  !

increased because the mechanical, thermal-hydraulic, and LOCA design criteria imposed on the fuel and

)

core maintain the level of protection previously achieved during anticipated operational occurrences and j accidents. The possibility of a different type of accident is not created because this change does not allow for a new fission product release path, result in a new fission product release failure mode, or create a new sequence of events that results in fuel cladding failures or fission product release. The equipment or l systems required to load and provide safe operation for the Cycle 19 fuel are identical to those used in I previous cycles. No plant modifications are required to accommodate the new fuel. No new activities are required for Cycle 19 operation. Therefore, the possibility of a malfunction of a difTerent type than previously evaluated is not created. The margin of safety is defined by the thermal limits in the Cycle 19 COLR. Since CNS is operated within the thennal limits specified in the COLR, the margin of safety for each of the limits is not reduced. The stability region dermed for Cycle 19 is the same as Cycle 18 and,  ;

therefore, there is no reduction in the margin of safety with regards to reactor stability.

1 I

J

CED 1998-0105 (USQE 1998-0034) l TITLE: Temporary Power and Control Cable Installation for GERIS/ Inservice Inspection Exams

! DESCRIPTION: This Temporary Configuration Change was installed to provide electrical power and control cable inside the Reactor Building and inside the General Electric trailers to accommodate the GERIS/ Inservice Inspection Program inspections. Power was supplied by the 12.5 KV electrical distribution system.

Cable was routed through a spare penetration into the Reactor Building. The temporary configuration l was removed following the completion of the subject examinations.

! SAFETY EVALUATION: The temporary power system and associated equipment is not an initiator of any event. The probability of an accident or fire is not increased by these changes because they are performed in accordance with site procedures. The electrical loads are supplied from non-essential 12.5 KV which has no impact on accident analysis or accident response / mitigation. The firt barrier and secondary containment seals will be demonstrated to provide an equivalent level of protection to the penetrations. The penetration seal will be restored to its previous operable status within the time allotted by the applicable Limiting Condition for Operation (LCO). The design of the distribution cabling and fusing, maintaining separation requirements, and supplying loads from non-essential 12.5 KV power will prevent any impact to safety equipment. The penetration seal will be tested to ensure secondary containment integrity. The consequences of a secondary containment or fire barrier failure are unaffected. Therefore, the probability ofoccurrence or consequences of a malfunction of equipment important to safety previously analyzed in the SAR are not increased. The seals are relied upon for accident mitigation and do not in themselves create the possibility for an accident or equipment malfunction not previously evaluated. The temporary cable is routed in accordance with applicable separation requirements and the temporary penetration seals are qualified for their containment integrity application. No new initiators or failure modes are being introduced. Therefore, no new types of accidents or equipment malfunctions are created. The 12.5 KV system is not governed by Technical Specifications. Applicable LCOs are complied with during installation and removal of this temporary configuration change and penetration integrity will be verified following installation and restoration. The four hour completion time to restore secondary containment is commensurate with its importance to safety and the penetration seal will be restored within the required time frame. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

CED 1998-012 I (USQE 1998-0038)

TITLE: Weld Repair of Senice Water (SW) Cross at Turbine Equipment Cooling (TEC) Ileat Exchanger Outlet DESCRIPTION: Erosion was identified at the cross of the SW outlet of the TEC heat exchangers. A Plant Temporary Modification (98-10) was completed to secure two through wall leaks. Ultrasonic inspection of the cross revealed an eroded area surrounding the pinholes. To restore the wall thickness to that required by code, a 3/8" thick reinforcement pad was welded to the outside of the cross and flange to encompass the -roded area. This CED documented the use of the welded plate as an adequate pressure boundary.

SAFETY EVALUATION: This activity restores the wall thickness of the 24" SW pipe to code required thickness. The repair will be done in accordance with the piping code, B3 L 1, and Cooper Nuclear Station procedures. The repair will not affect system performance or integrity and will not increase the probability of a previously evaluated plant event. This portion of the SW system has been previously analyzed in the USAR as not performing any accident or event mitigating function. A flow restricting orifice and procedures assure nonnal flow to essential systems. If SW header pressure drops to 20 psig, then the TEC system is automatically isolated from the essential components of the SW system. For this reason, the application l of the weld pad to abate any SW leakage will not increase the consequences of a previously analyzed event. This activity does not affect in any way existing protective features in the SW system Osigned to mitigate the consequences of equipment malfunction. It does not alter the operating parameters of the

-3l-

SW system or any ofits users. This portion of the SW system is not required to backup any nuclear safety related equipment. The SW line functic and configuration are not alTected; therefore, the possibility of creating a ditTerent type of plant event does not exist. If this activity does not stop the leak, the turbine building basement may flood; however, this is not a new or unanalyzed failure mode. It has been considered and addressed in the USAR. Therefore, this repair does not create any new faihire modes.

A failure of the TEC SW line in the turbine building basement will not result in a loss of cooling water to essential users, nor will it result in conditions beyond those previously analyzed. Therefore, this activity will not result in any reduction in the margin of safety.

CED 1998-0126 (USQE 1998-0030)

TITLE: Installation of Pre-Fabricated Buildings (Fab Shop and South Rad Material Storage Building)

DESCRIPTION: Two vendor-supplied pre-fabricated commere; ; metal buildings designated as the Fab Shop and the South Rad Material Storage Building were installed in the site yard by this CED. These buildings are not required to be designed to the requirements for Class I or 11 structures and equipment. The use of the South Rad Material Storage Building for storage of radioactive material will be controlled by the Radiological Protection Department and conform to Radiological Protection Procedures and USAR requirements. The material stored in the building will be restricted to material with a general dose rate ofless than 5 mrem /hr at 12",

SAFETY EVALUATION: These buildings are not required to be designed to the requirements of Class I or 11 structures and equipment as specified in the USAR because they will not house any equipment or components and are not structures: 1) whose failure or malfunction might cause or increase the severity of an accident which would endanger the public health and safety,2) which are required for safe shutdown and isolation of the reactor,3) which are important to reactor operation,4) which are essential for preventing an accident which would endanger the public health and safety, oi 5) which are required for the mitigation of the consequeaces ofdesign accidents. In addition, these buildings will not have any structural interfaces with j Class I or 11 buildings as described in the SAR. Therefore, the seismic, wind, and tornado loading criteria for Class I or 11 structures and equipment are not applicable to these buildings. The most likely etTect of the addition of the new buildings to the operation of the plant will be the failure of the roof or the siding due to high wind velocity above the Uniform Building Code design requu ement. This failure may cause the roofor siding to hit the main station transformers, the switchy .1, or Class I or 11 buildings which in turn may result in a Loss of OITsite Power. Loss of Offsite Power a an event already considered in the USAR. He South Rad Material Storage Building will be used to wre radioactive material that is now at various locations in the plant. He radiological consequences of pres iously evaluated accidents are not changed by storing the radioactive material inside the building. The use of the South Rad Material Storage Building will not affect any accident assumptions, nor will it cause a change to any system interface in a way that would increase the likelihood of an accident. There will be no essential equipment located inside the new buildings. The closest safety-related equipment or structures in the proximity of the two new buildings that can be impacted by the failure of the buildings are the Z-Sump and the Elevated Release Point (ERP) tower. The 100 mph wind load criteria, which is required for all Class I and 11 structures, does not include a requirement for an evaluation ofimpact loads caused by wind-borne or wind-blown objects. Because of this design criteria infonnation, the new buildings, which are not Class I or 11 designed structures, do not present a hazard to the safety-related function of the Standby Gas Treatment system or the Class I components of the Z-Sump and the ERP tower. The new buildings do not directly interface with equipment important to safety and do not affect the operation of accident mitigation systems. Therefore, there is no increase in the probability of occurrence or consequences of a malfunction of equipment important to safety. This activity does not produce any credible accident scenarios not previously evaluated in the SAR. The only potential failure modes of the buildings are during seismic and tornado events. These failures would not initiate any new accidents. Failure of the roof or siding panels due to high winds will not impact the capability of any equipment or structure to perform its essential function as required for various design conditions. No new types of malfunctions are created by the addition of the two new buildings. There are no margms of safety specified in any Technical Specification associated with this activity.

j CED 1998-0200 l 4

(USQE 1998-9902)

TITLE: 125 VDC Fuse Replacement with IIigher Interrupting Cunent Rating Fuses DESCRIPTION: This CED was developed to replace fourteen 125 VDC fuses. A Problem Identification Report identified that Bussman type "RIC" fuses installed in the 125 VDC fused disconnects had the incorrect fuse j intenupt rating. The required interrupt rating per design documents is 20,000 amperes; the rating for the j installed fuses was 10,000 amperes. The subject Bussman type RIC fuses were replaced with Gould l

Shawmut type AJT fuses with a 100,000 Amme Interrupting Current rating. A revision to this CED was l

generated to address the replacement of fuses in disconnects EE-DSC-125A and EE-DSC-125B. WM performed for the original CED identified that fuses installed in these applications were FRN-R-20s.

instead of RIC fuses as stated in the CED. This CED revision determined that the continued use of the FRN-R-200 fuses in these applications was acceptable.

SAFETY EVALUATION: Th replacement of these fuses will bring Cooper Nuclear Station (CNS) into compliance with its intended circuit protection design. The fuses and their associated equipment are not accident initiators.

During the replacement of the fuses, Shutdown Cooling (SDC) will be configured so that the DC Motor l Operated Valves are in their required positions as needed to support SDC. The disconnects will be I decnergized as determmed by Operations and plant conditions. Therefore, this work will not increase the probability of an accident previously evaluated in the SAR. The consequences of any postulated failure ofplant equipment is not affected since a postulated fuse failure is within CNS single failure analysis a circuits requiring single failure criteria design. The effects of a failure of a fuse or associated equipmem remains unafrected by this work. Procedures are in place that address the loss of various parts of the DC distnbution system. Therefore, this CED will not increase the consequences of an accident previously evaluated in the SAR. These fuses are required to protect safety related equipment in the event of an electrical fault. By replacing these fuses with the correct type for the application, the probability of an equipment malfunction is reduced. Failures of any equipment in the pre-modification configuration will have the same results in the post-modification configuration. No new accidents are postulated as the result of this fuse change as these fuses are not associated with any type of accident initiator. No new malfunctions are created by the replacement of the existing fuses with the proper fuses. This work will i bring CNS into compliance with its circuit protection design. The functions of the fuses remain unchanged and no new equipnw:nt is added by this work. Fuse replacement does not affect the Technical l

Specification margin of safety in any way. The fuses support the operation of safety related eqmpment l as required by Technical Specifications.

i CED 1998-0201 (USQE 1998-0061)

TITLE: 250 VDC Fuse Replacement With fligher Interrupting Cunent Rating Fuses DESCRIPTION: This CED was developed to replace twelve 250 VDC fuses. A Problem Identification Report identified that Bussman type "RIC" fuses installed in the 250 VDC fused disconnects had the incorrect fuse interrupt rating. The required interrupt rating per design documents is approximately 25,000 amperes for the 250 VDC system. The rating for the installed fuses was 10,000 amperes. The subject Bussman type RIC fuses were replaced with Gould Shawmut type AJT fuses with a 100,000 Ampere Interrupting Current rating. A revision to this CED was generated to address the replacement of fuses in disconnects EE-DSC-250A and EE-DSC-250B. Work performed for the original CED identified that the blades of these fuses were not fully engaged in the fuse holders. This CED revision replaced the " Class R" fuse holders with " Class J" fuse holders to properly accept the new AJT " Class J" fuses used in these applications.

SAFETY EVALUATION: The one for one fuse replacement will not increase the probability of occurrence or consequences of an accident previously evaluated in the SAR. The operation of the fuses and the 250 VDC system will not initiate an accident. The replacement fuses will bring Cooper Nuclear Station into compliance with its intended circuit protection design. A postulated single failure of a fuse is within the single failure analysis on circuits requiring single failure design criteria at CNS. The efTects of a failure of a fuse or associated equipment remams unaffected by this work. Disconnects EE-DSC-250A/B are used to connect the spare Battery Charger 1C to the 250 VDC Bus 1 A and iB, respectively. The impact of a loss of Charger !C while it is supplying the 250 VDC switchgear is the same as the loss of Charger 1 A or 1B. The impact on radiation monitoring equipment, procedures, and contamination control are not any different.

Therefore, the consequences of an accident or eqmpment malfunction previously evaluated in the SAR are not increased. The changes made by this CED will not alTect the normal operation of the 250 VDC batteries, buses, safeguards loads, or emergency loads supplied by the 250 VDC battery system. The new fuses are the correct type to protect the safety related loads on the 250 VDC system, and will decrease the probability of a malfunction of equipment important to safety previously evaluated in the SAR. The fuse clip modification will occur intemal to the 250 VDC disconnects for Battery Charger 1C. The fuse clips do not impact the operation of the disconnect. Failures of any equipment in the pre-modification configuration will have the same results in the post-modification configuration. No new accidents are postulated from this fuse replacement, as the fuses are not associated with any type of accident initiator.

The manner by which the 250 VDC disconnects are operated, controlled, or maintained is not altered by this change. No new malfunctions are created by the one for one fuse replacements with the proper fuses.

l The function of the fuses remains unchanged, and no new equipment is added to the design. The new fuses will support the operation of safety related equipment required by CNS Technical Specifications.

The spare battery charger is discussed in the Technical Specifications; however, operation of the disconnects is not discussed. The manner in which Battery Charger 1C is used and how it is cor.accted to the 250 VDC switchgear is not altered by the revision to this CED. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

l CED 1998-0217 i

(USQE 1998-0067)

TITLE: Overpower Relays Setting Change - EE-REL-lFE(32) and EE-REL-1GE(32)

DESCRIPTION: During the performance of the emergency diesel operability ran, it was recognized that the overpower relay setpoint was below the required operations procedural requirements. A revision to Nuclear

. Engineering Department Calculation (NEDC) 91-0220 was performed which changed the value to 115%

l of the normal 4000 KW loading. This CED raised the setpoint to the value in NEDC 91-0220 in order to ensure the overpower relays will perform their design function. This CED was performed during the refueling outage during a time when the 4160 VAC bus was considered part of the non-protected division. The USAR was subsequently updated per USAR Change Request 98-101 to change the setpoint for the diesel generator overpower relay from 110% to 115% of rated power.

SAFETY EVALUATION: This activity will result in separation of the e sential and essential buses for a given division; l however, all sources ofpower will remain capable %:rforming their design function. The safety related bus will maintain two sources of power aveilable during the performance of the surveillance procedures (the Emergency Transformer and its corresponding Diesel Generator). All AC electrical power distribution systems will remain capable of performing their design function and divisional separation will ensure the opposite division is not affected. Performance of this change will not affect any of the factors that contribute to the probability of an accident since the IF and IG bus, undervoltage protection, load l

shedding, and load sequencing circuitry that contribute to mitigate an accident are not accident initiators.

l The operation of this circuitry important to the protection of the IF and 10 bus and subsequently the plant, will not be altered and will be able to perform its intended function. All loads of the safety related bus will remain capable of performing their design function . Equipment used to mitigate an accident or its consequences is not compromised or defeated. There is no reduction in the protection of the public health and safety since the function of equipment designed to control the release of radioactive material l

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the diesel generator from the non-essential loads via the IFA or 1GB breakers to prevent overloading i

during the surveillances. The system operation and function will not be affected in a manner in which the j margin ofsafety will be changed.

l CED 1998-0228 l (USQE 1998-0094) l TITLE: l Permanently Install IIcat Trace on the Standby Liquid Control (SLC) Pump Discharge Relief Valves SLC-RV-10RV and SLC-RV-11RV and the Associated Relief Valve Piping DESCRIPTION: Plant Temporary Modification %-31 previously installed temporary heat trace and insulation to the SLC relief valves and their associated discharge piping in order to maintain the liquid control solution at least 10 degrees above the saturation temperature. This CED was generated to allow the installation of the SLC heat trace to become permanent with some modifications to the heat trace connections. The temporary heat trace was powered from an independent 120 VAC receptacle. The permanent installation l has the discharge piping heat trace spliced into the existing heat trace, utilizing installed junction boxes I and controlled by the heat trace control panel. I SAFETY l EVALUATION: Failure of the SLC system or failure of the boron solution temperature controls are not initiating events ,

to any of the design basis accidents or abnormal transients previously evaluated in the SAR. The l

radiological consequences of any previously evaluated accidents or equipment malfunctions are not increased by the addition ofheat trace to the SLC relief valves and associated piping. The addition of the j

heat trace for the relief valves and discharge piping to the previously installed heat trace for the balance 1 of the SLC system will not increase the probability of a malfunction since the heat trace will be supplied from a common power supply. Controls for the SLC heat trace will be from one common panel and no new malfunctions are introduced as a result of this design change. The loss of temperature controls for the SLC system has already been analyzed in the USAR. The heat trace addition is an enhancement to the system with respect to the temperature control of the boron solution. No new accident initiators are introduced. The malfunctions of the new heat trace are identical to those previously evaluated for the balance of the system heat trace. The currently installed electrical protection is adequate for the additional heat trace. The margin of safety as defined in the basis for any Technical Specification has not been reduced.

CED 1998-0253 (USQE 1998-0077) l TITLE: Main Steam Isolation Valve (MSIV) Seat Design Change DESCRIPTION: The MSIVs were having difliculty satisfactorily passing post-maintenance local leak rate testing. This was believed to be due to the existing seat design. Therefore, this modification was generated to allow the use of an interference seat angle design, where the body seat angle is ground and dressed to a 45*

angle, and the disk seat machined and dressal to a 43.5* to 44* angle, providing a 1/16" to I/8" final seat I

contact width. This ensures a tighter seal by increasing the seat stress and will compensate for an increased amount of mis-alignment between the disk assembly and the seat ring and still maintain full circumferential contact. This modification was performed on MSIVs 80A,80B,80C, and 86A during RFOl 8. The modification was not successful on MSIVs 80A,80C, and 86A; therefore, the disk seat angle was retumed back to 45* on these valves. MSIV 80B was the only valve left at 43.5*

SAFETY EVALUATION: This activity does not reduce the valve's basic temperature and pressure rating, nor does it alter the design l of the operator or its controlling component or circuitry. Therefore, it is incapable ofincreasing the pmbability of a valve body mpture or a random isolation due to stem / operator failure while under power.

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This activity has not made any changes to the stem or the disk's guides, ensuring that closure time will not be affected. By modifying the seat design, the probability of achieving the required seat leak tightness has been increased, ensuring that the assumptions used in any existing accident analyses or engineering assumptions remain valid. Even though the seat stress levels will increase, they will not exceed the metallurgical limits of either the seat's hard facing or its base material by a comfortable margin. By l developing a higher seat stress, the probability of obtaining a tight seal is enhanced, and the probability l of steam cutting reduced because it will reduce the likelihood of a seat leak. In the event the seat should fail, the basic geometry used to limit the consequences of this failure mode remain in place (i.e., the l . disk / body and seat / disk clearances remain the same). This activity cannot result in a common mode failure as the seat stress levels remain within acceptable limits, and will be fully bounded by normal f pressure stresses. Therefore, this activity will not increase the consequences of an equipment i malfunction. 'Ihis activity (k)es not create any new interfaces between any existing systems or structures, t

add new equipment, or alter any operating parameten or operating procedures. It does not alter the forces on any piping system or components, result in any new or unanalyzed forces either on the valve's operator or its intemals, increase or decrease steam flow or pressure, alter the valve's basic operation, or reduce l- its ability to close under pressure. Therefore, no new failure modes are created and no new types of l

accidents or equipment malfunctions are created. The ability of the MSIVs to reliably close and obtain a tight seal has not been reduced. Rather, by increasing the seat stress and incorporating a wear compensating seat design, the ability to obtain a tight seal is increased, thereby enhancing any existing margins of safety.

CED 1998-0282 (USQE 1998-0085)

TITLE: Main Steam Isolation Valve (MSIV) Riblet Guide Pads DESCRIPTION: The MSIVs experienced difIiculty in passing post maintenance local leak rate testing. This was believed to be due to the disk guide location and guide wear. Therefore, this CED authorized the installation of up to four additional short guide pads or riblets to the valve interior above the seat. These guide pads will ensure a tighter seal by reducing the lateral movement of the disk as it seats and increase the guide wear life. This modification was performed on MSIV 80A during REl8. This CED may be used to modify other MSIVs in the future.

SAFETY EVALUATION: This activity does not nx!uce the valve s basic temperature and pressure rating, nor alter the design of the operator or its controlling circuitry. Tnerefore, it will not increase the probability of a valve lxxly rupture or isolation. Valve closure time is not affected. By adding the guide pads the probability of achieving the required seat leak tightness will be increased, ensuring that the assumptions used in any accident analyses or engineering assumpons remain valid. MSIV design, operation, and function will not be adversely impacted; therefore, the probability of a malfunction of equipment important to safety is not increased. The guide pads will assure reliable and consistent leak tightness of the valves, which will in turn prevent the consequences of an accident from being increased. No new failure modes are created and the consequences of a malfunction are not increased. This activity does not create any new interfaces between any existing systems or structures, add new equipment, or alter any operating parameters or procedures; thus, no new types of accidents are introduced. Forces on any piping sy stems or components are not affected, no new forces are created on the valve operator or its intemals, steam flow or pressure are not increased or decreased, and the valve's basic operation is not altered. Therefore, no new types of equipment malfunctions are created. By increasing the ability to reliably obtain a tight seal and reducing guide wear, the existing margin of safety is enhanced.

CED 1998-0284 (USQE 1998M ;)

TITLE: eactor Equipment Cooling (REC) Quad Fan Coil Unit (FCU) Flow Control Valve Replacement l

DESCRIPTION: The minimum acceptance criteria established by Nuclear Engineering Department Calculation 97-087 for REC tlow to the Southeast and Northeast Quad FCUs was not being met. The Southwest and Nonhwest Quad FCUs narrowly met their acceptance criteria. The correct REC flow to the quads must be established to suppon the operability of the Emergency Core Cooling Systems (ECCS) located in those awas. Tnerefore, this CED modified all four FCUs to increase the REC flow rate to achieve the required acceptance criteria. This modification replaced the temperature control valve and bypass line sections ofpipe in each quad with a less restrictive 2* pipe and raanual throttling configuration. Balancing of the REC system flow was accomplished as pan of the aweptance testing to ensure that REC system cooling is available to support the design function for all applicable components.

SAFETY EVALUATION: The probability of any accident or transient event described in the SAR is not atTected by this change since all equipment design ratings are satisfied and all safety functions are maintained by the new design configuration. The consequences of any accident or transient event described in the SAR are not affected since no new failure modes are introduced for the FCUs and since the failure of a single train of ECCS has been evaluated in the SAR. This modification installs a single valve / pipe in place of a more complex configuration with a higher number of potential component failures. The new configuration meets all applicable codes and system design requirements. This modification does not rJd or cause any new ,

failure mechanisms than those previously evaluated for the REC system. Therefore, the consequences l cannot be any greater than those previously evaluated. There is no possibility of creation of a new accident since this modification merely establishes a simpler piping pathway between existing piping components.1Jo accident initiators are associated with REC supply lines. No new types of malfunctions will be created since the only credible malfunctions include rupture or blockage of the REC line to the FCUs, both ofwhich are passive failures and both of which are unchanged from the original design. The margin of safety is preserved since the FCUs will now be capable of performing their design safety function to support ECCS operation. The increased cooling water flow requirements are well within the REC pump capabilities.

l CED 1998 0289 Plant Temnorary Modification (PTM) 98-04 TITLE: Steam Jet Air Ejector (SJM) Steam Chest Pressure Instrumentation DESCRlPTION: PTM 98-04 installed sensing lines and pressure indicators on SJAE nozzle steam chests (" blue set" +

ejectors only). This PTM was installed during the 1998 mid-cycle outage to allow SJAE performance to be trended and evaluated during the remainder of fuel cycle 17.

CED 1998-0289 was subsequently implemented during the 1998 refueling outage to make PTM 98-04 j a permanent modification In addition,the CED added valves and tubing to allow lineup of the SJAE " red  !

set" ejectors to the pressure instruments installed under PTM 98-04. Measurement of inlet steam pressures is required to perform efTectise evaluation and troubleshooting of sub-standard SJAE performance.

SAFETY EVALUATION: All tubing and instrumentation will be installed using original design specifications for MS-1 piping.

Original design of the MS-1 system bounds the probability of occurrence of a plant event evaluated in the SAR. This change is passive and does not change the design function of the system. The SJAE is not relied upon for mitigating the consequences of a plant event. PTM equipment kication will not impact operator res[xmse to plant events. The SJAE system is not important to safety as evaluated in the SAR.

The tubing and gauges will not change the design function of the SJAE system. Steam leaks and missile generation are bounded by existing design criteria. The original system design will be maintained and i components do not affect equipment important to safety; therefore, no increase in the consequences of l

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f equipment malfunction will occur. Potential plant events include steam leakage. No new release paths are created by the addition of the tubing and gauges. Steam leakage is not a new type of failure. Failure of this tubing will not cause SJAE system failure. Separation criteria remain unchanged. The tubing does i

not interact with any equipmmt which can reduce the margin of safety. It is in a Seismic 11 area and does not impact equipment important to safety. Existing margins and analysis remain unchanged. The original safety evaluation performed for PTM 98-04 envelopes the work to be perfbrmed as a result of CED 1998-0289.

CED 1998-0302 Operatine License Chance Reauest 99-004 USAR Chance Regrest 99-007 Setnoint Chance Reauests 98-92 and 98-93 f

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(USQE 1998-0102) f l TITLE: Top of Active Fuel / Fuel Zone Zero DESCRIPTION: The fuel zone, wide range, containment spray interlock, and steam nozzle reactor water level instrumentation were previously scaled from a reactor vessel elevation called " Top of Active Fuel" l l (TAF). The original 144" fuel had a TAF elevation of 352.56". TAF was not changed when 150" fuel l was placed in the reactor. The current 150" fuel results in a TAF clevation of 358.56". This CED re-scaled the instnanents that were scaled from the 352.56" elevation to a point that will be defined as Fuel l

Zone Zero (358.56" of Bottom IIcad Invert). This new Fuel Zone Zero elevation provides a stable l l reference point that will encompass the parameters for 150" or shorter fuel. Instrumentation fbr fuel zone, wide range, and steam nozzle reactor water level indicators and recorders were re-scaled, transmitters were re-calibrated, and setpoint changes were generated to re-calibrate the containment spray interlock switches. In addition, the term " Bottom of Active Fuel" was changed to " Bottom of Fuel" as all fuel in the reactor is treated the same. The Technical Specification Bases and USAR were subsequently revised to reflect the changes made by CED 1998-0302.

SAFETY EVALUATION: The perfbrmance of this CED will climinate the error based on 150" fuel design and will ensure that operator actions based on water level in the vessel are performed at the proper level. This CED affects instrumentation used for indication and containment spray interlock instruments; no automatic actions are changed by the scale changes of these instruments. This CED affects Regulatory Guide (RG) 1.97 j instrumentation; however, the previous commitments for complying with RG 1.97 will not be affected. l The instrument re-scaling and re-calibration will ensure that the vessel level is accurately displayed, to assist in the decision making process of the operators. Therefore, performance of this CED will not increase the probability of occurrence or consequences of an accident previously evaluated in the SAR.

No instrumentation is being replaced by this CED. The existing instrumentation will be re-calibrated and have the scales for recorders and indicators replaced to ensure accurate readings of vessel level. The setpoint change associated with the containment spray interlock will not change the actual vessel level for actuation, but the reference input for this indication. All failure modes of the instrumentation will remain the same and no new failure modes are introduced. The instrumentation loops alTected by this l

CED will be removed from ser ice, calibrated, and returned to sersice using approved Surveillance '

Procedures. Therefbre, the probaility of occurrence or consequences of a malfunction of equipment important to safety previously evaluated in the SAR will not be increased. The existing instrumentation qualifications will not be affected by implementation of this CED. No difTerent types of accidents or equipment malfunctions are created as a result of this CED. This CED will take into account the 150" fuel design which is currently in use at CNS. The setpoint change associated with the containment spray interkick will not change the allowable value or the actual vessel level for actuation. Taking into account the fuel design will ensure that the actions based on Emergency Operating Procedures and Severe Accident Management Guidelines will be carried out at the proper vessel level in relation to the fuel.

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

i CED 1999-0055 (USQE 1999-0024)

TITLE: Fire Door Strike Plate Shim Addition DESCRIPTION: Building fire doors R501 and R502 did not meet the minimum latch bolt throw requirements as required by Cooper Nuclear Station Procedure 6.FP.604. Therefore, this CED installed shim plates under the j strike plates in order to ensure the 3/8" latch engagement requirement is satisfied. This modification was I necessary to maintain the fire resistance rating of the doors. In addition, Procedure 6.FP.604 was revised to allow maintenance of the minimum strike engagement.

SAFETY EVALUATION: The probability of a fire is not increased by this activity. Fire doors and their alteration are not accident precursors or initiators. No other special events can be caused by fire doors. By maintaining the fire resistance raung of the fire doors, the consequences of the fire event are maintained constant with respect to the current Fire llazards Analysis. The dimensional criteria being altered do not impact the analyzed consequences ofother events for which they are credited such as flooding response or building isolation.

Fire doors are not credited to mitigate the consequences of any other events as described in the SAR.

Only the fire doors are impacted by this activity. This activity is being taken to ensure that equipment failure does not occur. No additional failure modes are being introduced, including human error. No new types of accidents or equipment malfunctions are created. No other malfunction of a fire door can be pmduced beyond that which is currently analyzed. The fire doors are no longer contained in the Technical Specifications and the marg; f safety for the Technical Requirements Manual has not been reduced because the fire resistance rati , f the doors is being maintained.

SDC 95-051 TITLE: Plant Management Infonnation System (PMIS) Data Concentrator Replacement Software Modifications DESCRIPTION: 'litis SDC made modifications to the PMIS soihvare as follows: 1) changes dictated by the replacement of the existing PMis data concentrators,2) changes dictated by placing PMIS on a new computer, and

3) changes dicated by placing PMIS on an up;;raded version of VMS. All of these modifications can be traced to changes made as part of Minor Modification 93-179.

SAFETY EVALUATION: This SDC afTects non-safety related systems which do not have a direct interface with equipment important to safety. The Rod Position Indication System / Rod Worth Minimizer computer is not affected j

by this SDC. The 99% reliability factor required for PMIS by NUREG-0696 will be maintained. This SDC does not change the function of any safety related equipment. The existing PMIS hardware and software will remam in service until the new system has been verified. PMIS is not used to mitigate the '

consequences of a malfunction of equipment important to safety. PMIS has no direct control over plant systems. It serves as an information source only. PMIS and the implementation of this SDC cannot i create or contribute to a new or different type of accident or equipment malfunction than previously evaluated. The software modified by this SDC will not affect a margin of safety as defined in the basis for any Technical Specification.

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l PLANT TEMPORARY MODIFICATIONS (PTMs) i NOTE: PTMs 95-02,95-23,95-30, and %-22 were completed prior to this reporting period, but not previously  !

reported.

l PTM 95-02 '

TITLE: Drywell Personnel Airlock Door Strongbacks DESCRIPTION: This PTM authorized the installation of drywell personnel airkick door strongbacks to remain in place during the operating cycle. The strongbacks on the inner dryu ell airlock door were left in place after the airlock pneumatic test was performed The strongbacks were subsequently removed during a plant refueling outage. l SAFETY EVAL;UATION: Iraving the strongbacks in place on the drywell airlock door aller pneumatic testing is complete will not affect the capability of the door to maintain containment integrity under any normal or design basis accident condition. Neither the airlock door nor the strongbacks can initiate an accident when the l

strongbacks are left in place and the consequences ofpreviously evaluated accidents will not be increased. l Iraving the strongbacks in place will not reduce the sealing capability of the door seals and will not have any efTect on the door kicking mechanism. In addition, the strongbacks will not atreet the structural 1

capability of the door and no drywell airlock door components are modified during the installation of the '

strongbacks. Therefore, there is no incicase in the probability of occurrence or consequences of a malfunction of equipment important to safety. No new failure modes are introduced and the types of malfunctions that could occur are unchanged. Pneumatic testing of the drywell airlock is performed afler the strongbacks are installed to verify that the Technical Specification containment integrity requirements are met. Therefore, leaving the strongbacks in place does not reduce the margin of safety as defined in the basis fbr any Technical Specification. l PTM 95-23 TITLE: Temporary Power for Air Compressor DESCRIPTION: This PTM disconnected the plasma welder located on the 903' level of the weld shop and hard wired an air compressor in its place. The existing 150A fuses were replaced with 100A fuses. This PTM was implemented to provide power for the blast shack air compressor in support of the IE16 Turbine Generator Maintenance Outage. The PTM was removed when the outage work wa., completed.

SAFETY EVALUATION: The equipment and electrical protection are within the design of the facility and will not increase the probability of an accident. Equipment and electrical circuits are not safety related and provide adequate protection of safety related equipment. This equipment is not used to mitigate the consequences of an accident. It does not support safety related equipment, nor can it indirectly alTect such equipment. This PTM will not cause a failure of equipment important to safety, nor will it cause a failure of other equipment powered from the same mefor control center. All equipment is being installed within the design loadings of the system. All parameters are within design and will not increase the possibility of a new accident or malfunction of equipment important to safety. The logic and protective features of the electrical system remain unchanged by this PTM. This PTM will not change any safety limits, nor will it affect any Technical Specification limits, therefore, there is no reduction in the margin of safety.

l PTM 95-30 )

l l TITLE: Temporary Power for Turbine Rotor Bore Examinations I

DESCRIPTION: This PTM disconnected the plasma welder located on the 903' level of the weld shop and hard wired in  !

its place equipment to be used for high pressure turbine rotor bore examinations. The existing 150A 40- -

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r-fuses were replaced with 100A fuses. This PTM was implemented to provide power for equipment to l be used for rotor bore examination during the RE16 Turbine Generator Maintenance Outage. The PTM ,

l was removed upon completion of the examinations. I

! SAFETY EVALUATION: The equipment and electrical protection are within the design of the facility and will not increase the probability of an accident. Equipment and electrical circuits are not safety related and provide adequate l l protection of safety related equipment. This equipment is not used to mitigate the consequences of an j accident. It does not support safety related equipment, nor can it indirectly affect such equipment. This <

! PTM will not cause a failure of equipment important to safety, nor will it cause a failure of other equipment powered from the same motor control center. All equipment is being installed within the design loadings of the system. All parameters are within design and will not increase the possibility of l a new accident or malfunction of equipment important to safety. Installing a smaller rating fuse will not affect the system wiring and will provide protection for the high pressure turbine rotor bore equipment.

l The logic and protective features of the electrical system remain unchanged by this PTM, therefore, there l l is no reduction in the margin of safety.

l l PTM 96 22 TITLE: Isolation of Annunciation on C Phase Main Power Trusformer DESCRIPTION: This PTM isolated Control Room annunciation of low oil level or a lifted pressure relief device on the C Phase Main Power Transformer from its 125 VDC power source. This isolation was accomplished by lifling a lead and was required because the annunciation circuitry had a ground on it. The PTM was subsequently removed when the ground was repaired during a plant outage.

SAFETY EVALUATION: This activity will not cause any of the design basis accidents previously evaluated in the SAR and will not impact any other plant equipment that could cause such an accident. The main power transfonners are not used for the mitigation of any accidents described in the SAR. This change only impacts Control Room annunciation of a main power transformer which is not classified as important to safety. By isolating the associated annunciator circuits, the probability ofimpacting the 125 VDC B Train System with a second ground is reduced. This change does not impact any plant eqmpment that is important to safety. No new types of accidents are introduced as the main power transformers do not impact nuclear safety. The pressure rehef devices and local indication for the transformer remain operational. As the main power transformers are not included in the basis for any Technical Specification, there is no reduction in the margin of safety.

PTM 97 46 TITLE: "D" Sparger Pump Discharge Strainer Backwash Piping DESCRIPTION: This PTM installed a temporary patch over a pinhole leak on the "D" spirger pump discharge strainer backwash piping. This repair was performed to maintain the ability t.) backwash the strainer and to j prevent water intrusion into electrical equipment. The piping was sut sequently replaced when plant l configuration allowed.

SAFETY EVALUATION: This activity will not increase the probability of a Circulating Water (CW) or Service Water (SW) faihire.

l The patch will divert water away from plant components. Failure of the patch will not increase the consequences of a plant event because the sparger system is not needed for SW system operation nor relied upon to mitigate the consequences of an accident. SW does not rely on the sparging and screen wash system to perform its safety function. Failure of the patch will not cause a malfunction of equipment important to safety. The CW equipment is not important to safety as evaluated in the USAR nor is it relied upon to mitigate the consequences of an accident or equipment malfunction. Failure of the CW system bounds any failures of the patch. No new failures will occur as a result of the installation or failure of the patch. Loss of screen wash and sparger is the only appreciable malfunction and is already analyzed.

Installation of the patch will not create any new types of equipment malfunction. The function of the SW L

i system is not changed by this PTM. SW screen blockage up to 95% has been previously analyzed.

Existing Technical Specification margina of safety are unchanged.

PTM 97-47 1

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TITLE: Reactor Core Isolation Cooling (RCIC) Steam Supply Drip Leg Drain to the Main Condenser DESCRIPTION: This PTM facilitated a temporary steam leak repair on a 2" pipe elbow on the RCIC steam supply drip leg. A half-coupling was installed on the pipe / elbow around the leak hole. This PTM also authorized temporary removal of piping insulation in the immediate vicinity of the leaking elbow in order to allow for installation of the half-coupling. It was determined to be acceptable for the insulation to remain olT l for the duration of the PTM due to the dose that would be required for re-installation. Piping and j insulation were subsequently replaced during plant shutdown. l SAFETY  !

EVALUATION: The leaking piping is in a non-safety related portion of the RCiC system and is subject to Class 11 Seismic bading conditions. The probability of an accident is not increased due to the fact that the piping and half-

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c mpling will make up a pressure boundary that continues to meet all current design requirements. The probability of a steam line break is not increased and it does not challenge the bounding steam line break  ;

design basis accident. & PTM is located outside primary containment and does not affect the operation

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of the RCIC steamisolation valves A failure of the drip leg piping would affect secondary containment but this event is bounded by the main steam line break accident dose consequences. 'lhe loss of condenser vacuum event is also bounding for this activity. Contingency plans will be in place to minimize the possibility of a plant trip due to a loss of condenser vacuum. The alTected piping is not important to safety. The PTM will not affect the operation of the drip leg piping or its Class 1 Seismic ability.

Insulation being n: moved has been evaluated for its effect on equipment in the area and it will not result in excessive steam tunnel temperatures. The PTM will have no etrect on the RCIC system or any other system important to safety and no new failure mechanisms are introduced. The protective features for steam line break mitigation, described in the Technical Specifications, are not altered by this PTM. The existing margin of safety remains unchanged.

PTM 97-48 i

TFFLE: Modification of Manual llandwheel on RCIC-AO-A034 i DESCRIPTION: A replacement valw was installed for RCIC-AOV-A034 under a Replacement Component Evaluation.

Due to a difference in physical size, the replacement valve was relocated slightly with respect to the original valve. Ilowever, this relocation resulted in the manual handwheel on the RCIC-AOV-AO34 actuator slightly contacting the air actuator for RCIC-AOV-AO35. As a result, this PTM was issued to authorize alteration of the existing RCIC-AO-AO34 handwheel so that suflicient clearance (one inch) is maintained between the RCIC-AO-A034 handwheel and the RCIC-AOV-A035 air actuator. This PTM was removed when RCIC-AOV-A034 was subsequently replaced under CED 1998-0164.

SAFETY EVALUATION: The modification to RCIC-AO-A034 does not affect any plant event initiators. h Reactor Core Isolation Cooling (RCIC) system is a standby cooling system which is designed to respond to certain plant events. The function of RCIC-AOV-A034 is to remain in the open position when RCIC is in standby to provide a drainage pata for condensation which may have accumulated in the RCIC steam supply line.

This PTM does not atTect that function. Seismic qualification of the valve / operator assembly and piping system is not affected and the modification of the handwheel does not afTect the ability of the valve to open or close. There is no potential for an increase in ofTsite dose and no increased potential to exceed 10CFR100 limits. The nanual handwheel is only used as a manual down travel stop for such activities as maintenance. This PTM does not create the possibility of a plant event or equipment malfunction of a different type than previously evaluated because the modification to the handwheel does not affect the operation of the RCIC-AOV-A034 valve or the RCIC system. For these same reasons, there is no reduction in the margin of safety as dermed in the basis for any Technical Specification.

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PTM 98-05 TITLE: Installation of Patch on Reactor Feed Pump (RFP)"B" Casing Drain Line DESCRIPTION: This PIM installed a rubber patch and hose clamps on pipe elbows in the RFP room and the condenser area pit. The elbows are on lines which connect to the feed pump turbine casing drains. Through wall condenser piping leaks have resulted in excessive air in-leakage causing oft Gas flow rate to increase through the Augmcnted Off Gas (AOG) processing system. The temporary patch was installed to reduce the flow rate through AOO and increase system efliciency. The patch was removed when permanent repairs were implemented during plant shutdown.

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EVALUATION: The leak repair patch will function to prevent in-leakage during expected plant operation; however, j failure of the patch will not cause an increase in leakage in excess of design limits. The patch and clamp l were selected to perform under expected operating pressure and temperature conditions. This PTM will j not increase the chances of a loss of condenser vacuum. Furthermore, the Main Steam Isolation Valves i are credited with closure upon a Main Steam Line break. The afTected piping is outside primary containment ar,d does not affect systems relied upon to process radioactive releases following a plant i event evaluated in the SAR. The piping elbow and associated piping have no safety design basis. I Presence of the PTM will not alTect operation of the turbine casing drains or the condenser. Dose {

consequences of steam line break accidents remain bounding and are unaffected. The subject piping does j not afTect equipment important to safety. A loss ofcondenscr vacuum is bounded by a turbine trip without l bypass. None of the low condenser vacuum trip instrumentation is affected by this patch. Therefore, no new types ofplant events are created. Failure of the patch will not create a new equipment malfunction.

The only results of patch failure are a steam leak or air in-leakage, both of which are already within existing analpes. The logic of the protective features described in the Technical Specifications Ibr steam line break mitigation or loss of condenser vacuum have not been altered by this PTM; therefore, the margin of safety remains unchanged. I PTM 98-06 TITLE: Installation of Monitoring Equipment to Determine Cause of Feedwater Flow Oscillations DESCRIPTION: This PTM installed temporary monitoring instrumentation in the LoveJoy Pneumatic Assembly Cabinct l in order to monitor air signals on Reactor Feed Pump Turbine (RFPT) "A" to determine the source of control system excursions. Two AirCet pressure transducers were installed to support troubleshooting of the feedwater pump speed oscillation. The data obtained confirmed the problem was control system initiated, rather than pneumatic. The PTM was subsequently removed.

SAFETY EVALUATION: The pressure transducers will be connected to redundant 1-P modules in order to satisfy a single failure.

Reactor feedwater flow oscillations will be minimized because there are two pneumatic control signals and they will be tested one at a time. Therefore, the probability of a previously evaluated plant event or equipment malfunction is not increased due to the installation of the monitoring equipment. The non-essential reactor feed system is not relied upon to perform any accident mitigation functions or support any system performing mitigation functions. No change in radiological efTects will occur as a result of this PTM. A malfunction of the RFPT system due to a failure of any one of the pressure transducers is bounded by the USAR analysis for the loss of feedwater event, and the feedwater controller failure maximum demand. Therefore, no new types of plant events are created. This PTM will not introduce a difTerent failure mode to the RFPT system than those already analyzed This PTM does not afTect the validity of any assumptions, calculations, procedures, or design specifications used to determine the plant's margin of safety.

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i PTM 98-08 TITLE: Installation of Set Screw on CW-MOV-103MV DESCRIPTION: This activity documented the installation of a set screw in CW-MOV-103MV that altered the original configuration of the motor operated valve. CW-MOV 103MV is the inlet to the B2 condenser waterbox.

The set screw is acting as a stake to prevent movement of the spline adapter relative to the valve shaft

'Ihe set screw was previouslyinstalled perMaintenance Work Request 98-1035 this PTM appropriately documented the installation. The set screw was subsequently removed when a new valve was installed in the system in REl8.

SAFETY EVALUATION: This activity will not create new failure modes for the valve. It will not increase the potential for loss of j' condenser vacuum. All other design features of the valve remain unchanged. The Circulating Water (CW) system is not relied upon for mitigating the consequences of a plant event. The CW valve is not important to safety and does not alTect equipment important to safety. All logic features of the CW system J

remain unchanged. The limiting cross sectional area of the spline adapter and valve shaft remains in the areas of the keyway. Installation of the set screw does not alter the overall function of the CW system.

Any failures of this valve will not cause the system to operate in a way that afTects equipment important to safety. The valve operator will still function as designed and no new electrical loads are being added to the operator motor. Loss of condenser vacuum via loss of CW remains the bounding event. Thus, no new types of plant events are created. Installation of the set screw will not increase the probability of valve failure and no new types of malfunctions are created. No logic or protective features for safety systems are affected. All existing analyses and margins of safety remain unchanged.

PTM 98-10 TITLE: Temporary Patch on Service Water (SW) Piping on the Outlet of Turbine Equipment Cooling (TEC) Ileat l Exchanger "B"  !

DESCRIPTION: This PTM authorized installation of a temporary, mechanical patch to stop or abate two pinnole leaks found in the 24" SW discharge pipe from the TEC IIcat Exchanger "B" The PTM was temporarily removed to allow ultrasonic testing and then re-installed. The PTM was permanently removed when a permanent repair was initiated per CED 1998-0121.

SAFETY EVALUATION: While the failure of this pipe could result in a sudden loss of feedwater, the probability of a low pressure, low temperature, seismic Class IIS pipe suffering a rupture in the absence of an earthquake is very small.

As the application of a patch will do nothing to either strengthen or weaken the pipe, it cannot increase the probability of a plant event. This portion of the SW system has been previously analyzed in the USAR as not performing any accident or event mitigating functions. This activity does not relocate any safety-related cquipment, nor does it afTect existing protective features in the SW system designed to mitigate the consequences of equipment malfunction. It does not alter the operating parameters of the SW system or any ofits users. Therefore,it is incapable ofincreasing the probability of a malfunction of safety-related equipment. This portion of the SW system is not required to backup any nuclear safety-  ;

related equipment, nor is it required to provide support of any such equipment. The operating parameters j l

for plant systems, structures, and equipment remain unchanged, as well as their operating characteristics s'xlinterfaces. Therefore, no new and unanalyzed types of plant events are created. The consequences of a pipe break or unisolable leak flooding the Turbine Building basement would not result in any new _

and unanalyzed failure modes. The potential sudden loss of feedwater, which could be the result of such l a break, has been fully considered in the USAR and found to be acceptable. Therefore, no new types of malfunctions ofequipment important to safety are created. A failure of a SW line in the Turbine Building basement will not result in a loss of cooling water to essential SW users, nor will it result in conditions beyond those previously analyzed. Therefore, this activity, will not result in any reduction of margin of safety as defined in the basis for the Technical Specifications. I i

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SPECIAL PROCEDURES (SPs) AND SPECIAL TEST PROCEDURES (STPs)

SP 95-046 TITLE: Z Sump On Line Repair Procedure DESCRIPTION: During surveillance testing of the Z Sump operability, it was determined that the essential high level switch RW-LS-Z5 did not send a signal to start the Z2 Sump Pump. This SP provided guidance for entering the Z sump to examine and repair level switch RW-LS-Z5, while maintaining secondary containment integrity. The level switch was found to be defective and replaced.

SAFETY EVALUATION: his SP has the potential to atTect Standby Gas Treatment (SGT) alone and no other essential systems are atrected. He SGT system is an accident mitigation system and, as such, cannot increase the probability of occurrence of an accident. This SP does not degrade the operability of the SGT system, but only temporarily moves the secondary containment boundary. The boundary, which is normally at the sump cover and manhole, will be maintained by closing OG-V-84 to eliminate the open path from the Elevated Release Point tower to the sump and by establishing a minimum water level in the sump to ensure that all drain lines entering the sump are water scaled. These actions ensure that the consequences of an accident or malfunction of equipment important to safety are not increased. In addition, the sump cover could be put back in place upon occurrence of a Loss of Coolant Accident to ensure the integrity of the boundary. This SP does not affect the condition or operation of any equipment important to safety with the exception of the level switch which is to be repaired or replaced. Since the level switch is currently degraded, an administrative Limiting Condition for Operation has been entered to address the lack of single failure protection for accident mitigation. If required to be replaced, the level switch will be replaced with an identical model, so tbc probability of a malfunction of this component will not be increased. Temporarily mming the secondary containment boundary will not physically alter the SGT system and so will not increase the probability of occurrence of a malfunction of equipment of the SGT system or any other system important to safety as previously evaluated in the USAR. The SP will be tenninated and the sump cover put back in place upon initiation of Emergency Procedure 5.1.3, Flood, to prevent the malfunction ofcomponcats within the sump due to flooding. The new temporary boundary is sufficient to maintain secondary containment and the margin of safety as dermed in the Technical Specifications is not reduced.

SP 95-087 TITLE: Installation of Jumpers in Steam Tunnel Leak Detection Temperature Switches DESCRIPTION: This SP was prepared to facilitate maintenance on the steam tunnel fan coil unit. It identified and documented the steps required to installjumpers in the steam tunnel leak detection temperature switches.

He high temperature switches werejumpered during periods of the maintenance to eliminate inadvertent ,

Group I isolation initiations due to temperature fluctuations in the steam tunnel. I SAFETY ,

EVALUATION: The Main Steam Line Leakage Detection (MSLLD) System is an accident mitigation system that cannot I initiate a design basis accident, IIaving the Main Steam tunnel portion of the system bypassed will not j increase the probability of an accident previously evaluated in the USAR. The installation ofjumpers will l be procedurally controlled. At most, the consequences ofimproper installation of a jumper would be 1 limited to a one-half group isolation which would not initiate any protective functions or initiate the l

operation of any plant components. However, if a full group isolation should be received, closure of the {

Main Steam isolation Valves (MSIVs) would occur This is an analyzed operational occurrence. The  !

MSLLD system incomorates an automatic isolation of the MSIVs from equipment room high ambient i temperature switches located in the steam tunnel and heater bay. During the period when the MSLLD {

system automatic isolation portion located in the Main Steam tunnel is bypassed, the automatic function I will not be available to detect and terminate any leaks that may occur in the Main Steam tunnel. The  !

USAR Chapter XIV design basis accident analysis assumes MSIV closure will be initiated by the high  !

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steam flow Group I isolation. For breaks too small to initiate the high steam flow isolation, operator action serves as the pnmary initiation of closure of the MSIVs, steam tunnel switches proside backup to manual action and are bounded by the analysis for manual action. This operator action is procedurally controlled per station emergency procedures. Main steam line breaks are required to be isolated to limit the doses received both at the site bounday and in the control room. The dose analyses for both areas determined the bounding event is a large break that can be mitigated by the high steam flow isolation.

Therefore, the consequences of an accident presiously evaluated in the USAR are not increased.

Installation ofjumpers to bypass the MSLLD system portion in the Main Steam tunnel will reduce the challenges to the MSIV, reactor pressure vessel, vessel over-pressure protection systems, and inventory control safety systems by eliminating the potential for spurious Group 1 isolations. Jumper installation will not alTect the ability to manually close the MSIVs nor will it affect the high steam flow isolation as the logics are diverse. The fan motor replacement will not interfere with other equipment located in the Main Steam tunnel. The cooling fan units themselves perform no safety function. The loss of the high temperature actuation function of automatic closure of the MSIVs for a brief period of time was evaluated during initial approval of the Cooper Nuclear Station Final Safety Analysis Report. The allowable out of senice time is controlled by a Technical Specification Limiting Condition for Operation (LCO). The LCO requirements will be met during the performance of this SP. Therefore, the probability of a previously evaluated malfunction of equipment important to safety is not increased. Compensatog actions to manually close the MSIVs within five minutes will ensure the consequences of any accident remain within the USAR accident analysis. Compensatory action of a station operator stationed to monitor the steam tunnel area and heightened control room operatos attention to steam leak indicators will be instituted for the duration of this SP. The USAR Chapter XIV accident analysis states the second method of steam leakage detection relies on temperature sensors in the Main Steam tunnel. The accident analysis states that " response time of a sensor to a nearby steam line break would depend upon the size and location of the break in relationship to the nearest sensor." No MSIV closure time as the result of a temperature switch trip is assumed in the accident analysis. It is assumed that any trip caused by a high temperature switch in the steam tunnel in which a high flow trip is not present would occur prior to 30 minutes and is therefore bounded by the consequences of a manual closure of the MSIVs Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the USAR are not increased. Replacement of the fan motor will use existing pathways, platforms, and lifting points.

No new components will be introduced as a result of this activity. The existing work areas do not contain equipment that may be damaged or interfered with by the maintenance activity. Installation of the jmnpers involves manipulation of the MSLLD system logic. This logic is deenergize to trip. The worst case occurrence would be an inadvertent Group I isolation. MSIV closure at power is an analyzed transient.

The possibility of this occurrence will be minimized by verifying proper jumper installation in one channel before moving on to another. Repair of the fan motor does not reduce the margin of safety as defined in the basis for any Technical Specification. The cooling units themselves are non-safety related and are not credited in any transient or design basis event. Should a spurious trip occur, the closure of the MSIVs while at power is an analyzed operational occurrence.

STP 95-092 TITLE: IIigh Pressure Coolant injection (1IPCI) Exhaust Leg Drain Evaluation DESCRIPTION: Operation of the IIPCI system indicated that a quantity of water greater than expected may have been accumulating in the IIPCI turbine exhaust line. This STP provided for manually draining and measuring the content of the 1IPCI cxhaust line dripleg. The purpose of this test was to: 1) determine the source,

2) provide an estimate of the quantity of leakage into the IIPCI steam exhaust dripleg following surveillance testing, and 3) obtain a recording of the exhaust line pressure transient. The test was performed multiple times to evaluate the source and quantity of water present in the exhaust line until appropriate modifications were implemented during RE16.

SAFETY EVALUATION: The pnmay concern for this procedure is the breach of the Primary Containment boundary. Ilowever, Technical Specification 3.7.D and Procedure 2.0.2 allow for manipulation of manual containment boundary isolation valves within the specified administrative controls. The open containment isolation l

J valves will be monitonxl continuously and closed if plant conditions merit. Operation of the IIPCI system will be unaffected. IIPCI will be placed in PULL-TO-LOCK (auto-start disabled) if the system has been declared inoperable due to water hammer concerns. Administrative controls will mitigate the potential l for increased consequences beyond those already analped in the USAR. The resultant impact of draining the IIPCI exhaust line dripleg will have a negligible effect on equipment operation. The HPCI turbine will only be operated with the exhaust lines drained of all water to preclude water hammer events.

Draining the dripleg remotely is a routine operation. Manually draining the dripleg will have a similar i effect on the system and not impair the 1IPCI pump in any way. Primary Containment will be maintained admmistratively as per Procedure 2.0.2 and this STP. Any potential accident resulting from this activity ,

is bounded by the USAR. The Bases of the Technical Specifications allow manipulation of Primary j Contamment valves when administratively controlled. Operation ofIIPCI will be performed within the (

bounds of the Technical Specifications; therefore, there is no reduction in the margin of safety.

SP 96-093 TITLE: Reactor Vessel High Water Level Main and Reactor Feed Pump (RFP) Turbine Trip Relay Functional j' Test DESCRIPTION: The purpose of this SP was to perform contact testing of the relays associated with the Reactor IIigh Water Level Trip function (6A-KI A, B, and C) which were not previously tested. Previous testing demonstrated operability of the three trip circuits for the level transmitters to the trip relay coils. This SP tested the actual trip relay contacts in each of the trip circuits. Performance of this SP demonstrated that the subject trip relays were functioning as intended. The testing performed by this SP was subsequently incorporated into surveillance procedures.

SAFETY EVALUATION: None of the equipment involved in this SP is an accident initiator or credited with mitigating the consequences of an accident. The reactor coolant boundary is not affected, nor is any external pressure applied to the instrument lines associated with the instruments invo:ved in this actisity. Performance of this SP does not affect equipment in a way that would degrade the capability of systems required to j mitigate the consequences of an accident to perform their intended safety function. System performance will remain unchanged by this SP and no permanent modification to any plant equipment will result that could introduce a malfunction. Performance of this SP requires insertion of one vessel high level trip signal into the two-out-of-three high water level trip logic for both RFPs and the main turbine. The consequences of a malfunction of one of the two remaining high level switches in sersice are bounded by existing transient analysis, and are not increased. This activity requires operation with the feedwater control system selected to monitor "A" side level instrumentation. A failure of the comnmn sensing to the "A" side feedwater control level instruments during this time could result in a feedwater controller failure (maximum demand) transient. Moreover, two of the three level instruments required to initiate  ;

the main and RFP turbine high water level trips are not functional (due to the sensing line break) to i terminate the transient. Ilowever, the passive failure of this instrument sensing line is not considered credible as this SP performs no activities that would challenge the integrity of that sensing line.

Additionally, the possibility of a reasonable operator error that could result in a loss of the common sensing line is not created by performance of this SP. As a conservative measure, the SP directs Operations personnel to initiate a trip of the main and reactor feed pumps if vessel level approaches the ,

high level trip setpomt. Therefore, the ability to terminate the potential transient is maintained. No new l potential failure modes are introduced by the performance of this SP which could credibly create a transient, accident, or malfunction of equipment important to safety that has not already been evaluated. j Throughout this activity a sufficient number of main and RFP turbine high water level trip channels l remam functional to ensure high water level trip capability always exists; therefore, the margin of safety i is not reduced.

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SP 97-009 TITLE: Radiologically Contmiled Area (RCA) Teledosimetry Testing l DESCRIPTION: This SP documented the preinstallation testing of Science Applications International Corporation l Teledosimetry Units and associated signal repeater units in various locations in the RCA in order to I demonstrate that their operation will have no effect on the operation of plant equipment. This SP I

collected evidence that the teledosimetry equipment operating above 900MHZ at 500 mW did not have any effect on the operation ofequipment in the RCA.

SAFETY EVALUATION: This SP will be performed during reactor shutdown during which the only credible events involve non-transient scenaries. Potentially atrected equipment will be carefully monitored during the testing for any potential effects invohing the equipment's ability to perfonn its safety related function. Any etTects noted on that equipment will result in immediate cessation of the testing and the return of the equipment's normal functioning. The only credible etTect of this test on plant equipment is a transient effect that will j l stop when the test is stopped. This SP cannot initiate the permanent degradation of plant equipment

( function. All required plant equipment will be available to respond as required. Since the reactor will be shutdown during the performance of this testing, this climinates the requirements for the most urgent response of safety related equipment. No plant event initiators can be credibly affected by this testing.

)

This activity does not change any existing plant systems, structures, or components. It merely generates a low power localized signal which, at worst case, may induce electrical noise in nearby conductors or components resulting in a spurious trip or actuation. The response to, and the consequences of, these actions remains unchanged by this activity and are as previously analyzed. No new malfunctions are introduced. In fact,it is the intent of this SP to show that no new malfunction of equipment important to safety is introduced by the installation of the subject teledosimetry system. This SP does not involve any parameter included in the basis for any Technical Specification.

SP 97-012 TITLE: Post MP 97-068A Augmented Off Gas (AOG) System Operation with Recombiner A DESCRIPTION: This SP was issued to provide guidance for startup and operation of the AOG System A Train following installation of the Phase II Z-Sump Modification, MP 97-068A. It was developed to prmide the guidance necessary to return AOG Train A to sersice in a controlled manner while ensuring Standby Gas Treatment (SGT) operability was maintained. Specific objectives of this SP were: 1) Ensure monitoring and control of difTerential pressure between the Air Removal IIold-up Line and Z-Sump during startup of AOG Train A,2) test drainage of the 48" 1Iold-up Line into the Z-Sump,3) gather data to establish parameters that will be used to identify station procedures and required changes to be made eficctive for long-term AOG Train A operation, and 4) control of startup and operation of AOG Tnin A until applicable station procedures were revised, approved, and made effective. Results of the SP were as follows: 1) the maximum reading for the IIold-up Line to Z-Sump differential pressure v as far below the maximum allowable difTerential pressure,2) operability of SGT was not jeopardized,3) testing confinned the absence of air binding in the drain line from the Off Gas Hold-up Line into the Z-Sump and free drainage flow, and 4) operating procedures were revised for long-term AOG syste.n operation.

l EVALUATION: The SGT and AOG systems are not plant event initiators, therefore, additional acceptance testing and changing AOG system operating pressure per this SP does not increase the probabilky of occurrence of a plant event previously evaluated in the SAR. Monitoring and testing per this SP ensures SGT operability is maintained and the dose consequences associated with a previously evaluated plant event i remain unchanged. Dose consequences for the Control Rod Drop Accident have been evaluated to be l l within acceptable limits for AOG out-of-scr ice. Therefore, slight adjustments to AOG system I backpressure will not impact preve usly evaluated dose consequences. SGT operability will be i maintained by ensuring SGT lines are not challenged by flooding from Z-sump. In addition, slight AOG ,

backpressure changes will not afTect off gas isolation equipment, off gas efiluent radiation monitoring '

equipment, hydrogen control via AOG, off gas dilution, or elevated release. The possibility of an otT gas  ;

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hydrogen bum event is not created. This activity does not introduce any new equipment or direct AOG operation outside of system design; therefore, no ncw failure mechanisms are introduced The margin of safety as defined in the basis for the Technical Specifications remains unchanged.

SP 97-014 TITLE: Spent Fuel Pool Irradiated Hardware Shipping DESCRIPTION: This SP was developed to provide guidance for station and contract personnel to load irradiated test capsules from the Cooper Nuclear Station spent fuel pool into a shielded cask and ship them to General Electric.

SAFETY EVALUATION: This SP does not involve a change to any systems, structures, or components (SSCs) or involve unusual system operations that could initiate a previously evaluated plant event. Cask handling, as well as cask I drop events, are described in the USAR and bound the activities described in this SP. This SP does not affect the function of accident mitigation systems / components, nor does it prevent personnel from perfonning functions which may mitigate the consequences of an accident. The SP operates equipment within its analyzed design limits and per its intended function, utilizing approved operating precedures.

It does not modify plant SSCs or involve operation of the SSCs in a manner not previously evaluated.

Therefore, there is no increase in the probability of occurrence or consequences of a malfunction of equipment important to safety. No different types of plant events are created and no unusual system operations are introduced which could cause a different type ofequipment malfunction than previously i evaluated. Activities perfbrmed under this SP are conducted in accordance with Technical Specification conditions for operation and surveillance requirements. Technical Specification margin of safety is unaffected by performance of this SP.

SP 97-016 and Rev.1 TITLE: Sutorbilt Blower Functional Test DESCRIPTION: This SP was developed to determine whether Sutorbilt Compressor I A is functional. The Sutorbilt i system was previously allowed to be abandoned in place due to a lack of any requirements to maintain the system. The SP configured the Sutorbilt Compressor I A to take suction from the Reactor Building atmosphere and discharge to the Reactor Building atmosphere. Starting and running amperages were deternuned and recorded and flow rate was measured. The test also verified the Sutorbilt interlocks and trips. Two leaks were discovered during the test which were subsequently repaired. Based on the results of this SP, the system was made available for use during the performance of Surveillance Procedure 6.PC.503, Drywell to Suppression Chamber Leakage Test.

SAFETY EVALUATION: This activity will not change the state or function of safety related structures, systems, or components as l

described in the USAR and will not alter any of the inputs or assumptions for the probabilities of {

previously evaluated accidents. The consequences of previously identified accidents remain bounded by j the results contained in the USAR. The load from pump FPC-P-1B will be shed from Motor Control l

Center S before the load from Sutorbilt compressor I A is added. The load of Sutorbilt Compressor 1 A is bounded by.the load of FPC-P-1B. Primary containment is unalTected by the functional testing.

Therefore, the probability of occurrence and the consequences of a malfunction of equipment important  ;

to safety are not increased. The functional test procedure does not introduce new esent initiators or failure modes. No changes have been made to parameters that are bounded by the containment Technical l Specifications or Administrative Limits. Therefore, there is no reduction in the margin of safety. l l

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SP 97-017 TITLE: Augmented Off Gas (AOG) System Startup and Operation with Recombiner A for Significant Condition Adverse to Quality (SCAQ) 97-1391 DESCRIPTION: SCAQ 97-1391 documented the indication of a hydrogen burn, or potential flow blockage, in the OIT Gas system. This SP was used to restart AOG Train A to determine the functionality of the unit and to assist in establishing the cause of the event. The SP was modeled aller Station Operating Procedure (SOP) 2.2.58.3, but expanded to establish additional parameters, monitoring requirements, and contingencies ,

beyond the SOP to assist in controlling stanup of the unit, identifying the cause of the potential blockage, l and collecting suflicient data to ensure continued long-term operation. Based upon the results of this SP, the most probable cause of the AOG failure was due to a failure of CF-SOV-SPV2038A, resulting in a ]

loss of condensate to the post-recombiner condenser. Recommendations to prevent recurrence were subsequently identified.

SAFETY EVALUATION: Contingencies are in place to limit the consequences of a pressure transient during performance of this l SP such that a flow blockage will not result in damage to systems, structures, or components (SSCs) j assumed to function during an event previously evaluated in the SAR, or assumed to function during malfunctions ofequipment previously evaluated in the SAR. AOG is not an initiator or mitigator of any plant accident scenario. No new activities are being performed and no new methods of AOG operation are being introduced. Extra guidance has been added to abort an AOG startup and collect data to determine the cause if such an action is necessary. Procedural direction has been given to ensure steam / condensate is not admitted to unwanted portions of the system. Suction pressure will be monitored and steam sources isolated to prevent steam / condensate intrusion. No new or additional equipment lineups are specified. The logic, operation, and radioactive release control as provided in the Technica! l Specifications remains unalonxi as a result of this SP. The existing margins of safety remain unchanged.

SP 97-018 l l

TITLE: Condensate Filter Demineralizer Pleated Septa Testing and System Operation l DESCIUPTION: His SP authorized the installation ofMEMTEC pleated filter septa in four of the seven Condensate Filter Demineralizer vessels in place of Graver wound septa. It was generated because pleated filter septa may l allow significant reduction in resin usage and the associated disposal costs. The SP also provided l instructions for Operations personnel to operate the Condensate Filter Demineralizer system during operational testing of the altemate filter elements. SP 97-018 superseded Operations Procedure 2.2.5 l for the duration of the testing. Following successful testing of the pleated filters, CED 1998-0028 was l generated to document the pennanent installation of the filters and Procedure 2.2.5 was revised to reflect the new testing methodology.

SAFETY EVALUATION: Operational characteristics of the Condensate Filter Demineralizer system will not be alTected.

Operational parameters such as flow and differential pressure will be unchanged. No changes to the interlocks and setpoints that control the system are required. Therefore, the probability of an event previously evaluated in the SAR is not increased. Mechanical filtration will be enhanced, allowing removal of a larger portion of paniculate in the feedwater stream. The septa being tested are rated to withstand system operational conditions including differential pressure, temperature, and radiation levels, and function in the same manner as the current septa. No systems or components needed to mitigate or contml the consequences of any accident analyzed in the USAR are afrected. This SP does not involve any equipment or systems that are important to safety. Current methods of operation will be tmchanged, remairung in the manual made ofcontrol. The Condensate Filter Demineralizer system is not relied upon to mitigate the consequences of an accident described in the USAR and this SP will have no impact on those systems needed to mitigate or control the consequences of a malfunction of equipment important to safety. Should a failure occur, the results would be bounded by the Loss of Feedwater and Feedwater Controller - Maximum Demand events already analyzed. No new events are introduced. The pleated septa will reduce ir. soluble iron in the filter efliuent, allowing operation in the optimum water chemistry l

l band recommended by General Electric to minimize shutdown drywell dose rates. This SP does not introduce any failure mode of a afferent type that would cause a Condensate Filter Demineralizer system l failure that is not bounded by existing USAR analyses. The validity of assumptions, calculations, procedures, or design specifications used to determine margin of safety are not alTected.

SP 97-019 TITLE: Residual IIcat Removal (RIIR) lleat Exchanger Performance Data Collection DESCRIPTION: This SP was performed to gather additional heat exchanger performance data concurrently with the perfmnance of Procedure 13.17, RIIR IIcat Exchanger Performance Evaluation. This data was collected using non-intrusive temporary test instruments to record RIIR and Service Water (SW) temperatures and flows. The test was performed to proside information to support possible procedure and instrumentation changes. The use oftemporary ultrasonic flow instrumentation was determined to be desirable on IU IR and undesirable on SW. The temperature instrumentation installed by this SP produced mixed results.

SAFETY EVALUATION: This SP will not alTect the probability of occurrence of any previously evaluated plant event. The non-intrusive test instruments and temporarily modified insulation are not initiators for any plant events. The RIIR and SW systems' capability to mitigate any plant event that could occur is not affected by this SP.

The instrumentation is non-intrusive into either the RI1R or SW system. The attached instrumentatico has no negative effect on the RIIR and SW system boundaries and the temporary changes in insulation configuration have been evaluated. The instrumentation does not introduce any change in the probability l of a malfunction of the IUIR and SW systems It cannot initiate a failure of plant equipment or a plant event or transient. The data gathering lineup is controlled by approved plant procedures. Since the non-intrusive instrumentation will act affect the operation or failure modes of plant equipment, there is no change to the defined margin of safety.

SP 98-002 TITLE: Determination of Fuel Pool IIcatup Rate DESCRIPTION: This SP was developed to deternune how long the Fuel Pool Cooling system can be shut down before an operating temperature of 120*F is reached. This information was used to determine how long the Fuel Pool Cooling system can be removed from senice to perform maintenance on the system. Three thermocouples were installed at various elevations in the spent fuel pool to monitor the heatup rate with the Fuel Pool Cooling pumps temporarily shutdown. This SP subsequently formed the basis for permanent Procedure 13.20 to be used for future determination of fuel pool heatup rate.

SAFETY EVALUATION: Fuel pool design temperatures will not be exceeded during this SP. The assumptions for plant events or event mitigation previously evaluated in the SAR are not affected. This is a temporary non-intrusive test.

Loss of Fuel Pool Cooling has been previously evaluated in the USAR. Plant systems are designed for temporary isolation of Spent Fuel Pool Cooling during various operational evolutions. Equipment and systems are operated within their design basis. Therefore, the probability of occurrence or consequences of a malfunction of equipment important to safety are not increased. As plant systems will be operated per current plant procedures and within design, no new types of accidents or equipment malfunctions are introduced. A Spent Fuel Storage Pool temperature of 125*F was t. sed in the Standby Gas Treatment heater sizing calculation. This SP will not allow the Spent Fuel Pool temperature to increase above 125"F. If the temperature limit of 120 F is exceeded, Fuel Pool Cooling will be returned to senice.

Therefore, there is no reduction in the margin of safety.

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p band recommended by General Electric to minimize shutdown drywell dose rates. This SP does not intnxluce any failure mode of a ddferent type that would cause a Condensate Filter Demineralizer system failure that is not bounded by existing USAR analyses. The validity of assumptions, calculations, procedures, or design specifications used to determine margin of safety are not affected.

SP 97-019 1

I TITLE: Residual IIcat Removal (RIIR) IIeat Exchanger Performance Data Collection DESCRIPTION: This SP was performed to gather additional heat exchanger performance data concurrently with the performance of Procedure 13.17, RIIR IIcat Exchanger Performance Evaluation. This data was collected using non-intrusiw temporary test instruments to record RIIR and Senice Water (SW) temperatures and flows. The test was perfonned to provide information to support possible procedure and instrumentation changes. He use of temporary ultrasonic flow instrumentation was determined to be desirable on RIIR and undesirable on SW. The temperature instrumentation installed by this SP produced mixed results.

SAFETY EVALUATION: his SP will not atTect the probability of occurrence of any previously evaluated plant event. The non-intrusive test instnanents and temporarily modified insulation are not initiators for any plant events. The RIIR and SW systems' capability to mitigate any plant event that could occur is not aiTected by this SP.

The instrumentation is non-intrusive into either the RIIR or SW system. The attached instrumentation i

has no negative efTect on the RIIR and SW system boundaries and the temporary changes in insulation

! configuration have been evaluated. The instrumentation does not introduce any change in the probability of a malfunction of the RIIR and SW systems. It cannot initiate a failure of plant equipment or a plant event or transient. He data gathenng lineup is controlled by approved plant procedures. Since the non-intrusive instrumentation will not affect the operation or failure modes of plant equipment, there is no change to the dermed margin of safety.

SP 98-002 TITLE: Determination of Fuel Pool IIcatup Rate DESCRIPTION: This SP was developed to detemune how long the Fuel Pool Cooling system can be shut down before an operating temperature of 120*F is reached. This information was used to determine how long the Fuel l

Pool Cooling system can be removed from senice to perform maintenance on the system. Three thermocouples were installed at various elevations in the spent fuel pool to monitor the heatup rate with the Fuel Pool Cooling pumps temporarily shutdown This SP subsequently formed the basis for permanent Procedure 13.20 to be used for future determination of fuel pool heatup rate.

SAFETY EVALUATION: Fuel pool design temperatures will not be exceeded during this SP. The assumptions for plant events or event mitigation previously evaluated in the SAR are not affected. This is a temporary non-intrusive test.

less of Fuel Pool Cooling has been previously evaluated in the USAR. Plant systems are designed for temporary isolation of Spent Fuel Pool Cooling during various operational evolutions. Equipment and systems are operated within their design basis. Therefore, the probability of occurrence or consequences of a malfunction of equipment important to safety are not increased As plant systems will be operated per current plant procedures and within design, no new types of accidents or equipment malfunctions are introduced. A Spent Fuel Storage Pool temperature of 125 *F was used in the Standby Gas Treatment heater sizing calculation. This SP will not allow the Spent Fuel Pool temperature to increase above 125*F. If the temperature limit of 120*F is exceeded, Fuel Pool Cooling will be returned to senice.

Therefore, there is no reduction in the margin of safety.

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SP 98-005 (USQE 1998-0043)

TITLE: Testing of 32/1FE Relay DESCRIPTION: This SP was performed to verify the overpower relay will perform the design function of tripping (opening) the 4160 VAC Breaker IFA. It provided instructions to test and verify the proper wiring configuration for Relay 32/lFE after correction of a wiring error. The SP was perfonned during the refueling outage during the time when bus IF was considered as part of the non-protected division. The SP detemuned that the overpower relay will perform the design function of separating the diesel from the 1 A bus if an overpower condition is detected.

SAFETY EVALUATION: This activity will result in separation of the non-essential and essential buses for a given division; however, all sources ofpower will remam capable ofperforming their design function. The safety related  ;

bus will maintain two sources of power available during the perfonnance of the SP (the Emergency l

Transformer and its corresponding Diesel Generator No.1). All AC electrical power distribution systems will remain capable of performing their design function and divisional separation will ensure the opposite division is not affected. Performance of this SP will not afTect any of the factors that contribute to the l probability of an accident since the 1 F bus, undervoltage protection, load shedding, and load sequencing I circuitry that contribute to mitigate an accident are not accident initiators. All loads of the safety related i bus will remain capable of performing their design function and available during performance of this SP.

There is no reduction in the protection of public health and safety since the function of equipment designed to control the release of radioactivity is not affected and no new pathways for the release of .

radioactive materials are created. This SP will not alter the system operation. This is assured because l

the secondary of 6e B phase current transformer will be shorted and the 1FA breaker will be racked to i the test position making it incapable of interfering with either the operation of the l A or IF bus. ]

Equipment important to safety will remain operational and will not be challenged by performance of this i SP. No equipment will be modified in such a manner that would result in a loss of redundancy or independence to cause a malfunction. Shorting of the B phase current transformer and titling of the i 32/1FE relayleads prevents the faihire of the test equipment used within the SP from affecting the safety  ;

related function of the Diesel Generator No.1. Installation, operation, or failure of the test equipment will  !

not alTect any of the circuitiy necessary for operation of the IF bus. System operation and function will not change as a result of this SP or testing of the 32/lFE relay, nor will there be any other systems i affected, therefore, the mr.rgin of safety is not changed. I SP 98-006 (USQE 98-0045) l TITLE: Testing of 32/1GE Relay I

DESCRIPTION: This SP was performed to verify the overpower relay will perform the design function of tripping (opening) the 4160 VAC Breaker IGB. It provided instructions to test and verify the proper wiring configuration for Relay 32/lGE after correction of a wiring error. The SP was performed during the refueling outage during the time when bus 10 was considered as part of the non-protected division. The SP detemuned that the overpowerrelay will perform the design function of separating the diesel from the 1B bus if an overpower condition is detected.

SAFETY EVALUATION: This activity will result in separation of the non-essential and essential buses for a given disision; l

however, all sources of power will remam capable of performing their design function. The safety related bus will maintain two sources of power available during the performance of the SP (the Emergency Transformer and its corresponding Diesel Generator #2). All AC electrical power distribution systems will remain capable of performing their design function and divisional separation will ensure the opposite division is not affected. Performance of this SP will not affect any of the factors that contribute to the probability of an accident since the 1 G bus, undervoltage protection, load shedding, and load sequencing circuitry that contribute to mitigate an accident are not accident initiators. All loads of the safety related l

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a bus will remain capable ofperfonning their design function and available during performance of this SP. 1 There is no reduction in the protection of public health and safety since the function of equipment designed to control the release of radioactivity is not alrected and no new pathways for the release of radioactive materials are created. This SP will not alter the system operation. This is assured because the secondary of the B phase current transformer will be shorted and the 1GB breaker will be racked to the test position making it incapable of interfering with either the operation of the IB or 10 bus.

Equipment important to safety will remain operational and will not be challenged by performance of this SP No equipment will be modified in such a manner that would result in a loss of redundancy or independence to cause a malfunction. Shorting of the B phase current transformer and lifting of the 32/lGE relay leads prevents the failure of the test equipment used within the SP from alTecting the safety related function of the Diesel Generator #2. Installation, operation, or failure of the test equipment will not affect any of the circuitry necessary for operation of the 1G bus. System operation and function will not change as a result of this SP or testing of the 32/lGE relay, nor will there be any other systems affected, therefore, the margin of safety is not changed.

SP 98-007/CED 1998-0025 (USQE 1998-0006)

TITLE: Circulating Water (CW)/ Fire Protection (FP) Crosstie to Backflush CW Sparger DESCRIPTION: This SP/CED was a temporary configuration change to provide direction for crosstic of FP or CW sparger water from the intake structure to the CW butterfly sparging system in the condenser area. This was necessary to provide a higher pressure / flow flush to CW-MOV 110MV, the B-1 Condenser Backflush Valve. This valve is required to be opened when a condenser backwash is performed. The valve was suspected to be silted shut and the Service Water system pressure supplied to the butterfly valve sparging system did not provide adequate pressure or flow to dislodge the material deposited on the valve. This SP/CED was successful in de-silting the valve. j SAFETY EVALUATION: The CW and FP systems are not initiators of any accidents evaluated in the SAR. The conditions that -

must be met to accomplish the FP system functional requirements are not affected by this activity. The proposed changes do not change any of the initial conditions or assumptions used to evaluate accidents described in the SAR The CW and FP systems are not relied upon to mitigate the consequences of any accidents evaluated in the SAR. The ability of the FP system to mitigate the special event of a fire is unalTected. The CW system is not important to safety. Loss of the sparger pumps or system will not adversely atrect Service Water. Neither the design basis for the FP system, nor the initiating sequences or starting setpoints of safety related equipment, are afTected by this activity. This change does not induce failure of any equipment important to safety nor affect the consequences of equipment failure. This activity assures reliability of the FP system to function when needed through hardware limitations and continuous monitoring. The hose used to route water from the sparger system to the butterfly valve sparger system has an adequate design pressure to ensure pressure boundary integrity. Component failures and malfunctions which are analymibound any potential failures of this crosstic. No new failure modes are being introduced. Therefore, no new types of accidents or equipment malfunctions are introduced. This activity (krs not affect the acceptance limits for any system important to safety nor will l it affect the range allowed for operation under analyzed operational transients or the range required for the margin of safety to prevent approaching system limitations or design failure points. Therefore, no margins of safety as defined in the basis for any Technical Specification are reduced.

SP 98-029

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(USQE 1998-0081) i

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TITLE: Secondary Containment DifTerential Pressure Measurements DESCRIPTION: This SP provided instructions for obtaining absolute pressure measurements at various locations within the Rcactor Building, outside at grade level, and on the Reactor Building roof. The test was non-intrusive as measurements were taken using a portable pressure gage which was not connected to plant systems

or equipment. The data w?s gathered to measure the stack efTect on the Reactor Building. Currently, Procedure 6.SC.501, Secondary Containment Leak Rate Test, requires the addition of a stack effect conection factor to ensure there is -0.25" WG at the Reactor Building roof. The test data obtained by this SP will be evaluated under EE 1998-0124.

SAFETY EVALUATION: 'Ihe unit will be in Mode 4 or 5 when this test is performed and core alteration will be suspended while the roof hatch is opened. The test is non-intrusive and no components will be installed in the plant.

Therefore, the probability of occurrence or consequences of a previously evaluated accident are not increased. The user will be in attendance with the test equipment to prevent it from interacting with essential plant equipment. If the test equipment is left alone, the user will ensure it is secured and there will be no scismic interactions. Current plant procedures will be followed to access the Reactor Building roof and to run Standby Gas Treatment. Therefore, no new types of accidents of equipment malfunctions are introduced. When the Reactor Building hatch is opened. Procedure 6.SC.502 directs Operations to declare Secondary Containment inoperable. Limiting Condition for Operation 3.6.4. I will not be met and core alterations will have to be immediately suspended. Since Technical Specifications are being fbilowed, the margin ofsafety as defined in the basis for any Technical Specification will not be reduced.

SP 98-030 (USQE 1998-0091)

TITLE: Reactor Equipment Cooling (REC) Flow to the Quad Fan Coil Units DESCRIPTION: This SP was developed to balance the REC system flow rates to the four quads following implementation of a REC piping modification per CED 1998-0284. Flow balancing was necessary to ensure that individual component , supplied with cooling water from the REC system have adequate flow to meet their design requirements. The procedure also provided instructions for verifying that adequate flow is being delivered to the critical REC loop under conditions similar to those that would occur under a Loss of Coolant Accident without a Loss of OfTsite Power scenario.

SAFETY EVALUATION: The use of this SP does not alTect any accident precursors and does not introduce any operating modes not previously evaluated. It does not affect the ability of any system to perform its required function to mitigate the consequences of an accident. By establishing required REC flow rates to the quads, adequate cooling ofessential components is assured. This prowdure will not result in the challenge of any physical barrier and, therefore, will not result in an increase in radiological dose to the general public or plant personnel. This SP does not change the normal operating modes or operating parameters of any plant equipment, and (kies not change the operating environment of any equipment or components. Use of this procedure will not affect the capability of any equipment to perform its design function and no new plant equipment is added. Therefore, the probability of occurrence or consequences of a malfunction of equipment important to safety are not increased and no new types of accidents or malfunctions are created. The procedure does not affect the operability of the REC system and, therefore, does not alTect the operability of the systems supported by the REC system. Thus, the margin of safety as defined in the basis for any Technical Specification is not reduced.

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USAR CHANGE REQUESTS (UCRs), LICENSE CHANGE REQUESTS (LCRs),

AND DRAWING CHANGE NOTICES (DCNs)

NOTE: The USAR changes covered by LCRs 92-0040 through 95-0049 below were made prior to this reporting period, however, they were made without suppo ting i OCFR50.59 evaluatinns. A Problem Identification Report was generated and the 10CFR50.59 evaluations were subsequently completed during this reporting period.

LCR 92-0040 e . TITLE: USAR Change Service Water (SW) Supply DESCRIPTION: This LCR removed the requirement for SW-V 122, the critical SW header to the Reactor Equipment Cooling (REC) heat exchangers crosstie, to be kx:ked open.

SAFETY EVALUATION: 1he original design basis intent was to have the REC heat exchanger crosstie valve locked open to have the capability to operate both REC trains from either SW header. There is no postulated event within the Licensing / Design Bases of CNS which requires both REC trains to be operable to safely shut down. The plant has been analyzed for the most limiting accident case and determined to be capable of safe shutdown /cooldown, with one electrical division, one SW pump, one REC train, and one division of Emergency Core Cooling Systems available. Therefore, this change does not increase the probability of a plant event previously evaluated in the SAR. This activity does not impact any components or syste ns used to monitor, control, or report the radiation exposure of plant personnel or the general pr%c.

Resising the requirement for SW-V-122 to be locked open does not impact any equipment considered to be important to safety or safety related, and will not increase the consequences of a malfunction of equipment important to safety. It does not change the function of any system, structure, or component important to safety previously evaluated in the SAR. No new failure modes are introduced and no additional demands are placed on operations personnel. The activity associated with this LCR will not contribute to any accident initiators. This activity does not impact any Technical Specification or the basis for any Technical Specification. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

LCR 93-0052 TITLE: USAR Change - Reactor Equipment Cooling (REC) Activity DESCRIPTION: This LCR deleted a statement that said the REC system normally contains activity due to activation of added corrosion inhibitors. Chemicals have not been added to REC for several years; REC uses demineralized water. This LCR also made various adminktrative changes to incorporate standard abbreviations.

SAFETY EVALUATION: 1he integrity of the REC system is not affected by this change. The use of corrosion inhibitors in the REL system is not necessary. Corrosion in the REC system is limited by using demineralized water as make-up and the integrity of the system is assured by the Inservice Inspection program. The capability of the REC system to perform its intended function is not jeopardized by eliminating the use of corrosion inhibitors. This change will not increase the probability of occurrence or consequences of a presiously evaluated plant event. Elimination of the statement regarding the use of corrosion inhibitors will not contribute to a malfunction of any equipment in the REC system or increase the consequences of an equipment malfunction. There are no changes to the plant design bases and no new leak paths for radioactive materials have been introduced. No new equipment is added to the plant and the operation of any existing equipment is not changed. Eliminating the statement m te use of corrosion inhibitors will not impact the discussion of REC in the Technical Specifications and there is no reduction in the margin of safety as defined in the basis fbr any Technical Specification.

LCR 93-00(37 l TITLE: USAR Change - 12.5 KV System DESCRIPTION: 'this LCR revised the description of buildings fed from the north and south 480V outdoor switchboards to reflect the actual loads installed per the "as-built" design. No new electrical loads were added or deleted from these switchboards. It also corrected the KVA rating of the west warehouse transfwmer from 3000 KVA to 300 KVA. This reflects the actual rating of the transformer installed per the original 1

design. The four sheets of Burns and Roe Drawing 3009 presiously included in the USAR to reflect the 12.5 KV system were replaced with a single sheet of Drawing 3009 which is a 12.5 KV system ovmiew.

SAFETY EVALUATION: These revisions do not change the Auxiliary Electrical System function, operation, and performance or affect any system in any way that could increase the probability of an accident. Systems and equipment will not be affected by this change such that they are degraded or operated outside their design limits.

These changes do not impact any radiation monitoring equipment nor affect any parameters associated with radiation monitoring. The function of equipment designed to control the release of radiation is not affected by this activity. No physical changes to the plant are required. The changes do not decrease the reliability of any system or equipment important to safety assumed to function in the accident analysis, nor do they alter the assumptions made in evaluating the consequences of a malfunction of equipment important to safety. As this change does not afTect the ability of the Auxiliary Electrical System to perform essential auxiliary functions for each of the operating modes and plant events for which they are required, there is no increase in the probability of a different type of plant event than presiously evaluated.

No new type of malfunction is introduced to the Auxiliary Electrical System as a result of this change.

There is no change to the accuracy or response characteristics of the components of the Auxiliary Electrical System such that the margin of safety as outlined in the basis of any Technical Specification is affected.

LCR 93-0072 TITLE: USAR Change - Miscellaneous Changes DESCRIPTION: This LCR made various editorial changes, clarifications, and added or revised acronyms for existing plant terminology and componenti The Average Power Range Monitoring (APRM) system calibration frequency for electronic apparatus that have the potential to drill was changed from every three days to every seven days; this number is in accordance with the Technical Specifications. This LCR also revised the number of Scram Discharge Volume high level channels per trip system from two to four; this is consistent with Technical Specifications. A sentence referring to the vessel head spray line was deleted; the head spray line was removed under DC 86-78.

SAFETY EVALUATION: These changes do not impact the manner in which any system, structure, or component is maintained, operated, or controlled. The change concerning APRM system calibration frequency is in accordance with Custom Technical Specification 4.1 Bases and as shown in Custom Technical Specification Tables 4.1.1 and 4.1.2. These changes do not impact any components or systems used to monitor, control, or report the radiation exposure of plant personnel or the general public, or impact the method or process ofdealing with any radiological / contamination concerns. This LCR does not impact the function of any equipment considered to be important to safety or safety-related. No new control devices / schemes are added to the plant and no new equipment or structures are added. No new types of plant events or malfunctions of equipment important to safety are created. Changes are consistent with the Technical Specifications and do not reduce the margin of safety as defined in the basis for any Technical Specification.

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LCR 94-0001 l l

TITLE: USAR Change - Miscellaneous Changes DESCRIPTION: This LCRincluded the following changes: 1) corrected Control Rod Drive drive water presstire to 265 psi,2) corrected discussion ofinitial purging of primary containment with nitrogen and the method of nitrogen make-up,3) clarified logic ofIligh Pressure Coolant Injection auxiliary oil pump control switch,

4) added a statement to reflect a third narrow range reactor water level instmment,5) clarified Automatic Depressurization System control logic power transfer, 6) removed statement that the Reactor Recirculation (RR) pumps are tripped on e Loss of Coolant Accident by the Low Pressure Coolant Injection (LPCI) initiation logic,7) clarified response of Shutdown Cooling Mode to LPCI initiation,

8) corrected values of reactor pressure permissive interlocks for the closure of RR discharge valves and opening of LPCI injection valves,9) deleted statement regarding LPCI loop select logic, and 10) clarifted derription of Diesel Generator breaker interlocks. All of these changes resulted from the Station Operations Resiew Committee review of Revision 11 to the USAR.

SAFETY ,

EVALUATION: These changes correct USAR internal inconsistencies and terminology, clarify USAR text, and resise I discussion ofequipment operation. The changes are consistent with approved plant design and have no cITect on plant event initiating actions. They do not affect the way in which plant equipment is currently l

designed, operated, ormaintained. Equipment manipulations and functions, as clarified or resised by this LCR, are consistent with existing design and operation and do not afTect the equipment's ability to function as required in response to plant events. All equipment will continue to function as aesigned to prevent the release ofradiation to the public. USAR evaluations / discussions of equipment failures and malfunctions are not afTected These changes have no adverse efTect on equipment redundancy or reliability. None of the changes involve the addition or modification of any plant equipment and no new equipment failure modes are created. The corrections of USAR values and discussions of equipment operation are consistent with the Technical Specifications and their Bases. Therefore, these changes do not reduce the margin of safety as defined in the basis for any Technical Specification.

LCR 94-0011 TrfLE: USAR Change - CNS Operations Manual DESCRIPTION: The USAR was revised to update the discussion of the organization of the procedures contained in the CNS Operations Manual. This change reflects the organization of the procedures as described in CNS Procedure 0.1, Introduction to CNS Operations Manual.

SAFETY EVALUATION: These changes do not impact the manner in which any system, structure, or component, is maintained, operated, or controlled. They do not impact any components or systems used to monitor, control, or report the radiation exposure of plant personnel or the general public. Therefore, these changes do l increase theprobability or occurrence or consequences of a previously evaluated plant event. This LCR does not impact the function of any equipment considered to be important to safety or safety-related, or any equipment used to monitor, control, or detect radiation or contamination. Therefore, these changes do not increase the probability ofoccurrence or consequences of a previously evaluated malfunction of equipment important to safety. No new equipment, structures, or control devices are being added to the plant. The number of personnel required to be on site to fulfill specific functions is not impacted and personnel qualifications fbr specific job descriptions are not afTected. This change to the USAR is an administrative type change and does not impact the function, description, or operation of any system, stmeture, or component in the plant. Therefore, no new or difTerent types of plant events or equipment malfunctions are introduced. These changes do not impact the procedure discussions in the Administrative Controls section of the Technical Specifications. As there no impact on the Tecimical Specifications or the basis for any Technical Specification, there is no reduction in the margin of safety as defined in the basis fbr any Technical Specification.

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LCR 94-0012 l

TITLE: USAR Change - Recirculation Flow Control System DESCRIPTION: The description of the Recirculation Flow Control System in the USAR was revised to: 1) add a description of Reactor Recirculation (RR) pump runback on low reactor water level and less than two feed pumps in operation,2) correct the RR Motor Generator speed limiter setpoints, and 3) make various changes to reflect standard terminology and add cross references to other USAR sections. The RR runback circuit was previously shown on a USAR Figure, but was not discussed in the text. The speed limiter setpoints are consistent with those in Cooper Nuclear Station Operating Procedures.

SAFETY l EVALUATION: This change clarifies an existing drawing presently in the USAR and brings the USAR up to date with l

current setpoints. It provides consistency between the station operating procedur.:s and the USAR. It l

will not result in any physical changes to any plant equipment. Therefore, these changes will not increase

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the probability ofoccurrence or consequences of a plant event or malfunction of equipment important to safety and will not create any new types of plant events or equipment malfunctions. There will be no '

reduction in the margin ofsafety as defined in the basis for any Technical Specification as a result of these changes.

LCR 94-0013 TITLE: USAR Change - Iligh Pressure Coolant Injection 0 IPCI) Description DESCRIPTION: This LCR made various changes to the description of the IIPCI system in Section VI of the USAR for consistency with other USAR sections. Changes included: 1) addition of a paragraph to note that 1IPCI will operate during operational transients in which reactor water level decreases to the system initiation point at Level 2,2) discussion ofinadvertent IIPCI actuation with reference to Section XIV-5, " Analysis ofAbnormal Operational Transients," and 3) identification of a suppression pool suction strainer in the 1IPCI components discussion.

SAFETY EVALUATION: No physical change is being made to plant systems or to the plant transient analysis. Descriptions that occur in one location of the USAR are being referenced in another for clarity. The analyzed events and the IIPCI configuration remain the same. No equipment modifications are made with this USAR change.

This change adds a reference for two plant transient events to the IIPCI system discussion. Ilowever, these two transients (IIPCIinadvertent start and loss of feedwater) are described in USAR Section XIV.

The suction strainer was previously identified on a USAR Figure and is being added to the text for clarity.

The additions made by this USAR change have been previously analyzed and are part of the CNS design basis. There is no increase in the probability of occurrence or consequences of a plant event or malfunction of equipment important to safety and there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

LCR 94-0020 UCR 98-006 TITLE: Revision of USAR Table VII-2-1 DESCRIPTION: This LCR and UCR were generated to revise Table VII-2-1, Reactor Protection System Instrumentation Environmental Conditions, to reflect the current environmental conditions described in Equipment l

Qualification Data Package (EQDP) 46. LCR 94-0020 changed the Abnormal Condition maximum temperature for the iligh Pressure Coolant Injection (IIPCI) room and the Reactor Core Isolation Cooling (RCIC) equipment area to 290'F and changed the IIPCI toom maximum pressure to 7.7 psig. These values were subsequently determined to be incorrect and UCR 98-006 revised the subject iIPCI/RCIC maximum temperatures to 293 *F and the IIPCI room maximum pressure to 7.5 psig.

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SAFETY  !

EVALUATION: The identified temperatures establish the aging characteristics of electrical equipmeni imponant to safety. l The changes will provide consistency among station documents. The proposed changes are in EQDP 46 which is used to evaluate specific equipment performance for all service conditions. The changes to Table VII-2-1 raise the EQ profile temperature maximum which is used to evaluate equipment performance during accident conditions. Therefore, these changes will not increase the probability of occunence or consequences of a plant event or malfunction of equipment important to safety. These changes will not create the possibility of a plant event of a different type than those outlined in the USAR.

Design Basis Accidents, with respect to EQ, are defined as High Energy Line Breaks outside primary containment and a Loss-of-Coolant Accident inside primarf containment. There will be no possibility to create a ddTerent type ofequipment malfunction with the mcorporation of these changes. The changes do not impact the margm of safety as defined in the basis for any Technical Specification.

LCR 94-0027 TITLE: USAR Change - Miscellaneous Changes DESCRIPTION: This LCR made various changes to the USAR text for clarification of components under discussion, I correction of tenninology, deletion of redundant information, and correction ofinternal inconsistencies. l The changes are consistent with approved plant design and station procedures. Changes included: 1

1) deletion of Figure X-8-2 which was redundant to Figure IV-8-1,2) correction to indicate that the suppression pool cooling subsystem controls suppression pool water temperatures within the limits specified in the Technical Specifications, 3) correction to indicate that there is only one automatic High Pressure Coolant Injection valve in the suction line from the torus, 4) addition of statement that control room air conditioner is classified as a post-LOCA heat load,5) clarification that personnel radiation exposure records are maintained using thermoluminescent dosimetry instead of film badges, l

6) incorporation of correct document title for " Supplemental Reload Licensing Submittal," 7) revision of Section XIV discussion to indicate that the containment pressure of 62 psig in the Case D discussion is the ' maximum allowable' versus the ' design' pressure to be consistent with Section V, and

) 8) clarification that the results and consequences for a loading error accident are provided in the I

Supplemental Reload Licensing Submittal.

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! EVALUATION: The changes made by this LCR are administrative in nature and have no effect on plant event initiating actions. They do not afTect the way in which plant equipment is currently designed, operated, or i

maintained. Equipment functions, as clarified by these changes, are consistent with existing design and operation. All equipment required to suppat plant events will continue to function as designed to prevent the release ofradiation to the plant. The changes have no effect on equipment attributes associated with malfunctions and failures. They are not associated with the addition or modification of any plant equipment. No new equipment failure modes are created. The changes have no effect on the Technical Specifications or Bases. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

LCR 94-0030 LCR 95-0056 TITLE: USAR Change - Revision of Table VIII-5-1 DESCRIPTION: These LCRs revised USAR Table VIII 5-1 which reflects Diesel Generator loading. The USAR was revised to reflect changes based on approved Nuclear Engineering Department Calculation (NEDC) 87 104A, Plant AC Load Study. LCR 94-0030 was based on Revision 11 of the calculation and LCR 95-0056 us based on Revision 13 of the calculation. Changes to calculation NEDC 87-104 A, which is a design basis document, were initiated by various approved design change documents. Therefore, the calculation did not identify or document any adverse impact to the Diesel Generator loading.

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i SAFETY j

. EVALUATION: The loading and operation of the Diesel Generators are not adversely impacted by these changes. These i changes do not alter the operation of any radiation monitoring equipment nor alTect any parameters associated with radiation monitoring. The function of equipment designed to control the release of radiation is not alTected by this activity. The changes made to the USAR were the result of plant l approved changes that alTected calculation NEDC 87-104A. These changes do not degrade below the design basis the performance of a system assumed to function in the accident analysis in the SAR. They do not alTect the manner in which plant equipment is operated, controlled, or maintained. The changes are associated with the addition or modification of plant equipment as approved by other design change documents. The revision of the calculation does not decrease the reliability of a system or equipment important to safety assumed to function in the accident analysis. Plant equipment will function as j designed during planned operations, transients, accidents, and special events. No new type of plant event  !

or equipment malfunction is introduced by these changes. The Diesel Generators are discussed in the I

Technical Specifications but the changes initiated by these LCRs do not afrect the basis for the Limiting l Conditions for Operation or the Surveillance Requirements. Therefore, there is no reduction in the l margin of safety as defined in the basis for any Technical Specification.

LCR 94-0037 TITLE: USAR Change - Multipurpose Facility (MPF) Ventilation System DESCRIPTION: This LCR added a description of the Multipurpose Facility IIcating, Ventilation, and Air Conditioning System to the USAR. This change was based on the installation of the MPF ventilation system per approved design documents.

SAFETY EVALUATION: inis change does not alter any systems, structures, or components required to safely shut down the reactor )

and maintain it in a safe shutdown condition. It is not associated with any of the initiators of a plant event as described in the USAR; therefore, this change does not increase the probability of a previously evaluated plant event. The MPF ventilation system was installed by previously approved design documents; the addition of this information to the USAR does not impact the operation of the MPF ventilation system. Radiation monitoring components are calibrated / tested in accordance with plant procedures to verify proper operation. The changes to the USAR do not impact the Offsite Dose Assessment Manual or any Process Control Procedures. Therefore, this LCR does not increase the consequences of a plant event previously evaluated in the SAR. The MPF ventilation system is not associated with any safety-related equipment or any equipment required to meet the requirements of the i

Technical Specifications. This change does not alter any ventilation path that is monitored for gaseous  !

ellluent releases. It does not impact the method or process of dealing with any radiological / contamination  !

concems. This change does not impact any radiation monitoring equipment required per the Technical Specifications. It does not impact the description, control, or function of any equipment used to support reactor operation or support the generation of power. No different types of plant events or malfunctions l

ofequipment important to safety are created by this USAR change. This change does not impact any of l the radiation monitoring requirements discussed in the Technical Specifications. The MPF ventilation system is not addressed by the Technical Specifications or any of its bases. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

I CR 94-0041 1

TITLE: USAR Change - Service Water (SW) Flow DESCRIPTION: This LCR clarified the description of SW flow distribution to the diesel generator components during low pressure conditions. In order to remove the implication that orifice restriction and manual throttling are both normally used to ensure adequate distribution of SW to the served components, the discussion of manualthrottling of valves SW-V-268 and SW-V-269 was moved to a separate sentence to indicate that it would be a separate action, if required.

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SAFETY EVALUATION: This change is simply a clarification of SW flow balancing mechanisms for the diesels. There are no plant component changes associated with this USAR change, and no changes in plant activities or plant analysis. SW flow to each diesel generator remains at 1400 gpm, as assumed in the safety analysis; this flow is verified by procedure. The orifice function to maintain flow through the diesel generator remains unchanged. Therefore, this change does not increase the pmbability of occurrence er consequences of a plant event or malfunction of equipment important to safety and no different types of plant events or equipment malfunctions are introduced by this USAR clarification. This change is not the subject of any Technical Specification bases and the margin of safety is not reduced.

I CR 94-0049 TrfLE: USAR Change - Low Pressure Coolant Injection (LPCI) Valve Control DESCRIPTION: This LCR revised the stroke time for LPCI Injection Valves from 24 to 27 seconds, and indicated the stroke time applies only to the inboard injection valves, RIIR-MO-25A/IL It deleted the stroke time limit for contamment cooling valves and RHR system test line isolation valves. It also revised the discussion of automatic closure of the containment spray valves, shutdown cooling return valves, and containment cooling return valves to clarify that automatic closure is to assist the Control Room Operator in maximizing injection into the reactor vessel from the LPCI subsystem. These changes to the USAR were based on a review of design documents. The changes are consistent with stroke time requirements in support of LPCI operation as identified in Operating Plant Licensing Form OPL-4, Resolution of Emergency Core Cooling System Parameters, and also in USAR Table VI-5-1, Significant Input Parameters to the Loss of Coolant Accident Analysis.

SAFETY EVALUATION: This LCR corrected discrepancies in the USAR test which discusses LPCI operation to provide consistency with other licensing basis documents. The changes have no effect on plant event initiating actions or equipment response. They do not alTect the way in which plant equipment is currently designed, operated, or maintained. The changes simply provide a more correct description of the I parameters required in support of LPCI operation. All equipment required in support of plant events will continue to function as designed to prevent the release of radiation to the public. USAR evaluations of equipment failures ud malfunctions are not afTected. Stroke time testing of associated valves will continue to be performelin accordance with the CNS Inservice Testing Program. Since the changes are j consistent with LPCI opention parameters as identified in OPL-4, the changes have no adverse efTect on equipment attributes associawd with malfunctions or failures. Since the changes have no efTect on malfunctions of equipmen' iro portant to safety, there is likewise no effect on the consequences of a malfunction of equipment inrartant to safety. These changes are not associated with the addition or modification of any plant equipment. Equipment functions remain consistent with approved plant design and operation and no new equipment failure modes are created. The changes have no elTect on the Technical Specifications or their Bases. They do not impact any valves whose stroke times are listed in the Technical Specifications. Therefore, they do not reduce the margin of safety as defined in the basis for any Technical Specification.

LCR 94-0050 TITLE: USAR Change - Penetration X-218 DESCRIPTION: This LCR revbed the size for primary containment penetration X-218 from 4" to 2" on USAR Table V 2, Penetration Schedule. X-218 is a spare penetration. The actual installed size of the penetration was not altered. A review of station design documentation revealed that the actual penetration size was 2".

SAFETY EVALUATION: This change corrects a textual error in the USAR and has no efTect on plant event initiating actions or equipment response. it ckies not affect the way in which plant equipment is currently designed, operated, or maintained. All equipment required in support of plant events continues to function as designed to j prevent the release of radiation to the public. USAR evaluations / discussions of equipment failures and

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I I

I malfunctions are not affected by this text change. This USAR change is consistent with approved plant i

design and has no effect on equipment function. No plant equipment is added or modified as a result of l this change, and no new failure modes are created. This change has no effect on the Technical

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Specifications or their Bases. Therefore, there is no reduction in the margin of safety as defmed in the basis for any Technical Specification.

1 LCR 94-0062 TITLE: USAR Change - Locked-Open/ Sealed-Open Valves DESCRIPTION: This LCR revised the USAR to clarify the defmition of" sealed-open" and " kicked-open" pertaining to manual valve positions. The change indicated that the two terms are interchangeable.

SAFETY EVALUATION: His change was a clarification of a statement pertaining to valve position terminology. It has no effect on any system's function or design. It does not affect any plant equipment used to support plant operation. There is no impact to any system or component for monitoring radiological conditions.

Therefore, there is no increase in the probability of occurrence or consequences of a plant event previously evaluated in the SAR. This change has no effect on attributes associated with malfunctions or failures to safety-related or important to safety equipment. No structure, system, or component is being altered or added. Here will be no variation in the way that equipment or systems are operated as a result of this change. No new equipment related failure scenarios are created. No safety limits, limiting safety system settings, or design parameters are afTected. This change does not affect any Technical Specificatior' - r its basis; therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

I CR 94-00(>3 TITLE: USAR Change - Core Spray (CS) and Low Pressure Coolant injection (LPCI) System Tests DESCRIPTION: The USAR indicated that the CS and LPCI systems outside the drywell are checked for leaks during system flow tests. He associated surveillance procedures did not specifically require a general inspection for leaks during testing on the piping ortside the drywell. A statement was added to the USAR to take credit for periodic operator tours as a means of providing routine system leak checks.

SAFETY EVALUATION: This change (kx:s not imr et the manner in which any system, structure, or component is maintained. No accident initiators or plar.t ;; vent initiators are impacted by this change. It does not require the operator to perfonn any new/ additional functions. This change does not impact any components or systems used to monitor, control, or report the radiaticin exposure of plant personnel or the general public. It does not impact the function of any equipment considered to be important to safety or safety-related. No new control devices / schemes, equipment, or structures are being added to the plant. The change is based on existing surveillance testing / plant inspection requirements performed by Operations pasonnel. It de not impact any Technical Specification or the basis for any Technical Specification, and does not i wt the surveillance requirements or method of performing tests for the Residual IIcat Removal or CS systems. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

LCR 94-0065 TITLE: USAR Change - Alternate Shutdown Capability l DESCRIPTION: His LCR added clarifications to the USAR discussing Alternate Shutdown functions, capabilities, and limitations. The Alternate Shutdown System was installed via Design Change 86-021 and previously i

included in the USAR. Ilowever, this LCR clarified the interface of Alternate Shutdown with other systems and added cross references to the USAR section which specifically discusses Alternate Shutdown capability.

i I

SAFETY EVALUATION: These changes have no effect on actions / scenarios which initiate plant events. They do not afTect the way in which plant equipment is currently designed, operated or maintained. The changes simply provide clarification of the Alternate Shutdown System interface with associated systems. All equipment required in support of plant events will continue to function as designed to prevent the release of radiation to the public. USAR evaluations / discussions of equipment failures and malfunctions are not afTected. Current 3

methods of testing, along with surveillance program administrative controls, provide full compliance with i

Technical Specification testing and operability requirements. All equipment will continue to perform as required in providing redundsney and reliability. The changes are not associated with the addition or modification of any plant equipment Equipment functions remain consistent with approved plant design l

and operation, and do not create any new equipment failure modes. The changes do not alTect the information/ requirements presented in the Technical Specifications and the associated Bases; therefore, there is no reductie in the margin of safety.

LCR 94-0075

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TITLE: USAR Change - Core Spray (CS) Testing DESCRIPTION: 'Ihe USAR was revised to clarify that Cooper Nuclear Station is not required to test the CS sys'em with injection to the reactor vessel. The previous wordingjustified not injecting cold water to the vessel during l power operation; however, the wording could have been misconstrued to imply that injection was i

required if not in power operation. The change eliminated wording that discussed test conditions that l

were applicable to Preoperational System Tests that were performed prior to fuel load. No required plant testing was climinated.

SAFETY EVALUATION: These changes were made to prevent a misunderstanding that periodic testing by injecting water into the vessel is required. The changes are textual only and have no etTect on actions / scenarios which could initiate plant events. The manner in which the CS system is tested to verify operability is not altered. The changes do not afrect the way in which plant equipment is currently designed, operated, or maintained.

The current methods of testing components in the CS injection path are adequate to ensure that the system will deliver full flow to the reactor vessel in the event of an accident which requires the system to function. All equipment required in support of plant events will continue to function as designed to prevent the release of radiation to the public. No radiation monitoring equipment or procedures are im.pacted by this change. USAR evaluations / discussions of equipment failures and malfunctions are not l afTected by this change. No new test requirements are being added and no normal operation system

! testing requirements are being deleted. Therefore, there is no increase in the probability of a malfunction of equipment important to safety previously evaluated in the SAR. This change has o effect on I equipment attributes associated with malfunctions and failures. All equipment will continue to perform

) as required in providing redundancy and reliability. No new failure modes are created and no ditTerent types of plant events or equipment malfunctions are introduced. The changes do not alTect the requirements presented in the Technical Specifications and the associated Bases. Therefore, this USAR change does net reduce the margin of safety as defined in the basts for any Technical Specification.

LCR 94-0094 TITLE: USAR Change - Figures IV-8-2, VII-4-4, and VIl-4-6 j l

l DESCRIP TION: The subject USAR Figures were revised to add threaded plugs or pipe caps downstream of the following  ;

vent or test valves: RI1R-V-185, R1IR-V-145, RHR-V-146, R1IR-V-147, and CS-V-54. Also, valve '

RIIR-V-185 was not presiously shown and was added to Figure IV-8-2. These changes were the result of a primary containment walkdown which identified the subject salves located between containment isolation valves and containment without a second valve or cap indicated on the USAR figures. The use of a single, manually operated valve with a threaded cap or plug for vents or test connections is in acconlance with the USAR and has no etTect on the safe operation of the plant. These drawing changes are of an editorial /dratling nature.

i I

SAFETY EVALUATION: The subject changes will make the figures consistent with the design bases of the plant and other sections of the USAR. The changes do not impact the way any system, component, or structure is maintained, operated, or controlled. They ck) not contnbute to any accident initiators discussed in the SAR and do not impact the mitigation of an accident. There is no equipment important to safety impacted by this change.

No new equipment is added and no existing equipment is modified by this activity. The drawing changes will not create the possibility of a plant event or malfunction of equipment important to safety difTerent than previously evaluated in the SAR. The changes have no efTect on the margin of safety as described in the Technical Specifications or the basis of any Technical Specification.

I CR 94-0096 TITLE: US AR Change - Residual Ileat Removal (R)IR) Pump Control DESCRIPTION: This LCR revised the discussion of RIIR pump control logic to eliminate wording which indicated the pump control circuitry prevents a pump from starting unless a suction path is lined up. The actual design configuration is that with no suction path available, the pump can be started but will trip immediately.

Ilowever, the function of preventing RIIR pump damage due to overheating at no flow is still satisfied.

The suction valve control signals are a part of the " trip" circuit and not the " start" circuit. This change corrects a discrepancy in the USAR to reflect the original plant design configuration.

SAITsTY EVALUATION: This change has no effect on plant event initiating actions or equipment response. It does not revise the i manner in which the RIIR pumps are operated or tested. It does not afTect the way in which plant equipment is currently designed, operated, or maintained. The change simply corrects an error in the discussion of pump interlocks with associated suction valves. It has no efTect on the R1IR system's ability to operate in support ofplant events. The equipment will continue to function as designed to prevent the l release of radiation to the public. This change does not impact the function or operation of any {

radiation / contamination monitoring equipment or radiological control procedures. USAR evaluations / discussions ofequipment failures and malfunctions are not affected. The change is consistent with approved plant design as shown in preoperational tests and General Electric design drawings. It has no effect on the intended function of preventing pump damage due to operating the pumps with an incorrect suction path alignment. No equipment controls are altered and no plant equipment is added or ,

modified. The change does not affect existing plant procedures and does not create any new equipment J failure modes. The RIIR pump and valve control circuitry is currently tested by Procedure 6.RIIR.308, which is performed in order to meet Technical Specification requirements. Procedure 6.RIIR.308 is not impacted by this USAR change. The manner in which operability tests for the IU IR system are performed is not afTected. The change has no efTect on the information in the Technical Specifications or Bases; therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

LCR 94-0097 1

TITLE: USAR Change - Iligh Pressure Coolant Injection (IIPCI) Valve Control l

DESCRIPTION: This LCR corrected several discrepancies in the USAR section on IIPCI Valve Control as  !

follows; l) deleted discussion of turbine trip from sentence which discusses manual reset devices for valves which automatically ck>se, as the IIPCI turbine trip circuit does not scal-in and therefore does not require manual reset,2) correctly identified the IIPCI valve which has a stroke time requirement in support of HPCI automatic operation (i c., the iIPCI injection valve vs. the iIPCI pump discharge valve) and corrected stroke time from 20 seconds to 19 seconds; this is consis'ent with Operating Plant Licensing Form OPL-4, 3) deleted statement that a IIPCI actuating signal will open the IIPCI containment isolation valves if they are. closed, as this feature was previously removed by DC 87-191,

4) clarified / corrected the discussion 6f automatic opening of the IIPCI pump suction valve from the Emergency Corxlensate Storage Tank (ECST), and 5) revised the number of level switches used to detect

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the ECST low water level cor.dition from two to four which is in accordance with plant design drawings,  ;

the IIPCI preoperational test, and Technical Specifications. I SAFETY EVALUATION: These changes to the USAR pmvide c3nsistency with other licensing and design basis documents. They have no effect on plant event initiating actions or equipment response. The change in stroke time for the IIPCI injection valve is based on the value used for the Reload Loss of Coolant Accident Analysis.

Therefore, these changes do not increase the probability of a previously evaluated plant event. These changes do not afTect the way in which plant equipment is currently designed, operated, or maintained.

They do not impact the operation of any radiation monitoring equipment. The changes simply proside a more correct description of the parameters required in support ofIIPCI operation. All equipment required in support of plar t events will continue to function as designed to prevent and monitor the release ofradiation to the public. USAR evaluations / discussions of equipment failures and malfunctions are not affected. No new equipment is added to the plant. Stroke time testing of associated valves will continue to be performed in accoidance with the Cooper Nuc. lear Station Inservice Testing Program.

l Since the changes are consistent with IIPCI operation parameters as identified in OPL-4 and with current

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plant design, the changes have no adverse etrect on equipment attributes associated with malfunctions l

and failures. Therefore, there is no increase in the probability of occurrence or consequences of a l previously evaluated malfunction of equipment important to safety. Equipment functions remain consistent with approved plant design and operation, and do not create any new equipment failure modes.

The changes are consistent with the information presented in the Technical Specificatians and the associated Bases; therefore, there is no reduction in the margin of safety as defined in the basis for any ,

Technical Specification. I LCR 94-0033 UCR 98-042  !

1 l

TITLE: USAR Change - Emergency Condensate Storage Tank (ECST) l l DESCRIPTION: This change resised the ECST low level requirements to reficct Technical Specification requirements.

l Technical Specifications require 10,000 gallons usable water in the ECSTs for use by the Reactor Core i Isolation Cooling (RCIC) and/or Iligh Pressure Coolant Injection (HPCI) systems. Nuclear Engineering Department Calculation 92-050K determined that an instrument setpoint of 24 inches above the inside bottom of the tanks assures the Technical Specification requirement is satisfied. LCR 94-0098 incorrectly changed the low level switch setpoint elevation from 880'7" to 880'9". The correct elevation is 880'10" and was subsequently corrected by USAR Change Request 98-042.

SAFETY EVALUATION: These USAR changes bring the USAR into agreement with the Technical Specifications and approved plant design with respect to the low level setpoint and remaining water volume for the ECSTs. The chang, have no effect on actions / scenarios which initiate plant events. They do not affect the way in which plant equipment is currently designed, operated, or maintained. The HPCI and RCIC systems will continue to function as designed in support of plant events in order to mitigate the consequences of an event and prevent the release of radiation to the public. The change actually indicates an improvement m the time available (due to an increased volume of water) for transfer ofIIPC1/RCIC suction to the suppression pool without a loss of suction, in the event of a low ECST level. There is no effect on equipment attributes associated with malfunctions and failures. The changes are not associated with the I addition or modification of any plant equipment and no new equipment failure modes are created.

Therefore, no new types of plant events or equipment malfunctions are introduced. The changes bring the USAR information into agreement with Technical Specification requirements; therefore, there is no reduction in the margin of safety as dermed in the basis of any Technical Specification. j

( LCR 94-0103 TITLE: USAR Change - Reactor Water Cleanup (RWCU) System Surveillance Test DESCRIPTION: This LCR deleted the surveillance test requirement related to the RWCU system return isolation valves from USAR Table VII-3-5, Original Suiveillance Test Frequencies for Automatic Isolation Valves of the Primary Containment and Reactor Vessel Isolation Control System. This table indicated that the subject valves should be tested "by actuation of each isolation signal at each refueling outage." This change was warranted because only the supply isolation valves close upon receiving auto isolation signals. The retum isolation valve does not have any auto isolation signals associated with it.

SAFETY EVALUATION: This change removes a misleading reqmrement pertanung to testing of the RWCU return isolation valves It requires no physical changes to plant equipment or systems. The change in no way alters the system's function or design. It does not alTect the way in which any plant structure, system, or component is designed, operated, or maintained. The equipment required to support the plant will continue to function per design to prevent any increase in radiological consequences to plant personnel or the public. This LCR has no direct or indirect effect on equipment important to safety. There is no impact to any component or system that would increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR. No physical changes are being made to any plant equipment and no interfaces to any safety-related equipment are involved. No systems required to mitigate the consequences ofan accident or to bring the reactor to a safe shutdown condition are alrected.

The change ha no efTect on any existing Technical Specification or the basis for any Technical

{

j Specification, therefore, there is no reduction in the margin of safety.

LCR 94-0106 TITLE: USAR Change - Figures IV-3-3, VII-4-1, and X-12-1  !

I DESCRIPTION: The subject USAR Figures were revised to add threaded plugs or pipe caps downstream of the following vent, drain,or testvahrs: PC-V-199,IIPCI-V 115,IIPCI-V-260,IIPCI-V-218,IIPCI-V-221,IIPCI-V-289, IA-V-211, and SA-V-141. These changes were the result of a primary containment walkdown which identified the subject vahrs k)cated between containment isolation valves and containment without i a second valve or cap indicated on the USAR figures. The use of a single, manually operated valve with a threaded cap or plug for vents or test connections is in accordance with the USAR and has no etTect on the safe operation of the plant. These drawing changes are of an editorial /drailing nature.

SAFETY EVALUATION: The subject changes will make the figures consistent with the design bases of the plant and other sections of the USAR. They do not contribute to any accident initiators discussed in the SAR and do not impact l the mitigation of an accident. There is no equipment important to safety impacted by this change. No new equipment is added and no existing equipment is modified by this activity. The drawing changes will not create the possibility of a plant event or malfunction of equipment important to safety diiTerent than previously evaluated in the SAR, The changes have no efTect on the margin of safety as described in the Technical Specifications or the basis of any Technical Specification.

1 CR 94-0124 TITLE: USAR Change - Core Spray Testable Check Valves i

DESCRIPTION: This LCR clarified that the drywell must be de-inct ied to allow personnel access for manually testing the Core Spray testable check nives during reactor cold shutdown periods. This change is based on Inservice Testing (IST) Program Relief Request !!V-50 which was approved by the NRC in 1991. This i LCR also deleted a statement which could haye been misinterpreted to mean that the valve indicator position light in the Control Room is tested during each valve exercise test. The two tests (manual

exercising and remote position indication) have difTerent required frequencies. The remote position indication is (mly required to be tested once every two years per the IST program and, therefore, may not be performed each time the valves are manually exercised (quarterly during cold shutdown /de-inerted j

conditions). The testing of the position indication is in accordance with the requirements of the Technical j Specifications.

SAFETY  ;

EVALUATION: The subject valves are tested in accordance with the CNS IST Program which complies with the requirements of ASME Section XL The changes have no effect on plant event initiating actions. They do not afTect the way in which plant equipment is currently designed, operated, or maintained. All equipment required in support ofplant events will continue to function as designed to prevent the release of radiation to the public. USAR discussions of equipment failures and malfunctions are not affected.

The changes have no effect on equipment attributes associated with malfunctions and failures. Therefore, there is no increase in the probability of occurrence or consequences of a previously evaluated malfunction of equipment important to safety. The changes are not associated with the addition or mochfication of any plant equipment, and do not change the current method or frequency of valve testing.

Equipment functions remain consistent with existing design and operation, and these changes do not create any new equipment failure modes. Therefore, no new types of plant events or equipment  !

malfunctions are introduced. The changes to the USAR are in accordance with the Inservice Inspection requirements discus:*xiin the Technical Specifications. The changes have no etTect on existing Technical Specifications and their Bases. Therefore, there is no reduction in the margin of safety as dermed in the basis for any Technical Specification.

LCR 94-0127 TITLE: USAR Change - High Pressure Coolant Injection (HPCI) Testing DESCRIPTION: This LCR removed discussion from the USAR concerning testing IIPCI with suction from the suppression pool. This information was removed because such testing is not performed and is not required to be perfanned. Inservice Testing Program (IST) Relief Request RV-20, which was previously approved by the NRC, eliminated the requirement to perform a full system flow test on the 1IPCI Suction Check Valve,IIPCI-CV-11CV, because this would require taking a suction on the suppression pool and j result in the introduction oflow-quality water into the reactor vessel The NRC acknowledged that such j a test would be hannful to the system. The discussion removed by LCR 94-0127 pertains to the IIPCI ,

Injection Check Valve,IIPCI-CV 29CV. The basis for deleting the system flow test with the suction I from the suppression poolis the same as that accepted by the NRC for iIPCI-CV-11CV. Other methods are utilized for testing the system components in the IIPCI suction path from the suppression pool to assure proper operation. IIPCI system valves are tested in accordance with the IST program to ensure continued operability.

SAFETY EVALUATION: This change has no efTect on actions / scenarios which initiate plant events. The manner by which the design functional test of the !IPCI system is performed is not altered or impacted by this change. It does not affect the way in which plant equipment is currently desigt.ed, operated, or maintained. The current methods of testing components in the llPCI suction path from the suppression pool will ensure the equipment will perform properly. All equipment required in support of plant events will continue to function as designed to prevent the release of rediation to the public. This USAR change deletes a statement about iIPCI testing which, if perfonned, could result in fouling and degradation of equipment important to safety. The test methods currently in use are an enhancement with respect to the deleted USAR statement. This change has no etreet on equipment attributes associated with malfunctions and failures. It is not associated with the addition or modification of any plant equipment. Equipment functions remain consistent with approved plant design and operation, and no new equipment failure modes are created. The change does not alTect information/ requirements presented in the Technical Specifications and associated Bases. Therefore, there is no reduction in the margin of safety as defmed in the basis for any Technical Specification.

LCR 94-0130 LCR 88-0043 TITLE: USAR Change - Iligh Pressure Coolant Injection (HPCI) Steam Exhaust Line Vacuum Breakers DESCRIPTION: LCR 88-0043 was initiated to change the HPCI Steam Exhaust Line Vacuum Breaker description from

" swing" to" lift" 'lhe "lifl" type valve was a part of the original plant design. Prior to its incorporation, LCR 88-0043 was superseded by LCR 94-0130. This USAR change deleted the word " swing" from the l

discussion of the check valve function. It was determined that the inclusion of the specific type of check l

valve in the USAR was an unnecessary level of detail and the decision was made to remove the check l l

valve type rather than replace " swing" with " lift" Removal of the check valve type does not affect the component function of preventing the HPCI turbine steam exhaust line from being flooded by siphoned torus water. This LCR also made an editorial correction to the number of a reference document.

SAFETY l EVALUATION: These changes have no effect on plant event initiating actions or equipment response. They do not affect the way in which plant equipment is currently designed, operated, or maintained. The changes simply remove an unnecessary level of detail from a discussion of component function, and correct an error in the reference section. All equipment required in support of plant events will continue to function as designed to prevent the release of radiation to the public. USAR evaluations / discussions of equipment failures and malfunctions are not aflfected. The USAR changes are consistent with approved plant design which existed at the time the LCR was processed. They have no effect on single failure criteria attributes of the system design. The changes are not associated with the addition or modification of any plant equipment. Equipment functions remain consistent with approved plant design and operation, and no new equipment failure modes are created. Therefore, no new types of plant events or equipment malfunctions are introduced The changes have no impact on the information presented in the Technical  ;

Specifications or their Bases, therefore, they do not reduce the margin of safety as dermed in the basis for l any Technical Specification.

LCR 94-0140 TITLE: USAR Change - Emergency Transformer DESCRIPTION: The USAR was revised to indicate that relay 27/ET4 is a redundant alarm to relay 27/ET3 during emergency station senice transformer low voltage conditions. Relay 27/ET4 was already installed and was part of the plant design, but was not listed as a redundant alarm to relay 27/ET3 in the USAR. This change makes the USAR text agree with USAR Figure VIII-5-L SAFETY I EVALUATION: The subject change is administrative in nature. USAR Figure VIII-5-1 already shows both relays 27/ET3 l and 27/ET4 connected to the common bus of the emergency station senice transformer. This '

administrative change will not change the Emergency AC Power System function, operation, and performance or affect any system interface in a way that could increase the probability of an accident.

Systems and equipment will act be alTected by this change such that they are degraded or operated outside their design or test limits. This change does not afTect instrument accuracy or response characteristics in a manner that could increase the probability of an accident. It does not increase the probability of  ;

operator error or add complexity to human factor conditions. The circuit associated with the unden oltage  ;

relays is not connected to any radiation monitoring equipment and is not used to monitor or document any l parameters associated with radiation monitoring. The function of equipment designed to control the '

release of radiation is not affected. This activity does not change, degrade, or prevent actions required for any accident or alter the assumptions made in evaluating the consequences of an accident or i malfunction ofequipment important to safety. No physical work in the plant is required for this change.

It does not decrease the reliability of a system or equipment important to safety assumed to function in i the accident analysis, and does not affect the ability of the relays to perform essential auxiliary functions required for each of the operating modes and plant events they are required to function under. The l Emergency AC Power System will continue to function as designed because applicable design requirements for equipment have not been afTected. It will retain its safety objective and will function as i

l designed during planned operations, transients, accidents, and special events. No new type of malfunction is intrtduced for the Emergency AC Power System by this USAR change. The subject relay is not listed in Technical Specifications. This LCR does not change the accuracy or response characteristics of the l

undervoltage relays and their auxiliary relays such that it will alrect the margin of safety of the Auxiliary l

Electrical System as outlined in any Technical Specification Bases.

I CR 94-0155 TITLE: USAR Change - Multi-Purpose Facility (MPF) Vent Monitor I

DESCRIPTION: This LCR revised the instrument range noble gas upper value for the MPF Ventilation Radiation Monitoring System from 10'to 102 Ci/ml. This change brings the USAR value into agreement with the ,

associated equipment vendor manual and reference points as required by NUREG 0737 (Xc-133 l reference). This change does not afTect approved plant design requirements.

SAFETY l

l EVALUATION: This LCR has no efTect on actions / scenarios which initiate plant events. No new equipment or control '

functions are added and no alarm setpoints are impacted. This USAR change does not alTect the way in which plant equipment is currently designed, operated, or maintained. The existing plant procedure already utilizes the correct value for calibration of the radiation monitor. The change does not affect instrument setpoints or the ability of the equipment to alert Operations personnel in the event of an i

elevated activity level in the MPF ventilation exhaust. The change has no adverse effect on equipment l l '

' required in support of plant events to prevent the release of radiation to the public. This LCR has no effect on equipment attributes associated with malfunctions and failures. The afTected component will continue to fulfill the design function of monitoring, recording, and alarming for the associated MPF ventilation exhaust system. This change is not associated with the addition or modification of any plant equipment. Equipment functions remain consistent with approved plant design und operation, and the change does not create any new equipment failure mcxles. This change does not alTect information/ requirements presented in the Offsite Dose Assessment Manual or Technical Specifications l and the associated Bases; therefore, there is no reduction in the margin of safety.

I LCR 95-0001 TITLE: USAR Change - Spent Fuel Pool Level Channels DESCRIPTION: The USAR was revised to remove the allowable repair time of one month for the Spent Fuel Pool Level Channels, FPC-LS-60A/B. The original repair time was based on analyses performed at the time of originallicense application and was included in the Nuclear Safety Operational Requirements section of the USAR. Currently CNS uses the Technical Specifications to determine operation, maintenance, and surveillance requirements. There was no associated Limiting Condition for Operation specified in any Technical Specification for these switches.

SAFETY l EVALUATION: Spent fuel pool water level is still recorded daily by Station Procedure 2.1. I 1, and the minimum fuel pool level is not impacted by this change. The change does not impact any radiation monitoring equipment or procedures used to minimize radiation / contamination exposure to plant personnel or the public. The

USAR requirement is that at least one (of two) spent fuel pool water level indicators must be operable l during planned operation. The statement removed by this USAR change will not alTect this requirement.

l The calibration frequency and the preventive maintenance requirements for the level switches are not j impacted. No physical work in the plant is required to implement this change. The maintenance frequency of the associated level switches is not impacted by this USAR change. The removal of the one month allowable repair time for the Spent Fuel Pool Level Channels does not affect the Technical Specification requirement that when irradiated fuel is stored in the spent fuel pool, the water level shall be recorded daily. The minimum fuel pool level as stated in the Technical Specifications is not impacted by this USAR change. This change does not reduce the margin of safety as defined in the basis fbr any Technical Specification.

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i LCR 95 0006 l

l TITLE: USAR Change - Recombiner Catalyst l

l DESCRIPTION: This LCR deleted an item from the design precaution section of the Augmented OITGas System failure  !

analysis which indicated that recombiner catalyst would be replaced every five years. The information that was deleted is associated with pre planned maintenance and is not regarded as a design precaution.

Replacement of the catalyst at fixed intervals is not required or performed if the catalyst continues to I function properly. CNS system design and procedural requirements ensure that indications of decreased performance of the recombiners will be detected and resolved.

I SAFETY EVALUATION: This change has no effect on actions / scenarios which initiate plant events. It does not alTect the way in j which plant equipment is currently designed, operated, or maintained. The current methods of monitoring I recombiner operation ensure the equipment performs properly and any degradation of the recombiner l catalyst would be obsermi and corrected. All equipment required in support of plant events will continue l

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' to function as designed to prevent the release of radiation to the public. This change deletes an j inappropriate statement associated with a design precaution which does not affect the failure analysis. '

l This change has no adverse effect on equipment attributes associated with malfunctions and failures. It is not associated with the addition or modification of any plant equipment. Equipment functions remain consistent with approved plant design and operation, and this change does not create any new equipment failure modes. It does not affect the information/ requirements presented in the Technical Specifications I and associated Bases; therefore, there is no reduction in the margin of safety. l 1 CR 95-0009 l

TITLE: USAR Change - Figure VII-4-10 DESCRIPTION: This LCR corrected several errors in USAR Figure VII-4-10, Typical Core Standby Cooling System and System Instrumentation. The USAR Figure was resised to correct inconsistencies between the figure and

) USAR text, and with approved plant design. Changes meluded: 1) deleted high drywell pressure signal  !

from the Automatic Depressurization System Trip System Actuation Logic diagram,2) deleted the Low j

Pressure Coolant Injection (LPCI) loop select and reactor low pressure signal from the LPCI Trip System j l Actuation Logic diagram, and 3) deleted reactor low pressure signal frorn the Core Spray System j Actuation Logic diagram.

SAFETY EVALUATION: The changes bring the figure into agreement with other USAR sections and with approved plant design.

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They have no efTect on actions /secnarios which initiate plant events. The changes correct intemal USAR j discrepancies and do not affect the way in which plant equipment is currently designed, operated, or I maintained. All equipment required in support of plant events will continue to function as designed to l prevent the release of radiation to the public. The changes have no effect on equipment attributes associated with malrunctions or failures. They have no effect on equipment redundancy or reliability and are not associated with the addition or modification of any plant equipment. Equipment functions remain consistent with approved plant design and operation, and the figure changes do not create any new equipment failure modes. The changes do not alTect the information/ requirements presented in the Technical Specifications and associated Bases; therefore, there is no reduction in the margin of safety.

l LCR 95-0022 l l

- TITLE: USAR Change 1leating, Ventilation, and Air Conditioning Systems j l

1 I l DESCRIPTION: This LCR made the following changes in Section X-10, Heating, Ventilation, and Air Conditioning Systems: 1) revised a discussion of buildings supplied with steam from the station heating system to say I steam is supplied to various buildings rather than all buildings except those listed, 2) revised description of Off Gas Building iIcating and Ventilating System to remove " gas fired heating coil" from the unit ,

description, and 3) revised secondary containment section to indicate that signals from the four

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instruments which sense pressure difference between outside air and the secondary containment are averaged to provide control actions, as opposed to the worst case value from the individual instruments.

This nyeraging approach compensates for local pressure variations due to expected wind conditions around the Reactor Building. Changes 1 and 2 are associated with non-essential heating and ventilating systems and were made to provide a more accurate description of the equipment involved and are consistent with existing plant design. Change 3 corrects a discrepancy between the original as-configured i plant design and the SAR, SAFETY EVALUATION: This LCR corrected discrepancies in USAR descriptions of various heating, ventilation, and air conditioning systems. The changes have no effect on plant event initiating actions or equipment response.

They do not affect the way in which plant equipment is currently designed, operated, or maintained. All equipment required in support ofplant events will continue to function as designed to prevent the release of radiation to the public. The changes have no effect on equipment attributes associated with I malfunctions and failures. They are consistent with approved plant design and are not associated with the addition or modification of any plant equipment. Equipment functions remain consistent with j approved plant design and operation, cnd the changes do not create any new equipment failure modes. I The changes have no efTect on information presented in the Technical Specifications or their Bases; )

therefore, there is no reduction in the margin of safety as defined in the basis of any Technical Specification.

LCR 95-0034 TITLE: USAR Change - Reactor Protection System (RPS) Logic Description

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DESCRIPTION: This LCR revised the discussion of the RPS logic from "The automatic logics of trip system A are logics A1 and A2; the manual logics B1, B2, and B3" to "The automatic logics of Trip Systems A and B are logics A1, A2, B1, and B2. The manuallogics are A3 and B3." This was an administrative change as the logics for the RPS are correctly described in a number of places in the USAR and are correctly shown i on USAR figures.  !

SAFETY )

EVALUATION: This change is a documentation correction to be consistent with the correct RPS trip logic. No activity 1 is being perfbnned that would increase the probability of a previously evaluated plant event. It does not l impact any radiation monitoring equipment er procedures used to minimize radiation / contamination I exposure to plant personnel or the public. No actual work on plant congponents or procedure changes I are involved in changing the wording in the USAR. Therefore, there is no increase in the probability of occurrence or consequences of a malfunction of equipment important to safety. Since this change does not affect the manner in which plant equipment is operated, controlled, or maintained, it does not create i the possibility of a new type of plant event or equipment malfunction. The RPS is described in the Technical Specifications, but this USAR change does not affect the Bases for the Limiting Conditions for Operation or the Surveillance Requirements. The descriptions of the RPS system in the Technical Specifications are not impacted by this change; therefore, there is no reduction m the mar gin of safety as defined in the basis for any Technical Specification.

I CR 95-0039 TITLE: USAR Change - Correction and Clarification of Discussion of Site External Flomhng DESCRIPTION: This LCR trvised USAR Section II-4 conceming site ficoding to correct errors, provide clarification, and incorporate commitments made in response to FSAR questions. The conunitments were submitted as part of SAR amendments but never formally included in the original USAR. As a result, the USAR discussion of site external flooding did not adequately reflect the CNS licensing basis. The changes provided clarification and additional details of site design and planned measures for mitigating the effects of an extemal flood. These revisions brought the USAR into agreement with the approved plant design and existing operating and emergency operating procedures.

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SAFETY EVALUATION: This revision provides a more detailed discussion of potential conditions associated with flooding scenarios, and the plant response to such conditions. The changes have no efTect on actions / scenarios which initiate plant events. They do not affect the way in which plant equipment is currently designed, operated or maintained. They clarify the measures to be taken by plant stafT to ensure safety of die plant and the public in the event of a serious flooding condition. All equipment required in support of plant events will continue to function as designed to prevent the release of radiation to the public. The changes do not indicate any inadequacies in the existing plant design which would affect equipment attributes associated with failures and malfunctions. The USAR changes are not associated with the addition or modification of any plant equipment, and no new types of plant events or equipment malfunctions are introducal The changes do not afrect information/ requirements presented in the Technical Specifications or associated Bases. Therefore, there is no reduction in theinargin of safety as defined in the basis for any Technical Specification.

I.CR 95-0049 TITLE: USAR Change - Control Building Basement Flooding DESCRIPTION: This LCR clarified a statement associated with automatic actions which occur in the event of Control Building basement flooding. The resised statement reads,"Sersice Water Pumps Cross-tie Valve SW-MO-37MV is automatically closed and there is an alarm in the Control Room on water in the Control -

Building basement." The USAR previously stated that the two headers feeding Service Water to the plant have been modifiedto be automatically isolated and there will be an alarm in the Control Room on water level in the basement. This change clarified that only SW-MO-37MV closes on Control Buildmg basement flooding. This corrects an inconsistency within the USAR and with approved plant design.

SAFETY EVALUATION: This change provides consistency with other USAR sections and approved plant design, and has no effect on actions / scenarios which initiate plant events. It does not alTect the way in which plant equipment is currently designed, operated, or maintained. All equipment required in support of plant events will continue to function as designed to prevent the release of radiation to the public. The resised statement brings the USAR text into compliance with failure analysis provided in the Sersice Water System Design Criteria Document, which indicates that the existing plant design provides the necessary redundancy, l diversity, and separation requirements for the Control Building basement flooding scenario. There is no

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increase in the probability of occurrence or consequences of a malfunction of equipment important to  :

safety. This change is not associated with the addition or modification of any plant equipment. ,

Equipment functions remain consistent with approved plant design and operation, and the changes do not I create any new equipment failure modes. This change does not afTect the information/ requirements presemed in the Technical Specifications and the associated Bases; therefore, there is no reduction in the margin of safety.

I CR 96-0012 TITLE: USAR Change - Figure X-10-3 DESCRIPTION: 'Ihe LCR revised USAR Figure X-10-3, Flow Diagram of Radwaste Building iIcating and Ventilation (ll&V) System, to add two manual balancing dampers physically located downstream of fans llV-FAN-(EF-RW-1C) and IIV FAN-(EF-RW-1D). The drawing discrepancy was identified during activities to verify the as-built condition of the plant. The position of the dampers is controlled by engineering evaluation based on system flow requirements and is not changed during plant operation.

SAFETY EVALUATION: System operation is not changed from that reflected in the existing USAR description. The Radwaste Building II&V System components are not accident initiators and the system is not required to mitigate the consequences of an accident; therefore, this activity has no impact on the probability of occurrence or consequences of an accident described in the USAR. The radiation contamination function of the Radwaste Building 1I&V System as described in the USAR is not alTected. Since the dampers conform I

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to the system design requirements, the probability of a malfunction of equipment important to safety is l unchanged. As a failure of the dampers is bounded by the loss of the Radwaste Building II&V System, 1

this activity does not create the potential for an accident or malfunction ditTerent than previously evaluated. The Radwaste Building II&V System is not included in any Technical Specifications. This 4 activity does not reduce the ability of the Radwaste Building II&V system to maintain its minimum negative pressure requirement described in the USAR.

I UCR 97-100 l

l TITLE: USAR Change - Figure X-10-3 l

DESCRIPTION: This UCR revised the drawing of the Radwaste Building IIcating and Ventilation (II&V) System to i remove the notation " trip fans on low temp" from component IIV-TC-1014A, the freezestat for the

' Radwaste Building Il&V System supply. This change brings the figure into alignment with the current as-built operation of the Radwaste supply fans. '

SAFETY l EVALUATION: No change is being made to the Radwaste Building II&V System. The Radwaste supply fans are not

! required for the safe shutdown of the plant. Neither the Radwaste Building temperature nor the location of the temperature element is considered in any plant event evaluation. Procedural controls are currently in place to take actions if a low system temperature alarm is received from iIV-TC-1014 A. Freezestat i ilV-TC-1014A is not used as input to any radiation monitoring equipment and it is not located near any radiation monitoring equipment. It is not used to minimize the radiation exposure of personnel or equipment and is not used to minimize the spread of contamination. Therefore, this change does not l increase the probability of occurrence or consequences of a plant event previously evaluated in the ,

USAR. IIV-TC-1014A is used to provide an alarm function only. This UCR does not result in any I changes to any plant system, component, structure, or plant procedures therefore, it will not increase the probability ofoccurrence of a malfunction of equipment important to safety. This change will not impact the radiation exposure of personnel within or at the site boundaries as a result of a taalfunction of equipment important to safety. The operation of freezestat iIV TC-1014 A is not required for any safety-l related functions, and there are no automatic functions associated with IIV-TC-1014 A. It does not create the possibility of any new plant events or malfunctions of equipment important to safety. Neither the I

function of this component nor the function of the Radwaste ventilation system arc discussed in any Technical Specification or the basis for any Technical Specification. The Radwaste Ventilation Radiation Monitoring System is discussed in the Technical Specifications for Post Accident Monitoring, but these radiation monito ring functions are not impacted by this change. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

UCR 97-125 TFfLE: USAR Change - USAR Figure XI-7-1 l

DESCRIPTION: This UCR revised Figure XI-7-1 (Burns and Roe Drawing 2035, Sheet 2) to change valve number CF.

I 586 to CF-602 to reflect the as-built configuration. CF-602 is the Solution A Tank Drain Valve and CF-586 is the Condensate Supply to Augmented Radwaste Drain Shutoft Both valves were inadvertently identified as CF-586 on the subject drawing. In addition, CF-602 was also incorrectly identified as CF-586 in system operating procedures and the actual CF-586 valve was not listed in any system operating procedure. The alTected procedures were revised, concurrent with this activity, to correct the valve l number for CF-602 and to add CF-586 to the valve lineup checklist. Both manually operated valves are l part of the Condensate Filter Demineralizer System, a nonessential system not relied upon for any safety-related application, ar.d are maintained in the normally closed position.

SAFETY l EVALUATION: The purpose and function of the subject valves is not being changed. No physical changes to plant components are associated with this change. This portion of the Condensate Filter Demineralizer system is not assmiated with any plant events discussed in the SAR and is not used to minimize the consequences of any plant event. The related procedure changes will not change the normal position of valves CF-586

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and CF-602. The manner and sequence in which these valves are manipulated is not impacted. No equipment is added or modified. The function / design operation of the Condensate Filter Demineralizer system is not impacted by the valve number changes or procedure changes. Therefore, the probability ofoccurrence or consequences of a malfunction of equipment important to safety are not increased and

! no new types of plant events or equipment malfunctions are introduced. The Condensate Filter Demineralizer system is not discussed in the Technical Specifications and is not required to safely shutdown the reactor and maintain it in a safe shutdown condition; therefore, this UCR does not reduce the margin of safety as defined in the basis for any Technical Specification.

UCR 97-136 TITLE: USAR Change - Figure IV-7-2C DESCRIPTION: This UCR corrected a drafling error on Figure IV-7-2C (General Electric Functional Control Diagram for the Reactor Core Isolation Cooling (RCIC) System - Drawing 729E517BC, Sheet 3) that mislabeled I the Control Room panel number for the mode selector switch for the Residual Ileat Removal (RIIR) lleat i Exchanger level control circuit. The panel number was incorrectly identified as 9-4 versus 9-3. The panel number was verified during plant as-built verification activities.

SAFETY EVALUATION: This drawing change to correct a drafting error in the General Electric Functional Control Diagram will not afTect the design function of the RCIC system. There will be no physical or operational changes in the plant or plant procedures impacted by this change. OfTsite dose will not be affected. Therefore, there is no increase in the probability of occurrence or consequences of a plant event or malfunction of equipment important to safety previously evaluated in the SAR and no new types of plant events or equipment malfunctions are introduced. The panel number and operation of the mode selector switch for RIIR IIcat Exchanger level control an: not specifically addressed in the Technical Specifications and there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

l UCR 97-165 1

TITLE: USAR Change - Clarification ofliigh Pressure Coolant Injection (IIPCI) Pump Net Positive Suction Ilead (NPSH) Requirements DESCRIPTION: This UCR clarified the maximum suppression pool temperature at which adequate NPSil is available to l theIIPCIpump. Based on previously analyzed suppression pool temperature response during transient

) and accident conditions, the temperature was raised from 140"F to 170*F. In addition, discussion was

! added relative to the maximum expected pool temperature during transient / accident conditions and the expected pool temperatures when the IIPCI system is required to operate. Finally, a reference to Nuclear Engineering Department Calculation 92-100 was added to support these changes.

SAFETY EVALUA'IION: This change updates the IIPCI operating conditions and the suppression pool temperature response to be consistent with the assumptions of the General Electric transient and accident analysis. Presious analyses demonstrated the ability of the IIPCI system to perfbrm its intended function at increased suppression pool temperatures. The performance of the !IPCI system has not been changed, therefore, there is no increase in the probability of a malfunction of equipment important to safety and no increase in the consequences of an accident or equipment malfunction. The change does not involve any physical or procedural changes to any system, structure or component, or any changes to the method of plant operation; therefore, no new types of accidents ofequipment malfunctions are introduced. As this activity documents the ability of the IIPCI system to perform its intended function at increased suppression pool temperatures, there is no reduction in the margin of safety as defined in the basis for any Technical l Specification.

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UCR 97-169. UCR 98-071 TITLE: USAR Changes - Organizational Structure and Responsibilities DESCRIPTION: These UCRs made revisions to the USAR to reficct numerous organizational changes required to support both the short and long term goals for improved station performance. The changes varied from simple title changes to better describe positional responsibilities, to the restructuring of entire departments and divisions within the nuclear organization.

SAFETY EVALUATION: These changes do not adversely affect any programs, processes, procedures, or actisities which relate to nuclear safety. There are no configuration changes or changes to safety equipment or procedures used to mitigate the consequences of an accident. There are no changes to the maintenance or operation of any equipment important to safety. All of the programs intended to protect the margin of safety remain intact.

UCR 97-172 TITLE: USAR Change - Main Turbine Overspeed Trip Mechanism DESCRIPTION: This UCR revised the operational testing frequency of the main turbine overspeed trip mechanism from i "in accordance with provisions of the vendor's instruction manual" (i c., once every six months) to once l

per cycle. This change does not add any new equipment or change the operational trip testing process i or specified trip settings.

SAFETY EVALUATION: A postulated failure mechanism would be the varnishing of dynamic mechanical overspeed trip mechanism components due to the reduced cycling of the mechanism. This postulated varnishing would create binding or sluggish operation and therefore increase the actual trip above the 108% setpoint. Trip testing of the mechanical overspeed trip has been conducted at the once per cycle frequency for at least the last ten years and a review of the operational testing and functional testing history indicates that the postulated varnishing failure mechanism has not occurred. The mechanical overspeed trip mechanism has consistently tripped the main turbine at 1922 rpm (or 107%). Therefore, actual test data indicates that changing the test frequency will not increase the probability of a previously evaluated overspeed event. Quarterly performance of functional trip testing will continue to ensure that degradation of the mechanical overspeed trip mechanism is detected, thus the probability of occurrence of an accident or a malfunction ofequipment important to safety is not increased. Since any event caused by a failure of the l mechanical overspeed trip is bounded by the turbine overspeed event with missile generation, the consequences of an accident or equipment malfunction are not increased. As this change does not add any new equipment or change the operational trip testing process, the possibility of an accident or malfunction of a different type than presiously evaluated is not introduced. Finally, the turbine ourspeed margin of safety is not relied upon in the basis for any Technical Specification; therefore, the margin of safety as defined in the basis of any Technical Specification is not reduced.

UCR 97-173 TITLE: USAR Change - Increase in Spent Fuel Shipping Cask Lin Height DESCRIPTION: This UCR revised the maximum spent fuel shipping cask lin height from nine inches to sixteen inches.

The previously specified value of nine inches did not take into account the pedestal stud bolts which protrudc 61/4 inches above the floor elevation. The increase in maximum lin height will eliminate the potential interference, thus providing for additional safety when the cask is being moved. The Spent Fuel

! Cask Drop Analysis assumes a five-foot drop. Then: fore, the increase in maximum lift height established I

by this activity has no impact on safety.

SAIETY EVALUATION: This activity does not involve any equipment or systems credited as event initiators or adversely atTect any equipment or systems credited with mitigating the consequences of a plant event. Changing the vertical lin from nine inches to sixteen inches is bounded by the Spent Fuel Cask Drop Analysis and the l

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l probability is not increased. Lilling the cask an additional seven inches with an approved lifting device does not affect the probability of occurrence of a malfunction of equipment important to safety. This l activity does not affect the plant's ability to contain radioactive materials and will not result in increased I radiological consequences. No new types of plant events or equipment malfunctions are introduced since l

the Spent Fuel Cask Drop Analysis is bounding. This activity does not afTect the ability to shut down the reactor or to maintain it in a cold shutdown condition. As the maximum cask lift height is not addressed in the Technical Specifications or associated bases and no equipment functions are impacted, the margin of safety as defined in the basis for any Technical Specification is not reduced.

f UCR 97-177 l '

TITLE: USAR Change - Drywell Coating DESCRIPTION: This UCR revised the USAR to clarify and generalize the details associated with the protective coatings applied to the drywell. Details originally put in the USAR, such as the product name, manufacturer, and test lab name, have been removed or replaced with descriptors that better characterize the coating l material composition and requirements. The actual coating materials and performance requirements are unchanged by this activity.

SAFETY EVALUATION: This activity does not afTect any equipment or systems credited as event initiators or with mitigating the consequences of a plant event . The coating systems have been shown to satisfactorily withstand the temperatures and pressures of the steam environment postulated during a design basis loss-of-coolant accident for the drywell and wetwell. They have been qualified and have demonstrated that they are capable of surviving an accident without degradation that could result in detachment of the coating and subsequent impediment of the coolant flow path. This activity will not afTect the ability to shut down the reactor or to contain radioactive materials. No new types of plant events or equipment malfunctions are introduced by this USAR change. The margin of safety as defined in the basis for any Technical Specification is not cfrected since no equipment functions are impacted.

UCR 98-003 TITLE: USAR Change - Figure VIII-6-1 DESCRIPTION: This UCR revised USAR Figure Vill-6-1, DC One Line Diagram, to correct the component identification cale (CIC) for the Manual Transfer Switch EE-SW-LXTX (125) to Battery Charger 1C. The CIC w as l

correctly identified in the installing Design Change 87-73 (Replacement of 125VDC Batteries, Racks, and Chargers), but was incorrectly transferred to Burns and Roc Drawing 3058, the source drawing for the USAR figure.

SAFETY EVALUATION: This drawing change does not install, remove, replace, or modify any equipment installed and previously I

analymi under the original Design Change 87-73. It will not affect nuclear safety in a way not previously j evaluated in the USAR. The subject component performs and operates in the same manner as the component replaced and has been evaluated in the USAR. It performs the same safety function and intedaces with other plant systems in the same manner as evaluated in the USAR. The original quality star.dards and design criteria are met or exceeded. Therefore, there is no increase in the probability of occurrence or consequences of a plant event or malfunction of equipment important to safety as a result of this drawing error, and no new types of plant events or equipment malfunctions are introduced. This component will recharge the 125VDC system as required by the Technical Specifications and will not reduce the margin of safety as defined in the basis for any Technical Specification.

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UCR 98-004 DCNs 98-0133 and 98-0134 TITLE: Changes to USAR Figures IV-7-1 and VII-4 Burns & Roe Drawings 2043 and 2044 DESCRIPTION: Burns & Roe Drawing 2043 is the flow diagram for the Reactor Core Isolation Cooling (RCIC) system and Burns & Roe Drawing 2044 is the flow diagram for the fligh Pressure Coolant Injection (1IPCI) system. Subsequent to initiation of this UCR, these drawings were incorporated by reference into the 1

USAR. These drawings were revised to show changes to the Inservice Inspection (ISI) boundary flags.

l On Drawing 2043, the flag was moved from RCIC-MOV-MO33 to MO30. On Drawing 2044, the flag at HPCI-CV-18CV showed t. Class 2 to Class I boundary. It was corrected to show a Class 2 to Non-Code boundary. There are no changes to the design, function, or operation of these valves.

SAFETY EVALUATION: ISI is not a precursor to any accident or transient evaluated in the SAR and plays no role in accident mitigation. The boundary flags do not afTect the design or operation of any system, structure, or component. Therefore, there is no increase in the probability of occurrence of a previously evaluated plant event or malfunction ofequipment important to safety, there is no bearing on the dose consequences of any plant event or equipment malfunction, and no new types of plant events or equipment malfunctions are created. Although ISI is required by the Technical Specifications,it is not credited with providing any safety margin; therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

UCR 98-007 TITLE: USAR Change - Pressure Maintenance Systems for the Residual Ileat Removal (REIR) and Core Spray (CS) Systems DESCIUPTION: This change request revised the USAR to clarify the normal and backup pressure maintenance systems for the RHR and CS systems. As previously written, the USAR inferred that the Reactor Building Auxiliary Condensate pmnp was the primary source of pressure maintenance. This revision clearly identifies the Main Condensate system as the primary source of pressure maintenance and the Reactor Building Auxiliary Condensate system as the backup source. This clarification has no impact on plant configuration or operation.

SAFETY E s'ALUATION: This clarification aligns with existing operating procedures, design criteria documentation, and plant design. There are no accidents or transients evaluated in the USAR that are afTected by this activity and it will not cause the system to be operated outside its design basis. The pressure maintenance function for RIIR and CS is not being changed or eliminated, and system performance is unchanged. This resision does not change, degrade, or prevent actions described or assumed in plant events evaluated in the USAR; therefore, it does not increase the radiological consequences of a plant event. This actisity does not impact any safety related equipment. It does not change the manner in which the RHR and CS systems are maintained in standby or the manner in which they are operated. No new types of plant events, or failure modes which could result in a different type of malfunction, are created. There is no reduction in the margin of safety as defined in the Technical Specification Bases becuuse this activity does not change the requirement to establish pressure maintenance to support RHR and CS operability.

UCR 98-011

UCR 98-012 1

! TrfLE: USAR Changes - Figures III-9-5 and IV-7-2B 1

l DESCRIPTION: These UCRs incorporated drawing changes to the functional control diagrams for the Reactor Core Isolation Cooling and Standby Liquid Control systems to reflect the as-built status ofindicating lights on Control Room Panels 9-4 and 9-5 aller completion of the Detailed Control Room Design Review (DCRDR) modification. The required drawing changes were not all completed at the time of the modification.

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r SAFETY EVALUATION: The DCRDR modification upgraded the control room panels in order to make the displays and controls more human factored. These changes do not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety. No functional modifications were made to any of the atrected systems and all other design criteria are in accordance with governing codes, standards, and practices. Modifications were accomplished within existing Technical Specification limitations and there was no reduction in the margin of safety.

UCR 98-013 TITLE: USAR Change - Seismic Criteria for "Z" Sump DESCRIPTION: This UCR reversed changes made to Section XII that were incorrectly made by previous UCR 97-003.

UCR 97-003 was generated as a result of Design Change (DC)95-033," Sump Z Modification for Standby Gas Treatment (SGT) Operability," wherein the safety evaluation identified that the equipment necessary to ensure operation of the "Z" sump had been upgraded to meet essential requirements with the exception of seismic requirements. The safety evaluation stated that cot.pling a Design Basis Accident with a Design Basis Earthquake is not a credible scenario and, hence, the "Z" sump equipment required to support SGT operation was not required to be scismic class IS. This position was reevaluated and found to be non-conservative. As a result, the USAR was changed back to the original wording under this UCR. The scismic classification concems of the "Z" sump equipment were resolved by Modification Package (MP)97-100.

SAFETY EVALUATION: Since the USAR is being restored to the original, more conservative wording, this activity does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This activity does not authorize the change of any seismic classification or the change of any plant equipment. This change ensures that the USAR does not allow a change to the facility based upon the exception to the seismic classification and preserves the integrity of the seismic classification definitions. There is no reduction in the margin of safety because this change ensures that structures and equipment will be evaluated to the proper seismic classification.

UCR 98-014 Procedure 2.2.31. Revision 3 TITLE: Provision for Circulating Water Screen Wash System and Traveling Screen Out-of-sersice Time to Allow  !

for On-line Maintenance '

DESCRIPTION: The USAR was revised to explicitly allow the Circulating Water spray wash assembly and/or the traveling screens to be shut off for short periods of time (s 7 days) for maintenance while the unit is still in operation. While the USAR did not preclude this operating configuration, it did not presiously provide

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for it either. In addition to revising the USAR, Procedure 2.2.3.1, " Traveling Screen, Screen Wash, and Sparger System," was revised to recognize the same provisions and limitations. j EVALUATION: The Circulating Water system, the Sersice Water system, and the backup fire water pump will still bc l functioning to support the performance of the required safety functions of the systems. Discontinuing spray wash and turning oft the traveling screens for a maximum of 7 days does not increase the probability of a plant event or impact the boundaries or barriers to fission product release. There are no accident initiators associated with this equipment and the spray wash and traveling screens are not required to respond to an accident. Spray wash will be returned to sersice aller the maintenance activity is completed. This change does not intrcxluce any new accident initiators or precursors. No credit is taken for the system in any malfunction of equipment important to safety. The spray wash and traveling screens are not taken credit for in any accident analyses or Technical Specifications, thercibre, the margin of safety is not reduced.

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I UCR 98-019  ?

TITLE: USAR Change - Residual IIcat Removal (RIIR) Senice Water Booster Pump Fan Coil Unit

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1 DESCRIPTION: This UCR clarified statements regarding the fan wil unit in the Service Water " Power Generation Objective" and " Power Generation Design Basis" subsections. It also revised the discussion of the fan coil unit itself. The USAR previously indicated that Senice Water was the only cooling source to the fan coil unit and that its operation was expected in support of the Power Generation functions of the RIIR Senice Water Booster Pumps. This design criteria was changed in 1973 out of concerns for silting in i Senice Water process piping. Specifically, the USAR was revised to state that the IUIR Senice Water Booster Pump Fan Coil Unit is normally cooled by a closed cooling loop recirculating to a cooling tower.

Senice Water provides a backup cooling supply to the R1IR Senice Water Booster Pump Fan Coil Unit to allow operational flexibility for room cooling if the recirculation loop is unavailable. USAR text was also clarified to better describe the fan coil unit and recirculation loop power supply.

SAFETY .

EVALUATION: Normal RIIR Senice Water Booster Pump room cooling is provided by Control Building IIVAC, not the fan coil unit. The fan coil unit's backup function is not in any way degraded as a result of altering its ;

nonnal cooling supply. Therefore, there are no adverse effects on any precursors to events described in j the SAR and the consequences of plant events are not increased. Making Senice Water a backup fan coil  !

unit cooling supply provides a net safety benefit since there will be increased Senice Water flow to

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remaining essential k> ads. FC-C-1 A is the only "important to safety" equipment afTected by this change.

Nuclear Engineering Department Calculation 92-063 Revision I shows that FC-C-1 A as currently l designed will remove the expected heat load from the RIIR Senice Water Booster Pump room. In  !

addition, eliminating Senice Water as the primary cooling source to the fan coil unit resolves the issue

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of silt-induced degradation of the fan coil unit's function. A malfunction of the fan coil unit will have l

identical consequences whether stemming from a Senice Water-induced failure or a recirculation loop failure. There are no plant events discussed in the USAR that lead to a loss of RilR Senice Water i Booster Pump Room IIVAC. IIowever, the changes made by this UCR do not create any additional i

possibilities. The margin of safety is not reduced by eliminating the safety support function of the fan coil i unit to the operability of the Senice Water Booster Pumps.

UCR 98-025 1

TITLE: USAR Change - Residual Ileat Removal (RIIR) Inboard Injection Valve Automatic Isolations i

DESCRIPTION: This UCR revised Table VII-3-1, Process Pipeline Penetrations with Automatic Isolation Functions for Primary Containment, to correct information regarding the automatic isolation of the RIIR inboard I injection valves, and the conditions under which the isolation takes precedence over Low Pressure Coolant Injection (LPCI) system injection. It specifically added isolation signal "F" to R1IR LPCI Return Valve, Penetrations X-13A/B, MO Gate Valve, and revised the associated Note 17 accordingly and clarified that all of the stipulated conditions (both RIIR Shutdown Cooling supply valves not fully closed and reactor pressure below 75 psig) must bs, met for the valves to automatically isolate. The change reflects the as-built and original configuration of this logic arrangement, and corrects an error carried forward from the original FSAR. Further, the change brings Table Vll-3-1 into agreement with USAR j Figure VII-4-7b (also included in the FSAR) which complies with the system design requirements and, hence, correctly illustrates the functional control diagram for the IU IR system.

SAFETY EVALUATION: This activity revised Table Vil-3-1 to reficct the original and as-built design configuration of the IU IR I and Primary Containment Isolation (PCIS) systems. This logic is used to configure the LPCI inboard i injection valves in response to a plant event, but is not an event initiator. The ability of the RIIR and PCIS systems to mitigate the consequences of a plant event is unchanged from the original design. No actual change to component testmg, status, or maintenance is involved. In addition, this change does not involve any physical change to plant hardware. No change to system operating or emergency procedures results from this change. As there is no change to actual operation of the subject components, no new types of equipment malfunctions are introduced. This change is consistent with the design information

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! related to Shutdown Cooling isolation in response to a Group 2 isolation as specified in the Technical (

Specifications, therefore, there is no reduction in the margin of safety as defined in the basis fbr any j Technical Specification.

UCR 98-028 TITLE: USAR Change - Turbine Building Ver 41ation Radiation Monitoring System Description DESCRIPTION: The USAR was revised to clarify the fact that the Turbine Building Ventilation Radiation Monitoring System utilizes four sample points and not one as suggested by the original text. The system utilizes one sample point for each of the four Turbine Building exhaust ventilation ducts. This UCR also added discussion to describe the interlock between fan operation and the respective sample line root valve, which cnsums that effluent monitoring occurs for those exhaust fans / ducts actually in service. This USAR correction is consistent with the existing plant configuration and original design.

SAFETY EVALUATION: The methodology used to sample plant efiluents for normal and post-accident ellluent monitoring has no bearing on any plant event initiators. Operator actions to mitigate plant events with oft-site dose are predicated in part on accurate indication of efiluent release rates. This change more accurately desenbes the existing sampling methodology. It does not involve any hardware or configuration changes.

Therefore, the probability of occurrence or consequences of a malfunction of equipment important to ,

safety are not increased. This USAR correction is consistent with original equipment design and the operation of the Turbine Building Ventilation Radiation Monitoring Sampling System remains a unchanged. Failure modes and efTects of the equipment associated with the system are not affected and I no new types of accidents or malfunctions are introduced. This change accurately describes the design ]

provisions in place to ensure the accuracy of the Turbine Building ellluent monitoring sampling system,  !

which is an implicit assumption in the Technical Specification bases for ellluent monitoring. l i

UCR 98-029 TITLE: USAR Change - Removal of Pendant Control of the Reactor Building Crane DESCRIPTION: This UCR revised the USAR to climinate reference to the pendant control feature of the Reactor Building crane. Pendant control was removed by Minor Design Change (MDC) 77-56, making the main cab the only point ofcontrol. While the FSAR to USAR conversion captured this MDC in Section X-4.10.2, it failed to update the text in Sections X-4.10.3 and X-4.10.4. As originally installed, the Reactor Building crane deviated from Branch Technical Position (BTP) APCSB 9-1," Overhead Ilandling Systems for Nuclear Power Plants" The MDC served to bring the design into compliance with APCSB 9-1.

Although originally evaluated under the MDC, a supplemental safety review was done as part of this UCR.

SAFETY EVALUATION: The event ofinterest is a Spent Fuel Cask drop event. The potential for this event is minimized by enforcing the Restricted Mode of crane operation. Shifting control of this mode fmm the pendant to the main cab will have no effect on the probability of this event. Adherence to the intent of BTP APCSB 9-1 provides an acceptably low probability of a Spent Fuel Cask Drop event. The Reactor Building crane is considered to be equipment inrportant to safety when operated in the Restricted Mode. Control of this operating mode from the main cab instead of the pendant has no efTect on the reliability of the crane when operated in the Restricted Mode and does not affect the consequences of equipment malfunction. A Spent Fuel Cask drop is the only event discussed in the USAR associated with the Reactor Building crane.

Removal of the pendant does not create additional propensities for dropping other loads or for malfunction of the control circuit. Compliance with the BTP, along with the fact that the Restricted Mode l of operation is unchanged, ensures that the margin of safety as defined in the basis for any Technical l Specification is unchanged.

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I UCR 98-032 j TITLE: ' USAR Change - Figure IV-7-1 DESCRIPTION:. This UCR revised Figure IV-7 1 to change the valve numbers for MS-40 and MS-41 to RCIC-40 and >

RCIC-41, respectively. This change makes the figure consistent with the as-built configuration of the plant, the Equipment Data File, and plant procedures. The subject valves are clearly part of the Reactor I Core Isolation Cooling (RCIC) system and valve numbers MS-40 and MS-41 are assigned elsewhere in the Main Steam (MS) system.

SAFETY EVALUATION: ne changes will enhance the drawing by providing correct valve number information and will not alTect the design function of the RCIC system. There will be no physical or operational changes to the plant or plant procedures as a result of this UCR. Plant personnel or offsite dose will not be afTected. As no equipment will be added or modified, there is no increase in the probability of occurrence or consequences of a malfunction of equipment important to safety, and no new types of plant events or equipment malfunctions are created. Technical Specification discussion of RCIC testing and reactor overfill protection is not impacted, thus there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

UCR 98-033 l TITLE: USAR Change - Figure IX-S 1 DESCRIPTION: This UCR revised Figure IX-5-1, Sheet 2, Flow Diagram of the Augmented Off Gas (AOG) System, to identify two radiation monitors in the AOO system as " abandoned in place " The monitors were installed by the vendor under the original contract during the initial startup of the AOO system to verify proper system operation. A design review shows that the monitors have no control function and have never been added to the Equipment Data File. A historical review revealed that the units have always been valved out of service and have never had preventive or corrective maintenance. From an operational perspective, l

the position of the inlet and outlet valves for the radiation monitors is procedurally controlled in the closed '

position. These radiation monitors are not used to monitor the OIT Gas (OG) stream because the l performance of the OG system is measured at the Steam Jet Air Ejectors and the Elevated Release Point.

There were no changes to the plant configuration or operating procedures as a result ofclassifying the monitors as abandoned in place.

EVALUATION: This change is only a drawing enhancement to indicate that the two radiation inonitors are not used.

There will be no changes to procedures, components, or the operation of the plant as a result of this drawing change. This change does not add any equipment nor will there be any physical changes to the plant. Plant personnel and offsite dose will not be affected by adding the note to indicate that the monitors are no longer used. OG system monitoring will not be affected by this change. As this change will not affect the way in which plant equipment is currently designed, operated, tested, controlled, or maintained, l there is no increase in the probability of occurrence or consequences of a malfunction of equipment important to safety, and no new types ofplant events or equipment malfunctions are introduced. The 00 radiation monitoring process will not be affected by this drawing change. There is no efTect on the Technical Specifications and their bases and no reduction in the margin of safety as a result of this change.

UCR 98-034 TITLE: USAR Change - Figure X-12-1 DESCRIPTION: This UCR revised Figure X-12-1, Sheet Ia. Flow Diagram for the Instrument Air (IA) System in the Control and Turbine Buildings, to add valves IA-268 and IA-2008 to reflect the as-built configuration of the plant. The valves are the high and low side drain valves for IA-DPIS-605, the Dyer Outlet Filter Iligh Differential Pressure Alarm. They are K" globe valves, functionally closed, and non-safety related.

The valves were installed under Design Change (DC)90-161, " Instrument Air Dryer System Upgrade,"

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but not added to the USAR figure. There were no changes to the plant configuration or operation as a result of this activity to add the subject valves to the figure.

SAFETY EVALUATION: The change to the flow diagram is in accordance with field as-built conditions per DC 90-161. This change to provide missing information on the drawing will not affect the design function of the IA system.

There will be no physical or operational changes in the plant and no plant procedures are affected by this drawing change. Plant personnel or offsite dose will not be alTected. Sise no equipment is being edded or modified, there is no increase in the probability of occurrence of a malfunction of equipment important to safety, and no new types of plant events or equipment malfunctions are introduced The IA system is not discussed in the Technical Specifications and the system is not used to mitigate the consequences of an accident. It is not required to safely shut down the reactor and maintain it in a safe condition.

Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

UCR 98-035 TITLE: USAR Change - Figure XI-5-1 DESCRIPTION: This UCR revised Figure XI-5-1, Sheet 2 (Flow Diagram for the Main, Exhaust, and Auxiliary Steam Systems) to add various vent, drain, and shutoff valves associated with the main turbine first stage instrumentation. The valves were installed during original plant construction and were added to the drawing to reflect the as-built configuration ofthe plant. The valves are listed in the Equipment Data File and System Operating Procedures. There were no changes to the plant configuration or operation as a result of this activity to add the subject valves to the figure.

SAFETY EVALUATION: 'Ihe changes to the flow diagram will enhance the drawing by showing the existing instrument valves and will not atTect the safety or design function of the Turbine Bypass System, the Reactor Protection System, Turbine Protection Devices, Feedwater Control System, or the Turbine Control System. There will be no physical or operational changes in the plant and plant procedures are not affected. Plant personnel or offsite dose will not be alTected by these changes. As no equipment is being added or modified, there is no increase in the probability of a malfunction of equipment important to safety and no new types of plant events or equipment malfunctions are introduced. The drawing changes do not alter the function or operation of any instrument; therefore, there is no reduction in the margin of safety.

UCR 98-040 TITLE: USAR Change - Gland Steam Backup Capabilities DESCRIPTION: This UCR revised the function of the auxilian steam boilers with respect to their role in providing emergency backup steam to the Gland Steam System. The original USAR text implied that the auxiliary steam boilers would be available continuously while the main turbine is in operation. The USAR text was revised under this UCR to allow the auxilian steam boilers to be removed from senice for maintenance and during the summer months when auxiliary steam is no longer needed. An evaluation '

conducted by the main turbine manufacturer assuming the worst case situation (i.e., no gland steam or auxiliary steam for an assumed 434 startups and shutdowns during the 40 year rated life) concluded that an estimated 21.1% ofrotor life through the end of the plant life would be expended assuming worst case conditions. Since gland steam is always available during startups and is only lost due to the closure of the MSIVs, this evaluation is censidered to be very conservative. The study concluded that under worst case conditions, the material will remain within design requirements and be capable of performing its intended function for the expected life of the plant.

In addition, USAR Figure X-10-1 A was revised to remove pressure switch PS-808. The interlocks for the gland steam backup supply were originally intended to be controlled by a pressure switch. This switch was never installed during plant construction. Instead, a check valve was installed which sen es as a protection against backflow from the gland scaling steam line. When the pressure of the gland

sealing steam degrades below that of the auxiliary steam boilers, the check valve will pennit the passage of the steam to support gland sealing.

SAFETY EVALUATION: The auxiliary steam boilers are not an initiator of any plant event nor are they required Ibr the safe shutdown of the plant. Removal of the auxiliary steam boilers as a backup steam supply will not increaf e the frequency of any plant event associated with the Gland Scaling System. Offsite dose will not he affected and remains txxnided by the loss ofcondenser vacuum. The manufacturer's study conservatively estimated that the turbine rotor material will remain within design requirements and be capable of perfonning its intended function for the expected life of the plant without gland scaling steam. The Gland Scaling System is not a system important to safety. It is not required to mitigate the consequences of an accident. This activity makes no physical changes to any component in the Gland Sealing System or auxiliary steam boilers and all equipment remains within design requirements; therefore, no new types of accidents or equipment malfunctions are created. The logic of the protective features described in the Technical Specifications for steara line break and condenser vacuum mitigation have not been altered; therefore, the existing margin of safety remains unchanged.

5 UCR 98-04 i TITLE: USAR Change - Table 11-5-1 DESCRIPTION: This UCR revised Table 11-5-1 to correct math and editonal errors. A non-coruervative calculational error occurred when an incorrect conversion factor was used when converting cubic feet per second to gallons per minute. When the correct value was used, the derived value for the amount of activity released to the river in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> drepped by one order of magnitude. The second error had no effect on thmult, and was simply a transposition error from earlier in the table. The calculational error resulted in an overly-conservative analysis of the activity released to the river assuming a failure of the Radwaste Building. The error was also present in FSAR Q&A 2.37. The analysis and conclusions in the subject table are not used in any existing radiological reporting procedures and no other analysis has been fbund that uses this analysis as an input value.

SAFETY EVALUATION: This change corrects an error in the analysis which determines the potential ofTsite dose consequences of a failure of the Radwaste Building. The corrected value is an order of magnitude less than originally determined. No change to plant equipment, configuration, or procedures is involved. No change to equipment status, lineups, or maintenance results from this correction. Therefore, there is no increase in the probability of occurrence or consequences of equipment malfunction and no new types of plant events or equipment malfunctions are intnxluced. As the corrected results are more conservative than originally calculated, there is no reduction in the margin of safety.

UCR 98-045 TITLE: USAR Change - Reactor Equipment Cooling (REC) System Design and Safety Bases DESCRIPTION: This UCR incorporated various clarifications to the REC system operation, design, and safety bases. As a result of an NRC inspection, it became apparent that the existing USAR text hcked sufTicient detail and/or clarity. To address this weakness, the following clarifications were made:

The Safety Iksign Basis was revised to differentiate between REC design requirements for an active failure and a passive failure.

A note was added to Table X-6-2 to indicate that the REC heat exchanger heat transfer area and heat duty rate specified is the spec sheet data only, not the actual value. Actual heat duty rate and heat transfer area of the heat exchanger is a function of fouling and number of tubes plugged, which are both controlled by the applicable plant procedures and analyses.

A note was added to Section X-6,5.2 to reflect the fact that concurrent operation of all ibur REC pumps and both heat exchangers is allowed during normal operation.

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I The transient analysis in Section X-6.53 was deleted as it had been superseded by more rigorous calculations which were added by reference.

Discussion was added to Section X-6.6 concerning the Senice Water system backup capability.

This capability is for any pipe break, notjust for Class IS REC piping.

Discussion was added to Sections X-6.5 and X-6.6 to delineate how soon senice water must be introduced into the REC system in the event normal cooling cannot be maintained.

SAFETY EVALUATION: The design bases and functions of REC as described in the USAR are not changed by this activity. IUIC's nxxles ofoperation in response to a plant event as described in the USAR remain unchanged and continue to reflect the system's operational needs. The operation of the REC system, its level of maintenance and testing, and its service conditions as described in the USAR remain unchanged. Therefore, the probability of equipment malfunction is not increased. Service water backup to the REC system was provided prior to plant licensing to mitigate the consequences of a pipe rupture anywhere in the system.

This is currently reflected in the USAR's safety evaluation for the REC system. The subject changes bring the USAR sections into agreement. These changes do not infer any new functions or modes of operation for the REC system, nor do they suggest any diiTerent operating schemes than those originally designed. Therefore, no new types of plant events are created. The fc.ilure males of the REC system remain unchanged, thus no new types of equipment malfunctions are introduced. The changes provide a basis for a longer time until operator action is required to establish senice water backup in the unlikely event of a passive REC component failure. Therefore, the margins of safety have not been decreased.

UCR 98-047 TITLE: USAR Change - Iliennial Review of Flood Procedures DESCRIPTION: This UCR revised Section 11-4 to restore the administra'ive requirement for the Plant Manager to perfonn a biennial review of the floal procedures described in that section. This requirement had previously been deleted under LCR 94-0070 without proper evaluation per 10CFR50.59 requirements. The LCR was pc.rt of a larger administrative change that replaced the biennial resiew requirement with a requirement of"once every five years or more frequently." At the time, it was not realized that the review frequency of the flood procedures was formally addressed in FSAR Q&A 234, and that the biennial review requirement had been formalized through commitment to the NRC. Although the biennial rev: w requirement was deleted in July 1994, an administrative review of the flood procedures was conducted in March 1996. Therefore, the condition was discovered and corrected before the next biennial review period was exceeded.

SAFETY EVALUATION: The restoration of the biennial review of the flood procedure does not impact the manner in which any system, component, or structure is maintained, operated, or controlled. The steps within the flaxi procedure and the equipment utilized within the flood procedure are not altered by this change. There is no impact on any systems or components used to monitor, control, or report the radiation exposure of plant personnel or the general public. This change does not impact the function of any equipment considered to be important to safety or safety-related. No new equipment or control devices / schemes are being added to the plant. This change does not impact the maintenance of any equipment and does not alter the location of any equipment required to support the implementation of the flaxi procedure.

Therefore, thyrobability of a malfunction of equipment important to safety is not increased and no new types of plant events or equipment malfunctions are created. This change does not impact any of the discussions in the " Review and Audit" or " River Level" sections of the Technical Specifications, thus the margin of safety as dermed in the basis for any Technical Specification is not reduced.

UCR 98-050 TITLE: USAR Change - Fkxxiing Materials DESCRIPTION: This UCR revised the description of the type of materials to be used when responding to flooding conditions. The actions described in the CNS flooding procedure were inconsistent with the description 1

I in the USAR; however, the actions in the flooding procedure are equal to or better than the actions described in the USAR. The revised text describes a typical sandbag and wood barrier and permits future l changes to the flocxling procedure as long as they provide " equal to or better" protection than currently  !

described in the USAR. These revisions provide additional flexibility when preparing for 11oodmg j conditions while still meeting the intent of the USAR text. I SAFETY I EVALUATION: This USAR change will pennit newer technokigies to be utilized when building flood barriers, which will provide equal to or bc ter protection than previously described. Increasing the quality of the flood barriers will not increase the possibility of an offsite d, se release. No equipment is alli eted by this l

USAR change and flooding will still be controlled with appropriate barriers. The flooding barriers are 1 not related to any equipment malfunction, nor are they required to respond to any equipment raalfunction.

This USAR change does not permit any changes in the design or operation of the plant. The plant response to a ikxxiing event will romain equal te or better than the existing USAR description; therefore, l no new types ofplant events or equipment malfunctions are created. The basis for implementing the site flood procedure is not changed and there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

I UCR 98-051 1

TITLE: USAR Change - Clarification of the Diesel Generator (DG) Overspeed Protection Trips and Alarms i DESCRIPTION: This UCR clarified and corrected the USAR description of the DG overspeed trips and alarms. The previous text stated that the clarm was sounded before either the mechanical or electrical overspeed trip.

In actuality, an alarm is generated concurrent with either overspeed trip. This change brings the USAR into agreement with the installed configuration of the DG overspeed trip / alarm circuitry which was extensively revised by Design Change 93-024. The design change: 1) modified the electric trip from a limit switch component activated by the governor to an electronic trip activated by the relay tachometer,

2) added an alarm function to the mechanical overspeed trip, and 3) established a setpoint ditTerence of I 10 RPM between the two overspeed trip functions. As part of the design change, the USAR was resised to recognize the two overspeed trip functions, but failed to correct the text which specified that an alarm was received prior to the trip.

The previous text was originally provided by FSAR Amendment 13 (FSAR Q&A 8.7) which was not accurate in several respects. There have always been two overspeed trips, an electrical and a mechanical, t both of which were included in the emergency trip system for the DGs. Ilowever, prior to the aforementioned design change, an alarm was only included in the electrical overspeed trip. While the original configuration may have been such that the electrical overspeed tap (and resulting alarm) occurred at a lower overspeed value than the mechanical overspeed trip, the configuration at the time of DC 93-024 was that both overspeed trips were set to occur at the same point. Based on Safety Guide 9 (Regulatory Guide 1.9) and IEEE Standard 387-1977, no regulatory or industry guidance exists that would indicate an overspeed alarm should be generated prior to the trip. Therefore, revising the USAR i to accurately describe the configuration and operation previously evaluated and established by DC 93-024 l has no effect on safety.

SAFETY EVALUATION: The DGs are used to mitigate the consequences of analyzed plant events and are not accident initiators.

Consequently, the probability of an accident is not increased. The operation of the DGs to mitigate the consequences of a plant event is unafTected. An overspeed alarm involves information provided to i

operators following a malfunction and has no discernible cfTect on 1)G availability. The only malfunction {

ofinterest is one that results in a DG sullen k)ad rejection. The ability of the DGs to remain in operation j as a result of that scenario is unaflix:ted by overspeed alarm timing. There is no change to DG operations l or design bases; therefore, the possibility of a difTerent type of accident is not created. No new failure modes are introduced. As the operation of the DGs and their ability to respond to plant events is unchanged by a concurrent overspeed alarm and trip, the margin of safety as defined in the bases for any Technical Specification is not reduced.

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l UCR 98-053 TITLE: c USAR Change - Re onnatting of Table Vill-5-1, " Diesel-Generator Loading Table, Standby AC Power System" l

DESCRIPTION: This UCR revised and reformatted Table VIII-5-1," Diesel-Generator Loading Table, Standby AC Power System," to eliminate redundant information and detail not required to support the textual information.

The original USAR table was brought forward from the FSAR and was approximately one page in length.

During a 1987 NRC inspection, it was found that the table had not been maintained current. It was out-of-date and failed to consider motor efliciency in the determination of diesel-generator (DG) loads.

Consequently, CNS updated and corrected the table. Instead of updating only the out-of-date or incorrect infonnation, the table was replaced with an attachment from the updated DG loading calculation, twenty-three pages in length. As a result, the new table contained much information not required to support the USAR text and was entirely redundant to the calculation. This approach proved to be a source of confusion and created an unnecessary administrative burden. This UCR returned Table VIII-5-1 to the FSAR content level and provided a reference to Nuclear Engineering Department Calculation 87-104 A, Plant AC Load Study.

SAFETY EVALUATION: This activity made no changes to the plant load study or the DG loading analysis. Equipment performance and plant event consequences are not afTected by this change. No change to the operation, availability, or maintenance ofequipment important to safety results from this change; therefore, there is no increase in the probability ofoccurrence or consequences of a plant event or malfunction of equipment important to safety. No hardware or procedural changes are required and there are no changes to equipment operation, lineups, or maintenance as a result of this change; therefore, no new types of accidents or equipment malfunctions are introduced. The information being removed from this USAR table is completely contained in the design calculation of record. There is no change to the design inputs or outputs of existing event analysis; therefore, there is no reduction in the margin of safety.

UCR 98-052 TITLE: USAR Change Figure X-14-2 DESCRIPTION: This UCR revised Figure X-14-2, Flow Diagram for the Radiological Waste Drains / Vents, to reficct the as-built condition of the facility by adding two valves in the "Z" sump discharge line that were not previously shown on the drawing, not labeled, not in the Equipment Data File (EDF), and not in the system operating procedure. The physical configuration of the 1 inch drain and 3/4 inch vent valves was field verified. Engineering investigation and review determined the valves to have been installed during original plant construction and they were verified to meet the applicable system design criteria. While original drawings showed a 1 inch drain connection on the sump discharge line, no valve was shown The vent valve, while shown on original drawings, was incorrectly labeled and, as a result, not properly reflected in the system design documentation and operating procedures. Both valves are maintained closed and capped with no safety function, and are not required to be operated during accident conditions.

The valves were subsequently assigned Component Identification Codes RW-V-1258 and RW-V-1259.

SAFETY EVALUATION: This activity was a "de facto" change to the facility as described in the USAR. Since both valves were found to meet the system design criteria, no physical changes to the plant were required. The purpose and function of the valves are not being changed. Further, since the valves were already being maintained closed and capped, no operational or configuration changes were required. The valves have no safety function and are not required to operate during an accident condition. Plant personnel or olisite dose will not be affected by this change. For these reasons, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety has not been afrected. For these same J reasons, the possibility for an accident or malfunction of a different type than any previously evaluated has not been created and the margin of safety as defined in the basis for any Technical Specification has not been reduced.

UCR 98-061 (USQE 1998-001l)

TITLE: USAR Change - Clarification of the Testing and Inspection Requirements for Main Steam and Reactor Feed Piping DESCRIPTION: This UCR clarified the testing and inspection requirements for Main Steam (MS) and Reactor Feed (RF) piping. Originally, Section IV-11.4," Main Steam Line and Feedwater Piping," stated that the materials used were, as a minimum,in accordance with ANSI B31.1.0, with the following additional requirements for piping 2-1/2 inches in diameter and greater:

1, Full penetration welds without use of backing rings.

2. TIG root pass and shielded metal arc or submerged are welding of remaining passes.

3. 100% radiographic examination of welds.

4. Ultrasonic testing.

His appears to require a 100% radiographic examination of welds for all MS and RF piping at CNS 2 1/2 inches and greater. This was in apparent conflict with USAR Appendix A. " Pressure Integrity of Piping and Equipment Pressure Parts," which specifies that only Class IN and IIN piping requires a 100% radiographic examination of welds. I USAR Appendix A originated from General Electric (GE) Design Specification 22Al295AC, Revision I

1. This specification encompassed all piping and equipment pressure parts (i.e., Classes IN, IIN, IIIN, and IVP) at CNS and was a requirement for systems supplied by GE but only a recommendation for those systems supplied by other vendors. As it applies to the MS and RF systems, GE was responsible for the piping from the reactor vessel up to and including the isolation valve outside primary containment while Burns & Roc (B&R) was responsible for the balance. Therefore, the CNS design requirement is for 100% radiography of welds in piping from the reactor vessel up to and including the isolatior_ valve outside pnmary contamment only; there are no radiography requirements for welds in the remaining MS I and RF piping (i.e., class IVP). This agrees with the requirements of ANSI B31.1 While not a design requirement, B&R was contractually required during construction to perform 100% radiography for welds in the MS and RF piping, including that classified as IVP. These contractual requirements were later translated into Section IV-11.4, creating confusion as to the actual piping design requirements. To eliminate this confusion, this UCR revised Section IV of the USAR to eliminate the conflicting requirements and to reficct that Appendix A addresses the design, fabrication, inspection, and testing ,

requirements for MS and RF piping. A review of the original Safety Evaluation Report (SER) and I supplements verified that the nondestructive inspection requirements were not credited in the SER for j Class IVP MS or RF piping.  !

SAFETY EVALUATION: This activity will not change the overall design, function, or reliability of the MS or RF piping. This UCR

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j clarifies the piping design requirements for the MS and RF piping in Section IV to agree with Appendix l A of the USAR. This change does not affect the original design and fabrication requirements for the subject piping, nor will the integrity of the piping be decreased. In addition, this activity will not result in any increased radiological effects. Consequently, the probability of occurrence or the consequences I

of an accident previously evaluated in the SAR are not increased. This change will not induce any equipment malfunctions or failures. It will not adversely afTect any plant equipment or systems important to safety, nor willit affect their accident mitigation capability or change their failure modes. Clarifying

he testing and inspection requirements for the MS and RF piping does not create any accident scenarios different than currently evaluated in the USAR. This activity does not change the form, fit, function, or operating parameters of the MS and RF systems. The integrity of the piping will continue to meet ecxle requirements during any repair or fabrication work; therefore, the possibility of a ditTerent type of malfunction than previously evaluated is not created. As this activity does not alter the function, reliability, or accident mitigation capability of any equipment or systems, and does not effect any assumptions, calculations, procedures, or design specifications used to establish the basis for defining the plant's margin of safety, the existing margins of safety as defined in the bases for any Technical Specificatien are unafrected.

UCR 98-072 (USQE 1998-0035)

TITLE: USAR Change - Diesel Generator (DG) Starting Air Crosstie Line Operational Configuration DESCRIPTION: This UCR constitutes a"de fxto" change to the facility in that the USAR description of the DG Staning Air Crosstie Line valve line-ups has been revised to match the in-plant operational configuration. The USAR description is provided to demonstrate that single failure criteria are met for this DG support system, and originally described all five associated valves as closed. During normal operation, the four manual root valves are maintained closed, but the single manual valve associated with bypassing the air receivers is maintained open. This configuration does not invalidate the single failure criteria, and I therefore this change has no impact on safety. The text was revised to eliminate the detail relating to these valves

• normal status, and replaced with information pertaining to the design requirement to confonn to

{

I single failure criteria.

SAFETY EVALUATION: This "de facto" facility change involves a revision to the DG Starting Air Crosstie Line valve operational .

configuration as originally stated in the USAR. The DGs are used to mitigate the consequences of analyzed plant events and are not accident initiators. As a result, the probability of an accident is not increased and there is no possibility for an accident of a different type than previously evaluated. This change has no impact on DG reliability or performance. No change to the performance of the DG Starting Air subsystems results from this change because subsystem independence is unchanged.

Therefore, the consequences of an accident or equipment malfunction are not increased. This change has no efTect on the maintenance, testing, or performance of any structure, system, or component; therefore, there is no increase in the probability of a malfunction of equipment important to safety. Failure of one of the root valves would constitute a passive failure that is not with CNS' licensing basis. Finally, as this "de facto" change has no effect on the perfonnance or availability of the DG Starting Air system, the l margin of safety as defined in the bases for any Technical Specification is not reduced.

i UCR 98-073 '

(USQE 1998-0003)

TITLE: USAR Change - Protocols to Be Used for Incorporating Documents by Reference DESCRIPTION: This UCR established the protocols to be used for incorporating various documents into the USAR by I reference, as distinguished from the general references found at the end of each chapter. Fundamentally, there are three types of documents that can be incorporated into the USAR by reference: a) historical documents, b) documents that are subject to a review and update regimen outside the scope of 10CFR50.59, and c) documents that are subject to a review and update within the pursiew of10CFR50.59 and 10CFR50.71(c). Table 1-3-1 was created to identify specific documents that are incorporated by reference. Finally, these protocols were applied throughout the USAR resulting in numerous admmistrative changes, including the deletion of several inappropriate general references.

SAFETY EVALUATION: Applying the protocols to be used for incorporating documents by reference does not constitute either a real or a "de facto" change to the facility or to any underlying analyses. Consequently, there is no efTect on any accident precursors and the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is unaffected. This administrative activity does not introduce the possibility for an accident of a different type than previously evaluated and has no effect on potential equipment malfunctions. The act of formalizing the incorporation of documents into the USAR by reference may serve to better defmc what the margin of safety is for given Tecimical Specification, but the administrative process itself has no effect on any safety margins.

l UCR 98-080 Procedure 6 SI C.201 Revision 3 4 (USQE 1998 - 0024 l

TITLE: Standby Liquid Control (SLC) System Relief Valve Opening Acceptance Range DESCRIPTION: This UCR revised USAR Section 111-9 to clarify the SLC system relief valves'setpoints. A Procedure Change Request implemented similar clarifications in Surveillance Procedure 6.SLC.201, Revision 3, to ensure consistency between the two documents. Previously, the USAR identified the SLC relief valve setting as 1540 psig + 1% and the minimum lift pressure as 1478 psig with the assumed 3% setpoint tolcrunce. 'Ihe maximum lift pressure of 1602 psig was not stated. In addition, the USAR used the words "setpoint tolerance" in the discussion and reference to the 3% value. Under this actisity, the USAR discussion was revised to include the SLC relief valve maximum lift pressure of 1602 psig and replaced the words "setpoint tolerance" with the words " assumed inaccuracies" to more accurately describe the {

3% value. This change did not affect the safety objective of the SLC system or alter the design function of the SLC relief valves. Further, tbc current valve setting of 1540 psig + 1% or the specified reset pressure of s 1300 psig was not changed. Previously, Surveillance Procedure 6.SLC.201 stated that if the relief valve did not open between 1450 and 1680 psig or reset at s 1300 psig during the as found testing, the relief valve would be refurbished prior to commencing the as left testing. To be consistent with the above described USAR change, this activity revised the procedure to specify that if the relief valve does not open between 1478 and 1602 psig or reset at 51300 psig during the as found testing, the relief valve will be refurbished prior to commencing the as left testing.  ;

SAFETY i EVALUATION: The SLC system is designed to bring the reactor to a cold shutdown condition without the use of control i rods and to meet the requirements of the Anticipated Transient Without Scram (ATWS) rule per 10CFR50.62. The function of the SLC relief valves is to prevent system over pressurization and to remain closed during an ATWS event to prevent recirculation flow. These functions are not impacted by this activity as the opening setpoint and the reset pressure for the SLC relief valves are not changed.

As the described revisions do not alter any input parameters or precursors for any accident analyses described in the SAR, there is no increase in the probability of an accident previously evaluated in the SAR. Since the revisions do not change, degrade, or prevent actions required for any accident or alter the assumptions made in evaluating the consequences of an accident, the consequences of an accident previously evaluated in the SAR are not increased. Revising the acceptable opening range for the SLC relief valves to 1478 to 1602 psig for the as found testing provides for a more conservative acceptance range and approach to valve refurbishment, and piovides added assurance the valves will function as designed. Therefore, the probability of a malfunction of equipment important to safety previously evaluated in the SAR is not increased. This change does not increase radiological releases assuming a malfunction of equipment were to occur. As previously stated, this activity does not dTect any accident precursors or initiators. This change does not affect the perfonnance or reliability of the SLC system or the reliefvalves or any system interface which could lead to the possibility of a different type of accident occurring. Since the functioning of the SLC system is not affected by this activity and no new failure modes are introduced, the possibility of a difTerent type of malfunction than any presiously evaluated in the SAR is not created. Technical Specifications do not contain a surveillance requirement for a minimum or maximum lift pressure or a reset pressure for the SLC relief valves. The setpoints for the SLC relief valves are controlled through the Insersice Testing program and the Setpoint Control Program. l As the reliefvalve settings are not changed, the margin of safety provided by the current setpoints is not l affected.

1 J

F' UCR 98-081 UCR 98-098 (USQE 1998-0012)

TITLE: USAR Change - Incorporation of Controlled Station Drawings by Reference DESCRIPTION: UCR 98-081 provided for the physical removal of USAR Figures which are controlled station drawings in the CNS Drawing Control Program. The drawings are now incorporated into the USAR by reference.

Table 1-3-1 was added to the USAR to provide a listing of all controlled draings incorporated by reference. Text references to these USAR Figures were replaced with references to specific controlled drawings. Changes to drawings incorporated by reference will continue to be evaluated under 10CFR50.59 and will be provided to the NRC as part of the 10CFR50.71(c) updates. UCR 98-098 subsequently updated additional text references to USAR Figures that were inadvertently overlooked by the original UCR.

In addition, UCR 98-081 deleted USAR Figures Vll-1-8 (MSIV Test Switch) and Figure VII-l-9 (MSIV Test Piping Diagram). These figures were not referenced in the USAR and do not directly support the USAR text.

SAFETY EVALUATION: The physical removal of USAR Figures which are contained in the Drawing Control Program only constitutes a change in the way information is presented in the USAR. 10CFR50.32 recognizes that it is acceptable to incorporate information by reference into the SAR. No changes to plant event analysis, event initiators, or procedures used to mitigate plant events are involved. This change does not involve any physical change to equipment important to safety, and has no impact on system line-ups, plant operations, testing, or maintenance. Therefore, there is no increase in the probability of occurrence or consequences of an accident or malfunction of equipmert important to safety and no new types of accidents or malfunctions areintroduced. Technical Specification bases provide the reasoning to ensure that plant equipment is maintained and tested consistent with event analysis assumptions. Reference to current design basis documents (controlled drawings) in the USAR provides increased assurance that plant evolutions are evaluated for consistency with those assumptions. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

Figures Vll-1-8 and Vil-1-9 were originally included in the response to FSAR Q&A 7.6 as a way to illustrate how Automatic Depressunzation System (ADS) permissive switches could be tested, and have nothing to do with MSIV testing (as their title would imply). Testing methodology of these components is procedurally controlled under 10CFR50.59, and removal of these figures eliminates a potential source of confusion regarding ADS testing. There is no impact on MSIV or ADS operation as a result of this change. MSIV and ADS testing are governed by Technical Specifications and approved station procedures. No changes to system configurations or plant evolutions are involved. ADS or MSIV availability or performance is not alTected by removalof these USAR Figures. Therefore, there is no increase in the probability of occurrence or consequences of an ac:ident or malfunction of equipment important to safety, no new types of accidents or malfunctions are created, and there is no reduction in the margin of safety.

UCR 98-083 DCN 96 0855 (USQE 1998-0046)

TITLE: Change to USAR Figure XI-6 Burns & Roe Drawing 2006, Sheet 3 DESCRIPTION: Burns & Roe Drawing 2006, Sheet 3,is a flow diagram for the Circulating, Screen Wash, and Service Water Systems. Subsequent to initiation of this UCR, the drawing was incorporated by reference into the USAR. The drawing was revised to change the Component Identification Code (CIC) for four valves (CW-V-533 through CW-V-536) from Circulating Water to Service Water (SW-V-1497 through SW-V-1500). These valves are the Condenser Backwash Valve Disc Sparger Inlet Valves for the Senice Water

-90

1 I

system but were incorrectly assigned Circulating Water system identifiers aller original construction. In addition, an existing drain valve, CW V-809, was added to the drawing.

SAFETY j

EVALUATION: These changes will make the drawing information consistent with other plant documentation and the l existing configuration in the plant. They will not afTect the safety or design function of the Service Water or Circulating Water system. The subject valves are non-essential and provide no safety related function.

There will be no physical or operational changes in the plant or plant procedures as a result of these .

changes. The plant personnel or ofTsite dose will not be affected. There will be no equipment added or I any modification to existing equipment. Therefore, there is no increase in the probability of occunence I or consequences of a plant event or malfunction of equipment important to safety previously evaluated in the SAR and no new types of plants events or equipment malfunctions are created. The non-essential portion of the Senice Water and Circulating Water systems related to the subject valves are not discussed in the Technical Specifications; there is no reduction in the margin of safety.

)

US AR Chance Reauests (UCRs)98-090 and 98-116 l (USQE 1998-0055)

TITLE: USAR Change - Clarifications to the Reactor Equipment Cooling (REC) System Descriptions DESCRIPTION: UCR 98-090 revised Section X-6 to clarify the REC system description, and to make the terminology consistent with other plant documentation. Similar revisions were made to Section I-6 under UCR 98-116. Terminology revisions included changing: " loop" to " subsystem"; " constant loop" to " stable"

" head" to " surge"; " comer rooms" to " quads", "A" and "B" to " South" and " North"; and " Closed loot to " Critical loops." Clarifications were made to explain electrical independence and how leakage can be monitored, and to identify that REC makeup is non-essential and that the loops are interconnected.

Corrections were made to the listing of non-essential heat loads, and surge tank requirements were added as reflected in an engineering calculation. In addition, an outdated reference to the heatup time of a quad was deleted, as the conect information is reflected in other places in the SAR. The substance of the changes made to the USAR under this activity were initially made via other processes. As a result, these changes serve primarily to make the USAR consistent with the other station documents, including Technical Specifications. There were no changes to the plant configuration or operation as a result of these USAR revisions.

SAFETY EVALUATION: There are no changes to the design and operation of the plant. Initiators of plant events are not affected.

l As the design and operation of REC is not changed and there are no changes to other systems, the l consequences of plant events are not changed. These changes do not afTect the operation of plant l

equipment, including equipment important to safety. Therefore, the probability of occurrence or consequences of a malfunction of equipment important to safety are not affected. For the same reasons, the possibility for an accident or malfunction of a different type than any previously evaluated is not i created. These changes provide consistency with the Technical Specifications and do not reduce the ,

margin of safety as defined in the basis for any Technical Specification. l UCR 98-091 l (USQE 1998-0056) l TITLE: USAR Change to Correct Secondary Containment Internal Design Pressure Criteria DESCRIPTION: This UCR revised Section X11-2.3.4 to provide consistent and correct information relative to the l Secondary Containment internal design pressure criteria. Erroneous information was introduced under a previous USAR change (LCR 94-0152). This previous change revised Section Xil-2.3.4 to: 1) delete l the intemal design pressure criteria ofseven (7) inches of water as established by General Electric Design Specification No. 22Al 169 and the Burns & Roe Engineering Criteria Document, dated June 3,1970, l and 2) add a reference to the 75 psf tomado loading design criteria for the Reactor Building siding. Since the Reactor Building intemal design pressure of seven (7) inches water (36 psf) forms part of the design

( basis for Pressure and Thermal loads, this activity essentially retumed Section Xil-2.3.4 to the original 1

1

SAR description with regard to the Reactor Building internal design pressure. This activity also deleted the reference to the Reactor Building siding internal design pressure parameter from Section XII-2.3.4 as this information is discussed in detail in Section XII-2.3.3.2.4, " Tornado Loads and Additional Considerations." Finally, this activity clarified the reference in Section XII-2.3.4 to the 2 psi design

, criteria for the primary and secondary containment structural walls.

L While changes under this USQE are limited to USAR Section XII-2.3.4, the USQE also addressed and evaluated erroneous changes made to Section XIV-4.4.3 by LCR 94-0152 that were " unknowingly" corrected by a subsequent USAR change (LCR 95-0087). Since the corrections to Section XIV-4.4.3 l were made unknowingly, they were not adequately addressed in the USQE for LCR 95-0087.

l There have been no changes to Secondary Containment based on the erroneous information introduced by LCR 94-0152; therefore, this activity serves only to restore and clarify the original SAR text.

SAFETY EVALUATION: This activity corrects structural design parameters which are not credited as event initiators and, therefore, there would be no increase in the probability of a plant event as previously evaluated in the SAR. Since the USAR accident analysis does not rely on Secondary Containment / Reactor Building to maintain a positive or negative pressure beyond 0.25 inches of water, the consequences of an event or accident, as i.

' previously evaluated in the SAR, remain unchanged. As there are no physical changes to the facility as a result of this activity, the probability or consequences of a malfunction of equipment important to safety i is unaffected. For this same reason, along with the fact that the internal design pressure specification for the Secondary Containment / Reactor Building cannot initiate any postulated accidents or events, there is no possibility for an accident or malfunction of a different type than any previously evaluated. Finally, as this activity serves only to co Tect an erroneous change in the USAR, thus restoring the intemal design pressure criteria to actual design values, the margin of safety as defined in the basis for any Technical Specification is unchanged.

UCR 98-094 (USQE 1998-0041)

TITI.E: IIistoricalInformation Protocols DESCRIPTION: This UCR evaluated and established the necessary definitions and protocols for designating portions of the USAR as" historical." In addition, these protocols were applied to certain qualifying portions of the l USAR, resulting in the administrative reclassification of Section I-7, " Comparison of Principal Design Characteristics," and Section XIII-4, " Construction and Preoperational Test Program."

Where appropriate, the affected text has been changed back to the FSAR content level to provide for a consistent historical understanding. With the submittal of the FSAR in 1971, CNS provided the information required per 10CFR50.34(b)in order for the Atomic Energy Commission tojudge the safety of granting Nebraska Public Power District an Operating License. With the promulgation of the FSAR Update Rule J in 1981, the Code of Federal Regulations established requirements for incorporating the latest information into the USAR. Ilowever, Generic Letter 81-06 noted that not all USAR information needed to be updated. This UCR formally established the convention of designating certain USAR information as

" historical" and outside the scope of 10CFR50.71(e) for routine updating. As a result, infonnation in the USAR may now be classified as " historical" ifit is: a) related to a physical plant milestone that is inherently dated, b) outside the responsibility of the licensee to control or influence as a 10CFR50.59 change, test, or expenment, or c) information that has been retained to provide historical perspective, but has been superseded with equivalent up-to-date information. The protocols established are summarized as fallows:

1. / . . . font will be used to identify the " historical" portions of the text.

2. s rJ raseology will be employed to indicate the existence of" historical" infonnation within SAi .:ction or subsection.

3. L  : cal" text will be restored to its original form, on a case-by-case basis, from previous atta update the infonnation.

4. Contextual information will be added, as desinxi, so that the reader can understand the source or time frame of the information being designated as " historical."

5. Pointers to more up-to-date USAR information will be added when it is decided to leave information as dated for archival purposes.

6. " Historical disclaimers" will be removed that may have been previously added to identify h'storical information.

SAFETY EVALUATION: Fonnalizing the criteria and methodology for classification of qualifying USAR sections as " historical" has no effect on any accident precursors. The same is true for the restoration of the information pertaining to comparative plant data and the preoperational test program. Designating certain USAR information as " historical" does not constitute either a real or "de facto" change to the facility or any underlying j analyses; therefore, there can be no increase in the consequences of previously evalucted accidents.

Designating USAR information as " historical" (kies not diminish the need to consider that information

)

when it forms an underlying basis in the design of structures, systems, or components. The characterization ofinformation as " historical" has no physical, operational, or analytical effects on any plant equipment. Therefore, there is no increase in the probability of occurrence or consequences of a malfunction of equipment important to safety and no new types of accidents or equipment malfunctions are introduced. Information classified as " historical" may have at one time been used as a basis for the design ofstructures, systems, or components in the Technical Specifications. Ilowever, excluding such information from routine updating does not impact their function as CNS would still be obligated to evaluate the effects of significant new information that may be in conflict with the historical USAR discussions. Thus, there is no effect on the margin of safety as defined in the basis of any Technical Specification.

UCR 98-095 (USQE 1998-0053)

TITLE: USAR Change - Containment Isolation Valves DESCRIPTION: 'Ihis UCR clarified the purpose of two isolation valves in water sealed penetrations where nonnally only one valve is regmred. For most water sealed lines, only one primary containment isolation valve (CIV) is required in addition to the water seal; USAR Table VII-3-1 indicates which penetrations are water sealed. Ilowever, in a post-Loss of Coolant Accident condition, the water in the torus could be forced back up the penetration piping due to the excess pressure in the torus. Consequently, there are situations where the penetration piping drains to a sump and the failure of the single CIV could result in water being drained frun the torus to the Secondary Containment. As a result, water could be depleted from the torus, radioactive material could be released to Secondary Containment, and potentially contaminated water could be pumped from the sump to outside the Secondary Containment. For these penetrations, two isolation valves are required to prevent these effects in the event of the single failure of a single isolation valve. This UCR revised the USAR to reflect the above discussion and to clearly delineate the susceptible penetrations in Table VII-3-1.

SAFETY EVALUATION: There are no hardware or operating procedure changes associated with this USAR change; thus there is no effect on any accidem initiators. With the incorporation of this USAR change, the safety function of the subject valves is clarified and will be maintained. Therefore, the consequences of previously evaluated accidents are unchanged. The probability of a malfunction of equipment important to safety is actually decreased because with the safety function identified, the valves can be tested to ensure that they will fulfill their safety function in an accident. As this USAR change ensures that the isolation valves will continue to fulfill their safety function post-accident, the consequences of a malfunction of equipment important to safety are not increased. No new types of accidents or equipment malfunctions are introduced. The margin of safety will be maintained as the original design for these valves will be preserved.

m

UCR 98-096 (USQE 1998-0065) l TITLE: USAR Change - Residual IIcat Removal (RIIR)/ Low Pressure Coolant Injection (LPCI) Flow Rates DESCRIPTION: UCR 98-096 revised the USAR to clarify RIIR system limitations. As originally worded, the USAR implied that the RIIR system could not support RIIR/LPCI flow rates above 8400 gpm. In reality, the system can support flow rates in excess of 8400 gpm; however, the flows are procedurally limited in response to CNS commitments. The plant configuration and the method of operating the RIIR/LPCI system during normal plant or transient operation were not changed by this UCR. Maximum flow of the RIIR pumps is physically restricted by restricting orifices in the discharge piping of the RIIR pumps.

SAFETY EVALUATION: The operation of the RIIR/LPCI system during normal plant op:: ration or transient operation is not changed (i.e., the RIIR/LPCI pumps will not be allowed to run out) and the physical configuration of the RIIR/LPCI system is not modified. As a result, there is no change to any accident precursor that could increase the probability of occurrence or consequences of an accident or malfunction of equipment i important to safety. For these same reasons, no new types of accidents or equipment malfunctions are introduced. This change will not affect the ability of the system to mitigate the consequences of an accident, therefore, there will be no increase in radiological dose. While RIIR/LPCI system flow rates are factored into the margin of safety as defined in the basis for various Technical Specifications, the clarifications to the system limitations made by this activity have no impact on these margins.

UCR 98-100 (USQE 1998-0007)

TITLE: USAR Change to Correct Reactor Building IIVAC Related Errors DESCRIPTION: This UCR addressed two issues: 1) The as-built Reactor Building IIVAC System differed from the description provided in the USAR in that all of the non-below grade level air does not exhaust to the refueling floor. From a review of Burns & Roe Flow Diagram 2020, all of the 903'6" elevation airflow exhausts directly to the Reactor Building Plenum. This discrepancy is best characterized as conflicting USAR information since the B&R drawing showing the conect configuration is also in the USAR.

Ilowever, it was treated as a "de facto" facility change requiring a USQE since there is no explicit evidence that the Atomic Energy Commission acknowledged the as-built design. 2) The as-built Reactor Building IIVAC System differed from the description provided in the USAR in that all of the air flowing to the Refueling Floor does not exhaust via the embedded ducts in the Spent Fuel Pool, Reactor Cavity, and Dryer / Separator Pit. B&R Drawing 2020 shows several alternate exhaust paths besides the embedded ducts (as confumed by walkdown) leading to the Reactor Building Plenum. As a consequence, the ability of this air flow to prevent any contaminated air flow from egressing to an area oflower contammation is overstated. The actual capability is more of a thin air blanket that will cause low levels of airbome contamination to migrate to the embedded ducts. Ilowever, high airborne contamination levels will likely escape to the general area. As like the first issue, this condition was treated as a "de facto" facility change requiring a USQE.

SAFETY EVALUATION: Given the originallicensing basis assumption of zero unfiltered release due to the 6-second holdup time and iast-acting Secondary Containment isolation Valves, this activity would result in no safety impact since the Refueling Accident assumption of uniform mixing on the Refueling Floor would be equally conservative in oft-site dose consequences as a burst release through the embedded ducts. Under this ass'unption, all the exhaust would travel through the Standby Gas Treatment System. Ilowever, if a 90-second unfiltered release is assumed, this "de facto" change results in a dose that is consistent with the j accident analysis since the embedded ducts will not draw out a large unfdtered source term wi;hin the first :

90 seconds of the event. This USQE was conservatively based on the assumption of a 90-second  ;

unfiltered release. Neither change under this activity (i.e., changing the common point of entry for i Reactor Building liVAC exhaust from the Refueling Floor to the Reactor Building Exhaust plenum and l reducing the effectiveness of the embedded ducts to prevent the spread of contamination) can affect I I

I

)

accident precursors. Therefore, the probabihty of an accident is unaffected. Changing the common point ofentry for Reactor Building IIVAC exhaust from the Refueling Floor to the Reactor Building Exhaust plenum has no adverse efTect on accident consequences since the path of egress has no impact on the efTectiveness of the isolation signals to perfbrm their safety function within the time assumed in the accident analyses. Likewise, reducing the effectiveness of the embedded ducts to prevent the spread of contamination has no impact on the uniform mixing assumption in the Refueling Accident analysis and the otT-site dose conclusions reached in USAR Chapter XIV are unalTected. The Reactor Building IIVAC is classified as a non-safety related support system to equipment in the Reactor Building. As a result,its functions are not credited by any safety-related equipment. Therefore, the specific etTects of the ducting configuration changes associated with this activity are limited to Seismic 11/1 concerns. As the l

Seismic II/I evaluations during initial licensing and the recent Seismic Qualification Utility Group initiatives were based on physical walkdowns and as-built drawings, this "de facto" change to the facility does not increase the probability or consequences of a malfunction of equipment important to safety.

These changes do not create unique precursors to any new transients or accidents that are not currently described in the SAR. Analyses, operational controls, and maintenance practices are based on the as-built system; therefore, no new types of equipment malfunctions are introduced. While the Reactor Building IIVAC design and operation have input to the Technical Specification bases for Secondary Containment Integri'y, Standby Gas Treatment, Primary Containment Purging, and Reactor Building Isolation Radiation Monitoring, the "de facto" configuration changes made by this activity have no etTect on the margin of safety as established in these bases.

UCR 98-106 (USQE 1998-0059)

TITLE: USAR Change - Historical Information in USAR Section I-

6.1 DESCRIPTION

This UCR classified portions of Section 1-6.1.1, Site and Environs, as historical and either made no changes to the USAR text, or restored text that had been previously changed back to the original FSAR content level to provide for a consistent historical understanding. Section 1-6.1 has been expanded to provide an editorial explanation of the format used to identify the portions classified as historical USQE 1998-0041 evaluated the acceptability ofdesignating selected information in the USAR as historical, thus climinating the requirement for maintaining said information up-to-date.

SAFETY EVALUATION: This activity administratively classifies portions of Section 1-6.1.1, Site and Environs, as historical and no longer subject to updates under the n:quirements of 10CFR50.71(c). This administrative classification of the general station description as historical does not constitute either a real or a "de facto" change to the facility or to any underlying analyses. Consequently, there is no efTect on any accident precursors and the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is unatrected. Further, as this change has no physical, operational, or analytical elTects on CNS structures, systems, or components, this activity does not introduce the possibility for an accident or malfunction of a difTerent type than previously evaluated. While the information classified as historical may have at one time been used as a bases for the design of structures, systems, or components in the Technical Specifications (such as seismicity, hydrology, and meteorology), these inputs are controlled and evaluated under specific and detailed analyses or calculations. Should future changes to these inputs be required, changes to the related analyses or calculations would still be subject to 10CFR50.59 evaluation and 10CFR50.71(c) reporting if USAR changes were involved.

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f IJCR 98-109 Operatina License Chanoe Reauest 98-008 (USQE 1998-0086) t TITLE: USAR and Technical Specification Bases Changes to Remove Nuclear Safety Operational Requirements l l (NSOR)Information

{

DESCRIPTION: %e USAR'was revised to eliminate NSOR subsections, as well as related infonnation that established

- the NSOR concept. Information added to the NSOR subsections during the course of USAR updating over the years was also relocated as appropriate. The NSOR information was part of the original FSAR and was developed by General Electric along with the Nuclear Safety Operational Analysis (NSOA) to help derme the limiting conditions for operation and surveillances for the proposed CNS Tecimical Specifications. The Atomic Energy Commission rejected the NSOR-based proposed Technical Specifications. FSAR Amendment 22 subsequently resubmitted Technical Specifications which were l not based on the NSORs. In addition, the adoption ofImproved Technical Specifications has superseded j any current usefulness of the NSOR information. The NSOR information remains in the FSAR as a point j of historical reference. The Technical Specification Bases incorrectly referenced the NSOR USAR information in several cases; therefore, appropriate Bases changes were made to reflect these USAR i

changes. l SAFETY EVALUATION: Elimination of the NSOR concept from the USAR and Technical Specification Bases does not involve any physical changes to th design, analyses, or procedures pea umg the facility which could have any efTect on accident precursors or mitigation. This USAR change eliminates infonnation that has been deemed historical and is now superseded by current operational criteria. Compliance with the Tecimical l Specifications and the Technical Requirements Manual provide contemporary assurance that the i consequences of accidents remain bounded by the evaluations in the USAR. Removal of NSOR information has no elTect on any of the analyzed failure modes of equipment impoitant to safety. This change does not introduce any new accident precursors or failure modes. As the NSORs represent ,

supersedw historical information, they do not present any margin of safety information that is salient to I the present Technical Specifications.  ;

1 UCR 98-111 l TITLE: USAR Change - Figure X-8-4 DESCRIPTION: This UCR revised USAR Figure X 8-4, Water Jet Spraying System at Bottom ofIntake Structure, to remove fire pumps I A and IB from the figure. These pumps were previously removed from the plant by Design Changes 83-89 and 83-90. A new fire protection clean water supply was installed per Design Change 81-114 which eliminated the need for the original pumps. The Safety Evaluation Report for License Amendment 82 recognized and accepted the replacement of the old fire protection system with the clean water system.

SAFETY EVALUATION: There is no increase in the probability of occurrence or consequences of an accident or malfunction of equipment importai1 to safety because the previous electric fire pump 1 A and diesel fire pump 1B are no longer required for protection of safety related equipment. The new clean water fire protection system installed under Design Change 81-114 eliminated the need for these pumps. No new types of accidents or equipment malfunctions are introduced as this change is only removing equipment that is no 1,nger required. The new clean water fire protection system installed under Design Change 81-114 ensures that the margin of safety is not reduced.

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UCR 98-ll3 UCR 99-016 (USQE 1998-0097)

TITLE: USAR Change - Rebaseline of Section X-

8.2 DESCRIPTION

USAR Section X-8.2, Residual Ileat Removal Service Water Booster (RIIRSWB) System, was resised in ac' cordance with the USAR Rebaseline Project. Changes included the following: addition of safety objective and safety design basis to prevent uncontrolled releases to the environment; clarification that the RilRSWB system consists of two mechanically and electrically independent loops; clarification of RIIR heat exchanger cooling water discharge flow; correction of RIIRSWB pump rating; addition of l

conditions for windmilling of RIIRSWB pumps; correction of valve positions preventing inadvertent admission of Senice Water to the RIIR system from locked-open and locked-closed to sealed-open and scaled-closed; addition ofdescription of RIIRSWB pump electrical controls and interhicks; clarification that operation of both RIIRSWB pumps in either loop for periods of greater than one minute is prohibited except when required by Emergency Operating Procedures; clarification that two pumps are normally required to be available for each loop, clarification that RIIRSWB pumps are manually started from the Control Room; correction that only one RIIRSWB pump is required to remove the design RIIR system heat load during postulated transient or accident conditions; removal of reference to RIIR steam condensing mode; correction of discussion of common mode failures; addition of flooding effects of a failure of one Senice Water supply header; clarification of interlocks, operation, and contingencies relat:xi to basement flooding and the Senice Water cross-tie valve; clarification of seismic classification of 12" Fire Protection system supply header; clarification that valves controlling cooling water to the control room air conditioner are manually operated versus solenoid operated; clarification of post maintenance testing for RIIRSWB pump and valve testing; and various changes for consistency in terminology.

SAFETY

- EVALUATION: These changes do not involve any modification to the existing plant configuration or operation. There are no changes to admmistrative limits or acceptance criteria for equipment operability. Since the subject changes do not lead to any increase in the stress on any equipment, there is no increase in the probability of any equipment or system boundary failure which could initiate an accident sequence. These changes do not change the way associated valves are operated and do not change the requirements for RIIRSWB piping. Th:se changes will have no effect on system performance and will not introduce any failure mode in addition to those previously evaluated in the SAR. There is no increase in the consequences of any accident previously evaluated in the SAR. The inclusion of the design feature for mitigating fission product release to the environment more clearly states the safety design basis of the RI1RSWB system.

None of the changes degrade actions or design provisions required to mitigate the effects of presiously analped accidents. None of the changes can increase the probability or consequences of a malfunction ofequipment important to safety. The changes relating to the capability of a single RIIRSWB pump to provide the necessary cooling load for long term containment cooling are consistent with the accident analysis and there is no increase in the probability of a failure of the RIIRSWB system to perfonn its safety design basis function. The subject changes do not impact the ability of any structures, systems, or components to perform any safety functions described in the SAR. They will not induce failure of any equipment important to safety. The subject changes do not introduce any new accident initiators into the plant or create the possibility of a different type of malfunction than previously evaluated. No new failure nales or mechanisms are introduced into the plant. None of the changes reduce the margin of safety as defined in the basis for any Technical Specification; they have no elTect on existing acceptance limits, limiting conditions for operation, or surveillance requirements.

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UCR 98-114 l I (USQE 1998-0107)

TITLE: USAR Change " Standby" or" Spare" Condensate Filter Demineralizer Units i DESCRIPTION: The USAR was revised to more accurately describe the operating philosophy of the Condensate Filter I Dernincralizer system. The original design basis was for six condensate filter demineralizer units, with five in service during normal operation and the sixth in standby. Ilowever, to maintain optimum I

operation and performance, historically all six were routinely placed in service during full power j operation. To add additional operational flexibility and to further increase overall e0iciency, a seventh l condensate filter demineralizer unit was added under Design Change (DC) 86-008A. While the design I change clearly intended the seventh condensate filter demineralizer unit to routinely be in service during i full power operation, this operating philosophy was not translated to the USAR sections describing the I l Condensate Filter Demineralizer system operation.

l SAFETY EVALUATION: The Condensate Filter Demineralizer system is non-essential and is not credited with any action to ,

mitigate the consequences of an accident or transient. This system does not initiate or act as a precursor I for any accidents or transients described in the SAR. Operating with all seven condensate filter demineralizer units in senice would tend to enhance safety because: 1) the increased system capacity l l provides for a greater range of system response in the event of a feedwater disturbance, 2) the decreased l

pressure drop across the system provides increased suction head to the downsteam pumps, and 3) the l increased run times between unit backwash and re-coat reduces the potential of human error during i related equipment manipulations. No increase in challenges or degradation of safety system performance can be attributed to this activity. It affects only non-essential equipment not credited by any safety related equipment. Therefore, the probability of occurrence or the consequences of an accident or malfunction ofequipment important to safety are not increased. Revising the USAR to properly reflect the operating l configuration provided for by DC 86-008A does not create the possibility for an accident or malfunction l of a different type than previously evaluated. Although the overall system operation is being changed, i

subsystem and associated component manipulations are identical. There are no related Technical  ;

Specifications for this activity; therefore, it does not reduce the margin of safety as defined in the basis l for any Technical Specification.

UCR 98-ll5 (USQE 1998-0106)

TITLE: USAR Change - Air Ejector O!Tgas Monitors DESCRIPTION: This UCR revised Table Vll-12-1, Process Radiation Monitoring Systems Characteristics, to show the correct number of upscale trips per channel for the Air Ejector O!Tgas Monitoring System. It also revised associated text to provide a more accurate description of the system trip functions. The Air Ejector Offgas Radiation Monitors (AEORMs) have always had two upscale trip functions: the "high" provides an alarm in the Control Room and the "high-high" provides input to the protective action logic. However, the FSAR or USAR did not previously include the "high" upscale trip. In 1993, with the issuance of Amendment No.158, CNS committed to maintaining the "high" setpoint at approximately 1.5 times normal background as part of removing the Main Steam Line Radiation Monitors scram and Group I isolation functions. The AEORMs were also originally designed with an " inoperable" (i e., inop) trip which provides input to the protective action logic; however, the inop trip was not previously discussed l in the FSAR or USAR. l l SAFETY I EVALUATION: This activity revises the USAR to reflect the current design and licensing basis for the AEORMs and does not result in any change to the design or operation of any components in the plant. The non-essential i process radiation monitoring system does not initiate or act as a precursor to any evaluated accidents.

This activity does not affect the accident source term and accident mitigation systems continue to operate as credited in the accident analysis; therefore, there is no increased dose and no increase in accident consequences The addition of a high alarm on the AEORM and the inclusion of the inop feature as a trip

i. j

condition provides conservative warning and more likely isolation of this release path. No increase in challenges or degradation of safety system performance can be attributed to this actisity. These changes do not create unique precursors to any new transients or accidents that are not currently described in the USAR. Analyses, operational controls, and maintenance practices are based on the as-built system.

Revising the USAR to conform to the way the system was actually designed and analyzed does not create any new types of equipment malfunctions. The margin of safety can only be reduced if the Air Ejector Off Gas radioactive release rate could increase because of this activity. The changes reflected in this UCR provide for earlier release detection and isolation and thus can only serve to preserve this margin.

UCR 98-117 (USQE 1998-0079)

TITLE: USAR Change - Reactor Building Roof Response to a Tomado DESCRIPTION: The USAR was revised to clarify and provide additional discussion on the response of the Reactor )'

Building roof to a tornado. The original FSAR contained wording in Chapter Xil and Appendix C that implied that during a tomado, both the Reactor Building siding and the roof decking will blow oiTat 75 l psf. For the siding this was assured through the use of control release fasteners. The FSAR did not j explicitly state that control release fasteners were used on the roof decking and it is esident from design i

review and field inspection that the Reactor Building roof was never intended to be attached similarly by  ;

control release fasteners. Even though control release fasteners are not used (and a psf value cannot be predicted), further design review shows that at some point tornado generated winds between 100 and 300 mph winds will cause the roof decking to blow olT. The Reactor Building as a whole is designed to i withstand normal wind loading of up to 100 mph without any siding degradation, and up to 300 mph  ;

tomadic winds without over-stressing the structural members of the building. This is accomplished by I using the control release fasteners on the siding. The geometry of the roof does not require that its decking be similarly attached in order to preserve the structural steel members (Burns and Roe Civil / Structural calculations of record conservatively assume the roof decking remains intact during a tornado). <

SAFETY EVALUATION: Eliminating the implied requirement that the Reactor Building steel roof decking will blow oft at 75 psf has no efTect on any plant event except a tomado, which is the initiating event. As this change has no impact on any accident or event precursors, it does not increase the probability of any plant events analyzed in the USAR. The USAR has evaluated the consequences of a tomado hovering on the refueling floor, assuming the Reactor Building roof has blown oft. Therefore, whether or not the roof decking will blow off at 75 psf or at some intermediate wind velocity between 100 and 300 mph has no effects on the consequences of the event. Any changes in the probability of roof decking missiles is bounded by the tomado missile analysis and, hence, the probability of occurrence and the consequences of a malfunction of equipment important to safety are unaffected. This change does not create unique precursors to any new transients or accidents not currently described in the SAR. Not requiring the Reactor Building roof to blow off at 75 psfcreates an added reliance on the control release fasteners to shear as required so that the sid ng will blow offin order to preserve the structural integrity of the Reactor Building superstructure.

However, the fasteners are classified as Essential parts with required certified test results and certificates ofconformance. 'Ihus, the postulation that a sizable portion of the 300-400 control release fasteners will not shear and result in suflicient siding not being blown off is not credible. The only Technical Specifications that are remotely related to the Reactor Building roof are those involving Secondary Containment integnty, Scwndary Containment leakage is unalTected during non-tornado winds up to 100 mph. Secondary Containment is not credited during a tomado. Therefbre, no margins of safety are reduced by this USAR change.

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UCR 98-118 (USQE 1998-0090)

TITLE: USAR Change - IIistorical Information in USAR Section XIII-5 DESCRIPTION: His UCR classified Section Xill-5, Startup and Power Test Program, as historical and either made no changes to the USAR text, or restored text that had been previously changed back to the ariginal FSAR content level to provide for a consistent historical understanding. Section Xill-5.0 has b, en expanded to provide an editorial explanation of the format used to identify the portions classified a3 historical USQE 1998-0041 evaluated the acceptability of designating selected infonnation in the USAR as historical, thus eliminating the requirement for maintaining said infonnation up-to-date.

SAFE 1Y EVALUATION: His activity administratively classifies Section XIII-5, Startup and Power Test Program, as historical and no longer subject to updates under the requirements of 10CFR50.71(c). This administrative classification of the Startup and Power Test Program as historical does not constitute either a real or a "de facto" change to the facility or to any underlying analyses. Consequently, there is no effect on any accident precursors and the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is unaffected. Further, as this change has no physical, operational, or analytical eirects on CNS structures, systems, or components, this activity does not introduce the possibility for an accident or malfunction of a difTerent type than previously evaluated. The Startup and {

Power Test Program did not provide an input to the margin of safety as defined in the basis for any l

Technical Specification but instead was designed, in part, to verify that the margins of safety established j by the design and licensing bases were properly translated into the plant. Therefore, this activity has no impact on the margin of safety as defined in the basis of any Technical Specifications.

UCR 98-119 (USQE 1998-011!)

TITLE: USAR Change - Rebaseline of Section 111-5.

5.2 DESCRIPTION

USAR Section 111-5.5.2, Control Rod Drive (CRD) Ilydraulic System, was revised in accordance with the USAR Rebaseline Project. Changes included the following: correction of pressure value at which cooling water to the drives is required from 30 psi to 20 psi above reactor pressure; correction of minimum volume of the scram discharge headers from 1.1 to 3.34 gallons per drive; addition of i requirement for Scram Discharge Volume (SDV) drain and vent valves to isolate the SDV from I Secondary Containment atmosphere during scram such that no single active failure will cause uncontrolled loss of reactor coolant; addition of statement that the CRD pump can take its suction from Demineralized Water in addition to the condensate storage tank; clarification that CRD pumps have both a suction strainer and filter; correction that charging header pressure alarm is a high pressure alarm; addition of information to reflect that the piping between each SDV and its Scram Discharge Isolation Valve is designed to resolve issues docmnented in NRC Bulletin 80-17; and the addition ofinformation that k>ss of nitrogen or accumulator water leakage illuminates a light on a kical accumulator trouble panel.

SAFETY EVALUATION: The subject changes do not involve any modification to existing plant configuration or operation. There are no changes to administrative limits or acceptance criteria for equipment operability. Since the changes do not lead to any increase in the stress on any equipment, there is no increase in the probability of any equipment or system boundary failure which could initiate an accident sequence. The requirements .

i for CRD piping are not changed. The subject changes have no efTect on system performance and will not introduce arry new failure modes. They do not degrade actions or design provisions required to mitigate i the effects of any accident identified in the SAR. None of the changes can increase the probability of a malfunction of equipment important to safety because they do not impact any of the factors that could .

induce a malfunction. The changes do not increase the probability of a failure of the CRD Ilydraulic  ;

System to perform its safety design basi, functions, and do not impact the ability of any systems, l structures, or components to perform any safety functions described in the SAR. No new accident l initiators are introduced and no additional operational considerations are introduced by these changes. l

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r As no new failure modes or mechanisms are introduced, the possibility of a different type of malfunction than previously evaluated is not created. These changes will not reduce the margin of safety as they do not modify the existing plant configuration and have no efTect on existing acceptance limits, limiting conditions for operation, or surveillance requirements.

UCR 99-002 (USQE 1998-0074)

TITLE: USAR Change - Automatic load Following DESCRIPTION: 'lhis UCR disestablishes automatic load following as an allowable method for controlling recirculation flow, it clarifies that while Mode V Turbine Control (Automatic load Following) remains part of the CNS design, it cannot be used by CNS at this time. Mode V was not satisfactorily tested during CNS Startup Testing. Recirculation flow control is performed in the manual mode at CNS. In addition, this UCR replaced Figure Vll-11-1, Pressure Regulator and Turbine Generator Control, with Westinghouse Drawing 4590D53 Sheet' 4, BWR Digital Electro Hydraulic Control System Diagram (which was subsequently incorporated by reference into the USAR). The previous USAR figure was not based on CNS-specific design; the Westinghouse drawing reflects the actual configuration at CNS and is, therefore, a more appropriate reference drawing.

SAFETY EVALUATION: The USAR Chapter XIV transient analysis assumes manual recirculation flow control Therefore, revising the USAR to pmbibit current usage of automatic recirculation flow control is consistent with the current event frequencies analyzed in the USAR. Thus, there is no increase in the probability of i occtuence or conscquences of a previously evaluated plant event. Restricting recirculation flow control  !

to the manual mode is an operational prerogative that is consistent with the plant design. No new failure modes of equipment important to safety are introduced. There are no increased consequences of malfunctions of equipment important to safety since their failures have been analyzed as precursors to previously evaluated plant transients and are, therefore, bounded. Disallowing the current use of automatic recirculation flow control does not introduce any new accident precursors or failure modes.

There are no Technical Specifications that address recirculation flow control. Ifowever, the Bases to Technical Specification 3.4.1 (Recirculation Loops Operating) does state that the flow in each loop is manually controlled. Therefore, this revision to the USAR assures consistency with the Technical Specification Bases and there is no reduction in the margin of safety.

UCR 99-006 (USQE 1998-0088) 1 TrfLE: USAR Change - Feedwater Control Sy stem DESCRIPTION: This UCR made the following changes to the USAR description of the Feedwater Control System: 1) removed information conceming programming of reactor water level because varying reactor water level as a function of reactor power has been electronically canceled at CNS,2) added the word l

" range to the description ofoptimum water level,3) removed the word " corrected" from the discussion of the feedwater flow signal because Minor Design Change 74-80 previously removed the density compensation electronics for feedwater flow, and 4) reordered the discussion of electronic components as the flow control signal is processed through the feedwater control system because the presious description had the dynamic compensator and second proportional amplifier reversed.

SAFETY EVALUATION: The reduction in the total number of electronic components affecting the feedwater control system reduces the probability of an accident or transient occurring from a failure of the control system. Electronically canceling the effect of varying steam flow on reactor level demand signal also increases the reliability of the circuitry by eliminating this function. Removing the electronic components associated with density compensation has improved the reliability of the control system. With less components in the control circuit, the likelihood of a failure will not be increased. loss of Feedwater has been analyzed in Chapter XIV of the USAR as an Abnormal Operational Occurrence. These USAR changes make a loss of

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Feedwater no more likely to occur. A feedwater control failure - maximum demand has also been analyzed in the USAR. These changes have likely caused increased reliability of the feedwater level control system. The dynamic compensator was added to mitigate the effects of certain transients on reactor water level. By preventing tripping of the reactor feed pumps on a load reject, the possibility of a feedwater loss is reduced. None of the changes made contribute to the probability of an accident or transient. Since these changes do not affect the accident source term and accident mitigation systems continue to operate as credited in the accident analysis, there is no increased dose and no increase in accident consequences. Interfacing systems will not be more likely to malfunction as a result of these changes and no increase in challenges or degradation of safety system performance can be attributed to this activity. The functions of the feedwater level control system are not credited by any safety-related equipment. This activity will increase reliability and transient response causing the feedwater control system to better cope with malfunctions of equipment important to safety. These changes do not create unique precursors to any new transients or accidents not currently described in the USAR. Analyses, operational controls, and maintenance practices are based on the as-built system; thus, no different types ofmalfunctions than previously evaluated are introduced. This activity does not contradict any bases for the Technical Specifications, nor erode any margin of safety.

UCR 99-008 (USQE 1999-0018)

TITI.E: USAR Change - Multipurpose Facility (MPF)llVAC System DESCRIPTION: This UCR conected inaccuracies in the description of the MPF IIVAC System. Changes inchaled: 1) removing an automatic start feature of the MPF exhaust fan #2 which utilizes two redundant Iligh Efliciency Particulate Absolute (IIEPA) filters,2) removing an automatic stop feature (switching) of MPF exhaust fan #1, and 3) removing the automatic closing and opening of associated dampers - all in response to a radiation signal from the main ventilation exhaust radiation monitor. These changes bring the USAR discussion of the MPF IIVAC system in line with actual plant design / construction.

There is only one IIEPA filter train and the MPF effluent radiation monitor has no automatic control function (s); it provides indication and alarm functions only.

SAFETY EVALUATION: This activity ckx:s not increase the likelihood of precursor events to any transients or accidents described in the SAR such that their frequency classification is changed. The subject radiation monitor and associated 1IVAC system does not initiate or act as a precursor in any accident or transient. This activity does not degrade or prevent actions assumed or described in the accidents or transients discussed in the SAR. Since accident mitigation systems will continue to operate as credited in the accident analysis, there is no increase in accident consequences. Functions of the MPF radiation monitor or the MPF IIVAC system are not credited by any safety-related equipment, and there is no direct or indirect interface with safety-related equipment. 'Iherefore, there is no increase in the probability of occurrence or consequences of a malfunction ofequipment important to safety. These changes do not create unique precursors to any new transients or accidents that are not currently described in the SAR. Operational controls and maintenance practices are based on the as-built system. Revising the USAR to conform to the way the system was actually constructed does not offer new possibilities of a difTerent type of malfunction than previously evaluated. There are no Technical Specifications related to this activity; therefore, it does not contradict the basis for any Technical Specification, nor erode any margin of safety.

UCR 99-009 (USQE 1999-0011)

TITI,E: USAR Change - Appendix D(1), Quality Assurance Program DESCRIPTION: This UCR revised Appendix D(1) of the USAR, which describes the Construction Quality Assurance Program, to identify it as historical information and to restore the language to that in the FSAR. It also

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deleted infonnation pertaining to classification of systems, components, and structures that is redundant to Appendix A, and added a reference to Appendix A. l

5. 'TY ELUATION: This UCR has no cirect on any accident precursors and there is no increase in the consequences of previously evaluated accidents. The characterization of the designated text as historical has no physical, operational, or analytical efTect on any plant equipment. Therefore, there is no increase in the probability of occurrence or consequences of a malfunction of equipment important to safety and no new types of accidents or malfunctions are introduced. The purpose of the Construction Quality Assurance Program {

was in pad to verify that the presumed margins of safety existed in the plant. Ilowever, the program itself I did not define any margins of safety that formed the bases for any Technical Specification.

UCR 99-010 l (USQE 1999-0021)

TITLE: USAR Change - Organizational Structure DESCRIPTION: This UCR revised USAR Section XIII to reflect various changes made to the Nuclear Power Group organizational structure. This included changes in job titles, responsibilities, and reporting relationships.

Among the changes was the reassignment of responsibility for the corrective action program from the I Plant Manager to the Senior Manager of Site Support. This resulted in a change of a previous commitment made in NLS970215, Response to a Notice of Violation and Proposed Imposition of Civil J

Penalty dated December 31,1997. This change will allow the plant operating /line organization to better l

focus on the operation of the plant, yet maintain a strong management focus and attention on the  !

corrective action program.

SAFETY EVALUATION: The subject changes are admmistrative in nature and do not delete or afTect the overall functional aspects ofexisting programs, processes, procedures, or activities at CNS, nor do they afTect the design, function, system interface, or operating parameters of plant systems, stmetures, and components (SSCs) important to safety. They do not affect any assumptions for accidents evaluated in the SAR. The changes do not degrade or prevent mitigative actions described or assumed in the SAR, and do not impact fission product barriers. On-site and oft-site dose consequences are not affected. The changes will not degrade the perfonnance of safety systems assumed to function in the SAR accident analysis, increase challenges to safety systems, or revise system line-ups; therefore, the probability of occunence or consequences of equipment malfunction are not increawd No equipment is being modified or installed by this change.

No new accident initiators or equipment failures are introduced, therefore, the probability of a difTerent type of accident or equipment malfunction is not created. The changes do not alTect the margins of safety evaluated against Safety Limits, Limiting Safety System Settings, Limiting Conditions for Operation, or design parameters for SSCs. Organizational requirements and unit stalTqualifications are also unafTected by these changes.

UCR 99-014 (USQE 1999-0025)

TITLE: USAR Change - Reorganization of Engineering Depanment

! DESCRIPTION: 'Ihis UCR revised USAR Section XIII to relicct the reorganization of the Plant Engineering Department l l

(PED) and Engineering Support Department (ESD) to more efTectively align the departments with their l support ofplant operation functions. PED will support the Operations organization with systems issues j and ESD will support the Maintenance organization with components and programs issues. The changes involve changes in job titles, responsibilities, and reporting relationships.

SAFETY EVALUATION: The subject changes are administrative in nature and do not delete or alTect the overall functional aspects ofexisting programs, pnx: esses, procedures, or activities at CNS, nor do they afTect the design, function, system interface or operating parameters of plant systems, structures, and components (SSCs) important to safety. They do not afTect any assumptions for accidents evaluated in the SAR. The changes do not

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degrade or prevent mitigative actions described or assumed in the SAR and do not impact fission product barriers. On-site and off-site dose consequences of an accident are not affected. The changes will not degrade the performance of safety systems assumed to function in the SAR accident analysis, increase

{

1 challenges to safety systems, or revise system line-ups; therd;re, the probability of occurrence or consequences of equipment malfunction are not increased. No equipment is being modified or installed by this change. No new accident initiators or equipment failures are introduced; therefore, the probability of a difTerent type of accident or equipment malfunction is not created. The changes do not afTect the margins of safety evaluated against Safety Limits, Limiting Safety System Settings, Limiting Conditions ,

for Operation, or design parameters for SSCs. Organizational requirements and unit stafT qualifications are also unaffected by these changes.

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j US AR Chance Recuests (UCRS197-147.97-148. 97-149.97-150. 97-159.97-162. 97-167.97-168. 98-001.98-008.

l 98-017.98-018. 98 019 (nartialt 98-020.98-022.98-023.98-024.98-026.98-031.98-036.98-037.98-044.98-049. l 98-057.98-065.98-066.98-067.98-075  !

l TITLE: Editorial USAR Changes l DESCRIPTION These change requests made various editorial changes to the USAR. Included in these editorial changes 1

were the following: 1) correction of continuation points on USAR drawings,2) correction of drafling errors on USAR drawings,3) correction of typographical errors,4) correction to make USAR Figure  ;

match USAR text with regard to power supply configuration of Residual IIcat Removal pumps,

5) correction / addition of valve nomenclature on USAR Figures,6) correction of three Design Change numbers that are USAR references,7) correction of references to USAR Figure numbers,8) minor corrections to USAR Table VIII-5 1 to coincide with Calculation NEDC 87-104A, which is the source j document for the USAR Table,9) correction of a reference to a USAR subsection,10) establishment of i consistent nomenclature for the Residual IIcat Removal Service Water Booster Pump Fan Coil Unit,

11) restructuring of sentences for clarity, and 12) addition of a check valve symbol to the USAR figure I depicting symbols and abbreviations.

SAFETY EVALUATION: All changes are editorial in nature. They have no adverse effect on nuclear, industrial, or personnel safety.

They have no impact on any system, structure, or component (SSC) such that it would increase the probability ofoccurrence of an accident or transient or have any effect on the ability of an SSC to mitigate any plant accident or transient. Thus, there is no impact on the radiological consequences of a plant event. These UCRs do not make changes to equipment important to safety that have the potential of increasing the probability of occurrence or consequences of equipment malfunction. No new types of accidents or equipment malfunctions are introduced by the incorporation of editorial USAR changes.

l These UCRs result in no physical changes to the plant and do not alTect any margins of safety. l DCN 96-0111 DCN 96-Ol l2 (USQE 1999-0014) l TITLE: Changes to General Electric Drawing 729E589BB, Sheets 1 and 2 DESCRIPTION: Drawing 729E589BB is the functional control diagram for the IIigh Pressure Coolant Injection (IIPCI) system. Sheets 1 and 2 of this drawing are incorporated in the USAR by reference. This drawing series was revised to correct various logic discrepancies to make it agree with the as-built condition of the IIPCI system as shown in General Electric elementary diagram 791E271 series drawings, No physical or functional changes to the IIPCI system or any ofits components were involved.

SAFETY EVALUATION: These changes will not change the way the IIPCI system is operated or maintained. They will not add any automatic control function to the HPCI system or its components. These changes do not affect any radiation monitoring equipment and do not involve any parameters associated with radiation monitoring.

They do not afTect the safety or design function of the IIPCI system. No new equipment will be added in the plant. These revisions will not change, degrade, or prevent actions required for any accident or

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alter the assumptions made in evaluating the consequences of a malfunction of equipment important to safety. The reliability of the iIPCI system is not decreased. These changes will not impact the discussion of the IIPCI System Control and Instrumentation in the USAR. Therefore, there is no increase in the probability of occurrence or consequences of a plant event or malfunction of equipment important to safety previously evaluated in the SAR, and no new or ditrerent types of plant events or equipment malfunctions are created. The IIPCI system is discussed in the Technical Specifications, but not in the level of detail that will be alrected by this activity. The changes initiated by these DCNs do not impact the margin of safety as defined in the basis for any Technical Specification.

DCN 96-0113 DCN 96-0114 DCN 96-0116 (USQE 1999-0015)

TITLE: Drawing Changes to General Electric Drawing 730E140BB, Sheets 1,2, and 3 DESCRIPTION: Drawing 730El40BB is the functional control diagram for the Residual Ileat Removal (RI1R) system.

Sheets 1,2, and 3 of this drawing are incorporated in the USAR by reference. This drawing series was  ;

revised to correct various logic discrepancies to make it agree with the as-built condition of the RIIR l system as shown in the General Electric elementary diagram 791E26I series drawings. No physical or functional changes to the RIIR system or any ofits components were involved. l SAFETY '

EVALUATION: 'Ihese changes will not change the way the RIIR system is operated or maintained. They will not add any automatic control function to the RIIR system or its components. These changes do not alTect any radiation monitoring equipment and do not involve any parameters associated with radiation monitoring.

They do not affect the safety or design function of the RIIR system. No new equipment will be added in i the plant. These revisions will not change, degrade, or prevent actions required for any accident or alter l the assumptions made in evaluating the consequences of a malfunction of equipment important to safety.

The reliability of the RIIR system is not decreased. These changes will not impact the discussion of the RIIR System Control and Instmmentation in the USAR. Therefore, there is no increase in the probability of occum:nce or consequences of a plant event or malfunction of equipment important to safety previously evaluated in the SAR, and no new or difTerent types of plant events or equipment malfunctions are created. The RIIR system is discussed in the Technical Specifications, but not in the level of detail that will be alTected by this activity. These changes do not involve any parameters discussed in the Instrumentation and RIIR sections of the Technical Specifications. Therefore, the changes initiated by these DCNs do not impact the margin of safety as defmed in the basis for any Technical Specification.

DCN 96-0117 DCN 96-Ol l8 (USQE 1999-0013)

TITLE: Changes to General Electric Drawing 729E517BC, Sheets 1 and 2 DESCRIPTION: Drawing 729E517BC is the functional control diagram for the Reactor Core Isolation Cooling (RCIC) system. Sheets 1 and 2 of this drawing are incorporated in the USAR by reference. This drawing series was revised to correct various logic discrepancies to make it agree with the as-built condition of die RCIC system as shown in General Electric elementary diagram 791E264 series drawings. No physical or functional changes to the RCIC system or any ofits components were involved.

SAFETY EVALUATION: These changes will not change the way the RCIC system is operated or maintamed. They will not add l any automatic control function to the RCIC system or its components. These changes do not alTect any radiation monitoring equipment and do not involve any parameters associated with radiation monitoring.

They do not afTect the safety or design function of the RCIC system. No new equipment will be added in the plant. The revisions will not change, degrade, or prevent actions required for any accident or alter the assumptions made in evaluating the consequences of equipment malfunction. They do not decrease

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I the reliability of the RCIC system. These changes do not impact the discussion of RCIC System Control and Instrumentation in the USAR. Therefore, there is no increase in the probability of occurrence or consequences of a plant event or malfunction of equipment impodant to safety previously evaluated in the SAR, and no new or different types of plant events or equipment malfunctions are created. The changes initiated by these DCNs do not impact the values of instrument settings as discussed in the Technical Specifications. Hey do not involve any parameters discussed in the Instmmentation and RCIC sections of the Technical Specifications. Therefore, there is no reduction in the margin of safety as i defined in the basis for any Technical Specification.

I DCN 96-2234 i (USQE 1998-0009) ,

i TITLE: Change to Bums & Roe Drawing 2045, Sheet 2 DESCRIPTION: Drawing 2045, Sheet 2,is the flow diagram for the Standby Liquid Control (SLC) system. This drawing l is incorporated in the USAR by reference. The drawing was revised to reflect that the piping from the (

l SLC mixing tank to the SLC storage tank shall be seismic Class llS, as installed by Design Change (DC) )80-138.

SAFETY EVALUATION: This change is in accordance with the design requirement of DC 80-138 and will not cause a change to l l any system interface in a way that could increase the likelihood of an accident. There will be no change to any structure, system, or component in the plant, nor will there be any physical or operational changes to the plant. The SLC mixing tank and associated piping does not play a direct role in mitigating the radiological consequences of an accident described in the SAR. Dose to olTsite or plant personnel will not be affected. This change does not atrect the design function of the SLC system and no equipment will be added or modified, therefore, there is no increase in the probability of a malfunction of equipment -

impodant to safety. The SLC system is not an Essential system. The text discussion of the SLC system in the USAR is not impacted. No new types of accidents or equipment malfunctions are introduced by the classification of this piping. The SLC system is discussed in the Technical Specifications; however, l the SLC mixing tank and associated piping are not discussed. Therefore, this DCN does not redece the J

margin of safety as defined in the basis for any Technical Specification.

DCN 98-0414 (USQE 1998-0104)

TITLE: Change to Burns & Roe Drawing 2079, Sheet 2 DESCRIPTION: Drawing 2079, Sheet 2,is the Flow Diagram for the Augmented Liquid Radwaste system. Thh drawing is incorporated in the USAR by reference. It was revised to identify that only one process vent line for the radwaste tank vent system ties into the augmented oft-gas dryer beds located in the overhead of the Augmented Radwaste Building pipe gallery, instead of the two lines previously shown on the drawing.

The as-built configuration of the affected vent line is part of original construction under Contract E73-58.  ;

This DCN does not require any physical or operational changes to the plant.

, EVALUATION: ne Augmented Liquid Radwaste system is located in the Seismic Class 11 Radwaste Building extension.

l There is no essential or safety-related equipment in the area. There will be no change to any structure, system, or component in the plant. The plant personnel or offsite dose will not be alTected by this change.

There will be no equipment added or any existing equipment modified. No plant procedures are impacted. The process vent piping is non-essential and provides no safety related function. The change l

will not afrect the design function of the Augmented Liquid Radwaste system. The vent line for the Augmented Liquid Radwaste system is not discussed in the Technical Specifications. Therefore, there is no increase in the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR, no new types of accidents or equipment i malfunctions are created, and there is no increase in the margin of safety as defined in the basis for any l

Technical Specification.

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3 l

I I DCN 98-1479 (USQE 1998-0105)

TITLE: Change to Cosmodyne Drawing 6000302, Sheet 2 l

l DESCRIPTION: Drawing 6000302, Sheet 2, is the Piping and Instrumentation Diagram for the Augmented Off-Gas  !

- (AOG) system. This drawing is incorporated in the USAR by reference. It was revised to show the '

l addition of a drain valve in the radwaste tank vent system and to reflect the as-built configuration of the 1

process vent piping. The as-built configuration of the afrected vent line and drain valve is part oforiginal construction under Contract E73-58. This DCN does not require any physical or operational changes to the plant.

SAFETY EVALUATION: The AOG system is a Seismic Class II, non-essential system. The 3/4" drain valve being added to the drawing is normally closed, manually operated, and used to drain the vent header. There will be no change to any structure, system, or component in the plant. There is no essential or safety-related equipment in the area that can be impacted by this drawing change. Plant personnel or offsite dose will l

not be affected. There will be no equipment added or any existing equipment modified. The drain valve '

and process vent piping are non-essential and provide no safety-related function. The design function of the AOG system is not affected. The vent line and drain valve for the AOG system are not discussed in the Technical Specifications. Therefore, there is no increase in the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the j SAR, no new types of accidents or equipment malfunctions are created, and there is no increase in the margin of safety as defined in the basis for any Technical Specification.

l l

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p PROCEDURE CHANGES t

l Procedure 0.39 (Revision 14) I TITLE: Fire Watches DESCRIPTION: 'Ihis revision of the admmistrative procedure for regulating fire watch activities included: 1) creation of  ;

a new hot work permit,2) direction how to complete the hot work permit, 3) reference changes resultmg l from implementation ofImproved Technical Specifications, and 4) various administrative changes.  !

SAFETY.

EVALUATION: This activity does not change the state or function of safety related structures, systems, or components; therefore, it will not alter any of the inputs or assumptions for the probabilities of previously evaluated accidents. These changes do not change the function, performance, or integrity of any boundaries with which safety related systems form or support the primary protective barriers on which the consequences l

ofpreviously evaluated axidents are based. The changes to this procedure will not induce failure of any i equipment important to safety. The subject changes will not afTect equipment important to safety such I that it would operate differently than it has previously operated. This activity does not contribute to any of the initiation sequences or failure modes analyzed in the USAR and does not introduce any new event initiators or failure modes. The margin of safety which is established by the design and performance of i safety related systems is not reduced by this actisity. l Procedure 2.1.4.1 (Revision 0) l (USQE 1998-0009)

TITLE: Rapid Shutdown DESCRIPTION: This new procedure was developed to provide instruction for Operations personnel to perform a rapid shutdown from reactor power operation. It will only be used to meet LimiGig Condition for Operation Required Action Completion Times when shutdown completion time requirements cannot be met using Procedure 2.1.4, Normal Shutdown From Power. The procedure provides for a controlled shutdown to 30% power at which time the reactor is scrammed.

SAFETY EVALUATION: This activity will not change the design, function, or reliability of the Nuclear Boiler, Fuel, and Main Turbine systems This new rapid shutdown procedure implements the applicable steps out of two approved station procedures (i.e., Procedure 2.1.4 for normal shutdown and Procedure 2.1.5 for emergency shutdown), thus this new activity will not be an accident initiator. This procedure does not perform any accident mitigation functions nor will it result in any increased radiological efTects. This activity will not change the design, function, operation, or reliability of any equipment important to safety, nor willit induce any equipment malfunctions or failures. It will not change or affect the way equipment operates, only when it will operate. The effects of this procedure are bounded by the existing emergency shutdown procedure. Overall this procedure will place less stress on the asscciated equipment affected by shutdown than if the emergency shutdown procedure was used. No new failure modes are introduced since this procedure does not change the operation of the Nuclear Boiler, Fuel, and Main Turbine

- systems, nor does it change any existing interfaces between these systems and any other equipment and/or systems. Operating parameters will not be exceeded such as cooldou n rate of the reactor vessel, power

' reduction, or the load reduction rate to the main turbine. This activity does not affect any assumptions, calculations, procedures, or design specifications used to establish the basis for defining the plant's margm of safety. The effects of shutting down the plant at an accelerated rate from normal shutdown are bounded by the emergency shutdown procedure 2.1.5 and as such does not reduce the margin of safety as defined in the basis for any Technical Specification.

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Procedure 2.1.20 (Revision 33)

Procedure 2.2 84 (Revision 26) 4 Procedure 6 SC.201 (Revision 9) I (USQE 1998-0048,1998-0049,1998-0050)

TITL.E: Reactor Pressure Vessel (RPV) Refueling Preparation (2.1.20)

IIVAC Main Control Room and Cable Spreading Room (2.2.84) t Secondary Containment (Reactor Building II&V) Valve Operability Test (6.SC.201) I DESCRIPTION: ne failure of the Reactor Building isolation valves to meet the design basis requirements for Secondary i Contamment was identified as a nonconforming condition (reference License Amendment Request dated l August 6,1998). Operability Evaluation (OE) 2-24337 was subsequently prepared in accordance with Generic Letter 91-18, Revision 1. Temporary changes were made to the subject procedures to document l limitations and restrictions identified in the OE to support refueling activities. The change to Procedure 2.1.20 requires that the Control Room Emergency Filtration System (CREFS) be placed in sersice prior to the step to liA the vessel head. The change to Procedure 2.2.84 addresses the requirements to place CREFS in service and provides the actions to be taken in the event CREFS is lost. The change to Procedure 6.SC.201 changes the IST Retest Limit and the Operability Limit for valves IIV-MO-258 and  ;

IIV-MO-260 to 60 seconds to reflect the assumptions used in the calculations in support of OE 2-24337. 1 SAFETY EVALUATION: This change does not affect or alter the precursors to any accident or malfunction of equipment important to safety previously evaluated in the SAR. The limitation to restrict fuel movement or the movement of heavy loads during the first seven days following reactor shutdown does not affect any operating mode l or plant system. The restrictions to require the isolation of the Control Room or the isolation of the

)

Reactor Building and starting Standby Gas Treatment do not affect the method of fuel handling or handling of heavy loads therefore, they cannot inct, ase the probability of the refueling accident or an

, I equipment malfunction. The impact of the refueling acc_f nt on the CNS Control Room operators is not evaluated in the CNS USAR. A calculation was performed in order to address the issue of the impact on Control Room operator thyroid dose. In order to ensure that the guidelines specified in General Design Criteria (GDC) 19 are not exceeded, limitations and restrictions were identified. The consequences of the refueling accident on offsite dose would also be reduced by limiting the maximum Reactor Building iI&V exhaust flow rate, and by limiting the valve stroke time to 60 seconds or less. The consequences of an equipment malfunction in conjunction with the refueling accident are within the guidelines of GDC 19 and offsite dose would also be reduced by the limitations and restrictions. The limitations and restrictions imposed by the temporary procedure changes do not create any new operating modes and do not affect physical configuration of the plant. Therefore, the possibility of a different type of accident or equipment malfunction than previously evaluated in the SAR is not created. These changes do not affect the automatic function of the Reactor Building ventilation system for maintaining negative pressure on Secondary Containment. He restriction that CREFS be placed in service prior to the lifting of the reactor vessel head, and maintained in service until 266 hours0.00308 days <br />0.0739 hours <br />4.398148e-4 weeks <br />1.01213e-4 months <br /> after reactor shutdown, does not affect the ability of CREFS to meet its stated function as identified in the Technical Specification Bases. Based on this evaluation,it was determmed that the margin ofsafety defined in the basis of the 1echnical Specifications is not reduced.

Procedure 2.1.20.2 (Revision 01 Procedure 2 2 32 (Revision 28)

Procedure 2.2.69.2 (Revision 27)

(USQE 1998-0064)

TITLE: Cycle Specific Fuel Transfer and Alternate Cooling Guideline (2.1.20.2)

Fuel Pool Cooling and Demineralizer System (2.2.32)

Residual Ileat Removal (RIIR) System Shutdown Operations (2.2.69.2)

DESCRIPTION: Procedures 2.2.32 and 2.2.69.2 contained a limitation that during refueling operations, one Fuel Pool Cooling (FPC) pump could be removed from service provided the B Residual IIcat Removal (RHR) 109 E

subsystem RIIR-FPC System intertie was available. This limitation was removed and replaced with a '

reference to new Procedure 2.1.20.2. This new procedure provides information to support the cycle-specific transfer limitations for the movement of fuel from the reactor vessel to the spent fuel pool.

It provides cycle specific calculational results associated with fuel oft-load rates, alternate reactor casity !

and spat f el pool cooling, and alternate fuel pool cooling. It considers variables such as the number of buxb in the fuel pool fmm previous reloads, time after reactor shutdown, Reactor Equipment Cooling and Service Water temperatures, as well as various configurations of FPC.

{

SAFETY

{

EVALUATION: The changes to Procedures 2.2.32 and 2.2.69.2, and new Procedure 2.1.20.2, do not afTect any accident

{

precursors or initiators. The procedure changes do not affect the design basis or normal operating temperatures of the system. By maintaining temperatures within the design limits, no fuel damage can occur. Since there will be no inen ase in the probability for fuel damage due to overheating, there can be no increase in oft-site dose and thus no increase in the consequences of an accident. The ability of the spent fuel pool cooling system to maintain fuel pool temperatures within the design requirements is established and ckicumented in the USAR. These procedure changes do not change the method in which the fuel pool cooling system operates or introduce any new operational or failure modes. Therefore, the probability of a malfunction ofequipment important to safety is not increased. The loss of spent fuel poo!

cooling is described in the USAR and the NRC Safety Evaluation Report for License Amendment No.

52. The mitigation strategies for coping with the loss of spent fuel pool cooling remain unchanged. The time available to implement the licensing basis coping strategies and the flow capacities for the methods available for boil-offmakeup are not changed. No new types of accidents or equipment malfunctions are created. The margin ofsafety as defined in the basis for any Technical Specification is not reduced. The procedure changes do not affect the physical performance characteristics of plant equipment and do not affect any of the normal operating bands or temperature limits for the operation of plant systems. There are no specific Technical Specification requirements for the number of spent fuel pool cooling system components that are required to be operable or in operation for particular plant operational states. There are no requirements in the Technical Specifications relative to the operation of the RHR system spent fuel cooling intertie.

Procedure 2 2.10 (Revision 16)

Procedure 2 2.10A (Revision 51 Procedure 2 3 2 58 (Revision 11 Procedure 6 FP 302 (Revision 41 TITLE: Deluge and Sprinkler System (2.2.10)

Deluge and Sprinkler System Component Checklist (2.2.10A)

Oflice Building Fire Detection Panel FP-PNL-OF (2.3.2.58)

Automatic Deluge and Pre-Action Systems Testing (6.FP.302)

DESCRIPTION: These pmcwures were revised to allow the Air Compressor FP-CPSR-SACI to operate in the automatic maie. This air compressor provides supenisc7 af to pre-action systems 41 (Office Building Records Storage Vault) and 42 (Reactor Buildmg RailroaC hlock). Presiously, the air compressor was operated manually whenever air was required to be replenished due to air leaks in the system piping or dwing testing. The system was designed for auto operation as described on design / installation drawings. Auto operation minimizes operator distractions and nuisance alarms caused by minor system leakage which is undetectable through conventional methods, and restores the system to National Fire Protection Association Code conformance SAFETY EVALUATION: This activity does not change the state or function of safety related structurer. ,ystems, or components; therefore, it will not alter any of the inputs or assumptions for the probabilities of previously evaluated accidents. It does not change the function or performance of any boundaries with which safety related systems form or support the primary protective barriers. These changes are to procedures that are used to perform testing, valve line-ups, component checks, and alarm response for the subject pre-action sprinkler systems which do not contain safety related systems or components. The changes do not alter the design basis for the subject pre-action sprinkler systems, nor do they adversely affect initiating

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sequences or starting setpoints of safety related equipment. Thus, this activity does not increase the probability of a malfunction of equipment important to safety previously evaluated in the USAR. This activity assures reliability of fire barriers to function when needed to mitigate the consequences of the spread offire. It (kies not affect equipment important to safety such that it would operate differently than equipment important to safety has previously operated. No new initiators are introduced and the possibility of a different type of plant event than previously evaluated is not created. No new failure modes are introduced and the probability of a malfunction of a different type is not created. The margin of safety which is established by the design and performance of safety related systems is not reduced.

Procedure 2.218 (Revision 50)

(USQE 1998-0087)

TITLE: 4160V Auxiliary Power Distribution System DESCRIPTION: This system operating procedure was revised to remove the paddle from 32/1 FE (32/lGE) overpower relaysjust prior to synchronizing the F Bus (G Bus) to the A (B) Bus and the grid. After the two buses are synchronized and the Diesel Generator (DG) load is stabilized, the paddle will be returned to the relay. The total time the paddle is removed will be minimal. The paddle must be removed to allow DGl and the F Bus to be synchronized to the A Bus and the grid (DG2 and the G Bus to be synchronized to the B Bus and the grid). Otherwise, the 32/lFE (32/1 GE) relays will trip instantaneously and prevent the two buses from being synchronized.

SAFETY EVALUATION: The probability of occurrence of a previously evaluated accident is not increased. The change in the system operating procedure will not increase the probability of a Loss of Offsite Power (LOOP). The operating status of the diesel generator or the overpower relay is not an initiating event for a LOOP. This change in the operating procedure will not create a new pathway to release radioactive material and the function ofequipment required to control the release of radioactive material is not affected. There is no reduction in the protection ofpublic health and safety and no increase in the consequences of an accident or equipment malfunction previously evaluated in the USAR. This change will allow the F (G) Bus to be synchronized to the A (B) Bus, which is the normal power supply. Using the A (B) Bus as the normal power supply assures that the greatest number ofredundant power supplies to the F (G) Bus are available.

No other circuitry, such as the undervoltage protection, load shedding, and load sequencing, is afTected by the change to this procedure. The operability and function of the redundant diesel and bus is not alTected by the change to this procedure. No new or different type of accidents or equipment malfunctions areintrocluced by this procedure change. The LOOP is an event that has been analyzed in the SAR. The failure of a DG and the loss of a safety related bus has also been previously evaluated in the USAR. The margin of safety as defined in the basis for any Technical Specification has not been reduced. The redundant DG, safety related bus, and associated equipment can safely shut down the reactor, maintain the safe shutdown condition, and operate all auxiliaries necessary for station safety.

Procedure 2 2.20 (Revision 44)

(USQE 1998-0060)

TITLE: Standby AC Power System (Diesel Generator)

DESCRIPTION: This procedure was revised to be used in conjunction with the Clearance Order process to reconfigure the Diesel bencrator Starting Air (DGSA) compressor unloading valve assembly air supply line. The configuration change is needed to maintain the redundant compressor available and as a personnel safety boundary when mechanical maintenance is performed.

SAFETY EVALUATION: The DGSA compressors are non-essential components and are not contributors to any accident described in the SAR. They also do not mitigate the consequences of any accident described in the SAR. The reconfiguration of the DGSA compressor unloading valve assembly air supply line will maintain one compressor available to charge the air receivers. The DGSA compressor unloading function pr : vents excessive wear on the compressor during starting of the compressor. Even if a compressor sbcaid start

-111 L .

l with the air supply to the unloaders valved out, there is no failure mechanism associated with this function that could affect safety related equipn..nt in the area. There is no failure mechanism associated with this change that could create a different type ofaccident or malfunction ofequipment important to safety. The DGSA compressors are not described or relied upon by any Technical Specification.

Procedure 2 2.65A (Revision 15)

TITLE: Reactor Equipment Cooling Water System Component Checklist DESCRIPTION: This procedure was revised to maintain REC-V-269 in the normally closed position. A Caution Tag Order (CTO) was previously in place to maintain this Reactor Equipment Cooling (REC) valve in the ckised position. REC-V-269 is the lower root valve for the REC Surge Tank level indicator, REC-LG-488. The valve was ticing maintained in the closed position per the CTO because REC-LG-488 is a non-seismic qualified gage and the root valve must be closed as it acts as a pressure boundary for the REC system. This revision to the REC system valve checklist allows the CTO to be removed.

SAFETY EVALUATION: This activity will not alTect system performance and reliability or any system interface in a way that could lead to an accident. Changing the normal position of REC-V-269 to the closed position does not afTect the safety design basis fbr the plant as outlined in the SAR. This activity does not change any instrument accuracy or response characteristic in a manner that could increase the probability of an accident. It does not increase the possibility of operator error or add complexity to human factor conditions such that the probability of an accident is increased. The function of equipment designed to control the release of radiation is unaffected. The REC system will retain its design function and the REC Surge Tank level indicator will also function as designed. The REC Surge Tank level indicator is a nonessential component and has no active safety function which can be afTected. To ensure REC-V-269 is maintained normally closed to protect the REC system pressure boundary, necessary operator instructions were added to Proecdure 2.1.11, Station Operators Tour, to close the valve after a surge tank level reading has been taken. Therefore, this activity will not alTect the capability of any equipment important to safety to perform its design function and the probability of a malfunction of equipn'ent important to safety is not increased. No new types ofplant events or malfunctions of equipment important to safety are introduced.

REC Surge Tank level will continue to be monitored and recorded on a daily basis per Technical Specifications. In addition, the limit stated in Procedure 6. LOG.601, Daily Surveillance Log (Technical Specifications), for REC Surge Tank minimum vahtme and the requirement for daily monitoring of REC system inventory will be unafrected. The ability of the REC Surge Tank to provide sufficient volume for the REC system following a Loss of Coolant Accident is unalrected. Therefore, this activity does not l reduce the margin of safety as defined in the basis of any Technical Specification. t Procedure 2.2.69.2 (Revision 28)

(USQE 1998-0063) i TITLE: Residual iIcat Removal (R1IR) System Shutdown Operations l

DESCRIPTION: A new section was added to this procedure to establish Reactor Pressure Vessel (RPV) natural i circulation. A calculation was perfonned which demonstrates that with no forced circulation through the core, natural circulation is an altemate method of coolant circulation during refueling conditions. Ilaving an alternate means of coolant circulation is not only desirable to support maintenance activitics, but necessary to ensure compliance with Technical Specifications under certain conditions. Specifically, Limiting Condition for Operation 3.9.7, RIIR 1ligh Water Level, requires attemate cookint circulation to be established when no RIIR Shutdown Cooling (SDC) subsystems are operating.

SAFETY  !

! EVALUATION: During natural circulation, the plant will be in Mode 5 (refueling) with the RPV head removed, reactor l f well and pit flooded up, and the fuel pool gates removed. Based upon these specific conditions, many of l the accidents and n'o normal transients evaluated in the SAR are not relevant to the proposed procedure change. RIIR SDC is not required to mitigate any events or accidents evaluated in the SAR. Ilowever, the following were specifically considered due to their applicability in Mode 5: Refueling Accident.

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r Control Rod Removal Error During Refueling, Fuel Assembly Insertion Error During Refueling, Misoriented Bundle Accident, and Loss of RIIR Shutdown Cooling. The probability of dropping a fuel bundle on the core is not increased by this change. Refueling interlocks that prevent any condition which could lead to inadvertent criticality (due to a control rod withdrawal error or fuel assembly insertion error) i during refueling are not affected by this change. The design features associated with fuel orientation are I not affected. The SAR categorizes the Loss of SDC event as one which can cause a reactor vessel water temperature increase. This procedure change establishes the conditions for assuring that coolant temperature can be maintained within limits under reduced (natural circulation) core flow rates. This change does not increase the probability oflosing RIIR SDC. The consequences of dropping a fuel bundle are unaltered. The consequences associated with reactivity mismanagement events are not affected, all procedures and controls regarding fuel movement and control rod removal remain unchanged.

%e consequences ofreloading a fuel bundle incorrectly are unchanged because the event is bounded by the assumption of the most reactive core conditions. The consequences of a loss of SDC are unafTected, any form of water makeup can be used to maintain water level and assure core cooling. In fact, reliance on natural circulation is implicit during a loss of SDC flow. The probability of occurrence or consequences of a malfunction ofequipment important to safety previously evaluated in the SAR will not be increased. Fuel handling equipment, secondary containment, refueling interlocks, fuel handling processes, and RIIR SDC equipment are not adversely affected by this activity. The consequences of mechanical cladding damage as a result of a fuel bundle being dropped on the core remain the same. The consequences of reactivity mismanagement are not affected by the absence of forced flow. The consequences of a loss of SDC under the described plant conditions are not affected, because no forced flow is assumed upon a loss of SDC. The possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated is not created. During natural circulation operations, RIIR pumps and Reactor Recirculation pumps will be off. Cooling systems will be configured per approved operating procedures. Ample water will be covering the fuel in the RPV to

{'

assure continued cooling. The margin ofsafety as dermed in the basis for any Technical Specification will not be reduced. This change complies with the Required Actions of Technical Specification 3.9.7. If no RIIR SDC subsystem is in operation for greater than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, an alternate method of coolant circulation is required to be available and reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. Analysis indicates that measurable criteria for verifying natural circulation exist and that this ahernate coolant circulation, coupled with Fuel Pool Cooling operation, affords appropriate monitoring of coolant temperature.

Procedure 2.2 69.2 (Revision 29)

(USQE 1998-0069)

TITLE: Residual IIcat Removal (RIIR) System Shutdown Operations DESCRIPTION: This procedure was revised to provide guidance for opening the RIIR crosstie valve (RIIR-MOV-20) l to allow pumps in one kiop to discharge through the opposite recirculation loop. In this configuration, there will be no RllR loop available for the Low Pressure Coolant Injection mode of operation.

Therefore, a note has been added to the procedure which specifies this limitation and a step added to ensure compliance with Technical Specification Limiting Conditions for Operation.

SAFETY j

EVALUATION: This cb-nge in system configuration does not aiTect any accident precursors for accidents presiously l evaluated in the SAR. This procedure change does not afTect the design basis of the system, and does not I affect the ability of the Emergency Core Cooling System (ECCS) systems to perform their design safety function. This procedure change to allow the use of the RIIR crosstic valve is applicable only for Modes j 4 and 5,in which the reactor is shutdown and the average reactor coolant temperature is s 212 F. During '

Modes 4 and 5, the applicable mode of RIIR is the Shutdown Cooling (SDC) mode. There are no design basis accidents for which RIIR is credited in this mode This procedure change provides additional i ficxibility in re-establishing SDC. Since the ability of the RiiR system to remove decay heat from the I reactor is not afTected, there is no increase in dose from any accident previously evaluated in the SAR, I and the consequences of an accident are not increased. The change in operating configuration of the RIIR system does not change the operating environment of the system components; design temperatures and 1 113-

)

l I

l pressures of the system are not afTected. This change does not present challenges to any automatic functions of the system. Therefore, the probability of a malfunction of equipment important to safety previously evaluated in the SAR is not increased. Since the change does not adversely affect the ability of the system to remove decay heat from the reactor fuel, there will be no increase in dose from any malfunction of equipment important to safety previously evaluated in the SAR. The procedure change does not afTect any accident precursors or initiators which could cause a Loss of SDC. It does not change the physical characteristics of plant equipment and, therefore, cannot create the possibility of an equipment malfunction of a different type than presiously evaluated. The Technical Specification requirement for the minimum number of ECCS systems available has not changed. Cooling requirements and temperature limitations have also not changed. The Technical Specification Bases state that in Modes 4 and 5, the RHR system crosstic shutoff valve is not required to be closed. The Bases also include specific statements which provide that the crosstie valve may be used to allow pumps in one loop to  !

discharge through the opposite recirculation loop to make a complete subsystem. Therefore, this

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procedure change does not reduce the margin of safety as defined in any Technical Specifications. j Procedure 2.2.71 A (Resision 121 1

TITLE: Senice Water System Component Checklist DESCRIPTION: This revision changed the Component Identification Codes (CICs) of four Condenser Backwash Valve Disc Sparger Supply Valves. These valves were mistakenly assigned Circulating Water system CICs when they are actually Senice Water system valves. This revision also added eight Senice Water Pump Motor Oil Drain Valves. These valves appear to be original installation and were added to the checklist i in order to ensure proper configuration control.

SAFETY EVALUATION: These changes do not affect the precursors ofplant events presiously described in the SAR. They provide an enhancement to the barriers (i.e., configuration control) currently in place to prevent or mitigate the consequences of those events. None of the changes affect the manner in which the components are opeted, maintained, or configured. No new types of plant events or equipment malfunctions are intmduced as a result of this revision. The only Technical Specification equipment involved in this change is the addition of the eight oil drain valves for the station Service Water Pump Motors. These valves are added to the procedure for purposes of configuration control which supports the current analysis in the SAR and provides an enhancement to any margin of safety as the components will now be controlled per approved station documents.

Procedure 2.4.8.4.9 TITLE: Control Building Temperatures Above or Below Temperature Limits DESCRIPTION: This safety evaluation was perfbrmed to support previous revisions to this procedure (Revisions 8 and 9) which added steps to be taken to ensure the Control Building Basement does not exceed the 131"F temperature limit. Specifically, the procedure was revised to provide guidance to the operators to open the Control Building 903'-6" elevation equipment hatch to the basement by removing the hatch panels and opening doors Hi,11105, and H106 to establish a natural ventilation air tiow path to and fmm the Turbine Building in the event that the Control Building Basement temperature cannot be maintained I below 120'F. The procedure requires the operator to maintain the temperature below 130'F by shutting down all equipment except one Senice Water Booster Pump (SWBP) if necessary. Without this

operator action, the Control Building Basement bulk air temperature could exceed 131*F under a l postulated Design Basis Accident (DBA) Loss of Coolant Accident (LOCA). l SAFETY EVALUATION

The operator actions are in response to the identified need to establish natural ventilation cooling to the Contml Building Basement within eight hours following a postulated DBA LOCA to maintain the SWBPs operable. The operator actions are not a precursor to any postulated event. Only one SWBP is required l for all CNS design basis events. The procedure change ensures that temperatures in the Control Building Basement will remain below the limiting room temperature for SWBP operability. As a result, the 1

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SWBPs can perform their safety function during the course of the DBA LOCA event. Opening the Turbine Building door does not increase the radiation dose to the Control Room and has no efTect on offsite doses. The operator actions ensure that the Control Building Basement temperature will not l exceed limiting component operability temperature, thereby preventing the malfunction of the SWBPs.

No new equipment malfunctions are introduced as a result of the planned actions. Additional dose received by operators to take these actions is determined to be very low since the actions are not expected to be required until more than two hours after the initiation of the event. Station operators would not be ,

allowed to perform the actions during an event invohing a radiological release prior to establishment of 1 the dose rate by health physics event response personnel. No new plant events are created by the required actions to remove the hatch cover plates and open doors III,11105, and 11106. Ilowever, failure to perform these actions could result in a loss of the SWBPs. Such a failure is not credible based on the simplicity of the required actions. All actions to be taken are manual and involve evolutions previously performed at CNS by station personnel. More than single pump operation would be required only if event circumstances degrade beyond design basis at which time the operators would enter the Emergency Operating Procedures. Required post-event management of optional response equipment is consistent I with expectations for operators following the event. Since only one Residual IIcat Removal SWBP is I required to respond to the DBA, there is no reduction in the margin of safety as defined in the basis for I the Technical Specifications.

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l Procedure 2.5 3.7.A (Revision 1) l l Procedure 2.5.3 8. A (Revision 1)

TITLE: REC Filter Demineralizer Skid Component Checklist (2.5.3.7.A)

TEC Filter Demineralizer Skid Component Checklist (2.5.3.8.A)

DESCRIPTION: These procedures were revised to place three Reactor Equipment Cooling (REC) valves and three Turbine Equipment Cooling (TEC) valves to the CLOSED position to prevent the uncontrolled loss of i

water inventory from the REC and TEC systems through the filter demineralizer skid sampling valves.

l SAFETY l

EVALUATION: The placement of the REC filter demineralizer skid sample valves to the CLOSED position does not invohr adding new equipment or physically modifying existing equipment. Uncontrolled inventory loss which could prevent the REC system from performing its safety function is being eliminated. The TEC system uncontrolled inventory loss is also being eliminated, although the TEC system does not perfonn a safety function. The radiological consequences of events previously evaluated in the USAR are not

) increased. The valve position configuration is controlled by procedure. Any personnel failure to close the valves after sampling will be detected by Operations personnel through inventory loss monitoring.

There is no requirement to change the valve position post-accident. This change does not decrease the reliability of the REC /TEC systems, thus the probability of occurrence of a malfunction of equipment important to safety is not increased. The subject valves are nonessential and no new or additional equipment is being added. The potential impact associated with uncontrolled leakage has been eliminated. Therefore, radiological consequences are not increased. Placement of these valves to the i

I CLOSED position does not afTect the ability to maintain the REC and TEC system water chemistg. No di1Terent types of plant events or equipment malfunctions are created. Elimination of the uncontrolled inventory loss by closure of the valves restores system integrity to that required by the system design basis. Thus there is no adverse efTect on the health and safety of the public and the margin of safety is preserved.

Procedure 313 (Revision 19)

TITLE: Equipment Safety Classification DESCRIPTION: This change allows the use of nonessential passive equipment as a secondary containment boundary.

Active components will continue to be classified as essential. Justification for this change is based on the original CNS licensing basis as stated in FSAR Amendment 14, Response to Atomic Energy Commission Question 10.5b.

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SAFETY EVALUATION: The failure of secondary containment or any ofits components will not result in the creation of an initiator for a previously evaluated plant event. The consequences of a plant event are not increased because a failure of a non seismic passive component is not a credible event concurrent with a Design Basis Accident (DBA). This is based on the component not seeing significant pressure, nor is a Design Basis Earthquake considered to occur coincident with a DBA. Assuming the failure of passive components post accident is outside the original CNS design and licensing basis. The original Safety Evaluation Report (SER) accepted the design of secondary containment conditional on periodic testing to prove operability. Secondary containment operability is verified periodically by testing in accordance with an approved CNS testing program. It was recognized and accepted in the SER that the Main Steam lines beyond the outer containment isolation valve and consequently penetrating secondary containment were constructed to quality category D per Safety Guide 26. It is reasonable to assume that taking credit for nonessential passive components as a pressure boundary for secondary containment was recognized and accepted in the SER. Consequently, the use of nonessential passive materials for the purpose of a secondary containment boundary will not increase the probability of occurrence or consequences of a malfunction of equipment important to safety previously evaluated in the SAR. The failure of any J

nonessential component installed via this procedural change is bounded by the failure of the Main Steam l

Line, which was originally classified as nonessential. The Technical Specification I asis is the accident 1 analysis which already incorporates the provision for nonessential componento in the secondary {

containment penetration. Consequently, this activity does not reduce the margin of safety as defined m j the basis for any Technical Specification. '

Procedure 3. I 3 (Revision 2Q)

TITLE: Equipment Safety Classification DESCRIPTION: nis procedure revision removed the changes previously made by Revision 19 which allowed the use of nonessential passive equipment as a secondary containment boundary. This leads to more conservative and cautious evaluations.

SAFETY EVALUATION: This change is not associated with initiators of plant events discussed in the SAR. It deals only with equipment classification and involves a conservative method of classifying equipment. This change does i not alter the manner of operation or control of plant equipment, nor does it directly afTect the design I function of any component. This change does not cause equipment importt.nt to safety to be operated l differently than previously evaluated. Since this procedure change classifies equipment with a more I conservative classification process, equipment important to safety is identified more readily. Use of the revised procedure does not reduce any margins of safety as defined in the Techmcal Specifications.

Procedure 3.25 (Revision 6)

TITLE: Replacement Component Evaluation DESCRIPTION: This procedure revision included various changes to facihtate Replacement Component Evaluation (RCE) processing and work flow, and to climinate actions not required for implementation of a RCE.

SAFETY EVALUATION: This activity ensures the equivalency of a replacement component /part by identifying and evaluating the critical specifications to ensure that the component /part performs the same function to respond to the occurrence of a plant event as previously evaluated in the SAR. The evaluation of the replacement component /part ensures that it will perform its design and functional requirements and, therefore, not increase the probability of equipment malfunction. Since the component /part is evaluated to be equivalent by a review of the critical specifications, the consequences of a malfunction carmot be increased and no new types of failure modes are created. There is no impact on the margin of safety as defined in the basis for any Technical Specifications.

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Procedure 415 (Revision 20)

TITLE: Elevated Release Point and Building Radiation Monitoring Systems DESCRIPTION: This procedure was revised to clarify the role of the Auxiliaiy Sample Pump at the Reactor and Turbine Buildings. In response to a Condition Report, the procedure was revised to not inop the monitors for loss of the auxiliary pump. It was determined that if the Auxiliary Sample Pump would fail, the monitor would still monitor noble gasses correctly, but corrections to the particulate / iodine release data would need to be made to account for subisokinetic sampling. Procedure 8.8.8 provides the necessary corrections should this condition occur. In addition, the check source criteria was changed to " rises" as the Technical Specifications definition only requires a qualitative check and specifies no minimum response.

SAFETY EVALUATION: The loss of the Auxiliary Sample Pump only afTects sampling of particulate / iodine material. Procedure steps are in place to allow proper corrections to particulate / iodine concentrations. The new source check criteria are consistent with Technical Specifications and provide adequate indication of monitor response.

These monitors are passive monitoring systems and do not have the capability of causing an accident evaluated in the SAR nor do they impact any accident analysis assumptions. They do not have the capability to increase the consequences of a plant event or equipment malfunction evaluated in the SAR and do not create the possibility of a different type of plant event or equipment malfunction than previously evaluated. The procedure changes allow the monitors to perform their safety design basis function as described in the 'USAR by providing the methodology to correct analysis results ensuring proper determination of dose to the public via radiological efiluents. The changes do not operate the efiluent monitoring system outside ofits system design. Additionally, the procedure changes do not afTect other systems, subsystems, or components important to safety. Should a malfunction of equipment occur, provisions are in place to address the inoperability of this equipment as required by Technical Specifications. The changes do not affect the ability to obtain and calculate gaseous etliuent release data during normal operations or post-accident. The changes serve to maintain the equipment in an operable condition rather than unnecessarily declaring the equipment inoperable. Margins of safety remain unaffected by this procedure revision.

Procedure 6 FP 102 (Revision 51 Procedure 6 FP 605 (Revision 21 Procedure 6 FP 629 (Revision 21 Procedure 6. FP 635 (Resision 21 Procedure 15 FP 602 (Revision 21 Procedure 15 FP 603 (Revision 21 Procedure 15 FP.604 (Revision 21 Procedure 15 FP 606 (Revision 2) Procedure 15 FP 607 (Revision 21 Procedure 15 FP 608 (Revision 2L Procedure 15 FP.609 (Revision 21 Procedure 15 FP.610 (Revision 21 l Procedure 15 FP.611 (Revision 21 Procedure 15 FP 612 (Revision 21 Procedure 15 FP 613 (Revision 21 Procedure 15 FP 614 (Revision 21 Procedure 15 FP 615 (Revision 21 Procedure 15 FP 616 (Revision 2t Procedure 15 FP.617 (Revision 21 Procedure 15 FP.618 (Revision 21 Procedure 15 FP.619 (Revision 21 Procedure 15 FP.620 (Revision 21 Procedure 15 FP 621 (Revision 21 Procedure 15 FP 622 (Revision 2L Procedure 15 FP 624 (Revision 21 Procedure 15 FP.625 (Resision 21 Procedure 15 FP.626 (Revision 21 Procedure 15 FP.627 (Revision 21 Procedure 15 FP 628 (Revision 21 Procedure 15.FP 630 (Revision 21 l I

Procedure 15 FP.631 (Revision 21 Procedure 15 FP 632 (Revision 21 Procedare 15.FP.633 (Revision 21 Procedure 15 FP.634 (Revision 21 Procedure 15 FP 637 (Revision 21 Procedure 15 FP.638 (Revision 21 Procedure 15 FP 639 (Revision 21 Procedure 15 FP 640 (Revision 21 Procedure 15 FP.641 (Revision 21 Procedure 15 FP.642 (Revision 21 Procedure 15 FP 643 (Revision 21 Procedure 15 FP.644 (Revision 21 Procedure 15 FP.645 (Revision 2)

TITLE: Annual Testing of Fire Pumps (6 FP.102)

System Number 5 Flow Verification Test (6 FP.605)

System Number 29 Flow Verification Test (6.FP.629)

System Number 35 Flow Verification Test (6 FP.635)

System Number 2 Flow Verification Test (15.FP.602)

System Number 3 Flow Verification Test (15.IT.603)

System Number 4 Flow Venfication Test (15.FP.604)

System Number 6 Flow Verification Test (15.FP.606)

System Number 7 Flow Verification Test (15.FP.607) 117-

l System Number 8 Flow Verification Test (15.FP.608)

System Number 9 Flow Verification Test (15.FP.609)

System Number 10 Flow Verification Test (15.FP.610)

System Number 11 Flow Verification Test (15.FP.611)

System Number 12 Flow Verification Test (15.FP.612)

System Number 13 Flow Verification Test (15.FP.613)

System Number 14 Flow Verification Test (15.FP.614)

System Number 15 Flow Verification Test (15.FP.615)

System Number 16 Flow Verification Test (15.FP.616)

System Number 17 Flow Verification Test (15.FP.617)

System Number 18 Flow Verification Test (15.FP.618)

System Number 19 Flow Verification Test (15.FP.619)

System Number 20 Flow Verification Test (15.FP.620)

System Number 21 Flow Verification Test (15.FP.621)

System Number 22 and 23 Flow Verification Test (15.FP 622)

System Number 24 Flow Verification Test (15.FP.624)

System Number 25 Flow Verification Test (15.FP.625)

System Number 26 Flow Verification Test (15.FP.626)

System Number 27 Flow Verification Test (15.FP.627)

System Number 28 Flow Verification Test (15.FP.628) l System Number 30 Flow Verification Test (15.FP.630)

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System Number 31 Flow Verification Test (15.FP.631)

System Number 32 Flow Verification Test (15.FP.632) l System Number 33 Flow Verification Test (15.FP.633)

System Number 34 Flow Verification Test (15.FP.634)

System Number 37 Flow Verification Test (15.FP.637)

System Number 38 Flow Verification Test (15.FP.638)

System Number 39 Flow Verification Test (15.FP.639)

System Number 40 Flow Verification Test (15.FP.640)

System Number 41 Flow Verification Test (15.FP.641)

System Number 42 Flow Verification Test (15.FP.642)

System Number 43 Flow Verification Test (15.FP.643)

System Number 1 A Flow Verification Test (15.FP.644)

System Number 1B Flow Verification Test (15.FP.645)

DESCRIPTION: These procedures were revised to incorporate steps for utilizing "E" fire pump for flow testing and the  ;

jockey pump for system restoration following testing. The testing changes allow use of"E" fire pump  !

for flow verification without impacting pump or system operability. Use of the jockey pump to repressurize the system minimizes inadvertent pump starts.

SAFETY EVALUATION: Changes in the method of testmg and restoring the Fire Protection (FP) system do not present an accident initiator or precursor. Testing and restoration of the FP system ensures maximum availability to respond to a fire. Suflicient redundancy exists to satisfy credited event mitigation. The testing changes ensure this level of redundancy is maintained. The testing does not introduce additional failure modes of the FP equipment. Other equipment failures due to FP system maloperation are not increased by changes in testing since they result from extemal events, not the testing itself. Safety equipment failure consequences l due to FP system maloperation are unchanged and bounded by the existing fire suppression efTects  !

analysis. Changes to system testing sequences do not change the conclusions of the fire suppression effects analysis. No new failure modes or efTects are introduced by the changes in testing. The changes in testing ensure that Technical Specification margins for system operation and redundancy are maintained.

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l Procedure 6 FP 2Q3 (Revision 1)

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TITLE: Fire Damper Assembly Examination (Fire Protection System 18 Month Examination)

DESCRIPTION: This procedure was revised to eliminate preventive maintenance steps with regard to cleaning, 3 lubricating, and functional testing. The procedure is now only a visual examination per Technical l Specification requirements and does not perform additional work. This climinates preconditioning concerns associated with performing maintenance activities under a surveillance.

SAFETY EVALUATION: This change (kies not increase the probability of occurrence of a fire or other accident because no accident precursors or initiators are being introduced. The capability of the plant to mitigate an accident or event is unaffected by this change. The ability of a fire damper to close under fire conditions is ensured by this procedure change; however, its effectiveness as a fire barrier is unchanged. The procedure ensures that the assumption in the Fire llazards Analysis that a fire damper will close is valid by visually verifying damper condition. Probability ofequipment malfunction is unchanged or reduced by performance of this j revised procedure. Past performance has shown no correlation between component reliability and '

performance of the preventive maintenance activities. The consequences of a failed fire damper are unchanged by the performance of this procedure. This activity only impacts fire protection and ventilation system equiprnent. No new failure modes are introduced. The defense in depth fire protection program which fixms the basis fur the fire protection Technical Specifications is unaffected by these changes. The ,

operability of the fire dampers is still being verified by visual examination as described in Technical j Specifications. l Procedure 6 FP.304 (Revision 2)

TITLE: Fire Detection System Circuitry Operability DESCRIPTION: This surveillance procedure previously instructed operators to verify that an alarm LED light turned oft after tuming a key locked switch on the senice water pump room IIalon control panel to abort. This action could not be performed since the LEDs were already off because no fire alarm condition existed.

Therefore, specific steps of this procedure were revised to change the wording from " Alarm LED turns off" to " Alarm LED is off." The LEDs are for alann, pre-discharge, and discharge indication. These alarms do not come in during the abort switch activation. The abort switch activation is a test of the supenisory circuit trouble alarm only.

SAFETY EVALUATION: This activity does not change the state or function of safety related systems, structures, or components; therefore, it will not alter any of the inputs or assumptions for previously evaltiated accidents. The consequences of previously evaluated accidents remain bounded by the results contained in the USAR.

This activity provides changes to a procedure that is used to perform testing on a Halon suppression system control panel which does not contain se.fety related systems or compenents. Because this activity consists of procedural changes to correct crroneous procedure steps, the function of the Ilalon system is not adversely affected. The design basis for the service water pump room Halon system is not altered.

Initiating sequences or starting setpoints of safety related equipment are not adversely afTected. Thus, this change does not increase the probability of a malfunction of equipment important to safety. This activity assures reliability of the senice water pump room Halon syster u. function when needed to mitigate the consequences of a fire. 'niis change does not contribute to any meidents or failure modes analyzed in the l USAR and does not create any new initiators or failure modes. This activity has no efTect on the l acceptance limits for any system important to safety nor does it affect the range allowed for operation l under analyzed operational transients or the range required to prevent approaching system limitations or design failure points. Therefbre, the margin of safety is not reduced.

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Procedure 6 FP.305 (Revision 2)

TITLE: Ifalon 1301 Service Water Pump Room Fire Suppression Surveillance Checks I

Dl!SCRIPTION: his procedure was revised to appropriately identify acceptance criteria. Steps for perfbrming a visual  !

survey of the smoke detectors and cleaning them as required are not acceptance criteria, therefore the )

acceptance criteria designation was removed. Inspecting all the nozzles to assure they are unobstructed is a Technical Specification requirement; therefore, this examination was appropriately designated as part of the acceptance criteria for the surveillance of the Ilalon system.

SAFETY f

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EVALUATION: ne iIalon fire protection system was not identified as a component in an analyzed event that initiates or I mitigates abnormal operational transients or accidents analyzed in the USAR. The IIalon fire protection system (kies mitigate the consequences of a fire in the Service Water pump room. The subject changes i will not alter the design basis of the Service Water pump room Ilalon fire suppression system. Because this activity will not change the state or function of safety related structures, systems, or components, it j

will not alter any of the inputs or assumptions for the probabilities of previously evaluated accidents. It  !

will not change the function, performance, or integnty of any boundaries with which safety related systems form or support the primary protective baniers on which the consequences of previously identified accidents are based. This change does not adversely affect initiating sequences or starting setpoints of the fim pumps. It does not induce failure of any equipment important to safety. This activity will ensure reliability of the fire protection system to function when needed to mitigate the consequences of a fire.

The consequences of the Fire Suppression EITects Analysis are bounding and unchanged by this activity.

This change does not afTect equipment important to safety such that it would operate differently than equipment important to safety has previously operated. No new event initiators or failure modes are introduced This activity has no etrect on the acceptance limits for any system important to safety nor does it affect the range allowed for operation under analyzed operational transients or the range required to prevent approaching system limitations or design failure points. Therefore, the margin of safety is not reduced.

Procedure 6 FP 601 (Revision 5)

Procedure 6 FP 604 (Revision 41 l

TITLE: Fire Protection System 31 Day Examination (6.FP.601)

Fire Door Annual Examination (6.FP.604)

DESCRIPTION: These procedures were revised to identity Doors R210,11202, R1It, and R406 as doors which are nonnally locked and closed. They are not used for normal access / egress ar d do not require a latch / closure test. Security type fire doors that are locked and alarmed in the closed position do not require automatic door closures because security personnel are stationed at the door the entire time the door is open.

SAFETY EVALUATION: This activity does not change the state or function of safety related structures, systems, or components; therefore, it does not alter any of the inputs or assumptions for previously evaluated accidents. This revision (kes not change the function or performance of any boundaries with which safety related systems form or support the primary protective barriers upon which the consequences of previously identified accidents are based. Fire ckiors are passive components in fire barriers that are classified as safety related structures. He function of the subject fire doors is not adversely afrected and the fire doors will still be able to perform the separation of safe shutdown equipment function against a fire involving one division of safe shutdown equipment and cables. This change does not alter the design basis for fire barriers nor

, ck>es it adversely affect initiating sequences or starting setpoints of safety relrited equipment. It does not l induce failure of any equipment important to safety or cause it to operate differently than it has previously operated. This activity ensures reliability of fire barriers to function when needed to mitigate the consequences of the spread of fire. No new failure modes or accident initiators are introduced as a result of this change. The margin of safety which is established by the design and performance of safety related systems is not reduced by this activity.

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f Procedure 6 HPCI 204 (Revision 01 TITLE: IIPCI-SOV-SSV64 IST Closure Test DESCRIPTION: This new procedure was developed to perform inservice Testing (IST) ofIIPCI-SOV-SSV64 in order to satisfy ASME Section XI Code requirements Testing requirements for IIPCI-SOV-SSV64 were revised via Revision 2 of the CNS IST Program.

SAFETY EVALUATION: Testing of IIPCI-SOV-SSV64 in accordance with the IST Program demonstrates the operational readiness of the valve. The High Pressure Coolant Injection (1IPCI) system being out of senice during testing is not a precursor to any accident and is addressed in the USAR. The USAR states that components required for IIPCI are designed to allow functional testing during normal power operations.

The USAR further acknowledges that certain testing requires valves to be electrically isolated which would prevent auto retum to operating mode. However, the USAR concludes that due to the duration of these tests, there is no effect on system reliability or degradation of plant safety. This actisity does not alter the function or operation of HPCI-SOV-SSV64. Consequences of malfunctions are dictated by component safety function. Since the HPCI system is out of service during the test, a malfunction of HPCI-SOV-SS64 cannot affect system operation. The normal lineup is restored upon completion of l

testing. This activity does not degrade nor decrease the reliability of a HPCI component. It does not

! create the possibility of a different type of plant event or equipment malfunction. Testing ofIIPCI l

components is required by Technical Specifications. As such, this activity does not reduce the margin ofsafety.

Procedure 6. HPCI 313 (Revision 10)

(USQE 1998101)

TITLE: HPCI (150 PSIG) Beginning of Cycle Test DESCRIPTION: This procedure was revised to allow the High Pressure Coolant Injection (Hh'I) Automatic Actuation Test required by Improved Technical Specification Surveillance Requirement 3.5.1.9 to be performed at 150 psig during startup. This will verify that the HPCI system will perform its safety function as designed. The automatic actuation testing was previously performed at normal reactor operating pressure in accordance with CNS Custom Technical Specifications. This change also deletes the gathering of Inservice Testing (IST) data under this procedure as the pump will not be operating under baseline conditions to allow the IST data to be gathered. To facilitate automatic actuation timing during the testing, a recorder will be temporarily connected to the HPCI system discharge flow and discharge pressure instnanentation, and the annunciator system to record the HPCI system automatic actuation and performance.

SAFETY l EVALUATION: The HPCI system is the initiator of one previously evaluated Abnormal Operating Transient, inadvertent HPCI pump start. The consequences of this event are bounded by the loss of feedwater heating event.

l 'Ihe changes to this procedure do not increase the frequency ofits performance, nor do they climinate the procedural steps taken to eliminate the possibility of an inadvertent injection of HPCI. The temporary changes to the llPCI instrumentation will not increase the likelihood of initiating an inadvertent HPCI l pump start, as it is not part of the HPCI initiation logic. The changes to the procedure and the addition of the temporary recorder cannot create a new radioactive release path. The temporary changes to the instrumentation will not affect the performance of IIPCI or any other system. Connection of the temporary n: corder will not affect the ability of the flow controller to perform its function in controlling the HPCI turbine. Intemal fusing for the recorder provides adequate protection to the control circuit in the unlikely event ofrecorder failure. The temporary changes will not affect the divisional separation of any plant equipment. Therefore, there is no increase in the probability of occurrence or consequences of a malfunction of equipment important to safety. The changes to the procedure will not alTect the operation off IPCI. The HPCI system has been previously evaluated for inadvertent operation and failure to operate in an accident. The changes to the procedure will not affect any instrumentation or equipment for the Primary Containment Isolation function of the IIPCI steam supply valves. Therefore, the

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l possibility of a new type of accident or equipment malfunction is not created. The 1IPCI system is not operable during the testing. The temporary recorder will be removed when testing is completed. The requuunents for the Emergency Core Cooling System equipment will be controlled in accordance with l

existing CNS procedures and the Technical Specifications. Therefore, the margin of safety as defined in the basis for any Technical Specification has not been reduced.

1 Procedure 6.0G.201 (Revision 01 TITLE: OG-CV-8CV and OG-CV-12CV IST Open Exercise Test DES 991PTION: This new procedure was developed to perfbrm Inservice Testing (IST) of OG-CV-8CV and 12CV in order to satisfy CNS IST Program requirements. Design Change 97-068A added these valves to the plant and since they have an active safety function to open to prevent binding in the oft Gas (OG) hold up drain line to the Z sump, testing shall be performed accordingly.

SAFETY l EVALUATION: Testing of the OG check valves in accordance with the IST program demonstrates the operational l readiness of the valves. Testing of these valves is not a precursor to any design basis accident or I transient. Testing per the new procedure is performed on one check valve at a time. While one check l

valve is being tested, the other check valve allows the system to perform its design function. This activity i does not alter the function or operation of these valves or the OG system. The testing does not afTect any component's safety function. It does not degrade nor decrease the reliability of an OG system component.

Additionally, the procedure contains steps to install and independently verify check valve orientation and system lineup. Since the system is still capable of performing its safety function during the test, this

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activity does not reduce the margin of safety.  !

hpcedure 6.PC.50j (Revision 6)

TITLEi Primary Containment Local Leak Rate Tests DESCRIPTION: This revision made adjustments to leakage limits due to excellent performance history and in response to various Problem Identification Reports. It also enhanced the procedure by accounting for all tests performed on various expansion bellows. It added a new piece of test equipment and clarified instrument accuracy requirements.

SAFETY EVALUATION: This change does not cause the testing to be performed during a different mode of operation than was previously authorized. In addition, containment leak rate testing is not a precursor to any design basis accident or transient. The procedure change does not change the test pressure or boundary, as such, there will be no adverse efTect upon the containment boundary or to any components which may be contained within the test boundary. The ability to quantify containment leakage and to assess primary containment integrity are not adversely affected. Therefore, offsite dose determined for any design basis accident or transient or malfunction ofequipment important to safety is not affected. This change does not alTect the primary containment leakage acceptance criteria specified in the Technical Specifications or the ability to determine if that criteria is satisfied.

Procedure 6 PC 502 (Revision 2T TITLE: Primary Containment Instrumentation Local Leak Rate Tests DESCRIPTION: This procedure revision added a note to allow Instrument & Control personnel to perform packing / fitting adjustments on manual valves imulved with the test if a leak is detected. A note was also added to clarify pressure decay testing requirements. In addition, a note was added to indicate the proper annunciators and alanns that will trip in the Control Room while testing is being performed.

SAFETY EVALUATION: These changes do not cause the procedure to be performed during a different mode of operation than presiously authorized. In addition, containment leak rate testing is not a precursor to any design basis -

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l accident or transient. The test pressure boundary is not changed, as such, there will be no deleterious effect upon the cora nment boundary, or to any hardware which may be contained within the test boundary OfTsite dose determined for any design basis accident, transient, or malfunction of equipment important to safety is not afTected by this change. This revision does not affect the primary containment l leakage acceptance criteria specified in the Technical Specifications or the ability to determine if that 3 criteria is satisfied.

Procedure 6 PC 503 (Revision 4) i Encineerine Evaluation (EET 97-326 Safety and Relief Valve Setnoint Chance Reauest 97-007 {

4 TITLE: Drywell to Suppression Chamber Leakage Test (6.PC.503)

Non-Retum Valve Leakage Test Procedure Enhancement (EE 97-326)

Setpoint Change for PC-RV-14RV (97-007) 1 DESCRIPTION: Procedure 6.PC.503 was revised to use the Sutorbilt Pump Around System in conjunction with the Drywell Nitrogen Makeup System to develop the test parameters specified in the procedure. The previous method of establishing the difTerential pressure was the Drywell Nitrogen Makeup System.

Flow available from the Drywell Nitrogen Makeup alone is less than the flow needed to compensate for the total allowable leakage rate defined in the Technical Specifications. In order to provide more flow for pressurizing the drywell, the Sutorbilt Pump Around System will be used in conjunction with the Drywell Nitrogen Makeup System. EE 97 326 documented and evaluated the design basis of the existing Sutorbilt Pump Around system to ensure it is suitable fbr use in revised Surveillance Procedure 6.PC.503.

The system had been previously abandoned in place. EE 97-236 qualified one blower, PC-BLWR-A, in its existing configuration for use in Surveillance Procedure 6.PC.503. The blower was functionally tested under Special Procedure 97-016. The blower is protected from damage by relief valve PC-RV-  ;

14RV. The setpoint for PC-RV-14RV was reestablished by Safety and Relief Valve Setpoint Change Request 97-007. ,

SAFETY j EVALUATION: This activity will not change the state or function of safety related structures, systems, or components as l described in the USAR and will not alter any of the inputs or assumptions for previously evaluated '

accidents. Primary containment is protected by Group 2 Isolation Valves; use of the Sutorbilt Pump Around System does not affect these isolation valves. The drywell and suppression chamber pressure is manually monitored and adjusted. The previous procedure utilized the same Operator actions; thus, there is no change in the probabihty ofequipment malfunction because of Operator action. Some of the piping used is also used by Standby Nitrogen Injection (SBNI); however, there is a redundant SBNI train that will not be involved. Provisicms were added to the procedure to ensure the atTected Motor Control Center (kes not exceed load limitations. This actually lessens the probability of failure. This activity does not i affect or alter the accident mitigation capabilities or change the failure modes of any equipment important to safety. No new event initiators or failure modes are introduced. No changes have been made to parameters that are bounded by the containment Technical Specifications or Administrative Limits.

Therefore, there is no reduction in the margin of safety.

Procedure 6.PC 513 (Revision 5)

TITLE: Main Steam Local Leak Rate Tests l

DESCRIPTION: This procedure revision provides for a test method which will allow the outboard Main Steam Isolation  !

Valves (MSIVs) to be individually leak rate tested without flooding the Main Steam Lines (MSLs) and  ;

could possibly save time over other diagnostic test methods. With the vessel head installed and tensioned, l

, and the head vents closed, the MSLs can be pressurized to 28 psig. This pressure is well below the i

! Shutdown Cooling (SDC) low pressure isolation setpoint. Therefore, SDC will not isolate during this )

l test. This pressure is also less than the MSIV test pressure. It will force the inboard MSIVs closed and ]

will climinate any significant dp across the intioard valve. Thus all measured leakage will be attributed l l

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1 to the outboard valve. This test method is similar to thxxiing the steam lines, but will be quicker in re-establishing normal conditions if additional maintenance on the valves is required.

SAFETY EVALUATION: The test is perfomied when the plant is shutdown, the vessel head is installed and tensioned, and the head vents are closed (Mode 4). The water level is maintained below the MSLs and SDC is operating.

Suflicient margin exists between the test pressure and the SDC isolation setpoint. Essential components within the test boundary are designed for full reactor pressure. In this low energy condition, the slight increase in vessel pressure during the test will not increase the probability of any accident, transient, or equipment malfunction, and will not introduce any new types of accidents. The required core cooling systems are operable and the MSIVs are closed. Test personnel are able to isolate the manual fill, test, and vent connections if necessary. Therefore, the consequences of an accident or malfunction of equipment important to safety are not increased during this test. The SDC system will automatically ,

isolate on high pressure. With the water level below the MS nonles, the increase in pressure during the  !

test will raise the pressure at RR-PS-128A/B to approximately 40 psig. This will still provide a margin of ~32 psi to the trip point. Thus an unplanned isolation of SDC is unlikely and the possibility of a l

different type of malfunction is not increased. The purpose of the Residual IIcat Removal System in i Mode 4 is to remove decay heat and sensible heat from the reactor coolant. The SDC system will be {

operable during the test and suflicient margin exists between the test pressure and the SDC isolation i

setpoint. Therefore, this test does not reduce the margin of safety as defined in the Technical l Specifications.

1 Procedure 6 RCIC.309 (Revistou 10) l (USQE 1998-0100) ]

i TITLE: RCIC (150 PSIG) Beginning of Cycle Test l

l DESCRIPTION: This procedure was revised to allow the Reactor Core Isolation Cooling (RCIC) Automatic Actuation l

Test required by Improved Technical Specification Surveillance Requirement 3.5.3.5 to be performed  !

at 150 psig during startup. This will verify that the RCIC system will perform its safety function as designed. The automatic actuation testing was previously performed at normal reactor operating pressure in accordance with CNS Custom Technical Specifications. This change also deletes the gathering of insenice Testing (IST) data under this procedure as the RCIC pump will not be operating under baseline conditions to allow the IST data to be gathered. To facilitate automatic actuation timing during the testing, a recorder will be connected to the RCIC system discharge flow and discharge pressure instrumentation, and to the annunciator system to record the RCIC system automatic actuation and performance.

SAFETY EVALUATION: The RCIC system is the initiator of one previously evaluated transient, inadvertent RCIC injection. The consequences of this event are bounded by the loss of feedwater heating event and inadvertent IIPCI injection. The changes to this procedure do not increase the frequency ofits performance, nor do they eliminate the procedural steps taken to eliminate the possibility of an inadvertent injection of RCIC. The temporary changes to the RCIC instrumentation will not increase the likelihood ofinadvertent RCIC injection. The changes to the procedure and the addition of the temporary recorder cannot create a new radioactive release path. The temporary changes to the instrumentation will not afTect the performance ofRCIC or any other system. Connection of the temporary recorder will not affect the ability of the flow controller to perform its function in controlling the RCIC turbine. Internal fusing for the recorder provides adequate protection to the control circuit in the unlikely event of recorder failure. The temporary changes will not alTect the divisional separation of any plant equipment. Therefore, there is no increase in the probability ofoccurrence or consequences of a malfunction of equipment important to safety. The changes to the procedure will not afTect the operation of RCIC. The RCIC system has been previously evaluated for inadvertent operation and failure to operate in an accident. The changes to the procedure will not afTect any instrumentation or equipment for the Primary Containment Isolation function of the l RCIC steam supply valves. Therefore, the possibihty of a new type of accident or equipment malfunction is not created. The RCIC system is not operable during the testing. The temporary recorder will be removed when testing is completed. The requirements for the RCIC and Emergency Core Cooling

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System equipment will be controlled in accordance with existing CNS procedures and the Technical Specifications. Therefore, the margin ofsafety as defmed in the basis for any Technical Specification has not been reduced.

Procedure 6.RPS 302 (Revision 6)

TfrLE: Main Turbine Stop Valve Closure and Steam Valve Functional Test DESCRIPTION: This procedure was revised to allow testing of the turbine stop valve closure scram function below the 25% scram bypass setpoint. Steps were added to remove fuses to deenergize relays to defeat the scram bypass and allow testing of the turbine su>p valve closure scram function. Likewise, steps were added to reinstall the fuses upon completion of testing and to observe actions which verify the integrity of the fuses after installation. With the exception of steps added to defeat the scram bypass, the test methodology remains unchanged and only one reactor scram channel is tested at a time. This revision is necessary to demonstrate the main turbine stop valve Reactor Protection System (RPS) trip is operable prior to entering the mode or plant condition that requires the trip to be operable (i.e , >25% rated turbine first stage pressure).

SAFETY EVALUATION: In order for this activity to result in an event previously evaluated in the SAR, a single active equipment failure or single Operator error must occur. The steps added by this procedure do not increase the probability of an Operator error or single active component failure. No new equipment is added by this activity. Dose consequences are bounded by operations below the Technical Specification main turbine first stage pressure permissive setpoint where closure of the turbine stop valves from such low initial power levels does not constitute a threat to the integrity of any fuel barrier and a subsequent release of radioactive material endangering the health and safety of the public. This activity does not reduce the ability of any mitigation system to mitigate the consequences of an event. Decnergizing the appropriate relays will not affect RPS components that actually provide the turbine stop valve closure and turbine control valve fast closure scram functions. Verifying the relays pick up on reinstallation of the fuses ensures RPS function is restored and capable of performing its intended design function above 25%

turbine first stage pressure. Decnergizing the relays by pulling the respective fuses will not introduce any new failure mechanisms into the RPS components as they are normally deenergized when the turbine RPS functions are required to be operable. This activity does not increase the Technical Specification main turbine first stage pressure permissive setpoint of 179 psig, therefore, it does not reduce the margin of safety as defined in the basis for any Technical Specification.

Procedure 6 SC.501 (Revision 5)

USAR Channe Reouest (UCR 98-108)

(USQE 1998-0093)

TITLE: Secondary Containment Leak Test DESCRIPTION: The revision to Procedure 6.C.501 deleted steps associated with stack effect temperature correction. l Engineering Evaluation 1998-0124 described the application of stack effect correction as an overly I restrictive requirement beyond original design requirements for Secondary Containment or Standby Gas Treatment (SGT). It provided justification that CNS is not required to apply a stack effect correction to the Technical Specification required acceptance criteria of s .25" w.gls 1780 cfm.

The USAR was revised to clarify that the SGT system sizing was for an " average" pressure of

-0.25" w.g., with the average being across the whole building. It was further enhanced to reflect that the average of the four differential pressure transmitters located on the refueling floor are representative of ,

the average Reactor Building differential pressure as a whole. In addition, a USAR inconsistency was  !

corrected by clarifying that achieving the 0.25" w.g. vacuum is under neutral wind conditions. j SAFETY J EVALUATION: This change clarifies the USAR with regard to the pressure requirements within Secondary Containment f

such that it is understood that the pressure need not be s -0.25" w.g. at all locations in the building to i

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demonstrate leak tightness. This activity will have na impact on the probability of any accident described I in the SAR. Removal of the stack elTect from the Surveillance Procedure will have no efTect on the j probability of an accident. Evaluations of the consequences of accidents and equipment malfunctions are l based upon leak tightness of the Secondary Containment, which is demonstrated by the testing included in procedure 6.SC.501. The licensing basis for Secondary Containment is that adequate leak tightness is demonstrated by the capability to maintain s -0.25" w g. without a requirement to maintain this pressure at all locations in Seconday Containment. Therefore, this activity will have no impact on the consequences of an accident or malfunction of equipment. Similarly, removing the stack effect from Procedure 6.SC.501 will not impact the consequences of an accident or equipment malfunction since pressure within Secondary Containment should always be subatmospheric. Clarification of the USAR l with regard to the pressure requirements within Secondary Containment will not alTect the components which provide the subatmospheric containment and, therefore, the probability of a malfunction will not be increased. Removing the stack effect from Procedure 6.SC.501 will also not impact the functionality of the equipment, thus the probability of equipment n a! function is not increased. Clarifying the design basis (k)es not cn: ate the possibility of a different type of accident or malfunction of equipment important to safety than previously evaluated in the SAR. Removing the stack effect from the Surveillance Procedure has no impact on installed equipment nor does it change the method of operation. Therefore, removal of the stack effect will not create the possibility of an accident or equipment malfunction of a different type. The margin of safety as defined in the basis for any Technical Specification will not be reduced. The evaluations of margin of safety are based upon leak tightness of the Secondary Containment, which is demonstrated by the testing in Procedure 6.SC.501. Removing the stack eflbct from the Surveillance Procedure will not reduce the ability of mitigation systems to filter the release of radioactive material prior to release since pressure within Secondary Containment should always be subatmospheric as identified in licensing basis documents.

Procedure 6 SC 502 (Revision 3)

TITLE: Secondary Containment Penetration Examinatien DESCRIPTION: This procedure was revised to allow leak testing of Door R109 in the south vestibule of the Reactor Building. This door could not be tested previously during operation since the outer door in this airlock is normally scaled closed. Testing of this door will now be performed by temporarily opening a spare fire penetration seal, which allows the south vestibule to be open to extemal ambient pressure and establishes the required difTerential pressure across the door. This change is required to ensure the secondary containment boundary will still be intact if access to Standby Nitrogen injection controls in the vestibule is needed for post-accident mitigation. The sum of airlock path leakages, which is used for the acceptance criteria of the test, was conscivatively revised to use the higher of the estimated inleakage of the two doors for all of the tested penetrations.

SAFETY EVALUATION: No accident precursors are affected by this procedure change. A continuous fire watch will be established while the fire penetration sealis opened during the test and the penetration will be rescaled following the test. Secondary contamment integrity will be maintained during the test by procedurally ensuring that the inner south vestibule door is maintained closed while the spare penetration is opened. Secondary containment reliability will be improved since the procedure will allow leak testing of a door which was not previously tested during operation. The conservative change in the acceptance criteria calculation will ensure that secondary containment leakage does not exceed the Standby Gas Treatment (SGT) capability as long as one door in each access opening is closed, as required by Technical Specifications.

Existing dose calculations are unaffected by this change since the Reactor Building leakage characteristics used in these analyses are unchanged and SGT performance is not alTected. No new failure modes are introduced by this change and the probability of existing failure modes is not increased. The probability ofopening the inner door to the south vestibule while the spare penetration is open will be prevented by procedural guidance and both sides of the door will be under direct control of personnel involved in the test. Using a more conservative method to evaluate the amount of leakage from the secondary containment penetrations increases the margin of safety.

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Procedure 6.1 APRM.304 (Revision 21 Procedure 6.2 APRM.304 (Revision 21 l TITLE: APRM System Channel Functional Test (Mode Switch Not in Run) (Div 1)(6.1 APRM.304)

APRM System Channel Functional Test (Mode Switch Not in Run) (Div 2) (6.2APRM.304) l DESCRIPTION: In order to meet improved Technical Specification (ITS) requirements, these procedures were revised ,

to allow performance of the Average Power Range Monitor (APRM) downscale/ Intermediate Range l l

Monitor (IRM) upscale channel functional test prior to entering the mode of applicability (Mode 1).

l Steps were added to the procedure to install jumpers that simulate the reactor mode switch in the Run i position. This enables the APRM downscale trip and, thereby, allows testing of the APRM/IRM l

companion trip.

SAFETY EVALUATION: This change has no elTect on the initiation of any plant event. It enables a trip function that would ot envise be bypassed in the Run position of the reactor mode switch. Enabling the APRM downscale function to allow its testing prior to entering the mode of applicability does not increase - grobability ,

ofoccurrence of a plant event previously evaluated in the SAR. The ITS Bases states, "With the reactor I mode switch in run, an APRM downscale coincident with an associated Intermediate Range Monitor Neutron Flux-Iligh or Inop signal generates a trip signal. This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis." Therefore, any changes affecting only this function have no efTect on the mitigation of any accident. The jumpers that are installed enable the APRM downscale/lRM upscale trip function; however, they afTect no other function. Thejumpers are removed prior to entering i the made of applicability for the function and have no effect on the function when it is actually required.

This change does not increase the probability of occurrence or consequences of a malfunction of 4 equipment important to safety. Thejumpers only affect the APRM downscale/lRM upscale coincident

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scram function. Therefore, this change cannot create the possibility of a different type of plant esent or equipment malfunction than previously evaluated. Installation of the jumpers for testing is necessary to comply with Technical Specifications. Enabling an additional trip function in no way reduces the margin of safety. Any effect on the margin of safety is an increase.

Procedure 61DG 302 (Revision 91 Pmpsdure 6 2DG.302 (Revision 6)

(USQE 1998-0080)

TITLE: Undervoltage Logic Functional, Load Shedding, and Sequential Loading Test (Div 1)(6.1DG.302)

Undervoltage Logic Functional, Load Shedding. and Sequential Loading Test (Div 2)(6.2DG.302)

DESCRIPTION: These surveillance procedures were revised to remove the paddle from 32/lFE (32/lGE) overpower relaysjust prior to synchronizing the F Bus (G Bus) to the A (B) Bus and the grid. After the two buses are synchronized and the Diesel Generator (DG) load is stabilized, the paddle will be returned to the relay. The total time the paddle is removed will be minimal and these sur llance procedures are performed on a cyclic basis. The paddle must be removed to allow DG, .md the F Bus to be synchronized to the A Bus and the grid (DG2 and the G Bus to be synchronized to the B Bus and the grid). Otherwise, the 32/lFE (32/lGE) relays will trip instantaneously and prevent the two buses from being synchronized.

SAFETY EVALUATION: The probability of occurrence or consequences of a previously evaluated accident or malfunction of equipment important to safety are not increased. The change in the surveillance procedures will not increase the probability of a Loss of OITsite Power (LOOP). The operating status of the overpower relay is not an initiating event for a LOOP. The applicable DG is declared inoperable during the performance of the sun'eillance procedure ard the required Limiting Condition for Operation actions are performed.

The operability of the redundant and independent safety related electonic distribution system and the DG j is not afTected by the change to these procedures. The change in the procedures will not create a new pathway to release radioactive material. There is no reduction in the protection of public health and 127-

i safety, since the function of the operable equipment required to control the release of radioactive material is not affected. The change to the procedures will allow the F (G) bus to be synchronized to the grid and to the normal power supply after load sequencing tests are completed. No other circuitry, such as the undervoltage protection, load shedding, and load sequencing, is affected by the change to these procedures. No new or different type of accidents or equipment malfunctions are introduced by these procedure changes. The change to resynchronize the DG and associated safety related bus to the grid after testmg has been completed is not an accident initiating event. The failure of a DG has been previously evaluated in the SAR. The margin of safety as dermed in the basis for any Technical l Specification has not been reduced. The surveillance testing procedures comply with the required actions for inoperable equipment.

Procedure 6.1RHR 101 Revision 61 l

Procedure 6 2RHR 101 (Revision 7) i (USQE 1998-0057)

TITLE: RHR Test Mode Surveillance Operation (IST)(Div 1)(6.1RIIR.101)

RHR Test Mode Surveillance Operation (IST)(Div 2)(6.2RHR.101) l

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DESCRIPTION: These procedures were revised to remov( e Residual Heat Removal (RHR) pump ditTerential pressure  !

upper limit of 189 psid. This trpper differential pressure limit was derived from the upper flow limits l specified for the RilR pumps. However, maximum flow of the RilR pumps is physically restricted by j

orifices. Flow testing per Design Change 89-252 indicated maximum RHR pump flow with ,

RI1R-MOV-MO27A/B full open did not exceed 8900 gpm, which is well within the capabilities of the RHR pumps. There is little value added by testing the upper limit of flow with a quarterly surveillance test, given that the RHR pump maximum flow is dictated by the physical plant design (i.e., the restricting orifices). As long as the physicalintegrity of the orifices and system piping is maintained, the pumps cannot run out. Physical integrity of the orifices is periodically inspected via Preventive Maintenance.

An upper differential pressure limit was established per Inservice Testing Program guidelines.

SAFETY EVALUATION: The operation of the RHR/ Low Pressure Coolant Injection (LPCI) system during normal plant operation ,

or during plant transient operation is not being changed (i e., the RHR/LPCI pumps will not be allowed to run out). The physical configuration of the RHR/LPCI system is not being modified. This change does not affect any accident precursors. It will not afTect the ability of the system to mitigate the consequences of an accident; therefore, there will be no increase in radiological dose and no increase in the consequences of an accident. There is no increase in the probability of occurrence or consequences of a malfanction of equipment important to safety. No new operating modes are being introduced.

Additionally, the system operation during surveillance testing as specified by 6.lRHR.101 and 6.2RHR.101 will remain unchanged. No new or different types of accidents or malfunctions are introduced and there is no reduction in the margin of safety as defined in the basis for any Tecimical Specifications.

Procedure 61RPS 305 (Revision 3)

Procedure 6 2RPS 305 (Revir. ion 3)

TITLE: MSIV Channel Functional Test (Div 1) (6.1RPS.305)

MSIV Channel Functional Test (Div 2) (6.2 PRS.305)

DESCRIPTION: The revisions to these procedures added instructions to allow testing of the Main Steam Isolation Valve (MSIV) closure scram function when the reactor mode switch is not in RUN. These revisions are necessary to demonstrate that the MSIV closure scram function is operable prior to entering the mode that requires the scram function to be operable. The instructions pull fuses to deenergize SA K11 relays to defeat the MSIV closure scram bypass and install jumpers to preserve the Group I isolation bypass to allow testing of the MSIV closure scram function. These procedure changes make the procedures consistent with Current Technical Specification 1.0.J requirements for testing to ensure operability prior to mode change, as well as with Improved Technical Specification Surveillance Requirement 3.0.4.

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1 SAFETY EVALUATION: The steps and actions (i c., pulling fuses and installing jumper s) added to these procedures are typical of similar steps and actions in other plant procedures and do not present a new or unusual activity or install any new permanent equipment. The probability of an error in pulling fuses or installing jumpers is slightly increased, since the opportunities for such an error are increased, however, since the activity is limited to one Reactor Protection System (RPS) subchannel at a time, the increase in this type of error does not increase the probability ofoccurrence of a previously evaluated plant event. Closure of an MSIV or reactor scram from such low initial power levels does not constitute a threat to the integrity of any fuel l

banier and a subsequent release of radioactive material endangering the health and safety of the public.

This activity does not reduce the ability of any mitigation system to mitigate the consequences of an event.

The SA-K11 relays are permissive only when the Reactor Mode Switch is in RUN. Deenergizing 5A-K1 I relays will not alTect the RPS components that actually provide the MSIV closure scram function.

Preserving the Group I isolation low vacuum bypass by installation of a temporary jumper will not increase the probability of an equipment malfunction in the Group I isolation components as the low vacuum Group 1 isolation is already bypassed during the plant conditions which require the jumper installation. Dose consequences are bounded by operations below the RUN mode. No new types of plant events or failure modes are created. This activity does not change (increase) the MSIV closure scram permissive setpoint or the low vacuum Group I isolation setpoint; therefore, it does not reduce the margin of safety as defined in the basis for any Technical Specification.

Procedure 6 ISGT.401 (Revision 31 1

Procedure 6 2SGT.401 (Revision 3)

Procedure 6 I SGT.501 (Revision 31 hocedure 6 2SGT.501 (Revision 3)

TITLE: SGT A Charcoal Filter Leak Test, Fan Capacity Test, SGT H Cooling Flow Test and Check Valve IST (Div 1)(6.lSOT.401)

SGT B Charcoal Filter Leak Test, Fan Capacity Test, SGT A Cooling Flow Test and Check Valve IST (Div 2)(6.2SGT.401)

SGT A HEPA Filter and Component Leak Test (Div 1)(6.1 SGT.501)

SGT B llEPA Filter and Component Leak Test (Div 2) (6.1 SGT.50l) l DESCRIPTION: Additions were made to these procedures to direct operator action as a result of failed Standby Gas l Treatment (SGT) crosstic flow criteria. Also, steps were reordered to leave SGT-V-49 open when post j

work testing is performed so that cooling flow will be available to both trains of SGT. The Improved Technical Specification (ITS) acceptance criteria for SGT flow was lowered to < 1602 cfm with the ditTerential pressure control valve for the train full open. The requirement to declare both SGT trains inoperable during testing was deleted as this is not required by ITS.

SAFETY EVALUATION: The changes to reflect ITS requirements and operator actions for improper SOT crosstic flow results do ,

not affect plant components or the test process. The reordering of steps does not afTect any precursors I for events previously evaluated in the SAR. These revisions will result in better compliance with SAR l i requirements. Opening SGT-V-49 prior to performing the post maintenance testing fan run eliminates l l a potential condition that could have atrected the consequences of events evaluated in the SAR (i e , loss '

of coolant accident and fuel drop). During the test, the train being tested is inoperable. The test helps i ensure the long term probability of a malfunction remains low. Changing the sequence when SGT V-49 l is opened reduces the probability of a malfunction of the operable SGT train. It reduces the potential for a fire in a post-accident scenario. No new type of plant activity is created as a result of the changes to  ;

these procedures. The revised step sequence reduces the possibility of dangerous heat buildup which l

, could damage both trains of SGT. The loss of secondary contaimne! t evaluation in the SAR bounds any

! malfunction of the SGT system The o!Tsite dose effects of plant activitics and events are not changed by I

these procedure revisions. No other margins of safety are afTected.

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Procedure 7 0.7 (Revision 3i TITLE: Scaffolding Construction and Control DESCRIPTION: Changes included in this procedure revision included: 1) use of pre-approved seismic bracing options that will not require job specific Design Engineering scalfolding evaluations, 2) refinement of requirements for seismic separation, 3) implementation of a formal format for scaffold engineering evaluations, t

4) incorporation of applicable Occupational Safety and Health Administration standard requirements, l

5) identification of steps that implement industrial safety requirements versus Design Engineering l requirements, and 6) allowing use of composite plywood planking work platforms.

SAFETY EVALUATION: The USAR ckx:s not specifically identify any plant events associated with the failure of r,cafTolding. The i goveming plant scenario applicable to the nuclear safety evaluation of scalibiding is the potential for

)

equipment maltimctions caused by seismic interactions during a Design Basis Earthquake (DBE); i e.,

the etTects of a non-Class I component on nearby Class I equipment. The application of USAR code regarements to seismic bracing and separation requirements ensures that the scafrolding will not interfere with the proper functioning of any Class I equipment, and thus will not affect any Class I equipment nor influence any plaut events. The overall probability of an occurrence of a plant event during or following a DBE as previously evaluated in the USAR remains unchanged and the radiation doses associated with a DBE evaluated in the USAR mnain unchanged. The only potential interection between scafTolding and equipment important to safety would be the collapse of scatTolding onto Class I equipment during an carthquake; This procedure change directs that scaffolding be erected in a way that applies the same structural design margins applicable to Claw I equipment to scafrolding design. Thus, it will not increase the probability ofoccurrence of a malfunction of any Class I equipment and cannot infiuence radiological releases associated with Class I equipment malfunction. By directing the application of Class I design margin to the erection ofscafrokling, the procedure ensures that scaffolding will not fail during a scismic event, and thus will not create a new plant event or malfunction of Class I equipment. Adhering to the requirements of this procedure ensures that the margins of safety defined in the basis for the Technical Specifications, in this case resulting from seismically induced failures, will remain the same.

ProcedurrG o 15 (Revision 2)

TITLE: Station Painting Ouidelines DESCRIPTION: This procedure revision allows the use of a high solids coating material (98%

• 1% by volume)in the l areas where charcoal filters are present.

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SAFETY l EVALUATION: This activity does not involve any equipment or systems credited as event initiators or with mitigating the  !

consequences of a plant event or equipment malfunction. It does not affect the plant's ability to contain .

radioactive materials either during normal operation or post event. Excessive moisture or organic I materials will reduce the iodine absorption capability of the charcoal filter; however, most of these l

materials are removed by the demister, heater, or filter. The amount of Volatile Organic Compounds (VOCs) (3%) into the air is so small and with the further dilution of the VOC into the air, the amount reaching the charcoal fihers will have negligible impact. The other factor to be considered is the Reactor Building normal ventilation and how this system helps evacuate the VOCs from the building. High ,

EITiciency Particulate Absolute (HEPA) filters are installed before and after the charcoal adsorbers to I minimize potential release of pasticulates to the environment and to prevent clogging of the iodine j adsorbers. The charcoal adsorbers are installed to reduce the potential release of radiciodine to the j cmironment. The use of a high solids coating material in the areas where charcoal filters are present is I acceptable because coatings with high solid contents have 0.181b/ gal maximum of VOC content. The  !

requirements of 10CFR50.49 apply to the Standby Gas Treatment system so that it is capable of j withstanding the effects of a Loss of Coolant Accident (LOCA) to ensure that the offsite doses wi'll be '

within the limits of 10CFR100. Insignificant amounts of VOCs reaching the charcoal filters will have i no impact on the charcoal and should one train get clogged for some reason, the other train shall l automatically start and, therefore, the design basis accidents of a LOCA and a refueling accident will not !

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be compromised or altered. His activity does not introduce any other plant events than those previously evaluated It does not reduce the margin of safety as defined in the basis for any Technical Specification.

If painting, fire, or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals, or foreign materials, the same tests and sample analysis shall be performed as required for operational use.

Procedure 7.015 (Revision 41 (USQE 1998 0018) 1 TITLE: Station Painting Procedure J

l DESCRIPTION: Ris procedure resision addresses the potential impact of Volatile Organic Compounds (VOCs) on the  !

chamoal fdters of essential air handling equipment. It provides for strict control of all coating materials that will be used in areas where contamination of the charcoal filters by VOCs is possible.

SAFETY EVALUATION: This activity does not involve any equipment or systems credited as event initiators nor does it adversely affect any systems credited with terminating transients. It does not affect the plant's ability to contain radioactive materials either during normal operation or post event. Excessive moisture or organic )

materials will reduce the iodine adsorption capability of the charcoal fdters; however, most of these materials are removed by the demister, heater, or filter. The amount of VOC (3%) into the air is so small and with the further dilution of the VOC into the air, the amount reaching the charcoal filters will have negligible impact. Iligh efficiency particulate absolute filters are installed before and after the charcoal adsorbers to minimize potential release of particulate to the emironment and to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential release of radiciodine to I the environment. Because the radionuclide absorption is unaffected, the consequences of previously evaluated accidents are unchanged. Control of VOCs to 3% maximum by weight of all approved coating materials in the areas where charcoal filters are present precludes introduction of a new common mode failure of the charcoal filters. No new event initiators or precursors are introduced by this activity. The margin of safety as defined in the basis for any Technical Specification is not reduced. Technical Specifications require testing of the Control Room Emergency Filter System and the Standby Gas Treatment System in accordance with the Ventilation Filter Testing Program. To minimize the need for these tests, painting controls are established. Testing of the charcoal adsorbers shall be required any time there is reason to suspect the filters may have adsorbed more than 3% VOCs.

D wrdure 7.2.24 (Revision 20) it N)E 1998-0068)

'lTI1.H: Main Steam Isolation Valve (MSIV) Maintenance DESCRIPTION: This procedure was revised to allow installation of an inflatable pipe plug in the Main Steam (MS) line piping as a secondary containment boundary when the outboard MSIV is disassembled. Removal of the intemals of the MSIVs with no compensatory action could cause the Secondary Containment to become inoperable. The use of an inflatable pipe plug in each of the MS lines,just downstream of the MSIVs, will ensure that the Secondary Containment remains operable during refueling operations.

SAFETY EVALUATION: Ris procedure change will not increase the probability of an accident since installation of the pipe plug in place of the MSIV will not increase the probability of a Loss of OITsite Power, Loss of Shutdown Cooling, nor a refueling event, because there is no interaction of the pipe plug with any of these systems.

There is no increase in the consequences of the refueling accident since Secondary Contaimnent integrity will be maintained following and for the duration of a Refueling Accident. This change assures that Secondary Containment integrity, which is the only safety system affected, will be maintained in both normal and accident conditions. The pipe plug assembly is a passive component; therefore, the probability of a malfunction of equipment important to safety is not increased. There is no credible malfunction of the pipe plug not covered by the proposed normal and post accident surveillance requirements which will result in a loss of Secondary Containment integrity and a resulting increase in

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the con.w of a refueling accident. The installed pipe plug will not cause an accident of a difTerent l

type than previously evaluated. Installation of the pipe plug in place of the MSIV cannot affect the reactor core or any of the required safety systems. It will not alter the normal or accident operating parameters of any system, structure, or component. An arbitrary failure of the pipe plug is outside of the CNS licensing basis. The demonstration of Secondary Contain nent operability when the pipe plug is installed ensures that the margin of safety in the Technical Specifications is maintained.

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Procedure 7.2 24 (Revision 21) l (USQE 1998-0075)

TITLE: Main Steam Isolation Valve (MSIV) Maintenance DESCRIPTION: This procedure was revised to reduce the MSIV disc-seat contact from 3/16" to 1/8". This will ensure a tighter seal by increasing the seat stress. Due to years of maintenance on the MSIV valve seats, the seat width contact had become wider, thus increasing the seat area and reducing the sealing stress, and making it more difIicult to achieve an acceptable local leak rate test. A reduced seat width does not in itself alter the basic design of the seat as specified by the valve's manufacturer.

SAFETY EVALUATION: 'Ihis change does not reduce the valve's basic temperature and pressure rating, nor does it alter the design of the operator or its controlling component or circuitry. The change in seat width does not introduce any i accident initiators or precursors as described in the SAR. This activity has not made any changes to the  !

l valve's stem or the disc's guides, ensuring that closure time will not be afTected. Rather, by sligh'Jy altering the valve's basic seat width to increase seating stress, the probability of obtaining a tight seal is l enhanced, ensuring that the assumptions used in any existing accident analyses or engineering I assumptions remain valid. Even though the seat's stress levels will increase as the result of this activity, they will not exceed the metallurgical limits of either the seat's hard facing or its base metal by a comfortable margin. In the event the seat should fail, the basic geometry used to limit the consequences of this failure mode remain in place. This activity cannot result in a common mode failure as the seat stress levels remain within acceptable limits, and are fully bounded by the normal pressure stresses.

Therefore, this activity is incapable of increasing the consequences of a malfunction of equipment important to safety. No new interfaces are created between any existing systems or structures, no new equipment is added, and no operating parameters or operating procedures are altered. Therefore, the probability of any new and unanalyzed accidents is not increased. It does not alter the forces on any piping systems, result in new and analyzed forces on the valve's operator or internals, increase or decrease steam flow or pressure, alter the valve's basic operation, or reduce its ability to close under pressure. Therefore, this activity cannot result in a new and dilTerent type of equipment malfunction. The ability of the MSIVs to obtain a tight seal has been increased, thereby enhancing any existing margins of safety.

Procedure 7.2 2tl 12 (Revision 01 TITLE: Atwood and Morrill Extraction Steam Non-Return Valves DESCRIPTION: This is a new Maintenance Procedure to provide instructions for maintenance of the Extraction Steam non-return check valves and actuators.

SAFETY EVALUATION: This activity will not affect system performance and reliability or any system interface in a way that could lead to an accident. Systems and equipment will not be afTected by this activity such that they are degraded or operated outside their design or test limits. In addition, this activity does not increase the possibility ofoperator error or add complexity to human factor conditions such that the probability of an accident is increased. This activity does not affect or increase the radiation dose associated with the plant's response to any accident or equipment malfunction. The function of equipment designed to control the release of radiation is unalTected. The Extraction Steam system will retain its design function and the non-return valves (NRVs) will also function as designed aller maintenance has been performed.

The NRVs are considered nonessential components and maintenance work completed on these

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l components with the new procedure will not affect the capability of any equipment important to safety to perform its design function. Use of this procedure will not create the possibility of a different type of plant event or malfunction of equipment important to safety. This activity does not involve any margin of safety as defined in the basis for any Technical Specification.

Procedure 7 3.7.2 (Revision 0)

TITLE: REC-REL-lFR and REC-REL-lGR Functional Test l

DESCRIPTION: This new procedure provides instmetions to functionally test time delay relays REC-REL-lFR(IGR).

This test provides post-maintenance testing to demonstrate the relays will perform their function after calibration / replacement. The procedure allows on-line maintenance to be performed on these components.

SAFETY i

EVALUATION: The afrected component is tested while the applicable Limiting Condition for Operation is entered. The

) test switchrjumper is placed in parallel with the isolation relay contacts that operate the time delay relay ;

and does not affect other functions of the isolation relay or the other operable Reactor Equipment Cooling ,

I (REC) subsystem. The testing only affects the time delay relay and the associated motor operated valves j in the inoperable REC subsystem. The test aligns the inoperable subsystem in its safety configuration and I does not affect the operable REC subsystem or the isolation logic circuit. The circuit jumpered is isolated from the power source via a fuse dedicated to the time delay circuit. This test does not increase the l

probability of occurrence or consequences of a plant event or malfunction of equipment important to l l safety and does not reduce the margin of safety as defined in the basis for any Technical Specification.

l Procedure 7.3 27.2 (Revision 0)

(USQE 1998-0028) i

{

TITLE: 125/250 VDC System Ground Locator i 1

L DESCRIPTION: This new procedure was developed to provide guidance for locating grounds in the 125 and 250 VDC l systems. It utilizes the 'DC Scout" ground locator to locate resistive or non-resistive current paths from  !

DC battery distribution systems to building ground, without deenergizing components or loads in the l system.

SAFETY EVALUATION: The 125/250 VDC system provides safety system auxiliary functions, but is not an initiator of any of the l accidents previously evaluated in the SAR. The introduction of the pulsating milliampere current along the same path as the existing ground current will not cause the 125/250 VDC battery distribution system l

l to be operated outside any loading limit. The disabling of a ground detection system to perform the l

investigation is prudent due to the fact that the ground detection system has performed its intended function and now a detailed investigation must be performed. The location and correction of grounds will ensure the 125/250 VDC systems will be available to provide the essential auxiliary function required during previously evaluated accidents. This procedure does not cause the loss of a single DC system, which is analyzed for in the USAR, and therefore will not increase the consequences of a previously evaluated accident. The procedure will not increase challenges to safety systems. Precautionary steps taken in the procedure will ensure no malfunction of any equipment important to safety previously  !

evaluated in the SAR can be increased beyond that already created by the existing fault. Performance of this procedure (k>es not cause the associated DC subsystem to deviate from the design basis requirement of providing adequate and acceptable source power to the equipment performing accident mitigation j functions and, therefore, cannot affect a malfunction of equipment important to safety previously evaluated j in the SAR. Loss of one of the DC subsystems is analyzed for in the SAR. Therefore, this ground  !

investigation does not cause an increase in malfunctions of dose mitigation equipment and cannot  !

increase the consequences of a malfunction of equipment important to safety previously evaluated in the  !

SAR. The level of current which is to be used will not be at an amplitude which will cause the inadvertent pickup of any coil that was not already energized by the fault, nor would the disabling of the ground detection system on the afTected bus while the investigation is going on cause a different type of

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l malfunction. The basis of the Technical Specifications for the DC subsystems recognizes that the " loss of any DC electrical power subsystem does not prevent the minimum safety function from being performed." ne performance of this procedure does not cause any DC subsystem to become inoperable and therefore does not reduce the margin of safety as defined in the basis of any Technical Specification.

Procedure 8 4 (Revision 14)

TITLE: Routine Sampling and Sample Valve Control DESCRIPTION: This change added Component Identification Code (CIC) numbers to components that did not have CIC numbers since original constmetion. It also added additional components that were identified as not being a part of any other checidist, deleted components that were already included in Operations checklists, and modified component descriptions. More detail was also included in generic sample instructions.

SAFETY EVALUATION: The addition / deletion / revision of sample valve information in the valve position list does not alTect the ability of systems, subsystems, or components (SSCs) important to safety to perform as described in the SAR. Rese changes do not affect any of the assumptions made for plant events previously evaluated in the SAR. They do not affect the ability of SSCs to mitigate the consequences of plant events evaluated in the SAR. Containment isolation functions are not impacted by this procedure revision. Configuration control of these valves, and operation of the SSC during normal operation, is maintained via Station Procedures to prevent operation of the SSCs outside of their design. These changes do not create any new plant event assumptions which could increase the consequences of a plant event due to equipment malfunction. The possibility of a different type of phmt event or malfunction of equipment important to safety is not created. This revision does not afTect the margin of safety for any Technical Specification.

The safety functions of SSCs described in the Technical Specifications are unafTected by the procedure changes since valve configuration is maintained consistent with that assumed in the CNS Design Basis Documents and Technical Specification Bases.

hpcedure 9.4 TELDOS (Revision h TITLE: Teledosimeter Installation, Relocation, and Removal DESCRIPTION: Changes included in this procedure revision included: 1) allowing the teledosimeter and power supply I to be mounted to nonsensitive plant components using Velcro or Ty-rap mounting,2) adding a requirement to obtain concurrence from Civil Engineering ifinstalling teledosimeter and power supply  ;

in Reactor Building within five feet of sensitive essential components,4) relaxed requirement for '

Electrical Engineering to approve the proposed and final teledosimeter wiring installation location, l

5) added note to allow use of extension cords for temporary installations provided a transient '

combustibles permit is obtained, and 6) added precautions and limitations related to installation.

SAFETY EVALUATION: This procedure provides the necessary limitations and has adequate controls to assure that plant 1 equipment will not be adversely affected. A review of Appendix G of the USAR indicates that this system i does not adversely affect any structures, systems, or components (SSCs) which can fail and cause a plant event or increase any plant event consequences. The installation of the system does not afTect any radiological release barriers and provides no new release pathways. The probability of occurrence of a previously evaluated malfunction of equipment important to safety is not increased and the teledosimetry system will not adversely affect any SSC in such a way as to increase the radiological consequences associated with a malfunction of the equipment. The results of Special Procedure 97-009 provide the l assurance that SSCs will not be adversely affected by the radio frequency signals emitted. No new types i ofplant events or equipment malfunctions are introduced. Since the system does not adversely afTect any

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plant SSCs, the procedure does not reduce the margin of safety as defined in the basis for any Technical Specifications.

Procedure 15. AR 301. Revision 0 6 (USQE 1999-0023)

TITLE: Condenser Air Removal isolation Valves Functional Test i

DESCRIPTION: This new procedure was developed to functionally test the condenser isolation air removal valves. A j

similar procedure, Surveillance Procedure 6.4.8.11, previously existed to demonstrate operability of these j valves. Ilowever, when Design Change 91-088 was implemented during RF15 to move the condenser l isolation safety function from the air removal isolation valves to the mechanical vacuum pump suction j and discharge valves, Procedure 6.4.8.1I was deleted. No routine maintenance had been performed on i these valves since the design change. Thermography activities to troubleshoot poor condenser / Steam Jet l

Air Ejector (SJAE) perfonnance identified at least one condenser air removal isolation valve that may be j closed. Therefore, this procedure was developed to individually stroke these valves to demonstrate that l the valve actuator and butterfly orientation is correct. This procedure has no set frequency and may be .

performed at the request of Operations or Engineering when orientation of applicable components is in l question. This procedure, if successful in identifying a closeCsolation valve, will provide the basis for l improving SJAE and main condenser performance.

l SAFETY i EVALUATION: The loss of condenser vacuum is a transient described in the SAR. Ilowever, this procedure does not increase the possibility of a loss of condenser vacuum because appropriate controls are in place to l maintain control of and minimize changes to condenser vacuum. Momentary cycling of the condenser air removal isolation valves will maintain the system within normal design and operation. The off-gas isolation function will remain unchanged, so any mitigation activities associated with accidents and transients will remain unchanged. The SJAE system (with the exception of the mechanical vacuum pump air removal isolation valves and the system pressure boundary) is not relied upon in any accidents or operational transients to mitigate the consequences of those events. This activity will not have any impact on the mechanical vacuum pump air removal isolation valves or pressure boundary. If this actisity is successful in obtaining better SJAE and condenser performance, it may result in higher off-gas radiation in the SJAE room during the winter. This dose increase is due to a reduction in non-condensible hold-up (decay) time inside the main condenscr. Dose rates that are experienced in the summer months would be common throughout the year. With the 30 minute hold-up line and Augmented Off Gas (AOG) charcoal bed hold-up, sufIlcient delay time occurs and no change in dose released to the public will occur.

No permanent hardware changes are being perfbrmed so no new equipment failure mechanisms are being introduced. Individual valve closures will be conducted in a slow, controlled fashion with provisions for restoring system operation to a stable state should undesirable responses be observed. All equipment used in this activity is being operated within a normal range and will not be challenged in an adverse way.

Because operation of the system is within normal design and because no new equipment is being added, the possibility of a new type of accident of malfunction of equipment important to safety is not created.

Technical Specification 3.7.5 requires a gross gamma activity rate limit of 1.0 Ci/sec noble gas and the Bases section states that a gross gamma activity rate limit of 0.39 Ci/sec noble gas release rate at the air ejectors (after 30 minute delay) is used as the source term for the accident analysis of the AOG system.

None of the rates specified for AOG are being altered and the highest expected operating rate level is not being altered. Therefore, the margin between the Technical Specification rates and expected operating rates is not being altered.

Procedure 15 pRM 304 (Revision 11 TITLE: Multi-Purpose Facility (MPF) Kaman Monitor Functional Test DESCRIPTION: This procedure was revised to make the procedure methodology consistent with testing performed on other Kaman efIluent monitors. Recording of various parameters was deleted since the monitor will now be restored by performing monitor reset which will ensure proper parameters are present. This monitor

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f is not included in the Technical Specifications, but is described in the USAR. The use of a radioactive )

l source for the soun:e check on the particulate channel was deleted as this is only a Tecimical Specification l requirement on gas channels. The MPF gas channels have radioactive sources employed for the built-in l source check. The particulate channel utilizes a LED source check which provides adequate assurance l that the detector is functional.

SAFETY EVALUATION: The subject monitor is a passive monitoring system and does not have the capability of causing an accident evaluated in the SAR nor impacting accident analysis assumptions, and does not have the capability to increase the consequences of a previously evaluated plant event. The use of the LED check ,

source installed in the detector provides adequate indication of monitor response. Procedural guidance l

' has been provided to ensure the features described in the USAR are tested and proven to be functional.

Ute changes to the procedure provide the methodology to ensure proper functional testing and do not alter the monitor's ability to perform its safety design basis function. The changes do not operate the efIluent monitoring system outside of its system design and do not afTect other systems, subsystems, or components important to safety. The monitors are not capable of creating a difTerent type of plant event  ;

or malfunction of equipment important to safety. The MPF monitor is not included in the Technical  ;

Specifications. Ik) wever, the same alarm setpoint required for Technical Specification related monitors is employed for the MPF. The procedure changes do not afTect the ability to obtain and calculate gaseous ef11uent : lease data during nonnal operations or post-accident. Margins of safety remain unafTected.

Procedure 15 RF 601 (Revision 0)

(USQE 1999-0002) l TITLE: Reactor Feed Pump Turbine Overspeed Test i DESCRIPTION: This new procedure was developed to provide the necessary controls to conduct a Reactor Feed Pump Turbine mechanical overspeed test while the plant is operating on the remaining Reactor Feed Pump.

The test will be performed utilizing extraction steam with the respective Reactor Feed Pump uncoupled from the turbine.

SAFETY EVALUATION: The probabdity of the Feedwater Controller Failure - Maximum Demand abnormal operational transient is not increased because the affected feedwater pump will be uncoupled from its associated turbine during conduct of the overspeed functional testing. As such, the atfected feedwater pump will be incapable of delivering feed flow, regardless of the controller signal. The probability of the Loss of Feedwater Flow abnormal operational transient is not increased because the likelihood of operator error, controller failure, or pump failure causing a loss of feedwater flow during implementation of this procedure would be similar to that which currently exists during normal plant startup and shutdown maneuvers. Ilowever, since the plant will be operated on only one feedwater pump, any failure of the operating pump will result in a loss of normal feedwater. The probability of the Anticipated Transient Without Scram - Loss of Feedwater Flow transient is also not increased for the same reasons. The probability of a missile event is not increased since this procedure contains rigorous controls to ensure that turbine speed is tightly controlled during functional testing. The procedure also contains direction to temporarily reset the electrical overspeed trip device above the desired mechanical overspeed trip device trip setting to provide

added protection in the unlikely event of a significant turbine speed increase transient. In addition, the procedare directs plant operators to manually trip the turbine should speed exceed 6200 RPM. The consequences of accidents previously evaluated in the SAR are not increased since the seventy of the l previously analyzed events is not changed. Procedure 15 RF.60I will ensure that feedwater pump turbine overspeed setpoints are accurately set so that proper reliability is maintained and spurious feed pump trips due to overspeed do not occur during plant operation; i.e., this activity is intended to decrease the probability of equipment malfunction. The overspeed testing will be perfbrmed by carefully increasing turbine speed in small increments via manual control. This activity does not increase the radiological consequences of a malfunction ofequipment important to safety. The consequences of a malfunction are bounded by existing USAR abnormal operational transient analysis. There are no difTerent types of transients or accidents that can be postulated based on a single equipment failure or operator error as a result of the implementation of Procedure 15.RF.601. This activity will reduce the possibility of a

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spurious reactor feed pump turbine trip and will also ensure that proper turbine protection is prosided upon an actual overspeed event. Similar testing is periodically conducted on other CNS turbines Therefore, no new types of equipment malfunctions are introduced. This procedure functionally checks the reactor feed pump mechanical overspeed setpoints to ensure vendor design requirements are met There are no specific Technical Specification requirements associated with reactor feed pump overspeed trip functional checks or trip settings. As such, the margin of safety is not reduced by the implementation of Procedure 15.RF.601.

Procedure 15 RFI)01 (Revision 11 (USQE 1999-0003)

TITLE: Reactor Feed Pump Turbine Overspeed Test DESCRIPTION: Procedure 15.RF.601, Revision 0, was originally generated to test the Reactor Feed Pump Turbine mechanical overspeed. This revision added the necessary steps to also perform testing of the electrical overspeed mechanism. In addition, steps were added to temporarily open sliding links in the turbine control pneumatic assemblics to allow stroking of the steam nozzle block valves with the turbine trip present.

SAFETY EVALUATION: The probability of the Feedwater Controller Failure - Maximum Demand abnormal operational transient is not increased because the affected feedwater pump will be uncoupled from its associated turbine during conduct of the overspeed functional testing. As such, the affected feedwater pump will be incapable of delivering feed flow, regardless of the controller signal. The probability of the Loss of Feedwater Flow abnormal operational transient is not increased because the likelihood of operator error, controller failure, or pump failure causing a loss of feedwater flow during implementation of this procedure would be similar to that which currently exists during normal plant startup and shutdown maneuvers. However, since the plant will be operated on only one feedwater pump, any failure of the operating pump will result j in a loss of normal feedwater. The probability of the Anticipated Transient Without Scram - Loss of f Feedwater Flow transient is also not increased for the same reasons. The probability of a missile event is not increased since this procedure contains rigorous controls to ensure that turbine speed is tightly controlled during functional testing. During the testing of either the electrical or mechanical overspeed function, the function not being tested is still intact and able to trip the turbine. In addition, operations personnel are closely monitoring speed during testing and are instructed to trip the turbine manually if necessary. The consequences of accidents previously evaluated in the SAR are not increased since the severity of the previously analyzed events is not changed. This procedure will ensure that feedwater pump turbine overspeed setpoints are accurately set so that proper reliability is maintained and spurious feed pump trips due to overspeed do not occur during plant operation; i c., this activity is intended to decrease the probability ofequipment malfunction. The overspeed testing will be performed by carefully increasing turbine speed in small increments via manual control. This activity does not increase the radiological consequences of a malfunction of equipment important to safety. The consequences of a malfunction are bounded by existing USAR abnormal operational transient analysis. There are no different types of transients or accidents that can be postulated based on a single equipment failure or operator error as a result of the implementation of this procedure. This activity will reduce the possibility of a spurious reactor feed pump turbine trip and will also ensure that proper turbine protection is provided upon an actual overspeed event. Similar testing is periodically conducted on other CNS turbines.

Therefore, no new types of equipment malfunctions are utroduced. This procedure functionally checks the reactor feed pump overspeed trip setpoints to ensure vendor design requirements are met. There are no specific Technical Specification requirements associated with reactor feed pump overspeed trip functional checks or trip settings. As such, the margin of safety is not reduced by the implementation of this procedure.

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MISCELLANEOUS CllANGES l Setnoint Chance Renuest 98-02 TITLE: Setpoint Change for iIPCI-LS-91 A/B, Suppression Pool liigh Ixvel DESCRIPTION: The setpoint ofIIPCI-LIS-91 AA3 was decreased from 875' 3.75" to 875' 3.0" to allow the setpoint to be within the adjustable range for both switches. A setpoint of 875' 3.0" corresponds to a level of 3" 110 2 above normal suppression pool water level. The setpoint change only affects the actuation of the trip (Iligh Pressure Coolant Injection [IIPCI] suction transfer) and not the performance or reliability ofIIPCI injection or the level switches themselves.

SAFETY EVALUATION: The high suppression pool level setpoint is not associated with accident initiators or precursors.

Lowering the setpoint ofIIPCI-LS-91 A/B will not affect the ability of HPCI to start and inject to supply cooling water to the vessel when required to mitigate the consequences of a plant event, nor will it affect thrust loading or pnssure suppression capabilities of the suppression chamber during an event. The new setpoint does not invalidate or afTect any of the assumptions used in the determination of consequences for plant events. The operation ofIIPCI-LS-91 A/B remains unchanged by this activity oflowering the setpoint. No new failure modes are introduced that could affect the availability ofIIPCI. Therefore, this setpoint change will not increase the probability of occurrence or consequences of a presiously evaluated equipment malfunction. No hardware changes are made as a result of this setpoint change and it will not affect the way in which the level switches function or operate. This setpoint change does not create the possibility of a different type of plant event or equipment malfunction. Lowering of the setpoint from 875' 3.75" to 875' 3.0" increases the margin from the Technical Specification limit of 875' 5". Therefore, this activity does not reduce the margin of safety as defined in the basis for any Technical Specification.

Setnoint Chance Reauest 98-07 Setnoint Chance Reauest 98-10 Setnoint Chance Reauest 98-11 Setnoint Chance Reauest 98-12 Setnoint Chance Reauest 98-13 Setnoint Chance Reauest 98-14 Setnoint Chance Reauest 98-15 Setnoint Chance Reauest 98-16 Setnoint Chance Reauest 98-17 Setnoint Channe Reauest 98-18 Setnoint Chance Reauest 98-19 Setnoint Chance Reauest 98-20 Setnoint Chance Reauest 98-21 Setnoint Chance Reauest 98-22 Setnoint Chance Reauest 98-23 Setnoint Chance Reauest 98-24 Setnoint Chance Reauest 98-25 l

Setnoint Chance Reauest 98-26 TITLE: Setpoint Change Requests (SCRs) for More Restrictive Setpoint Changes for Improved Technical l Specification (ITS) Implementation DESCRIPTION: The setpoint changes covered by this safety evaluation are a result of revision to various setpont i calculations in an effort to calculate allowable values, consistent with the philosophy in NUREG-1433 for Improved Technical Specifications. The calculations were revised in ace,rdance with the General Electric Instrument detpoint Methodology and CNS procedures. Analytical limits usd in the calculations an: derived from the limiting values of the process parameters obtained from the safety analysis or other appropriate documents. The more : 'rictive setpoint changes covered by this analysis are for ITS i

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l functions in the following systems: Reactor Protection System (RT S), Feedwater and Main Turbine liigh Water Level Trip, Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT),

Emergency Core Cooling System (ECCS), Reactor Core Isolation Cooling (RCIC), Primary Containment i

Isolation, Secondary Containment Isolation, and Low-Low Set System. The specific setpoint changes I covered by this evaluation are as follows:

SCR 98-07 NBI-PS-SSA/B/C/D, Reactor Vessel Iligh Pressure - RPS Scram Original Setpoint = 1049.0 psigt Revised Setpoint = 1047.0 psigi SCR 98-10 PC-PS-12A, DrywellIligh Pressure - Reactor Scram, Groups 2 and 6 Isolations, and Emergency Diesel Generator Start Original Setpoint = 1.84 psig1, Revised Setpoint = 1.625 psig1 SCR 98-ll PC-PS-101 A/B/C/D, Drywell High Pressure - Ifigh Pressure Coolant Injection (HPCI), Core Spray, and Residual IIeat Removal (Low Pressure Coolant Injection Mode) Initiation Original Setpoint = 1.84 psig1, Revised Setpoint = 1.625 psig1 SCR 98-12 PC-PS-12B/C/D, Drywell High Pressure - Reactor Scram, Groups 2 and 6 Isolations, and Emergency Diesel Generator Start l Original Setpoint = 1.84 psigt, Revised Setpoint = 1.725 psigI SCR 98-13 RFC-LA-121 A/B/C, Iligh Vessel Water Level - Main Turbine and Reactor Feed Pump Turbine Trip Original Setpoint = 54.9 inches 1, Revised Setpoint =- 52.50 iides t l SCR 98-14 NBI-PS-102A/B/C/D, Reactor Vesselliigh Pressure - Alternate Rod Insertion and

• ATWS RPT Original Setpoint = 1088.0 psigI, Revised Setpoint = 1073.0 psigi SCR 98-15 CS-AM-45A/B, Core Spray Pump Discharge Flow Alarm Module Originai Setpoint = 1340.0 gpml, Revised Setpoint = 1613.0 gpm!

SCR 98-16 NBI-LIS-101 A/B, High Reactor Vessel Water Level - Trip IIPCI and RCIC Turbines Original Setpoint = 54.00 inchest , Revised Setpoint = 48.75 inches 1 SCR 98-17 NBI-LIS-101C/D,Iligh Reactor Vessel Water Level - Trip iIPCI and RCIC Turbines Original Setpoint = 53.60 inches t , Revised Setpoint = 48.75 inches 1 SCR 98-18 IIPCI-FIS-78 (Switch 1), HPCI Pump Discharge Low Flow - Minimum Flow Bypass Valve Open Signal Original Setpoint = 485.00 gpml, Revised Setpoint = 517.50 gpmi SCR 98-19 MS-REL-K5A/B, Automatic Depressurization System Initiation Timer Original Setpoint = 105.00 seconds t , Revised Setpoint = 98.08 secondsi SCR 98-20 MS-PS-134A/B/C/D, low Main Steam Line Pressure - Group 1 Primary Containment isolation Original Setpoint = 840.00 psig!, Revised Setpoint = 849.00 psigi SCR 98-21 MS-DPIS-116,117,118,119A/B/C/D, Main Steam Line High Flow - Group 1 Primary Containment Isolation Original Setpoint = 144.17% t , Revised Setpoint = 140.12 %1 SCR 98-22 IIPCI-DPIS-76n7 (Switch 2), HPCI Steam Line Iligh Flow - Group 4 Isolation and HPCI Turbine Trip Original Setpoint = 270.0%, Revised Setpoint = 241.0%

SCR 98-23 IIPCI-DPIS-7607 (Switch 1),IIPCI Steam Line High Flow - Group 4 Isolation and HPCITurbine Trip Original Setpoint = -270.0%, Revised Setpoint = -241.0%

SCR 98-24 IIPCI-PS-68A/B/C/D, HPCI low Steam Line Pressure - Group 4 Isolation and iIPCI Turbine Trip Original Setpoint = 135.00 psig1, Revised Setpoint = 137.20 psig1 SCR 98-25 RCIC-PS-87A/B/C/D, RCIC Iow Steam Line Pressure - Group 5 Isolation and RCIC Turbine Trip Original Setpoint = 80.00 psig1, Revised Setpoint = 94.40 psig t

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SCR 98-26 RR-PS-128A/B, Reactor Recirculation Pressure Switch Original Setpoint = 83.75 psigi, Revised Setpoint = 79.20 psigi SAFETY EVALUATION: Application of the CNS instrumentation setpoint methodology results in selected instrument setpoints which more accurately reflect total instrementation loop accuracy, as well as that of test equipment and setpoint drift between surveillances. The methodology ensures the afTected instrumentation remains capable of mitigating design basis events as described in the safety analyses and that the results and consequences described in the safety analyses remain bounding. The instrumentation included in the evaluation is not an initiator of any analyzed event. The subject changes provide more stringent requirements for operation of the facility. nese requirements do not result in operation that will increase the severity of an analyzed event and do not alter assumptions relative to mitigation of an accident or transient event. The setpoint changes will not cause any malfunctions or failures of equipment important to safety other than those already postulated and analyzed in the SAR. The setpoint changes are conservative and as a result there are no equipment limits impacted. Revision of the previous plant setpoints to more conservative values will cause the essociated safety / trip functions to occur earlier in the transient and/or accident scenario. The proposed setpoint changes do not introduce a new mode of plant operation and do not involve physical removal and replacement of any component, system, or structure in the plant. Plant equipment will not be operated in a manner ditTerent than previously operated with the exception that the setpoints will be conservatively changed. Since operational methods remain unchanged and the operating parameters have been evaluated to maintain the plant within existing design basis criteria, no different type of accident or malfunction of equipment important to safety is created.

Revision of the cetpoints to more restrictive values will not reduce the current margin of safety.

Setnoint Chance Reauest 98-08 (USQE 1998-0016)~

TITLE: Setpoint Change for Average Power Range Monitor (APRM) Neutron Flux - Iligh (Startup) Scram, NM-NAM-AR2, 3, 4, 7, 8, 9 l

DESCRIPTION: This more restrictive setpoint change was the result of Nuclear Engineering Department Calculation 98-024 which was developed to calculate an allowable value in consideration of the transition to Improved Technical Specifications. This change is also based on an increase in surveillance inten al to six months (+25% grace period). The calculation was developed in accordance with the General Electric I Instrument Setpoint Methodology and CNS Procedure 3.26.3. Analytical limits used in the calculation are derived from the limiting values of the process parameters obtained from the safety analysis or other appropriate documents. This Setpoint Change Request changed the setpoint for the APRM Upscale (Startup) Scram from 14.1%f to 13.0%f. This instrument setpoint provides an input signal to the Reactor Protection System for reactor scram on high neutron flux (startup) from the APRM.

SAFETY EVALUATION: Application of the CNS instrumentation setpoint methodology resulted in en instrument setpoint that more accurately reflects total instmmentation loop accuracy, as well as that of test equipment and setpoint drill between surveillances. The methodology ensures the a!Tected instrumentation remains capable of mitigating design basis events as described in the safety analyses and that the results and consequences

)

described in the safety analyses remain bounding. The function of the APRM included in this analysis l

is not an initiator of any analyzed event. This change provides more stringent requirements for operation of the facility. The more stringent requirement of a lower power level for the startup scram function does not result in operation that will increase the severity of an analyzed event and does not alter assumptions relative to mitigation of an accident or transient event. This setpoint change will not caese any malfunctions or failures of equipment important to safety other than those already postulated and analyzed in the SAR. The setpoint change is more conservative and as a result there are no equipment limits impacted. Revision of the current plant setpoint to a more conservative value will cause the associated safety / trip function to occur earlier. This change does not introduce a new mode of plant operation and does not involve physical removal and replacement of any component, system, or structure in the plant.

Plant equipment will not be operated in a manner different than previously operated with the exception that the setpoint will be conservatively changed. Since operational methods remain unchanged and the 140-1 l

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operating parameters have been evaluated to maintain the plant within existing design basis criteria, no different type of accident or malfunction of equipment important to safety is created. Revision of the setpoint to a mon: restrictive value will cause the associated safety function to occur earlier and will not reduce the current margin ofsafety.

Setnoint Chance Recuest 98-09 (USQE 1998-0017)

TITLE: Setpoint Change for Average Power Range Monitor (APRM) Upscale (Startup) Rod Block, NM-NAM-AR2, 3, 4, 7, 8, 9 DESCRIPTION: This more restrictive setpoint change was the result of Nuclear Engineering Department Calculation 98 024 which was developed to calculate an allowable value in consideration of the transition to Improved Technical Specifications This change is also based on an increase in sun eillance inten al to six months (+25% grace period). The calculation was developed in accordance with the General Electric Instmment Setpoint Methodology and CNS Procedure 3.26.3. Analytical limits used in the calculation l i

are derived from the limiting values of the process parameters obtained from the safety analysis or other appropriate documents. This Setpoint Change Request changed the setpoint for the APRM Upscale (Startup) Rod Block from 11.1%f to 10.0%t . This instrument setpoint provides an input signal to the Control Rod Block System for rod block trip on an upscale (startup) condition from the APRM.

SAFETY EVALUATION: Application of the CNS instrumentation setpoint methodology resulter; m an instmment setpoint that more accurately reflects total instrumentation loop accuracy, as well as tha: of test equipment and setpoint drill between surveillances. The methodology ensures the affected instrumentation remains capable of mitigating design basis events as described in the safety analyses and that the results and consequences described in the safety analyses remain bounding. The function of the APRM included in this analysis is not an initiator of any analyzed event. This change provides more stringent requirements for operation of the facility. The more stringent requirement of a lower power level for the startup rod block Iunction does not result in operation that will increase the severity of an analyzed event and does not alter assumptions relative to mitigation of an accident or transient event. The purpose of this rod block function is to avoid conditions that would require Reactor Protection System action if allowed to proceed.

This setpoint change will not cause any malfunctions or failures of equipment important to safety other than those already postulated and analyzed in the SAR. The setpoint change is more consen ative and as a result there are no equipment limits impacted. Revision of the current plant setpoint to a more conservative value will cause the associated safety / trip function to occur earlier. This change does not introduce a new mode of plant operation and does not involve physical removal and replacement of any component, system, or structure in the plant. Plant equipment will not be operated in a manner different than previously operated with the exception that the setpoint will be conservatively changed. Since l operational methods remain unchanged and the operating parameters have been evaluated to maintain I

the plant within existing design basis criteria, no different type of accident or malfunction of equipment i important to safety is created. Revision of the setpoint to a more restrictive value will cause the

) associated safety function to occur earlier and will not reduce the current margin of safety.

I l Setnoint Chance Reauest 98-69 Sgtnoint Chance Recuest 98-70 TITLE: Setpoint Change Requests (SCRs) for Reactor Vessel Level Indicating Switches, NBI-LITS-73 A and NBl-LITS-73B DESCRIPTION: These more restrictive setpoint changes were the result of a revision to Nuclear Engineering Department Calculation 92-050N to calculate allowable values in consideration of the transition to improved Technical Specifications (ITS) in conjunction with an increase in surveillance intenal to 18 months l (+25% grace period). The calculation was revised in accordance with the General Electric Instrument Setpoint Methodology and CNS Procedure 3.26.3. Analytical limits used in the calculation are denved from the limiting value of the process parameter obtained from the safety analysis or other appropriate

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documents. SCR 98-69 changed the setpoint for NBI-LITS-73A from -24.48"1 to -21.10"1. SCR 98-70 changed the setpoint for NBI-LITS-73B from -24.70"1 to -21.30"1. These instrument setpoints provide Low Reactor Water Level (Level 0) permissive signal to the Residual IIcat Removal System to allow I containment spray.

SAFETY EVALUATION: Application of the CNS instrumentation setpoint methodology resulted in instrument setpoints that more accurately reflect total instrumentation loop accuracy, as well as that of test equipment and setpoint drift between surveillances. The methodology ensures the afTected instrumentation remains capable of mitigating design basis events as described in the safety analyses and that the results and consequences described in the safety analyses remain bounding. The subject instrumentation is not an initiator of any analyzed event. This change provides more stringent requirements for operation of the facility. The more stringent requirement of a higher level does not result in operation that will increase the severity of an analyzed event and does not alter assumptions relative to mitigation of an accident or transient event.

These setpoint changes will not cause any malfunctions or failures of equipment important to safety other than those already postulated and analyzed in the SAR. The setpoint changes are more consen ative and as a result there are no equipment limits impacted. Revision of the Containment Spray Reactor Level permissive setpoint to a more conservative value will cause the loss of permissive at a higher water level in the transient and/or accident scenario. Raising the permissive setpoint ensures that the core is flooded (2/3 core height) before allowing Low Pressure Coolant Injection flow to be diverted to provide containment spray. These changes do not introduce a new mode of plant operation and do not involve physical removal and replacement of any component, system, or structure in the plant. Plant equipment j

will not be operated in a manner different than previously operated with the exception that the setpoints 1 will be conservatively changed. Since operational methods remain unchanged and the operating parameters have been evaluated to maintain the plant within existing design basis criteria, no different type of accident or malfunction of equipment important to safety is created. Revision of the setpoints to more restrictive values will cause the associated safety function to occur earlier and will not reduce the current margin of safety.

Setnoint Chance Reauest 98-81 (USQE 1998-0015)

TITLE: Setpoint Change for Rod Block Monitor (RBM) Low Power Range Downscale Trip, NM-NAM-ARS,

-AR6 DESCRIPTION: This more restrictive setpoint change was the result of Nuclear Engineering Department Calculation 98-024 which was developed to calculate an allowable value in consideration of the transition to Improved Technical Specifications. This change is also based on an increased surveillance inten al of siunonths (+25% grace period). The calculation was developed in accordance with the General Electric j instmment Setpoint Methodology and CNS Procedure 3.26.3. Analyticallimits used in the calculation are derived from the limiting values of the process parameters obtained from the safety analysis or other appropriate documents. This Setpoint Change Request changed the setpoint for the RBM Downscale Trip from 92.2W to 94.0W. This instrument setpoint provides an input signal to the Control Rod Block System for rod block downscale trip.

SAFETY EVALUATION: Application of the CNS instrumentation setpoint methodology resulted in an instrument setpoint that more accurately reflects totalinstrumentation loop accuracy, as well as that of test equipment and setpoint drift between surveillances. The methodology ensures the affected instrumentation remains capable of mitigating design basis events as described in the safety analyses and that the results and consequences desenbed in the safety analyses remain bounding. The instrumentation included in this analysis is not an initiator of any analyzed event. This change provides more stringent requirements for operation of the facility. The more stringent requirements do not result in operation that will increase the severity of an analyzed event and does not alter assumptions relative to mitigation of an accident or transient event.

This setpoint change will not cause any malfunctions or failures of equipment important to safety other than those already postulated and analyzed in the SAR. The setpoint change is more consen ative and as a result there are no equipment limits impacted. Revision of the current plant setpoint for the RBM

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Downscale Trip to a more conservative value will cause the associated safety / trip function to occur i carlier. This change does not introduce a new mode of plant operation and does not involve physical  !

removal and replacement of any component, system, or structure in the plant. Plant equipment will not be operated in a manner different than previously operated with the exception that the setpoint will be conservatively changed. Since operational methods remain unchanged and the operating parameters have been evaluated to maintain the plant within existing design basis criteria, no different type of accident or malfunction of equipment important to safety is created. Revision of the setpoint to a more restrictive value will cause the associated safety function to occur earlier and vill not reduce the current margin of safety.

Protective Relav Setnoint Chance 97-001-EE-REL l

TITLE: Setpoint Change for Time Overcurrent Relay for Breaker CSP 1B DESCRIPTION: The time overcurrent relay for breaker CSPIB was replaced with a different model; the old relay is no longer available and has been superseded by a new model. The overcurrent setpoint for the new relay is 3.0 amps while the setpoint for the relay it replaced was 3.1 amps. The function of the relay is to protect Core Spray Pump 1 B motor by providing overcurrent protection.

SAFETY EVALUATION: The specific long time overcurrent trip setpoint for a Core Spray pump motor is not included in accident I and transient analysis and is not credited as an accident initiator. The degree of protection and i

coordination remains unchanged. The relay provides a protective function in response to a fault or l degraded condition and serves to limit its effect on other equipment, that function remains unchanged.

The change in setpoint is a reduction of about 3% in the long time overcurrent trip setpoint which still provides adequate margin above the normal operating current value to prevent an invalid trip. This change does not change the design or function of any accident mitigating equipment. The setting of this relay is not credited in the basis of any Technical Specification. The protective function provided by this I

relay remains unchanged and, therefore, does not represent a reduction in any margin of safety.

, Problem Identification Report 2-25884 i

! TITLE: Disabling Ronan Point 4314 Because of Nuisance Alarms 1

DESCRIPTION: The Ronan point 4314 on panelAvindow A-1/G-1 was disabled because electrical noise in the wiring from MUX-14 was causing numerous alarms resulting in a disturbance or nuisance in the control room.

Compensatory measures were established which included notification of the Shill Supenisor and visually verifying both power supplies are on twice a shift SAFETY EVALUATION: Disabling of this Ronan point has no impact on the assumptions of the Nuclear Safety Operational Analysis for accident / transient initiators or precursors. It does not increase the consequences of a plant event since it has no effect/ impact on the boundaries and barriers to fission produce release. The annunciator does not cause or affect automatic initiation of safety systems. No annunciator failure can impact the performance of any safety related equipment function since their power feeds are electrically isolated from all safety related equipment. Disabling of Ronan point 4314 does not introduce any new accident initiators or precursors and does not introduce any new failure modes. It does not impact the indication or sensing of any parameters (such as water level, water temperature, etc.) that are bounded by the containment Technical Specifications or Administrative Limits. Therefore, there is no reduction l in the margin of safety.

' Disabline of Annunciator TITLE: Disabling IIPCI Turbine Exhaust Drip Leg Alarm 9-3-2/F-2 DESCRIPTION: The annunciator for the 'llPCI Turbine Exhaust Drip Leg Level fligh' was disabled because it was alarming approximately every 30 minutes and causing a distraction to the operating crew. Supply steam

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is leaking through the iIPCI Steam Supply Valve and is condensing in the HPCI Turbine Exhaust Drip 1 Leg. This condition was previously identified by a Problem Identification Report and compensatory actions have been taken. Disabling the annunciator does not disable the automatic features of the fligh

!cvel drain valve nor does it disable the annunciator and drain valves for the I figh-1ligh level.

i SAFETY EVALUATION: This change removes an annunciator alarm only. It does not remove any active automatic features associated with the Exhaust Drip Leg Drain. The annunciator response procedure does not require any operator action for this alarm. Disabling the annunciator will not change the consequences of a previously evaluated event. IIPCI-SOV-SSV64 will still function to maintain the exhaust line free of condensate. Panel 9-3-2 has other diverse indications that IIPCI-SOV-SV64 is operating correctly.

Since disabling the annunciator (kies not change any manual or automatic actions, it does not increase tl e probability of a malfunction of the Drip Leg Drain system. Since Operations has other diverse indications that the Drip Leg Drain system is operating correctly, it is not possible for this actisity to create a ditTerent type of event or malfunction. There is no required operator action until the Drip Leg High-Iligh ,

annunciates. Since IIPCI-SOV-SSV64 still operates as designed, the margin of safety is not reduced. j Problem Identification Renort 2-25011 TITLE: Disabling Annunciator R-2/B-7 DESCRIPTION: Annunciator R-2/B-7, which is the reactor recirculation motor generator (RRMG) A/B no air flow alarm, was disabled because high wind conditions were causing numerous alarms and creating a nuisance.

Compensatory measures were established which included notification of the Shin Supersisor, monitoring RRMG-TR-66 winding temperatun s hourly, and use of another annunciator to alert the Operator to rising temperatures. With multiple indications in place for monitoring temperature, as an indicator that adequate air flow was reaching the windings on the RRMGs, there was no degradation of the reactor recirculation l system.

SAFETY EVALUATION: Disabling the annunciator does not in itself cause an increase in the probability of occurrence of a plant event because it has no impact on the assumptions of the Nuclear Safety Operational Analysis for accident / transient initiators or precursors. It does not increase the consequences of a plant event because it has no effect/ impact on the boundaries and barriers to fission product release. There is no impact on important to safety equipment reliability. The annunciator does not cause automatic initiation of safety systems and the failure of the safety systems has been evaluated for probability in the USAR. The annunciator has no effect on the operation of the reactor recirculation system. Disabling of the annunciator does not introduce any new accident initiators or precursors and does not introduce any new failure modes. It does not cause a change to any parameters (such as water level, water temperature, etc.)

that are bounded by the contamment Technical Specifications or Administrative Limits. Therefore, there is no reduction in the margin of safety.

Condition Renort (CR) 98-0849 and Generic Letter 98-04 (USQE 1998-0070)

TITLE: Unqualified / Degraded Coatings in Containment DESCRIPTION: While preparing the response to Generic Letter 98-04, it was determined that a change to the facility as described in the USAR had occurred without conducting an unreviewed safety question evaluation. CR 98-0849 was written to document the following conditions. A containment coatings walkdown during RE16 had identified the application of unqualified coatings since original construction. In addition, peeling paint was removed inside the drywell. Therefore, the interior surface of containment no longer matched the description in the USAR. This safety evaluation was performed to determine whether unqualified and degraded coatings within the containment involved an unreviewed safety question.

SAFETY EVALUATION: Paint and coatings are not considered part of the containment pressure boundary under Code rules because they do not contribute to the pressure retaining function or leak tightness of the containment. The

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l containment shell is a passive component n'ot subject to single active failure criteria. Failure of the containment shell is not an initiator of design basis accidents or transients previously evaluated in the {

SAR. The effects of paint debris were considered in the design of the Emergency Core Cooling System l' (ECCS) suction strainers. The strainers are passive components not subject to single active failure criteria. Strainer failures are not initiators of design basis accidents or transients. Introduction of paint debris into systems inside containment and accumulation of paint debris on exposed moving parts of active safety-related components are not considenxl credible because no transport mechanism exists. The -

radiological consequences of the postulated design basis Loss of Coolant Accident (LOCA) or other pipe breaks remain the same as previously evaluated in the SAR. Unqualified or degraded coatings do not affect primary containment integrity and leak tightness characteristics. The replacement ECCS suction strainers installed under Modification Package 96-132 improved the efliciency of ECCS during postulated design basis LOCA conditions; the new strainers reduce head losses and are capable of accommodating an increase in the allowable debris loading compared to the previous design. These features ensure the consequences of an accident are not increased because ECCS is capable of fulfilling its safety function. .

This change does not intmduce any new failure modes. Corrosion allowance was included in the design I of the containment shell. The capability of containment to meet its safety function including confining l the reactor coolant that would be released during a postulated pipe rupture and limiting the release of fission products to the environment are not affected by this change. The design corrosion allowance ensures a malfunction of a different type is not created. The Containment Systems Technical Specifications and associated bases define requirements for isolation valves, penetrations, leakage rate testing, and visual exanunation to demonstrate the leak-tight characteristics and structural integrity of the primary containment. The unqualified and degraded coatings will not alTect the leakage characteristics or stmetural integrity ofprimary containment. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

USOE 1998-0036 & Revision 1 TITI.E: Reactor Pressure Vessel (RPV) Shell Weld Inspection Using the GERIS 2000 Manipulator .

DESCRIPTION: The purpose of this safety evaluation was to evaluate the inspection of the IU'V Shell Welds using the GERIS 2000 ID Manipulator during Refueling Outage 18. The mechanical computer controlled manipulator system, coupled with a General Electric Nuclear Energy designed Ultrasonic Test (UT) system, has the capability to conduct UT inspection from the inner surface (ID) of the RPV above the core shroud support plate. This evaluation addressed all activities related to the installation / removal and use of this equipment. Revision 1 to this USQE removed a statement that specifically excluded movement of the GERIS 2000 equipment over the equipment storage pit.

SAFETY EVALUATION: Damage to fuel or damage to a safety system were considered in the design and controls for the use of the GERIS 2000 equipment and are not considered credible events. Use of the GERIS can be performed concurrently with fuel handhng within the RPV. The initiators, assumed failures, and sequences of events for all other accidents in the SAR are not afTected by the use of the GERIS 2000. The installation, use, and removal of this equipment does not affect any structure, system, or component assumed to operate during an accident, or to mitigate the consequences of any accident in the SAR. The components for lifting / moving / lowering the GERIS 2000 meet the NUREG-0612 safety factor criteria. All ECCS, shutdown cooling, and other equipment required to be operable during a refueling outage will not be affected and will be available to mitigate any potential accident. The UT examinations using the GERIS 2000 cannot affect a radiological consequence of any accident in the SAR. All movement, rigging, and lifting will be performed in accordance with approved plant procedures. The lifting and lowering of the GERIS 2000 is well within the design load of the crane, and the probability of a crane, cable, or connector failun:is not increased. The weight of the GERIS 2000 is within the load capacities of the reactor vessel and shroud and will not cause any degradation to these components. The operation of the GERIS 2000 will not degrade the quality of any system or component of the reactor coolant pressure boundary. In addition, there is no component of a safe shutdown system located on the refueling floor with which the GERIS 2000 could interact. The installation, use, Lnd removal of the GERIS 2000 involves no process or interaction with a fission product barrier or a radioactive material control function that is used to

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mitigate the consequences of any malfunction in the SAR. The GERIS 2000 is used during a refueling outage, its associated activities are not significantly different than other refueling activities, and the I installation, use, and removal of the GERIS does not involve any potential new fission product release l path or create any new failure mode. No adverse material interactions will result from the tne of the GERIS 2000. Iost parts evaluations show there is no safety concern with respect to flow bk)ckage to the fuel bundles, interference with the safety function of the control rod operation, fuel damage, or adverse chemical reaction with other reactor materials. Operation of the GERIS does not create a new failure mode, nor does its operation place CNS in a condition of operating outside any safety limit. There is no Technical Specification basis associated with inservice examination of the RPV. Those specifications associated with inservice inspection are included in the Technical Requirements Manual. These I requirements will be met by performing the beltline inspection using the GERIS 2000 manipulator.

Technical Specification requirements for when the plant is shutdown and the associated bases are not afrected by the use of the GERIS. Therefore, the margin of safety will not be reduced. l USOE 1998-0054 l

TITLE: Evaluation of Temporary Laydown Arrangement for Reactor Pressure Vessel (RPV) Beltline Inspection DESCRIPTION: Assembly requirements of the GERIS 2000 inspection system used to perform RPV beltline inspections during Refueling Outage 18 resulted in the need to relocate refueling floor laydown equipment to accommodate assembly of the GERIS system. Temporary changes were made to Procedure 7.4.1 to incorporate these temporary laydown equipment relocations on the refueling floor. Engineering Evaluation 1998-0059 concluded that floor loading allowables were not exceeded and this equipment rearrangement was acceptable.

SAFETY EVALUATION: Equipment evaluated under this USQE is cunently on the refueling floor and relocation of this equipment i does not damage fuel, the RPV, or a system, structure, or component (SSC). This equipment is not considered an initiator or contributor for an accident evaluated in the SAR. Relocation of this equipment ckles not affect any SSC assumed to operate during an accident. Equipment important to safety will not be affected. An Engineering Evaluation determined that relocation of this equipment was acceptable fbr the refueling floor from a loading standpoint. No adverse malfunctions of equipment are introduced by this temporary change. The laydown equipment does not alTect the Emergency Core Cooling Systems, Reactor Protection System, or the containment isolation functions. Relocation of the laydown equipment does not interact with the fuel or create a new release path. It cannot create the possibility of a malfunction of a different type than previously evaluated in the SAR. There are no Technical Specifications associated with the laydown equipment or basis for same. Therefore there is no reduction in the margin of safety as dermed in the basis for any Technical Specification.

USOE 1998-0099 TITLE: CNS Plant-Specific Technical Guidelines and Severe Accident Technical Guidelines (PSTG/SATGs)

DESCRIPTION: The PSTG/SATGs implement severe accident guidance at CNS. This Safety Evaluation was performed to ensure that the changes due to the replacement of the CNS Emergency Procedure Guideline (EPG),

Revision 3, with the CNS PSTG/SATGs are consistent with the basis of Revision 4 of the Boiling Water Reactor Owners Group (BWROG) EPGs, and do not conflict with the licensed design basis or operation of CNS. It was determined that the CNS PSTG/SATGs conform to the intent of the guidelines of Revision 4 of the BWROG EPGs and the applicable NRC Safety Evaluation Report dated September 12, 1988, and do not conflict with the licensed design basis or operation of CNS. The revised Emergency Operating Procedures (EOPs) and Severe Accident Guidelines (SAGS) will provide for improved operation of CNS during transients, accidents, and emergency conditions, and are designed to prevent and mitigate severe accidents.

SAFETY EVALUATION: Since the implementation of the CNS PSTG/SATGs will not modify the design basis of the plant and they are only used after plant conditions during an operational transient or accident have reached the EOP

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entry conditions, it is concluded that they cannot increase the probability of occurrence of any event analyzed in the SAR. Although the EOPs address events within the design and licen. sing basis of the plant, the probability of occurrence of these events is not afTected by the EOPs. Based upon evaluation of the effects of the subject changes on accidents and malfunctions of equipment important to safety previously evaluated in the SAR, it is concluded that the radiological consequences of accidents and equipment malfunctions are not increased. The PSTG/SATGs may require that some systems and equipment are operated in ways which may affect their future reliability or performance; however, the PSTG/SATGs are intended for use during conditions beyond those analyzed in the licensed design basis.

It is concluded that the PSTG/SATGs do not increase the probability of a malfunction of equipment important to safety evaluated in the SAR because they are only used after the event has commenced. As the PSTG/SATGs do not modify the design basis of the plant, the possibility of accidents or equipment malfunctions of a difTerent type than previously evaluated in the SAR cannot be created. Although the PSTG/SATGs contain operator actions for plant conditions not analyzed in the SAR, to reach these conditions multiple equipment failures or operator errors must be assumed. 'Ihe CNS licensing basis does not require that accidents requiring the assumption of multiple equipment failures and operator errors be considered. For plant conditions which exceed the licensed design basis, the question of a reduction in the margin ofsafety as dermed in the basis of the Technical Specifications is not applicable.

A review of the SAR found that no operator actions in the EOPs and SAGS conflicted with any assumed actions in any analyzed event or design basis accident.

Oncratine License Chance Reauest 97-008 TITLE: Technical Specification Bases Changes DESCRIPTION: Technical Specification Bases 3.5.0, " Maintenance of Filled Discharge Pipe," was revised to delete the sentence which states that if a water hammer were to occur at the tune at which the system was required, the system would still perform its design functions. This portion of Bases 3.5.G was deleted because the CNS Core Spray (CS), Low Pressure Coolant Injection (LPCI), Iligh Pressure Coolant Injection (1IPCI),

and Reactor Core Isolation Cooling (RCIC) systems are not specifically designed for dynamic loading with the exception of seismic events.

SAFETY EVALUATION: There are no accidents or transients evaluated in the USAR that are affected by this activity. It will not L cause the CS, LPCI, IIPCI, and RCIC systems to experience water hammer piping loads or cause the system to be operated outside ofits design limits. The change is proposed because the CS, LPCI,1IPCI, and RCIC system piping is not designed for water hammer loads. Dynamic water hammer loads are not included in the design analysis of CS, LPCI,IIPCI, and RCIC per USAR Appendix A. As such, removal of this statement from Bases 3.5.G does not increase the probability of occurrence of a plant event previously evaluated in the USAR. That portion of Bases 3.5.G which discusses the basis for actions to minimize the potential for a water hammer event (keep fill requirement) remains unchanged. This change does not alter any assumptions previously made in evaluating the radiological consequences of a plant event described in the USAR and will not afTect any fission product barriers. The probability of occurrence of a previously evaluated equipment malfunction is not increased because no new loads are being added to the system piping, the activity does not delete or modify system / equipment features, and the pressure maintenance support system is not being downgraded. Radiological consequences are not i increased since the manner in which the CS, LPCI, IIPCI, and RCIC systems are operated or maintained in standby is not changed. No new failure modes are created. Since the CS, LPCI, IIPCI, and RCIC piping systems ' vere not designed to withstand water hammer loads and the Technical Specification Bases has been incorrect since it was originally issued, there is no reduction in the margin of safety. i 1

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l Operatine License Chance Reauest 98-007 (USQE 1998-0052)

TITLE: Revision to OITsite Dose Assessment Manual (ODAM) Bases DESCRIPTION: This revision clarified ODAM Bases B 3.2.7 such that it is consistent with the wording in ODAM Specification DLCO 3.2.7. The wording in B 3.2.7, Primary Containment Venting and Purging, was being interpreted to imply that Standby Gas Treatment must be used to ventilate primary containment for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown; this interpretation was incorrect. DLCO 3.2.7 states that venting and ]

~

purging of the primary containment shall be through the Standby Gas Treatment System. However, as described in Notes I and 2 of DLCO 3.2.7, this specification does not apply when normd ventilation is established and while performing inerting during a startup following a shutdown of greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SAFETY L

EVALUATION: This change only serves to make the ODAM Bases wording consistent with the ODAM Specification. j No accident initiators or assumptions are affected by this ODAM Bases change and no changes are made to any accident mitigation assumptions. This revision does not affect the safety function, operation, or  !

design ofplant equipment or systems assumed to function in any SAR accident analysis. Once primary l containment purging / venting is complete, initiation of primary containment normal ventilation in MODE 4 or 5 has no impact on any accident analysis. There are no increased challenges to safety systems such that their performance is degraded below the design basis nor is there an increased probability of a ,

malfunction of equipment iroportant to safety. No new accident initiators, failure modes, or equipment operating modes are introduced by this change. The margin of safety as defined in the basis for any i Technical Specification is not reduced by this change. Design parameters / assumptions for systems and I components are unaffected. No Technical Specification, Technical Requirements Manual, or ODAM requirements have been changed. i Oncratine 1 icense Chance Renuest 98-002 (USQE 1998-0103)

{

TITLE: Technical Specification Bases Change h

DESCRIPTION: Tnis Operating License Change Request revised the Technical Specification Bases for Surveillance l

Requirements 3.3.1.1.5 and 3.3.1.1.6 concerning Source Range Monitor (SRM) and Intermediate Range j Monitor (IRM) overlap. The Bases previously stated that overlap between the SRMs and IRMs existed j when " prior to withdrawing the SRMs from the fully inserted position, IRMs are above mid-scale on  !

range 1 before SRMs have reached the upscale rod block." This was revised to indicate that there is i suflicient overlap between the IRMs and SRMs when the IRMs are on scale, and therefore operable prior j to withdrawal of the SRMs from the ful:y inserted position. During startup from RE18, mid-scale on l range 1 of the IRMs could not be achieved prior to the SRMs reaching the upscale rod block. The i function of the IRM/SRM overlap is to ensure that reactor power will not be increased into a neutron flux l

region without adequate indication. This change continues to ensure that this function is fulfilled. A I change was also required to Procedure 2.1.1, Startup Procedure, to conform with the Technical i Specification Bases change.

SAFETY l

EVALUATION: There are no changes to any of the accident initiators associated with this Bases change. The limiting accident ofconcern with the reactor at very low power is the Control Rod Drop event. Since this Bases change does not involve any changes to fuel, rod patterns, or rod withdrawal sequence, the probability l

of occurrence of the Control Rod Drop accident remains unchanged. Accident cc'nsequences are not affected. There are no physical changes to any equipment important to safety and the associated l procedural change does not change the manner in which the equipment functions. The IRM/SRM overlap is maintained so that the neutron flux is continuously monitored; therefore, no new types of accidents or equipment malfunctions are introduced and the margin of safety is preserved. Since the Bases change preserves the operability of both the IRMs and the SRMs during the overlap phase, adequate indication of neutron flux will be maintained.

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i Operatine License Chance Reauest 99-002

{

(USQE 1999-0007) I l

l TITLE: Revision to Offsite Dose Assessment Manual (ODAM) Section D 3.3.2, Gaseous Emuent Monitoring 1

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DESCRIPTION: This change revised Condition D 3.3.2.1 Required Actions and Completion Times such that building {

ventilation does not have to be secured whenever the building efiluent particulates and iodines cannot be I sampled via installed sampling systems. Securing building ventilation will not necessarily terminate building release. Unmonitored, unfiltered, ground level releases could occur once building ambient pressure becomes positive with respect to its surroundings. During the relocation of the radiological effluents section of Custom Technical Specifications to the ODAM as part of the conversion to improved Technical Specifications, Required Actions and Completion Times were added which actually defeated the ultimate intent of the specification. The Required Action stated "1.1 Continuously collect samples with auxiliary sampling equipment as required in Table D3.2.3-1" with a Completion Time of "Immediately." This Operating License Change Request revised the Completion Time for this Action to"4 Hours" and an altemative action was added which states "QR I.2.1 If auxiliary sampling equipment cannot be established within the specified completion time, initiate a Problem Identification Report to evaluate particulate and iodine efiluent releases, MQ l.2.2 Report this event in the Annual Radioactive Emuent Release Report." This change is consistent with License Amendment 89 which provides the basis for the ODAM.

SAFETY EVALUATION: This is an admmistrative change that does not affect the design, function, system interfaces, or operating parameters of the radiobgical emuent monitoring system or any equipment important to safety. It does not affect any assumption for accidents evaluated in the SAR or cause previously analyzed accidents to shill to a higher frcquency class. This change does not degrade or prevent mitigative actions described or assumed for accidents discussed in the SAR, and does not impact fission product barriers. The particulate and iodine filters are not designed to terminate building efiluent releases or play a direct role in mitigating the radiological consequences of an accident. This change actually reduces the radiological detriment to personnel and the environment by ensuring that building releases take place through designed release points which in some cases provide filtering capabilities. This change will not degrade the perfonnance of safety systems assumed to function in the SAR accident analysis or increase challenges to safety systems such that system perfonnance is degraded below design basis. It does not add or modify any equipment and does not create a new system lineup. No new accident initiators or equipment faihtre modes are introduced by this change. It does not all'ect the margins of safety evaluated against Safety Limits, Limiting Safety System Settings, Limiting Conditions for Operation, and design parameters for systems, structures, and components.

Technical Reauirements Manual (TRM) Chance Reauests 1998-002 and 1999-001 (USQE 1998-0098)

TITLE: Revision of Sections T3.3.3 and B3.

3.3 DESCRIPTION

TRM Section T3.3.3,"Non-Type A, Non-Category 1 Post Accident Monitoring (PAM) Instnunentation",

did not adequately reflect the surveillance requirements or which instruments were required for the given modes of applicability. The applicable instrument channels are required to be operable per other Technical Specifications and TRM requirements. TRM Section T3.3.3 was revised to identify existing l

channel calibration surveillances. TSR 3.3.3.2 was revised to replicate SR 3.3.1.1.2 for periodic calibration of the Average Power Range Monitors (APRMs). TSR 3.3.3.6 was revised to replicate SR 3.3.1.1.12 and SR 3.3.1.2.7 for calibration of the Intermediate Range Monitors (IRMs) and Source Range Monitors (SRMs). Function 7 of Table T3.3.3-1 was revised to include SRM, IRM, and APRM with applicable mode and surveillance requirements. TRM Bases B3.3.3 was revised to indicate that Function 7 applies to the SRMs, IRMs, and APRMs and changed the range to 3 cps - 100% power. Associated Surveillance Procedures were also revised to reflect the additional surveillance applicabilities. In addition, the Component Identification Code Cross Reference Table contained in the TRM was updated accordingly, l -149

SAFETY EVALUATION: The subject instruments provide the function of monitoring reactor power following an accident or transient, and are not accident initiators. The instruments ofinterest are currently being calibrated at the Technical Specification frequency proposed by this change. The addition of the information to TRM Section T3.3.3 provides a link to identify neutron monitoring subsystems with their applicable mode and surveillance requirement. The probability or consequences ofinstrument malfunction have not changed.

This activity provides additional information for adherence to a TRM Limiting Condition for Operation.

No new types of accidents or equipment malfunctions are introduced. The changes to the TRM currently exist as Technical Specification requirements for the same instruments; therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

Technical Recuirements Manual (TRM) Chance Reauest 1999-002 (USQE 1999-0026)

TITLE: Revision of TRM Table T3.3.1-1 DESCRIPTION: TRM Table T3.3.1-1, " Control Rod Block Instrumentation", was revised to change the calibration frequency of the Reactor Recirculation Loop Flow Transmitters to 18 months. The flow transmitter calibration frequency had been inappropriately changed during the implementation ofImproved Tecimical Specifications (ITS); this change corrects that error. Revising the calibration frequency to once every 18 months will permit the calibrations to be performed during refueling outages. The configuration of the flow transmitters creates a risk of a reactor scram if the calibration is performed on line.

SAFETY EVALUATION: Changing the calibration for the flow transmitters to 18 months may change the drill experienced by the transmitters. Ilowever, this is accounted for in approved Nuclear Engineering Department Calculation 98-024. In addition, the 18 month calibration of the flow transmitters was thu standard calibration frequency prior to the implementation ofITS. Changing the calibration frequency does not change any of the precursors assumed in the accident analysis or increase the probability of an accident evaluated in the SAR. The transmitters provide an input to the Average Power Range Monitor (APRM) flow biased upscale alarm and trip. The APRM flow biased alarm is not specifically credited in the CNS Safety Analysis. The operation of the transmitters is not associated with accident initiators or with accident mitigation as evaluated in the SAR. The APRM flow biased alarm and trips are not associated with plant equipment important to safety as evaluated in the SAR. Changing the calibration frequency of the flow transmitters does not change any structures, systems, or components and does not change the form, fit, or function of any plant equipment credited in the accide-t analysis. Therefore, no new types of accidents or equipment malfunctions are created. The APRM flow biased rod block and the APRM flow biased upscale alarm are not mentioned in the Technical Specifications. This change in calibration frequency does not reduce the margin of safety as defined in the bases for any Technical Specification.

Technical Reauirements Manual (TRM)

TITLE: Administrative Changes to the TRM DESCRIPTION: This evaluation applies to purely administrative changes made to certain Specifications rek)cated to the TRM as part of the Improved Technical Specification project (ic., reformatting, renumbering, nontechnical rewording, elimination of redundancy, and elimination of excessive detail).

SAFETY EVALUATION: These changes contain no technical changes and in no way impact any event-assumed initial conditions, event initiators, or event mitigators. Therefore, they do not increase the probability of occurrence or consequences of a previously evaluated plant event. These changes do not introduce a new mode of plant operation and involve no physical alteration to any plant equipment. Therefore, they do not increase the probability ofoccurrence or consequences of a malfunction of equipment important to safety and no new or different types of plant events or equipment malfunctions are created. These purely administrative changes do not involve a reduction in the margin of safety.

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Technical Reauirements Manual (TRM)

TrrLE: Change to Fire Protection System Specifications in the TRM DESCRIPTION: Fire Protection System Specifications were relocated from the Technical Specifications to the TRM in the conversion to improved Technical Specifications. This change deletes from the required actions of T 3.11.1 through T 3.11.5 the requirement o prepare and submit a special report to the NRC should the I subject system, structure, or component be inoperable in excess of the required completion time. This

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requirement is replaced with a required action to initiate a Problem Identification Report for evaluation of the degraded condition with a completion time of"immediately." An additional change was made to T 3.11.2 which previously required a report be made to the NRC by both telephone and in writing in the event the fire suppression water system is degraded beyondjust a single fire pump being inoperable. This requirement was changed to report the condition to the SORC Chairman by telephone with a completion time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. These changes meet the NRC guidance provided in Generic Letter 88-12 for Removal of Fire Protection Requirements from Technical Specifications.

SAFETY EVALUATION: The changes are administrative in nature. They are in compliance with regulatory requirements and do not compromise the corrective action program for inoperable fire protection equipment and systems.

Except for the requirement to prepare and submit special reports, the fire protection program is unchanged. Requirements and limitations regarding safe shutdown paths, fire barriers, emergency lights, fire detection systems, water supplies, fire water pumps, spray and sprinkler systems, carbon dioxide l systems, fire hose stations, fire hydrants, and halon systems remain the same. Also unaffected are  !

I personnel requirements, including stafling, training, and drill requirements. Since the requirements and limitations on combustible loading and ignition sources are unchanged, the probability of a previously evaluated fire is not increased. The consequences of a fire are also not increased. Operability and surveillance requirements remain the same. Also unafrected are fire protection equipment operation, maintenance, and emironmental requirements and standards. Consequently, the subject changes cannot increase the probability of occurrence of a malfunction of fire protection equipment. Since the same compensatory and corrective actions will be taken regardless of whether or not a special report is written, the subject changes will not increase the consequences of equipment malfunction. No new types of accidents or equipment malfunctions are created. These changes do not afTect the Technical Specifications or Bases; there is no reduction in the margin of safety as defined in the basis of any l Technical Specification.

Technical Reauirements Manual (TRM)

TITLE: Change to Fire Protection System Functional Tests in the TRM DESCRIPTION: Fire Protection System Specifications were relocated from the Technical Specifications to the TRM in the comersion to Improved Technical Specifications. This change revises Specifications T 3. I 1.2, (Fire Suppression Water System) and T 3.11.3 (Sprinkler Systems), Functional Testing requirements to allow crediting an actual signal for initiation and to eliminate discussion of testing valves which do not exist in the plant.

SAFETY EVALUATION: This change allows a system functional test to include actual or simulated automatic actuation. This does not impose a requirement to create an actual signal, nor does it climinate any restriction on producing an actual signal. It does not affect procedures goveming plant operations or the probability of creating an actual signal. It does not affect the acceptance criteria of the functional test. The removal of testing requirements for nonexistent equipment is administrative and has no effect on any aspect of plant

. operation. Therefore, the probability of occurrence or consequences of a previously evaluated plant event are not increased. The subject changes do not involve a physical modification to the plant or change the requirements for operating, testing, or maintenance of equipment. Therefore, they do not increase the l probability ofoccurrence or consequences of a previously evaluated malfunction of equipment important to safety. The possibility of a new or durerent kind of event or equipment malfunction is not created since the change does not introduce a new mode of plant operation. Operability is adequately demonstrated

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with either an actual signal or a simulated signal since the system cannot discriminate between actual or simulated signals. Therefore, the changes do not involve a reduction in the margin of safety.

Technical Reauirements Manual (TRM)

TITLE: Change to Diesel Engine Driven Fire Pump Battery Check in the TRM 1

DESCRIPTION: The requirements for performing a diesel engine driven fire pump battery 18 month deterioration check I (TSR 3.11.2.9) were revised. The requirement previously stated: " Verify the batteries, cell plates, and battery racks for the diesel engine driven fire pump show no physical damage or abnormal deterioration."

This was changed to: " Verify the battery racks for the diesel engine driven fire pump show no visual  !

indication of physical damage or abnormal deterioration, and: (a) Verify the batteries and cell plates show no visual indication of physical damage or abnormal deterioration; or (b) Replace the batteries." The batteries in use at CNS are opaque type batteries; therefore, an adequate internal visual check is not possible. Thus, the option of battery replacement is prosided.

SAFETY EVALUATION: This change adds an allowance to replace the subject batteries versus checking the degradation of the l

batteries. This is a conservative action in that the intent of the specification is to determine if battery l maintenance or replacement is required. Replacing the batteries at an 1 fs month or shorter frequency in i no way affects the inidal conditions of any analyzed event nor does it affect the success path in the mitigation of any analyzed event. This change can only enhance equipment reliability and does not I introduce a new mode of plant operation. Therefore, there is no increase in the probability of occurrence .

or consequences of a malfunction of equipment important to safety and no new or different kinds of plant i events or equipmmt malfunctions are created. The diesel driven fire protection pump is not in Tecimical Specifications. Changes to requirements for these batteries do not involve a reduction in the margin of safety.

Technical Recuirements Manual (TRM)

TITLE: Change to Fire Suppression Water System Specification in TRM DESCRIPTION: The Fire Suppression Water System specification was clarified in its translation to the TRM. Ambiguity regarding the allowance for a 7 day restoration time for a single inoperable fire pump was resolved.

Allowing a 7 day Completion Time with a single pump inoperable is appropriate. In this condition, the remaining operable pump is adequate to perform the fire suppression function. The 7 day Completion Time is based on the availability of the other pump and the low probability of a fire occurring with the failure of the second pump. This position is further supported by the Nebraska Public Power District Cooper Nuclear Station Appendix A to Branch Technical Position 9.5-1 Commitment Summary for Section E.2.(c), titled " Fire Pumps."

SAFETY EVALUATION: This clarification of a Required Action, providing a reasonable restoration time while retaining function consistent with the single failure assumptions of Technical Specifications, will not affect the capability of the components or systems to perform their intended functions. The fire suppression water system is not assumed in the mitigation of any analyzed accident. There is no change in the capability of the system to perform its intended function. Therefore, the consequences of a malfunction of equipment important to safety are not increased. This change does not introduce a new mode of plant operation and does not involve physical modification to the plant. Therefore, no new or different kinds of plant events or equipment malfunctions are introduced. There is no margin of safety associated with establishing the backup fire suppression system and associated reporting requirements. Therefore, this change does not involve a reduction in a margin of safety.

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Technical Reauirements Manual (TRM)

TITLE: More Restrictive Changesin TRM DESCRIPTION: This evaluation applies to those items being relocated from the Technical Specifications to the TRM that are modified to be more restrictive than the verbatim relocation. These changes impose more stringent requirements for operation of the facility.

EVALUATION: Rese more stringent requirements do not result in operation that will increase the probability of uuttatmg '

an analyzed event and do not alter assumptions relative to mitigation of any event. The more restrictive requirements continue to ensure process variables, structures, systems, and components are maintained consistent with the safety analyses and licensing basis. These changes do not introduce a new mode of plant operation and do not involve a physical modification to the plant. Therefore, they do not increase the probability of occurrence or consequences of a malfunction of equipment important to safety. The possibility of a new or different kind of event or equipment malfunction is not created. These more stringent requirements do not involve a reduction in a margin of safety.

Technical Reauirements Manual (TRM) l TITLE: Post Accident Monitoring Instrument Identification Number Relocation l

DESCRIPTION: This change relocates to the TRM Bases the details relating to the Non-Type A, Non-Category 1 Post Accident Monitoring instrument identification numbers associated with instruments relocated from the Custom Technical Specifications to the TRM. These details are not necessary to ensure the Post Accident Monitonng instrumentation is maintained operable. This change is consistent with the ITS requirements established by NUREG 1433.

SAFETY EVALUATION: This change is admmistrative in nature and will have no effect on the probability of occurrence of a plant event. This change will not affect the capability of the components or systems to perform their intended functions. Furthermore, these details are not assumed in the mitigation of any analyzed accidents. This change does not modify any equipment and does not introduce a new mode of plant operation. There is no margin of safety associated with relocating administrative details associated with the Post Accident i Monitoring instrumentation identification numbers.

Technical Reauirements Manual (TRM)

TITLE: Turbine Building Main Steam Line Leak Detection Instrumentation DESCRIPTION: This change extends the completion time to close the Main Steam Isolation Valves (MSIVs) from 8 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to remain consistent with similar changes made to Improved Technical Specification 3.3.6.1, Pnmary Contamment Lsolation Instrumentation, for these same valves. This provides the necessary time to close the MSIVs in a controlled and orderly manner that is within the capabilities of the unit, assuming the nummum required equipment is operable. The extra time reduces the potential for a unit upset that could challenge safety systems. The MSIV closure time extension applies to the Turbine Building Main Steam High Temperature Detection System.

SAFETY EVALUATION: His change allows for a more controlled shutdown which reduces the chances for a plant transient. This allowance to increase the completion time will not affect the capability of the components or systems to perform their intended functions. Credit is not taken for actuation of these bi-stable temperature switches to mitigate the offsite dose consequences of the Design Basis Main Steam Line Break (MSLB), or a MSLB oflesser magnitude. The Design Basis MSLB is mitigated by MSIV closure due to actuation of time MSL high flow instrumentation. The Turbine Building area temperature switches are provided to detect a MSL leak in the Turbine Building iIcater Bay of a magnitude from 1-10% rated flow. For small breaks of this size, no credit is taken for automatic isolation of the MSIVs by the area temperature switches. Operator action is credited for terminating leaks of this magnitude because the offsite dose

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consequences of this type of event have been calculated to be orders of magnitude less than those of the

[ MSLB Design Basis Accident, since virtually no water carryover (and resulting iodine) is expected to l occur. This change does not introduce a new mode of plant operation and does not involve physical modifications to the plant. It does not create the possibility of any new or difTerent kinds of accidents or malfunctions ofequipment important to safety. There is no margin rf safety defined in the basis for any Technical Specification associated with extending the completion time for closure of the MSIVs due to failure to immediately place one or more inoperable channels in trip. Therefore, this change does not involve a reduction in a margin of safety.

Technical Reauirements Manual (TRM)

TITLE: Reactor Coolant System (RCS) Chemistry l

DESCRIPTION: This change modifies the time line for bringing the reactor to cold shutdown when the RCS chemistry limits are exceeded. Current Technical Specifications allow 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to reach Mode 3, then require the l reactor to be in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> aller reaching Mode 3. Consistent with the philosophy ofImproved Technical Specifications (ITS), this change starts all required actions from time of discovery and, so, is stated: "Be in Mode 3 and Be in Mode 4" with completion times of 12 and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />

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respectively. So stated, the full 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to reach cold shutdown is always allowed.

SAFETY l EVALUATION: This change allows for a more controlled shutdown and cooldown which reduces the chances for a plant l l transient. This allowance to increase the completion time will not afTect the capability of the components I or systems to perform their intended functions. Furthennore, RCS chemistry is not assumed in the mitigation of any analyzed accident. The limits for chloride concentration and conductivity become less stringent in Mode 4 while the pillimit remains virtually unchanged. The major benefit of cold shutdown is to reduce the temperature dependent corrosion rates and provide time for the cleanup system to i reestablish the purity of the reactor coolant. General Electric stress corrosion test data substantiate that stainless steel specimens exposed in this abnonnal environment at high temperatures for thousands of hours resulted in no cracking or failure. This change will only increase the minimum time to reach cold I

shutdown by < 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This small increase is of no significance in relation to these tests. This change does not impact the ability to monitor or limit RCS chemistry. Therefore, it does not increase the probability of a malfunction of equipment important to safety. This change does not introduce a new mode of plant operation and does not involve physical modification to the plant. Therefore, it does not j create the possibility of a new or difTerent kind of accident or equipment malfunction. There is no margin j of safety defined in the basis of any Technical Specification associated with RCS chemistry. Therefore, l this change does not involve a reduction in a margin of safety. (In the conversion to ITS, RCS chemistry l is relocated from the Technical Specifications to the TRM).

Technical Reauirements Manual (TRM)

TITLE: Liquid Nitrogen Minimum Quantity Requirement DESCRIPTION: The liquid nitrogen system functions in support of maintaining the primary ccotinment oxygen concentration limitations that were previously modified in the Technical Specific as. The oxygen concentration limitation governed by LCO 3.6.3.1 required a minor applicabilit, hange to allow j adherence to begin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after thermal power reaches >l 5% Rated Thermal Power (RTP), instead of l Run Mode (approximately 5% RTP). Likewise, TRM Specification T 3.6.1 requires the same l applicability following relocation from the Current Technical Specifications to the TRM. In order to maintain consistency between the two specifications, the TRM applicability 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> start time is changed i from Run Mode to 15% RTP. The justification used in Technical Specification 3.6.3.1 adequately l supports identical changes to T 3.6.1 in regard to this requirement.

SAFETY EVALUATION: This change maintains consistency with the requirement for oxygen concentration in the Technical Specifications. 'Ihis change is administrative for the technical requirements already justified by Technical Specification 3.6.3.1. It does not impact the capability to inert the drywell. Since the content of the 154-l L

nitrogen storage tank is not a mitigator of any plant event, this change cannot impact the consequences of any plant event. This change does not modify any equipment nor does it modify the operation of equipment. Therefore, it does not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated. There is no change in the capability of the system to perform its intended function. Since this change does not introduce a new made of plant operation and does not involve physical modification to the plant, it does not create the possibility of a new or different kind of accident or equipment malfunction. This change does not involve a reduction in a margin of safety since the change is simply an administratively revised presentation, to maintain consistency with the ITS requirements previouslyjustified.

j l Technical Reauirements Manual (TRM)

TITLE: Liquid Nitrogen Minimum Quantity Requirement l DESCRIPTION: The liquid nitrogen system functions in support of maintaining the primary containment oxygen concentration limitations that were previously modified in the Technical Specifications. The oxygen concentration limitation governed by LCO 3.6.3.1 required specification of the Condition A Completion Time which was previously not indicated. Likewise, TRM Specification T 3.6.1 requires the same changes following relocation from Current Technical Specifications. In order to maintain consistent Completion Times between the two specifications, the TRM Condition A Completion Time will be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to restore liquid nitrogen quantities to within the limit prior to requiring a power reduction. During this time, the drywell will normally be inerted and the Standby Nitrogen Injection System and containment venting will be available to backup the Nitrogen Supply subsystem, thus a means to control hydrogen exists. This new action may possibly prevent an unnecessary shutdown and the increased potential for transients associated with a shutdown. The justification used in Technical Specification 3.6.3.1 adequately supports the identical change for T 3.6.1 in regards to this particular requirement.

i SAFETY EVALUATION: The quantity ofliquid nitrogen in the storage tank is not assumed to be an initiator for any analyzed event.

The requirement for the dqwell to be inerted is governed by Technical Specification 3.6.3.1, Primary Containment Oxygen Concentration. The nitrogen storage tank limits provide the capability to satisfy LCO 3.6.3.1, but do not satisfy the criteria of 10CFR50.36 for inclusion in the Technical Specifications.

Therefore, per NRC guidance, the nitrogen storage requirements do not proside a mitigating function and this activity does not increase the consequences of a plant event previously evaluated in the SAR. This

, change in no way lessens the dgwell ineding requirements governed by LCO 3.6.3.1. Additionally, this l

change does not machfy any equipment nor does it modify the operation of equipment. There is no change in the capability of the system to perform its intended function. This change does not introduce a new mode of plant operation and does involve physical modification to the plant; therefore, it does not create the possibility of a new or difTerent kind of accident or equipment malfunction. It does not involve a reduction in a margin of safety since the change is simply an administratively revised presentation, to main consistency with the ITS requirements previouslyjustified.

Technical Reauirements Manual (TRAD TITLE: Deletion ofInstrument Identification Numbers DESCRIPTION: This change deletes the details relating to the Emergency Core Cooling System (ECCS) and Reactor Core Isolation Cooling (RCIC) instrument identification mimbers associated with instruments relocated from the Custom Technical Specifications to the TRM. This change is consistent with presious ITS changes j to comply with NUREO 1433 table formats previously established. Instrument identification will be provided on a separate list supplemented with the TRM.

SAFETY EVALUATION: This change is administrative only and will have no effect on the probability of occurrence of a plant event. An allowance to deletc instnnnent identification number details will not alTect the capability of the components or systems to perform their intended functions. These details are not assumed in the mitigation of any analyzed accident. This change does not introduce a new mode of plant operation and

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does not involve a physical modification to the plant. There is no reduction in the margin of safety as there is no margin of safety associated with deleting administrative details related to instrument identification numbers.

Technical Reauirements Manual (TRM)

TITLE: Control Rod Block Instrwnentation Time Delay Allowance for Surveillance Requirements j DESCRIPTION: This change adds a note to the Surveillance Requirements stating: "When a channel is placed in an inoperable status solely for performance ofrequired Surveillance, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains control rod block capability." This allowance is consistent with that allowed for the instrumentation retained in j the Technical Specifications in the conversion per NUREG 1433.

SAFETY EVALUATION: This change has no effect on any accident initiator and doea not increase the probability of an accident  ;

or transient. The allowance is only applicable if function is maintained. Furthermore, the accident and transient mitigation function of these instruments is governed by Technical Specifications. This spec,ification governs the insttwnents only for their control rod block functions. Therefore, Technical '

Specifications ensure that the accident and transient mitigation function is operable, and this change, for the specific function, does not increase the consequences of a previously evaluated plant event. Since this

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change ensures function is maintained, it (kies not increase the probability of occurrence of any equipment l

malfunction. This change does not introduce a new mode ofplant operation and does not involve physical l modification to the plant. It does not impose any new or ditTerent requirement and does not eliminate any I requirement. Therefore, it does not create the possibility of a new or different kind of accident or  !

equipment malfunction. The affected function does not meet the criteria of 10CFR50.36 for inclusion in Technical Specifications and has no effect on any assumptions of a safety analysis. Therefore, this change does not reduce any margin of safety.

Technical Reauirements Manual (TRM)

TITLE: Deletion ofInstrument Identification Numbers DESCRIPTION: This change deletes the details relating to the Turbine Building Main Steam Line Leak Detection instrument identification numbers associated with instruments relocated from the Custom Technical Specifications to the TRM. This change is consistent with previcus ITS changes to comply with NURtG 1433 table formats previously established. Instrument identification will be provided on a separate list with the other deleted instruments from both the TRM and ITS.

SAFETY EVALUATION: This change is administrative only and will have no efTect on the probability of occurrence of a plar.1 event. An allowance to delete instrument identification number details will not affect the capability of the components or systems to perform their intended functions. These details are not assumed in the mitigation of any analyzed accident. This change does not introduce a new mode of plant operation and does not involve a physical modification to the plant. There is no reduction in the margin of safety as there is no margin of safety associated with deleting administrative details related to instrument identification numbers.

Technical Reauirements Manual (TRM)

TITLE: Emergency Core Cooling System (ECCS) Required Action Time Delay Allowance for Surveillance Requirements DESCRIPTION: This change adds a Note to the Surveillance Requirements to allow a delay of six hours in entering the associated Action during performance of Surveillance Requirements. The six hours are allowed provided the associated Function or redundant Function maintains ECCS or Reactor Core Isolation Cooling initiation capability. This allowance is consistent with that allowed for the instrumentation retained in

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the Technical Specifications in the conversion per NUREG 1433. Current Technical Specifications Bases state,"When necessary, one channel may be made inopeable for brief intervals to conduct required

' functional testsand calibrations." This change clarifies the allowance and provides restrictions on its use.

SAFETY EVALUATION: Clarifying these TRM requirements in the same manner as previously approved by the NRC for those functions retained in the ITS does not change the initial conditions or initiating action of any analped plant event. An allowance to delay entering an associated Action will not afTect the capability of the components or systems to perform their intended functions. Furthermore, these details are not assumed in the mitigation of any analyzed event. This change does not modify any equipment or its operation, does not introduce a new mode of plant operation, and does not involve physical modification to the plant.

Therefore, it does not create the possibility of a new or di1Terent kind of accident or equipment malfunction. Given that function is maintained, there is no margin of safety associated with the allowance to delay entering an asax:iated Action. Therefore, this change does not involve a reduction in the margin ofsafety.

Technical Reauirements Manual (TRM)

TITLE: Post Accident Monitoring (PAM) Required Action Time Delay Allowance for Surveillance Requirements DESCRIPTION: This change allows a six hour delay before entering an associated Action during performance of Surveillance Requirements. The six hours are allowed provided the associated Function or redundant Function maintains Non-Type A, Non-Category 1 PAM capability. This allowance is consistent with that allowed for the instmmentation retained in the Technical Specifications in the conversion per NUREG 1433. Current Technical Specifications Bases state, "When necessary, one channel may be made inoperable for briefintervals to conduct required functional tests and calibrations." This change clarifies the allowance and provides restrictions on its use.

SAFETY EVALUATION: This allowance is applicable only if function is maintained furthennore, this instrumentation (indication only) has no bearing on the initiation of any plant event. An allowance to delay entering an associated Action will not afTect the capability of the components or systems to perform their intended functions.

Therefore, the consequences of a previously analyzed accident are not increased. This change does not modify any equipment nor does it modify the operation of equipment. Therefore, it does not increase the probability of a malfunction of equipment important to safety. Since this change does not involve a physical modification to the plant and does not introduce a new mode of plant operation, no new or difTerent kinds of accidents or equipment malfunctions are created. This instrumentation is not assumed as an initial condition of, nor credited in the primary mitigation success path of, any design basis accident or transient analysis. Therefore, this change does not involve a reduction in a margin of safety.

Technical Reauirements Manual (TRM)

TITLE: Applicability of Limiting Condition for Operation (LCO) 3.0.4 to Post Accident Monitoring (PAM)

Instrumentation DESCRIPTION: LCO 3.0.4 would preclude Mode changes if a Non-Type A, Non-Category 1 PAM instrument is inoperable. This change adds a note stating that LCO 3.0.4 is not applicable to proposed TRM LCO 3.3.3 Actions. This note excludes this LCO from the Mode change restriction of LCO 3.0.4. This i exceptim i acceptable due to the passive function of the Non-Type A, Non-Category 1 PAM instrtr aents e operator's ability to diagnose an accident using attemative instruments, and the low -

probabljty u an event requiring this system. The Non Type A, Non-Category 1 PAM instruments function similar to the Type A, Category 1 PAM instmments which were previously modified in the Technical Specifications to delete the application of LCO 3.0.4.

SAFETY l EVALUATION: The Non-Type A, Non-Category 1 PAM instrumentation is not considered to be an initiator for any previously evaluated accident. This instrumentation does not meet the criteria of 10CFR50.36 for inclusion in the Technical Specifications. Therefore, it is not a part of the primary success padi in l

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1 mitigating a design basis accident or transient. Furthermore, ITS provides this same allowance for those PAM instruments that are a part of the primary success path. This change does not modify any equipment or change the mode of operation of any equipment; therefore, it does not increase the probability of occurrence of a malfunction of eqaipment important to safety. As this change does not introduce a new mode of plant operation and does not involve physical modification to the plant, it does not create the possibility of a new or difTerent kind of accident or equipment malfunction. This change does not involvc a reduction in a margin of safety since no margin of safety is associated with the monitoring function of these instnmients.

I Technical Reauirements Manual (TRM) l TITLE: Control Rod Block Specification >

DESCRIPTION: This change adds a Note to the Channel Calibration surveillances within the Control Rod Block TRM Specification which states," Neutron detectors are excluded." Channel calibrations typically include the sensor initiating the signal. IIowever, in the case ofincore neutron monitoring instrument channels, calibration of the neutron detectors is not viable. Improved Technical Specifications provide an exception i to calibration of the neutron detectors of the Average Power Range Monitor, Intermediate Range Monitor, and Source Range Monitor channcis. Similarly, the TRM is revised to explicitly allow the exclusion of the neutron detectors from the channel calibration for the Control Rod Block functions.

SAFETY EVALUATION: The Control Rod Block functions are not assumed in the initiation of any analyzed event. This change i does not impact the capabihty of the system to perform its required function; i.e., initiate a control rod block at the specified setpoint. Furthermore, this function is not assumed in the mitigation of any analyzed event in the SAR. This clunge does not involve a physical alteration to the plant (no new or ditrerent type of equipment will be installed) and there are no changes in methods governing normal plant operation.

It does not impose any r.ew or di1Terent requirements and will not eliminate any requirements. Therefore, i it does not increase the probability of occurrence or consequences of a malfunction of equiprnent important to safety previously analped in the SAR and ck>es not create the possibility of a new or different kind of accident or equipment malfunction. This change will not reduce a margin of safety since it has no impact on any s.ifety analyses assumptions. The rod block function of this instnunentation does not meet the criteria of 10CFR50.36 for inclusion in Technical Specifications. Changes to the Surveillance Requirements con sistent with those in the ITS submittal do not impact the margin of safety as defined in the basis for Technical Specifications.

Technical Reauirements Manual (TRM)

TITLE: Source Range Monitor (SRM) Count Rate During Spiral Loading and Unloading DESCR.IPTION: This change madifies when the SRM Control Rod Block allowable value of 23 cps must be met from the requirements of Current Technical Specifications to those in ITS. The 3 cps SRM minimum count rate is not required to be met on an SRM that has less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.

SAFETY EVALUATION: The probability of an accident is not increased because under these conditions, the SRM will be in the optimum pcsition for monitoring changes in neutron flux levels resulting from the core alteration.

Additionally, the reactivity addition accidents are assumed to be initiated at the lowest level of source range detect <r sensitivity and, therefore, are independent of any changes in the ability to monitor changes in the source range flux level. The SRMs are not credited for mitigation of any accidents. This change does not involve a physical alteration of the plant and there are no changes in methods governing normal plant operation. There is no change in the capability of the instrumentation to perform its intended function. Therefore, the probability of occurrence or consequences of a malfunction of equipment important to safety are not increased and no new or different kinds of accidents or equipment malfunctions are created. This change does not involve a reduction in a margin of safety because the SRMs are not credited in aly safety analysis and the use of a spiral pattern provides assurance that the SRM will be in

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l the optimum position for monitoring changes in neutron flux resulting from core alteration. As a result, '

this change does not affect current analysis assumptions. ,

Technical Reauirements Manual (TRM)

TITLE: Control Rod Block Instrument Identification Number Deletion DESCRIPTION: This change deletes the details relating to the Control Rod Block instmment identification numbers associated with instruments relocated from the Custom Technical Specifications to the TRM. This change is consistent with previous ITS changes to comply with NUREG 1433 table fonnats previously estab!ished. Instmment identification numbers will be provided on a separate list supplemented with the TRM along with other deleted instruments from both the TRM and ITS.

SAFETY EVALUATION: This change is purely admmistrative and has no etTect on the probability of occurrence of a plant event.

An allowance to delete instrument identification number details will not affect the capability of the components or systems to perform their intended ftmetions. Furthermore, these details are not assumed in the mitigation of any analyzed accident. This change does not introduce a new mode of plant operation and does not involve a physical modification to the plant. There is no reduction in sa'ety margin as there is no margin of safety associated with deleting administrative details related to instrument identification mimbers.

Technical Reauirements Manual (TRM)

TITLE: Post Accident Monitoring (PAM) Specifications DESCRIPTION: This change deletes the shutdown requirements and NRC reporting requirements from the TRM PAM specifications for PAM instrumentation that is inoperable in excess ofits allowed outage time. These requirements are replaced with an action to initiate a Problem Identification Report (PIR) for evaluation of the degraded condition. The PIR process will allow proper review and attention to assure that .

necessary and prudent actions are undertaken. I SAFETY EVALUATION: This instmmentation was previously approved for relocation from the Technical Specifications and is not assumed in the initial conditions for any plant event. As such, its inoperability has no effect on the initiation of any plant event. The PAM instrumentation relocated from Technical Specifications is by dermition not part of the primary success path for the mitigation of any design basis accident or transient, noris it assumed in any other analyzed event. This change does not involve a physical alteration to the plant, nor changes in methods governing normal plant operation. Therefore, it does not increase the probability ofoccurrence or consequences of a previously evaluated malfunction of equipment important to safety and does not create the possibility of new or different kinds of accidents or equipment malfunctions. This change will eliminate shutdowns and NRC reports required solely due to the loss of the indication only non-Technical Specification PAM instrumentation. Shutting the plant down due to loss of this indication only equipment could unnecessarily challenge systems important to safety.

Additionally, there is no margin of safety associated with reporting requirements. Therefore, any change in a margin of safety is an increase.

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Technical Reauirements Manual (TRM)

TITLE: Changing the Average Power Range Monitor (APRM) Upscale (Startup) Control Rod Block Applicability to Mode 2 Only DESCRIPTION: The APRM Upscale (Startup) Control Rod Block Applicability is changed to Mode 2 only. Current Technical Specifications state: "For the startup and run positions of the Reactor Mode Selector Switch, the Control Rod Withdrawal Block Instrumentation trip system shall be operable for each function." No allowance is given for the APRM Upscale function. This function is bypassed when the reactor mode switch is placed in "Run"; therefore, it is meaningless to require the function in Mode 1.

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l SAFETY EVALUATION: 'lhe APRM Upscale function is not assumed as an initial condition of any accident or transient analysis and, therefore, does not increase the probability of occurrence of a plant event previously evatusted in the SAR. This function is not assumed in the primary success path in the mitigation of any accident or transient. This change does not involve a physical alteration to the plant nor change methods governing normal plant operation. There is no change in the capability of the instrumentation to perform its intended function. Therefore,it does not increase the probability ofoccurrence or consequences of a malfunction of equipment important to safety, nor create the possibility of a new or different type of accident or equipment malfunction. This change simply corrects an ambiguity in the Current Technical Specifications which,in cfTect, require the operability ofinstrumentation which is automatically bypassed in the required Applicability. This change does not involve a reduction in a margin of safety.

Technical Reauirements Manual (TRM)

TITLE: Extension of Channel Calibration Surveillance Frequency for Non-Technical Specification Post Accident Monitoring System Instrumentation DESCRIPTION: This change extends the channel calibration surveillance intervals from 184 days to 18 months for Function 1 (Drywell Temperature), Function 3 (Suppression Chamber /forus Water Level -4' to +6'),

Function 4 (Suppression Chamber /forus Water Level -10" to +10"), and Function 5 (Suppression Chamber /forus Pressure). The affected instruments were previously approved for relocation from Technical Specifications, not having met the criteria for inclusion as defined in 10CFR50.36. A resiew ofmaintenance history has shown that these instruments are highly reliable, and they provide indication only. No automatic actions are perfonned by this instrumentation. The sensors and recorders are similar to others that are calibrated every 18 months and retained in the Technical Specifications.

SAFETY EVALUATION: An increase of the surveillance interval will not afTect the capability of the instruments to perform their intended function. The affected instruments are not considered initiators for any previously analyzed accidents; therefore, this change does not increase the probability of occun ence of a previously evaluated plant event. Instrumentation relocated from Technical Specifications, by definition in 10CFR50.36, is not part of the primary success path for the mitigation of any design basis accident or transient, nor is it assumed in any other analyzed event. This change does not involve a physical alteration of the plant, nor change methods goveming normal plant operation. The afTected instruments are for " indication only."

Therefore, this change does not increase the probability of occurrence or consequences of a presiously evaluated malfunction of equipment important to safety, nor does it create the possibility of a new or different type of accident or equipment malfunction. The afTected instrumentation was approved for relocation from the Technical Specifications in the conversion to the improved Technical Specifications; therefore, there is no Technical Specification margin of safety associated with this instrumentation. A channel check is also n: quired for this instrumentation; this check will identify any significant degradation ofindication.

Technical Reauirements Manual (TRM)

TITLE: Control Rod Block Instrumentation Time Delay Allowance for Performing Functional Tests and Calibrations DESCRIPTION: This change adds a note to the 31 day channel functional test and the 92 day,184 day, and 18 month channel calibrations for the Source Range Monitor (SRM), Intermediate Range Monitor (IRM), and Average Power Range Monitor (APRM) upscale control rod block functions in TRM Specification T 3.3.1. The note allows the plant to enter Mode 2 from Mode 1 and, for SRMs, enter the IRM range applicability from the higher applicability, without performing the required surveillance. The surveillance, however, must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering Mode 2 or, for the SIWs, the IRM range applicability. 'lhis change is consistent with the allowance proposed for the IRM and APIW

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Reactor Protection System functions retained in Technical Specifications in the Improved Technical Specification conversion and eliminates the need for utilizing jumpers, lifted leads, or movable links.

SAFETY i EVALUATION: The rod block function of this instrumentation is not an assumed initial condition of any accident or I l transient analysis. This change has no efTect on any accident initiator and, therefore, does not increase l the probability of an accident or transient. The rod block function of this bstrumentation is not in the primary success path for any analyzed accident or transient. The change provides an allowance to reach I proper plant conditions for performance of the surveillances. In doing so, the need to use jumpers, lifted leads, opened links, etc. is eliminated. Any effect on the probability of occurrence of a malfunction of l

equipment is a decrease. This change does not introduce a new mode of plant operation and does not j involve physical modification to the plant. The change does not impose any new or different requirement l

and does not eliminate any requirements. Therefore,it does not create the possibility of a new or difTerent l

kind of accident or equipment malfunction. The affected function does not meet the criteria of 10CFR50.36 lbr inclusion in Technical Specifications. The change has no effect on any assumptions of a safety analysis and, therefore, does not reduce any margin of safety. l Technical Reauirements Manual (TRM)

TITLE: Relocation of Current Technical Specification (CTS) Requirement for Operability of Secondary Containment Instrumentation '

l DESCRIPTION: CTS Table 3.2.D, Note 1.B contained a requirement that secondary containment instrumentation be maintained operable when moving loads inside secondary containment which have the potential to j damage irradiated fuel. Our Improved Technical Specification submittal proposed that this requirement i be relocated to the TRM. The CTS requirements for secondary containment during movement ofloads that could damage irradiated fuel, for secondary containment isolation valves during movement ofloads that could damage irradiated fuel, and for Standby Gas Treatment Subsystems during movement of loads that could damage irradiated fuel, were relocated to the USAR versus the TRM.

SAFETY l

EVALUATION: This change places the requirement for secondary containment instrumentation to be maintained operable when moving loads inside secondary containment which have the potential to damage irradiated fuel, in the same section of the USAR as the equipment supported by the instnunentation. While the subject instrumentation requirement does control equipment that would mitigate a dropped heavy loads event, this change does not change the requirement nor does it lessen the controls on the equipment. It simply changes the k) cation of the requirement. This change in no way afTects the operation or maintenance of any equipment or lessens the ability of any equipment to perform its function. This change does not increase the probability of occurrence or consequences of a plant event or malfunction of equipment important to safety, does not introduce any new failure modes or new event initiating conditions, and does not reduce the margin of safety as defined in the basis for any Technical Specification.

Technical Reauirements Manual (TRM)

TITLE: Removal of Specification for Reactor Pressure Instruments RFC-PI-90A and RFC-PI-90B DESCRIPTION: Current Technical Specification Table 3.2.F, Primary Containment Surveillance Instrumentation, contained reactor pressure instruments RFC-PI-90A&B. Our Improved Technical Specification (ITS) submittal stated that these instruments are not credited as Category 1 or Type A variables and were proposed to be relocated to the TRM. However, per Revision 3 of the CNS response to Regulatory Guide (RO) 1.97, these instruments are not credited as any Type or Category of RG 1.97 instruments. As such, they should not be included in either the Technical Specifications or the TRM Post Accident Monitoring instrumentation.

-SAFETY EVALUATION: The RFC-PI-90A&B pressure instruments are not credited in any accident or transient analysis or in the mitigation of any accident or transient, nor are they credited as RG 1.97 Post Accident Monitoring instrumentation This change afTects only the location of controls for the subject pressure instruments.

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It does not affect the operation or maintenance of any other equipment. Since the afTected instrumentation is not important to safety as defmed in the SAR and RG 1.97, this change does not increase the probability of a malfunction of equipment important to safety. No new failure modes or new event initiating conditions are introduced by this change. His change affects a requirement approved for relocation from  :

the Technical Specifications per ITS. The subject instrumentation does not meet the criteria of I 10CFR50.36 for retention in the Technical Specifications, nor is it credited as any Type or Category of RG 1.97 instrumentation at CNS. Derefore, this change does not reduce the margin of safety as delined in the basis for any Technical Specification.

Technical Reauirements Manual (TRM) l (USQE 1998-0021)

TITLE: TRM Development Package DESCRIPTION: This evaluation applies to changes made to the TRM development package aller SORC approval of Revision 0 but prior to the document being made etrective as part ofImproved Technical Specification implementation. It specifically applies to changes needed due to information being incorrectly or incompletely relocated from Technical Specifications to the TRM in the Revision 0 package. It also addresses editorial changes and the addition of information to the TRM required by the Technical Specification Bases. These changes are administrative in nature and affect only the requirements that are relocated from the Technical Specifications to the TRM in the conversion to ITS.

SAFETY EVALUATION: The changes have no efTect on any accident or transient initiating conditions and have no efTect on any event mitigators. They have no clTect on the operation or maintenance of any equipment and cannot impact the severity of any analyzed event. No new modes of operation are introduced. These requirements are all relocated from the Technical Specifications and were made to make the reh>cated requirements consistent with the Technical Specifications. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

Technical Reouirements Manual (TRM)

(USQE 1998-0021)

TITLE: Deletion of Average Power Range Monitor (APRM) Rod Block Allowable Value DESCRIPTION: This change deletes the maximum of s 108.2% of rated power for the Average Power Range Monitor (APRM) Rod Block Allowable Value in the TRM. This change is necessary since CNS is not licensed for core flevs above 100% ofrated flow and, therefore, a maximum APRM Rod Bkick Allowable Value is not applicable to CNS.

SAFETY EVALUATION: Since CNS is not licensed for core flows greater than 100%, this limit serves no design function and the absence of this limit will not affect CNS accident probability. There is no direct credit taken for the flow biased nxi block in the CNS accident analysis to mitigate the consequences of any accident. In particular, the increased flow event is discussed in the USAR. There is no credit taken to mitigate the consequences of this event for the 108.2% of rated power axl block trip. By limiting the core flow to 100%, the existing ral block trip ensures that during normal power operation, Minimum Critical Power Ratio is greater than the safety limit and, therefore, the initial conditions assumed in the accident analysis are unchanged. This change does not alter the method of operation or maintenance of equipment important to safety. It does not alter the reactor coolant boundary or any other fission product boundary. This change does not affect any other systems, structures, or components and, therefore, does not create the possibility of a different type ofmalfunction than any previously evaluated in the SAR. The design function of the flow biased axi block trip to alert the operator prior to a reactor scram during an abnormal flow transient is still preserved. Since CNS does not permit core flows above 100%, the design function is achieved without the necessity of an actual rod bhx:k trip. An additional safety factor is due to the General Electric methak> logy for determination of Critical Power Ratio which assumes a core flow of 102.5% Assuming

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a core flow of 102.5% rather than 100% provides an additional degree of conservatism. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

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Technical Recuirements Manual (TRM)

(USQE 1998-0021)

TITLE: Average Power Range Monitor (APRM) Rod Block Functions DESCRIPTION: This change revises the Allowable Values for the APRM Upscale Flow Bia:ed Rod Block, the APRM Upscale (Startup) Rod Block, and the APRM Downscale Rod Block. The revisions resulted from calculations perfonned to establish Technical Specification Allowable Values and surveillance frequencies for the APRM related functions. These calculations also included the APRM rod block functions. The new Allowable Values are more conservative than the existing values. l SAFETY EVALUATION: The changes in the TRM Allowable Values are the result of applying approved CNS instrument setpoint .

methodology. This methodology is based on GE Instrument Setpoint Methodology, NEDC-31336. I Application of this methodology results in instrumentation selected Allowable Values which more accurately reflect total instrumentation loop accuracy as well as that of test equipment and setpoint drill between surveillances. The subject changes do not result in any hardware changes. The instrumentation in the TRM is not assumed to be an initiator of any analyzed event. The function of these instruments is a rod block;it is not any function that results in a plant transient. This change in no way alters the plant's ability to detect and mitigate events. It administratively changes the APRM rod block Albwable Values in the conservative direction. This does not increase the likelihood of any equipment malfenction. The use of the subject Allowable Values does not impact safe operation of CNS in that the safety analysis limits will be maintained. Plant equipment will not be operated in a manner different from previous operation, except that setpoints may be changed. Since operational methods remain unchanged and the operating parameters have been evaluated to maintain the station within existing design basis criteria, no dilTerent type of failure or accident is created. These changes do not involve a reduction in the margin of safety of any Technical Specification. He APRM rod block functions are approved for relocation from Technical Specifications in the conversion to ITS. Furthermore, the methodology used to determine these Allowable Values is the same methodology used for those APRM scram functions retained in the Technical Specifications.

Technical Requirements Manual (TRM)

(USQE 1998-0021)

TITLE: Surveillance Frequency for Automatic Depressurization System (ADS) Inhibit Function DESCRIPTION: This change extends the surveillance frequency for the Channel Functional Test of the ADS inhibit function from 31 days to 92 days. He inhibit switch function was relocated to the TRM in the conversion to ITS. Calculations were performed for all of the other ADS functions to extend their Channel Functional Testing frequency from 31 days to 92 days. However, the inhibit switch function was relocated with the Current Technical Specification frequency of 31 days. It was determined to be acceptable to extend this frequency to 92 days because the function is not assumed in any accident or transient analysis, the function is manual, and there have been no past maintenance problems with this switch.

SAFETY EVALUATION: This change to the testing frequency of the ADS inhibit function does not increase the probability of a previously evaluated accident because this function is not assumed in the initial conditions for any accident or transient. The ADS inhibit function does not serve to mitigate any accidents or transients.

The maintenance history of the ADS inhibit switch supports a surveillance frequency extension. Given this history, a change from 31 days to 92 days will have no impact on the reliability of the function. No new failure modes are created by this change. The ADS inhibit function is not associated with the margin of safety in any Technical Specification.

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Inchnical Reauirements Manual (TRM)

(USQE 1998-0021)

TITLE: Surveillance Frequency for the Reactor Core Isolation Cooling (RCIC) Minimum Flow Function DESCRIPTION: This change extends the surveillance frequency for the Channel Functional Test of the RCIC minimum flow function from 31 days to 92 days. The RCIC minimum flow function was relocated to the TRM in  !

the conversion to ITS. Cr.lculations were performed for all of the RCIC functions retained in Technical Specifications and for all of the Emergency Core Cooling System (ECCS) minimum flow functions to extend their Channel Functional Testing frequency from 31 days to 92 days. Ilowever, the RCIC mirumum flow function was relocated with the Current Technical Specification frequency of 31 days. It was determined to be acceptable to extend this frequency to 92 days because: 1) the function is not assumed in any accident or transient analysis,2) no credit is taken for operation of the RCIC system in the safety analysis,3) the minimum flow function is tested while performing the Channel Functional Tests i of the other RCIC functions; to retain the 31 day Channel Functional Test of this function would I necessitate a 31 day run of the RCIC system which would be contrary to the intent ofITS and actually I increase the amount of time RCIC is inoperable, and 4) a logical conclusion can be drawn that the RCIC minimum flow function should be tested at the same frequency as the ECCS minimum flow functions, since RCIC serves as a far lesser capacity backup to the ECCS.

SAFETY EVALUATION: This change to the testing frequency of the RCIC minimum flow function does not increase the probability of a previously evaluated accident because it is not assumed in the initial conditions for any accident or transient. The RCIC minimum flow function does not serve to mitigate any accidents or transients. The RCIC minimum flow function serves as equipment protection only. It does not impact the RCIC system's ability to start and inject rated flow into the vessel on an automatic initiation signal. Furthermore, this change only extends the surveillance interval consistent with the Technical Specification related RCIC functions and ECCS muumum flow functions. It does not alter any plant equipment and does not create any new modes of plant operation or new types of failure modes. The RCIC minimum flow function is not associated with the margin of safety in any Technical Specification.

Technical Reauirements Manual (TRM)

(USQE 1998-0021)

TITLE: Surveillance Testing of Nuclear Instrumentation (NI) Inoperative Rod Block I DESCRIPTION: This change added a clarification to the Bases for the Control Rod Block TRM Specification, as follows: l "The ' circuit boards not in circuit' inoperative trip for the SRMs [ Source Range Monitors), IRMs

[ Intermediate Range Monitors], and APRMs [ Average Power Range Monitors] is inherent to the design  !

of the equipment and is not required to be tested." The rod block functions were relocated to the TRM  !

in the conversion to Improved Technical Specifications. A literal translation of Current Technical Specifications subjected the ' circuit boards not in circuit' inoperative trip to testing. It was determined '

to be acceptable to exclude this function from surveillance testing for the following reasons: 1) this is not j an automatic trip function that can fail to occur at the proper setpoint, 2) the removal of the circuit card from the NI drawer, by design, results in the drawer being inoperable, producing an inop trip, 3) to truly I test the function, each card would have to be pulled and receipt of the trip verified; the negative I consequences due to the likelihood damaging the equipment by pulling the card every 31 or 92 days are ,

significant,4) the rod block functions are not assumed in any safety analysis, and 5) pulling the cards can cause problems with the NI functions that are assumed in the accident analysis. -

SAFETY EVALUATION: This change to the surveillance requirements for the NI inoperative rod block functions does not increase the possibility of a previously evaluated accident because these functions are not assumed in the initial conditions for any accidents or transients. The NI inoperative rod block functions do not serve to mitigate any accidents or transients. This is simply an administrative change to surveillance testing. No new modes of plant operation or new types of failure modes are created. The NI inoperative rod block j functions are not associated with the margin of safety in any Technical Specification. There is no safety

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l impact due to excluding thia function from rigorous surveillance testing; however, there is a possible negative impact to safety related functions that would result from actually performing this testing.

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l Therefore, any efTect of not subjecting the ' circuit boards not in circuit' rod block inoperative trip to testing is a net increase in margin of safety.

Technical Reauirements Manual (TRM)

TITLE: Application of Improved Technical Specification (ITS) Surveillance Requirement 3.

0.3 DESCRIPTION

This change applies the ITS new Surveillance Requirement (SR) 3.0.3 to those specifications relocated l

l l

from the Technical Specifications to the TRM. This is a " rules of usage" surveillance requirement which l allows up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for performance of a missed surveillance prior to declaring the associated equipment inoperable.

SAFETY EVALUATION: Surveillance frequencies are not assumed to be the initiator of any analyzed event. This change will not allow continuous operation such that a single failure will preclude the associated function from being performed. It is overly conservative to assume that systems or components are inoperable when a SR has not been performed. In fact, the opposite is the case; the vast majority of SRs demonstrate that systems or components are operable. This is consistent with Generic Letter 87-09 and the allowance granted for Technical Specification surveillances in the conversion to ITS. This change does not involve a physical alteration of the plant or introduce any new modes of plant operation. There is no change in the capability of the equipment to perform its intended function. The equipment atTected by allowing this SR to be applied to relocated requirements is of far lesser significance than the equipment retained in the Technical Specifications for which this allowance is approved under the ITS Amendment. The possibility of a new or different type of plant event or malfunction of equipment important to safety is not created. Without the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> delay, it is possible that a missed surveillance could force a plant shutdown; thus, the plant could be shutting down while the missed surveillance is being performed. Any net efTect on the margin of safety will be a gain due to the avoidance of unnecessary plant transients and shutdowns. Furthermore, since these specifications are relocated from Technical Specitications, there is no efTect on Tecimical Specifications margin of safety.

j Technical Reauirements Manual (TRM)

TITLE: Application of Improved Technical Specification (ITS) Limiting Condition for Operation (LCO) 3.

0.5 DESCRIPTION

This change applies the ITS new LCO 3.0.5 to those specifications relocated from the Technical Specifications to the TRM. This is a " rules of usage" LCO which allows an exception to LCO 3.0.2 for instances where restoration of equipment to an operable status could not be perfonned (i.e., returning to senice to demonstrate operability) while continuing to comply with Required Actions.

SAFETY EVALUATION: The addition of LCO 3.0.5 allows restoration of equipment to senice under administrative controls when it has been removed from senice or declared inoperable to comply with Required Actions. The likely effect of returning equipment to senice to prove operability is an increase in safety due to increased system availability. Retuming the equipment to senice will promote timely restoration of the operability of the equipment and reduce the probability of any events that may have been prevented by such operable equipment. Since the equipment to be restored is already out of service, the unavailability of the i equipment has been previously considered in the evaluation of consequences of events. Temporarily l returning the equipment to senice will restore the capabilities of the equipment to mitigate the consequences of any events as previously analyzed. This change does not involve a physical alteration of the plant or introduce any new modes of plant operation. Furthermore, it serves to facilitate timely restoration of required equipment. Ther e is no change in the capability of the equipment to perform its intended function. The equipment affected by allowing this LCO to be applied to relocated requirements is of far lesser significance than the equipment retained in the Technical Specifications for which this allowance is approved under the ITS Amendment. Operation with the inoperable equipment temporarily restored to senice is not considered a new mode of operation since existing procedures and

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administrative controls prevent the restoration of equipment to ser ice until it is considered capable of providing the required safety functions. For those times when equipment which may be temporarily returned to senice under ad.ninistrative controls is subsequently determined to be inoperavie, the resulting condition is canparable to the equipment having been determined to be inoperable. For those times when equipment which may be temporarily returned to senice under administrative controls is subsequently determined to be inoperable during operation, continued operation for a specified time is allowed to complete Required Actions. This condition was previously evaluated in the development of the Technical Specifications. The equipment has been determined to be in a condition which provides the previously determined margin of safety for its associated function. There is no reduction in any margin of safety associated with this change. Furthermore, since these specifications are relocated from Tecimicai Specifications, there is no efTect on Technical Specifications margin of safety.

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OfTsite Dose Assessment Manual (ODAM) '

TITLE: Administrative Changes to ODAM Specifications Relocated from Technical Specifications

]

DESCRIPTION: This evaluation applies to purely administrative changes made to items relocated from the Technical I Specifications to the ODAM and to administrative / editorial changes to the body of the ODAM. The changes involve reformatting, renumbering, nontechnical rewording, elimination of redundancy, elimination of historical information, and elimination of excessive detail from the relocated Specifications and the body of the ODAM.

SAFETY l

EVALUATION: The subject changes contain no technical changes and in no way impact any event assumed initial j conditions, event initiators, or event mitigators. Therefore, they do not increase the probability of j occurrence or consequences of a previously evaluated plant event. They do not introduce any new modes l of plant operation and involve no physical alteration to any plant equipment. Therefore, they do not increase the probability of occurrence or consequences of a previously evaluated malftm: tion of equipment important to safety. The possibility of a new or different kind of event or equipment malfunction is not created as the changes are administrative only. These purely administrative changes do not involve a reduction in a margin of safety.

OfTsite Dose Assessment Manual (ODAM)  !

l TITLE: More Restrictive Changes in ODAM i

DESCIUPTION This evaluation applies to those items being relocated from the Technical Specifications to the ODAM that are modified to be more restrictive than the verbatim relocation. These changes impose more stringent requirements on operation of the facility.

SAFETY EVALUATION: These more stringent requirements do not result in operation that will increase the probability of initiating an analyzed event and do not alter assumptions relative to mitigation of any event. The more restrictive requirements continue to ensure process variables, structures, systems, and components are maintained consistent with the safety analyses and licensing basis. These changes do not involve any physical alteration to plant equipment and do not introduce a new mode of plant operation. Therefore, they do not increase the probability of occurrence or consequences of a malfunction of equipment in'portant to safety.

The possibility of a new or difTerent kind of event or equipment malfunction is not created. These more stringent changes do not involve a reduction in a margin of safety.

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l Offsite Dose Assessment Manual (ODAM)

TITLE: Application ofITS Surveillance Requirement (SR) 3.0.3 to Relocated Specifications DESCRIPTION: This change applies the ITS new SR 3.0.3 to those specifications relocated from the Technical Specifications to the ODAM. This is a " rules of usage" SR which allows up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for performance of a missed surveillance prior to declaring the associated equipment inoperable.

SAFETY EVALUATION: This change does not result in any hardware or operating procedure changes. The surveillance frequencies are not assumed to be the initiator of any analyzed event. The change will not allow continuous operation such that a single failure will preclude the associated function from being performed. The probability of occurrence or consequences of a previously evaluated event are not l incmased since the probable outcome of performing a surveillance is that it does, in fact, demonstrate the l system or component is operable. This is consistent with NRC Generic Letter 87-09 and the allowances granted for Technical Specification surveillances in the conversion to the Improved Technical

{

Specifications. This change does not involve a physical alteration of the plant or intreduce any new modes of plant operation. There is no change in the capability of equipment to perform its intended function. The possibility of a new or different kind ofplant event or equipment malfunction is not created.

l The requested time allowance will provide sufficient time to perform the missed surveillances in an I orderly manner. Without the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> delay,it is possible that the missed surveillance would force a plant I shutdown, thus, the plant could be shutting down while the missed surveillance is being performed. As I a result of the delay, the potential for human error will be reduced. As such, any net effect on a margin of safety ' vill be a gain due to the avoidance of unnecessary plant transients and shutdowns. Furthermore, since these specifications are rekrated from the Technical Specifications to the ODAM, there is no effect on Technical Specifications margin of safety.

1 Offsite Dose Assessment Manual (ODAMD )

TITLE: Application ofITS Limiting Condition for Operation (LCO) 3.0.5 to Relocated Specifications DESCRIPTION: This change applies the ITS new LCO 3.0.5 to those specifications relocated from the Technical Specifications to the ODAM. This is a " rules of usage" LCO which allows an exception to LCO 3.0.2 for instances where restoration of equipment to an operable status could not be performed (i.e., returning to senice to demonstrate operability) while continuing to comply with Required Actions.

SAFETY EVALUATION: The addition ofLCO 3.0.5 allows restoration of equipment to senice under administrative controls when it has been removed from senice or declared inoperable to comply with Required Actions. The likely i effect of returning equipment to service to prove operability is an increase in safety due to increased system availability. Retunung the equipment to senice will promote timely restoration of the operability of the equipment and reduce the probability of any events that may have been prevented by such operable equipment. Therefore, this activity does not increase the probability of occurrence of a plant event previously evaluated in the SAR. Since the equipment to be restored is already out of service, the unavailability of the equipment has been previously considered in the evaluation of consequences of events. Temporarily returning the equipment to senice in a state which is expected to function as required to mitigate the consequences of a previously analyzed accident will promote timely restoration of the operability of the equipment and restore the capabilities of the equipment to mitigate the consequences of any events as previously analyzed. Therefore, this change does not involve an increase in the corsequences of an event previously evaluated. This change does not involve a physical alteration of the plant or introduce any new modes of plant operation. Furthermore, it serves to facilitate timely restoration of required equipment. There is no change in the capability of equipment to perform its ir. tended function. Operation with the inoperable equipment temporarily restored to senice is not considered a new mode of operation since existing procedures and administrative controls prevent the restoration of equipment to senice until it is considered capable of providing the required safety functions. Performance of the surveillance is considered to be a confirmatory check of that capability which demonstrates that the equipment is indeed operable in the majority of the cases. For those times

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when equipment which may be temporarily returned to service under administrative controls is subsequently detemuned to be inoperable, the resulting condition is comparable to the equipment having been determined to be inoperable. For those times when equipment which may be temporarily retumed to senice under administrative controls is subsequently determined to be inoperable during operation, continued operation for a specified time is allowed to complete required actions. Since this condition has been previously evaluated in the development of the current Technical Specifications (information subsequently relocated to the ODAM), the possibility of a new or different type of event is not created.

There is no reduction in any margin of safety associated with this change. Furthermore, since these specifications are relocated from the Technical Specifications to the ODAM, there is no efTect on Technical Specifications margin of safety.

Offsite Dose Assessment Manual (ODAM)

TITLE: Conditions and Required Actions for Senice Water E111uent Monitor Inoperability DESCRIPTION: Current Technical Specifications state, for the Senice Water Elliuent Line Radioactivity Monitors: "With the munbers of channels operable less then required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue provided that at least once every day a grab sample is collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection not greater than 104microcuries'ml." This is translated it,to the ITS format as an ODAM Specification. Ilowever, the action required should the grab sample not be taken is in question. Verbatim translation of the CTS results in securing releases via the pathway if the sample is not taken. Securing releases means securing service water. This would be neither safe nor logical, as it would subject the plant to a scram and a demanding condition (loss of senice water). Instead, the ODAM will state a required action for this condition as: " Initiate a Problem Identification Report (PIR) for investigation of compliance faihire." So, should the proper sampling not be performed, upon discovery the sampling would be conducted and a PIR initiated to evaluate the root cause to prevent recurrence.

SAFETY EVALUATION: This change defines a reasonable approach should grab samples not be taken as required for an inoperable service water radiation monitor. In requiring a PIR be initiated to investigate the event and prevent recurrence versus securing senice water, a plant transient is avoided and the probability of occurrence of a plant event is decreased. As this change will prevent unintentionally securing senice water and subjecting the plant to an unnecessary transient, it retains senice water capability to event mitigating equipment. This change decreases the probability of occurrence of a malfunction of equipment important to safety in that it relaxes a requirement to intentionally secure plant senice water when there is no reason to suspect a problem. This action does not introduce any new modes of operation or negatively impact any equipment and does not involve physical modification to the plant. Furthermore, it does not create any new event initiating conditions. Subjecting a faihire to perform required sampling of the senice water system to the PIR process versus inducing a plant transient by securing senice water results in an increase in the margin of safety. Unnecessarily challenging the plant and the operators with a diflicult scenario when there is no reason to suspect a problem is not a conservative, safe approach.

Therefore, any change in a margin of safety is an increase. l OfTsite Dose Assessment Manual (ODAM)

(USQE 1998-0020)

TITLE: Revision 1 to Appendix D of the ODAM DESCRIPTION: This evaluation applies to changes made to the ODAM development package following SORC approval of Revision 0, but prior to the document being made effective as part of Improved Technical Specifications (ITS) implementation. These changes are administrative and affect only the specifications that are relocated from the Technical Specifications to the ODAM in the conversion to ITS. The changes are made to correct minor inaccuracies or omissions in the initial development of the ODAM Appendix D. In making these corrections, the specifications directly translate the Current Tecimical Specifications.

There are no changes made to any existing requirements.

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SAFETY EVALUATION: The changes have no effect on any accident or transient initiating conditions or event mitigators. They do not affect the operation or maintenance of any equipment and cannot impact the severity of any analyzed event. No new modes ofoperation are introduced. There is no reduction in the margin of safety as defined in the basis for any Technical Specification as these changes make the rek)cated requirements consistent with the Current Technical Specifications. The changes maintain the level of radioactive elliuent control required by 10CFR20.1302,40CFR190,10CFR50.36a, and 10CFR50 Appendix I and do not adversely impact the accuracy or reliability of cilluent, dose, or setpoint calculations.

l Safety Function Determination Procram (SFDP)

TrfLE: l Revision I to Safety Function Determination Program '

DESCRIPTION: This revision to the SFDP made changes that resulted from program validation in training or changes made to the ITS submittal since the SFDP was initially approved. They all reflect actual site configuration and the ITS submittal thmugh Revision A.

SAFETY EVALUATION: These changes enhance the program and make it match the actual plant configuration and the ITS l

submittal. They in no way negatively impact any assumed event initiating conditions or accident mitigators. No changes are made to any plant equipment or the capability of any equipment to perform l

its intended function. The possibility of new or different types of plant events or equipment malfunctions are not created because this change does not introduce a new mode of plant operation and involves no physical modification to the plant. Any change in the margin of safety as defined in the basis for any Technical Specification is an increase.

Technical Reauirements Manual Chance Reauest 1998-001 (USQE 1998-0031)

TITLE: Revision 2 to Safety Function Determination Program (SFDP) l ,

DESCRIPTION: This revision incorporates the Improved Technical Specifications (ITS) readiness assessment comments, comments resulting from site validation of the program, and changes due to Revision B of the ITS submittal.1hese changes are admmistrative in nature, correcting inaccuracies and updating the program l

to the approved ITS.

SAFETY l EVALUATION: These changes enhance the SFDP consistent with ITS. There is no efTect on any accident or transient initiating conditions and operator response to transients is unalTected. The changes have no efTect on any

event initiators, other than conecting when a kiss of function occurs. They have no efTect on the operation I

or maintenance of any equipment and, therefore, do not increase the probability of a malfunction of equipment important to safety. The changes have no effect on accident mitigating equipment. They do affect the administrative controls of the equipment, but in doing so define when a loss of safety function occurs which will drive conservative action. No new modes of operation are introduced. The program

uses the Technical Specifications to define losses of function. It will be used to determine when l inoperable supponed systems' Required Actions must be taken. The changes are in accordance with the approved ITS and the direction of LCO 3.0.6 and its Bases Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification y

USOR 1999-0028 l

TFFLE: Changes to Improved Technical Specification (ITS) Bases DESCRIPTION: A number of changes were made to the ITS Bases subsequent to the receipt of the ITS License Amendment Safety Evaluation Report dated July 31,1998, but prior to implementation of ITS on l August 15,1998. These changes were collectively screened as not requiring an Unreviewed Safety t

Question Evaluation (USQE)in a single 10 CFR 50.59 screening. A Problem Identification Repon was

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f generated because the screening did not specifically address each Bases change to justify why a USQE was not needed. A detailed screening was subsequently performed and it was determined that seven issues warranted a full USQE. Each of these issues are discussed below.

TIEM 1 Bases B 3.3.3.1, Post Accident Monitoring Instrumentation DESCRIPTION: This change raised Limiting Condition for Operation (LCO) Section 4, Primary Containment Gross Radiation Monitors, to clarify that the common recorder is not pan of the channel requirements. The Bases clearly state that both channels of radiation monitoring instrumentation are required to be operable for compliance with the LCO. However, the common recon'er is not part of the " channels" required for operability as it simply provides a means to obtain readout / historical data.

SAFETY EVALUATION: Excludmg the common recorder from the channel operability requirements does not alter the function of the radiation monitors or the ability to comply with Regulatory Guide 1.97 requirements. Therefore, there is no increase in the probability of a malfunction of equipment important to safety. Since the common I

recorder and radiation monitoring system are not accident or transient initiators, there is also no '

probability ofincreasing the possibility of an accident or transient. In addition, there is no impact on  ;

fission product barriers, therefore there is no increase in the consequences of an accident, transient, or l equipment malfunction. No new failure modes are introduced, nor does this change create the possibility of an accident, transient, or equipment malfunction of a different type than previously evaluated. The common recorder does not affect the radiation monitoring function in terms of when the detectors pick up high radiation and send signals to the monitors for action / annunciation. Adding the statement that the common recorder is not included in the channel requirements does not impact the ability to perform this i function or meet Technical Specification requirements. Therefore, there is no reduction in the margin of I safety as defined in the basis for any Technical Specification.

ITEM 2 Bases B 3.3.3.2, Alternate Shutdown System DESCRIPTION: This change revised the LCO discussion to indicate that it is necessary to verify operability of the instnnnentation and controls required for " cooling water" systems that perform a safety suppon system function for vessel pressure control, inventory control, and decay heat removal. The text previously referred to Reactor Equiptr.ent Cooling (REC) v1 cooling wc*.cr. I lowever, not all REC is needed for safe shutdown from outside the Control Room, as REC has both critical and non-critical supply loops, and cooling water does not need to be limited to REC. This change in no way affects the list of instmmentation/ controls required to be operable to ensure safe shutdown from outside the Control Room.

SAFETY EVALUATION: There are no impacts on accident or transient initiators, nor is there an increased probability of a l malfunction of equipment important to safety because the requirements of cooling water as a support system (which provides assurance that safety related equipment will not malfunction) have not changed.

This change does not afTect fission product barriers or the ability to ensure safe shutdown of the reactor and, therefore,(krs not increase the consequences of an accident, transient, or malfunction of equipment important to safety. Operability requirements and operation of the equipment is not being changed, the change by itselfis admmistrative in nature and does not create the possibility for accidents, transients, or equipment malfunctions of a different type than previously evaluated. This change does nu affect the support system operability requirements for meeting safe shutdown criteria from outside the Control Room. Margin of safety is not affected because the change from REC to cooling water has no impact on the need to maintain certain instmmentation and controls operable and functional in the event a shutdown of the reactor is required from outside the Control Room. There are no impacts on the key elements of this Technical Specification, which are reactor vessel pressure control, vessel inventory control, and decay heat removal.

ITEM 3 Bases B 3.6.1.3, Primary Containment Isolation Valves (PCIVs) l DESCRIPTION: This change deleted the following statement from the Background section: "To ensure that a vent path is available, a 12 inch bypass line is provided around the valve leading directly to the SGT [ Standby Gas l -170

Trea' ment] system." This statement was deleted because the 12" bypass line is not related to the PCIVs

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or the 24" contamment purge valves. A description of the vent path is provided earlier in Bases B 3.6.1.3

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as follows: "The isolation valves on the 24" vent lines have bypass lines around them for when the 24 inch J

valves cannot be openal." The CNS design as described in the USAR does not discuss a 12" bypass line  !

to ensure a vent path is available.

SAFETY EVALUATION: The removal of this information does not result in an increase in the probability of a malfunction of

)

equipment important to safety because there is no change in how vent and purge is actually accomplished, l

nor are there any additional requirements placed on systems, structures, or components important to safety such that an increase in equipment malfunctions could be expected. The change does not affect Primary Containment (PC) as the venting design and operation is not being changed, and the 12" bypass line description is not applicable to PC integrity, nor does it atrect any other fission product barriers.

Therefore, it does not result in an increase in the consequences of an accident, transient, or equipment malfunction. In addition, since the removal of this description has no impact on accident or transient initiators, nor does it affect equipment required for primary / secondary containment integrity, the possibility of an accident, transient, or equipment malfunction of a difTerent type than previously evaluated is not created. The Technical Specification Bases clearly discuss how vent and purge of PC is accomplished and the removal of the subject statement does not impact the margin of safety as defined in the basis for any Technical Specification.

ITEM 4 Bases B 3.6.1.3, Primary Contairunent Isolation Valves (PCIVs)

DESCRIPTION: This change revised Surveillance Requirement 3.6.1.3.2 to delete the statement "the primary containment

, is inerted" as one of the reasons why verification of valve position by the use of administrative controls

! is acceptable for PCIVs and blind fianges located outside primary containment. The Bases allow this to l be accomplished by administrative means primarily because the valves and flanges are located in high  !

l radiation areas and access to these areas is typically limited. Since the Surveillance Requirement deals l with components located outside primary containment, the fact that primary containment is inerted is l irrelevant.

l SAFETY l EVALUATION: Removing this statement does not change the method by which administrative verification is performed I l

or the reasons why administrative verification is allowed. This change does not impact accident or J transient initiators, nor is there an increase in the probability of equipment malfunction because this is i j- an admmistrative change which atrects neither form, fit, function, operation, or the description of how the PCIVs are to perform their function. As there are no changes in verification requirements, there is no reduction in the ability of containment to perform its safety function, and thus there is no increase in the I consequences of an accident, transient, or equipment malfunction; nor is there a reduction in the margin j of safety. The ability to administratively verify certain PCIVs are closed is not a new allowance and is i specifically allowed by Technical Specifications; therefore, no new types of accidents, transients, or equipment malfunctions are introduced.

ITEM 5 Bases B 3.7.2, Service Water (SW) System and Ultimate IIcat Sink (Ulls)

DESCRIPTION: Surveillance Requirement 3.7.2.4 was revised to change the referenced setpoint value for when SW pressure switches will actuate to ensure cooling water is exclusively supplied to essential loads from

" greater than or equal to 20 psig" to"appmximately 20 psig " The actual setpoint value was not changed, an engmeenng setpoint calculation demonstrates that a " nominal" setpoint of 20 psig is acceptable, with tolerances both above and below this value. Therefore, it is appropriate to state "approximately."

SAFETY EVALUATION: "Ihe pressure switches are not accident or transient initiators, and the calculation supports the appropriate setpoint values to avoid an increase in the probability of equipment malfunction (due to spurious trips, inappropriate settings, etc.). Since these pressure switches do not afTect fission product barriers and the l calculation supports the appropriate setpoint values to ensure accident mitigation capability of supported systems, there is no increase in the consequences of an accident, transient, or equipment malfunction as a result of this change. As the calculation determines the appropriate nominal setpoint and tolerances,

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E this change to "approximately" does not create the possibility of an accident, transient, or equipment malfunction of a difTerent type than previously evaluated, nor does it result in a reduction in the margin of safety as defined in the basis for any Technical Specification.

ITEM 6: Bases B 3.7.3, Reactor Equipment Cooling (REC) System I

DESCRIPTION: Surveillance Requirement 3.7.3.4 was revised to change the setpoint values listed for the REC heat exchanger outlet pressure switches, which provide isolation of non-critical loads on low header pressures, from "less than or equal to" to "approximately." This change is supported by an engineering setpoint calculation which infers "nommal" setpoints with the appropriate tolerances. Therefore, it is appropriate to state "approximately."

SAFETY EVALUATION: 'Ihe pressure switches are not accident or transient initiators, and the calculation supports the appropriate setpoint values to avoid an increase in the probability of equipment malfunction (due to spurious trips, inappropriate settings, etc.). The response or safety function of the switches is not changed. Since these pressure switches do not afTect fission product barriers and the calculation supports the appropriate setpoint values to ensure accident mitigation capability of supported systems, there is no increase in the '

consequences of an accident, transient, or equipment malfunction. The calculation determines the appropriate nominal setpoints and tolerances; therefore this change to "approximately" does not create the possibility of an accident, transient, or malfunction of a different type than previously evaluated, nor does it result in a reduction in the margin of safety as described in the basis for any Technical Specification.

ITEM 7: Bases B 3.8.1, AC Sources - Operating DESCRIPTION: This change deleted the limit of"4700 kW - 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year" from the listing of Diesel Generator ratings. It was determined this was an unnecessary limit to have listed in the Bases since the limit of "4400 kW 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per day," which is also included in the Bases, is more limiting. The vendor rating of 4700 kW for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year is still reflected in the USAR.

SAFETY EVALUATION: Deleting the description of this rating from the Bases does not change the rating and has no impact on the Diesel Generators' design, operation, or testing. This change will not allow the Emergency Diesel Generators (EDGs) to exceed any vendor-recommended limits; therefore, there is no increase in the probability of a malfunction of the EDGs. The EDGs are not accident or transient initiators, thus this change has no impact on the possibility ofincreasing an accident or transient. The EDGs do not affect fission product barriers; therefore, there is no impact on the consequences of an accident, transient, or malfunction ofequipment important to safety. As the ratings, design, operation, maintenance, and testing of the EDGs remain unchanged, there is no possibility of creating an accident, transient, or equipment malfunction of a ddTerent type thar previously evaluated, and there is no reduction in the margin of safety as defined in the basis for any Technical Specification.

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REGULATORY COMMITMENT CilANGES j

The following regulatory commitments were revised based upon evaluations performed in accordance with the NEI Guideline for Managing NRC Commitments. '

Source of Commitment: 1xtter NSD930526 from G. R. Ilorn (NPPD) to NRC Document Control Desk dated f April 29,1993 - Response to Notice of Violation and Proposed Imposition of Civil i Penalties (IR 93-06) l Revision to Conunitment: The response to IR 93-06 stated that 10 CFR 50.9 training would be integrated into NPPD's Management and Technical StafT continuing training programs. This commitment was revised to state that CNS will maintain a procedure (0.42,"NRC Correspondence Control") that requires verification and validation of information submitted to the NRC in accordance with 10 CFR 50.9 requirements. Review of conespondence to the NRC in accordance with CNS Procedure 0.42 will prevent violations of 10 CFR 50.9 more consistently than providing training to Management and the Technical Staff.

Source of Commitment: Letter NLS960213 from P. D. Graham (NPPD) to NRC Document Control Desk dated November 6,19% - Reply to a Notice of Violation NRC Inspection Report 50-298/96-12 Revision to Commitment: The response to IR 96-12 indicated that the period of demand for measuring Iligh Pressure Coolant Injection (IIPCI) system unavailability under 10CFR50.65 will be determined by the position of the reactor head vent. Unavailability will be accrued for l IIPCI if unavailable when the reactor head vent is closed. This commitment was revised to add "EXCEPT during vessel hydrostatic test, pressure test, or during cold shutdown " 1IPCI is unable function during cold shutdow n, vessel hydrostatic test, or pressure test due to lack of adequate steam pressure, so IIPCI is not in a period of demand at these times.

Source of Conunitment: Irtter NLS960034 from J.11 Mueller (NPPD) to NRC Document Control Desk dated March 14,19% - Reply to a Notice of Violation, NRC Inspection Report 50-298/95-17 Revision to Conunitment: The response to IR 95-17 stated that CNS Procedure 2.0.10 was revised to include a sign-oft ror the reconciliation of the Foreign Material Exclusion (FME) log prior to torus close out. This commitment was resised to specify that CNS Procedure 0.45.1 was written and implemented to provide FME controls for work in the torus.

Procedure 0.45.1 requires FME log reconciliation and inspection of the torus prior to torus closure. The revision to Procedure 2.0.10 was originally completed in March 19% to address the Notice of Violation. Procedure 0.45.1 was first implemented in November 19%. There is no need to have essentially redundant requirements in two different procedures.

Source of Commitment: letter NLS950069 from G. R. Ilorn (NPPD) to NRC Document Control Desk dated March 16,1995 - Resubmittal of a Reply to a Notice of Violation and Proposed Imposition of Civil Penalties, NRC Inspection Report Nos.50-29f;/94-14, 50-298/94-16, and 50-298/94-19

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l Revision of Commitment: The response to Violations A.1 and A.2 stated that the current licensing basis for Appendix J testing would be formally captured in an Appendix J Testing Basis l Document. This commitment was revised to str.te that the curcent licensing basis for Appendix J testing is currently captured in the Appendix J Program Document, l l USAR, and the Primary Containment Design Criteria Document. The original commitment has been overcome by events and the need to replicate the information from the above referenced documents is not necessary.

l Source of Commitment:

Letter LQA8200019 from J. M. Pilant (NPPD) to G. L. Madsen (NRC) dated September 24,1982 - NPPD Response to IE Inspection Report 50-298/82-20 Revision of Commitment: The response to inspection Rep rt 82-20 stated that Procedure 9.1.1.4, "Special Work Pennit"(SWP), was being revised to require a beta survey whenever a SWP is written due to smearable contamination levels in excess of 2200 dpm/100 cm2 (subsequently l corrected to 22,000 dpm/100 cm2 and acknowledged by the NRC in closure of j associated violation). This commitment was revised to state that in areas where '

2 general contamination levels are found to be > 100,000 dpm/100 cm , an initial beta survey shall be conducted. Subsequent beta surveys in these areas shall be performed per Radiation Work Permit requirements and when Radiation Protection supenision l deems necessary. It was determined that requirements for air sampling and beta dose 2

rate surveys at contamination levels less than 100,000 dpm/100 cm are generally low l value activities from a radiological hazard and ALARA standpoint. The deletion of ,

the phrase "whenever an SWP is written" is due to the fact that an SWP will have i already been issued as a result of SWP highly contaminated area issuance criteria. The beta survey requirement may show up in several Radiation Protection procedures, therefore, the specific procedure number was removed from the commitment. l Source of Commitment: Enforcement Conference held on October 17.1997 Revision of Commitment: The original commitment was to provide training to all CNS personnel on the lessens learned from the Residual Heat Removal (RHR) Heat Exchanger issue. This commitment was revised to state that tailgate training on the lessons leamed from the l RHR Heat Exchanger issue will be provided to appropriate personnel. Training "all" l

CNS personnel is inetTective and costly. Only personnel making safety-related decisions should be required to attend this training.

Source of Commitment: Letter CNSS907024 from L. G. Kunci (NPPD) to NRC Document Control Desk dated  ;

January 29,1990 - Response to Generic Letter 89-13 Revision of Commitment: Rcccatmended Action II of Generic Letter 89-13 required licensees to conduct a test program to verify the heat transfer capability of all safety related heat exchangers cooled by senice water. A portion of our response to this item indicated that the i District currently verifies the heat transfer capability of the Diesel Generator Jacket Water (DGJW) and Diesel Generator Lube Oil (DOLO) safety related heat exchangers by testing at least once each operating cycle. In 1997 this commitment was I subsequently revised to state that visual inspection and cleaning, as required, would

! be performed each operating cycle. This commitment has been revised again to state that the District will verify the heat transfer capability of the DGJW and DGLO heat exchangers by performing visual inspection and cleaning, as required, every other operating cycle. llistory demonstrates that as-found conditions in the DGJW and DGLO heat exchangers justify less frequent cleaning.

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Source of Commitment: Letter CNSS870060 from G. A. Trevors (NPPD) to J. E. Gagliardo (NRC) dated

]

January 27,1987 - NPPD Response to IE Inspection Report 50-298/86-27 i

Revision of Commitment: The response to Inspection Report 86-27 stated that Procedure 10.25, Refueling, was revised to add a new admmistrative limit and a note prior to the applicable step which I stated that when traversing the refueling bridge between the fuel pool and the reactor I cavity, the refuel mast shall be fully retracted. This commitment was revised to

{

indicate that this procedural precaution and associated equipment operating steps will I

be moved by procedure changes from the 10.25 refueling procedure to Procedure j 2.2.31, Fuel Ilandling-Refueling Platform. Procedure 2.2.31 contains specific  !

operating instructions for the refueling bridge equipment; therefore, this precaution '

and operating steps are better suited in the equipment operating procedure.

I i

Source of Commitment: CNS Licensee Event Report 93-031, Revision 1, dated February 17,1994 - High )

Pressure Coolant Injection System Inoperability Resulting from a Dislodged Motoi Operated Valve Pinion Gear Key and Motor Starter Contaminants Revision of Commitment: The corrective action for LER 93-031 stated that procedures used to develop and j implement design changes (including temporary modifications) would be reviewed to i ensure that post-modification inspections address cleanliness. Procedures 2.0.7 (Plant Temporary Modification Control) and 3.4.3 (Design Changes) were subsequently I revised. The revisions made to these procedures are no longer required because the Foreign Material Exclusion (INE) Program (Procedure 0.45) now performs the same function to a more stringent level. The steps added to Procedure 3.4.3 were removed in July 1996. The remaining references / steps should not be transferred to the revised 3.4 procedure, which will replace 3.4.3 and 2.0.7, because the FME program is l adequate.

Source of Commitment: CNS Licensee Event Report 97-014 dated December 12,1997 - Station Modification Creates Pctential Inability to Mitigate Accident Consequences Revision of Commitment: The response to LER 97-014 stated that CNS would perform a Safety System FunctionalInspection (SSFI) type evaluation of two risk significant systems that have (

had significant design modifications by the end of April 1998 to further evaluate past modifications. This commitment was revised to state," Complete actions 3.3.f and

)

j 3.3.g of the Strategy for Achieving Engineering Excellence [ Revision 2]. Develop and l implement a program to ensure that appropriate safety margin exists." The original commitment was replaced by the subject actions from the integrated Engineering plan.

Action 3.3.f of the Strategy fbr Achieving Engineering Excellence states," Conduct primary and secondary containment SSFI-like inspection" and has a due date of September 30,1999. Action 3.3.g states," Safety analysis inputs and assumptions translation to processes and programs" and has a due date of December 31,2002.

Source of Commitment: CNS Licensee Event Report 97-014 dated December 12,1997 - Station Modification

)

Creates Potential Inability to Mitigate Accident Consequences Revision of Commitment: In the response to LER 97-014, CNS committed to conduct a review of past design changes to risk significant systems to identify and evaluate station modifications with the potential to impact the ability of safety related systems to perform their design basis functions by June 30,1998. In addition, the licensec committed to implement

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additional corrective actions as necessary based on the results of this review, including expanding the scope of the review as appropriate. This commitment was revised to I perfonn the same activity by September 30,1998, as Action No. 3.4.e of the Strategy for Achieving Engineering Excellence, Revision 1. This commitment was revised a second time to perform the same activity by December 31,1998, as Action No. 3.3.d q of the Strategy for Achieving Engineering Excellence, Revision 2. This commitment I was revised a third time to complete the same activity by June 30,1999 (reference NLS990009 dated February 2,1999, from NPPD to the NRC). Action No. 3.3.d -

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states," Perform a review of selected past mulifications."

Source of Commitment: Letter LQA8000347 from J. M. Pilant (NPPD) to K V. Seyfrit (NRC) dated July 14, 1980 - IE Bulletin 80-14, Degradation of BWR Scram Discharge Volume (SDV)

Capability Revision of Commitment: The CNS response to IE Bulletin 80-14 stated, "If the SDV vent and drain valves are not operable or are closed for more than I hour in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period during operation, the reason will be logged and the NRC will be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." d This commitment was deleted as 10CFR50.72 requirements and our improved reporting process now prop:rly document and report these types of deficiencies. In addition, CNS installed modifications to ensure the SDV system addresses issues of safety function capability, including redundancy and reliability for detection of an isolated condition.

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